3rd floor (Clarion Congress Hotel Prague)

3rd floor

Clarion Congress Hotel Prague

Freyova 945/33, 190 00 Prague 9 - Vysočany

The 7th International Conference on
Advancements in Nuclear Instrumentation Measurement Methods and their Applications

Dear Colleagues, dear friends, dear ANIMMA community members,

ANIMMA 2021 is the seventh of a series of conferences devoted to endorsing and promoting scientific and technical activities based on nuclear instrumentation and measurements. The main objective of the conference is to unite the various scientific communities not only involved in nuclear instrumentation and measurements, but also in their wide fields of applications such as fundamental physics, space technologies, nuclear medicine, medical and environmental sciences. The conference is all about getting scientists, engineers, students, and the industry to meet, exchange cultures and identify new scientific and technical prospects to help overcome both current and future unresolved issues.

Furthermore, in light of the current worldwide sanitary situation, we would like to reassure you that the Program and Organizing Committees are devoted to making this seventh edition a successful meeting place shared by all those working in nuclear instrumentation and measurement and their applications, as we continue to strongly believe that cross-border exchanges between scientists, engineers and industrialists can only lead to the most developed ideas, the best solutions and the most efficient collaborations and partnerships. 

With my kind regards,

Abdallah LYOUSSI

ANIMMA General Chair and Program Committee Chair


ANIMMA 2021 Organizers and Partners

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Exhibitors Registration Form
  • CAO, Hongrui
  • DAS, Biswajit
  • ELEON, Cyrille
  • HODAK, Rastislav
  • LALLEMAN, Anne-Sophie
  • LEFEUVRE, Gwenaëlle
  • STEKL, Ivan
  • TANCHAK, Anhelina
    • 08:40 18:30
      Short Courses: SC Day #1

      For the Short Courses, a very complete and remarkable program is proposed for two days (Sunday June 20 and Monday June 21) mixing basics and advanced lectures as well as lab sessions (exercises on nuclear instrumentation & measurement).

      Convener: VERMEEREN, Ludo (SCK CEN, Belgium)
      • 08:40
        Welcome 5m
        Speaker: VERMEEREN, Ludo (SCK CEN, Belgium)
      • 08:45
        Introduction by the participants 10m
        Speaker: ALL PARTICIPANTS
      • 08:55
        Radiation detection and measurement methods – part I 1h 5m

        Starting from the physical principles, the course will discuss the performances and the limitations of various radiation detectors that can be used in nuclear reactors and in the subsequent stages of the nuclear fuel cycle. Two specific applications will illustrate the practical applications of the detection techniques: neutron dosimetry in fission reactors and nuclear monitoring of the fuel cycle (passive and active neutron measurements, sensing via photo fission, coupling of measurements and combined interpretation).

        The course will be structured as follows:

        Interaction of radiation with matter
        Physical principles of radiation detectors
        (a) Gas-filled detectors
        (b) Scintillation detectors
        (c) Solid-state detectors (semiconductor, diamond, TLD)
        (d) Activation detectors

        Speaker: LYOUSSI, Abdallah (CEA)
      • 10:00
        Coffee Break 15m
      • 10:15
        Radiation detection and measurement methods – part II 45m

        Starting from the physical principles, the course will discuss the performances and the limitations of various radiation detectors that can be used in nuclear reactors and in the subsequent stages of the nuclear fuel cycle. Two specific applications will illustrate the practical applications of the detection techniques: neutron dosimetry in fission reactors and nuclear monitoring of the fuel cycle (passive and active neutron measurements, sensing via photo fission, coupling of measurements and combined interpretation).

        The course will be structured as follows:

        Interaction of radiation with matter
        Physical principles of radiation detectors
        (a) Gas-filled detectors
        (b) Scintillation detectors
        (c) Solid-state detectors (semiconductor, diamond, TLD)
        (d) Activation detectors

        Speaker: LYOUSSI, Abdallah (CEA)
      • 11:00
        Neutrino and Dark Matter Detection 1h

        Our Universe is composed of about 68 % of dark energy, 27 % of dark matter, and less than 5 % of baryonic matter including < 1 % of neutrinos. Neutrinos and dark matter share many similar features. The same way neutrinos are considered as elementary particles, the nature of dark matter is widely believed to be based on the existence of new type of elementary particles as well. They both interact very weakly with visible matter, and therefore their direct detection is extremely difficult requiring ultra-low background conditions and advanced statistical methods for data analysis. Many theories explaining tiny neutrino masses predict the existence of heavy sterile neutrinos which are serious candidates of dark matter particles. Generally, this is accompanied by lepton number violation which is a necessary ingredient of explanation of matter-antimatter asymmetry of the Universe. That relates neutrinos and dark matter to the cosmological concept of the leptogenesis. In the context of detection of neutrinos and dark matter, decay or annihilation of dark matter particles could present a key role producing neutrino fluxes detectable for large volume neutrino telescopes.

        The short overview of the recent status of the neutrino and dark matter physics, with emphasis to the experiments where IEAP researchers take an active role in the detector construction, data processing or MC simulations, will be introduced. Namely, experiments Baikal-GVD (neutrino telescope – high energy cosmic neutrinos detection), SuperNEMO (search for neutrinoless double beta decay), S3 (reactor antineutrinos detection & search for sterile neutrinos), DUNE (neutrino oscillations) and PICO (direct dark matter search) are going to be presented.

        Speaker: FAJT, Lukáš (IEAP CTU in Prague)
      • 12:00
        Sandwich Lunch 1h
      • 13:00
        Lab session 1 - Introduction to gamma dose rate, spectroscopic measurements, and source identification 1h 45m

        Detection of gamma emitters is one of the main issues in most of the radiological fields, from dismantling and decommissioning to safety and environment. The important aspect to understand are dose rate measurements indicating the seriousness of the situation, but it is also important to identify the source type to better handle the identified material. This short course will focus on an introduction to gamma radiation focusing on the identification of gamma sources based on spectroscopy. The course will end with a hands-on to let the students measure using a handheld instrument based on a specific example, the tagging and characterization of a nuclear waste drum.

        Speakers: FANCHINI, Erica (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy) , MANGIAGALLI, Giacomo (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy)
      • 14:45
        Lab session 2 - Environmental gamma radiation meaurements with SiPM based instrumentation 1h 45m

        The course will describe the Naturally Occurring Radioactive Materials (NORM) describing the natural background around us and the principal sources of radioactivity. NORM potentially includes all radioactive elements found in the environment without human intervention. These elements have always been present in the Earth's crust and atmosphere and in any case, exposure to NORM radiation is responsible for the majority of an average person’s yearly radiation and is therefore not usually considered of any special health or safety significance.

        The second step of the course will be analyzing a gamma spectra of a NORM sample to recognize the track and, after this review, the students will use a radiological sensor (if possible) to detect different sources of natural radioactivity to evaluate their content, identify the radioisotopes comparing results between different measurements.

        Speakers: FANCHINI, Erica (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy) , MANGIAGALLI, Giacomo (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy)
      • 16:30
        Coffee Break 15m
      • 16:45
        Lab session 3 - Pixel detector hands-on exercises with MX-10 particle camera 1h 45m

        The short course is focused on the progressive detector technology of pixel detectors. Medipix is a family of read-out chips and detector assemblies for particle detection and imaging developed by the Medipix Collaborations. The original concept of Medipix is that it works like a camera, detecting and counting each individual particle hitting the pixels when its electronic shutter is open. This enables high-resolution, high-contrast, noise hit free images – making it unique for imaging applications.

        Pixelman SW for control and evaluation of Medipix/Timepix detectors was developed and licensed by the Institute of Experimental and Applied Physics (IEAP) of the Czech Technical University in Prague. The R/O interface MX-10, based on the Medipix/Timepix detector, is produced by the Czech company Jablotron in cooperation with IEAP.

        Participants will get familiar with the pixel detector Timepix through the session of practical hands-on exercises performed with MX-10 Edukits. Participants will learn how to build simple experimental set-up, how to configure measurement parameters, how to run acquisition and how to evaluate own measured data. The basic types of radiation of several emitters will be studied and distinguished according to measured tracks via the pixel detector. They will also examine the natural radiation background including cosmic rays muons. In addition to the main focus of the course, participants will get introduced into detection of neutrons posing an example of a more advanced experiment practicable with pixel detectors.

        Speakers: Dr HOLIK, Michael (IEAP CTU in Prague, UWB in Pilsen) , VICHA, Vladimir (IEAP CTU in Prague)
    • 09:00 18:30
      Short Courses: SC Day #2

      For the Short Courses, a very complete and remarkable program is proposed for two days (Sunday June 20 and Monday June 21) mixing basics and advanced lectures as well as lab sessions (exercises on nuclear instrumentation & measurement).

      Convener: VERMEEREN, Ludo (SCK CEN, Belgium)
      • 09:00
        Hybrid Pixel Detector Applications in Space 1h

        Hybrid pixel detectors of Timepix technology have become increasingly interesting for space applications. Several devices are for example flown on the international space station providing a real-time measurement of the radiation levels. A key strength of Timepix is that ionizing particle interactions in the sensor are seen as imprints in the pixel matrix (tracks) with a rich set of features which can be exploited for the identification of impinging particles and particle trajectory reconstruction. The first Timepix (256 x 256 pixels, pixel pitch: 55 µm) used in open space is SATRAM (Space Application of Timepix Radiation Monitor) attached to the Proba-V satellite launched to low earth orbit (LEO) in 2013. In this contribution, measured dose, electron, and proton flux maps are shown. It is demonstrated how the use of machine learning tools improves particle separation. Further, the development and space qualification of a MIniaturized RAdiation Monitor (MIRAM) based on the latest-generation chip of the Timepix family, Timepix3, are described.

        Speaker: BERGMANN, Benedikt (Institute of Experimental and Applied Physics)
      • 10:00
        Counting neutrons in a nuclear reactor by reactor dosimetry 30m

        The aim of reactor dosimetry is to determine neutron fluences for various types of applications. One of the most commonly used techniques for this purpose is the neutron activation technique. Surveillance capsules present in nuclear power reactors, for example, are typically equipped with a set of activation wires or foils (dosimeters). After irradiation, the neutron fluence received by the surveillance capsule can be determined from the measured dosimeter activities. Also, many experiments in research reactors (reactor physics experiments, material test irradiations, ...) are equipped with neutron dosimetry in order to provide neutron flux and spectrum information.

        This course will first present the example of surveillance dosimetry for power reactors. Then some basic aspects of neutron physics will be shortly described. The main part of the course will focus on the neutron activation technique. This will be complemented with practical examples.

        Speaker: Dr WAGEMANS , Jan (SCK-CEN)
      • 10:30
        Coffee Break 15m
      • 10:45
        Advanced diagnostic concepts for fusion reactors 1h 15m
        Speaker: LAYET, Jean-Marc (Aix-Marseille University, France)
      • 12:00
        Sandwich Lunch 1h
      • 13:00
        Lab session 4 - High resolution X-ray radiography with proton counting detectors 1h 45m

        Short course will introduce the micro radiography system installed at the IEAP imaging laboratory. The system is equipped with the nano-focus X-ray source (FXE-160.51) and large area hybrid pixel detector (Widepix 4×5). Basics of the high-resolution imaging with cone beam geometry will be explained as well as commonly used image corrections (flat field and beam hardening). Students are encouraged to bring their own small visually attractive samples for inspection (less than 1 cm in thickness).

        Speakers: DUDAK, Jan (Institute of Experimental and Applied Physics, Czech Technical University in Prague) , ZEMLICKA, Jan (IEAP CTU in Prague)
      • 14:45
        Lab session 5 - Computed tomography 1h 45m

        Short course explains the principles of computed micro-tomography and demonstrates them using a custom-built micro tomography system installed at the IEAP imaging laboratory. The system is equipped with micro-focus X-ray source (Hamamatsu L12161-07) and flat-panel detector (Dexela 1512). Used system reaches resolution of 5 µm. Data acquisition, processing, reconstruction, and 3D visualization will be demonstrated on test samples. Students are encouraged to bring their own small samples with interesting internal structure (less than 5 cm in diameter).

        Speakers: DUDAK, Jan (Institute of Experimental and Applied Physics, Czech Technical University in Prague) , ZEMLICKA, Jan (IEAP CTU in Prague)
      • 16:30
        Coffee Break 15m
      • 16:45
        Optional Examination 1h


      • 18:15
        Proclamation 15m


    • 14:00 17:45
      Workshops #1: Nuclear instrumentation and measurement for major research facilities in the fields of fusion and fission

      The development of nuclear instrumentation and measurement methods for large research facilities in the fields of fission and fusion is crucial and is reinforced by the construction of new facilities, such as the ITER Tokamak and the Jules Horowitz Reactor, which will offer new more extreme conditions requiring accurate, reliable and miniaturized hardened sensors and detectors. Nuclear aspects and constraints will increase in particular in the field of magnetic confinement fusion due to the use of the Deuterium-Tritium reaction and the challenge of controlling and monitoring the in situ Tritium breeding for future generations of nuclear fusion reactors. The workshop will focus on recent advances in nuclear instrumentation and scientific and technological breakthroughs through invited oral presentations. The workshop will cover experimental and numerical studies carried out under laboratory conditions and in real media. The workshop will also detail issues related to future challenges. Finally, one of the objectives of this workshop will be to contribute to cross-fertilization between the fission and fusion scientific communities through discussions also.

      Conveners: Dr KOHSE, Gordon (Massachusetts Institute of Technology, Nuclear Reactor Laboratory, Cambridge, Massachusetts, USA) , REYNARD-CARETTE, Christelle (Aix-Marseille University)
      • 14:00
        Introduction 30m


        Speaker: REYNARD-CARETTE, Christelle (Aix-Marseille University)
      • 14:30
        Recent improvements in instrumentation for French experimental nuclear reactors 30m


        Speaker: Dr DESTOUCHES, Christophe (CEA, DES, IRESNE, DER, Physics Safety Tests and Instrumentation)
      • 15:00
        Measurement of nuclear absorbed dose rate by calorimeter: from present work dedicated to a future irradiation campaign in the MIT's reactor to key challenges 30m

        Nuclear research reactors support the nuclear power industry in their research and development by allowing experiments under real conditions. A wide range of research reactors exists and their modularity and flexibility permit the study of fuel or material behavior and the establishment of specific incidental or accidental conditions. For example, a Material Testing Reactor (MTR) provides a high neutron and gamma fluxes as well as strong displacements per atom. This extreme radiation environment coupled with advanced instrumentation enables the collection of data for safety issues, the lifetime of existing nuclear power plants and their developments with new concepts. The Jules Horowitz Reactor, a new 100 MWth nominal power MTR with unequalled performance in Europe, is built in this perspective. Thanks to a high fast neutron flux of 5.5x10^14 (from 1 MeV), a high nuclear absorbed dose rate (up to 20 W.g-1 in aluminum as against 13W.g-1 in graphite for OSIRIS) and a high accelerated ageing (up to 16 dpa/year) will be induced. New instrumentation is needed to measure online key parameters both before the experiments for the device design and during the experiments for the result interpretation (inside the core or in the periphery). Hence, new collaborative research work has been carrying out since 2009 between Aix-Marseille University (AMU) and the CEA within the framework of the joint laboratory LIMMEX* and several research programs. The aims are to implement and develop innovative instrumentation but also advanced online measurement methods for the quantification of key nuclear parameters such as neutron and photon fluxes and nuclear absorbed dose rates. The harsh environment and extreme conditions associated with reproducible and accurate measurement requirements (5% on the nuclear absorbed dose rate for nominal power of the JHR) implies particular sensors. As a result, non-adiabatic calorimeters are used to measure the nuclear heating rate in MTRs thanks to temperature measurements and a crucial preliminary calibration under non-irradiation conditions (steady thermal states with an integrated heating element or transient thermal states in other cases) either for differential or single-cell calorimeters.
        An overview of experimental, 1-D thermal predictive model and 3D thermal numerical works under laboratory and real conditions will be presented for the CALORRE type differential calorimeter patented in 2015. This review focuses on the recent works for both low and high nuclear absorbed dose rate (up to 20W/g and therefore never quantified). It will emphasize the advantages and drawbacks (sensor geometry and size, sensitivity, linearity, range, reproducibility and response time) and compare them to other calorimetric sensors (such as CARMEN, CALMOS, KAROLINA and the GAMMA Thermometer) with regard to the nuclear absorbed dose rate range targeted. Finally, work conducted for a new research program, called CALOR-I and involving AMU, the Nuclear Reactor of the MIT and CEA, will be detailed. The two main aims, which are the design of a new CALORRE prototype and the mapping of an in-core water-loop in the MIT reactor, will be developed.
        * Laboratory for Instrumentation and Measurement in EXtreme Environments
        compact-CALORimeter Irradiations inside the MIT research reactor

        Speaker: VOLTE, Adrien
      • 15:30
        Sensor Technologies for the European TBM Program in ITER and DEMO Relevancy Aspects 30m

        The TBM (Test Blanket Module) Program, which represents a relevant part of the experimental campaign in ITER, is the main step to validate the basic technologies to be adopted in the breeding blankets for DEMO. In the ambit of the TBM Program, Europe is engaged in the design, construction, installation and operation of the WCLL (Water Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) - TBS (Test Blanket System). The main differences between the two TBS concepts lie in the tritium breeder-neutron multiplier (molten lead lithium alloy vs. pebble beds of ortho-meta silicate and beryllium) and in the main coolant (water in PWR conditions vs. pressurized helium).
        One of the main technologies to be validated in the TBM Program refers to sensors and instrumentation, which have to operate in an environment rather different from the one traditionally experienced in the fission reactors. Their importance is not only related to the TBS operational and safety aspects but also to the reliability of the data that will be recorded all along the TBS experimental campaigns in ITER and post-processed for scientific purposes.
        The criticality of the sensors to be used in the TBS and, on a longer perspective, in the breeding blanket comes essentially from:
        - the environmental conditions, mainly high-energy neutrons/gamma fluxes and magnetic field;
        - the process fluids, particularly flowing Pb-16Li;
        - the need to carry out an accurate tritium accountancy in the whole TBS.
        All this has triggered an important R&D effort, carried out by the main European Fusion Laboratories, in the last period under the EUROfusion coordination.
        This paper is intended to provide the status of the development for two sensor families: sensors for the measurement of tritium concentration and other sensors operated in Pb-16Li. The development of these two families is of particular interest because:
        - the performance (i.e. accuracy, precision, linearity, response time) of the tritium concentration sensors in the different TBS fluids must be maximized for a correct reconstruction of the tritium distribution in the different TBS sub-systems and components. This will make possible the validation of tritium transport modelling tools, which are ones of the major RoX (Return of Experience) elements for DEMO;
        - the Pb-16Li environment is particularly critical because of the high Pb-16Li corrosion potential, high density and metallic nature. This makes the adaptation of the classical sensors of pressure, temperature and mass flow-rate rather challenging.
        In this paper, beyond the status of the development of these two classes of sensors, is included also a brief gap analysis with the objective to provide a preliminary work-plan in these areas for the next years. One can find also considerations on the DEMO relevancy of the different technologies presented.

        Speaker: RICAPITO, Italo (F4E at ITER Organization)
      • 16:00
        Coffee Break 15m
      • 16:15
        Contributions to the development of nuclear instrumentation for the EU test blanket modules for ITER 30m

        Fusion power reactors will rely on the internal production of the fuel tritium from lithium in the tritium breeding blanket (TBB). There are various concepts for TBB, in general they are heterogeneous structures including structural materials such as steels, cooling and breeder materials as well as neutron moderators and multipliers. A necessary condition for the operation of fusion reactors is a sufficient tritium production which must cover the fuel consumption, losses of tritium and an excess amount of tritium for startup. Currently there is no experimental facility available which could test a complete tritium breeding system (TBS), the present designs rely heavily on radiation transport simulations. In the last decades, experiments were performed in which very simple models of breeding blankets were irradiated with DT neutrons from neutron generators with the aim to provide an experimental data base for testing the radiation transport simulations including the utilized nuclear data libraries. The ITER reactor under construction at Cadarache in France is expected to allow testing of TBB concepts in a real fusion reactor environment. The radiation fluxes will be of comparable magnitude. Since the fusion plasma as intense neutron source extends over a reactor-typical volume, it is possible to insert so called Test Blanket Modules (TBM) with a heterogeneous structure very similar to TBB. The measurement of neutron fluxes with good accuracy in these TBM during ITER operation will be of high importance since the amount of produced tritium as well as other measured parameters such as heating for example will be related to these neutron fluxes. This will allow checking calculation tools and models utilized for TBB development with experimental data obtained under real fusion reactor conditions. The conditions in a TBM are harsh and characterized by high temperatures (>400oC), scarce space and difficult access, and, depending on the operation scenario, intense radiation levels.
        We present an overview of work conducted in the last years on three types of neutron flux monitors under development as candidates for the TBMs with focus on the KIT activities: A neutron activation system (NAS) with pneumatic probe transport, self-powered (neutron) detectors (SPD), and silicon carbide detectors. The intense DT neutron generator of Technical University of Dresden has been used for most of the irradiation tests.

        Speaker: KLIX, Axel (Karlsruhe Institute of Technology, Germany)
      • 16:45
        Neutron Detection using Silicon-Carbide- and Diamond-based Sensors: 3D Thermal Simulations and Instrumentation 30m

        The principal role of neutron detectors in nuclear applications is to get information about the actual neutron yield and reactor environment. Harsh radiation environments near the nuclear reactor core or in TBM (Test Blanket Modules) for tokamaks require the radiation detectors to be resistant to high temperature (up to 500°C) and radiation level. Currently, Si is the most used semiconductor material but its drawbacks make it impossible for harsh media applications.
        Silicon Carbide (SiC) is one of the most attractive materials for neutron detection. Thanks to its outstanding properties, such as high displacement threshold energy (22-35 eV), wide band gap energy (3.27 eV) and high thermal conductivity (4.9 W/(cm·K)), which is an advantage as it leads to low temperature gradients inside the semiconductor.
        The aim of this work is to evaluate the ability of SiC to detect neutrons, focusing on the 3D Thermal Simulations and the specific Instrumentation.
        For this purpose, the thermal neutron irradiation tests of detectors were implemented at MINERVE nuclear research reactor at CEA Cadarache (for a neutron flux equal to 9.4×10^8 n·cm-2·s-1), and fast neutron irradiation tests were carried out at TUD - Helmholtz-Zentrum Dresden- Rossendorf (HZDR), using a deuterium-tritium neutron generator (for a neutron flux equal to 3.2×10^7 n·cm-2·s-1). The results will be presented.
        Then a parametric thermal study thanks to 3D thermal numerical simulations with COMSOL Multiphysics was realized to define and design a new housing for the SiC-based sensor for future irradiation tests with a higher neutron flux value (1.2×10^13 n·cm-2·s-1) and another thermal environment in the Triangular Irradiation Channel of the Jožef Stefan Institute TRIGA Mark II reactor (dry air channel). This study was achieved by considering different heat source values covering a wide range of expected nuclear heating rate and by considering only conductive heat transfer inside the detector, natural convective heat transfer as boundary conditions and neglecting radiative heat transfer. The influence of several parameters (housing material nature, gas nature in the gap around the SiC-based sensor, height of the gas gap and thickness of the housing) was estimated on the temperature field and thus the maximum temperature inside the detector. The main technological breakthroughs will be provided, as well as the perspectives of nuclear measurements using wide-band-gap materials.

        Speaker: Dr OTTAVIANI, Laurent (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 17:15
        To be defined 30m


        Speaker: USA
    • 14:00 17:45
      Workshops #2: Crystalline Materials for Radiation Detection

      The development of efficient and environmentally-friendly technologies for radiation detection is a great challenge. Among the materials of present and future perspective are crystalline materials, wide-bandgap semiconductor crystals, in particular. The recent progress in crystal growth, theoretical modelling, understanding of radiation induced defects, and radiation hardness has offered a new perspectives for radiation detection. Wide bandgap semiconductors such as SiC, GaN, Ga2O3 and diamond are expected to be applied to radiation detectors with extremely radiation hardness, and huge efforts to develop radiation detectors based on wide bandgaps semiconductors has been spent. This Workshop is dedicated to all aspects related to the growth, characterization, and applications of crystalline materials for radiation detection with the aim to provide an overview of the issues of current interest and future perspectives.

      Conveners: Dr CAPAN, Ivana (Ruđer Bošković Institute) , Dr MAKINO, Tahakiro (National Institutes for Quantum and Radiological Science and Technology)
      • 14:00
        Radiation defects in the silicon sublattice of 4H-SiC: electronic structure and annealing behavior 30m


        Speaker: COUTINHO, José (University of Aveiro, Portugal)
      • 14:30
        Silicon-Carbide- and Diamond-based Neutron Sensors : Detection Stability under Irradiation 30m


        Speaker: Dr OTTAVIANI, Laurent (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 15:00
        Conversion pathways of primary defects by annealing in proton-irradiated n-type 4H-SiC 30m


        Speaker: KARSTHOF, Robert (University of Oslo, Norway)
      • 15:30
        Silicon carbide neutron detectors: sensitivity, energy selectivity and new converting materials – overview of research activities in the scope of the E-SiCure and E-SiCure2 projects 30m


        Speaker: RADULOVIĆ, Vladimir (Jožef Stefan Institute, Slovenia)
      • 16:00
        Coffee Break 15m
      • 16:15
        Proposal and Development of Novel Neutron Detector using Boron Gallium Nitride 30m


        Speaker: NAKANO, Takayuki (Shizuoka University)
      • 16:45
        TBD 30m


        Speaker: KRAUS, Hannes (NASA JPL, USA)
      • 17:15
        Closing remarks 30m
        Speakers: Dr MAKINO, Tahakiro (National Institutes for Quantum and Radiological Science and Technology) , Dr CAPAN, Ivana (Ruđer Bošković Institute)
    • 14:00 17:45
      Workshops #3: Forum on Prospective technologies for the future PET imaging

      This half-day forum will be oriented around the technical evolution of Positron Emission Tomography (PET) starting from the present to extrapolate to the future. It will be discussed the various technological progresses of tools and techniques in the component of this device from picosecond electronics, Time of Flight, new scintillators, use of Compton event detection, software and analysis of the data using IA as well as its use in various application in Cancer diagnostic, real time patient dose monitoring in particle therapy, small animal experimental state of the art small devices...

      We propose to invite few eminent worldwide scientific experts of that domain who made great outstanding development of tools and techniques in their respective fields that made a breakthrough in medical applications. After a short introduction of their technical work, we will ask them to develop their vision for the future. At the end, there will be a forum between them to reply to the question of the audience. We want to transform this kind of forum in an open access accessible by the maximum of students and scientific interested by the subject. We want to promote it like a webinar type where participants should register and be able to ask question remotely...

      Conveners: BOHM, Christian (University of Stockholm, Sweden) , Prof. DA VIA, Cinzia (University of Manchester, UK) , LE DÛ, Patrick (IEEE NPSS, France) , NOMACHI, MASAHARU (OSAKA University, Japan)
    • 09:00 10:40
      Opening Session
      Conveners: LYOUSSI, Abdallah (CEA) , STEKL, Ivan (IEAP CTU in Prague) , Dr HODÁK, Rastislav (IEAP CTU in Prague)
    • 10:40 11:00
      Coffee Break 20m
    • 11:00 12:40
      Keynote Lectures
      Conveners: Dr POSPÍŠIL, Stanislav (IEAP CTU in Prague) , Dr CARETTE, Michel (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 11:00
        Recent developments in sensor materials 40m MERIDIAN


        Clarion Congress Hotel Prague

        Freyova 945/33, 190 00 Prague 9 - Vysočany

        Early studies of the photoconductive properties of materials eventually led to an understanding of the solid-state and particularly of a new class of materials – the semiconductors. Their sheer number and diversity of properties give rise to a wide range of sensitivities covering almost the entire electromagnetic spectrum − from terahertz to gamma-ray wavelengths. The application of semiconductors to radiation detection and measurement through an understanding of the electronic properties of matter has come to dominate radiation physics – a field that did not exist 150 years ago. Of special interest has been the rise of compound semiconductors and more recently the organic semiconductors. Compound semiconductors have a number of distinct advantages over their elemental counterparts, Si and Ge. For example, wide-gap materials offer the ability to operate in a range of hostile thermal and radiation environments while still maintaining spectral resolutions <2% FWHM at X and gamma-ray wavelengths. Narrow-gap materials offer the potential of exceeding the spectral resolution of Ge by a factor of three. However, while compounds are routinely used as detection media at infra-red, optical and UV wavelengths, for hard X- and gamma-ray applications their development has been plagued by material and fabrication problems. So far only a few have evolved sufficiently to produce commercial detection systems. Recently there have been considerable advances in the development of organic semiconductors largely spurred on by the smart phone and tablet industries. While not destined to replace silicon-based technologies, they offer the promise of low cost, fully flexible devices for large-area displays and solid-state lighting. Much research currently focuses on solar cell and energy harvesting applications, although the hybrid organic-inorganic perovskites (HOIPs), such as formamidinium lead triiodide (CH(NH2)2PbI3), are now being explored as X- and gamma-ray detection media.

        As well as on-going activities in semiconductors, scintillator research has undergone a renaissance since the discovery of the cerium doped lanthanum halides in the early 1990’s − the most notable outcome of which, has been the recent availability of large volume (>350 cm3) LaBr3(Ce3+), LaCl3(Ce3+) and CeBr3 high resolution scintillation detectors, with FWHM energy resolutions of 3−4% being achieved at 662 keV. Recent experiments with co-doping, using activators such as Eu and Pr, suggest that a further improvement to 2% is possible. These results are comparable with the best results obtained with a compound semiconductor such as CdZnTe. At present, the outstanding issue in scintillation physics is understanding and mitigating the effects of energy non-proportionality which currently limits the attainable energy resolution in all scintillators.

        In this talk we summarize recent developments in scintillator and semiconductor research from a materials perspective. To improve performance, underlying efforts in both should continue on achieving material perfection. In the near and intermediate terms, performance improvement for semiconductors lies in the controlled and directed manipulation of charge, for example employing single carrier sensing techniques to neutralize the degrading effects of the poorest carrier, while for scintillators, it lies with the understanding and reduction of non-proportionality effects. In the longer term, it is clear that for semiconductors conventional detection techniques based on the manipulating the electrons charge using electric fields will reach an impasse in terms of sensitivity. A major step forward should be possible, by exploiting other obscure internal degrees of freedom of the electron in addition to its charge for nonvolatile information processing. For example, utilizing its spin or alternately the valley degree of freedom implicit in the band structure of some semiconductors.

        Speaker: OWENS, Alan (European Space Agency, ESTEC, Noordwijk, The Netherlands)
      • 12:00
        Getting to the point in particle therapy 40m MERIDIAN


        Clarion Congress Hotel Prague

        Freyova 945/33, 190 00 Prague 9 - Vysočany

        The usage of protons or even heavier particles like energetic carbon ions for radiation therapy was proposed 75 years ago by Robert R. Wilson. Since then, pioneering developments and applications have been pursued for several decades at several specialized institutions hosting powerful accelerators for beam production. With the advent of first hospital-based facilities in the 1990s, ion therapy, especially with proton beams, has been increasingly adopted in the clinical setting, making it rapidly emerging as a promising treatment modality in external beam radiation therapy. In particular, the favorable physical interaction properties of ion beams in matter enable concentrating the energy deposition in a well-localized maximum in depth, the so-called Bragg peak, offering optimal coverage of the tumour target volume along with better sparing of normal tissue and critical organs compared to the widely established photon therapy. Ions heavier than protons can also offer advantages from their reduced lateral scattering and localized enhancement of biological damage in the Bragg peak region due to their elevated ionization density when coming to rest.

        Over the last years, considerable developments in accelerator technology, beam delivery, treatment planning and in-room volumetric image guidance have entered the clinical routine and enabled achieving even further improved conformation of the therapeutic dose to the tumour target volume. However, full exploitation of the advantageous properties of ion beams for therapy is still hampered by several sources of uncertainties in treatment planning and delivery. This talk will thus review the state-of-the-art and ongoing efforts tailored to reduce such uncertainties, with emphasis on new instrumentation aiming at improving the pre-treatment patient model, in particular for determination of the tissue stopping power relative to water, along with in-vivo verification of the actual dose delivery, or at least beam range, during or shortly after treatment. Moreover, it will address promising new directions paving the way to innovative concepts of rapid/real-time adaptive particle therapy, along with the development of precision pre-clinical research platforms to unravel the complex underlying biological mechanisms toward the development of novel therapeutic strategies of improved efficacy with reduced side effects.

        Speaker: Prof. PARODI, Katia (Ludwig-Maximilians-Universität München)
    • 12:40 14:00
      Lunch 1h 20m
    • 14:00 16:00
      01 Fundamental Physics

      Oral Session

      Convener: GROSSMANN, Martin (PSI, Switzerland)
      • 14:00
        #01-247 ATLAS LAr Calorimeter Commissioning for LHC Run-3 20m

        Liquid argon (LAr) sampling calorimeters are employed by ATLAS for all electromagnetic
        calorimetry in the pseudo-rapidity region |η| < 3.2, and for hadronic and forward
        calorimetry in the region from |η| = 1.5 to |η| = 4.9. In the first LHC run a total luminosity of
        27 fb$^{−1}$ has been collected at center-of-mass energies of 7-8 TeV. After detector consolidation
        during a long shutdown, Run-2 started in 2015 and about 150fb$^{−1}$ of data at a center-of-mass
        energy of 13 TeV was recorded. With the end of Run-2 in 2018 a multi-year shutdown for the
        Phase-I detector upgrades was begun.
        As part of the Phase-I upgrade, new trigger readout electronics of the ATLAS Liquid-Argon
        Calorimeter have been developed. Installation began at the start of the LHC shut down in
        2019 and is expected to be completed in 2020. A commissioning campaign is underway in
        order to realize the capabilities of the new, higher granularity and higher precision level-1
        trigger hardware in Run-3 data taking. This contribution will give an overview of the new
        trigger readout commissioning, as well as the preparations for Run-3 detector operation and
        changes in the monitoring and data quality procedures to cope with the increased pileup.

        Speaker: Dr ANDEEN, Tim (CERN)
      • 14:20
        #01-122 The spectrometer MULTI for measurement of pygmy dipole resonances in exotic nuclei 20m

        The construction of new portable spectrometer "MULTI" is designed for the measurement of β-γ-neutron coincidences in search for pygmy dipole resonances in exotic nuclei. It is based on many years of experience with spectrometer for direct measurement of the total reaction cross section with radioactive beams. Experimental technique includes measurement of the ratios between gamma-ray emission and multi-neutron emission, following beta-decay of the implanted radioactive nucleus. The spectrometer consists of an in-beam multi-detector telescope for identification of the secondary beam projectiles and a gamma-ray spectrometer with neutron counters array, surrounding the β-detector. The in-beam part consists of active collimators, silicon dE0 detector for particle identification by dE:TOF (time of flight) method and β-detector for implantation of the nuclei of interest and measurement of electrons emitted in β-decay. Gamma-rays and neutrons are measured by eight CeBr3+NaI(Tl) phoswich scintillation detectors, surrounded by 40 neutron counters in form of 3He tubes in polyethylene blocks. The design was optimized with Monte Carlo method in Geant4. Data acquisition electronics is based on the digital pulse processing modules in VME standard, including pulse-shape analyzing techniques for identification of the particle type. Project and construction of the spectrometer, calibration measurements of its fundamental characteristics and the method of measurement with radioactive beams will be presented.

        Speaker: Dr SIVÁČEK, Ivan (Flerov Laboratory of Nuclear Reactions, Joint Institute for Nuclear Research, 141980 Dubna, Russia)
      • 14:40
        #01-228 A high speed data link optimization for digitalized data transfer to processing FPGAs 20m

        State-of-the-art arrays of detectors, that require digital processing, may have a sizeable number of digitalized signal links. This is the case in several experimental nuclear physics instruments. Moreover, the sampled signals data rate, defined primary by the signal bandwidth of the individual detector, may not exhaust the capabilities of a single FPGA transceiver input.
        This is a critical issue in position sensitive HPGe segmented detectors as AGATA, where each crystal detector has 36 segment signals plus a core signal. The 37 signals for each crystal detector are digitalized and processed with an independent electronics. The sampling is performed with 14 bit (about 12 ENOB) at 100MHz due to the bandwidth limitation to about 30 MHz of the detector and preamplifier response. Each ADC has a link to the preprocessing system at 2Gbps in a JESD204 protocol.
        The preprocessing is carried out in a modern FPGA with transceiver data rate capabilities over 10Gbps. Moreover, cost effective FPGA have a limited number of transceivers for a given FPGA processing capabilities. The investigation of a cost-effective and efficient solution to the mismatch between both data rates, optimizing the use of the FPGA resources, is the topic of the present work.

        We have developed a solution based on the Time Domain Multiplexing link aggregation, in the form of a Mezzanine board. This mezzanine combines four channels from an optical or copper input up to 2.5 Gbps to one up to 10Gbps, and serves them to the FPGA via the mezzanine connector. The board itself is controlled by a small FPGA by Two Wire Interface (TWI) protocol as a standalone intelligent device, with minimum slow control needed. An associated firmware has been developed to de-aggregate the data in the FPGA and recover the original digitalized data, with the JESD204 cores, inside the FPGA. The method has been validated and applications, beyond the development of the AGATA electronics, may be envisioned.

        Speaker: Dr COLLADO, Javier (Instituto de Física Corpuscular (IFIC) - CSIC / Universidad de Valencia - ETSE)
      • 15:00
        #01-98 The Data Acquisition System for the ATLAS Phase-II Tile Calorimeter Demonstrator 20m

        he Tile Calorimeter (TileCal) is the central hadronic calorimeter of the ATLAS detector at the Large Hadron Collider (LHC). This subdetector is a sampling calorimeter composed of steel plates as absorber material and plastic scintillators as active material. The complete readout of the detector is done using approximately 10,000 photomultipliers (PMTs).
        The LHC will undergo a series of upgrades in 2025 leading to the High-Luminosity LHC (HL-LHC). The HL-LHC will provide an instantaneous luminosity 5 to 7 times larger than the nominal LHC design value. The ATLAS Tile Calorimeter Phase-II Upgrade (2025-2027) will completely replace the readout electronics with a new clock distribution and readout architecture with a full-digital trigger system.
        In the upgraded data acquisition architecture, the on-detector readout electronics will transmit digitized signals from the PMTs for every bunch crossing (~25 ns) to the Tile PreProcessor (TilePPr) boards located in the counting rooms. The TilePPrs will store the detector data in pipeline memories until the reception of a trigger acceptance signal activating data transmission to the ATLAS Front End LInk eXchange (FELIX) system. The new readout system will require 32 TilePPr modules to read out the entire detector for a total bandwidth of 40 Tbps. In parallel, the Trigger and Data Acquisition Interface boards will receive reconstructed cell energies from the TilePPrs and will transmit pre-processed trigger objects to the first level trigger with improved precision and granularity.
        As part of the Phase-II Upgrade Demonstrator program, a full-size Demonstrator module containing all the upgraded readout electronics was installed into the ATLAS experiment in 2019 with the aim of validating the new clock and readout strategy for the Phase-II Upgrade.
        The Demonstrator module consists of four independent mini-drawers capable to operate up to 12 PMT blocks. A mini-drawer is composed of a mechanical aluminum structure that supports one Mainboard, one Daughterboard, one high voltage regulation board, and up to 12 PMT blocks equipped with 3-in-1 cards. The PMT signals are shaped and amplified in two gains by the 3-in-1 cards and digitized in the Mainboard by 12-bit dual ADCs. The Daughterboards transfer the digitized data to the off-detector electronics through high-speed optical links. In addition, the lower gains of the 3-in-1 cards are summed in towers and transmitted to the Level-1 Calorimeter trigger system for trigger decision.
        In the off-detector electronics, the Tile PreProcessor prototype receives the detector data for every bunch crossing through optical links with fixed and deterministic latency which provide a total data bandwidth of 160 Gbps. The main FPGA of the PreProcessor prototype processes and buffers the detector data in pipeline memories capable of storing up to 10 $\mu$s of samples. Upon the reception of a trigger acceptance signal, the selected events are extracted from the pipeline memories, packed, and transmitted to the legacy Read-Out Drivers (RODs) keeping backward compatibility with the ATLAS DAQ system, and to the FELIX system.
        The PreProcessor prototype also distributes the LHC bunch-crossing clock embedded with configuration commands to the on-detector electronics for synchronization with the accelerator. In the current Demonstrator, the Preprocessor prototype interfaces with the legacy Timing, Trigger and Control (TTC) system to receive the LHC clock and configuration commands. The recovered clock is then cleaned with dedicated jitter cleaner chips before driving it to the PreProcessor FPGA transceivers to ensure stable high-speed communication with the Daughterboards.
        The Demonstrator module is fully integrated with the ATLAS TDAQ software and Detector Control System through the PreProcessor prototype, which translates the legacy commands into Phase-II commands, and transmits the triggered data to the RODs using the G-Link protocol. Therefore, the PreProcessor prototype enables the operation Demonstrator module for taking calibration and physics runs using the current ATLAS software tools. From the point of view of the ATLAS TDAQ system, the Demonstrator module behaves as a legacy TileCal module.
        The Demonstrator module has been operated together with the rest of the TileCal modules since its installation in June 2019. The performance of the upgrade electronics has been studied with Charge Injection, Laser, and Cosmic runs, showing excellent performance in terms of low noise, signal quality, and timing. One of the goals of the Demonstrator program is to keep the Demonstrator module in ATLAS during Run-3 (2021-2024) to continue studying its performance in real conditions.
        This contribution describes in detail all the hardware, firmware, and software components of the clock distribution and data acquisition system for the Demonstrator module, focusing on the PreProcessor prototype developments, as well as the results of the Demonstrator module operation in the ATLAS experiment and future plans.

        Speaker: CARRIÓ ARGOS, Fernando
      • 15:20
        #01-191 Preliminary results of the CGEM-IT of the BESIII experiment 20m

        The BESIII (Beijing Spectrometer III) experiment is hosted at the leptonic collider BEPCII (Beijing Electron Positron Collider II) at IHEP in Beijing and it is successfully running since 2009. Recently, an extension of the data taking for the next 10 years has been approved, so an upgrade program has started. A central element of the upgrade is the replacement of the present inner tracker with a new detector based on the cylindrical GEM (Gas Electron Multiplier) technology. The CGEM-IT detector is composed of three layers of cylindrical triple-GEMs with time and charge readout to guarantee excellent performance in a wide range of incident angles in 1 Tesla magnetic field. Two of the final layers are already at IHEP and are undergoing commissioning with cosmic rays events. The third layer is being completed by end of 2020 and will be taking data with the other two by the summer of 2021. The installation is foreseen in Summer 2022. In this presentation, the general status of the project will be presented with a particular focus on the recent preliminary results from the cosmic data taking and future plans.

        Speaker: MEZZADRI, Giulio (INFN - IHEP Fellow)
      • 15:40
        #01-84 Measurement of anomalies in angular correlation of electron and positron internally produced in excited 8Be and 4He 20m

        Theoretical prediction for the distribution of the angle between electrons and positrons originating in internal pair creations is a monotonic featureless decrease with the opening angle. Recent studies on excited states of 8Be and 4He nuclei, made in ATOMKI, Hungary, however, revealed deviations from this expectation. If true, such a result may have a fundamental impact: the anomaly can be explained by introducing a new short-lived neutral boson that can still fit into known experimental and theoretical constraints. Although serious work has been done on the theoretical side, an independent laboratory has not yet verified these results yet, although related experiments are being prepared worldwide. We present the ongoing construction of a suitable time-projection-chamber-based (TPC) spectrometer for light charged particles, utilizing magnetic field as a means for energy measurement and also Multiwire Projection Chambers (MWPC) together with Timepix3 pixel detectors, for unprecedented spatial and angular resolution. The experimental effort will be operated at the Institute of Experimental and Applied Physics (IEAP) Van-de-Graaff accelerator facility in order to confirm or either refute the above-mentioned anomaly. Details of the detectors will be shown, together with simulations that provide information on the expected performance of this system.

        Speaker: Dr PETOUSIS, Vlasios
    • 14:00 16:00
      04 Research Reactors and Particle Accelerators

      Oral Session

      Conveners: MUTTI, Paolo (Institut Laue-Langevin) , HORTVATH, Akos
      • 14:00
        #04-173 Progress and prospects on nuclear instrumentation development achieved through JSI – CEA collaboration 20m

        The collaboration on nuclear instrumentation started more than ten years ago between the CEA Experimental Physics, Safety experiment and Instrumentation Section and the Reactor Physics Division of the Jožef Stefan Institute in the frame of bilateral agreement between CEA and the Slovenian Ministry of Higher Education, Science and Technology started in 2008.
        More than ten development projects have then been successfully achieved on several subjects: miniature fission chambers, Self-Power Neutron Detector (SPND) and gamma-ray measurement techniques (using ionization chambers and Thermo-Luminescent Detectors (TLDs)), kinetic parameter measurement techniques (eff), reactor dosimetry unfolding techniques and nuclear data improvements.
        Since 2018, three common projects have been started: the first is dedicated to the calorimetry measurement techniques testing different materials sensitive to nuclear heating, the second project concerns improvement of neutron and gamma-ray measurement during power transient pulses and the last aims at establishing an experimental benchmark for a modelling scheme of neutron and gamma-ray sensors. After a short presentation of the two organizations’ expertise in nuclear instrumentation and the aims and status of these three projects, the paper presents several themes identified as possible topics for future joint projects. The following topics will be addressed to cover the needs of experimental measurements for the future Jules Horowitz reactor reactor and fusion facilities: improvement of off-line neutron field characterization with the search for new inelastic scattering nuclear reactions for epithermal neutron dosimetry, and on-line, with the study of solid-state silicon carbide (SiC) based neutron sensors and the use of optical fibers as neutron or gamma-ray sensors. The development of liquid neutron filters, containing low concentration aqueous solution of boron, is considered as promising for experimentally simulating energy shifts in the thermal neutron peak in the neutron spectrum and thus allowing to obtain experimental data to support nuclear data evaluation and validation for various actinides.

        Speaker: Mr SNOJ, Luka (Jožef Stefan Institute)
      • 14:20
        #04-167 First in-core gamma spectroscopy experiments in the zero power reactor CROCUS 20m

        Gamma rays in nuclear reactors, arising either from nuclear reactions or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data can contribute to models, code validation and photo atomic/nuclear data evaluation. To date, most experiments have relied on flux measurements using ion chambers or spectrometers set into low flux areas. The CROCUS reactor allows for flexible detector placement in the core, and has recently been outfitted with gamma detection capabilities to fulfill the need for in-core gamma spectroscopy, as opposed to flux. In this paper we report on the experiments and accompanying simulations of gamma spectrum measurements in a zero power reactor core. The CROCUS reactor is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). With a maximum power of 100 W, it is a zero power reactor used for teaching and research. Herein we also introduce, in detail, the new LEAF system: A Large Energy-resolving detection Array for Fission gammas. It consists of an array of four detectors -- two large 127x254 mm Bismuth Germanate (BGO) and two smaller 12x50 mm Cerium Bromide (CeBr$_3$) scintillators. We describe the calibration and characterization of LEAF followed by first in-core measurements of gamma ray spectra in a zero power reactor at different sub-critical and critical states and different locations. The spectra are then compared to code results, namely MCNP6.2 pulse height tallies. We were able to distinguish prompt processes using photon production tracking, and delayed peaks from decay databases. The results indicate the possibility of on line isotope tracking and burn-up validation. We provide the data as validation means for codes that attempt to model these processes for energies up to 10 MeV. We finally draw conclusions and discuss the future uses of LEAF.

        Speaker: Dr PAKARI, Oskari (EPFL)
      • 14:40
        #04-190 High resolution measurements with miniature neutron scintillators in the SUR-100 zero power reactor 20m

        Three 0.95-mm$^2$ miniature fiber-coupled scintillators have been used to perform cm-wise resolution measurements of the thermal neutron flux within experimental channels of the SUR-100 facility, a zero power thermal reactor operated by the Institute of Nuclear Technology and Energy Systems at the University of Stuttgart. The detection system is developed at the École polytechnique fédérale de Lausanne in collaboration with the Paul Scherrer Institut.

        Reaction rate distributions are measured along the experimental channels I and II of SUR-100, which cross the reactor at the center and tangentially to the core, respectively. The results for the experimental channel I are compared to neutron activation measurements. In addition, reaction rate gradients across the 2.6 cm and 5.4 cm diameters of the channels are measured. The reactor was modelled with the neutron transport code Serpent-2.1.31, and the collected experimental data are compared with Monte Carlo simulations.

        The comparison of experimental and computed reaction rate distributions showed a good agreement within the core region, with discrepancies within 2$\sigma$. An unexpected discrepancy, probably caused by a geometric inconsistency in the computational model of the reactor, was observed in the reflector region of the experimental channel I, where a 20$\%$ difference (i.e. 8$\sigma$) was found between experimental and simulated results. Significant discrepancies, respectively worth 10$\sigma$ and 15$\sigma$, were noticed at distance, in the lead shielding region, for both experimental channels I and II.

        An horizontal reaction rate gradient of (9.09 $\pm$ 0.20)$\%$ was measured within 2.4 cm across the diameter of the experimental channel II, with a difference from computed results of 2$\%$. The absence of a vertical reaction rate gradient inside the experimental channel I was confirmed by measurements.

        The performed measurements highlighted the suitability of the miniature fiber-coupled scintillators for high resolution reaction rate distribution measurements and for the characterization of highly localized gradients in reaction rate. The negligible flux perturbation induced by the miniature fiber-coupled detectors has been found to be a great advantage compared to the previously used manganese activation techniques. In the case of the activation technique, the presence of the samples holder caused a maximum 35$\%$ increase of thermal neutron flux in the core region of experimental channel I.

        Speaker: BRUNETTO, Edoardo Luciano (École polytechnique fédérale de Lausanne (EPFL))
      • 15:00
        #04-185 Calibration of CFUL01 fission chambers in the standard neutron fields of BR1 reactor at SCK CEN 20m

        Recent subcritical VENUS-F experiments showed that fission chambers with a threshold deposit like U-238 can essentially improve the on-line sub-criticality monitoring with the beam interruption method, which is currently supposed to be the main method for the ADS MYRRHA. To suppress the uncertainty caused by fissions in the U-235 impurities, the fraction of U-235 in the U deposit should be accurately known. Three PHOTONIS CFUL01 type fission chambers with a high purity of U-238 were purchased for sub-critical experiments in the VENUS-F reactor. To verify the nominal purity (U-235 content) of the deposits ofthese fission chambers, the effective U-235masses were measured in 2017 in the empty cavity of the BR1 reactor with a well-known thermal neutron spectrum. The measured effective mass was determined as the mass of the isotope corresponding to a certain discrimination level and is obviously expected to be smaller than the total mass of the isotope, especially for heavy, thick deposits, which is the case for the fission chambers under investigation. It turned out that the measured effective U-235 mass in two fission chambers is lower than the nominal (as it should be), but this is not the case for the U-235 content in the third fission chamber. To obtain the correct of U-238 content in these three FCs, the effective U-238 mass was measured in the well-known fast spectrum of the MARK-III convertor in the BR1 reactor. Finally, the isotopic composition was obtained by dividing the U-238 effective mass (measured in MARK-III) by the U-235 effective mass (measured in the empty cavity) at the same discrimination levels. It was found that the purity of two CFUL01 FCs is in agreement with the values declared in the certificates but it is not the case for the third fission chamber. The statistical and systematic uncertainties and corrections applied in the data analysis are discussed. The developed procedure using the BR1 standard irradiation fields can be of interest for calibration and impurity determination of large fission chambers.

        Speaker: Dr KOCHETKOV, Anatoly
      • 15:20
        #04-184 Local and high distance neutron and gamma measurements of fuel rods oscillation experiments 20m

        Within the COLIBRI program, reactor noise related to fuel vibrations is investigated in the CROCUS zero power reactor. It consists in experiments on rod lateral displacement (static) and oscillation (dynamic) with different rods’ numbers at various relevant amplitudes and frequencies. Its main motivation is the increased amplitudes in the neutron noise distributions recorded in ex- and in-core detectors that have been observed in recent years in Siemens pre-Konvoi type of PWR reactors. In particular, with this program, the Laboratory for Reactor Physics and System behaviour (LRS) is contributing to the Horizon 2020 European project CORTEX, which is dedicated to the understanding and simulation of reactor perturbations for the development of novel core monitoring techniques.

        During the first phase of COLIBRI, the observation of a spatial dependence of the induced noise, also called neutron modulation, was demonstrated. In the second phase of COLIBRI starting 2021, it is planned to use a core mapping array of neutron detectors to record its propagation. It consists in about 150 miniature scintillators coupled to optical fibers and SiPM readouts, to be distributed in the reactor core. As a feasibility test, experiments were performed using a miniature scintillator prototype placed on one moving fuel rod, or the one directly adjacent to it. In addition, it is theoretically possible to measure branching or induced reactor noise using gamma radiation. Following recent developments on gamma measurements in CROCUS, the fuel oscillation was simultaneously recorded with a gamma detection array, LEAF. Its two large and high efficiency BGO detectors were used by placing them at the maximum distance to the core, i.e. seven meters away with a clear line of sight using an experimental channel through the reactor cavity.

        We report here on the successful observation of the lateral oscillation of one fuel rod ±2.5 mm around nominal and 0.1 Hz frequency, using the miniature neutron scintillator at the rod level, and the BGO gamma detectors seven meters away from the reactor core.

        Speaker: LAMIRAND, Vincent (Ecole Polytechnique Fédérale de Lausanne)
      • 15:40
        #04-176 Design of a 150-miniature detectors 3D core-mapping system for the CROCUS reactor 20m

        The interest of the scientific community in studying the three-dimensional propagation of induced perturbations inside nuclear reactor cores and their prediction with deterministic codes triggered the need for experimental full-core mapping systems. In this optic, the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the École polytechnique fédérale de Lausanne (EPFL), Switzerland, has been designing a 3D core-mapping system for its zero-power research reactor CROCUS. The system is composed of 150 miniature neutron detectors distributed within the core double lattice at three different axial levels. The detector technology is based on the well-proven coupling of a miniature ZnS:Li$^{\text{6}}$(Ag) scintillator to a state-of-the-art silicon photomultiplier (SiPM) via jacketed optical fibers. The challenges in the design of such an extensive mapping system are numerous and the synergy between detector physics, core criticality, mechanical design and electronics acquisition is fundamental.
        In CROCUS, the limited space between fuel rods and the need for positioning tools required a downsizing of the detector dimensions compared to the prototype version. The scintillator screen volume was reduced from 0.2 mm$^{\text{3}}$ to a volume <0.1 mm$^{\text{3}}$ and the fiber diameter was decreased from 3 mm to 1 mm. This detector arrangement was chosen as a cut-off between the need for size reduction and the preservation of the minimum level of neutron sensitivity required. In addition, smaller scintillator screens reduce the neutron absorption and consequently the reactivity impact on the core. Preliminary results showed that a single of these <0.1mm$^{\text{3}}$ detector positioned at the core periphery contributes to a reactivity reduction of around 0.7 ± 0.2 pcm. Based on this experimental result, a modelling of the full-system via the MCNP6.2 Monte Carlo code allows to assess the total reactivity impact given by the 150 detectors spread in the CROCUS core and the eventual need for reactivity compensations.
        The 150 detectors, split among three identical axial layers, will be arranged to have the best mapping of the induced perturbations in the CROCUS core, e.g. fuel rod oscillations within the COLIBRI program, absorbers insertions or rotations, etc. The mechanical system to held in positions the 150 detectors was designed to be as practical as possible in the manipulation, of easy manufacturing and respectful of the safety standards for nuclear reactors. The detectors will be held in place through 3-mm thick plastic guides running between fuel rods and never entering in contact with the fuel cladding. The optical fibers will be glued to the plastics guides with epoxy and their front-end, consisting in the detector active zone, will be glued and embedded within the plastic itself in order to avoid any unwanted movement of the detectors or loss of neutron absorbers in the core.
        An essential role is played by the acquisition system, which was considerably upgraded with respect to the analog read-out used in the prototype. The need for the processing and acquisition of data from 150 channels induced the development of in-house electronics tailored for the application to the 3D mapping system. Stand-alone electronics modules, each able to process the light signal coming from 32 detectors, were designed and built at LRS. The output signals of these 32-channels modules, consisting in the photon counting in every channel, is sent to an FPGA board equipped with a custom firmware that performs a software-based processing of the photon counting into neutron counting.
        The installation of the full-core mapping system for the CROCUS reactor is planned for spring 2021 and it will mark a milestone in proving the feasibility of such first-of-a-kind system, and in providing valuable localized experimental data for the validation of high-fidelity neutronics codes.

        Speaker: Ms VITULLO, Fanny (École polytechnique fédérale de Lausanne (EPFL))
    • 14:00 16:00
      07 Nuclear Fuel Cycle, Safeguards and Homeland Security

      Oral Session

      Convener: FAVALLI, Andrea (Los Alamos National Laboratory, USA)
      • 14:00
        #07-133 Unmanned Aircraft Systems Based Radiological Mapping of Buildings and Objects 20m

        This paper is focused on the radiological mapping of buildings and structures/objects by utilizing unmanned aircraft systems (UAS), including multicopters in particular. These platforms make possible measuring data in close distance to studied objects and with adjustable altitudes, in contrary to terrestrial and other aerial assets. Tasks such as inspection of illegal transportation or storage of radioactive nuclear material, search for uncontrolled radioactive sources, detailed survey of buildings and objects for possible contamination, monitoring of nuclear facilities (e.g., nuclear repositories) require accurate and up-to-date topographic information on the area of interest.

        Performing the detailed radiological characterization or mapping requires the aircraft to be equipped with a camera and a radiation detection system. The camera is utilized to take a series of aerial pictures of the object in order to reconstruct its three-dimensional (3D) model via photogrammetric techniques. In addition, in the case automatic flight mode is employed, an accurate georeferencing is essential.

        The radiation data is preferably acquired following a regular grid around the building/object whenever the circumstances allow it. Knowledge of the model and georeferenced data points enables to project the measured values onto surfaces where they are then subjected to interpolation based on the Delaunay triangulation. The mapping of the gamma dose rate can be superposed to the reconstructed 3D model of the inspected area or building/object. The map may further be converted in the form of isodose areas. Radionuclide identification is also possible when using a detection system with gamma spectroscopy capabilities.

        We report in this paper on the development and testing of a radiological mapping system based on the use of a DJI Matrice 210 v2 drone equipped with Zenmuse Z30 camera and a lightweight radiation detection system. The latter device is based on NaI(Tl) and CeBr3 scintillators, coupled with silicon photomultipliers. Its design includes a digital pulse processor, a laser altimeter and a GNSS receiver providing precise synchronization of the dose rate values with position and altitude or distance from the object, respectively. Real-time data processing is also feasible. A built-in RF module and a small battery ensure the complete independence of the detection system from the used aircraft.

        The goal of the research is to provide experimental verification of simulation results performed during earlier studies. The utilization of UAS in the field of radiation mapping shows promising results, however, it is still challenging to find an optimal trade-off in the size and weight of the detector. Heavier instrumentation grants better sensitivity, conversely, it reduces the flight time and therefore challenges the data collection process. Poor reception of GNSS signal in the vicinity of walls also remains an open issue.

        This work demonstrates the capabilities of using the UAS based radiation detection technology for radiological mapping of buildings and objects with possible applications in areas such as nuclear safety and security, radiation protection and environmental monitoring.

        Speaker: Mr LAZNA, Tomas (Nuclear Science and Instrumentation Laboratory, Physic Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency Seibersdorf, Austria)
      • 14:20
        #07-206 Safeguards investigations based on gamma spectrometry to determine the activity ratio of fission products in spent fuel assemblies of VVER-440 Nuclear Reactor 20m

        Safeguards investigations deal with the inspection of fission materials and their goal requires a reliable experimental verification method to determine the power history of spent fuels. In nuclear power plants it is a key task to continuously monitor the fuel burnup, for this purpose different types of verified deterministic and stochastic codes are applied. For both tasks, it is absolutely necessary to validate experimentally the theoretical models and calculations, for which one of the most appropriate methods is in-situ gamma spectrometry.
        High resolution gamma spectrometry measurements have been carried out on the VVER-440 type spent fuel assemblies of Paks Nuclear Power Plant, Hungary. The purpose of the measurements was to verify the final burnup values calculated by the method of the power plant’s staff. Until now, gamma spectra of more than 100 fuel assemblies were collected and analysed by using coaxial high-purity germanium semiconductor detectors. The majority of the measured spent fuel are so called work assemblies, but few of the absorber follower assemblies were also measured. The measured assemblies have varying initial enrichments (1.6 – 4.7%), cooling times (1 – 5 y) and operational histories, the latter resulting in differing levels of burnup. All of the hexagonal assemblies were measured from either all six or three non-neighbouring sides, which means that some information regarding the whole assembly can be derived. Additionally, the data collected throughout the years can be used for safeguards method investigation and development as well.
        Due to the measurement conditions and the relatively short cooling times, only prominent gamma emitting fission products are able to be measured. However, since the production of these isotopes depends solely on the fuel’s initial composition, geometry and operational parameters, their final activity is connected to the relevant safeguards quantities, namely the burnup, initial enrichment and cooling time. By using high resolution gamma spectrometry, we were able to detect the 137Cs, 134Cs, 154Eu, 106Ru, 144Ce, 125Sb, 110mAg, 95Zr, 103Ru and 91Y fission products. The last three isotopes could only be measured in assemblies with cooling times shorter than two years, due to their short half-lives. The difficulties raised by an absolute calibration procedure limit us, at the present, to use a relative efficiency calibration method. With such a “self-calibration” technique, only activities relative to some other isotope can be calculated.
        In the present study, we examine our dataset to reveal connections between the measured activity ratios and the three relevant quantities by utilizing model functions, dependent on the burnup, initial enrichment and cooling time, and statistical methods such as cluster analysis. The clustering of certain isotopic activity ratios shows at a glance which assemblies have similar or dissimilar initial compositions and operating histories. We also investigate the axial distribution of fission product activity ratios from assemblies measured in several different axial positions. From this, we can determine the burnup level seen by the detector at a given axial position of the fuel and compare that to the known average assembly burnup.

        Speaker: Mr KIRCHKNOPF, Péter (Centre for Energy Research (Hungary))
      • 14:40
        #07-166 Simulation and Development of Prototype Simplified Neutron Scatter Camera for Nuclear Safeguards Applications 20m

        Key in the international safeguards regime is the use of radiation detectors to track and characterize nuclear material. A relatively new area of interest in detector design is directional detectors: detectors that can report information on radiation source location and distribution along with data on source strength and identity. Neutron scatter cameras are a type of directional neutron detectors that rely on multiple neutron scatters to generate images that can reveal the direction and distribution of neutron sources. Fast neutron cameras which have recently been developed rely on multiple detector volumes and make use of neutron time-of-flight measurements. These designs, though effective in localizing the source direction, relies on a large amount of detection and electrical equipment, thus increasing size, cost, and complexity of the systems to unreasonable levels for some applications. This project seeks to develop a compact scatter camera that is less expensive than systems relying on multiple detector volumes. Crucially, two components and capabilities are needed to achieve this: fast scintillation detection materials and picosecond electrical pulse timing. Utilizing such electronics, distinguishing between scintillation light pulses generated by the same neutron within one detector volume is possible. An MCNPX-PoliMi model of such a system has been developed to guide prototype designs. A cube of EJ230 fast plastic scintillator and six photomultiplier tubes (PMTs) were used to construct the prototype camera that localizes neutron sources based on the principle of cone back projection. Neutron scattering positions within the detector volume are found by comparing the timing and quantity of light arriving at PMTs mounted to opposing faces of the scintillator volume. The use of a digitizer with a sampling rate of 5 GSPS allows for the identifications of secondary features on light pulses, indicative of secondary scattering events. Prototypes of the system in one, two, and three dimensions have shown promising initial results when coupled with a script that algorithmically identifies candidate neutron double scatter events and back projects probability cones in the direction of possible sources. Imaging resolution/quality, double scatter efficiency, and cost for the system are quantified. Paths forward for further improvement of a future system based on this camera’ operating principles are discussed.

        Speaker: HARVEY, Taylor (University of Florida)
      • 15:00
        #07-88 Neutron time-of-flight technique for nuclear material localization using plastic scintillators 20m

        Reprocessing, nuclear fuel fabrication, or uranium enrichment require large facilities that contain many glove-boxes, tanks and pipes where uranium and/or plutonium can build up. During the cleaning and dismantling of these facilities, the quantity of fissile materials need to be evaluated for nuclear material accounting and control, criticality safety, radiation safety and waste management. Holdup measurements are usually performed using gamma-ray techniques, although neutron measurements with 3-He detectors are also used. During the measurement, many factors can influence the result, such as the location of the source, its distribution and isotopic composition.
        Neutron time-of-flight event detection and fast coincidence counting are powerful techniques that can be used to discriminate neutron from gamma and establish the location of nuclear materials. MCNPX-PoliMi is used to model Pu-240 source in a glove-box and an array of plastic scintillators around the inspected object. A post-processing macro based on ROOT simulates the response of the acquisition system (energy threshold, energy resolution, time jitter, etc.). The neutron time-of-flight technique is performed to identify fission signatures. Afterwards, using time-of-flight distribution, the location is established and activity calculated.

        Speaker: Dr DEYGLUN, Clément (IRSN)
      • 15:20
        #07-78 Characterization of a Special Nuclear Material fast identifier in respect to ANSI and IEC standards to be used for nuclear interdiction to detect the presence of threat items 20m

        The increase of concerns in global nuclear security has led to the development of advanced systems for the measurement and identification of radioactive materials. Nowadays different radioisotope identification devices (RIIDs) are commercially available, aiming at deterring and combating the illicit trafficking of radioactive materials and their possible misuse in criminal acts, with catastrophic scenarios if used in public areas. For this reason, the most efficient and rapid instruments possible are highly required in homeland security applications in order to detect hidden radioactive materials and identify them, in case of alarm.
        This paper describes the latest tests performed with Sniper-GN, a novel portable RIID. This instrument has been designed as a mobile backpack-based device for the detection and identification of both gamma and neutron sources, even multiple ones. Its singular feature is the capability to identify sources through the detection of neutrons, discriminating spontaneous fission sources (like 252Cf), gamma-n type sources (like AmBe and AmLi), and nuclear material containing plutonium or uranium, in different enrichment grade. The entire system is a rugged, battery-powered device, and it has been optimized to fit in a backpack, with a total weight smaller than 10 kg. It contains the detectors and all the electronics necessary for their operation and the consequent data processing.
        The presence of only passive detection units considerably simplifies the system, if compared with the common methods of special nuclear materials investigation. Nevertheless, Sniper-GN has the capability of identifying neutron sources thanks to the novel specifically created algorithm, based on the simultaneous detection of fast neutrons and gamma rays coming from the target sources. Currently, this device is the only one in the market showing this unique feature. Several measurement campaigns have been carried out in different laboratories to test the device for the identification of nuclear material. In particular, plutonium and uranium sources as well as $^{252}$Cf, Am-Be and Am-Li samples have been measured with 1 minute long identifications. Thanks to its novel algorithm, Sniper-GN has proved to outreach the requirements of international standards. In particular, plutonium samples have been identified up to 10 and 2.5 times the reference source-to-device distance, according to ANSI N42.34-2015, IEC62327:2017, and the more restrictive ANSI N42.43, respectively. Nuclear material has also been measured by applying different shielding materials (both lead shielding and neutron moderators) or gamma masking sources. With a novel patent-pending procedure, Sniper-GN is able to distinguish the different kinds of shielding scenarios surrounding the target source. It has succeeded when gamma spectroscopy has failed due to the presence of the shielding materials. For example, plutonium sources, shielded with 5 cm of lead and 5 cm of polyethylene, have been identified up to 5 times the reference distance for neutron sources (IEC62327:2017). Furthermore, Sniper-GN warns the user of the possible presence of gamma sources masking the nuclear material and provides prompt information regarding the enrichment grade of uranium or plutonium sources, with particular concern on weapon grade uranium and plutonium.

        Speaker: MORICHI, Massimo (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy)
      • 15:40
        Empty 20m
    • 14:00 16:00
      08 Decommissioning, Dismantling and Remote Handling

      Oral Session

      Conveners: Prof. JOYCE , Malcolm (Lancaster university ) , Prof. MÉGRET, Patrice (University of Mons)
      • 14:00
        #08-222 RADLOC: A Fast Responsive Handheld Radiation Compass 20m

        This paper introduces a new approach for radiation localization and its implementation in the context of the CEA project called RADLOC (RADiation LOCator). The RADLOC development tends to ease the radioactive source localization by the use of a complete handheld direction-sensitive sensor based on user movements.
        Localization of radioactive sources is currently a main issue addressing different fields, including nuclear industry, nuclear decommissioning as well as homeland security. The objective of the RadLoc project is to propose a system capable of quickly locating a radioactive source. When talking about localization, the first step is to go in the right direction. Existing systems can be divided in two categories. The first one, focused on the sensor part, is composed by expensive solutions, which require a high level of expertise to be handled. Most recent examples are based on multiple sensors [1], [2], [3] to achieve coincidence processing or covers 360 degrees measurements [4]. The second one, most commonly encountered, is based on the use of ictometer-type detector (e.g. Geiger-Müller) by observing the count rate displayed on a screen, a visual signal (LED) or audible signal (buzzer). With the detector in hand, the user scans the surrounding space in order to vary the count rate indicated and thus obtain an initial indication of direction. In this case of concrete application, the ability of the user to gauge the evolution of the signal (visual or audible) over time and to take an arbitrary direction is necessary for the method to work. However, these systems never benefit from the movement that naturally occurs when brought by the handler. Based on this idea, this paper proposes a low-cost and lightweight measurement system able to perform the same function automatically by adding the notion of movement, resulting on a direction-sensitive detector able to find the closest minimum distance between the starting point and the location of the radioactive source.
        The RADLOC system is built around four orthogonal sensors, a four channel ad-hoc digital pulse processing (DPP) board, an embedded PC, an Inertial Measurement Unit (IMU), a screen and all required voltage supplies. Each measurement channels is composed by an YSO(CE) scintillator (0.28x0.28x0.83 inch) coupled with SiPM, directly connected to Analog to Digital Converters (ADC) sampled at 250 MHz on the DPP board. A Field Programmable Gate Array (FPGA) achieves sampled data acquisition and pulse processing. Configurable DPP processing units such as smoothing, Baseline Restoration (BLR) or fast timing discrimination for pulse feature extraction are performed (count rate, pulse height or width). Then, features are transmitted to a microcontroller for frame building and serial communication with the embedded PC, a Raspberry Pi. Then, the Raspberry Pi carries out the angular position reading with an IMU and displays on the screen the direction to take following the example of a compass. For the first version of the prototype, two measurement criteria are used to make an elementary decision-taking algorithm and plot the right direction. The first is the difference of counting rates between each measurements channels, with calculation of the barycentre. The second is the counting rate variation between the detectors during the user's movements. These criteria can ideally benefit from the shielding effect between the detectors aligned with the radioactive source. Different count rates granularities are used to allow the system an automatic adaptation to the activity i.e. the proximity of the source in order to process more statistics or make the display of the direction more reactive.
        Preliminary results obtained with the first prototype are very promising for radioactive sources at very short range (<2m). The entire DPP architecture is flexible enough to be used with many other sensors. Therefore, with an objective of performing localization at least 5 meters away, Monte-Carlo simulation are accomplished to design a system with larger sensors while respecting the constraints of portability and will be detailed in the full paper.


        [1] C. B. Hoshor et al., “Real-time neutron source localization and identification with a hand-held, volumetrically-sensitive, moderating-type neutron spectrometer,” Nucl. Instruments Methods Phys. Res. Sect. A Accel. Spectrometers, Detect. Assoc. Equip., vol. 866, no. May, pp. 252–264, Sep. 2017, doi: 10.1016/j.nima.2017.05.033.
        [2] D. E. Newman, “Radiation Detector With Two-dimenstional Directionality,” 2018.
        [3] E. D. Greenlee, “Real Time Fusion of Radioisotope Direction Estimation and Visual Object Tracking,” University of Tennessee, 2018.
        [4] E. M. Becker, “A Direction-Sensitive Radiation Detector for Low-Altitude, UAV-based Radiological Source Search,” Oregon State University, 2016.

        Speaker: Dr CARREL, Frédéric (CEA/DRT/LIST/DM2I/SCI/LCAE)
      • 14:20
        #08-179 Updates on a real-time in-line monitoring of tritium in water for Fukushima 20m

        Tokyo Electric Power Company (TEPCO), in order to enhance the monitoring capability for site contamination and surrounding environment is looking to continuously monitor Tritium (3H) concentration in seawater. We hereby present the latest developments of a system by CEA List, in partnership with TEPCO, to achieve the challenging real-time/in-line monitoring of Tritium (3H) in water.
        During Phase 1 of the project, a prototype has been developed, based on the SAFEWATER real-time in-line beta/gamma contamination monitoring system. Initially based, for SAFEWATER, on coated scintillating fibers, the TEPCO-dedicated design is based on ZnS-coated PMMA plates to make it sensitive to 3H. Digital pulse processing is performed to improve the 3H detection. An experimental campaign undertaken at Laboratoire National Henri Becquerel (LNE-LNHB), France’s National Metrology Laboratory in the field of ionizing radiation, showed promising results. This experimental campaign run during the first Phase of the project, while conclusive with 3H measurable at a concentration of 1.5 kBq/L has highlighted some possible paths for improvement.
        The focus of the current Phase 2 work, is to take into account the feedback and lessons learned from the promising Phase 1 results, in order to improve the design, for a better measuring efficiency and robustness of the system.
        The first part of the present study is dedicated to the design of the new measurement apparatus, using computer models and simulations carried out using the MCNP6.2 code and Solidworks CAD software; in order to design and eventually confirm assumptions made during Phase 1. The second part is focused on two measurements campaigns, and the adjustments that were made in-between, both undertaken at LNHB, in order to achieve metrology-grade assessment of the 3H concentration.

        Speaker: Dr MICHEL, Maugan (Institut LIST, CEA, Université Paris-Saclay)
      • 14:40
        #08-169 Radiation Detection Instruments Ideally Suited for Robotic Autonomy and a Novel Approach to Reduce Collimator Field-of-View 20m

        There are many applications in nuclear environments that benefit from the use of robotics, such as the response to accidents, decommissioning and routine monitoring. Robots have some advantages over humans in the aforementioned contexts, for example: they are often better at repetitive tasks or working in confined spaces. Such situations in nuclear environments can also present a radiological or chemical risk, which can be mitigated by deploying robots in the place of people. As the needs of the situation can be very diverse, the robots available match this with a multitude of shapes and methods of locomotion, wheeled, tracked, aerial, submersible etc. To navigate their environment, robots are equipped with sensors, which can be optical and infrared cameras or Light Detection And Ranging (LiDAR), for example. Radiation detection instruments are almost essential when operating in a nuclear environment, as they can provide valuable information when decommissioning or responding to an accident. When these are combined with a tether-less robot with a degree of autonomy, more complex tasks can be undertaken. Robotics in nuclear environments have already demonstrated significant use: e.g., submersible use at Sellafield, aerial mapping over the Chernobyl Red Forest and tracked robot deployment at the CEA etc. Unfortunately, radiation instruments are seldom designed to fit within the strict limits for robot use, such as power, weight, size and operation by onboard computer. Along with improvements in hardware, there is a wealth of opportunity for better onboard data analysis going beyond mapping count-rate data.
        This paper will report on the testing and deployment of two $\gamma$-ray detectors adapted for robotics and a novel data analysis method for reducing the Field-of-View (FOV) of collimation. One of the detectors is a 10-mm cubic CeBr$_{3}$ scintillator with photomultiplier tube (PMT), paired with a Red Pitaya STEMlab 125-14 which is repurposed as a multichannel analyser (MCA). The second detector is a combination of NaI:Tl with a PMT and a Brightspec bMCA MCA. These were selected as they are capable of spectroscopy, compact and well suited to a variety of robot platforms and applications. The associated software for them has been adapted and written to fit a number of criteria, these are: to use a widely adopted robot middleware, minimal performance loss on the robot and negligible additional work for the operator. The robot middleware used is the Robot Operating System as it is used by the robots in the collaboration. Both detector setups have been tested in a laboratory and demonstrated excellent performance, for example, achieving 5.4% and 8% resolution for CeBr$_{3}$ and NaI:Tl respectively when counting $^{137}$Cs.
        The detectors underwent preliminary field-testing at a low-level waste (LLW) drum store using a Clearpath$^{ \text{TM} }$ Husky robot in a routine monitoring application. During field-testing, the CeBr$_{3}$ detector was placed inside a thin-walled collimator and mounted onto the Pan-Tilt Unit on the robot to be able to point around the environment. The NaI:Tl detector was mounted in a fixed location without collimation. Field-testing was successful in satisfying the above criteria, and the preliminary data from the CeBr$_{3}$ detector was combined with LiDAR. This combination was used to overlay the radiation data onto the 3D LiDAR map to highlight the activity of the adjacent drum.
        For $\gamma$-ray detectors to be able to localise a source of radiation it is usually necessary for them to be surrounded by a dense material such as lead or tungsten. Producing a narrow FOV collimator means adding additional dense material to reduce the angular magnification, but this increases the mass of the system significantly. This increase in mass is a particular problem if the collimator is mounted to a Pan-Tilt Unit, which often have limited payload restrictions. An alternative technique has been developed here which processes the output from the barrel collimated detector by fitting it to a transform function, to test the hypothesis that this approach can aid source location. The arrangement is a 2-mm lead barrel collimator, the detector being panned over a single $\gamma$-ray point source. The detector is panned incrementally using only a single dimension as a proof-of-concept for the technique. This method relies on the response of the transform function of a step input, with the accumulated counts at each increment fitted to this response. The transform is then applied in reverse, to recover the response (now being a step), revealing the location of the point-source in the data. Preliminary results demonstrate that the transform technique is successful. This is currently being tested using a collimator with ~90° FOV and a point-source at 0.5 m.

        Speaker: Dr ALTON, Tilly (Lancaster University, UK)
      • 15:00
        #08-68 1D OSL/FO dosimeter array for remote radiological investigations in hard-to-access zones 20m

        Remote dosimetry (RD) is an essential tool for Decommissioning and Dismantling (D&D) as it enables to predict the impact of decontamination procedures in terms of worker exposition and to set up a cost-effective dismantling scenario associated with a risk analysis and waste management strategy. However, long-range RD in hard-to-access zones (e.g. tanks, reactors, storage ponds etc.) is still challenging. A critical case concerns inspections through small-diameter pipes (< 1 cm) of small radius of curvature that require miniaturized dosimeters. The miniaturization of conventional active dosimeters leads to a reduction in detection volume that in turn degrades the limit of detection (LOD) in dose rate. Furthermore, miniaturization is limited by electromagnetic (EM) shielding and embedded electronics (power supply, signal conditioning, thermal regulation) and signal degradation arises due to cable length as well. Geiger-Mueller detectors are often used to this purpose but they are still too bulky and fragile and require a modeling to retrieve dose rate from counting data.
        During the past 20 years, CEA investigated an alternative RD technique based on OSL/FO probes (Optically Stimulated Luminescence/Fiber Optics), affixed at the extremity of armored optical cables, robust enough to be pushed and removed within pipes. OSL/FO is a passive online dosimetry technique which does not require local electronics to operate (the readout unit is out of radiological zone). The LOD may be as low as several µGy/h (for daylong integration times), therefore making OSL/FO useful for the assessment of cleaning processes for instance. This technique provides wide range in dose rate detection (6 decades) by a suitable combination of integration time and dose. Other main advantages are EM and Cerenkov immunities, long-range remote operation, high miniaturization (sensor head Ø = 5 mm), high radiation resistance and very low fading (the OSL signal changes linearly with the integration time at constant dose rate). Finally, OSL/FO sensors are waterproof and easy to decontaminate.
        The INSPECT Project
        The INSPECT project started 11/2016 for 54 months. It aims at developing two multichannel OSL/FO readout units, a human-machine interface, and innovative 1D OSL/FO detectors (10- and 16-point detectors) for the monitoring of dose rate profiles within the range [μGy/h – Gy/h]. They are connected to a readout unit through a MIL-38999 hybrid (electrical + FO) connector which also incorporates an EEPROM for immediate data transfer of manufacturing and calibration data. Finally, field tests are planned on several French pilot sites (Orano La Hague, CEA Marcoule, CEA Cadarache).
        OSL/FO dosimetry provides radiological investigations in hard-to-access zones in the range [µGy/h - Gy/h]. It is now a proven RD technique for D&D, sparing heavy duty that would otherwise be necessary to provide access to conventional dosimetric means. Dose rate data coupled with topographic modeling of the inner infrastructure enable activity reconstructions with the help of Monte-Carlo modeling. The use of 1D detectors speeds up the investigation process and improves localization uncertainty in comparison with point dosimeters (0D). A performance assessment of the INSPECT device will be presented during the conference.
        INSPECT is supported by Andra under the French “investing for the future” Program (Programme Investissements d’Avenir), organized in cooperation with the French National Research Agency (ANR).

        Speaker: Dr MAGNE, Sylvain (CEA List)
      • 15:20
        #08-154 Automated radiological characterization of unknown objects with TRACER 20m

        Photopeak Efficiencies play an important role in the characterization of radioactive material. The presented approach coupling 3D sensors for geometry modelling and mathematical photopeak efficiency calculation offers an automated procedure for a wide range of applications, e.g. in drum measurement devices, clearance of bulk material on a conveyor belt, radiological sorting system as well as in-situ measurements on standing structures. For activity calculation, the modeling of an unknown object for a gamma spectrometric measurement to calculate the photopeak efficiency is highly individual and is currently mostly done by a human operator of the measurement facility. Thus, for more complex geometries, conservative assumptions are necessary, which sometimes oversimplify the geometry. Besides the high personnel expenditure for the modelling, this is associated with an overestimation of activity. The improvement of our developed nuclear simulation code TRACER (TRAck-based Calculation of Efficiency for Radiation) is based on generating the geometric models automatically from precise 3D point clouds recorded by a laser scanner for mathematical photopeak efficiency calculation. TRACER is a fast code for multicore photopeak efficiency calculation for a wide energy range from 50 keV to 10 MeV, primarily for High purity Germanium (HPGe) detectors. Thereby TRACER takes into account the objects surrounding the sample under investigation in an efficient hybrid approach combining Monte Carlo and deterministic methods including variance reduction. For the purpose of modelling the geometry of the objects, a 3D laser scanner is used which records position and dimensions of the sample and generates a 3D point cloud. From the point cloud data, TRACER automatically models the sample geometry, saving personnel expenditure and reducing uncertainty in the activity estimation. The additional information from the 3D point clouds offer e.g. the possibility to calculate the uncertainty for drum measurement systems more precisely and to detect errors in the measurement setup automatically.
        In the R&D project VIRERO (VIrtual REmote RObotics for Radiometric Sorting), conducted by Framatome, FAPS and AiNT, a robotic system is developed, which uses the recorded 3D point cloud to create a 3D model of the measured material for radiological sorting. The choice of the model is adapted to the application purpose and offers the possibility to calculate a conservative prognosis activity as well as a realistic activity (best estimate). In the automatically generated model, geometric details for the efficiency calculation are retained without affecting the required computing power to a large extent. These models offer the possibility to create a 3D-resolved nuclide specific activity profile with the measurement data from the HPGe detector. The information, where the nuclide specific activity is located, is used to achieve much higher accuracy levels in radiological characterization of unknown objects.

        Speaker: HANSMANN, Tim (AiNT GmbH)
      • 15:40
        #08-10 UAV prototype for localization and identification of radioactive contamination and emitters. 20m

        The Dragon (Drone for RAdiation detection of Gammas and Neutrons)
        prototype aims at designing and developing an unmanned aerial vehicle (UAV)
        equipped with a detection system able to identify radioactive materials, spread
        over an area or located in a specific position. The system is focused on the
        localization of the unknown emitter and its subsequently identification.
        The proposed prototype is made up of two easily interchangeable detection systems, one will be used as a counter while the second will be aimed to perform good-resolution gamma spectroscopy. Both solutions have neutron - gamma discrimination capability in order to be suitable for special nuclear materials (SNM) detection in gamma contamination background.
        The data acquisition module is made up of a compact digitizer board (RedPitaya, sampling rate of 125 MHz and 14 bits of resolution.), a mini computer (Raspberry, for example). This combination allows to install an embedded operating system (e.g. Linux) that can run the necessary software for the Data Acquisition (DAQ), like the ABCD distributed DAQ.
        Our contribution will be aimed to show a comprehensive characterization of the two detection systems, a medium size CLLB scintillation detector, and a large plastic scintillator, EJ-276, in order to assess their potential use in a UAV-based radiation monitoring system.

        Speaker: MORETTO, Sandra (Padova University)
    • 16:00 16:20
      Coffee Break 20m
    • 16:20 17:20
      Exhibitors and Industrials plenary session
      Convener: LYOUSSI, Abdallah (CEA)
    • 17:20 18:20
      Visit to Exhibitors
      Conveners: Dr HODÁK, Rastislav, LYOUSSI, Abdallah (CEA)
    • 10:40 11:00
      Coffee Break 20m
    • 11:00 12:40
      05 Nuclear Power Reactors Monitoring and Control

      Oral Session

      Convener: JIANG, Jing (Western University (Canada))
      • 11:00
        #05-65 Iterative method in modelization of gamma radiotracer measurements on the Colentec loop in Cadarache for the quantified analysis of the clogging phenomena in Steam Generators 20m

        The Tube Support Plate blockage, also named clogging, is a complex phenomenon that can occur in the steam generator of Pressurized Water Reactors. This deposit mainly composed of iron oxides, may induce several consequences (thermohydraulic flows changes inside SGs, vibrational or oscillatory risks, mechanical resistance of SG tubes and internals, impact on water inventory …). . Since 2014, the representative dedicated equipment, named COLENTEC loop at Cadarache had already provide a large number of results on the physico-chemical properties of the deposit formed in specifics thermohydraulic and chemical stable conditions. A major improvement was the switch to an active configuration allowing the injection of a gamma radiotracer 59Fe inside the circuit. Thanks to three dedicated gamma-measuring stations specifically designed by our laboratory, a first test showed that it was possible to determine on line the influence on clogging of parameters such as the chemistry, the temperature, the pressure without waiting for the opening and the dismounting of the test section.
        This paper presents the iterative method of modelling performed for the analysis of the results obtained during this first test on the three gamma detectors.
        The analysis starts with the simplest low background gamma measuring station in place and allows quantification of the 59Fe measurements of the fluid inside the loop.
        The modelling, with a Monte Carlo code MCNP, of the second on-line measuring station positioned on a section of the loop provides quantitative results on the residual 59Fe deposit along tubes.
        Finally, the sophisticated modelling of the geometrical configuration of the main on line measuring station on the Tube Support Plate allows determining the extent of the clogging.
        The presented method is then able to discriminate but also quantify primary and secondary phenomena taking place in the specific selected thermohydraulic and chemical conditions. It therefore improves the understanding of clogging and could lead to a better management of the maintenance of steam generators in power plants.
        These works are carried out within the framework of a scientific and financial collaboration between CEA and EDF (France) that the authors gratefully thank.

        Speaker: LOUBET, Laurent (CEA Cadarache)
      • 11:20
        #05-9 Numerical simulations in support of the design of an ultrasonic device for subassembly identification 20m

        In this paper, it is shown how numerical simulations can help designing an ultrasonic instrument operating in harsh conditions. To prevent fuel handling errors in sodium cooled fast reactors, the identification of fuel subassemblies using ultrasound is being investigated. It is based on the interpretation of a code (aligned notches) engraved on the subassembly head using an emitting/receiving ultrasonic sensor. This reading is performed in liquid sodium with high temperature (up to 600°C) transducers.
        A first experiment in liquid sodium demonstrated the feasibility of this method. The reading quality and robustness depend on various parameters related to the ultrasonic beam (spectral response, focal distance, focal spot size), the code geometry (especially the notches’ dimensions) and geometrical alignments.
        In order to avoid numerous experiments, two numerical models are developed. The first one is a finite element simulation of the sensor providing its radiated field. Unlike well-known analytic one-dimensional models, the finite element model is able to take into account the curved geometry of a focusing sensor. Moreover, it allows to model complex geometries of transducer. Finally, with the continuous growth of computing power, the finite element model allows the calculation of the radiated field with reasonable computational cost. This model is validated with the well-known analytic solution of the Rayleigh integral; then it is applied to the sensor used in the sodium experiment. The focal distance and focal spot diameter are close to the expected values.
        The second simulation, using CIVA software, provides the ultrasonic scan of the code. The latter is computed by a ray tracing technique, using the pencil method to derive echoes with their amplitude. It allows computational cost much lower than the finite element modelling. The result is in good agreement with the sodium experiment and a first comparison with a water experiment shows that this numerical tool is relevant for easily taking into account misalignment and misorientation of the scan.

        Speaker: PAUMEL, Kevin (CEA)
      • 11:40
        #05-203 Characterization of high harmonics frequencies in reactor noise experiments within the CORTEX project 20m

        The Laboratory for Reactor Physics and Systems Behavior (LRS) of the école Polytechnique Fédérale de Lausanne (EPFL) has in recent years been involved with the European CORTEX project, which aims to characterize reactor perturbations such as local bubbling, mechanical vibrations, etc. by neutron flux measurements. The goal of the LRS is the preparation and execution of experiments and experimental data analysis, to provide quantities of interest for newly developed neutron transport code validation. In this paper, we present a method of discriminating the origins of spectral power peaks at different harmonic frequencies: either due to higher modes of the reactor perturbation inducing device or due to effects of different harmonics.
        The European CORTEX project aims to develop an instrumentation system framework with the capability of localization and identification of different reactor neutron noise sources, leading to improved reactor diagnostics and ultimately safety. The project involves development stages from ground up: from experiment design, execution and analysis for code development and validation to instrumentation development supported by machine learning.
        As part of the experimental data analysis is to provide the data for code validation. Initial developments enabled simulations of frequency effects of the same order, which changed during the CORTEX project. In terms of the actuall experiments, a more rigorous approach was employed by including the position signal of the induced perturbation device in the form of oscillating absorber or fuel. This enabled for an estimate of two different contributions: due to higher harmonics of the oscillator itself, or higher order harmonics of lower order oscillations, which have different origins.
        An estimate on the two contributions can be made by comparison of spectral power peaks of the detector signals and the oscillator position signal. This can help distinguish between the two effects and provide higher quality data for code validation.
        This is possible by utilizing a high frequency resolution analysis technique based on bootstrapping with replacement. The oscillation and the detector data are synchronized in time. The data is then sectioned by well defined oscillator position data features (e.g. crossing neutral position), and used for spectral density calculations using Fourier transform and the bootstrapping re-sampling process. The developed analysis scripts also use modern graphics processors (GPU’s) to speed up the re-sampling process by a factor of ~10.
        The methodology has already been applied to the data from the 1st experimental campaign of the CROCUS reactor, yielding higher harmonics up to the 14th order. The data is already being utilized for the code validation and further developments.
        The improved capabilities of reactor noise propagation codes identified the need for higher harmonics experimental oscillator and detector data. The newly developed analysis techniques enable the distinction between higher order effect of the oscillator movement itself and higher order effects due to lower harmonics oscillations. The analyzed experimental data is available to the CORTEX consortium members for code validation and comparison.

        Speaker: AMBROŽIČ, Klemen (LRS, EPFL)
      • 12:00
        #05-114 Monte-Carlo determination of correction factors for the absolute measurement of the activity of solid-state dosimeters using γ- and X-ray spectrometry 20m

        The MADERE platform (Measurement Applied to DosimEtry for REactors) is a metrology facility whose main purpose is to determine the specific activity of solid-state dosimeters irradiated in nuclear reactors, using γ- and X-ray spectrometry. The platform is accredited by the French accreditation Committee (COFRAC) for specific activity measurement of γ- and X-ray emitters.
        Analysis methods used at MADERE platform for deriving absolute activities of a dosimeter from the spectrometry measurements induce to take into account several perturbing inherent phenomena of γ and X-ray spectrometry experiments such as self-attenuation inside the dosimeter or escape peak inside the detector. Thus, in order to perform a reliable and accurate analysis, the usual method to determine these correction factors is to use semi-empirical formula and dedicated software. Associated uncertainties are calculated using a Bayesian approach and can reach up to 10 % of the correction value.

        In the framework of the continuous improvement of the MADERE platform, an integral Monte-Carlo simulation approach has been initiated to model all the interactions occurring in the whole measurement system (sample, holder device and semiconductor detector).

        In order to improve the accuracy on the correction factor determination, a simulation scheme using the combination of the Monte-Carlo codes TRIPOLI-4® (CEA) and GEANT4 (CERN) is developed. In this work the different method developed to determine the correction factors due to the photon interaction inside the detector and inside the dosimeters are presented. Monte-Carlo correction factors are determined for different dosimeters configurations (matrix, geometry) and measurements setup (detector type, dosimeter-detector distance). To validate the results provided by the Monte-Carlo methods, simulations have been compared with the factors routinely applied at the MADERE facility. Then a qualification process of these simulated correction factors will be set up based on the comparison with results of dedicated experiments.

        Speaker: Dr SUBERCAZE, Alexandre (CEA)
      • 12:20
        #05-144 Novel Model-Based approach for instrumentation and control of nuclear reactors 20m

        Technological platforms dedicated for digital instrumentation and control of nuclear reactors are quite complex in terms of functionalities and devices. Hence, the design of these platforms requires high-level abstraction layers able to reduce the complexity, to rise the automation and to check the consistency between different development stages. The development of such systems is a challenging task that requires modeling of various components at different levels of abstraction and viewpoints, notably functional, hardware and software levels. In this paper, a new system engineering methodology is proposed to provide high-level models of different components and inter/intra- communication between them. These models are used for system specification, architecture design, performance evaluation or verification and validation. This approach focuses on the internal behavior of different components at different levels of abstraction in order to enable the interoperability of these components and to enhance cooperation between different stakeholders of the development process. An experimental setup has been carried out to validate this approach by customizing an open source model based engineering tool, Eclipse Papyrus, towards a significant reduction of system development cost in terms of engineering resources and equipment devices.

        Speaker: OUNI, Bassem (Université Paris-Saclay, CEA, List, F-91120 Palaiseau, France)
    • 11:00 12:40
      08 Decommissioning, Dismantling and Remote Handling

      Oral Session

      Conveners: MORICHI, Massimo (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy) , BOUDERGUI, Karim (CEA, LIST, Laboratoire Capteurs et Architectures Électroniques)
      • 11:00
        #08-54 Deep Learning for Compton image reconstruction with Caliste miniature imaging spectrometer: An innovative approach to enhance detection sensitivity 20m

        Caliste is a miniature CdTe-based pixelated imaging spectrometer. The detector is a single plane 1 cm² crystal readout by 8 low noise full custom ASICs enabling high energy resolution. Thanks to its position sensitivity and its spectrometric performances, this detector is able to perform Compton imaging in order to localize radioactive point sources, emitting high-energy gamma-ray photons, and its compactness is advantageous to be the detection component of a portable Compton gamma camera. However, our current prototype is equipped with a small detection volume of 100 mm3 (1 mm thick). Because our detector is thin and the photon interaction positions are determined in 2D, Compton reconstruction in such a miniature single plane detector is challenging. More specifically, the 3D interaction position uncertainty in the crystal and the small detector volume are both critical limitations for efficiency and sensitivity of the detector for Compton reconstruction. Nonetheless, Compton reconstruction is possible and reported in the paper. Classical backprojection and more advanced algorithms in the literature, such as LM-MLEM (List-Mode Maximum Likelihood Expectation Maximisation) and SOE-RR (Stochastic Origin Ensemble with Resolution Recovery), require to detect at least an order of magnitude of 2000 Compton events to reconstruct a radioactive point source. With Caliste geometry, this necessitates at least 7 hours of acquisition to localize a 137Cs point source of an activity of 1.6 MBq at 30 cm from the detector, with a corresponding dose rate of 1.4 µSv/h at the detector level. In order to enhance the sensitivity of our tiny system, we develop in this work a Deep Learning approach to reconstruct Compton images from spectroscopic data and identify the position of point sources over 2π steradians field of view and we demonstrate the feasibility of this new algorithm for high-energy imaging. We train a Convolutional Neural Network (CNN) with a synthetic database to localize a point source from direct backprojections, computed by Monte-Carlo simulations. We apply this CNN to real data at the 662 keV line of 137Cs, acquired with Caliste and we show that we are able to reconstruct point sources with an order of magnitude of 200 Compton events. This corresponds to a sensitivity of about 150 nSv/h with only a 0.1 cm3 detector, which surpasses the performances of classical algorithms.

        Speaker: Dr DANIEL, Geoffrey (CEA)
      • 11:20
        #08-33 Novel thermometry approaches to facilitate safe and effective monitoring of nuclear material containers 20m

        Many established nuclear power producing countries are currently decommissioning first and increasingly second-generation power producing plants and fuel processing facilities. This has led to a growing inventory of different containers and packages containing radioactive waste and other nuclear materials, as well as storage of spent fuel. Here we describe establishment of in-situ yet remote health monitoring techniques based on novel temperature measurement methods for different containers and stores used to hold different radioactive waste forms, such as intermediate level waste (ILW), special nuclear materials (SNM) and spent fuel.
        ILW and SNMs are usually held in high quality stainless steel containers. In the case of the containers used by Sellafield Ltd the ILW suspended in cementitious grout and held in 500 litre containers (drums), whilst in the future heterogenous ILW will be stored wet in approximately 3 m3 packages (i.e. covered with water inside the container). SNMs are generally held in nested stainless-steel containers. Advanced gas cooled reactor (AGR) spent fuel will be held in ponds awaiting final disposal.
        Each of these container waste forms and spent fuel storage has particular monitoring challenges to ensure the on-going integrity of the storage. Particular challenges are; the high level of gamma radiation emitted by the ILW and spent fuel, the gamma and neutrons emitted by the SNM, and the general inaccessibility of the store and pond configurations.
        Here we report the use of quantitative remote temperature measurement techniques to assess the condition of waste package or temperature of spent fuel racks. Specifically the following measurement requirements are addressed; ILW container health and internal temperatures, determination of the surface temperature of SNM containers in-situ with the future prospect of identifying surface and sub-surface defects, development of surface thermometry approaches for a new long term storage container of SNM, development of a method to determine the temperature of racks holding spent AGR fuel and a remote thermal method to identify anomalously high temperature 3 m3 packages in stores.
        Key to the success of this work is that the thermometry approaches, where needed, are traceable to the current temperature scale the International Temperature Scale of 1990 (ITS-90). Robust traceability to ITS-90 gives a stable baseline against which to compare all subsequent measurements and confidence in the actual measured temperatures. Temperature measurement uncertainties are evaluated according to the internationally accepted Guide to the Expression of Uncertainties (GUM).

        Speaker: Prof. MACHIN, Graham (NPL)
      • 11:40
        #08-22 Nondestructive active characterization of large concrete nuclear waste packages using photofission technique and high-resolution delayed gamma spectroscopy 20m

        Nondestructive characterization of large concrete radioactive waste packages is a major challenge. The fissile mass inside has to be estimated very precisely in order to transport and store the wastes in the safest way. Ultimately, knowing the alpha activity of the waste packages could enable to lower their activity category and thus to significantly reduce their final repository cost, e.g. in a surface vs. a geological disposal. The 870 L concrete waste drums, produced and stored at CEA Cadarache in France, represent an issue since they cannot be characterized using Active Neutron Interrogation. Indeed, their large dimensions and poorly known hydrogenous content lead to extremely penalizing uncertainties associated to high attenuation effects. This article focuses on the development of a nondestructive characterization method using Active Photon Interrogation (API) to estimate the fissile mass contained inside these large concrete waste barrels. This technique exploits the photofission reaction of nuclear materials occurring with photons of energy larger than 6 MeV. Since all actinides have similar photofission cross sections, the method cannot directly estimate the fissile mass (for example 235U, 239Pu) or the alpha activity (mainly Pu isotopes), and therefore a discrimination method is needed to identify the main nuclides of interest. This method is based on the difference of photofission product yields in actinides, and consequently on their delayed gamma ray intensities. In a first step, we present here new experimental data of more than 30 photofission product yields for 238U and 235U obtained by irradiating bare uranium samples with a Bremsstrahlung photon flux produced by a 16 MeV electron linear accelerator at the CINPHONIE facility of CEA Cadarache, France. The fission contribution of parasitic neutrons produced by photonuclear reactions in the accelerator conversion target, in the surrounding materials, and in the sample itself, was also calculated and subtracted. These measurements lead to select the most suitable gamma-ray lines for discriminating between 238U and 235U in a dense environment like a concrete package. In a second step, we use these measured photofission yields in our Monte-Carlo (MCNP6) simulations to assess which delayed gamma-rays are detectable in concrete radioactive waste drums. Since matrix effects produce a differential attenuation of these delayed gamma rays, they affect the discrimination of actinides and increase the uncertainty on the fissile mass. Therefore, we also study actinide localization methods via Monte-Carlo simulations, using specific delayed gamma rays from photofission as a probe, which will be qualified experimentally with a mock-up of a concrete waste matrix.

        Speaker: Mrs DELARUE, Manon (CEA, DES, IRESNE, Nuclear Measurement Laboratory)
      • 12:00
        #08-16 The Gamma and Neutron Monitor Counters for the MICADO Project 20m

        In the framework of the MICADO EU project, aimed at the full digitization of the low and intermediate level radioactive waste management, we have produced and characterized a full set of radiation counters suitable for monitoring gamma rays and neutrons. In particular the goal of the Work Package 7 is to set up a granular radwaste monitoring system to be used during the final demonstration. The proposed system for the online real-time monitoring consists of an array of many radiation sensors to be deployed all around a number of radioactive waste drums, in order to collect counting-rate data in real time and to make them available to a software platform named DigiWaste. In order to be suitable for mass deployment these detectors have to be small, reasonably inexpensive, robust, easy-to-use and reliable. The sensors developed possess all of these features.
        Continuous radiological monitoring of radwaste has to be based on the measurement of gamma and neutron radiation, since these are the penetrating types of radiation more easily detectable out of the drums. This is the reason why the focus was placed on the development of detectors for gamma rays and neutrons. In view of a possible mass deployment the foreseen sensors have to be reasonably low-cost. The overall system must be modular, so that one can easily modify number and placement of the sensors around the drums, and it has to be scalable in order to make it possible to tailor it to small, medium and large scale storage configurations without conceptual limitations.
        The proposed system is based on detectors which can be easily installed and/or reassembled in different geometrical configurations, as they are mechanically very simple and are based on commercial electronics (see the attached file with figures). The detection efficiency the SiLiF thermal neutron counter is 4%, whereas the SciFi gamma counter has a ≈2.5% efficiency. Moreover, these detectors do not need special calibration procedures, and their ease of use and installation makes them suitable for short, medium and long term (as long as possible) monitoring.
        A continuous automatic monitoring of the radwaste drums after their characterization represents an added value in terms of safety and security, and the availability of continuous streams of counting-rate data around each drum is a comfortable tool toward the transparency, which now more than ever is a hot topic of the nuclear industry with respect to the common people environment-aware.

        Speaker: FINOCCHIARO, Paolo (INFN Laboratori Nazionali del Sud)
      • 12:20
        #08-141 Simulation of the physical processes associated to the detection of beta particles with scintillating fibres 20m

        In this study, we present an exhaustive model to simulate the detection of beta particles with scintillating fibres, based on a chronological follow-up of the information carrier. We manufactured a detector composed of a bundle of 100 BCF-10 scintillating fibres, which will be used to compare simulation and experimental results. A Monte-Carlo simulation model of the detector was generated using the MCNP6.2 code. To simulate more accurately the generation of scintillation photons inside the detector, particle tracking was used which allows us to use the incident beta particle stopping power instead of the energy deposition. A small energy deposition from a high-energy beta particle should not be processed as an incident low energy beta particle, as they will not yield the same number of scintillation photons. Using stopping power over energy deposition for beta particles enables to use theoretical scintillation laws, which take into account the ionisation quenching effect. Indeed, for low energy electrons, there is a discrepancy of the scintillation linearity. By integrating the quenching ionisation law into our model, we can evaluate more precisely the expected experimental spectrum at very low incident energies. Based on the literature, a more detailed photon trapping efficiency is calculated, which was found to be underestimated in the scintillating fibre technical sheet. The manufacturer value corresponds to a trapping efficiency based on meridional rays solely, while the calculated value also factors in the skew rays. Self-absorption in the fibre was estimated using an exponential loss at the scintillation wavelength. Interface between the fibre and the photomultiplier was evaluated using Fresnel and transmission equations. Finally, the probability distributions associated with the photocathode quantum efficiency and the photomultiplier gain were both taken into account to generate expected acquisition spectra. To do so, we used the thermionic signal, comparable to the amplification of a single photoelectron, to evaluate the photomultiplier gain distribution. This model will help in evaluating the detector applicability to specific experimental scenarios such as nuclear decommissioning and dismantling operations.

        Speaker: DUFOUR, Nicolas (Commissariat à l'Energie Atomique)
    • 11:00 12:40
      09 Environmental and Medical Sciences

      Oral Session

      Convener: LE DÛ, Patrick (IEEE NPSS, France)
      • 11:00
        #09-237 Applicability of large-area single-photon counting detectors Timepix for high-resolution and high-contrast X-ray imaging of biology samples 20m

        High-resolution X-ray imaging techniques, usually known as micro-radiography and micro-CT, have become highly required and frequently used tools for biology, biomedical and pre-clinical research. State-of-the-art micro-CT scanners are capable of achieving spatial resolution of few micrometers or even less thanks to the constant development of compact micro-focus X-ray sources with simultaneous progress in detector technologies.
        X-ray micro-CT is frequently used for phenotyping, study of zoomorphology, drug development and others. Moreover, it is nowadays used for non-destructive inspection of ex vivo soft biology tissue. Such technique has become known as virtual histology. High-resolution X-ray imaging currently becomes a competitor to conventional research techniques used in biology research like optical microscopy and histology.
        The current standard in X-ray detection is a digital read-out chip coupled with a scintillation sensor. Such detectors are available in variety of different sizes, they are easy to use and relatively affordable. Nevertheless, the mentioned technology suffers from inherent technology limitations, like for example undesirable generation of dark-current, that compromise the quality of the provided data.
        This work demonstrates the advantages of large-area hybrid-pixel photon-counting detectors Timepix for high-resolution X-ray imaging in biology research. Photon-counting detection technology provides dark-current-free quantum-counting operation. Therefore, enhanced contrast-to noise ratio is of the acquired data is achieved. Furthermore, the biased semiconductor sensor achieves almost ideal point-spread-function resulting in high spatial-resolution of images. And finally, the detectors are operated with user-adjustable detection threshold opening possibilities for energy-sensitive X-ray imaging. Abovementioned features make photon-counting detectors to be excellent tools for high-resolution X-ray imaging of samples with low intrinsic absorption contrast like for example soft biological tissue.
        We evaluate the imaging performance of large-area Timepix detectors compared to widely used scintillation-based X-ray imaging detectors dedicated for high-resolution X-ray imaging. Further, we summarize and demonstrate the applied results in the field of biology and pre-clinical research achieved at Institute of Experimental And Applied Physics, Czech Technical University in Prague. The institute, as a member of Medipix Collaboration, has actively participated in the development of Timepix technology from its introduction. The presented data obtained in cooperation with Charles University demonstrate the versatility of the used detectors as it covers a wide range of samples from laboratory animals to single-cell marine organisms. Finally, practical experience from long-term usage is discussed and the limitations of Timepix technology for X-ray imaging are mentioned.

        Speaker: DUDAK, Jan (Institute of Experimental and Applied Physics, Czech Technical University in Prague)
      • 11:20
        #09-235 The Wearable PET project for a compact clinical exam for an early diagnosis 20m

        The Wearable PET (WPET) project was successfully supported by EU-ATTRACT ( Its objective is to demonstrate the feasibility of a light PET systems for cancer preventive screening which can be hosted by wearable vest.
        Positron Emission Tomography or PET is a very precise imaging modality commonly used in hospital oncology units to test patients with cancer to establish possible metastasis or other complex diseases difficult to diagnose with traditional methodologies. In PET a radioactive positron emitter is injected into the patient’s body. Sugar is attracted by infected human cells and cancer cells, which therefore accumulate the positron emitter. The emitted positrons annihilate producing two high energy gamma rays back to back that can be detected outside the human body.
        Traditional PET systems detect the two emitted gammas and reconstruct the emission point. These devices produce a 3D image of the full body reconstructing the cancer location with spatial resolution of the millimetre. The disadvantages of these systems are that are expensive, massive and can only be used for a limited number of tests daily.
        On the contrary, the WPET, that performs a similar, but less detailed, measurement for a longer time, would allow one to use this modality for routine screening for a large fraction of the population at risk, improving early stage detection.
        The WPET project aims at building a wearable PET scanner with a modular, flexible design, able to image different body parts according to specific needs, as well as to track detector motion and correct its effects in real-time. This is made possible not only by lighter and compact LYSO-SiPM detector modules, but also by accompanying developments in miniaturized readout electronics, battery technology, wearable sensors, wireless data transmission and fast PET image reconstruction algorithms. WPET requires a novel system design to combine wearable electronics with novel materials, controls, batteries and data transmission components to be hosted in a comfortable wearable support. Storage, handling and visualization are also part of the package on the medical side for an easy diagnosis of the collected data.
        Within the project Geant4 MonteCarlo simulations were performed to optimize the jacket design and evaluate the feasibility. Reconstruction algorithms were used to demonstrate the proof of concept performances. With this paper we would like to present the WPET concept and the early results obtained. Using simulation and tests with single modules we can conclude that the fabrication and use of a WPET system is feasible. The key to feasibility is scalability, miniaturization, data storage and handling and precise position monitoring. With current technology, WPET can detect tumours as small as 2 mm with 10 kg crystal weight in 6 hours, using the same tracer dose as conventional clinical PET.

        Speaker: FANCHINI, Erica (CAEN s.p.a.)
      • 11:40
        #09-150 Development of a proton bunch monitor for accurate particle therapy treatment verification 20m

        Treatment verification is expected to improve targeting precision in particle therapy. A promising technique to achieve this goal is the detection of prompt gamma rays emitted along the particle tracks inside the patient. The range of the particle beam can be inferred by determining the time distribution of these gamma-rays relative to the radio frequency of the accelerator, a method commonly referred to as Prompt Gamma-Ray Timing.

        However, the translation of this method into a clinical setting is currently hindered by instabilities of the phase relation between the arrival of the proton bunches and the radio frequency of the accelerator. These instabilities include two effects, which have been studied at the clinical treatment facility of the University Proton Therapy Dresden. Firstly, a long-term drift of the proton bunch phase relative to the radio frequency in the order of several hundred picoseconds per hour was observed, which may be caused by small temperature changes in the cyclotron’s magnet resulting in magnetization variations in its iron parts. Secondly, strongly damped oscillations in the mean of measured prompt gamma-ray timing spectra with an amplitude in the order of few hundred picoseconds occur for about two seconds after each change of the particle energy during pencil beam scanning. This oscillation is caused by ramping the acceleration voltage back to its nominal value, which is reduced between energy layers to minimize the dark current of the accelerator and the resulting excess dose to the patient.

        While the former effect is only of secondary importance for the treatment due to its comparably long time scale, the phase oscillation has a considerable negative impact on the accuracy of the Prompt Gamma-Ray Timing method, which has to detect time shifts in the order of a few picoseconds for the detection of millimeter range changes. Therefore, the development of a method to monitor the arrival time of the proton bunches independently from the accelerator radio frequency, a so-called proton bunch monitor, is crucial.

        To this end, a bunch monitor prototype was developed consisting of scintillating fibers placed in the halo of the proton beam. The fibers were read out on both ends by silicon photomultipliers. A thick acrylic glass target with cylindrical air cavities of varying thickness and different tissue-equivalent inserts was irradiated with protons of clinically relevant energies and typical beam currents. The mean proton arrival time, determined from the time spectra of the proton bunch monitor, was used to correct the prompt gamma-ray timing spectra acquired by Ø2”x2”CeBr3 high-resolution scintillation detectors. This correction allowed to resolve differences in the prompt-gamma ray timing spectra acquired with the different cavities and inserts.

        In conclusion, the developed proton bunch monitor was successfully integrated to the Prompt Gamma-Ray Timing method and is expected to enable the clinical application of this method for clinical treatment verification in particle therapy.

        Speaker: Dr SCHELLHAMMER, Sonja M. (OncoRay, Helmholtz-Zentrum Dresden - Rossendorf)
      • 12:00
        #09-205 Multi-particle detection and imaging with quasi-monolithic organic detector arrays for treatment verification in particle therapy 20m

        Particle therapy is a modality for treating cancer using ionizing radiation from, e.g., protons or carbon ions. A growing number of patients worldwide receive particle therapy as a part of their cancer treatment due to its dosimetric advantages over the more conventional external beam radiotherapy using photons. The finite range of particles in tissue results in sparing of healthy tissue surrounding the tumour and thereby a reduced risk of adverse effects compared to conventional treatment. This opens possibilities for a more intensified treatment by dose escalation to the tumour and thereby an increased probability for controlling the disease. However, charged particles, when used for external beam radiotherapy, are much less forgiving than photons in case of treatment deviations. Treatment deviations may be caused by (1) sensitivity of charged particles to anatomical or density changes along their radiological path in the patient, e.g., due to organ motion or tumour shrinkage and (2) uncertainties associated with the estimation of the exact position of the Bragg-peak (maximum dose deposition) in tissue, e.g., due to uncertainties in the estimation of stopping power ratios. Collectively, these are referred to as “range uncertainties”. The most important consequence of range uncertainties is that the tissue sparing potential of charged particles cannot be fully exploited. Therefore, there is an urgent need for reliable and robust treatment verification systems that can identify potential treatment deviations in real-time.
        Despite recent advances in state-of-the-art prompt gamma-ray imaging, spectroscopy and timing systems as well as in-beam and offline PET imaging systems, the development of treatment verification systems still lags and as such, there is no system yet in wide routine clinical use. Common to all existing solutions is the fact that they rely on a single feature of a single particle species and will therefore suffer from limited counting statistics. This is an important limiting factor in applying state-of-the-art systems in a treatment verification system.
        Recently, a collaboration of detector, nuclear and medical physicists, nuclear engineers and mathematicians initiated the NOVO (Neutron and gamma-ray imaging with quasi-monolithic organic detector arrays – a novel, holistic approach to real-time range assessment-based treatment verification in particle therapy) project. The aim of the NOVO project is to develop a sophisticated quasi-monolithic organic detector array (QuDA) that will combine detection and imaging of secondary fast neutrons and prompt gamma-rays produced in nuclear interactions of incident particles with tissue, the profiles of which show a strong correlation with the incident particle beam range. In addition to imaging spatial profiles of secondary fast neutrons and prompt gamma-rays, the QuDA will be used to acquire information on additional features of each particle species, such as timing, energy and intensity, that can provide supplementary information on the range of particle beams in tissue, thus representing a major shift from single-feature, single-particle systems to a multi-feature, multi-particle system.
        In this contribution we will report on first estimates of the imaging properties of a potential QuDA design, based on Monte Carlo radiation transport models with MCNP6.2, Geant4 and GATE, considering fast neutrons and prompt gamma-rays for various, realistic combinations of timing, energy and position resolution of the individual sensing elements. Furthermore, we will report on the expected detection efficiencies and the resulting range monitoring precision for proton beams with clinically relevant energies and intensities incident on homogeneous polymethyl methacrylate phantoms. The preliminary results demonstrate the potential to obtain a range monitoring precision of approximately 1 mm down to proton beam intensities of about 1 – 2 x 107 [protons/pencil beam] when fast neutron and prompt gamma-ray data are combined.

        Speaker: Dr MERIC, Ilker (Western Norway University of Applied Sciences, Bergen, Norway)
      • 12:20
        #09-164 Multi-Feature Treatment Verification in Particle Therapy 20m

        Particle therapy constitutes a promising and rapidly developing method in modern cancer treatment. In order to exploit its full potential, however, it requires detailed dose verification.
        Although the applicability of in-beam positron emission tomography and prompt gamma rays has already been demonstrated in patients, range verification is not yet part of the clinical routine in particle therapy. This is due to not only the methodological limitations of previous systems, but also to commercial, clinical and physical boundary conditions.
        In pencil beam scanning, the state-of-the-art treatment method in particle therapy, the number of secondary particles (essentially positrons, prompt gamma rays and fast neutrons) available per spot ($\Delta t = 10$ ms to $100$ ms) is limited. This leads to statistical accuracy limits for verification systems exploiting these secondary particles as range probes. The development of a clinically useable treatment verification system requires gathering as much information about the local dose, as possible.
        The instantaneous fluence rate of prompt gamma rays reaching $5\times10^{6}$ cm$^{-2}$s$^{-1}$ to $10^{8}$ cm$^{-2}$s$^{-1}$ challenges modern data acquisitions connected to monolithic inorganic scintillators with typical sizes used in present verification systems. In order to reduce the detector load, and also with regard to the ever higher intensities of next generation medical accelerators, future systems have to be more granular.
        Multi-Feature Treatment Verification combines and extends established methods (prompt gamma-ray imaging, spectroscopy, timing, etc.) in order to achieve higher reliability and performance. This idea was taken up by the NOVO project and expanded by a multi-particle approach. The NOVO (i.e. Neutron and gamma-ray imaging with quasi-monolithic organic detector arrays – a novel, holistic approach to real-time range assessment-based treatment verification in particle therapy) consortium is a large collaboration of medical, nuclear and detector physicists, nuclear engineers, and mathematicians, which aim to develop a holistic real-time treatment verification system in particle therapy.
        Elements of a potential multi-feature/multi-particle treatment verification multi-channel system were characterized in a double time-of-flight experiment at the pulsed photo-neutron source nELBE (neutrons @ Electron Linac for beams with high Brilliance and low Emittance). The essential properties (time resolution, light yield, detection efficiency and pulse shape discrimination) of an EJ-276 plastic scintillator were determined.
        The very first experimental results show that the time resolution ($\Delta T< 400$ ns) of a $20\times20\times200$ mm$^{3}$ EJ-276 plastic scintillator with double-sided readout will reach the high demands of such a proposed range verification system. However, the determined quality of the pulse-shape discrimination, the energy resolution and the quite high neutron detection threshold of above $200$ keV show that the light yield of this type of scintillator is not high enough to be used in a multi-feature-based treatment verification system. Particle transport calculations with MCNP6 and GEANT4 were performed to confirm the experimental results of a single detector element. Furthermore, they also show a promising measurement accuracy of a multi-channel overall system.

        Speaker: Dr KÖGLER, Toni (OncoRay – National Center for Radiation Research in Oncology, Faculty of Medicine and University Hospital Carl Gustav Carus, Technische Universität Dresden, Helmholtz-Zentrum Dresden - Rossendorf, Dresden, Germany)
    • 11:00 12:40
      10 Education, Training and Outreach

      Oral Session

      Convener: GARCÍA, Carmen (Instituto de Física Corpuscular (CSIC-Universidad de Valencia))
      • 11:00
        #10-130 On Teaching Experimental Reactor Physics in Times of Pandemic 20m

        Practical exercises or hands-on experiments in Reactor Physics at a research reactor are an essential part of education and training in Nuclear Engineering. They are mostly performed at low power research reactors and/or critical and subcritical assemblies. Institutions that do not have their own research experimental facilities often visit other facilities to perform hands-on experiments. In times of pandemic, when mobility of people is very limited, such visits are practically impossible to perform hence other options has to be considered.
        The Jožef Stefan Institute (JSI) TRIGA Mark II research reactor is regularly used for education and training. However, since spring 2020 the access to the reactor has been limited due to worldwide Covid-19 pandemic. In spring 2020 the reactor was completely shut down, practical exercises on experimental reactor physics had to be performed online using videoconference software, cloud services and research reactor simulator. In June 2020 the reactor started normal operations, however due to travel restrictions around the world and in Europe, students could not come to Slovenia. Hence a practical educational curse ‘’Experimental reactor physics’’ for students from Uppsala University was organised remotely.
        A five-day practical Educational Course was organised and performed using off the shelf but advanced software and hardware components such as: a remotely controlled dome camera in the control room, portable cameras that the lecturers could take to the reactor, two video-conference setups, remotely controlled laptop used to operate the data acquisition software and the Digital Reactivity Meter, a remotely controlled common whiteboard, a remotely-operated camera showing the reactor core, cloud document system.
        The educational course program consisted of the same exercises that are offered during in-person exercises: Introduction to JSI TRIGA reactor, Critical experiment, Reactor response to step reactivity changes, Reactor operation, Void reactivity coefficient, Control rod worth measurements and Temperature reactivity coefficient.
        The performance of the course was evaluated using an online anonymous survey taken by all the students and their mentors. The organizers provided both open-ended questions and questions that were answered using a rating scale. The aspects being evaluated included the technical content, quality of material, performance of the individual lecturers and the quality of the remote session. In general, the response was overwhelmingly positive with most questions with a rating scale answered with “excellent”. Multiple participants complained about the occasional problems with the sound quality, as they could not clearly hear the speakers when they stepped too far away from the conference microphones. The organizers of the course agree that the first remote exercises on a research reactor in Slovenia were well organized and useful but observed difficulties in achieving the same level of student involvement as the in-person exercises.
        In the paper we describe the education course and its implementation. This is followed by evaluation of the course and outlook for future improvements. The paper can serve as demonstration and a guide on how to organise remote hands-on exercises at a research reactor, using relatively cheap and widely available components. It is important to note that the purpose of remote hands-on experiment is not to replace the in-person ones but to provide an alternative in difficult times, such as pandemic or any other situation affecting free movement of people.

        Speaker: MALEC, Jan (Jožef Stefan Institute)
      • 11:20
        #10-201 The European Nuclear Experimental Educational Platform – ENEEP: Progress, Prospects and Remote Education Capabilities 20m

        The European Nuclear Experimental Educational Platform – ENEEP is currently being established by five European educational / research organizations in the framework of a Horizon 2020 project, initiated in 2019. The ENEEP partner institutions are the Jožef Stefan Institute (JSI) in Ljubljana, Slovenia, the Slovak Technical University (STU) in Bratislava, Slovak Republic, the Czech Technical University (CTU) in Prague, Czech Republic, Technische Universitaet Wien (TU Wien) in Vienna, Austria and the Budapest University of Technology and Economics (BME) in Budapest, Hungary. ENEEP is intended as an open educational platform, offering experimental hands-on educational activities at the ENEEP partner institution facilities: four research reactors and one Radiation Physics Laboratory.
        ENEEP educational activities will be offered in different formats: group educational activities (“package” and “custom” courses) and individual activities, and are targeted at university students at the bachelor, master and Ph.D. educational levels and young professionals in the nuclear field wishing to deepen their knowledge and gain valuable practical experience in nuclear facilities.
        This paper gives an overview of the ENEEP project activities and the progress achieved so far, highlighting the experimental capabilities which will be offered. In the first implementation phase, ENEEP will be based on a comprehensive set of experiments which constitute the basics in Reactor Physics and Nuclear Engineering curricula, e.g. approach to criticality, reactor response to changes in reactivity, neutron flux mapping, as well as more specific experiments focusing on particular aspects – investigated phenomena, types and working principles of detectors, etc., e.g. neutron emission rate measurements with the manganese sulphate bath technique, radiation measurements with semiconductor detectors. Subsequently, novel education activities will be introduced and implemented in ENEEP, following scientific development in nuclear science and technology and nuclear instrumentation detectors, stemming from research activities. Attention will be devoted to the development and optimization of remote education capabilities at the ENEEP partner institutions, of particular relevance during the current Covid-19 pandemic which is responsible for major changes in education activities worldwide.

        Speaker: RADULOVIĆ, Vladimir (Jožef Stefan Institute)
      • 11:40
        #10-230 Development of Active-Learning Units in Nuclear Engineering 20m

        Active learning engages students in activities that could enhances their ability to analyze, synthesize, and evaluate the material being learned. The students participate in doing things instead of just listening. Evidence-based studies have shown that active learning increases student performance in Science, Technology, Engineering, and Mathematics (STEM) courses. The goal of this project is to develop active-learning units to enhance students learning and technical skills to improve their preparation for success in pursuing STEM graduate programs and careers in nuclear engineering. Three modes of active learning that are of interest include: problem-solving, lab-based hands-on activities, and simulation. This paper focuses on the development of problem-solving interactive units aimed at mastering fundamental principles and concepts, and better understanding of how equations translate and apply to real-life engineering situations. It also enhances the understanding of how different parameters in an equation interact with each other (such as dependency relationships). The practicality is in understanding how different components of an engineering systems function together to accomplish the goals for which they are designed. The design approach to each problem-solving unit starts with a brief review of the fundamental concepts, and identification of key parameters. Students participates in determining independent and dependent variables, constant parameters, and equations/formulas associated with the problem. The steps in each unit include understanding the problem, identification of the equations/formulas needed to solve the problem, listing the knowns and the unknowns, solving the problem, testing/checking the solutions/answers, and write down any difficulties encountered. Part of the post-activity evaluation include identifying any difficulties encountered, how were they resolved, and any lessons learned. The problem units presented in this paper are in the areas of neutron interactions, cross-sections, and attenuation.

        Speaker: EGARIEVWE, Stephen (Alabama A&M University)
      • 12:00
        #10-257 Visualization of ionizing radiation with MiniPIX-EDU turns boring theory into wonderful exploration 20m

        The MiniPIX EDU is miniaturized imaging detector visualizing traces of individual particles of ionizing radiation in the normal environment such as classroom. It is simplified but still very powerful version of the MiniPIX detectors operated by NASA in space for radiation monitoring on the orbit. This CERN based technology gives very illustrative and easy to understand insight into complicated processes of particle physics.
        This miniature particle tracking detector can visualize even very small amounts of ionizing radiation which is present everywhere. Students can “see” the different types of radiation emitted by common materials such as piece of granite or ash, radioactivity accumulated in paper bag from vacuum cleaner or deposited on the face mask used for protection against infections. They can explore the variation of the air radioactivity during the day, hunt for cosmic muons and check their directions, see how altitude affects presence of radiation types. They can try to prepare their own (safe) radioactive source and try to construct the shielding against the radiation it emits. Students can directly observe how different radiation types interact with matter and what happens then. They can watch the laws of radioactive decay.
        The MiniPIX EDU device and its utilization for educational extents experiences collected by earlier projects such as “CERN technology in schools” of Mrs. Becky Parker. About 450 schools with more than 15.000 students participated since 2007. Similar project of the Institute of Experimental and Applied Physics (IEAP) of Czech Technical University in Prague and Jablotron company resulted in the prize winning MX-10 detector device. All these projects were done in close collaboration with CERN Medipix group. The new MiniPIX EDU device is based on the same Timepix chip as MX-10 but it is much smaller, more stable and less expensive.

        Speaker: JAKUBEK, Jan (ADVACAM)
      • 12:20
        Empty 20m
    • 12:40 14:00
      Lunch 1h 20m
    • 14:00 16:00
      03 Fusion Diagnostics and Technology

      Oral Session

      Convener: GIRARD, Sylvain
      • 14:00
        #03-251 Long Term Neutron Activation in JET DD Operation 20m

        The Joint European Torus (JET) is the largest operating fusion tokamak in the world. This allows performance of unique experiments that cannot be performed anywhere else. One such experiment is long term irradiation of different fusion relevant materials with neutrons produced in deuterium-deuterium (DD) and deuterium-tritium (DT) plasmas. During the last DD plasma campaign at JET in 2019 and 2020 several different ITER relevant materials and activation foils were irradiated in a specially design long-term irradiation station located inside the vacuum vessel with the purpose of testing activation of ITER materials by fusion neutrons. The samples were exposed to neutron fluence of 1.9×10$^{14}$ n/cm$^2$ during JET discharges performed in the experimental campaign over a period of 5 months. After the irradiation with DD neutrons, activation of materials was determined with measurements of long-lived isotopes in the samples such as Co-58, Co-60, Fe-59, etc.
        In order to support the long-term irradiation experiments at JET a series of Monte Carlo neutron transport calculations is required to calculate neutron fluxes, neutron flux spectrum, reaction rates, etc. A detailed computational model of the complex JET tokamak structure and the experimental position was developed in MCNP code and validated on experiments previously performed at JET. Another important parameter for computational analysis is the detailed description of the materials in the JET structure and activation samples. The detailed material composition together with the nuclear data libraries is used to properly describe physics of neutron interaction with materials during Monte Carlo simulations. The last important parameter for the Monte Carlo simulation is the source of neutrons. In DD plasma some DT reaction take place due to production of tritium during the DD reaction. Due to this the fraction of neutrons produced by DT reaction is of significant importance in computational analysis.

        Speaker: Mr ŽOHAR, Andrej (Jožef Stefan Institute)
      • 14:20
        #03-242 In-situ gamma irradiation testing of radiation hardened chips till 1MGy 20m

        Most of today’s commercial off-the-shelf (COTS) electronics are not specified to meet the demanding requirements of advanced nuclear applications requiring MGy-level TID tolerance. Examples are maintenance and diagnostics tasks in future burning plasma fusion reactors, for example the ITER. Applications such as interventions during nuclear accidents, dismantling of old nuclear power plants and disposal of radioactive waste also call for rad-hard electronics. Therefore, first, the development and, secondly, the testing of custom tailored MGy hardened integrated solutions become necessary for use in these environments. It will not only reduce the shielding but will make it possible to place electronics closer to front-end sensor transducers and actuators in a radiation environment.
        To address this need, Magics Instruments has developed a set of 5 chips which are highly resistant to ionizing radiation of over 1 MGy. The pioneering technology of Magics Instruments offers unprecedented opportunities for the use of electronics. Furthermore, Magics Instruments developed a modular test system to assess the radiation hardness of these chips even during irradiation (in-situ). Due to the modularity, this test setup can be customized for any other chip.
        The set of these 5 chips developed by Magics Instruments on behalf of F4E deploys an RS485 communication network for reading out sensors and driving actuators. The performances of 3 of these chips were tested on a set of samples during an irradiation assessment. In total 30 samples at once were measured, from which 24 in-situ. The first chip is a BiSS (RS485) interface communication Application Specific Integrated Circuit (ASIC) used for building up a sensor/actuator network. Another chip is a resistive bridge sensor signal conditioner ASIC capable of reading out resistive based sensors like temperature sensor, strain gauges, potentiometers and so on. The last chip tested is a resolver or Linear Variable Differential Transformer (LVDT)-to-digital converter that can be used to read out these sensors typically used in closed-loop motor control applications. The remaining 2 chips out of the set of 5 were tested during a previous irradiation assessment. One chip is a 10-channel relay driver chip and the second one a 10-channel limit switch sensing ASIC.
        The irradiation assessment was performed in the Co60 gamma underwater test facility at SCK CEN (called RITA) in Mol. The 30 chips were loaded in a closed container that was lowered down in a pool of water with the Co60 radioactive sources at the bottom. Prior to the actual irradiation assessment of these chips a reference dosimetry was performed to determine the dose rate distribution inside the irradiation volume. Axial variations of the dose rate along the chip-support were ~20%. On average, the exposure dose rate was 484 kGy/h. The chips themselves were connected to a Data-Acquisition system, using cables pulled through 4 tubes of 10 meters length what made it possible to perform a challenging task of the in-situ measurements.
        To achieve in-situ measurements, Magics developed a modular test rack computer making it possible to measure the chips fully autonomous in a remotely controlled setup with only limited lab equipment. Data was locally stored in a database on a hard drive and replicated continuously to a database in a cloud storage server for reasons of redundancy. A test setup failure warning mechanism was built in the test setup to lower the risk of data loss. When the test software would crash or power to the setup was interrupted, a warning email and text message was sent to undertake fully remotely necessary actions. An uninterruptable power supply was included to overcome short power interruptions. Even a self-restoring functionality, restarting measurements after failure, was built in.
        Multiple tests were continuously performed in a loop on all samples for 97 days to reach at a total ionizing dose of at least 1 MGy for each sample in the test. The higher located samples received even 1.2 MGy. The irradiation itself was interrupted twice to investigate the dynamic balance between defect generation due to irradiation and recovery. These interruptions were performed for 1 hour when average TID of 10 kGy and 100 kGy were reached. After the end of the irradiation, all samples were measured in-situ for 2 more weeks to investigate the recovery during the annealing phase. First, 1 week of annealing at room temperature was performed followed by another week of high temperature annealing at 100 °C. These tests were performed in a climate chamber. The full assessment was performed in-line with the ESCC 22900 standard: Total dose irradiation test standard . Furthermore, the ITER EEE Nuclear Radiation Compatibility Handbook was used as the reference throughout the full project, from design in the very beginning till the final post-processing of the data.
        The collected data was processed into readable plots showing the performance of the chips along the TID they received and their recovery. Almost 9000 plots were generated. Some of these data plots show a drift in the measured data, but all of these were acceptable. Therefore, it can be concluded that performance was proven over the full TID range. The data obtained during this irradiation assessment complete the data yet obtained for the 2 other chips, out of the set of 5, tested in a previous irradiation assessment. The performance of these chips was also proven over the full TID range. This means that the radiation hardness of the RS485 communication network to read out sensor and drive actuators, build up by the chips of Magics Instruments is proven.
        DISCLAIMER: The work leading to this publication has been funded by Fusion for Energy under specific contract F4E-OFC-0358-2-01-03. This publication reflects the views only of the author, and Fusion for Energy cannot be held responsible for any use which may be made of the information contained therein.

        Speaker: GEYS, David (Magics Instruments, Belgium)
      • 14:40
        #03-174 Assessment of radiation effects on the plasma current measurement at JET using fibre optics current sensor 20m

        The Fibre Optic Current Sensor (FOCS) is a system that will perform plasma current measurements at ITER during long plasma discharges under intense nuclear radiation. Plasma current measurements are important for safe machine operation. The FOCS must perform measurements is harsh tokamak environment, which include strong magnetic fields, high temperatures, and also significant levels of neurons and high energy photons. JET is a unique machine where the impact of such conditions on the sensor performance can be investigated experimentally. In particular, future DT operation will be fully ITER-representative in terms of 14 MeV neutron fluxes.
        Considering FOCS radiation hardness assessment, we have installed polarisation detection based FOCS systems at JET and performed measurements in various machine operating scenarios at currents up to 3.5 MA. One system uses the Fibrecore Ge-doped core spun fibre SLB 1250 with a 10 mm spun period, and the another system the Crystal Techno pure silica core LB1300 from with a 5 mm spun.
        Spun fibres may be more sensitive to radiation due to internal stresses induced as a result of the preform spinning. No information about the RIA in spun fibres could be found in literature. To address the possible problem of the RIA in spun fibres, radiation testing using Co60 gamma-radiation was performed. Four pure silica core fibres drawn from the same preform without and with different spinning rates, corresponding to the spun period of 5, 20, and 80 mm were exposed up to doses of ~200 kGy, relevant for the ITER environment. A general conclusion from the irradiation is that in case of pure silica fibres the spinning influence on the radiation sensitivity is not significant when the FOCS performance is of concern. The RIA at 1550 nm, FOCS operation wavelength, reached saturation at a dose of ~100 kGy at a level acceptable for ITER.
        To address the radiation effect on the FOCS operating in a tokamak environment operating with hydrogen, which is relevant for the ITER first phase operation, analysis of the performance of FOCSs installed at JET provides necessary information. Data accumulated during more than four years of operation were analysed. The radiation dose was estimated based on scaling of the integrated plasma current. Possible transient effect have been observed, but on a long term the performance variation are within measurement uncertainties.

        Speaker: GOUSSAROV, Andrei (SCK CEN)
      • 15:00
        #03-134 Fusion fuel ion ratio measurements at JET using neutron spectroscopy 20m

        The fuel ion ratio, $n_T/n_D$, is an important plasma parameter in magnetically confined fusion, especially for future fusion devices (ITER, DEMO), which plan to achieve fusion through deuterium-tritium (DT) plasmas. In order to maximize the fusion power output, it is necessary to ensure equal number densities of tritium and deuterium (i.e. $n_T/n_D = 1$). Neutron spectroscopy offers the possibility to measure the fuel ion ratio by comparing the contribution to the neutron spectrum from 14 MeV neutrons generated by the $\text{T(D, n)}^4_2\text{He}$ reaction, and 2.5 MeV neutrons from the $\text{D(D, n)}^3_2\text{He}$ reaction. Consequently, neutron spectroscopy has been identified as one of the primary diagnostics for measuring $n_T/n_D$ at ITER. Measurements of the fuel ion ratio is required at ITER over a range of $0.01 < n_T/n_D < 10$ with a precision of 20% and time resolution of 100 ms. In future fusion power plants, the ion ratio will likely be a continuously monitored parameter to ensure the reactor is running at maximum capacity. In this paper we describe a method for measuring the fuel ion ratio for low concentrations of tritium ($n_T/n_D < 1$) at the Joint European Torus (JET) using the time-of-flight (TOF) neutron spectrometer TOFOR. TOFOR has a vertical sight-line of the JET plasma and consists of 37 plastic scintillation detectors which when used in coincidence provide TOF spectra from which information on the energy spectrum of the different neutron components can be inferred. Measurements of the fuel ion temperature are commonly performed using TOFOR data. In this paper, the fuel ion ratio is determined by comparing the TOF peaks corresponding to the two different neutron energies (2.5 MeV & 14 MeV) after removing background and scattered components. Finally, an estimate of the upper limit of $n_T/n_D$ at which the ratio can be accurately determined using a forward scattering TOFOR-like spectrometer is presented.

        Speaker: ERIKSSON, Benjamin (Uppsala University)
      • 15:20
        #03-94 Designed A Real-time Full-Range Digitizer for Neutron Flux Monitor On EAST 20m

        The neutron flux monitor (NFM) is a standard diagnostic for fusion neutron yield measurement on experimental advanced superconduct tokamak (EAST), and the neutron yield is a most important parameter to research of plasma auxiliary heating. After the upgrade of EAST, higher requirements for neutron flux measurements are put forward. Based on fast ADC (14bit, 500MSPS) and a high-performance FPGA, a four-channel real-time full-range digitizer for NFM was designed. In order to meet the requirement of fusion neutron flux wide-range measurement during high-parameter operation, the advantages of pulse counting modes, Campbell modes and current modes were combined to expand the dynamic range, so that the measuring range could reach 8 orders of magnitude. In addition to reduce the interference of gamma ray backgrounds on neutron flux measurements, a pulse signal processing technology was used to real-time distinguish between neutrons and gammas. Furthermore, to meet the requirements of mass data storage and high-speed transmission in the case of high neutron flux, a DDR3-based controller and PXIe bus technology-based DMA mode were separately designed. Moreover, a series of test were performed on laboratory and EAST device. Those test results have reached the expected design indicators, and the time resolution could reach 1ms, which proved the feasibility of NFM for EAST neutron flux measurements.

        Speaker: Ms YANG, Li (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Science)
      • 15:40
        #03-79 Metrology of acquisition chains and signal processing of LMJ experiments 20m

        Since the first experiment in 2014, more and more plasma diagnostics are being deployed on the Laser MégaJoule (LMJ) facility manufactured by CEA/DAM. These diagnostics aim at measuring radiations or particles emitted during laser experiments to study high-energy physics, especially inertial confinement fusion (ICF). Up to 176 laser beams, converging on a millimeter-sized target placed at the center of a 10-meter diameter chamber will be installed. Different types of sensors surround the LMJ target chamber (coaxial diodes for soft x-rays, photomultipliers and scintillators for neutrons, etc.…) and realize the conversion of the quantities of interest to an electric signal. The signal is then transmitted via coaxial cables, acquired by a broadband oscilloscope, and digitally post-processed. Each step of this typical acquisition chain adds measurement errors and increases the global uncertainty. In this paper, a study of the whole acquisition chain, including signal processing algorithms, will be presented. First, a numerical model of the digitizer alongside a specific hardware system designed to perform its metrology in situ will be presented. This system has been customized to be included inside an acquisition rack, and it allows quantifying several parameters before each laser experiment. It computes errors sources such as offset, gain and skew, and provides a measurement of the effective number of bits (ENOB) of the digitizer. This number gives insights concerning thermal noise, phase noise and non-linearity, and the ENOB is measured at different frequency values in the range of interest. This range can go up to a few GHz for the measurement of a hohlraum radiation temperature for example. The experimental characterization of the electrical chain via its transfer function measurement will also be detailed. Finally, the numerical methods deployed to handle the inverse problem, based on deconvolution processes, will be introduced, including future developments exploiting Bayesian inferences and statistical approaches.

        Speaker: Dr TRAUCHESSEC, Vincent (CEA)
    • 14:00 16:00
      04 Research Reactors and Particle Accelerators

      Oral Session

      Conveners: Dr KOHSE, Gordon (Massachusetts Institute of Technology, Nuclear Reactor Laboratory, Cambridge, Massachusetts, USA) , REYNARD-CARETTE, Christelle (Aix-Marseille University)
      • 14:00
        #04-80 Sensitivity Analysis of an Advanced Measurement Method for Thermal Neutrons Absorbers Detection in Irradiated Beryllium 20m

        Operation experience in many Material Testing Reactors, where beryllium moderator or reflector is used, has shown that accumulation of neutron absorbers, so-called poisons, has a non-negligible impact on both core performance and safety aspects due to n-Be interactions, and needs to be addressed properly in calculations. As a result, depletion of beryllium, called also beryllium poisoning, was implemented in core follow-up deterministic calculations in several facilities. However, Material Testing Reactors are designed mainly for performing irradiation tests of material behaviour in severe conditions, resulting in frequent repositioning of the core elements in order to host experiments. This specificity makes almost impossible to follow the exact irradiation history of each beryllium elements and consequently inferring the exact accumulation of poisons. Thus, there is a strong need for developing a methodology enabling measurement and assessment of beryllium poisoning.
        We addressed the problem of quantifying poisons accumulation using neutron transmission method. It is based on measuring transmission and absorption of thermal neutrons in beryllium elements. The measurement set is composed of a neutron source, a neutron moderator and a neutron detector, mounted on a bench, with beryllium sample placed between detector and moderator. In the study, we used the MARIA reactor’s beryllium moderator blocks as material samples. Their characteristic truncate shape contributes to uneven axial neutron flux distribution in the core. To enable evaluation of axial distribution of poisons, both neutron source and detector move simultaneously along the block. A non-irradiated beryllium block has been measured and constitutes a reference point. We then performed series of irradiations in the NCBJ’s MARIA reactor on several beryllium blocks delivered in the same batch. Comparing measurement results of irradiated blocks to the reference results, allows evaluating thermal neutrons absorbed in irradiated beryllium. Conducting such experiments requires taking into account several effects that might affect the measurement accuracy. In order to evaluate these effects, we performed a sensitivity analysis and all the above steps were simulated with the SERPENT2 Monte-Carlo code, using JEFF3.1.1 nuclear data library. We focus on analysis of expected uncertainties related chosen detector, impurities in beryllium, nuclear data uncertainty and gamma activation of irradiated beryllium.
        In our study, we analysed the possibility of using following detectors: fission chamber with 235U deposit, 3He detector, 10B lined-detector. As no pure thermal neutron source exists, polyethylene-moderated neutron sources have been tested to be used for measurements, taking into account characteristics of the system. Neutron sources considered in this study include Am-Be, Pu-Be, Sb-Be, 252Cf. For each of these sources, we studied the optimal polyethylene thickness.
        In this paper, we present the newly developed method for measurement of these neutron absorbers’ concentration, we discuss calibration and sensitivity study of the above described measurement setup. The methodology and codes used for simulations are described. The results of the optimization study are detailed, and choice of specific type of detector and neutron sources are justified. Summarizing the results, we also give some recommendations.

        Speaker: WROBLEWSKA, Malgorzata
      • 14:20
        #04-126 CEA-JSI Experimental Benchmark for validation of the modelling of neutron and gamma-ray detection instrumentation used in the JSI TRIGA reactor 20m

        Constant improvements of the computational power and methods as well as demands of precise and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modelled in order to provide selective and precise neutron and gamma-ray measurements.
        With this aim in mind, the French Alternative Energies and Atomic Energy Commission (CEA) and ”Jožef Stefan” Institute (JSI) have started an experimental program dedicated to the set up of a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and uncertainty evaluation for the main neutron and gamma-ray sensor types used in the JSI TRIGA (Training, Research, Isotopes, General Atomics) reactor. This benchmark should also provide recommendations for instrumentation modelling and analysis for mixed gamma-neutron fields.
        CEA has setup a simplified TRIPOLI-4 (three-dimensional polykinetic code) modelling scheme of the JSI TRIGA reactor based on the information available in the International Reactor Physics Experiment Evaluation Project (IRPhEP) and International Criticality Safety Benchmark Evaluation Project (ICSBEP) benchmarks in order to simulate the sensor responses for neutrons and gamma-rays. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme taking into account the geometry of the sensor and of the irradiation device as well as the surrounding irradiation conditions.
        This paper presents main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained with the Monte-Carlo N-Particle transport code (MCNP) – Evaluated Nuclear Data File B-VII.0 (ENDF/B-VII.0) - International Reactor Dosimetry and Fusion File (IRDFF) calculation scheme. This paper will conclude by giving some information on the second step of the experimental program consisting of a dedicated experimental campaign to be carried out in the following months in the TRIGA reactor core.

        Speakers: FAUSSER, Clément (CEA) , Mr THIOLLAY, Nicolas (CEA) , Mr DESTOUCHES, Christophe (CEA)
      • 14:40
        #04-75 3D thermal and radiation-matter interaction simulations of a SiC solid-state detector for neutron flux measurements in JSI TRIGA Mark II research reactor 20m

        In the nuclear reactor, it is crucial to measure key parameters such as neutron and photon fluxes or nuclear heating rate for a better understanding of the behavior of nuclear fuels or materials subjected to nuclear radiation. The coupling of different on-line measurement sensors is at the heart of these scientific objectives. The LIMMEX laboratory (Laboratory of Instrumentation and Measurement Methods in EXtreme media, a joint laboratory between Aix-Marseille University and the CEA) has been carrying out experimental and numerical works (innovation, design, characterization, calibration and optimization for laboratory and nuclear conditions) for accurate on-line measurements by means of differential calorimeters and semi-conductor detectors for many years. A new project, called SiC-CALO, aims to develop a multi-sensor device by coupling a Silicon Carbide (SiC) – solid-state based detector and a calorimeter for measuring simultaneously neutron fluxes and nuclear heating rates. Indeed, one objective is to know accurately the contribution of each radiation on nuclear heating rate thanks to the coupling of these two kinds of sensor and their combined interpretation. The finality of this project is to use this device in the core of the Jules Horowitz Reactor (JHR, under construction at the CEA Cadarache research center). The JHR core characteristics are, for a nominal power of 100 MWth, a high fast neutron flux (5.5 ×10^14 n·cm^-2·s^-1 for an energy > 1 MeV) leading to a high nuclear heating rate (up to 20 W·g^-1 in Aluminum). To be able to endure all of these harsh conditions, a SiC-based detector has been studied within the framework of a previous joint European project, called I_SMART (Innovative Sensor for Material Ageing and Radiation Testing). SiC material was chosen, among other wide band-gap semiconductors, thanks to its several advantages like its fast response (a few ns), its stability under radiations thanks to its wide bandgap energy, its low leakage current and its low thermal resistance and temperature gradient. The I_SMART project was intended to develop an innovative system made of SiC for radiation detection at very high temperature (> 500 °C) and under high integral neutron flux (about 10^14^-2.s^-1). The main result of this project was the development of SiC p+n junction diodes with implanted 10-B thermal neutron converter layer to measure thermal and fast neutrons fluxes. With regard to the short-term outlooks, the SiC-based detector performances can be improved through the using of a 6-Li converter for a better neutron/gamma discrimination, by decreasing the metal contact thickness or by coupling the detector with a suited current amplifier and a treatment process of pulses. The middle-term outlook is to adapt the SiC-diodes for fast neutrons measurements inside tokamaks (high temperature about 400 °C - 500 °C and extreme magnetic fields about 4 T). The last outlook is to extend the measurement range to values higher than those which was measured previously during the I_SMART project, i.e. 9.4 ×10^8 n·cm^-2·s^-1. Consequently, in 2021, an irradiation campaign at the Jožef Stefan Institute TRIGA Mark II research reactor in Slovenia is scheduled to measure higher neutron fluxes inside an in-core dry air triangular irradiation channel with a SiC detector and a commercial pCVD (polycrystalline Chemical Vapor Deposition) diamond detector (CIVIDEC Instrumentation). Main feature of this channel is its size, as it measures almost 6 cm in diameter. Because of the harsh conditions within this channel: a maximum integral neutron flux of about 1.2 ×10^13 n·cm^-2·s^-1, a nuclear heating rate lower than 0.1 W·g^-1 and a low heat transfer coefficient due to air natural convection, it is essential to predict the most precisely the behavior of the SiC detector for this environment.
        Consequently, this paper will focus on 3D numerical studies dedicated to the definition of the most optimized device for the measurement campaign.
        In a first part, a state of the art of the SiC detectors will be realized.
        In a second part, 3D simulations dedicated to the interactions between nuclear radiations and matter will be performed by means of MCNP calculation code (Monte-Carlo N-Particle transport code). The goal of these simulations is to estimate the flux attenuation and the nuclear heating rate in each part of the studied system for different axial positions by considering input source spectra, which represent the irradiation conditions of the triangular dry-air channel of the JSI TRIGA Mark II reactor core. The studied system will contain all of the components of the detector such as the housing, the duralumin support, the SiC diode putted on its alumina support, the connectors, the screws and the instrumentation cables.
        In the last part, 3D thermal simulation results will be presented. A 3D thermal model built and solved with COMSOL Multiphysics finite element code will be detailed for the same system as that used for MCNP code. The thermal model will take into account heat sources estimated by MCNP simulations, conductive, convective and radiative heat transfers inside the system and natural convection boundary conditions by applying specific heat transfer coefficient values determined thanks to a correlation formula depending on the heat sources. The aim of these thermal simulations is to determine the temperature field inside the system and in particular in the SiC diode. A parametrical study will be given and discussed in order to define the optimized detector design.

        Speaker: Mr VALERO, Valentin (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 15:00
        #04-37 Numerical and experimental characterization of the neutron flux and spectra in the core of the CNESTEN’s TRIGA Mark II research reactor 20m

        Education, training and isotopes production are the most important uses of the Moroccan 2 MW TRIGA Mark II reactor situated at the National Center for Energy Sciences and Nuclear Techniques (CNESTEN, Morocco). To develop new R&D projects in research reactors, the particular and advanced knowledge of neutron and photon flux distribution, within and around the reactor core, is crucial.

        In order to precisely prepare the forthcoming experiments in the reactor, a detailed model of the reactor core was developed using the 3D continuous energy Monte Carlo code TRIPOLI-4 [1] and the continuous energy cross-section data from the JEFF3.1.1 nuclear data library. In particular, all the geometries and compositions of materials were described in the TRIPOLI-4 simulation model with good details. This new model was used to carry out preliminary neutron and photon calculations to estimate flux levels in the irradiation channels as well as to calculate kinetic parameters of the reactor, core excess reactivity, integral control rods worth and power peaking factors. As a first step of the validation of the model, the obtained results were compared with the experimental ones available in the Final Safety Analysis Report (FSAR) of the TRIGA reactor [2]. Fairly good agreement was found, which indicates that the Monte Carlo model is accurate enough to perform criticality calculations of TRIGA reactor.

        Furthermore, this work aims at experimentally characterizing the neutron flux and energy spectra in various irradiation channels inside and outside the reactor core. Absolute flux measurements will be carried out using the neutron activation technique [3] [4]. To set up the experimental design for the activation experiments a series of preliminary calculations were performed using the TRIPOLI-4 model to calculate the expected gamma flux/intensity levels of various materials after irradiations in different positions in the core and reflector. Different activation foils with known characteristics will be then irradiated and the activity of several isotopes will be measured with the Gamma Spectrometry Method. During the irradiation process, the reactor power will be monitored via two fission chambers (FCs) adjacently fixed in the reactor cavity. The count rate is proportional to the neutron flux, which is proportional to the reactor power in the operation range. Sources of uncertainty in reactor power can therefore be either the calibration of the FCs, or the uncertainty in the count rate. Along with FCs, gold monitors will be irradiated in an ex-core position in order to monitor the neutron flux leading to estimate the reactor power during each irradiation. Experimental measurements will be compared with calculated neutron flux evaluated with the new TRIPOLI-4 model of the reactor, as well as against the reference calculation scheme using MCNP5 [5] and ENDFB-VII, developed by the CNESTEN [2] [6].

        The final results of the above-mentioned experimental campaign will be presented in this study only after the betterment of the current situation of the Covid-19 pandemic. With the gradual opening of the airports and boundaries, we will be able to carry out our experiments with full concentration and in good conditions.


        [1] E. Brun, F. Damian, C. Diop, E. Dumonteil, F. Hugot, C. Jouanne, Y. Lee, F. Malvagi, A. Mazzolo, O. Petit, J. Trama, T. Visonneau et A. Zoia, « TRIPOLI-4®, CEA, EDF and AREVA reference Monte Carlo code », Annals of Nuclear Energy , vol. 82, pp. 151-160 , 2015.
        [2] B. El Bakkari et al., « Monte Carlo modelling of TRIGA research reactor », Radiation Physics and Chemistry, vol. 79, pp. 1022-1030, 2010.
        [3] A. Borio di Tigliole et al., « TRIGA reactor absolute neutron flux measurement using activated isotopes », Progress in Nuclear Energy, vol. 70, pp. 249-255, 2014.
        [4] V. Radulovic et al., « Validation of absolute axial neutron flux distribution calculations with MCNP with z197Au (n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor », Applied Radiation and Isotopes, vol. 84, pp. 57-65, 2014.
        [5] « MCNP- A General Monte Carlo N-Particle Transport Code, Version 5 », X.-5. Monte Carlo Team, Los Alamos National Laboratory, 2003.
        [6] E. Chham et al., « Neutronic and thermal–hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core », Applied Radiation and Isotopes, vol. 116, p. 178‐184, 2016.

        Speaker: Mr GHNINOU, Hamza (CEA, IRESNE, Cadarache, DER/SPESI, F-13108 Saint-Paul-lez-Durance, France.)
      • 15:20
        #04-224 Optimization of Next-Generation Fast-Spectrum Self-Powered Neutron Detectors using Geant4 20m

        Self-powered neutron detectors (SPND) have been a common diagnostic tool for intra-core neutron flux mapping in thermal nuclear reactors for more than 45 years. They are attractive flux monitors as they are compact, simple, and produce a signal proportional to local neutron flux without the need for an external source of power. Signal in these detectors is driven by electrons generated from nuclear reactions within the emitter in the detector. As the next-generation reactors are on the horizon and some of them are designed to be operating with harder neutron spectra, it is imperative to develop diagnostic tools tuned to their neutron spectrum, peaking around 0.5 MeV. However, the current state-of-the-art for SPNDs is optimized for thermal neutron interactions. We will be discussing our efforts to develop fast-spectrum SPNDs (FS-SPND) that are sensitive to neutrons with energies ranging from thermal up to 1 MeV. We have performed an in-depth analysis of ENDF VII.1 neutron-capture cross sections and have identified 4 novel materials that are suitable emitter candidate materials to measure fast neutrons, all are stable mid-shell nuclei in the region between doubly-magic 132Sn and 208Pb. We will be discussing the results of Geant4 simulations for each emitter candidate with detector parameters optimized to maximize overall detector efficiency as well as an exploration of complex detector geometries.ocess.

        Speaker: GOETZ, Kathleen (UTK/ORNL)
      • 15:40
        #04-158 Fragment production in 200MeV proton reaction on 232Th, experimental and simulated data. 20m

        This paper shows the results from the experiment at Joint Institute for Nuclear Research in Dubna and was carried out in 2014. The sample, using a special device, were placed in the accelerator chamber at a radius corresponding to the energy of protons 200 MeV at the current of 0.3 A. The position of the 200 MeV beam was determined by placing an aluminium foil inside the accelerator and perpendicular to the proton beam and evaluating the activity of the foil. For the target, foils of Th232 were used with a thickness of 100 microns and a weight of 149.5 mg placed between two Al foils with thickness of 50mkm. The foil area was 1.5 cm2. After target irradiation, foils where removed from the device and moved to YSNAPP-2 complex where was separately measured the spectra of radiation from the Th and Al foils by HPGe detectors of the CANBERRA company with efficiency 18% and resolution of 1.9 keV in the line 1332 keV. The processing of the gamma spectra was carried out using the DEIMOS32 program to find positions of peaks, their areas and other parameters. The identification of the nuclei formed in Th232 samples as a result of nuclear reactions with protons and their reaction rates was carried out using a set of scripts based on the Ruby programming language. For modelling of the experiment MCNP 6.1 was used. The main purpose was to simulate creation of parent residual nuclei during proton irradiation, their distribution and escaping from the Th foil and calculate the cross-section of the residual nuclei. The experimental data on the fragmentation of the Th232 nucleus under the influence of protons in the energy of 200 MeV have been processed As a result we compared the calculated and simulated results, found residual nuclei and their cross-sections.

        Speaker: HOLOMB, Robert (Brno University of Technology, Faculty of Electrical Engineering and Communication, Department of Electrical Power Engineering, Brno, Czech Republic, Uzhgorod National University)
    • 14:00 16:00
      06 Severe Accident Monitoring

      Oral Session

      Conveners: LYOUSSI, Abdallah (CEA) , RUDDY, Frank (Ruddy Consulting)
      • 14:00
        #06-240 Reactor core accidental conditions effects on infrared multispectral pyrometry temperature sensors 20m

        The Loss Of Coolant Accident (LOCA) is one of the most considered accidental scenario in nuclear power plant design. It occurs due to a break of the primary cooling circuit, which induces a strong pressure decrease, this leads to an overheating of the fuel cladding. The water steam and high temperature environment undermine the claddings and can lead to their burst with the release of fission products.

        The Light-water One Rod Equipment for LOCA Experimental Investigations (LORELEI) is a device that will be dedicated to such accident simulations and studies. It will be implemented in the Jules Horowitz research Reactor (JHR) which is under construction at the CEA Cadarache center. The main objective of the LORELEI device will be the study of fuel claddings behaviors under extreme reactor conditions. In such experiments, the cladding surface temperature monitoring is essential; it allows linking the burst conditions with the temperature. However, to minimize the perturbation and avoiding any change of the burst conditions, this measurement needs to be non-invasive and preferably contactless, which excludes the use of conventional temperature measurement techniques such as thermocouples and thermistors. The IR pyrometry based temperature measurement technologies present a suitable solution for this application. It allows a contactless temperature measurement with a good accuracy, high speed and reliability. However, the conventional pyrometers are not adapted for harsh environments application and could not survive under the extreme core conditions. In order to overcome this issue, we developed a sensor prototype based on the infra-red multispectral pyrometry with an optical collection head (figure 1) that can withstand both reactor core nominal and accidental conditions; high radiations levels (about 1020 fast neutron/cm² fluence and several GGy of gamma dose), high temperatures (up to 1000°C during the accident) and pressure (about 90 bars).

        The use of multispectral pyrometry technique allows to estimate both temperature and emissivity, which makes possible to overcome the issue of variation in the emissivity of the cladding and the changes of the system transmission during the operation. The infra-red wavelength range is selected due to the high performances of silica-based materials and their high resistance to ionizing radiations compared to the UV-VIS domain.

        Before the sensor integration in the LORELEI experience, validation and qualification procedures are required. This will estimate the reliability and durability of the system under conditions that are representative of the final application field.
        In order to provide preliminary results on the reactor core environment effects on our system, we separated the core constraints and the device was tested for each one using different CEA facilities. In addition, we performed several studies to estimate the combined effects of those constraints on the system. We focused in particular on the combined high radiation/temperature effects on the optical element response and the temperature/pressure effects on the sensor body. Finally, we reached a system accuracy with less than ~ 2% temperature measurement error in a range of 700 to 1200 °C.

        Speaker: LADACI, Ayoub (CEA)
      • 14:20
        #06-239 X-Ray Imaging Calibration for Fuel-Coolant Interaction Experimental Facilities 20m

        During a severe accident in sodium-cooled fast reactors, jets of molten nuclear fuel may penetrate into the coolant resulting in fuel-coolant interactions (FCI). Experimental programs are being conducted to study this phenomenology and to support the models development for evaluating consequences of severe accident. Due to the optical opacity of the test section walls and sodium coolant, high-speed X-ray imaging is the preferred technique for FCI visualization. The configuration of these X-ray imaging systems, whereby the test section is installed between a fan-beam X-ray source and a scintillator-image intensifier projecting an image in the visual spectrum onto a digital camera, entails certain imaging artefacts and uncertainties, not limited to vignetting, geometric distortion and noise in the detected photon flux. Although the X-ray imaging configuration can observe FCI process qualitatively, it ideally requires precise calibration to enable detailed quantitative characterization of the FCI. Calibration tests have been conducted for a new, enlarged, sodium test section at the MELT facility. To this end, ‘phantom’ models have been fabricated using polyethylene, either steel or hafnia powder, and empty cavities to represent the sodium, molten fuel and sodium vapor phases, observed during FCIs, respectively. The checkerboard configuration of the phantom enables calibration and correction for distortion artefacts which magnify features towards the edge of the field of view with fan-beam X-ray imaging. Polydisperse steel ball configurations enable precise determination of the minimum object size detectable by the camera, and the estimation of parallax errors which introduce uncertainty in an object’s silhouette dimensions due to the uncertainty in its position within the depth field. Analysis of these calibration tests is presented with the objective of establishing a universal procedure for the optimization of FCI visualization experiments.

        Speaker: JOURNEAU, Christophe (CEA)
      • 14:40
        #06-208 Optimized High Temperature Irradiation Resistant Thermocouple for Fast Response Measurements 20m

        The High Temperature Irradiation Resistant thermocouple (HTIR-TC) is the only temperature probe proven to withstand both the high temperatures (e.g. >1290 °C) and high radiation (e.g. up to a fluence of 1E21 n/cm2) of nuclear reactor fuel design tests and/or over-temperature accident conditions. The HTIR-TC heat treatment, calibration, and in-pile performance during the Advanced Gas Reactor 5/6/7 fuels tests inside the Idaho National Laboratory’s Advanced Test Reactor have been previously shared. The current work describes the improved performance of the molybdenum versus niobium thermocouple by utilizing a coaxial design—i.e. single-wire grounded to the outer sheath. The thermocouple junction is formed at the end of the coaxial cabling by swaging the sheath down and welding to the inner wire. This optimized HTIR-TC is more concise by simplifying the design while allowing for more robust individual components. The niobium and molybdenum thermoelements can be interchanged as sheath or wire depending on the application. Using a plunge test in flowing water, the coaxial build of the HTIR-TC was found to be 30x faster in response time than ungrounded Type K TCs, and 10x faster than both grounded Type K TCs and traditional ungrounded HTIR-TCs (i.e. two-wire configuration). Further, by capitalizing on the coaxial design, a multi-core HTIR probe has been proven with multiple ‘single-pole’ wires down the length of the sheath. Each wire is then terminated and formed into a junction with the inside of the outer sheath at the location where temperature measurement is desired. This multi-core thermocouple design has been dubbed “demicouple.” The primary application of the HTIR demicouple is during fuel experiments to achieve multiple fuel pin centerline temperature measurements in one compact sensor. However, the shared ‘common’ leg between demicouple junctions reduces error propagation in secondary measurements such as temperature differentials.

        Speaker: SKIFTON, Richard (Idaho National Laboratory)
      • 15:00
        #06-62 Design and characterization of a gamma imaging system for fuel rod deformations 20m

        The study of accidental situations is one of the major asset for nuclear industry in guaranteeing the so-called regulated safety of its electricity production facilities. The Loss Of Coolant Accident (LOCA) is tested in experimental reactor and hot cell to verify the mechanical properties of the fuel rod, the safety criteria and the associated calculation codes. The VINON-LOCA experimental set-up, currently under development in CEA Cadarache, aims at observing, online, the Fuel Fragmentation, Relocation and Dispersal (FFRD) during a LOCA sequence. The tests will consist in placing a fuel rod in a shielded cell, which will be heated by magnetic induction in order to reproduce the temperature conditions of the first phase of a LOCA transient. Instrumentation positioned in the cell will be placed close to the fuel rod in order to monitor its behaviour during heating. A proposed technique to quantify the fuel rod deformation is the use of gamma imaging. In this context, we have designed and developed, in partnership with CEA Cadarache, EDF and Framatome, a gamma imager to monitor the deformation of the fuel rod. Preliminary experimental tests carried out at LAMIR Laboratory of CEA Cadarache have estimated that the variation in the diameter of the fuel rod during heating is 2 to 3 mm. The gamma emitters, which will enable the swelling to be monitored, are derived from the fission products of the pre-irradiated rod. Only the most energetic gamma rays passing through the cladding of the rod, and then through the experiment's containment, will be imaged.
        The work presented therefore concerns, firstly, the design of a collimator offering the best compromise between mechanical feasibility, cost, signal-to-noise ratio and angular resolution, thus making it possible to visualise the deformations of the fuel rod, and secondly, a preliminary evaluation of the experimental performance of the proposed gamma imager.
        The chosen detector is the Widepix pixel detector developed by ADVACAM s.r.o. The Widepix detector is based on the Timepix technology resulting from the Medipix international collaboration founded at CERN. It consists in 512 × 1280 square pixels of 55 µm-pitch, hybridised to a 1 mm-thick semiconductor, i.e. a pixelated detection surface of 20 cm². In addition to positions of interactions, the detector also provides their energies, which can be used to reduce the intrinsic resolution of the pinholes in the collimator.
        The collimator was dimensioned using Monte-Carlo simulations with MCNP6.2. During the design of the collimator, the material (tungsten), the pattern (with parallel holes) and the geometry (thickness, septa and pinhole diameter) were studied. Among the strong hypotheses, we chose to design the collimator with the most energetic gamma rays emitted by the radioisotope with the highest photon flux, namely Cs-137 emitting gamma rays of 661.7 keV. We finally ended up with a collimator of 6.2 cm thickness, pierced with an arrangement of pinholes of 1 mm diameter and 1 mm septa. This design results in an angular resolution of 0.48° for a field of view of 2.8 × 7 cm² at 25 cm.
        Preliminary experimental measurements were then carried out on a Cs-137 irradiator located at the LNHB of CEA Saclay, and validated the possibility of identifying a variation in the incoming gamma rays flux diameter of the order of 2 mm using the proposed imager.

        Speaker: Dr AMOYAL, Guillaume (Université Paris-Saclay, CEA, List, F-91120 Palaiseau, France.)
      • 15:20
        #06-49 MERARG Experimental Loop: A new deconvolution Method for FGR analysis 20m

        MERARG experimental loop hosted at the LECA-STAR Hot Laboratory (CEA Cadarache) allows characterizing nuclear fuels with respect to the behavior of fission gases during thermal transients representative of accident conditions such as for example loss of coolant accident (LOCA). MERARG loop consists of three main parts: 1- an induction furnace 2- a gamma spectrometry detector 3- a glove box containing a micro gas chromatograph (μ-GC). The furnace is located in a high activity cell while gamma detector and μ-GC are both located in the back zone of the hot cell. The fuel pellet to be characterized (generally with its cladding) is introduced into a metal crucible. The sealed enclosure of the furnace is swept by the circulation of a carrier gas (argon or dry air) with a low flow rate (about 60 cm3 /min). As the temperature rises, the released gases are driven to the back zone to the gamma spectrometry station measuring their activities on-line. After passing in front of this on-line gamma station, the gases are then driven to a glove box where they are analyzed by µ-GC and then collected in storage capacities. In order to simulate the thermal characteristics of a loss of coolant accident, the annealing test consists in a first stage of thermalization at 300°C followed by an increase in temperature up to 1200°C (according to a ramp of 0.2°C/s or 20°C/s).

        Consequently, fission gas release measurement is carried out at a distance of several meters from the sample where the release actually occurs, in a dedicated counting chamber located outside the hot cell itself (i.e. at the so-called back zone). It is therefore necessary to deconvolve the acquired experimental data in order to “rebuild” the release rate data at the sample position. This requires taking into account fission gas flow out of the furnace to the counting chamber and its dilution in the carrier gas. Up to now, a previous deconvolution procedure ran well for the major part of the experiments performed but showed, in very few cases, some drawbacks such as “negative release” or “unappropriate noise” yielding in unusable-reconstructed spectra.
        This paper aims to expose a new mathematical deconvolution method for re-processing the experimental data obtained using Laplace transformation. This new method is based on a simulation of the measurement made during the heating of a glass capsule containing 85Kr that opened abruptly during a heat treatment. The release of the 85 Kr during capsule opening was considered as a Dirac function, which allows the fitting of the 85Kr measured at the gamma counter position by an analytically calculated transfer function. The obtained numerical transfer function is used to build elementary functions whose combination allows to describe the experimental release in a satisfactory manner while overcoming numerical instability issues. This deconvolution method provides an accurate modelling of the measured release and allows the analysis of the annealing test data making it possible to describe physical phenomena directly related to the gas release of the samples such as the release of Kr85 by bursts at different temperature levels.

        Speaker: Mr LOSFELD, Jimmy (CEA, DES, IRESNE, DEC/SA3E)
      • 15:40
        #06-47 Analysis of the Signal over Noise Ratio of the hodoscope determined by Monte Carlo calculation 20m

        The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA) for the purpose of the CABRI International Program (CIP), managed by IRSN in the framework of an OECD/NEA agreement. The hodoscope equipment installed in the CABRI reactor is a unique online fuel motion monitoring system, thanks to the measurement of the fast neutrons emitted during a power pulse by a tested rod positioned inside a dedicated test loop reproducing PWR conditions. This system is dedicated to the analysis of fuel displacement and degradation of the tested rod during the power excursion. Hence, one of the most important parameter measured by the hodoscope detectors is the Signal over Noise Ratio (SNR), characterizing the fraction of neutrons directly coming from the test rod (“signal”) over neutrons coming from the core (“noise”). Nowadays, this value is measured during a power plateau of the CABRI reactor, a few days before the experiment.
        It is interesting to calculate the SNR in order to define some quantitative criterions to improve hodoscope measurements and to understand if any modification linked to the test loop may significantly change this essential parameter.
        In this article, the method used to calculate the SNR using MCNP6.2 Monte Carlo code will be detailed. Because the hodoscope detectors are located far away from the test rod (up to 4 meters), a 2D model of CABRI core and instrumentation has been implemented. No variance reduction techniques have been used to solve this problem in order to record the place of birth of neutron which contributes to the different scores with the goal to perform a detailed analysis of the SNR.
        Another parameter of interest which will be evaluated by means of this new model is the so-called “scattering coefficient”, which corresponds to the fraction of neutrons coming from the test rod and being scattered between their birth and their detection. This parameter is used to enhance the analysis of the fuel displacement which may happen during the power transient.
        Finally, the comparison between calculated and measured SNR for a case study will be carried out. A quite good agreement between the 2D simulations and experiments recently performed in the CABRI reactor has been obtained.

        Speakers: MIROTTA, Salvatore (IRSN) , Dr DI SALVO, Jacques (IRSN)
    • 14:00 16:00
      07 Nuclear Fuel Cycle, Safeguards and Homeland Security

      Oral Session

      Conveners: KOUZES, Richard (Pacific Northwest National Laboratory, USA) , MOUTIERS, Gilles (CEA, France)
      • 14:00
        #07-226 Study of natural uranium fuel for new reactor design TEPLATOR 20m

        The TEPLATOR is a new type of nuclear reactor which the main purpose is producing heat for district heating. It is designed as a special thermal reactor with 55 fuel channels for fuel assemblies, which is moderated and cooled by heavy water and operated around atmospheric pressure. The TEPLATOR DEMO is designed for using irradiated fuel from PWR or BWR reactors. Using heavy water as the moderator and coolant in this reactor concept allows to use natural uranium as an alternative fuel in case that the irradiated fuel is not available for some reasons. This solution is suitable because of the price of natural uranium and the absence of costly fuel enrichment. This article is focused on deeper analyses of alternative suitable fuel for TEPLATOR based on natural uranium and new fuel geometries. This work builds on previous research on alternative fuel material and geometry for the TEPLATOR. It is mainly concerned with the neutronic development of fuel assemblies, the possibility of manufacturing of developed fuel types and optimization of fuel management and uranium consumption. This article contains predetermined candidates of suitable fuel geometries and new untested types of fuel geometry with some new advantages. Finally, the optimization of the whole reactor core and number of fuel channels in terms of higher safety and higher fuel burnup were made. Presented calculations were performed by Monte Carlo code Seprent.

        Speaker: Mr PELTAN, Tomáš (University of West Bohemia)
      • 14:20
        #07-142 Ruggedized High Purity Germanium Detectors for in-situ Gamma Spectroscopy 20m

        This paper presents the design and performance of turnkey and compact HPGe solutions, developed
        by Mirion Technologies (CANBERRA) for radionuclide identification outdoor and under harsh
        environmental conditions. Surveys can be undertaken under various weather conditions, in
        contaminated areas, underground or immersed under water (sea, rivers, pools), with fast on-site
        deployment and without compromising the performances and reliability experienced with
        laboratory-grade HPGe instruments.
        In situ measurement is a privileged way of detecting radioactive contamination compared to
        analyzing samples in a distant, specialized laboratory. On the other hand High Purity Germanium
        (HPGe) spectrometers provide unmatched nuclide identification capability with the lowest minimum
        detectable activities thanks to its excellent energy resolution and high stopping power. However,
        HPGe instruments are not always of practical use on the field (because of the liquid nitrogen, weight
        and bulkiness).
        These systems relies on advanced technologies such encapsulating the HPGe crystal under ultra-high
        vacuum (UHV), different low vibration electrical cooler adapted to the crystal size, and advanced
        digital spectroscopy processor. Besides, their design includes hardened pressure housing,
        minimization of footprint and weight, sealing and water-tightness allowing easy cleaning from dirt
        or contamination.
        Several examples of such ruggedized HPGe detectors will be described, illustrating the wide new
        range of applications permitted by these technologies: they are respectively designed for borehole
        measurement, high efficiency spectrometry from an aircraft or other vehicles, in situ sea and river
        contamination monitoring, as well as an ultra-compact detector for D&D or high count rate
        The sealed probe is an assembly consisting of a 80 mm in diameter shock proof and watertight
        external housing, including a 20% relative efficiency HPGe crystal mounted in an ultra-high vacuum
        (UHV) cryostat (CANBERRA proprietary technology) along with a compact cryocooler. A detailed
        view of the probe is shown in Figure 1.
        The UHV encapsulation of the HPGe crystal allows partial thermal cycling without harming the
        crystal and degrading the detector performances, thus extending the life of the detector. The HPGe
        crystal cooling relies on a new compact inline cryocooler, operated with active vibration reduction in
        order to keep the excellent intrinsic crystal energy resolution. This technology allows an increased
        portability, smaller footprint and safety of operation without the use of any flammable gas. It is also
        maintenance free and the reliability has also been probed with a large MTBF.
        Higher efficiency versions of water tight detectors have also been designed and manufactured for
        continuous monitoring of contaminants in rivers, lakes or sea water (Figure 2). Such configurations
        can accept germanium crystals of several kg and relative efficiencies in excess of 100%). As the
        sealed probe, the heat generated by the electrical cooler is dissipated passively through the outer
        housing of the detector. A full set of monitoring and readout equipement and software is also
        These systems provide solutions to perform high resolution gamma spectroscopy similar to the
        performance achieved in laboratories with regular High Purity Germanium detectors (HPGe), but
        where no current products are compact or robust enough to be installed.
        They features a FWHM of 2 keV at 1.33 MeV, 1.7 keV at 662 keV and below 1 keV at 122 keV. A
        complex mixture of nuclides can therefore be analysed in order, for instance, to distinguish
        anthropogenic radioactive sources from natural ones. The excellent resolution also allows for a
        significant increase in the minimum detectable activity (MDA), 3 to 5 times higher than
        scintillator-based detectors of similar sizes. For instance, simulations lead to a MDA of Cs-137 in
        water of less than 0.5 Bq / L for 600 seconds acquisition time.

        Speakers: MARIAN, Vlad (Mirion Technologies (Canberra)) , Dr GINSZ, Michael (Mirion Technologies (Canberra))
      • 14:40
        #07-135 Qualification Test System for Radiation Detection Devices - QuTeSt 20m

        Measurement equipment for the detection and identification of radioactive and nuclear (RN) material has a wide application area. The main application aspects are monitoring, search, and identification. Measuring systems are divided into different device classes. There are both portable systems like the handheld systems radiation isotope identifiers (RIID) or personal radiation detectors (PRD) and their spectroscopic versions (SPRD) as well as stationary systems like radiation portal monitors (RPM). A common goal is to gain reliable measurement results. In the past, the only way to assess the performance of a measuring device was to rely on the data given by the manufacturer of the device itself. This situation is unsatisfactory, which is also due to the lack of an international seal of approval. Reliable test results from an independent third party are more than welcome. These tests can be performed against consensus standards in order to have reproducible test results, independent of the testing location and the performing laboratory. This is especially relevant for the procurement of new devices, as well as for the comparison between different systems. Fraunhofer INT has conceived and built a test environment to perform dynamic and static test measurements using neutron and gamma sources. These qualification systems were part of a round robin test during Phase II of the project “Illicit Trafficking Radiation Assessment Program (ITRAP+10)”, initiated by the European Commission to capacitate several laboratories to conduct qualification tests. Several tests were performed in accordance with the ITRAP+10 test procedures, which are based on ANSI and IEC standards, as well as in accordance with the ANSI standards themselves. In this paper both parts, the static one and the dynamic one, are presented, and exemplary results are shown. This includes qualification tests of truck portal monitors with the dynamic test system. Generally, the effects of one test parameter on other test parameters are not considered in the test procedures. For example, the accuracy of the dose rate may depend on the energy range of the radioactive source used. This will then also affect the over range tests. So far, the latter are only intended for one single nuclide. Besides the overview of the test systems the paper will address restrictions, problems and limitations of the possible qualification measurements as well as potential limitations arising from the given test procedures themselves.

        Speaker: RISSE, Monika
      • 15:00
        #07-86 Design and Optimization of 4π Directional Radiation Detector based on Compton Effect 20m

        Detection of gamma ray sources is a challenging task in many applications, where monitoring and mapping of radiological and nuclear materials is required. Obtaining directional information is required for nuclear homeland security (HLS) and safety, mapping for post-accident decontamination of nuclear incident or radiological event, etc. Many of directional radiation detectors are based on directional shielding, made of lead or tungsten collimators, which introduces two main drawbacks. The first is the size and weight, that makes those detectors too heavy and irrelevant for utilization in HLS handheld devices, drone mapping or space applications requiring cosmic gamma ray directional measurements. The second drawback is the small field of view (FOV), which requires multiple detectors to cover all the required FOV or machinery to rotate the limited FOV detector.
        We propose a novel 4π directional detector based on Compton Effect interactions. Instead of using various shielding methods for directional information, we use the Compton scattering event, with two or more interactions within the detector. Based on the interactions locations of a single gamma ray event, we can back-project a probability cone of source locations. Using enough events, a collection of such cones converges into a single point source. One of the main advantages of this method is the ability to see more than just one source, just like a Compton Camera (CC). Whereas collimated detectors are erroneously calculated the mean location of a few sources, missing most of them. Additional advantage of the proposed detector is the 4π directional ability, which is missing in both the shielding collimated detectors and the CC devices. Moreover, shielded detectors undergo decrease in signal to noise (SNR) at high background radiation. Nevertheless, the proposed solution is not affected severely, since non relevant background events are identified as of incorrect direction events and rejected, without entering the calculation algorithms.
        In order for the defector to be accurate, we need small voxels for accurate localization of an interaction location and a large distance between the interactions of a single event for high directional angle accuracy. We suggest an efficient geometrical solution that spreads the radiation detection voxels as far as possible in the detector volume, on the faces of the cubical directional detector. Doing so, we allow the Compton scattered gamma ray photon to travel from one side to the other side of the cube, obtaining maximal distance between the voxels. Further geometrical optimization is required for an optimal voxel size. Each voxel has an impact area and an interaction depth. Maintaining the total detector size, a small voxel area provides high interaction accuracy, but requires more amplification and sampling ADC channels which introduces complexity of the electronics. For high detection efficiency an increased voxel depth is required. On the other hand, this causes an interaction location uncertainty along the depth axis and Compton interaction probability to decrease, lowering the detection efficiency.
        The 4π directional radiation detector provides significant advantages over the traditional directional detectors. It presents compact and lightweight structure without heavy collimators enabling the usage in drones and satellites; Introduces a radiation background rejection abilities; Detects gamma rays in a widest FOV of 4π with the ability to detect multiple radioactive sources, simultaneously, in a single measurement without mistakenly combining them into one. Such abilities enable utilization of such detectors for HLS and mapping of post-accident decontamination.

        Speaker: GHELMAN, Max (Electronic and Control Laboratories NRCN, Ben-Gurion University of the Negev)
      • 15:20
        #07-7 Low-resolution gamma spectroscopy to estimate the concentration of uranium ores and the radioactive disequilibrium with two energy bands 20m

        The estimation of uranium content in ore samples by high resolution gamma-ray spectroscopy requires long measurement times and expensive high-purity germanium (HPGe) detectors. In this work, we present the possibility to measure uranium by low-resolution gamma-ray spectrometry with an easy-to-operate and cost effective NaI(Tl) detector. This method is based on the analysis of two energy bands of the NaI gamma spectrum, which allows estimating a possible “U/Rn” imbalance between the top (238U and its daughters up to 230Th) and the bottom (226Ra, 222Rn and their daughters) of the 238U decay chain, due to differential leaching in roll-front deposits. In case of secular equilibrium, more than 95 % of gamma rays emitted by uranium ores come from 214Pb and 214Bi isotopes, which are in the back-end of 238U chain. Consequently, a disequilibrium in the chain might produce an overestimation of the uranium concentration if U/Rn < 1, or an underestimation if U/Rn > 1 (activity ratios). Therefore, the estimation of the disequilibrium between the beginning and the end of the chain represents a key objective in uranium mining exploration and exploitation. The disequilibrium is estimated with a count ratio between a low-energy band of the NaI(Tl) gamma spectrum around 100 keV that includes a significant signal coming from the top of the chain (92-keV gamma rays of 234Th and 235U and uranium X-rays), and a higher energy band around 609 keV, characteristic of the back end of the chain (gamma ray of 214Bi). Finally, the uranium concentration is derived from this U/Rn disequilibrium (activity ratio) and from the “Rn” activity of the back-end chain. This last is directly obtained by conventional gamma-ray spectroscopy with the 609 keV peak of 214Bi. The MCNP model of the NaI(Tl) detector has been validated using experimental data measured with a representative ore sample, with known uranium concentration and U/Rn disequilibrium. Then a linear relationship has been established by numerical simulation between the ratio of the two energy bands and the disequilibrium. More than 800 simulations have been performed to establish gamma corrections according to the density of the ore, filling height of the sample, and uranium concentration itself (gamma self-absorption in the sample), resulting in a total relative uncertainty smaller than 30 % on the uranium concentration. Experimental results obtained with a series of 38 ore samples provided by ORANO Mining, with acquisition times ranging from a few dozen seconds to a few minutes (depending on uranium concentration, which ranges from 100 ppm to 10000 ppm) show a very good agreement with ICP-MS analyses performed on real samples. For these samples, the gain in measurement time is between a factor 20 and 50 with respect to HPGe measurements.

        Speaker: MARCHAIS, Thomas (DES/IRESNE/DTN/SMTA/LMN)
      • 15:40
        #07-15 Characterization of Inelastic Proton Scattering on Carbon for Active Interrogation Applications 20m

        Passive detection of special nuclear material is challenging because of its inherently low rate of spontaneous emission of penetrating radiation, the relative ease of shielding, and the fluctuating and frequently overwhelming background. Active interrogation, the use of external radiation to increase the emission rate of characteristic radiation from special nuclear material, has long been considered to be a promising method to overcome those challenges. Current active-interrogation systems that incorporate radiography tend to use bremsstrahlung beams, which can deliver high radiation doses. Low-energy ion-driven nuclear reactions that produce multiple monoenergetic photons may be used as an alternative. The inelastic scattering of protons on carbon — specifically, $^{12}$C(p,p′)$^{12}$C — is one such reaction that could produce large yields of highly penetrating 4.4- and 15.1-MeV gamma rays. The gamma-ray energies produced by this reaction would provide higher penetration capability than existing bremsstrahlung systems (photon-energy endpoints of 6 and 9 MeV) while enabling robust material discrimination by means of dual-energy photon radiography. This reaction does not directly produce neutrons below the approximately 19.7-MeV threshold, and the 15.1-MeV gamma-ray line is well matched to the photofission cross-section of $^{235}$U and $^{238}$U. Rate and spectral measurements were made with liquid organic scintillators and NaI(Tl) scintillators, while the radiation dose was measured using thermoluminescent dosimeters. We use Geant4 simulations to simulate the detector responses and background radiation production in the experimental area, which allows the estimation of the yields of the 4.4- and 15.1-MeV gamma rays from near the reaction threshold up to 30 MeV. The yields in all experimental configurations are greater than in a comparable deuteron-driven reaction that produces the same gamma-ray energies and which has been considered for use in active interrogation — $^{11}$B(d,nγ)$^{12}$C. However, a significant increase of the neutron radiation dose accompanies the proton energy increase from 19.5 to 30 MeV. This source could potentially address some of the key technical specifications required for new active-interrogation systems. By taking advantage of the emission of gamma rays into a large solid angle, even multiple cargo scanning streams may be feasible with a single source, thereby increasing the container throughput. With no direct neutron production at proton energies below 19.5 MeV, the $^{12}$C(p,p′)$^{12}$C reaction should further reduce the neutron-shielding requirements and lower the total radiation dose imparted to cargo.

        Speaker: JOVANOVIC, I. (University of Michigan)
    • 16:00 16:20
      Coffee Break 20m
    • 16:20 16:50
      01 Fundamental Physics: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #01-248 Development of the ATLAS Liquid Argon Calorimeter Readout Electronics for the HL-LHC 5m

        To meet new TDAQ buffering requirements and withstand the high expected radiation doses
        at the high-luminosity LHC, the ATLAS Liquid Argon Calorimeter readout electronics will be
        upgraded. The triangular calorimeter signals are amplified and shaped by analogue
        electronics over a dynamic range of 16 bits, with low noise and excellent linearity.
        Developments of low-power preamplifiers and shapers to meet these requirements are
        ongoing in 130nm CMOS technology. In order to digitize the analogue signals on two gains
        after shaping, a radiation-hard, low-power 40 MHz 14-bit ADCs is developed using a
        pipeline+SAR architecture in 65 nm CMOS. Characterization of the prototypes of the
        frontend components show good promise to fulfill all the requirements. The signals will be
        sent at 40 MHz to the off-detector electronics, where FPGAs connected through high-speed
        links will perform energy and time reconstruction through the application of corrections and
        digital filtering. Reduced data are sent with low latency to the first level trigger, while the full
        data are buffered until the reception of trigger accept signals. The data-processing, control
        and timing functions will be realized by dedicated boards connected through ATCA crates.
        Results of tests of prototypes of front-end components will be presented, along with design
        studies on the performance of the off-detector readout system.

        Speaker: ANDEEN, Tim (CERN)
      • 16:25
        #01-215 Experiments with fast neutrons at nELBE 5m

        The neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf features the first photo-neutron source at a superconducting electron accelerator. The electrons are focused onto a liquid lead target to produce bremsstrahlung which in turn produces neutrons via photo-nuclear reactions. The emitted neutron spectrum ranges from about 10 keV up to 15 MeV with a source strength of above $10^{11}$ neutrons per second. The very precise time structure of the accelerator with a bunch width of a few ps enables time-of-flight measurements at very short flight path and experiments to investigate the time response of novel detector concepts.
        The high repetition rate of 100 to 400 kHz in combination with the low instantaneous flux and the absence of any moderating materials provide favorable background conditions.
        The very flexible beam properties at nELBE enable a broad range of nuclear physics experiments. Examples for the versatility of nELBE will be presented: From transmission measurements and inelastic neutron scattering and fission experiments to determine nuclear reaction cross sections with relevance for fundamental nuclear physics, reactor safety calculations, nuclear transmutation or particle therapy to experiments to investigate the response of novel particle detectors e.g. for dark matter search experiments, nuclear instrumentation or the range verification in cancer treatment.

        Speaker: BEYER, Roland (Helmholtz-Zentrum Dresden-Rossendorf, Dresden, Germany)
      • 16:30
        #01-211 Production and Monitoring of Neutron Flax by Activation Detectors 5m

        At the microtron M-10 of Uzhgorod National University there are carried out experimental investigations of photonuclear reactions as well as applied studies of irradiation influence on the properties of new technological materials and electronical components. It is possible to use electron, bremsstrahlung and neutron fluxes.
        The bremsstrahlung is generated by output electron beam at few millimeter thick tungsten or titan converter. Neutrons can be generated on 9Be target in (γ,n) reaction (with a reaction threshold 1.67 MeV). To measure neutron flux we use 59Co as activation detectors. Cobalt detector is activated by neutrons to 60Co ground state and to isomeric level 60mCo with half-life 10.47 minutes and following isomeric transition Eγ = 58.6 keV to the ground state. After β-decay (half-time 5.27 years) the ground state gives well known two lines Eγ = 1173 keV and Eγ = 1332 keV.
        The neutron flax was tested at the microtron for 8.6 MeV output electron beam, which was converted to bremsstrahlung by 2 mm thick tungsten plate. The cobalt detector (boxed powder CoCO3Co(OH)2nH2O, mass 31 g) was irradiated during 10 min by neutrons generated by bremsstrahlung at beryllium block of size Ø10x14 cm and weight 2 kg.
        Measurements of γ-radiation from the activated cobalt were performed with a spectrometer, which includes a NaI(Tl) scintillation detector with an SBS-40 amplifier-to-digital converter board, which is connected to a computer (the spectrum is built using AkWin). The energy resolution of the spectrometer was 7% (FWHM) for line Eγ = 1173 keV. The flux density of thermal neutrons was 2∙107 n/(cm2·s).
        Due to the presence of isomer state 60mCo cobalt as neutron detector is very convenient.

        Speaker: HAYSAK, Ivan (Uzhgorod National University)
      • 16:35
        #01-156 Investigation of Alpha Particle Production in Photonuclear Reactions at Photon of Low Energies 5m

        The production of alpha particles in photonuclear reactions on $^{65}$Cu, $^{115}$In, $^{92}$Mo and $^{207}$Pb targets in the beam of accelerated electrons with low energies was studied. The experiments were run on the Linear Electron Accelerator NLA (The Yerevan Physics Institute). The data on production of alpha clusters in photoproduction reactions were obtained in low energies of photon as a result of measurement of yields of residual products by means of activation technique.The measurement results represent the yields of reactions studied as functions of the mass number of nuclei and the photon energy. The yields of ($\gamma$, $\alpha$) reactions as measured in the present work are given in Fig.1. The data obtained enable one to suggest a change in the mechanism of alpha particle photoproduction with increasing energy of incident photons and target mass.


        Fig. 1.Dependence of yields on mass number of targets. Dark and light circles are experimental points and calculations according to TALYS program, respectively, in case of photon energies of 21 MeV. Dark and light squares are respectively the experimental points and calculations according to TALYS program, in case of photon energy of 40 MeV.

        Speaker: GAGINYAN, Susanna
      • 16:40
        #01-146 Measurement of 93Nb(n,2n)92mNb reaction cross-section at different neutron energies for reactor application 5m

        The natural niobium metal is well suited for atomic reactors due to its high temperature resistant, corrosion resistant, low long term induced radioactivity properties. Niobium alloys such as NbTi and Nb3Sn are used as super magnets in fusion reactors due to its super conducting properties. The 93Nb(n,2n)92mNb reaction is also used as monitor reaction. Therefore accurate data of 93Nb(n,2n)92mNb reaction cross-section needed. The cross-section of 93Nb(n,2n)92mNb reaction was determined at the neutron energies of 16.99±0.53 and 20.00±0.58 MeV by using activation and off-line γ-ray spectrometric technique. The neutron beam of different energy was produced by 7Li(p,n)7Be reaction and the neutron flux was determined using the 27Al(n,α)24Na monitor reaction. The measured cross-sections are compared with the theoretically calculated values based on TALYS-1.9 code and also with the evaluated and literature data.

        Speaker: Prof. SINGH, Nand Lal (M.S.University of Baroda, Vadodra,INDIA)
      • 16:45
        #01-52 Towards a data processing software strategy for Cherenkov Telescope Array 5m

        The Cherenkov Telescope Array (CTA) is an international initiative to build the next generation ground-based observatory for gamma-ray astronomy at very-high energies, ranging from a few tens of GeV to more than a hundred TeV. With more than one hundred of telescopes, of three different sizes, located in the northern and southern hemispheres, CTA will detect high-energy radiation with unprecedented accuracy and a sensitivity that is approximately 10 times better than current and is expected to generate approximately 100 petabytes (PB) of data by 2030, thus CTA is a Big Data project. Data processing is therefore as important as telescope hardware to the scientific performance of CTA. This report describes the preliminary architectural design concept for the CTA data processing software system that will be responsible to manage all the functionality of the CTA array, thereby enabling CTA to reach its scientific goals.

        Speaker: Dr BENEKOS, Nektarios
    • 16:20 17:20
      04 Research Reactors and Particle Accelerators: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #04-276 Characterization of the external pulsed beam at CMAM 5m

        The Centro de Micro-Análisis de Materiales (CMAM) is one of the two research centres with an ion accelerator in Spain. It belongs to the Universidad Autónoma de Madrid (UAM) and the building that hosts the laboratory is at the university campus.

        The equipment of the facility consists on an electrostatic ion accelerator with a maximum terminal voltage of 5 MV and six beam-lines dedicated to various application areas such as the analysis and modification of materials, the study of the nuclear reactions or archaeometry studies. The accelerator, built by High Voltage Engineering Europe (HVEE), is of the tandem type with a Cockroft-Walton acceleration system. It is provided with two sources: a plasma source for gaseous substances and a sputtering source for obtaining any element from H to U from a solid target.

        The accelerator feeds up six beam-lines after the bending magnet at different angles: the standard multi-purpose line which ends in the internal micro-beam line, the time of flight line, the implantation line, the external micro-beam line and the nuclear physics line. For a more detailed description of each line, see Ref. [1]. There are two lines where the beam can be extracted from vacuum to air: the implantation line (IMP) and the external micro-beam line (EuB). Until now, the first has been devoted to the implantation or irradiation in large areas for material modification. The second has been focused onto archaeometry studies. Now, a new project has started to use this two lines for proton-therapy preclinical studies.

        What will be presented at the conference is the characteritzation of the pulsed beam in air obtained in the IMP line. The main aim of this beam-line is to perform homogeneous implantations or irradiations in large areas (up to several cm$^2$). For this purpose, the line has a HVEE electrostatic scanner consisting of four plates for vertical and horizontal beam deflection and the corresponding power supplies. This system is capable of scanning Hydrogen of 10 MeV at a size of 100×100 mm$^2$. Until recently, this system, called RASTER, has been used merely for scanning and irradiating large areas. However, it can open the door to obtaining external pulsed beams for pre-clinical studies on proton-therapy in the FLASH regime.

        1. A. Redondo-Cubero et al., Eur. Phys. J. Plus 136:175 (2021)

        Speaker: VIÑALS, Sílvia (Instituto de Estructura de la Materia, CSIC, Madrid)
      • 16:25
        #04-221 Measurement of fast neutron spectra using activation detectors with local moderation 5m

        During recent years, there has been increasing interest in innovative reactor concepts, which would bring better efficiency and ability to “burn” actinides from spent nuclear fuel in order to minimize the amount of radioactive waste to be disposed in deep geological repositories. One of the most promising ones are molten salt reactors, where the salt in molten form serves not only as coolant, but can also be a fuel carrier, so the fuel would be dissolved in the coolant. In order to determine the fundamental characteristics of salt environment for use in nuclear reactor, the physical and chemical features of salts have to be determined very precisely. There were mainly investigated two groups of salts – fluorides and chlorides. According to obtained results so far, both are suitable for use in molten salt reactor, but chlorides are much easier to work with since they are in general less chemically aggressive. Despite great progress in research, there are still some gaps which need to be addressed. One of them is neutron data for most common chloride salts, especially in energies above 10 MeV. In order to perform such research activities, experimental salt assembly has been created and neutron spectra in several experimental channels has been investigated by using gold activation detectors. Since gold shows relatively low cross-section for n, reaction at higher energies, the local moderation has been considered. The main aim of this study is to identify the most suitable moderator material which will perform local moderation around the detectors and therefore provide better results through increased reaction rate, but without moderating of surrounding fast neutron field. In order to achieve that, Monte Carlo calculations using MCNP code together with various nuclear data libraries have been performed. Following to that, calculated results have been validated by experimental results which provided sufficient data for meaningful discussion.

        Speaker: ŠŤASTNÝ, Ondřej (Brno University of Technology)
      • 16:30
        #04-194 Diamond detectors for neutron flux measurement during transient neutron flux changes 5m

        Precise measurement of neutron flux is crucial for reactor operation, especially during fast reactor transients. Measurement is valuable not only in reactor operation itself, but also for the prediction of reactor behavior via calculation or experiments. Especially fast transients with significant change in neutron flux are interesting and in order to analyse them, an advanced detection system is needed.

        Since 2017, diamond-based detectors are being developed in collaboration with CEA Saclay. Detectors are now in the testing phase at the VR-1 reactor. Diamond detectors are rarely used in reactor core measurements but are promising for this particular application.

        To cover changes in neutron flux during a very short time, a large number of interactions must be recorded, especially if significant changes in neutron flux are measured. Due to this prerequisite, the detection system and data processing must be advanced.

        Several approaches are available to process the signal from diamond detectors. In a nuclear reactor, detector signal contains plenty of different pulses. Some are created by noise or by other electronic influences should be separated. Others that have an origin in particle interactions must by sorted by the source particle, as not only neutrons interact in the detector volume. For example, gamma radiation creates a significant amount of impulses that needs to be identified and separated, so only the count rate from neutron detection can be obtained. In order to solve the separation problem, several approaches are used in current detection systems. For diamond detectors, we tested pulse shape analysis, where pulses are separated by their shape. This approach is very accurate, however, it requires a robust detection system, including a fast AD convertor that allows continuous data recording, if the experiment is performed in reactor core. Alternatively, the subtractive gamma compensation may be a better solution for online measurement signal analysis.

        In order to examine the subtractive gamma compensation, we are testing a set of diamond detectors that allows to collect data that can be analysed by subtractive gamma compensation. Experimental work is performed at the VR- 1 reactor and adjacent laboratories, where various radiation sources are used. Preliminary data should be available at the beginning of the year 2021.

        This detection approach should be used in detection part of a new experimental device that focuses on the rod SCRAM transient. This device is currently under development at the Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague and should be installed at the VR-1 reactor.

        Speaker: NOVAK, Ondrej
      • 16:35
        #04-192 Measuring gamma doses over the mGy-MGy range with a single type of TLD detector 5m

        Thermo-luminescent detectors are currently used to measure gamma doses in the medical and environmental surveillance fields. During the past few years, the CEA Reactor Studies Division tested and validated the use of these detectors for gamma flux characterization and nuclear heating measurements in mixed neutron/gamma fields of low power reactors. Doses were comprised between a few mGy and a few Gy for dose rates up to a few Gy/h. However, in MTR or TRIGA reactors, the gamma flux level is much higher (> 10$^{12}$ n/cm²/s) and the TLD currently in use (TLD400 (CaF$_2$:Mn) and TLD700 ($^7$LiF:Mg,Ti)) and their readout protocol are no longer suitable for the resulting doses. In order to extend the applicable dose range up to ~1MGy (dose rate of a few kGy/h), several options were explored. On one side, some adjustments were made to the readout protocol of TLD400 and TLD700, notably by testing the use of filters to reduce the amount of light received by the reader PMT to avoid saturation. On the other side, a new type of TLD (LiF:Mg,Cu,P) with different Li enrichments (natural – MCPN or enriched in $^7$Li – MCP7) was tested.

        This paper presents the calibration measurements results performed in pure gamma fields, at the irradiation platform of the CEA Cadarache Radioprotection Division, between 250 mGy and 3 Gy for all detector types. In addition to the calibration, these measurements also studied the MCPN and MCP7 response: reproducibility, dose rate dependence, incoming photon energy dependence, high temperature effect when reading TLD, etc. In parallel, a numerical model of the thermoluminescence process was developed. This numerical model was validated using the calibration measurements and used to optimize the heating laws of all TLD types.
        Using these optimized heating laws, TLD detectors will be used to characterize the gamma field and nuclear heating in the TRIGA Mark II reactor at the Jožef Stefan Institute (JSI) in Slovenia. Measurements are expected to take place in the second half of 2021, in the frame of a CEA-JSI collaboration dedicated to the benchmark of the TRIGA reactor neutron and photon 3D transport model.

      • 16:40
        #04-187 Design and optimization of an irradiation housing and shielding of a DT neutron generator 5m

        The International Atomic Energy Agency (IAEA) is implementing a Neutron Science Facility
        (NSF) based on the use of deuterium-deuterium (DD) and deuterium-tritium (DT) neutron
        generators to expand its domains of application and its training capabilities using neutrons
        and associated techniques. Targeted applications include neutron radiography, neutron
        activation analysis and production of radiotracers for training purposes.
        This study presents the optimization of the design and the preliminary characterization of the
        irradiation housing and shielding of a DT generator with a neutron yields of up to 4x108 n/s.
        The irradiation housing increases the effectiveness of the neutron irradiation and contributes
        to the shielding.
        The optimization of housing and shielding has been performed using three types of Monte
        Carlo Codes, namely Monte Carlo N-Particle (MCNP6.2), MCS Monte Carlo neutron/photon
        transport code 0.1.6, and Monte Carlo Code for Advanced Reactor Design and analysis
        Factors that drove the optimization were safety, performance, cost, mass, and commercial
        availability of materials, as the final design is expected to serve as a good practice for IAEA
        Member States implementing a similar facility.
        Results related to the neutron spectrum and flux in the irradiation chamber of the housing as
        well as neutron and dose rates outside of the shielding and in the facility will be presented.
        Radiation dose rates are further decreased by the shielding, surrounding the irradiation
        housing. The paper will include intercomparing the calculations performed with the three
        codes and preliminary experimental measurements assessing the accuracy of the
        Acknowledgements: The authors thank Valentin Blideanu who has provided, under a consultancy
        contract with the IAEA, a valuable contribution to the very first modelling on shielding configurations.

        Speaker: MELLARD, STEVEN (International Atomic Energy Agency)
      • 16:45
        #04-160 Online optical refractive index measurement in research reactor core 5m

        There is a growing interest in fiber optic measurements for applications in radiation environments. They can be used to monitor environmental parameters such as temperature, size, pressure, chemical composition, irradiation doses and dose rates.... Often, the developed systems imply no propagation of the light beam outside the fiber, but for some applications, fiber optic is combined with an optical system that collects or focuses the light beam. The question then arises not only about the impact of the RIA (Radiation Induced Attenuation), which has already been largely studied, but also of the radiation induced change of the refractive index of the used glasses, which is a determining value for the optical function. Studies of refractive index variations under irradiation are already being carried to evaluate their impact on imager performances [1]. Also, for the development of optical sensors for applications in research reactors core, CEA is preparing an irradiation in the BR2 reactor of SCK.CEN in Belgium. We implemented an on-line refractive index measuring device in order to test various glasses that can be good candidate to take place in a hardened optical system. The assembly uses an interferometric measurement system, which is a challenge because of its very small size (diameter of 9 mm), the impossibility of making optical adjustments once installed, and the goal of monitoring these changes at an elevated high temperature. The refractive index variation of silica for example under high fluence – 1019 n/cm² (E>1MeV) and several GGy in gamma- is in the range of some 10-3 [2]. The precision required for the measurement is then very ambitious for such difficult conditions. In addition, the variation in density observed at high neutron fluence on silica causes an elongation of the optical components [3], the targeted online measurement therefore becomes an optical path measurement (product of the length by the refractive index) and no longer just the one of the refractive index. Apart from this online monitoring, we expect to perform post-mortem measurements of irradiated sample dimensions which could help to dissociate the refractive index and length variations. In this work, we will present the planned irradiation and the first results of the index change measurement under the effect of temperature. The index variation between 20°C and 350 ° C is of the same order of magnitude as the variation of the refractive index of silica expected under reactor core neutron fluence. This work is now going on in the framework of a PHD at the French Alternative Energies and Atomic Energy Commission (CEA) in collaboration with the Laboratoire Hubert Curien (LabHC) of the University of Saint-Etienne and also the STIL company [4].

        [1] C. Muller, « Conception optique pour les environnements radiatifs et application à une caméra résistante à des doses élevées (MGy) », PhD thesis, University of Saint-Etienne, France, 2020.
        [2] W. Primak, “Fast neutron induced changes in quartz and vitreous silica,” Phys. Rev. B, vol. 110, no. 6, pp. 1240–1254, 1958.
        [3] L. Remy, G. Cheymol, A. Gusarov, A. Morana, E. Marin, and S. Girard, “Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence,” IEEE Trans. Nucl. Sci., vol. 63, no. 4, pp. 2317–2322, Aug. 2016.

        Speaker: FOURNEAU, Gary (Université Paris-Saclay, CEA, Service d'Études Analytiques et de Réactivité des Surfaces)
      • 16:50
        #04-148 Cross Section Measurement and Comparison of Methods for Background Subtraction in Quasi-monoenergetic Spectrum from p(Li) reaction 5m

        Quasi-monoenergetic neutron spectrum generated in p(Li) reaction has great importance for cross-section measurement. But only half of the neutrons are produced in monoenergetic peak and the rest is in the background without any significant peaks. For measuring nuclear reactions with threshold energy lower than the energy of neutron peak is necessary to subtract this background.

        This contribution shows experimental cross sections for different pure materials and a comparison of two methods of subtraction. The first uses a combination of Monte Carlo simulation and cross-section simulation in Talys code. The second method uses the fact that the angular distribution of generated neutrons rapidly changes with increasing distance from the proton beam axis.

        The experiment was done in the CANAM facility where were irradiated different materials in quasi-monoenergetic neutron field with a peak energy of neutrons 33.5 MeV. The comparison was done for Au and Co samples which have a lot of different reactions for this energy range. The experimental results were obtained via the dosimetry foils activation method including the gamma-ray spectroscopy method.

        Speaker: Mr KRAL, Dusan (Brno University of Technology, Faculty of Electrical Engineering and Communication, Department of Electrical Power Engineering, Brno, Czech Republic)
      • 16:55
        #04-136 Reactor pulse operation for nuclear instrumentation detector testing – preparation of a dedicated experimental campaign at the JSI TRIGA reactor 5m

        The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors (such as fission and ionization chambers and self-powered neutron detectors) relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications. This is thanks to extensive experimental and computational efforts in the past to characterize the experimental conditions in the reactor, in large part in collaboration with the Instrumentation, Sensors and Dosimetry Laboratory of the French Atomic and Alternative Energy Commission (CEA) – Cadarache. The maximal achievable neutron flux in the reactor in steady-state operation mode is approximately $2 \times 10^{13}$ n cm$^{-2}$ s$^{-1}$, in contrast to the MTR-relevant neutron flux range $10^{14}$ – $10^{15}$ n cm$^{-2}$ s$^{-1}$. The JSI TRIGA reactor can also operate in pulse mode, where one of the reactor control rods is ejected from the reactor core, thus introducing a sufficiently large reactivity to achieve prompt supercriticality. This operation mode is made possible by the TRIGA fuel, in the form of a dispersion of uranium and zirconium hydride, which gives rise to a large, prompt and negative temperature coefficient of reactivity. The resulting time dependence of the reactor power is a pulse, the peak power and duration depending on the initial inserted reactivity. In pulse mode, the maximal achievable peak power is approximately 1 GW, corresponding to a peak neutron flux level of the order of $10^{17}$ n cm$^{-2}$ s$^{-1}$, for a duration of a few milliseconds.
        Pulse operation mode at the JSI TRIGA reactor has been investigated mainly in the academic context, for the validation and improvement of the Fuchs-Hansen model describing the reactor behaviour during reactor pulses and for educational activities, however its application for irradiation experiments was of limited scope. Recognizing the potential of reactor pulse mode for nuclear instrumentation detector testing, in particular the possibility of extending the useful neutron flux range up to MTR-relevant levels, in 2019, a bilateral collaboration project between the CEA and JSI was initiated. The aim of the project is the performance of absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. The measurements will be made possible by special CEA-developed miniature fission chambers (fissile deposit mass targeted at 50 ng) and by modern, validated, wide dynamic range data acquisition systems, in particular the CEA-developed MONACO system. In addition to the JSI TRIGA nuclear instrumentation, providing information on the peak power and released energy, activation dosimetry will be employed as a reference for the normalization of the recorded detector signals (neutron flux integral) for reactor pulses with differing characteristics. Finally, as an alternative experimental method enabling the measurement of the reactor power dependence during pulse operation, measurements of the intensity of Cherenkov light are proposed and being investigated.
        This paper presents the 2019-2020 preparatory activities for an exhaustive experimental campaign to be carried out at the JSI TRIGA reactor jointly by CEA and JSI researchers during the first semester in 2021. A series of test measurements using not fully appropriate fission chambers in reactor pulse operation was performed using a Keithley electrometer and the MONACO fission chamber acquisition system. Activation dosimetry measurements were performed for several thermal and fast neutron sensitive nuclear reactions. Photodiodes and silicon photomultipliers (SIPMs) have been tested in steady state and pulse operation modes. The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

        Speaker: RADULOVIĆ, Vladimir (Jožef Stefan Institute)
      • 17:00
        #04-123 Characterization of calorimeter responses under laboratory conditions thanks to an optimized transient thermal test bench 5m

        The nuclear heating named also “absorbed dose rate” is a quantity that it is essential to predict numerically in order to optimize the design of different elements of a nuclear research reactor and of its experimental irradiation devices for instance for safety, thermal and mechanical aspects and then to measure it accurately in order to analyze results associated to in-pile experiments. Heat flow calorimeters are the sensors that permit real time measurement of this key parameter.

        Since 2009, Aix-Marseille University and the CEA have been conducting research dedicated to improving such sensors in terms of their size, range, out-of-pile calibration and in-pile measurement methods. In 2015, a new differential calorimeter (called CALORRE) composed of two compact calorimetric cells with predominantly radial heat transfer was patented and tested successfully in the Polish MARIA reactor. At present, a new joint research project, which also involves the Nuclear Reactor Laboratory of the MIT, is in progress. This project, called CALOR-I, focuses on the characterization of a new reduced-height design of the CALORRE calorimeter and its use for the mapping of the nuclear heating rate inside the MITR water loop. The study will be from laboratory calibration conditions to real conditions inside this loop located in the reactor core (a cylindrical pressure vessel of 38 mm in diameter, forced or natural convection, a fluid flow temperature up to 300 °C and a nuclear heating rate up to 2 W/g (in stainless steel) at a thermal power of 6 MWth).

        The method of calibration of a calorimeter varies according to the type of calorimeter used. For single-cell or differential calorimeters with an electrical heating system, a steady state thermal calibration is usually applied, whereas for calorimeters without an electrical heating system a transient thermal calibration is used.
        Within the framework of the CALOR-I project, the aims of this paper are the optimization of a transient thermal calibration test bench, called BERTRAN, and the study of the response of different calorimeters by using this tool and a 1-D transient thermal model.

        First of all, the test bench and its improvement will be detailed. The bench is composed of two temperature-controlled thermostatic baths filled with a heat-transfer fluid which can reach a temperature up to 250°C without boiling (silicone oil). The hot bath hosts three heating cartridges and the cold bath integrates a heating cartridge and a cooling coil. In addition, inside each bath, there are a cylindrical tube with an internal diameter similar to that of experimental channels and a propeller located at the bottom of the tube in order to create an upward silicone oil flow (adjustable speed). This configuration allows the generation of thermohydraulic conditions close to those existing inside experimental channels of research reactors. Moreover, this bench has a mechanical system used to transfer the calorimeter automatically from the hot cavity (at a temperature from 100°C to 250°C) to the cold cavity (at a temperature from 20°C to 50°C) (or vice-versa). The bench improvements concern the implementation of two new K-type thermocouples to measure the temperature around the calorimeter and an infrared sensor to determine precisely the position of the calorimeter (the inlet or outlet time in each bath); and the change of the acquisition device to improve the sample time.
        Next, the operating protocol allowing the determination of the transient response of the sensors will be described. In fact, after the thermal stabilization of the two baths previously tuned for two different temperature set-points, the calorimeter is inserted inside the cold cavity. When a first stationary state is reached inside the calorimeter, the data acquisition is started. After 10 minutes to ensure stabilization, the calorimeter is transferred automatically and quickly (1-2 seconds) from the cold cavity to the hot cavity (heating phase). After a 10 minutes immersion inside the hot cavity, a second stationary state inside the sensor is achieved, then the reverse transfer of the calorimeter is done (from the hot cavity to the cold cavity, cooling phase). Finally, the calorimeter remains for 10 minutes in the cold cavity in order to reach a final thermal equilibrium.

        Then the experimental results obtained by applying this procedure for a parametrical campaign carried out in July 2020 for a single-cell calorimeter (called KAROLINA) will be presented. The response curves of this calorimeter, for the heating phase and then the cooling phase, will be shown for various experimental conditions, with consideration of repeatability and reproducibility. The influence of the fluid flow conditions (temperature and velocity) and of the operating phase will be given on the response time, the thermal constant and the deduced sensitivity of the calorimeter.
        In addition, a 1-D transient thermal model will be described and applied to predict the temperature inside the calorimeter versus time and to analyze the experimental results.

        Finally, the results obtained with the KAROLINA calorimeter will be compared to forthcoming experimental results of other calorimeters (CALORRE calorimeter and a Gamma Thermometer). If the health situation causes difficulties in carrying out these experiments, the experimental results already obtained will be compared to 3D thermal simulation results obtained with COMSOL Multiphysics.

        Speaker: REBAUD, Jérémy (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 17:05
        #04-63 CAREDAS: a Comprehensive Architecture for a Redondant and Evolutive Data Acquisition System for JHR reactor and measuring bench for CEA Cadarache facilities 5m

        A new material testing reactor Jules Horowitz Reactor (JHR) is under construction at CEA Cadarache. The materials to be irradiated will be placed into experimental devices around the reactor. Process and measurements of experimental devices will be carried out by command control. Programmable Logic Controller (PLC) dedicated on each experimental device will realize this function. Some experiment devices will need to achieve complex real time processing that PLC cannot reach. A data acquisition system (DAS) having these processing performances will be associated to the PLC. This system will realize measurements and processing. The challenge is to design and realize for twenty experiment devices a high availability data acquisition system architecture for 50 years of sustainability. The real time target will achieve 24/7 data acquisition and real time processing. The other components will be in charge of data storage, online and offline data visualizations, experiment setting and processing without modifying software core.
        A proof of concept have been realized by the end of 2019. This scalable architecture could be use as well for JHR experimental devices with high availability as for testbed. This architecture could be run on a standalone station for a measuring bench or deployed on cluster with redundant data acquisition system for high availability. CAREDAS's design is modular and use proven widely used open source solutions of IIoT world (Industrial Internet of Things). All parts are independent from each other and can be replaced with another technology with the same functionalities. This ensures sustainability and control of software sources.
        The heart of CAREDAS is the communication between functions. It is based on MQTT (Message Queuing Telemetry Transport) for message transmission and Google Protocol Buffer for data serialization. Data are stored in InfluxDB time series database. Offline data visualization is achieve with Grafana to compose dashboards that query directly InfluxDB databases. Only three software in C++ language are homemade to ensure sustainability: real time data acquisition, writing data into InfluxDB, tools for experiment setting and supervision. All software run on station or cluster nodes on the Linux operating system (Red Hat distribution).
        Real time data acquisition software is modular and technology independent. The operating system is Linux with real time functionality. For instance, technologies implemented are PXIe bus, CompactRIO/CompactDaq from National Instruments, Ethercat, various measurement units over Ethernet.
        Now, CAREDAS is under development and a first stable version is planning for September of 2021. Meanwhile, a beta version was deployed on five measuring bench on some CEA Cadarache facilities.

        Speakers: Mr LEROUX, Fabrice (CEA Cadarache) , Mr DUCOBU, Lionel (CEA Cadarache) , Mr MILLEVILLE, Frédéric (CEA Cadarache)
      • 17:10
        #04-61 PISTIL, a reactivity modulation device to probe the transfer function of the research nuclear reactor CROCUS 5m

        The present article summarizes the development and testing of a reactivity modulation device developed by the French Atomic Energy Commission (CEA). It was installed in the CROCUS reactor of the Swiss Federal Institute of Technology in Lausanne (EPFL). Experimental tests were performed in the framework of a collaboration between CEA and EPFL.

        The so-called PISTIL reactivity modulator (Periodic reactivity Injection System Transients Induced Locally) aims at measuring the nuclear reactor transfer function in the frequency range of interest, between 1 mHz and 200 Hz. The device is made up of several aluminum tubes (outer diameter 10 mm), two inner ones holding Cadmium foils. The center tube can be rotated by means of a brushless motor, while the outer one remains static. During the rotation, the radiative capture reaction rate of Cadmium can be varied. This gives rise to a controlled reactivity modulation when the device is within or close to a nuclear reactor core, as the in-core neutron population is varied. The height of the ensemble, as well as the relative position between rotary and static parts are adjustable, offering flexibility on reactivity worth and reactivity amplitude.

        Thanks to its compactness, PISTIL can be inserted at the center of the fuel lattice of CROCUS. In the experimental setup, the maximum reactivity worth, as estimated using TRIPOLI-4 Monte Carlo code, was less than 0.15 $\$$ and the maximum amplitude of the reactivity modulation was about 0.02 $\$$.

        Out-of-pile tests inside a dummy mock-up of CROCUS were conducted for the mechanical characterization of PISTIL, while the device was driven periodically to rotate up to 100 Hz. Recorded motion profiles and induced vibrations were analyzed for performance and safety studies. In-core reactivity calibration was then performed for several reference configurations of PISTIL. The measured reactivity worth and modulation amplitude were consistent as compared to TRIPOLI-4 estimations.

        Speaker: Mr JIANG, Yifeng (DES/IRESNE/DER/SPESI/LP2E, CEA Cadarache, F-13108 Saint-Paul-Lez-Durance, France)
      • 17:15
        #04-44 Application of a TlBr detector to on-line monitoring of fuel element failure in the experimental fast reactor Joyo 5m

        Fuel element failure detection is important for FBR plants not only to achieve a high availability for a reactor safety operation and to secure operational reliability, but also to conduct special test attended with a fuel failure. Thus, the experimental fast reactor Joyo uses HP-Ge detector, called OnLine Gamma-ray Monitor (OLGM), to monitor Xe and Kr, which are rare gases as fission products, in its cover gas.
        The OLGM is composed of a charcoal bed, gamma-ray shielding, a HP-Ge detector, valves, and piping that connects the charcoal bed to the cover gas inlet and outlet piping and a fresh argon gas supply line. As cover gas flows through the charcoal bed at room temperature, Xe and krypton isotopes are selectively adsorbed in the charcoal bed at higher efficiencies than those of 41Ar and 23Ne, which obstruct a precise gamma-ray analysis of FP nuclides, because of the difference of the adsorption coefficients between noble gases. Then, the charcoal bed is flushed with a small amount of fresh argon to release the undesirable radioactive Ar and Ne isotopes because of due to the same principle. Joyo OLGM is equipped with two charcoal beds to enable measurement in a short interval of time, and with a HP-Ge detector cooled by liquid nitrogen. In the measured gamma-ray spectrum, it can be seen that the distinct photo-electron peaks of the activated nuclides 125Xe, 127Xe, 129mXe, 131mXe, 133Xe and 135Xe are distinguished with high energy resolution.
        However, the HP-Ge detector requires the time and man power to feed liquid nitrogen regularly. In this study, we examined the application of a thallium bromide (TlBr) detector to OLGM as a room temperature semiconductor detector that has the potential to achieve high gamma-ray energy resolution. The room where the OLGM is installed will be approximately 40 °C at the rated output of the reactor. And Joyo's rated operating days are 60 days. There are two major challenges here. First, the energy resolution needs to be as high as possible to identify the isotopes of Xe. The thallium bromide detector operates at room temperature and slightly higher temperatures, but the energy resolution decreases as the temperature rises. Therefore, it was decided to cool the detector with a Peltier element. The second is the stability of energy resolution and detection efficiency during continuous operation for 60 days. In this study, as the beginning of the study of the applicability of the thallium bromide detector to OLGM, we conducted experiments using a prototype detector on the temperature dependence of energy resolution and the stability of resolution and detection efficiency during long-term operation. It was confirmed that the detector can be stably cooled with a Peltier element in an environment of 50 °C using a constant temperature bath. In addition, the energy resolution was confirmed at a detector temperature of about 10 °C., and the required level for improving the energy resolution of the detector itself was determined. In addition, the time dependence of energy resolution and detection efficiency was confirmed by continuous operation for 60 days, and it was confirmed that TlBr detector could be applied to OLGM.

        Speaker: Dr MAEDA, Shigetaka (Japan Atomic Energy Agency)
    • 16:20 17:20
      07 Nuclear Fuel Cycle, Safeguards and Homeland Security: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #07-252 Passive neutron multiplicity counting with PVT plastic scintillators for plutonium characterization in radioactive waste drums 5m

        In the framework of plutonium characterization in radioactive waste drums by passive neutron coincidence counting, the Nuclear Measurement Laboratory of CEA Cadarache is studying plastic scintillators as a cheaper alternative to $^3$He gas proportional counters. Plastic scintillators offer a three order of magnitude faster time response than $^3$He detectors, larger volumes and a similar neutron detection efficiency. However, the high sensitivity to gamma rays and crosstalk make this technology difficult to use in neutron-gamma mixed field and for large detectors without PSD capabilities. A new patent-pending data processing, based on a time discrimination of triple coincidences, permits to isolate useful fission coincidences from parasitic ones, such as those due to (α,n) reactions and gamma-ray cascades. The performances are studied with an experimental setup designed for 100L to 200L waste drum measurements, and composed of sixteen 10x10x100 cm$^3$ plastic scintillators positioned in circle around the package. MCNPX-PoliMi simulations are also performed and compared to experimental data, first to validate the numerical model, then to optimize data processing, and finally to study main causes of uncertainties. For instance, we investigate the linearity of the method with the plutonium quantity, in presence of increasing background noises and matrix effects. To this purpose, different neutron and gamma calibration sources ($^{252}$Cf for the useful signal, AmBe, $^{60}$Co and $^{137}$Cs for the background) are placed in mockup drums filled with metallic (iron) or organic (wood, polyethylene, PVC…) materials mimicking common technological waste. Without unfolding real and accidental coincidences at this step, the new data processing increases the signal-to-noise ratio (SNR) recorded with bare calibration sources (no drum) by up to 50% compared to classical triple coincidence calculation with the shift register method. On the other hand, MCNPX-PoliMi simulations are in good agreement with experiment, with a relative difference between the measured and calculated number of coincidences lower than 20%, whatever the coincidence order from total counting to triples. Then a linear response is observed by simulation with the equivalent mass of $^{240}$Pu. In addition, this linearity is preserved up to an “alpha ratio” of 100 between (α,n) and fission neutrons. We are currently studying the deconvolution of real and accidental coincidences at high count rate, especially due to high gamma-ray fields, as it is frequently the case with radioactive waste ($^{60}$Co, $^{137}$Cs). We will also discuss the benefit of crosstalk rejection algorithms, as well as the $^{240}$Pu detection limit with different waste matrices, plutonium localization and background noise level. The ultimate goal of this work is to reduce the cost of passive neutron coincidence collars for radioactive waste drums by approximately a factor five, compared to $^3$He-based systems.

        Speaker: BOTTAU, Vincent (CEA)
      • 16:25
        #07-241 The Mini Labyrinth – A Simple Benchmark For Radiation Protection And Shielding Analysis 5m

        Since World War II there has been a significant development of methods and approaches used in the calculation of radiation shielding. This development was directly supported by the needs of industry (military technology, nuclear power plants, food processing, medical applications, accelerators, etc.). Over time, modelling and simulation of relevant effects shifted from an analytical modelling to methods based on the so-called primary principles and their stochastic nature. Even nowadays it is necessary to know the accuracy of available computation codes, used nuclear data and it is desirable to evaluate the influence of the user on the final calculated parameter. One of the most effective ways of gaining user experience and minimizing user effects on the results of calculation is international collaboration comprising the designing and constructing of relevant benchmark experiments, following simulation with tools available at the workplace (engineering tools vs. high-fidelity methods), comparison of work group results and subsequent identification of the source of observed deviations from the experiment. The proposed paper comprises a definition of the simple neutron and gamma shielding benchmark, inspired by the ALARM-CF-AIR-LAB-001 ICSBEP experiment. The experimental setup consists of the PuBe neutron source, several NEUTRONSTOP C5 shielding blocks (polyethylene with 5 % boron), H2O filled PLA tank, plastic source holder and the active and passive detectors. The measured quantities are compared to values calculated by MONACO (as a part of SCALE 6.2.4 system) and MCNP 6 stochastic codes. The influence of different cross section libraries and propagation of cross section uncertainties is studied through the shielding analysis. The achieved results are included and finally, some discussions on further needed development are also included.

        Speaker: Dr VRBAN, Branislav (Slovak Technical University in Bratislava)
      • 16:30
        #07-212 Evaluation of the characteristics of CdZnTe-detectors for the quality of non-destructive assays of nuclear fuel using passive tomography methods 5m

        Non-destructive assays (NDA), carried out to account for nuclear materials, generally include the measurement of the initial enrichment of fresh fuel, the burnup, the colling time and the initial enrichment of the irradiated nuclear fuel, and distribution of nuclear materials and fission products over the volume of fuel assemblies in order to identify the fact of extraction of nuclear materials. In addition, the distribution of fission products makes it possible to detect a damaged fuel rod during refueling of the reactor core. Now the amount of research aimed at developing such systems has increased significantly. One of the universal technologies for such control is passive reconstructive tomography. We have previously presented systems designed to solve the listed tasks in real-time based on spectrometric measurements with CdZnTe detectors. All implementations require a multi-detector system that is installed on the mast of the refueling r machine. It has been shown that the implementation of the SVD decomposition algorithm for passive reconstructive tomography using several energies allows for two to three orders of magnitude to improve the signal-to-noise ratio to improve the quality of measurements in comparison with traditional algorithms. Now there are publications that the use of machine learning methods can provide a similar result. However, the implementation of spectrometric measurements in real-time places high demands on hardware and software.
        We have developed a method for evaluating and optimizing the parameters of the measuring cycle based on the analysis of the amount of information. At the first step, a measurement model is created that formalizes the relationship between physical quantities describing the measured object and the results of observations. As a rule, this is a system of equations in which the instrumental spectrum of one detector is represented by the sum of the products of the sensitivity and the radiation flux density in a given energy range, taking into account the background component. In this case, the measurement channel is understood as the digital radiation spectrum for one detector for a given spatial configuration of radiation sources. It is obvious that the complex dependences of per-channel sensitivities (here it is a complex indicator depending) on the parameters of the detectors and the measured fields limit the possibilities of obtaining them analytically. Therefore, the experimental results on the dependence of per-channel sensitivities were supplemented by simulations in GEANT. Then, for each experiment, we build an information matrix and use it to determine the amount of information in the experiment. The analysis of the previously obtained results showed that this matrix can be used to analyze the quality of NDA by passive tomography methods. In spectrometric measurements, the information matrix explicitly contains the registration efficiency, energy resolution, peak/Compton ratio, and the range of measured energies. In the future, we solve the optimization problem for one of the parameters. To check the obtained calculated dependences, laboratory experiments were carried out. The experimental setup consists of a fuel assembly simulator and a multi-detector measurement system. The fuel assembly simulator consists of aluminum tubes located at the vertices of a regular triangle and thorium sand, the number of tubes may vary. The measuring system is based on μSPEC micro spectrometers (ZRF Ritec SIA) and SDP310, SDP500 CdZnTe-detectors of several models with different values of the studied parameters. A detailed analysis of the obtained dependencies will be presented in the presentation.

        Speaker: Prof. MASLOV, Oleg (Odessa National Polytechnic University)
      • 16:35
        #07-209 New development from Orano Mining in the field of nuclear instrumentation to improve Uranium extraction and recovery. 5m

        Facing more complex mining extraction issues, Orano Mining has implemented a development plan on implementing new instrumentation techniques. These new techniques are mainly in support of the geological control of our mine at SOMAIR in Niger and to respond to the problem of radioactive imbalance in the roll-fronts for our ISR (In Situ Recovery) mines of KATCO in Kazakhstan and in support of our projects in Mongolia and Uzbekistan.
        A. In support of our KATCO site and our ISR mining projects, we can cite the following ongoing projects:
        - The development of a LaBr borehole probe for measurement by gamma spectrometry using an algorithm patented by Orano with CEA to separate Uranium more quickly from its radioactive decay products
        - The same type of detector is also used to characterize cores or ore samples

        B. In support of our SOMAIR mine, we have the following ongoing projects:
        - a new connected stick coupling a very precise differential GPS (10cm uncertainty in altitude) and a gamma measurement, make it possible to ensure the selectivity and traceability of the ore, while avoiding having to go through a gantry measured.
        - a gamma measurement on a belt conveyor after crushing of the ore, allowing to have a precise mine balance and to perform a good grade selectivity between the ore leached in heaps and that of higher grade entering our processing plant
        - an X-ray fluorescence measurement to characterize the uranium, the penalizing agents (carbonate and clay) in order to optimize the quantity of reagent necessary to leach the ore in the front end workshop of the ore processing plant.

        All these developments allow Orano mining to better characterize these ores with an objective of industrial performance aimed at:
        - a reduction in local uncertainties on Uranium reserves of our ISR deposits, for a better positioning of our production cells
        - an improvement in mining selectivity
        - an increase in leaching yields

        Speaker: TOUBON, Hefvé (Orano Mining)
      • 16:40
        #07-202 A comparison of bounding approach with isotopic correction factors and Monte Carlo sampling in burnup credit method 5m

        In nuclear criticality safety analysis, burnup credit is an approach that credits the reduction in reactivity due to fuel burnup. The calculation using burnup credit method consists of two main steps: a burnup calculation, which estimates nuclide concentrations in the spent fuel, and criticality calculation, which uses the nuclide concentrations determined in the first step. However, computational prediction of fuel composition introduces an additional source of bias or uncertainty. There are several methods that deal with these uncertainties and in general, they can be divided into two main categories: bounding methods and best-estimate methods.
        The bounding approach aims to get the most conservative result by adjusting the concentrations of nuclides in a system. This can be done using isotope correction factors (CF), which are determined based on the difference between the experiment and the calculation results (E/C) and are part of validation of each calculation code. The concentration of selected nuclides in the model is then multiplied by these factors so that concentrations of fissile nuclides are increased and concentrations of absorbing nuclides are decreased. This leads to a more reactive system with the conservative estimate of multiplication factor keff.
        The best-estimate methods aim to determine the uncertainty in spent fuel in a more accurate and realistic way. This can be done using Monte Carlo sampling of the different parameters in a model based on their uncertainty and selected distribution. Other options, such as direct-difference method or sensitivity calculations, are also possible.
        In this work, the comparison of bounding approach with correction factors and Monte Carlo sampling was made on the model of spent fuel pool. The isotopic composition of the fuel was calculated for different values of burnup using TRITON code from SCALE code system. In a bounding approach calculation, we have considered 48 nuclides to which the correction factors were applied. The criticality calculation was then made by KENO-VI code from SCALE. In a Monte Carlo sampling calculation, the concentrations of nuclides were multiplied by a factor, which was sampled with normal distribution based on average E/C value and its uncertainty. A total of 5,000 sampled input files were then calculated using KENO-VI code. The average multiplication factor keff of the results and its standard deviation σ were subsequently determined.
        The comparison of results of both methods shows how much conservative the bounding approach is. In this case, the multiplication factor is higher by 0.02935, which is about 5.4× more than the value of the standard deviation of the sampled Monte Carlo tasks. The using of Monte Carlo sampling method can reduce the maximum multiplication factor for models where the coefficient is slightly above the legislative limit based on the current conservative burnup credit methodology. This could lead to an alleviation of possible restrictions resulting from the exceed of the limit. However, the Monte Carlo sampling approach requires a calculation of large number of tasks and is therefore more computationally demanding.

        Speaker: Ms HAROKOVÁ, Pavlína
      • 16:45
        #07-199 Optimising Detecting Geometry for Improved Pulse Shape Discrimination Performance in Plastic Scintillation 5m

        Pulse Shape Discrimination is a technique used extensively for the detection and analysis of neutron emissions in mixed fields of ionising radiation. Pulse Shape Discrimination utilises a single detector and digital processing setup that can discriminate between incident γ-rays and neutrons based on the decay rates in the scintillation light caused by the different radiations interacting in the detector media. Extensive work has been carried out on optimising Pulse Shape Discrimination through improvements in digital signal processing, however few works focus on the physical geometry of the detection medium and how this influences discrimination.

        This work uses Monte Carlo codes (OpenMC, FLUKA, GEANT4) to model the ratio of energy deposition from incident gamma and neutron isotropic radiation fields (1 MeV – 9 MeV range) in a series of EJ299 plastic scintillator geometries. The general shape types simulated included, spherical, cuboid, cylindrical, and conical geometries in a range of volumes. The simulated shapes were compared with regard to their volume and their surface-area-to-volume (SAV) ratio. Where SAV offers a normalised parameter to assess the cross-section of the detector covering the solid angle of the emitted radiations.

        Simulated results indicate that at lower particle energies (< 3 MeV) conical detectors demonstrate a lower gamma/neutron energy deposition ratio, i.e. a higher proportion of neutron energy is deposited compared to the gamma energy deposition, when compared to other detector shapes with similar volumes and SAV ratios. A conical detector with an approx. 40 cm3 volume (height 6 cm, 2.5 cm base radius) has a gamma/neutron energy deposition ratio of 8.31%, compared to 10.21% for a cylindrical detector (height 2 cm, 2.5 cm base radius). Similar differences are found throughout a range of comparable volumes. Though modest, this difference would stimulate greater production of optical photons with a slower decay times which enhance the ability of digital signal processors to discriminate between the signals produced by the γ -rays and neutrons. The difference in energy deposition ratios between shape geometries diminishes as radiation energy increases.

        Further work is being conducted to model additional geometric refinements, different detector media, and simulations of optical photon production occurring as a result of the energy deposition.

        Speaker: Dr PARKER, Andrew (Engineering Department, Lancaster University)
      • 16:50
        #07-181 Timepix3 detector network for nuclear waste monitoring 5m

        Distributed networks of hybrid semiconductor pixelated detectors have proven their usefulness in characterization of mixed radiation fields, e.g. the ATLAS TPX network, which has provided data on particle flux within the ATLAS cavity at LHC over a span of several years by now; or the similar case at the MoEDAL experiment in the LHCb facility. In the framework of the Horizon 2020 funded MICADO project (Measurement and Instrumentation for Cleaning and Decommissioning Operations), we propose a long-term nuclear waste monitoring system based on Timepix3 technology as an alternative/complementary solution to the conventionally used detectors. Particle detectors based on Timepix3 are capable of particle discrimination and provide spectrometric information in common with precise timing information. The particle flux encountered by these detectors, depending on nuclear waste composition, volume, geometry and the distance between the waste packages and detectors, is expected to cover a challenging range and to be composed of mixed radiation field of ionizing and non-ionizing radiation. Moreover, in some cases, part of the detectors will be attached directly to the waste drums, while others are fixed to the storage walls. To maximize the versatility of the setup, the Timepix3 detectors employed are equipped with a range of different sensors, namely Si sensors with thicknesses of 100, 300, and 500 µm, as well as CdTe sensors with a thickness of 1 mm. To enable measurement of neutrons, two quadrants of the detectors are covered by neutron converters. Thermal and epithermal neutrons are detected below a $^6$LiF foil through products of a $^6$Li(n,α)$^3$H reaction and fast neutrons are visible through recoil protons under a polyethylene layer.
        Energy calibration of all detectors was performed in the Institute of Experimental and Applied Physics of Czech Technical University in Prague, using an X-ray tube with several fluorescence foils and gamma rays from an $^{241}$Am source. The neutron detection efficiency calibration was performed at the Czech Metrology Institute by thermal neutrons from $^{238}$PuBe source moderated in a graphite prism and by neutrons from $^{252}$Cf and $^{241}$AmBe sources. A specific software for the Timepix3 network control and data acquisition was developed in-house. This software allows to reproduce and visualize particle fluxes and dose on-line in all detectors separately, as well as to combine quantities from some of them on demand. An early warning system in case of a specified increase of radiation is incorporated as well. First results of long-term laboratory test where various radiation sources were used to simulate radwaste will be presented. However, comprehensive measurements on radwaste sites are currently being prepared.

        Speaker: BISKUP, Bartolomej (Institute of Experimental and Applied Physics, CTU in Prague)
      • 16:55
        #07-178 URANIA-V: An innovative solution for neutron detection in homeland security applications 5m

        Detection of neutrons is becoming of the utmost importance, especially in the studies of radioactive waste and in homeland security applications. The crisis of $\mathrm{He}^3$ availability has required the development of innovative techniques. One solution is to develop light gas detectors for neutron counting to be used as portals for ports and airports. The neutron is converted on the Boron-coated cathode, releasing a charged particle, which passage can be identified by the gas detector. While several technologies have been deployed in the past, the project $\mu$RANIA-V ($\mathrm{\mu}$Rwell Advanced Neutron Identification Apparatus) aims to detect thermal neutron by means of the $\mu$Rwell technology, an innovative gas detector. The goal is to produce tiles to operate as portals in homeland security or for radioactive waste management. The technological transfer towards the industry has started, thus the production can be cost-effective also owing to a construction process relatively easier compared to similar apparatus. By reading directly the signals from the amplification stage, the neutrons can be counted with simplified electronics further reducing the total cost. The first part of the project, more dedicated to the optimization of the neutron detection technique, has been developed within the ATTRACT-EU call. In this presentation, the project will be described, with details on the $\mu$Rwell technology and on the neutron counting, on the test beam performed, and on the future plans.

        Speaker: MEZZADRI, Giulio (INFN - IHEP Fellow)
      • 17:00
        #07-151 Tests of various scintillator detectors in selected monoenergetic neutron beams 5m

        The neutron detection using organic scintillators is a perspective technique for online neutron detection independent of the use of 3He detectors. This is an important issue due to the increas-ing need of neutron detection in homeland security, nuclear energy and also in nuclear nonpro-liferation applications. One of the important parameters in organic scintillation spectroscopy is the energy resolution of a scintillation detector, which for example in homeland security can distinguish between fission neutron source and radioisotope source. The measurement of pro-ton recoil spectra in monoenergetic neutron fields can be used for both validations of neutron response function and also for the determination of resolution of new neutron scintillation ma-terials. These responses were compared with the stilbene response. The stilbene was used for comparison because it is the best scintillation material for fast neutron spectral flux measure-ments in neutron gamma mixed fields. A set of monoenergetic neutron sources in the range of 1.5 – 19 MeV in PTB Braunschweig were used in the testing of detectors. Hamamatsu R329-02 photomultiplier was used to collect the scintillation light. It is connected to the preamplifier that uses different amplification to increase the dynamic range. This also allows us to check on the linearity of the output using calibration sources Co-60 and Cs-137. A verified measurement apparatus NGA-01 was used for the evaluation of the signal of the photomultiplier. This sys-tem can work with a high impulse count per second (>10^5) without any dead time. NGA-01 contains integrated PSD circuits based on the integration method. The measurements show that the best resolution was reached with stilbene, a satisfactory resolution with p-terphenyl, why in the case of the EJ299-33A scintillator the resolution is not very good. On the other hand, due to the low price of plastic scintillator EJ299-33A, the resolution is good enough for large area de-tectors applicable in homeland security applications.

        Speaker: MATEJ, Zdenek (Masaryk university)
      • 17:05
        #07-113 The natIn(n, γ)116mIn reaction cross-sections - An important neutron monitor reaction 5m

        The (n, γ) reaction cross-sections of Indium (In) isotopes are of prime interest for the study of neutron-induced nuclear data. The (n, γ) reaction cross-sections are crucial for upcoming nuclear technologies, like Accelerator Driven Subcritical Systems (ADSs) and Advance Heavy Water Reactors (AHWR) [1, 2]. Indium has extensively been used for the flux measurements in the neutron-induced reaction experiments. There is a need for improvement in the neutron monitor reaction data as the uncertainties in flux directly goes into the measured sample reaction cross-sections. For this purpose, production cross-sections of 116mIn isotope were measured using neutron activation [3] of natIn target following off-line γ-ray spectroscopy by using a pre-calibrated HPGe detector. The 27Al(n, α)24Na reaction is used for neutron flux monitoring purposes. Appropriate energies of proton beams were used from the 14UD BARC-TIFR Pelletron facility, Mumbai, India to generate 10.95 ± 0.67, 13.97 ± 0.0.97, 16.99 ± 0.88, and 20.00 ± 0.94 MeV average energy quasi-monoenergetic neutrons using 7Li(p, n)7Be reaction. Experimentally measured data from the present work have been compared with the existing data libraries such as ENDF/B-VII.1, JENDL-4.0, JEFF-3.2 and CENDL-3.1 [4]. The uncertainty and correlation between the present experimental data have been determined using the ratio technique of covariance analysis [5]. The results were also reproduced and compared with the theoretical nuclear modular codes like TALYS-1.95 [6] and EMPIRE-3.2.3 [7]. The present results show a good agreement with the theoretical as well as with the existing experimental data in different data libraries. The present findings are important for the improvement in the nuclear reaction data, advanced reactor design, dose estimation, and flux measurements.

        Keywords — natIn(n,γ)116mIn reaction, natLi(p,n)7Be reaction for neutrons, cross-section, γ-ray spectroscopy, EMPIRE-3.2.3, TALYS-1.95, Covariance analysis.

        [1]. C.D. Bowman, Annual Review of Nuclear and Particle Science, 48, 505 (1998).
        [2]. C. Rubbia, et al., CERN-AT-95-44-(ET) (1995).
        [3]. K. Shibata, et al., Activation cross-section file for decommissioning of LWRs. JAEA, 47-52, (2016).
        [4]. EXFOR,
        [5]. S. Parashari et al., European Physical Journal A, 55, 51 (2019).
        [6]. A. J. Koning, S. Hilaire, S. Goriely, TALYS user manual, A nuclear reaction program, NRG-1755 ZG PETTEN, The Netherlands (2015).
        [7]. M.Herman et al., "EMPIRE: Nuclear Reaction Model Code System for Data Evaluation", Nucl. Data Sheets, 108 (2007) 2655-2715.

        Speakers: Dr PARASHARI, Siddharth (The Maharaja Sayajirao University of Baroda) , MUKHERJEE, Surjit (The Maharaja Sayajirao University of Baroda)
      • 17:10
        #07-112 A computational methodology for estimating the detector pulse-height spectrum in gamma-ray spectrometry of irradiated nuclear fuel 5m

        Gamma-ray spectrometry using collimated detectors is a well-established method to acquire information about the state of irradiated nuclear fuel. However, the feasibility of examining a particular nuclide of interest is subject to constraints; the peak must be statistically determinable with the desired precision, which is governed by the peak count rate, and the continuum background in the region of interest. Furthermore, the total spectrum count rate in the detector will pose constraints, by potentially causing pile-up or dead time issues, that may paralyze the detector. Therefore, the prediction capability of these spectral parameters is needed.
        Methods were assembled for spectrum prediction with the intended usage in the optimization of gamma emission tomography, and to enable a priori feasibility evaluation of determination of the intensity of a peak in an energy spectrum of gamma-rays emitted from an irradiated nuclear fuel rod. The focus was on finding reliable results regarding total spectrum and peak count rates with faster computation time, as compared to exclusively using Monte Carlo. For this purpose, the method is based on depletion calculations with SERPENT2, a point-source kernel method for the collimator response, and a detector response matrix pre-computed with SERPENT2. The computational methodology uses as input the fuel properties (dimensions, materials and power history and cooling time), and the instrumental setup (collimator and detector dimensions and materials).
        The prediction method was validated using measured data from a high-burnup, short-cooled test fuel rodlet from the Halden reactor. Absolute count rates and ratios of characteristic peaks were compared between predicted and measured spectra. The comparison showed a total count rate underestimation of 6%. The trend is also confirmed for the single peaks, showing a minimum discrepancy of 4-5% for the characteristic gamma lines of 137Cs and 140La, and a maximum discrepancy of 30% for the peaks present in the energy interval between 400-600 keV.

        Speaker: SENIS, Lorenzo (Uppsala university)
      • 17:15
        #07-108 UO2 thermal diffusivity measurement with laser techniques 5m

        The knowledge of the thermal conductivity of nuclear fuel and its evolution as a function of temperature and burn up is a major challenge in the context of the evaluation and understanding of irradiated fuel performances in current reactors. It is also the case for the development and qualification of fuel for future reactors. Indeed, numerical simulations of the fuel behaviour under various conditions require the accurate knowledge of thermal conductivity over a wide range of temperature (from 293 K to melting point temperature) but also at the scale of few tens of micrometers to take into account the microstructural effects on the thermomechanical evolution of the fuel in normal or incidental irradiation conditions. For example, the decrease of thermal conductivity involves an increase of the fuel pellet temperature, which can generate thermal expansions as well as cracking of the fuel due to internal stresses resulting from the radial thermal gradient. Thermochemical equilibria within the pellet, and hence fission product release, may also be affected.

        The laboratory reference method can deduce the thermal conductivity from a thermal diffusivity measurement. One of the most common and non-intrusive method is the laser flash method. This technique is based on a laser-flash excitation and infrared thermography non-contact measurement of the thermal response of a studied sample. The sample can be heated to different temperatures (for instance with lasers) to determine the dependence of thermal conductivity on temperature. The thermal diffusivity measurements obtained with this technique present typically millimeter spatial resolution. Another method is the photoreflectance microscopy, based on the measurement and analysis of the periodic temperature increase induced by the absorption of an intensity modulated laser beam (“pump” beam). By detecting the thermally induced reflection coefficient variations with the help of a secondary continuous laser beam (“probe” beam), the temperature increase at the sample surface is measured as a function of time. Unlike laser flash method, this technique has a micrometric spatial resolution. A laser method with intermediate spatial resolution (few tens of microns) is the Infrared Microscopy technique, based on the detection, by means of infrared thermography, of the surface sample temperature rise distribution induced by the absorption of an intensity-modulated focused laser beam. All of these approaches, using laser, allow reaching thermal diffusivity measurements for temperatures up to at least 3000 K.

        The potential of each of these techniques in the context of the determination of the fuel thermal conductivity will be discussed. Furthermore, the experimental bench being developed at CEA Cadarache, implementing laser techniques to obtain thermal conductivity, from thermal diffusivity measurements, at different scales of interest and for different range of temperature will be presented. The aims of the development of this new experimental set-up in a “test” glove box and the associated methodology are to assess the potential of these techniques by applying them first on non-irradiated UO2 or model materials before its transposition in a hot cell for measurements on irradiated fuels.

        This work is also supported by EDF and Framatome in the frame of Tripartite Institute (CEA/EDF/FRA).

        Speaker: Mr DOUALLE, Thomas (CEA, DES, IRESNE, DEC/SA3E)
    • 16:20 17:20
      08 Decommissioning, Dismantling and Remote Handling: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #08-243 An advanced blind-tube monitoring instrument to improve the characterization of subsurface radioactive plumes 5m

        Nuclear legacy waste storage facilities, such as ponds, tanks and silos, at nuclear waste management facilities or waste disposal sites, often contain radioactive waste under water. Some of these structures in the UK were not lined and, over time, cracks and slits have occurred providing paths for egress of radioactivity to the environment. Consequently, a mixture of radionuclides and water, defined as liquor, can leak into the subsurface dispersing downward and outward in the unsaturated ground from the waste source, contaminating the soil and sediment during its path.

        The scenario on which this research is focussed concerns the potential for nuclear material to have leaked into the ground beneath the Magnox Swarf Storage Silo (MSSS), which was built in the 1960s at Sellafield (UK). In the main, solid waste from the reprocessing of nuclear fuel is stored in the MSSS underwater but it also contains residual material held within the water that is highly mobile and thus has the propensity to migrate into the ground. Sellafield Ltd. is in the process of retrieving the waste and decommissioning the MSSS facility, which requires radiological surveillance of the ground. The current techniques of measurement of the subsurface radioactivity generally involves periodic campaigns to collect soil and groundwater samples from several boreholes, and subsequent laboratory analysis. In addition, ground surveys employing geophysical logging probes have also been deployed in specific boreholes/blind-tubes, in preference to the laboratory analysis. The available radiometric sensors operate at only relatively low dose rates, have low radiation resistance and have low spectroscopy potential. Thus, they are not satisfactory for deployment in a high dose rate environment and constitute only a short-term solution. Therefore, an innovative technological solution is necessary to achieve increased resolution and better understanding of soil and groundwater contamination around the MSSS building. It will also allow to identify and predict spatial migration pathways and guide site remediation programs.

        A design for a resilient radiometric logging probe developed in this research will be presented that fulfils these requirements for down-hole rapid monitoring of long-lived contaminants, such as caesium-137 and strontium-90 (and hence yttrium-90). The device (Figure 1) holds a CeBr3 scintillator detector and a digital data transmitter (MCA), in a resistant and waterproof housing. The probe is then connected to a cable and lowered down in the blind-tube by means of a winch system, and the position in the hole calculated by a depth encoder. The logging cable has usually three functionalities: transmit electrical signals to recording instruments at surface, provide power, and support the lower and raise of the probe. The assessment of caesium-137 and strontium-90 in the surrounding soil formation (but closed to the borehole), is done by detecting the high penetrating gamma-ray photons and Bremsstrahlung photons, emitted by the deceleration of the energetic beta particles in the steel blind tubes, respectively. Moreover, it is anticipated that the extension of a single detector to a blind-tube string network will provide 3D spatial arrangement in order to characterize radioactivity in the ground system, at interim storage sites for high-level radioactive waste. Modelling and laboratory experiments are underway to validate the concept and calibrate the system in a soil-filled phantom arrangement that replicates the in-ground blind-tubes set on site which will be reported on in this work. The ultimate goal is to achieve a validation of the complete system in the existing blind-tube network at the Sellafield site.

        Speaker: ELÍSIO, Soraia (Lancaster university)
      • 16:25
        #08-227 Handheld Spectrometers: From CZT To NaI, an Embedded Template Analysis for Isotope Identification 5m

        This paper presents a solution for embedding an isotope identification algorithm on portable gamma spectrometry systems based on template approach. There exist many algorithms able to identify isotopes in a spectrum. These algorithms can be classified into two large families. The first family is based on peak search or Region Of Interest (ROI) in the spectrum, the second on a global template analysis or deconvolution [1]. All of these algorithms can be evaluated using off-line solutions as described in a large amount of papers [2-4]. In the case of algorithm embedded on handheld spectrometers, the first family is commonly used due to their low requirements in term of computing resources and memory. However, the new generation of microcontrollers or small board computers allows template analysis algorithm that need more computing resources to be embedded.
        In this paper, we present a template analysis, based on MLEM algorithm [5], and its implementation on two different targets, a STM32 microprocessor and a Raspberry PI. Time performances show that this kind of approaches can be used in handheld system such as spectrometers and portable isotope identifier. Experimental results are based on two use cases, the first applies the requirement of ANSI42.34 [6], the algorithm can be performed at the end of the measurement, the second needs to provide evolution of the isotope identification as the progress measure.
        Some promising results have already shown that this kind of algorithms can be uses for handheld spectrometers. Both implementation will be described and results of their implementation will be presented for a set of different detectors from CZT to NaI. The study will be focused on CZT, LaBr3/CeBr3 and NaI detectors. For each detectors, a spectra database is performed based on isotopes described in ANSI 42.34 norm. We have developped pocket spectrometer providing spectral information on an ultra-portable and wireless CZT prototype capable of providing identification information on the radioelement being detected. The system includes a battery, a screen allowing the visualization and an on-board PC (raspberry PI) that perform online identification of the detected. Base on this system and others developments for others detectors, the impact of energy resolution and algorithm implementation will be discussed in the full paper.

        [1] Monterial, M., Nelson, K. E., Labov, S. E., & Sangiorgio, S. Benchmarking Algorithm for Radio Nuclide Identification (BARNI) Literature Review. United States. doi:10.2172/1544518.
        [2] M. H. Zhu, L. G. Liu, Z. You and A. A. Xu, "Least-Squares Fitting of Gamma-Ray Spectra with B-Spline Basis Functions," in Image and Signal Processing, 2008. CISP '08. Congress on DOI - 10.1109/CISP.2008.744, 2008.
        [3] L. J. Meng and D. . Ramsden, "Improved quantitative gamma-ray spectroscopy using a standard 3 inch NaI detector," , 2000. [Online]. Available: [Accessed 6 12 2018].
        [4] M. J. Neuer, "Spectral identification of a 90Sr source in the presence of masking nuclides using Maximum-Likelihood deconvolution," Nuclear Instruments & Methods in Physics Research Section A-accelerators Spectrometers Detectors and Associated Equipment, vol. 728, no. , pp. 73-80, 2013.
        [5] L. J. Meng and D. Ramsden, "An inter-comparison of three spectral-deconvolution algorithms for gamma-ray spectroscopy," in IEEE Transactions on Nuclear Science, vol. 47, no. 4, pp. 1329-1336, Aug. 2000, doi: 10.1109/23.872973.
        [6] "American National Standard Performance Criteria for Handheld Instruments for the Detection and Identification of Radionuclides," in ANSI N42.34-2015 (Revision of ANSI N42.34-2006) , vol

        Speaker: CORRE, Gwenolé (CEA LIST)
      • 16:30
        #08-182 Follow-up on characterizing low-activity waste containers: a case study for Compton Suppression Systems under challenging signal-to-noise ratio 5m

        Gamma-ray spectrometry is a reference technique for measurements carried out during the decommissioning process. Compton continuum is a major contribution to the signal-to-noise ratio when considering low-energy peaks measurement. Lead shielding is oftenly used to reduce background contribution but its weight is a major drawback.
        This work investigates a compact Compton Suppression System (CSS) based on a High-Purity Germanium (HPGe) diode as a primary spectrometer. Among a wide range of guard detectors, including BGO, NaI(Tl), and LaBr3, a plastic scintillator is selected on account of its lightweightness.
        For this purpose, CEA investigates a compact Compton Suppressor System “CSS” consisting of an HPGe “High Purity Germanium” spectrometer coupled with a plastic “EJ 200” scintillator. Hence, when an event temporally coincides in both detectors, corresponding to at least on Compton scattering event, it is rejected by an anti-coincidence filter. The signal-to-noise ratio improvement of total-absorption peaks over Compton continuum hence betters the Minimum Detectable Activity “MDA”.
        The present article is focused on the experimental study of said Compton Suppression System. It consists in using HPGe detectors of various relative efficiencies (10 %, 20 % and 40 % at 1.3 MeV), set inside a plastic annulus scintillator from Scionix Holland. The acquisitions were carried out using “Hexagon” and “DSPEC 502A” digital MCAs, from CAEN SpA and AMETEK Ortec respectively. The sources and peaks of interests shall be 59 keV (for 241Am, 238U’s 63 keV and 133Xe’s 81 keV), 122 keV (for the 100 – 200 keV region of Plutonium), 137Cs’s 662 keV, 60Co’s 1.17+1.33 MeV.
        The figures of merit of interest quantized with and without anti-Compton filtering are the overall Compton continuum “C(E)” reduction at different energies “E”, and the P(E)/C(E) ratio between total absorption peak “P” and the Compton continuum; the latter corresponding to the overall improvement of the measurement.

        Speaker: Dr MICHEL, Maugan (Institut LIST, CEA, Université Paris-Saclay)
      • 16:35
        #08-147 Radiation Stability of Gadolinium Zirconate: A Nuclear Waste Immobilization Matrix 5m

        The ease of formation of defect fluorite structure of Gd2Zr2O7 pyrochlore oxide addresses the remarkable radiation tolerance for the nuclear waste immobilization. The rare earth, particularly Gd, zirconates are effective neutron absorbers, thus advantageous for the disposal of plutonium. Several studies have been conducted on the Gd2Zr2O7 ceramic, the exact nature of grain size dependent structural phase transformation upon irradiation are not well captured. In this report, the grain size dependent radiation effects of microcrystalline Gd2Zr2O7 ceramic upon irradiation of 100 MeV I7+ ion at various fluences are examined and discussed. The grazing incidence X-ray diffraction, Field emission scanning electron microscopy, Raman spectroscopy, and high-resolution transmission electron microscopy are employed to investigate the microcrystalline Gd2Zr2O7 ceramic. FE-SEM results indicate that grain size and grain boundaries are clear in pristine Gd2Zr2O7. GIXRD results demonstrate that the amorphization fraction appears to be grain size and irradiation ion dose dependent. The Raman spectroscopy analysis exhibits that there is a distortion in atomic order/local disorder and increases with enhanced fluence. HRTEM results confirm the partial amorphization upon ion irradiation. We conclude that grain size plays a crucial role in the irradiation resistance of microcrystalline Gd2Zr2O7.

        Speaker: Prof. SINGH, NANDLAL (M.S.University of Baroda, Vadodara,INDIA)
      • 16:40
        #08-125 Modelling the response of an in-situ CdTe detector to radionuclides in groundwater. 5m

        This research examines the potential deployment of a cadmium telluride strontium-90 detector in groundwater boreholes at nuclear decommissioning sites. This is represents a novel approach to monitoring strontium-90 contamination at decommissioning sites such as Sellafield, and has the potential to reduce lifetime monitoring costs while providing information on a significantly reduced timescale. A Geant4 simulation was used to model the deployment of the detector in a contaminated groundwater borehole. It was found that the detector was sensitive to strontium-90, yttrium-90, caesium-137 and potassium-40 decay, some of the significant beta emitters found at Sellafield. However, the device showed no sensitivity to carbon-14 decay, due to the inability of the weak beta emission to penetrate both the groundwater and the detector shielding. The limit of detector for such a sensor when looking at solely strontium-90 decay would be 323 BqL$^{-1}$ after a 1 hour measurement and 66 BqL$^{-1}$ after a 24 hour measurement. Existing techniques are capable of examining strontium-90 decay below the World Health Organisations safe drinking water limit of 10 BqL$^{-1}$. A GaAs sensor with twice the surface area, but 0.3 % of the thickness was modelled for comparison. Using this sensor, sensitivity was increased, such that the limit of detection for strontium-90 was 91 BqL$^{-1}$ after 1 hour and 18 BqL$^{-1}$ after 24 hours. However, this sensor sacrifices the potential to identify the present radionuclides by their end-point energy.

        Speaker: TURKINGTON, Graeme (University of Glasgow)
      • 16:45
        #08-89 Efficient System for Small Waste Containers Activity Estimation 5m

        A new system design and a method to estimate and classify small waste containers is presented. A homogenous radioactive waste, placed in these small containers. Each container activity needs to be assessed in order to classify its activity level, e.g. Very Low-Level Waste etc. We propose a measurement system in the shape of a rectangular box made of several plates of scintillator detectors. The waste-container is placed inside the scintillators box to measure the emitted Gamma-photons. This detection configuration is the most efficient due to the full 4π coverage of the radioactive material with detectors. In this configuration the photons count rate is not affected by the 1/R2, since all photons emitted from the radioactive material and not absorbed by the container matrix reach the detectors. These photons are counted with high efficiency if the scintillator is thick enough relative to the emitted photons energies. The total count rate is predominantly affected by the matrix attenuation due to absorption and scattering.
        To estimate the activity accurately, one needs to know the matrix attenuation and the spatial distribution of the radioactive material or 'hotspots' position. For known homogenous materials the attenuation can be estimated by weighting the container and calculating its density. Otherwise, we validate the homogeneity by measuring the attenuation along the three container axes using a collimated external source on one side and collimated detector on the other side.
        When all the activity is emitted from a single concentrated source, in the center of the container, then the count rate of the detector is minimal due to the maximal attenuation. Any other source with the same activity in a different position or spatial distribution (such as several sources with the same total activity), will give a higher count rate. Thus, for a measured count rate and attenuation, if we assume that all photons were emitted from the center of the container, the estimate of the total container activity is the highest possible activity. In other words, the upper bound of the activity in the container is calculated by assuming a single source in the center of the container. Hence, a better estimation of the source spatial distribution will lower the upper bound of the activity estimation and will make it more accurate.
        For a better spatial distribution estimation, we 'break' the box detector into many smaller detectors (pixels). For example, when each side is measured independently, it gives six detectors readings. If each face is divided into several pixels, then the fuller data will provide a better spatial estimation, and hence more accurate activity estimation. However, this leads to a more complex and expensive system. The purpose of this work is to find, using simulations, the relation between the number of pixels in the detection box, to the total activity estimation accuracy, for different attenuations (the medium type, density and energy emitted). The results will be used in the system prototype design in order to meet the requirement for the allowed activity estimation error.

        Speaker: VAX‬‏, ‪Eran (Electronic and Control Laboratories NRCN)
      • 16:50
        #08-69 A simple determination of 41Ca in decommissioning of nuclear power plant: a strategy to manage a large number of samples 5m

        The activity of several radionuclides has to be estimated during the decommissioning of nuclear facilities for assessing accurately its total inventory and accounting for its variation with time. An accurate characterization of materials is a key point for categorising the produced waste. In a nuclear reactor facility, many radionuclides are produced by neutron activation: among these 40Ca is transformed in 41Ca by a (n,γ) reaction. 41Ca has a half-life of (1.03E+05 y). It is characterized by a high mobility in the environment and decays to the ground state of 41K by pure electron capture, emitting X-rays and Auger electrons of very low energy (between 0.3 and 3.6 keV). For its identification and possibly quantification, mass spectrometry techniques such as accelerator spectrometry (AMS) and resonance ionization mass spectrometry (RIMS) have also been used. Also X-ray spectrometry and liquid scintillation counting are possible, but, due to the poor energy resolution of beta spectra and the low energy of X-rays and Auger electrons of 41Ca, all other radionuclides have to be removed, after the decomposition of the samples ,with time-consuming and potentially expensive chemical purification.
        The challenge for radiochemical analyses generally required for the decommissioning processes, is to deal with several different type of materials some of which are difficult to decompose, or conventional analytical methods cannot be applied for various reasons. The aim of this work is to propose a simple and reliable procedure to determine 41Ca by ICP-MS equipped with a reaction or collision cell, in samples produced during a nuclear power plant decommissioning processes. Evidences will be provided that the method is general and robust enough to be applied to different contexts too.

        Speaker: Dr COZZELLA , Maria Letizia (ENEA )
      • 16:55
        #08-67 Non-destructive Gamma Spectrometry characterization of Radioactive Waste Packages: integration of different techniques for a smart and autonomous measurement station 5m

        The EU Project MICADO, Measurement and Instrumentation for Cleaning And Decommissioning Operations, has the purpose to propose a cost-effective solution for non-destructing characterization of nuclear waste, implementing a digitization process that could become a referenced standard, facilitating and harmonizing the methodology used for the in-field waste management, dismantling and decommissioning.

        Among research activities within MICADO, an integrated station for gamma measurements have been conceived. The main idea is to combine different gamma detection technologies supporting each other to a comprehensive non-destructive characterization. e.g. the cooperation of techniques to reduce the overall analysis time if results acquired from one technique can be used to direct the others, by skipping up some step in the traditional characterization procedure, or by running some measurements simultaneously when it is technically feasible.

        The gamma measurement station combines, for the first time, multiple non-destructive techniques based on gamma-ray detection systems in a single device. The techniques that are taken into consideration are:

        • dosimetry measurements in Open Geometry detector;
        • gamma imaging in Open Geometry detector;
        • Segmented Gamma
        • Scanning, SGS;
        • Angular Scanning, AS;
        • Emission and Transmission tomography, ETC and TCT.

        The characterization station will be able to provide:

        • multiple Radioactive Waste Package (RWP) sizes accommodation and characterization;
        • an estimate of the matrix material and density based on the evaluation of linear attenuation coefficients;
        • radioactive “hot-spot” localization;
        • radioisotopes identification based on high resolution gamma spectrometry.

        All technologies chosen are state-of-the-art detection techniques available on the market or novel techniques near to the industrial level. They only need to be updated or integrated to get the expected results. One innovation of MICADO comes from the fact that these measurements, provided by three separate and independent systems, will be integrated to work together.

        Within the MICADO project the three technologies that will perform the previously described measurements are:

        • the ENEA SEA Radioactive Waste Gamma Analyzer (SRWGA) for high resolution gamma spectrometry in different geometries, from open geometry up to tomography, to complete the RWP characterization;
        • the CEA Nanopix: gamma camera detector for hot spot identification and localization;
        • the CAEN RadHAND for fixed dosimetry and initial spectrometry measurements in open geometry.

        The integration of these technologies will make the RadHAND and the Nanopix gamma camera the starting point of the inspection phase for the assay, being those instruments able to provide input information for the following step, i.e. full radionuclides identification and quantification by means of the “gamma-scanner” system SWRGA. Another significant advantage put in place by the gamma station regards the user contribution to the whole characterization. Currently, measurements are typically run by expert user.
        Have given,

        1. dosimetry information about the RWP (magnitude and values distribution of the contact dose-rate at the side of the waste package);
        2. information about radionuclides identified by handheld detectors;
        3. information from the RWP “history”, when available;
        4. RWP weight compared with the expected weight of a “standard” RWP;
        5. some preliminary measurements of the RWP to collect information about spatial distribution of material and density of the waste matrix;
        6. some preliminary measurements of the RWP to collect information about spatial distribution of radioactivity,

        the expert user decides about the best gamma characterization technique to apply to the complete the RWP non-destructive assay, based on his experience and judgement.

        The aim of the MICADO gamma station here is to overcome this approach via an “expert system” able to autonomously controlling the gamma station. Decision parameters will be decided and set-up so that experimental values collected from real measurement could produce decision about most suitable gamma characterization technique to apply.

        The “expert user decision and control option” will be still allowed for taking into account the case of particularly complex items or other special cases if the programmed expert system fails.

        Here, such a conception is presented and illustrated, and the technical integration of all systems described. Given the gamma station so conceived, RWP characterization examples will be finally provided and discussed.

        Speaker: Dr LEPORE, Luigi (ENEA)
      • 17:00
        #08-57 Characterization via the RadHand device integrated into the REACh system for a low-cost in-situ waste characterization of nuclear waste 5m

        A key aspect involving both operational activities as well as radioprotection in NPP (Nuclear Power Plant) is the management of nuclear waste. Providing efficient and reliable real-time radionuclide concentrations contributes invaluable information while processing nuclear waste as it can in turn reduce costs with packaging, transportation, and disposition for NPPs.
        Typical NPPs procedures can involve characterizing nuclear waste by using an HPGe detector to determine the radionuclide concentrations and the individual isotope abundances. In most instances, this requires NPPs to be equipped with a facility on site (e.g., laboratory) which can perform these measurements. These analyses require a radiological expert to analyse the results. Additionally, this process often taken an immense amount of time and effort to obtain the results for well-defined gamma-emitting radionuclides which are easily detectable.
        Due to the integration of software and hardware technologies in the smart era, we are able to provide a faster and more user-friendly waste package characterization system than what current methods are offering; this saves time, improve flexibility, movable, and is a more cost effective solution.
        Current characterization practices involve obtaining waste stream specific distributions which are then applied to all waste packages of the same waste stream type (e.g., DAW (Dry Active Waste) waste).
        It is important to remember that this type of analysis is typically performed by HPGe detector. The main drawbacks for a system such as this are that it typically involves long measurement times, is a very expensive solution, and requires post process analysis to be performed to determine the radionuclide information. Equipping a laboratory with this type of instrument requires also a supporting structure for the shielding weight of the HPGe, a liquid nitrogen filling for the operations, and a budget on the order of 100 kEuro for buying it.
        The REACh system represents a step change in the way low-level radioactive waste is characterized and classified. The system is intended for all waste packages for open geometry measurements by providing a low-cost detection technology with an easily transportable device.
        The Reach system directly measures gamma dose rates and gamma emitting activity by radionuclide for packaged radioactive material to perform characterization and classification. To obtain an effective and complete characterization and management of the waste packages (WP) the REACh system is equipped with the RadHand detection system used for the gamma characterization, a software database for data storage and handling and it has the capability to use the RFID technology. The system can track in every moment the uniquely identify WP using the attached RFID tag and restore in every moment the WP characterization history using the developed database.
        The RadHand system was specifically re-designed to perform static measurement using a tripod and to work in an open geometry in front of the waste package. The system allows users to work with LLW (Low Level Waste) activities up to which can read up to 1 mSv/h at contact. Additionally, the system is equipped with external shielding to reduce the intensity at the detector surface should it be required.
        This presentation will focus on the laboratory test made with the RadHand system to evaluate its performances in the two fixed configurations. The feasibility to substitute the old typical procedure based on HPGe characterization with this new system will be analysed.

        Speaker: PEPPEROSA, Andrea (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy)
      • 17:05
        #08-56 Spid-X: A Gamma camera with spectro-identification and dosimetry embedded functions 5m

        A gamma camera, allowing the localization of radioactive sources, is a very useful device in various fields of the nuclear industry: monitoring, D&D or waste management are some examples. However, the information of the source position is not enough and should come with the sources identification and dosimetry information, which are provided by additional devices, such as radiameters. In this paper, we present the various functions and performances of the Spid-X Gamma camera, which overcomes this problem by performing in real time the 4 following functions: automatic identification of radioactive sources, proportion measurement of each detected sources, imaging of sources and dosimetry at the camera level. This is possible thanks to the embedded Caliste-O technology, which is a CdTe-based pixelated imaging spectrometer using a single plane 1.4 x 1.4 cm² crystal and 8 low noise ASICs as a readout enabling high energy resolution. Thus, Caliste-O allows position sensitivity and fine spectroscopic performances that are a perfect fit to develop the Spid-X camera. The spectro-identification is performed thanks to advanced Convolutional Neural Network (CNN) trained on synthetic data, which can determine from a measured spectrum made of a few photons which radioisotopes created the signal, even in the case of multi-sources detection. In this last case, it also provides the information of the relative proportions of the different sources. This identification capability is coupled to imaging algorithms, allowing precise Coded Mask and Compton imaging on a wide energy range from 2 keV to 2 MeV and localization of each point and extended detected sources. Finally, dosimetry measurement at the device position is automatically calculated in order to provide to the used the answers to the three following questions: what do we detect, where is it, and how much there is, enabling efficient and safe use in the nuclear industry.

        Speaker: Dr SERRANO, Paul (3D PLUS)
      • 17:10
        #08-46 Non-Destructive characterization of activate materials from fusion reactors 5m

        Nuclear fusion reactors will produce activated materials containing both activation products and tritium; the former is the main source of the radioactive inventory of the generated waste, because of the reactor components that are directly exposed to neutrons. Activated structures have to be replaced during the operation of future fusion power plants, such as ITER. Decommissioning of the plants generates activated metals and concrete, that require treatment and conditioning which, in turn, will generate secondary waste. Strategies are already considered for reducing the amounts of activated waste by adopting recycling, interim storage, and clearance.

        In terms of waste management processes, each single step implies detailed quantitative and qualitative knowledge of radionuclides occurring in materials and components involved, making pivotal the implementation of appropriate measurement techniques, protocols, and modeling for characterization.

        Radionuclides with significant impact in the mid- and long-term management of activation waste are, among the others, Fe-55, Mn-54, Ni-59, Co-60, Zr-93 and Nb-94. With the exception of the gamma emitters such as Mn-54, Co-60 and Nb-94, which are relatively easy to detect and quantify by means of gamma spectrometry, these nuclides are hard to measure, given their little-to-none emission of gamma radiation. In particular, Fe-55 and Ni-59 decay by electron capture and are traditionally detected by destructive characterization techniques, either mass spectroscopy or Liquid Scintillation Counting, by exploiting the emitted Auger electrons. The same approach is usually adopted for the beta emitter Zr-93.
        Given the potential large amount of fusion waste to be generated in the reactor lifetime, non-destructive characterization techniques are preferred since they may require less time and efforts with respect to the destructive approach and are safer from the radiation protection point of view. Techniques useful in the fusion context can be borrowed from those waste streams generated by fission, accelerators or by the industrial and the medical fields.

        Here we present the case study of a decommissioned fusion research reactor which provides useful insights for quantifying advantages and drawbacks of the non-destructive characterization techniques in the context of the management of activated materials. Non-destructive techniques based on semiconductor crystals, such as CdTe, for X-ray spectrometry along with gamma spectrometry will be illustrated. Considerations for application to the case of the future ITER reactor will also be provided.

        Speaker: MARZO, Giuseppe A. (ENEA Fusion and Technology for Nuclear Safety and Security Department, Casaccia (Roma), ITALY)
      • 17:15
        #08-11 Design of MICADO passive and active neutron measurement system for radioactive waste drums 5m

        In the frame of the MICADO H2020 project, a passive and active neutron measurement system is being developed to estimate the nuclear material mass inside legacy waste drums of low and intermediate radioactivity levels. Monte-Carlo simulations have been performed to design a modular and transportable neutron system, with the main objective to reach a good tradeoff between the performances in passive mode, i.e. neutron coincidence counting, and in active interrogation mode with the Differential Die-away Technique. Different designs are compared, which mainly differ in their moderation materials, graphite and polyethylene. This parametric study allowed us to define a prototype taking into account practical constraints in view of its final implementation in a wide range of in-situ locations and nuclear facilities. The total neutron detection efficiency of the prototype is 6.75%, as calculated for an empty drum without waste matrix. The detection limit in terms of nuclear material equivalent mass have also been estimated by calculations based on assumptions for a homogeneous distribution of nuclear materials inside the drum, filled with four types of matrices covering the range of nuclear waste drums defined in the frame of the project. The most penalizing matrix is made of polyethylene with an apparent density of 0.7, which leads to a mass detection limit of 519 mg of 240Pu in passive mode, and 564 mg of 235U or 349 mg of 239Pu in active mode. Measurement time is 30 min for both passive and active modes. The most favorable matrix is made of stainless steel in passive mode and of polyethylene in active mode, with an apparent density of 0.7 and 0.1, respectively. The calculated mass detection limits are respectively 68 mg of 240Pu, 62 mg of 235U and 39 mg of 239Pu. Next steps will be a complete investigation of matrix effects based on intensive Monte Carlo calculations and an experimental design to build appropriate corrections. Experiments will also be conducted at CEA Cadarache Nuclear Measurement Laboratory with the neutron system prototype, and mock-up drums filled with different matrices.

        Speaker: Dr DUCASSE, Quentin (CEA, DES, IRESNE, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance, France)
    • 16:50 17:00
      02 Space Sciences and Technology: Parallel Intensive Oral Sessions

      Oral Session

      • 16:50
        #02-115 Gamma-ray Detection and Localization at high angular resolution 5m

        The recent association of gamma-ray bursts with neutron star mergers highlighted the need for sensitive gamma-ray detectors with high localization accuracy. Identification of the host galaxy of the gravitational wave event GW170817 and the gamma-ray burst GRB170817A was only achieved 11 hours after initial detection, due to difficulties in surveying the uncertainty region of the gravitational wave event. Gamma-ray bursts are transient events with a rapidly diminishing afterglow at longer wavelengths. A delay in the precise localization means crucial data were lost until observations at longer wavelengths could begin. This problem can only be mitigated using detector systems with large viewing angles, preferably the entire sky, and far improved angular resolution of a few degrees, which is required for prompt follow-up.
        We present a novel gamma-ray detector concept aimed at improved angular localization with respect to the current state of the art. The presented detector system relies on a non-uniform pattern of small scintillators coupled to silicon photomultipliers. The pattern utilizes the mutual occultation between detectors to reconstruct the gamma-ray burst’s direction in the sky with good angular accuracy. Our simulations show that the achievable localization accuracy for such a configuration is considerably better than those obtained by larger scintillator assemblies while maintaining a field-of-view of the entire sky. We show that even when the total effective area is decreased, our detector system still shows better angular sensitivity compared to designs based on the current state of the art. We present laboratory experiments on various configurations of 90 small detectors. We show the effects of changing the configuration and of changing the scintillator type. The systems are also experimentally compared with a larger detector system built with a traditional approach. Both simulations and experiments clearly show our novel concept can achieve a considerable improvement in angular sensitivity, without compromising sensitivity or field-of-view. The proposed concept can be easily scaled to fit into small satellites, as well as larger missions.

        Speaker: Mr RAHIN, Roi (Technion - Israel Institute of Technology)
      • 16:55
        #02-73 A test-bench for characterizing SDD-scintillator coupled detectors within the context of HERMES TP/SP nanosatellite mission. 5m

        The main objective of the HERMES-Technologic and Scientific Pathfinder (HERMES-TP/SP) mission is to develop a cheap and scalable network of 3U nanosatellites to promptly detect, localize and probe high-energy astronomical transients such as Gamma-Ray Bursts and electromagnetic counterparts to gravitational waves events.

        HERMES will be able to detect GRBs prompt emission over a broad energy band ranging from a few keV to MeV with high temporal resolution and localization accuracy. The key to achieve these ambitious goals is a simple yet innovative miniaturized detector design, in which Silicon Drift Detectors (SDD) play the double role of sensor for scintillation light and independent detector for low energy X-Ray.

        In the present we illustrate the HERMES payload architecture and discuss the implementation and performances of a test-bench detector developed for characterizing GAGG:Ce scintillators - SDD coupling. This test-bench allow us to perform X and gamma-ray measurements using HERMES-like detector.

        Speakers: DELLA CASA, Giovanni (University of Udine) , DILILLO, Giuseppe (Università di Udine)
    • 17:00 17:20
      06 Severe Accident Monitoring: Parallel Intensive Oral Sessions

      Oral Session

      • 17:00
        #06-172 Long-term transmission characteristics of CYTOP fiber exposed by gamma radiation 5m

        Polymer optical fibers (POFs) have been attracting substantial attention from research and industry thanks to particular mechanical properties, simple handling, biomedical compatibility and safety. Among POFs, cyclic transparent amorphous fluoropolymer (CYTOP) fiber demonstrates radically low attenuation in the telecom transparency windows of 850 and 1310 nm, thus becoming an efficient solution for short distance communication links. Besides fiber communications, CYTOP fiber has been also intensively studied for various sensing applications: fiber Bragg gratings, Brillouin scattering and intermodal interference are good examples of investigated sensing principles.
        A particular topic of research is the use of POFs in radiation environments, including possible solutions for gamma radiation dosimetry. A basic step in this direction is the study of gamma radiation influence on fibers’ properties aiming for evaluation of their transmission characteristics degradation. Recently, the gamma-radiation induced attenuation (RIA) in CYTOP fibers was extensively investigated. The RIA was measured in the visible and near-infrared spectral bands, the influence of temperature and relative humidity on irradiated fibers was studied, and the distributed gamma-radiation sensor was proposed. However, evolution of the RIA properties versus time was not considered.
        In this work, we focused on the long-term (residual) RIA of the CYTOP fiber. We irradiated 2-m samples of the graded-index CYTOP fiber with core diameter of 50 µm (Chromis Fiberoptics) to 1, 5, 20 and 50 kGy doses, and measured the RIA five months after irradiation. The experimental setup was based on broadband supercontinuum light source Super K Compact, 450-2400 nm (NKT Photonics) and optical spectrum analyzer Yokogawa AQ6374, 350-1750 nm. We utilized cut-back technique and the attenuation spectrum of unirradiated CYTOP fiber, measured in advance. The results demonstrated significant residual RIA in the visible range as well as at wavelengths above 1300 nm. For example, values of 40 dB/m at λ = 600 nm and 68 dB/m at the RIA spike in the vicinity of λ = 1450 nm were measured for the sample irradiated to 50kGy.
        To investigate the evolution of the RIA with time, we irradiated two additional fiber samples to 5 and 17.5 kGy doses and continuously measured the RIA up to 80 hours with the measurement started shortly after irradiation. The results demonstrated contrasting RIA time evolutions for visible and near-infrared spectral ranges: partial recovery of transmission properties in the visible range, while in the near-infrared spectral range the RIA increased with time.
        Thus, we have found that the post-irradiation time-evolution of the RIA in gamma-irradiated CYTOP fibers strongly depends on the wavelength range. The transmission partly recovers in the visible range, while at wavelengths above 1300 nm the RIA grows and saturates. We conclude that the RIA induced by gamma irradiation in CYTOP fibers persists for a long time, and can be considered as permanent in the NIR.

        Speaker: Dr CHAPALO, Ivan (University of Mons)
      • 17:05
        #06-66 The VINON-LOCA test facility: exploring the LOCA phenomenology through an out-of-pile thermal sequence on irradiated pressurized fuel rod 5m

        Since the Halden loss-of-coolant accident (LOCA) test series IFA-650, a major safety interest has raised for Fuel Fragmentation, Relocation and Dispersal (FFRD) during a LOCA sequence. Besides the characteristics and the behavior of the fuel ejected from the rod after the clad burst, the occurrence of the ballooning, of the fuel fragmentation and of the possible fuel relocation within the rod before the clad rupture is still to be investigated. For this purpose, the VINON-LOCA program is devoted to perform Out-Of-Pile heating tests on irradiated repressurized fuel rod, in the framework of a trilateral agreement between EDF, Framatome and CEA.
        The experimental set-up is implemented in a hot cell of the LECA-STAR facility of the IRESNE Institute of CEA, at the Cadarache research center. After a thermal reconditioning phase at 300°C, the thermal transient sequence consists in a ramp test (typically 5°C/s from 300°C to 1000°C) performed in an induction furnace under an inert gas circulation. The tested rodlet is initially pressurized at around 50 bar (ambient conditions) and equipped with thermocouples attached to cladding (rather than welded) for monitoring the clad surface temperature.
        One of the specificity of the VINON-LOCA experiment is the online gamma measurement systems: the expected ballooned region is examined both with a classical collimated gamma spectrometry station and with an innovative 2D fast gamma imaging system to have on-line measurement of fission product activity (specifically developed for this application with CEA/DRT). Another gamma spectrometry station targets the experimental rodlet plenum region in order to estimate Fission Gas Release (FGR). In this region, the gas composition is also measured online with an acoustic sensor (developed by the IES Electronic Institute at Montpellier) coupled to pressure and temperature on-line measurements. . In addition, FGR after clad failure is addressed thanks to both online and post-test gas analyses (microchromatography, online gamma spectrometry, gas sampling). Pre and post-test characterizations of the rod include in situ gamma scanning (i.e. with the rod still in the furnace) using the capabilities of the collimated spectrometry gamma station. Additional measurements are performed on a dedicated instrumented bench: gamma scanning performed with a better accuracy, possibly gamma emission tomography, and laser diameter measurement. The fuel fragments ejected from the rod are collected, weighted and sieved after the test in order to evaluate their size distribution.
        An extensive and substantial qualification campaign has been performed for validating the test protocol and conditions and for qualifying the instrumentation. It has included tests on an out-of-cell twin mockup, tests on dummy inactive rods in the hot cell, and was completed by some preliminary modellings and calculations. This allowed achieving successfully the first experimental qualification test of the program end of 2019 on irradiated UO2. A second qualification test is planned in 2021.
        This contribution will highlight the experimental set-up and its associated instrumentation, illustrated by some results issued from the qualification campaign and preliminary tests.

        Speaker: BIARD, Bruno (CEA, DES, IRESNE, DEC, Cadarache F-13108 Saint-Paul-Lez-Durance, France)
      • 17:10
        Empty 10m
    • 17:20 18:20
      Visit to posters

      Poster Session

    • 10:40 11:00
      Coffee Break 20m
    • 11:00 12:40
      02 Space Sciences and Technology

      Oral Session

      • 11:00
        #02-229 Introduction of hybrid silicon/scintillator detector for space experiment BION-M2 20m

        Space experiment BION-M2 focuses on research of effects of ionizing radiation to biological samples such as geckos and mice. The return module of BION-M2 carries the life support systems for all living organisms. The launch of the satellite is planned on the year of 2023, the average altitude will be 800 km.
        The newly developed hybrid silicon/scintillator detector will support this experiment with measurement of radiation doses in combination with other detectors such as thermoluminescence detectors, track-etched detectors CR-39 and silicon-based detectors.
        The detector is based on silicon strips which are deployed in 4 layers. The layers are placed in such pattern that when a charged particle crosses all four layers the incident angle of the particle can be calculated. Each layer contains 64 strips. Altogether there are 256 strips in all layers. The block of plastic scintillator EJ-276 is placed between second and third layer. Plastic scintillator is coupled with silicon photomultiplier (SiPM) which converts scintillation photons to electrical signal. EJ-276 supports technology of pulse shape discrimination (PSD) which can be used to estimate the linear energy transfer (LET) of the particle. The considerable effort has been made to optimize the PSD performance with the SiPM component. The detector is designed to manage large fluences of radiation which can be encounter for example when satellite passes through South Atlantic Anomaly (SAA) which is located at the coast of Brazil. Such arrangement of the detector provides detection capabilities to calculate radiation quantities related to biological response to ionizing radiation. The goal is to obtain time resolved LET spectrum, absorbed dose and dose equivalent rates.
        We would like to introduce newly developed hybrid silicon/scintillator detector and present the preliminary results from experiments performed at high energy accelerators. The type of the particles and their energy were chosen so it was close to the ionizing radiation at Earth orbit.

        Speaker: Mr SOMMER, Marek (Department of Radiation Dosimetry, Nuclear Physics Institute, The Czech Academy of Sciences)
      • 11:20
        #02-124 Wide bandgap SiC and GaN power technologies for space applications: impact and modeling of aging with electrical stress on the radiation sensitivity 20m

        Silicon carbide and GaN power MOSFETs are very promising candidates for space power applications requiring high efficiency, fast switching and reduced mass and volume. However, to meet those requirements, such technologies have also to demonstrate a good tolerance to space radiations, in particular regarding Heavy Ion-induced Single Event Effect and Total Ionizing Dose effects. From their material properties point of view, SiC and GaN technologies are superior compared to Silicon based Power devices. Both technologies exhibit indeed higher breakdown electrical field and higher carrier saturation velocity. This raises the question about the sensitivity to radiation of those technologies. Some studies have focused on TID effects, displacement damage, electron and heavy ions. However, one cannot consider that radiation is the only stressor during space missions. In fact, the switching operating of power devices acts as a background electrical stress which might affect the overall lifetime of the power system, through mutual effects with radiation constraints. The purpose of this study is therefore to assess the impact of electrical stress on the sensitivity to TID under x-ray irradiation. Both SiC and GaN technologies will be investigated. TCAD simulation will support experimental results as well as a compact model that takes into account the eventual drift of threshold voltage after irradiation due to the creation of traps in the gate oxide for SiC power devices and in the AlGaN layer or even at the AlGaN/GHaN interface when considering GaN power technologies.

        The devices used in this study were commercial power transistors manufactured by CREE (SiC MOSFET) and EPC (GaN HEMT). SiC MOSFET is a vertical device whereas the GaN HEMT is a lateral device. Maximum drain to source ratings are respectively 900 V for SiC Power MOSFET and 200 V for GaN power HEMT. Looking at the drain to source current ratings, SiC devices exhibit a 23 A maximum drain to source current where GaN devices exhibit a 3 A current rating.
        Characterizations before and at each irradiation step included IDSVG and IGSVG measurements, while VGS was swept from 0V to the maximum rated gate voltage. During characterizations VDS was held at 100mV with sources grounded.
        At first the variability of the batch was investigated as well as the stability of the threshold voltage for each device. Then, a constant current stress (CCS) has been performed on a first batch of devices in order to weaken them and to investigate the impact of electrical stressing on TID sensitivity. Electrical stress consisted in a 100μA constant current applied during 100 s and resulting in a final injected charge of 10 mC for all the stressed devices. Such injected charge has been chosen in order to be far below the average charge to breakdown QBD found to be about 160 mC.
        TID irradiation have then been performed at room temperature, using the on-site X ray source (X-RAD 320 x-ray cabinet) with 320 keV electron gun energy and current of 12.5 mA. Devices were all irradiated at 80 % of their maximum VDS rating while source was grounded. Maximum negative bias was applied to the gate to make sure that the transistor was off, even in case of a strong VTh shift. SiC and GaN devices were respectively irradiated till 1 MRad and 500 kRad with a dose rate of 14.75rad/s. Characterization of each device was performed before and after each irradiation step allowing to investigate any variability among the batch after electrical stress and during irradiation.

        Experimental results obtained after electrical stress and subsequent TID irradiations will be presented (see attached file). Results will be analyzed through TCAD simulation, using a dedicated TCAD tool allowing to consider the trapping dynamics. Then, we will detail a compact model that takes into account TID effects on static electrical characteristics.

        Speaker: Dr TOUBOUL, Antoine (IES - Univ. Montpellier)
      • 11:40
        #02-30 Development of a Neutrino Detector Operated in Space 20m

        If a neutrino detector can be operated in space it would facilitate at least three future missions for new science in heliophysics, astrophysics and dark matter research. By going very close to the Sun the neutrino flux can increase more than ten thousand times that on Earth and by going away from the Sun the solar neutrino backgrounds for Dark Matter searches in the low energy band below 10 GeV/c2 can be reduced by fifty thousand or more. The Sun could be used as a gravitational focus of neutrino sources detectable with either a spacecraft out at 20 to 40 Astronomical Units or by using the atmosphere of a gas giant such as Uranus or Neptune. This could be used to image the Galactic Core, the second brightest neutrino source in the sky. All of these ideas rely upon making a neutrino detector in space function without the large shielding typically needed on Earth; results from a NASA funded research showing how a special double pulsing method developed would work to permit operating a neutrino detector in space.

        Speaker: SOLOMEY, Nickolas (Wichita State University)
      • 12:00
        #02-23 Ultra Fast timing detectors with applications in cosmic ray physics, medical science and other domains 20m

        We were developing a cheap concept of a fast commercial silicon pixel detectors combined with a state of the art readout electronics based on ultrafast sampling and waveform digitizers of the signal giving simultaneously its time and amplitude.
        This concept was used originally in High Energy Physiscs and we transpose it
        with two applications from cosmic ray physics and particle therapy medicine.
        For the astrophysics program, the principle is to perform direct particle identification and energy measurement using a few layers of cheap detectors that will be sent to space in collaboration with NASA. We will discuss the performance of our detectors as well as the results from simulation before launch in few months.
        For the medical application in particle therapy, the principle is to measure with high accuracy the beam dose delivery during proton beam treatment individual bunch by bunch to know with high accuracy especially for flash high intensity beam treatment. Results performed at the hospital of Dublin, Ireland, will be presented.

        Speaker: ROYON, Christophe
      • 12:20
        Empty 20m
    • 11:00 12:40
      05 Nuclear Power Reactors Monitoring and Control

      Oral Session

      • 11:00
        #05-225 Dynamic Response Analysis of Fast-Spectrum Self-Powered Neutron Detectors 20m

        Self-powered neutron detectors (SPNDs) are attractive online flux monitoring tools as they are compact, simple, and produce a signal proportional to local neutron flux without the need for an external source of power. They are however optimized for thermal-neutron reactions and many do not have the required sensitivity to neutrons with energies approaching 1 MeV. Because neutron-capture cross sections in the fast-energy region are often orders of magnitude lower than those at thermal energies, the device performance (i.e., electrical current output at a given neutron flux level) will inevitably deteriorate in a neutron flux peaking at 0.5 MeV as is common in fast-reactor spectra. Therefore, any improvement in reaction rates within the emitter will help increase device sensitivity and boost the signal-to-noise ratio of the instrument. To date, the published efforts surrounding the development of SPNDs for use in fast reactors—what we are terming a fast- spectrum SPND (FS-SPND)—have centered around the examination of currently developed SPND technology for sensitivity to 0.5 MeV neutrons. In this work, we have performed a survey of Evaluated Nuclear Data File (ENDF) cross section data in an effort to identify materials, known and unknown, suitable for use as an emitter in a FS-SPND. As SPNDs generate an electrical current as a result of nuclear reactions within the neutron-sensitive portion of the detector, termed the emitter, we have focused our initial efforts on the identification of fast-neutron sensitive materials. To this end, we have identified five new emitter materials: tantalum, terbium, thulium, lutetium, and iridium. We have also identified two currently used emitter materials as suitable for use in a fast-spectrum reactor: rhodium and silver. Each of these emitter materials exhibits a sensitivity on par with currently used emitters operating in a thermal-neutron field. In this paper, we first derive the radiative capture and decay-chain rate equations for each emitter material to develop their dynamic response models. We then assess their anticipated performance as neutron monitoring instruments. Finally, we discuss dynamic compensation approaches to improve the tracking performance for local neutron flux changes in the vicinity of the device.

        Speaker: Dr CETINER, Sacit (ORNL)
      • 11:20
        #05-32 The analysis of different physical mechanisms during cladding failure evolution and detection in sodium cooled fast reactors 20m

        To keep the dependability of Sodium Cooled Fast Reactor, the ‘‘clean sodium concept’’ is demanded, which means that the sodium is free from contamination. The release of fission products is searched for by a contamination measuring system. We need to have a comprehensive description of cladding failures and the detection of contamination, including the failure occurrence on the fuel pin, the transfer process through the sodium and cover gas, the measurement efficiency, etc. We aim to identify the important parameters of physical phenomena, with modelling and simulations based on the return of experiments from past reactors such as Phenix. There have been a total of 15 open pin failures in Phenix reactor. Through studying these detected signals, we can get a better physical explanation and description of the evolution of failures.

        The detection system is related to different stages of the evolution of fuel pin, with different types of fission products, various release mechanism and physical properties. During the evolution of the failed fuel pin, gaseous fission products is released on the first stage of failure followed with other fission fragments including delayed neutron precursors. There are two main parts of the whole detection system. DRG (Détection des Ruptures de Gaine) system is functioned for detecting the crack of the cladding, with only gaseous fission products release. DND (Détection des Neutrons Différés) system is functioned for ruptures with solid fission products leakage.

        We propose a block diagram that identify the physical phenomena occurring at each stage of the detection system, and have a focus on two parts for illustration. For gas signal, we propose a qualitative modeling of transfer function to describe the time broadening of the gas release from the fuel pin to the detector. The result matches well with Phenix experiment data, with the same order of magnitude of the time broadening and the same shape of exponential decreasing. For delayed neutron measurement, we study the complex evolution inside the fuel pin to interpret the different release mechanism and neutron signal variation.

        Speaker: Ms DING, Congjin (CEA, DES, IRESNE, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache F-13108 Saint-Paul-Lez-Durance, France.)
      • 11:40
        #05-27 Multi-parameter discrimination of partial-discharge-induced pulses in fission chambers designed for sodium-cooled fast reactors 20m

        Fission chamber technology has been identified as the most suitable method for neutron detection to be used in the vessel of a sodium fast reactor and may possibly also be used for neutron diagnostics in fusion power (ITER, DEMO). This type of detector, namely High Temperature Fission Chamber (HTFC), must be able to operate under high irradiation of up to 10^10 n/cm².s, must have high sensitivity (~1) to detect fast neutrons, and must operate at high temperatures, up to 650°C in sodium fast reactors and probably at even higher temperatures in fusion power reactors. It has been observed that an effect of the high temperature environment is an additional signal, ascribed to partial electrical discharges, which is of similar amplitude and duration as the useful signal caused by neutron interaction with the fissile layer inside the fission chamber. Partial discharges (PD) in gases are often referred to as “Corona discharges”, and are associated with the phenomenon of an electronic avalanche in the presence of strong electric field. Previous work by Hamrita et al. demonstrated how mono-parameter discrimination based on signal full width at half maximum values can discriminate neutron pulses from PD pluses, but only for small diameter fission chambers.
        In the presented work, neutron and PD pulses are collected using two large diameter HTFCs, designed to be used in the vessel of a sodium fast reactor, tested at high temperature and under neutron beam irradiation and off beam. Firstly, starting from preliminary results of Hamrita et al., the collected pulses (neutron and PD) are separated using a mono-parameter discrimination method based on full width at half maximum. Then the same data are separated using the proposed multi-parameter discrimination method by means of KNN and SVM algorithms.
        Finally, the results obtained from the tested discrimination methods were compared and a PD-neutron discrimination process for HTFC was proposed.

        Speaker: LYNDE, Clément (CEA)
      • 12:00
        #05-163 Fiber Bragg Grating based sensors for applications in harsh environments. 20m

        Nowadays, the optical fiber sensors (OFSs) have attracted the interest not only of researchers but also of industries, for their applications in harsh environments characterized by high temperatures and/or radiations. The Fiber Bragg Grating (FBG) based technology is one of the most widely studied under radiation [1]. Even today it presents qualities unbeatable by other OFSs, such as its high acquisition repetition rate, that can reach at least 20 kHz with an adequate acquisition system. Even if it is not a distributed sensor, as the ones based on scattering phenomena, it is easily multiplexable, which means that several point sensors can be placed in series on a same fiber. However, the research on the radiation-response of different types of gratings is still active, since the radiation degrades their performances through two phenomena:
        - Radiation-Induced Attenuation (RIA), which reduces the signal transmission,
        - Radiation-Induced Bragg Wavelength Shift (RI-BWS), which introduces an error on the sensing parameter measurements.
        In this work, we will compare two of the best FBG technologies that are considered for applications in severe environments with the more classical FBG type.
        o The Rad-Hard type II gratings (RH-type II-FBGs) are manufactured with the patented procedure developed by Areva and the Laboratoire H. Curien [2]: these type II FBGs were written with a femtosecond laser at 800 nm in radiation-resistant fibres, having a fluorine doped or pure-silica-core. Before the irradiation they were subjected after their writing to a short thermal annealing at 750°C. The chosen manufacturing process parameters combined with this pre-treatment enable to reduce the RI-BWS below 10 pm (error of about 1°C on the temperature measurements performed by the sensor) after MGy dose of irradiation [3].
        o The regenerated gratings (R-FBGs) are obtained by treating at high temperatures (higher than 650°C) a strong type I grating inscribed in a H2-loaded Ge-doped fiber by an UV laser (continuous wave or pulsed in the nanosecond regime) [4]. The seed grating quickly disappears during the annealing and, in its place, another grating, the R-FBG, appears. The new grating is resistant to high temperatures and, preliminary results show that its radiation resistance increases at high temperatures, with a RI-BWS of only 10 pm, at irradiation temperature higher than 150°C at 0.1 MGy dose [5].
        o The type I-UV FBGs are the most classical grating type: the FBGs were written with a cw UV laser (at 244 nm) in a H2-loaded Ge-doped fiber and thermally stabilized at 120°C. Contrary to the two previous gratings, this one does not resist to high temperature or to radiations: the Bragg peak shifts with the dose; moreover, the higher is the temperature, the lower is the BWS [1].

        It is worth noticing that both R-FBG and RH-type II FBG undergo a treatment at high temperature, however, the mechanisms at the origin of these gratings are very different and still under investigation. Moreover, whereas the R-FBG has to be written into a photosensitive fiber, as the Ge-doped one, the RH-type II-FBG can be inscribed into a rad-hardened fiber, as the F-doped one, which is less affected by the RIA compared to the Ge-doped one.
        We report in Fig. 1 a first example of the Bragg wavelength shift induced by the gamma-rays at room temperature (RT) on the three FBGs. The experiments have been conduced at the IRMA facility of the IRSN (Saclay, France).

        Fig. 1. Bragg wavelength shift, corrected for temperature fluctuation (recorded by thermocouples) induced by gamma-rays on the three types of gratings at RT, dose-rate being 1.3 kGy/h, up to TID of 500 kGy.

        The BWS of the RH-type II-FBG is only few pm, as expected for this rad-hard solution. The type R and type I-UV FBGs, instead, red-shift of ~30 pm and 50 pm, at the total ionizing dose (TID) of ~500 kGy. The reported curves are a bit noisy, probably because of a non-perfect correction of the nights and days temperature fluctuations.
        In the final paper, these three grating types will be irradiated under gamma- and X-rays at different temperatures up to the MGy dose level, in order to investigate the real performances of the sensors based on these technologies. We will also investigate different pre-loadings of the Ge-doped fiber for the R-FBGs inscription, since our group recently showed the possibility to regenerate FBGs written in D2-loaded Ge-doped fibers [6].

        [1] A. Gusarov and S. Hoeffgen, “Radiation effects on fiber gratings,” IEEE Trans. Nucl. Sci., vol. 60, no. 3, pp. 2037-2053, 2013.
        [2] AREVA and Laboratoire Hubert Curien, "Procédé de fabrication d’une fibre optique traitée pour capteur de température résistant aux radiations,” France Brevet 13 62691, 16 December 2013.
        [3] A. Morana et al., Opt. Lett., vol. 39, no. 18, pp. 5313-5316, 2014.
        [4] J. Canning, Laser Photonics Rev., vol. 2, no. 4, pp. 275–289, 2008.
        [5] T. Blanchet et al., J. Light. Technol., vol. 37, no. 18, pp. 4763–4769, 2019.
        [6] T. Blanchet et al., submitted to Trans. Nucl. Scie.

        Speaker: Dr MORANA, Adriana (Laboratoire Hubert Curien)
      • 12:20
        #05-234 Advanced sensor technologies for real-time temperature measurement in nuclear reactors 20m

        In 2012 the US Department of Energy Office of Nuclear Energy (DOE-NE) initiated the Advanced Sensors and Instrumentation (ASI) program as part of the Nuclear Energy Enabling Technologies (NEET) research as part of crosscutting RD&D activities to advance the state of nuclear technology, improve its competitiveness, and promote continued contribution to meet the nation’s energy and environmental challenges. The objective of the ASI program is to provide reliable, cost-effective, real-time, accurate, and high-resolution measurement of the performance of existing and advanced reactor core and plant systems. Instruments are designed, fabricated, and tested in relevant and operational conditions to advance their technological readiness to a point in which they can be integrated in I&C systems without the significant cost and risk associated with development activities. This technology maturation is made possible by the deployment of developmental instrumentation in Material Test Reactors (MTRs) irradiation experiments aimed at the characterization of advanced reactor components, such as advanced fuel forms. In the near term therefore the requirements on sensor technologies are driven by the irradiation test conditions and their research objectives.
        This paper reports on the effort to develop technological solutions to the measurement of temperature in irradiation experiments aimed at the development of core components of nuclear reactors, including its fuel assemblies. The scope of the work extends from the current fleet (primarily Pressurized Water Reactors) to advanced reactor concepts, including Small Modular Reactors and micro-reactors. Considering experiments that address both operational conditions and design basis accident cases, the requirements on temperature measurements extend up to 2000°C under a wide range of neutron flux and integrated fluence depending on the facility under consideration – from a peak thermal flux of 1x1015 n/cm2-s and fast flux of 5x1014 n/cm2-s in the Advanced Test Reactor to the sub-second burst of neutron and gamma radiation in the Transient Reactor Test Facility (TREAT), both located at the Idaho National Laboratory. In addition to operating temperature and radiation resistance, continuous or discrete distributed sensing capability is another important requirement to consider for technologies development.
        Progress in three areas of research are reported: High Temperature Irradiation Resistant (HTIR) thermocouples, Ultrasound Thermometers and optical fiber sensors. HTIRs are thermocouples based on Molybdenum and Niobium alloys and they have been investigated at INL for more than 10 years and have demonstrated reliable performance up to 1600°C. The recent deployment in irradiation test aimed at the characterization of TRISO fuel, in parallel with commercialization efforts, enabled the collection of sufficient data to satisfy the Preliminary Design Review towards their qualification process. Ultrasound Thermometers are acoustic sensors that utilize a radiation resistant transducer and a passive waveguide to derive temperature from the measured sound speed in the waveguide materials. Discrete features in a solid waveguide or a waveguide design based on a bundle of thin wires with stacked ends allows discrete multi-point sensing. The selection of refractory alloys for the waveguide (Molybdenum or Tungsten) allows temperature measurement beyond 2000°C, providing the transducer is maintained at or near reactor coolant temperature. Optical fiber sensing is now considered for many irradiation tests in TREAT, and through accurate selection of fiber materials (radiation hardened fibers) and interrogation techniques their performance characterization has been extended to neutron fluences that are compatible with extended testing in high flux facilities such as ATR. Temperature and radiation limits depend strongly on the sensing method and the fiber material, limiting their current applicability to less than 800°C. Several examples of the use of optical fibers for temperature measurement in irradiation experiment are presented and their results discussed.

        Speaker: CALDERONI, Pattrick (INL)
    • 11:00 12:40
      09 Environmental and Medical Sciences

      Oral Session

      • 11:00
        #09-55 The CORSAIR Project. Characterization of a portable instruments for NORM characterization of stone blocks 20m

        The CORSAIR (Cloud Oriented Radiation Sensor for Advanced Investigation of Rocks) project was born to meet the EU guidelines 2013/59/EURATOM and now the Italian legislation decree D.Lgs.101/2020 on safety standards for protection against ionizing radiations.
        With this project there is a specific focus on the detection of the NORM (Naturally Occurrence Radioactive Materials) contribute coming from stone blocks used in the engineering field.
        As a matter of fact, radiological measurements and traceability of materials are two aspects becoming more and more important in this and for the green building sectors.
        In Europe exist regulatory restrictions on the maximum level of radiological emissions for materials used in public and private building structures. A similar approach is also adopted in other countries all over the world. These legislations not only are fixing the radiological limits, but forbid purchasing and incoming of not controlled or above limits materials in the country. These legislations are acting on import/export of the stone market, one of the Italian sectors already affected by the incoming of low cost not natural (Okite, high pressure laminate or ceramic stones) and not certified natural stones.
        Alle these aspects drove the CORSAIR project to design an automated cyber-physical system capable of providing a real-time measurement of the radioactive activity concentration index for building materials according to regulations of more than 20 different countries. It quantifies activities, abundances and related effective dose-rates of natural radionuclides ($^{40}$K, $^{232}$Th, $^{238}$U) and their progenies).
        The system is empowered by cloud-based technology consisting of sensing nodes, data collection gateways and a centralized cloud application. These components are interconnected in a star-of-stars topology, exploiting GPS, Bluetooth, LoRaWAN, WIFI network connection and providing specific user interfaces with an Android app running on smartphones.
        Measurements are conducted through in situ $\gamma$-ray spectroscopy techniques on blocks of rock at quarries or processing centres.
        The system is designed for providing an autonomous, fast, repeatable, real-time and non-destructive method to measure the radiometric indexes with a 30 min measurement.
        The innovative aspects of the system are due to the integration of several elements: its autonomous operation that does not require expert operator to provide results, the easy fruition of the results of the material characterization thanks to the cloud database, the presence of the RFID tags used to uniquely identify each analysed block, the energy calibration and the fully-automated results computation.
        All system outputs are finally synced to a cloud database, where they can be easily accessed by operators and clients, enabling to trace the materials along the full market chain, from extraction to the final customer, with modern technologies preventing the placing on the market of blocks hazardous to public health and as a warranty of the origin of the block thanks to the RFID tag used.
        This presentation after describing the system will provide tests results performed in-situ on real blocks and the comparison with laboratory measurement performed to characterize the radiological system.

        Speaker: FANCHINI, Erica (CAEN s.p.a.)
      • 11:20
        #09-64 Deconvolution methods used for the development of a neutron spectrometer. 20m

        This presentation stems from a thesis project in progress, leading to a unique multi detectors Bonner sphere. Neutron spectrometers like Bonner spheres have been studied and developed for more than 60 years for astrophysics and radiological protection applications. In particular, in radiological protection, it is a crucial challenge to determine the initial energy of the incident neutrons with a sufficient accuracy, to evaluate their damage to human body.

        During the simulation part of the development, GEANT4 simulations were performed to study the response of Bonner spheres instrumented with 30 to 50 active detectors. The aim of these studies is the optimization of the number and size of the active detectors and the diameter of the sphere. One of the most significant issues was the unfolding of the different simulated “measures” obtained by the active detectors to get the original energy spectra with neutron energies between 0 and 20 MeV and to evaluate the errors related to these spectra. This unfolding is a key point for the optimization of the multi detector Bonner sphere.

        In particular, three deconvolution methods have been developed and tested : the least-squares, the maximum entropy coupled to maximum likelihood and the maximum likelihood methods, which require a certain number of parametrical adjustments to become optimal. In general, every method needs a converging algorithm to be accurate.

        • For instance, in the maximum likelihood methods, where the likelihood function is based on Poisson distribution, it is due to the nonlinearity of the involved equations, which cannot be analytically solved. However, these methods work for underdetermined – in case of an insufficient number of detectors in comparison to the number of groups of energies - as well as overdetermined problems, and can be coupled to the maximum entropy method ;

        • In the least-squares methods, it is due to the important sensitivity to perturbations related to the matrix inversion and to non-positive solutions. To resolve this, the iterative Levenberg-Marquardt’s method can be used. It implies nonlinear equations as well but leads in particular to only positive solutions, by inserting exponential functions in the solution and a Levenberg-Marquardt’s parameter λ which has to be modified after each iteration. But this method does not work for underdetermined problems.

        The results obtained with these methods have been compared to the unfolding package distributed by the Nuclear Energy Agency, UMG 3.3 - Unfolding with MAXED and GRAVEL. This package consists of two unfolding programs based on maximum entropy (MAXED) and least-square methods (GRAVEL), one spectrum analysis program (IQU) and one graphical display program (UMGplot).

        The different methods will be introduced, then some examples of unfolding spectra will be presented and the performances of the different methods discussed. Finally, firsts results of optimizations of the dimensions of the sphere and the dimensions and number of active detectors will be exposed. First developments and tests of neutron detectors dedicated to the multi detector Bonner sphere will also be presented.

        Speaker: SIMONETTI, Claude-Alexandre (EAMEA Cherbourg / ENSICAEN / LPC Caen)
      • 11:40
        #09-218 Particle identification and tracking by the use of a pixel-based semiconductor radiation detector coupled with voltage controlled oscillators 20m

        In this paper, an approach based on a semiconductor radiation detector exhibiting a matrix organization coupled with the use of rings oscillators for current detection is developed. This approach is based on the reading of the information related to indirect output parameters of the detection chain instead of directly measuring the current from the sensor. This solution is interesting because it avoids most of the design problems related to the use of a charge sensitive amplifier and keeps the current shape. Indeed, in applications such as gas prospection or medical therapy, the knowledge of the electrical signature of the particles is required. Moreover, knowing the actual shape of the generated current at the output of the detector allows easier post processing of the signal. Then, in our system, the information is extracted by correlating the initial oscillation of the system with that of the system after the particle has passed the detector (Fig. 1). To be able to make a diagnosis, the requirement is then to link the output information (i.e. the mean voltage of the oscillator) to the input information (current stimuli). The initial version of the chain was implemented in bulk technology. As the full integrated circuit (detector + readout electronic) is supposed to work in a radiative environment, the chain has been implemented on a 130nm partially depleted Silicon On Insulator (SOI) process, in order to make it more tolerant to radiation.
        The detection chain has been simulated at circuit level using “Spectre” simulator (SPICE-based) under Cadence DFII ©. The readout circuits is composed of a ring voltage controlled oscillator (VCO) working at high frequency. Each VCO is composed of three specific delay cell in CMOS-SOI technology making their implementation very simple. This simplicity allows several configurations of use. The current source at the input of the VCO comes from the realistic simulation of the 3x5 matrix using TCAD device simulation tools (Synopsis ©). The effect of the ion strike is simulated using the Heavy Ion module of Synopsis [1], considering an electron-hole pair column centered on the ion track axis. The Linear Energy Transfer is defined as the energy lost by the particle, by unit of length [2]. An actual variation of the LET was integrated in our simulations, based on the value given by SRIM tables [3]. The VCO based chain presented here is optimized for the detection and identification of particle fluxes lower than 109 particles per detection area and per second. This chain could be particularly suitable for the detection of low energy particle such as alpha particle crossing the device with an initial energy of 0.9MeV, corresponding to a range of 2.3 m in the silicon matrix . The aim is to study the end of the path corresponding to an alpha particle of 1.47MeV generated by the initial interaction of thermal neutron with boron-10. The case of a more energetic particle is also considered. This is a 50MeV Aluminum which could be produced by the interaction of fast protons with silicon, for instance [4].
        In a previous work, the various currents generated in the matrix of detection were studied and several ways of improvement of the detection ability were explored [5]. We propose here a complementary solution to extract the total collected charge from the detector what should facilitate particle identification. Other particle species will be presented in the final paper.
        After being generated by the sensor, the signal has to be conditioned in order to allow particles counting and/or recognition. The key point is how the output parameters of the VCO chain can give information related to the input current. This could be done through the analysis of various characteristics extracted from the VCO output voltage (Fig. 1) that is the variation of the average output signal (ΔVmax) versus the maximum of the input current (Imax) for different particle charges. In [6], a linearity curve linking input to output parameters was determined for a 2GHz VCO. A new linearity curve has been obtained for this new 4.3GHz VCO [5]. Through calibration curves, the output parameters can be linked to the input currents, which could allow the incoming particle tracking. The matrix studied here contains fifteen contacts (Fig. 2). All the currents of the matrix, generated using Synopsis have been injected in the fifteen VCOs contacts. The novelty in this work is that one more VCO is added in order to collect the the total collected charge. Indeed, the linearity of the VCO response may be affected if too high magnitude level of current is produced by the pixel detector, leading to a saturation of the VCO response. In order to overcome this limitation, we increased the size of the bias transistor of the delay cell composing the oscillator, leading to a high sensitive readout circuit but working a lower frequency. This second topology has been used for the total charge calibration. Then oscillator used at the bottom contact oscillates at 2.98GHz.
        Fig.3 presents first results obtained with this VCO on four cases: an Al particle an alpha particle following two configuration corresponding to an horizontal crossing and a diagonal crossing of the cell. On the axis, the total collected charge at the bottom contact of the matrix is plotted. Its corresponds to the amount of charge collected in the whole structure. This gives a direct information about the deposited charge. Then the charge deposited by the Al particle is higher than the one of the alpha, and the charge collected for an horizontal track crossing 5.05m is higher than the diagonal case crossing 2.82m. Theses analysis will be details in the final paper. This trend is respected by the integral of the mean voltage value of the VCO. Then the linearity between the output and the input response is preserved (Fig.4). This will be checked for other particles species in the final paper. The tracking ability will be also detailed through the analysis of the currents from the fifteen other VCO.

        Speaker: COULIÉ, Karine (IM2NP - Aix-Marseille Université)
      • 12:00
        #09-145 Autonomous submarine radiometrics for offshore particles surveillance 20m

        Underwater robots rarely have the ability to dig themselves into sediment below a body of water,
        as might be required for analytical purposes. However, several aquatic animals can do this, in search
        of food or to evade predators. This project considers the design of an innovative underwater robot
        to dig unaided into the sediment.
        One use of this robot is to detect low-level, low-energy, gamma-ray emitting radioactivity present
        under the seabed or ocean floor, autonomously. The material to be detected is americium-241 (and
        related actinides such as uranium, plutonium), the fission product caesium-137 and potentially
        short-lived debris.
        The domain of autonomous robotic devices for digging and drilling on seabed has not been
        explored extensively despite having a significant number of devices and designs suggested for robots
        which have applications in the domain of sample collection and exploration. Several digging devices
        have been reported in the research often inspired by how aquatic animals dig and drill on the
        seabed or ocean floor such as clams or worms as well as digging robot for the planting of trees and
        adapted screw burying themselves. Planetary expeditions have also been a stimulus for the
        development of these types of machines [1].
        Nevertheless, it is inspiring to observe the performance of these designed robots especially when
        these robots are fully submerged through their digging process. In this project, the aim is to build an
        autonomous robot that can dig itself into the seabed for the purpose of low-level, low-energy
        gamma-ray emitting radioactivity characterisation.
        [1] Nagaoka, K., Kubota, T., Otsuki, M., & Tanaka, S. (2008). Experimental study on autonomous
        burrowing screw robot for subsurface exploration on the Moon. In 2008 IEEE/RSJ International
        Conference on Intelligent Robots and Systems (pp. 4104–4109). IEEE.

        Speaker: MFUNYI, Patrick
      • 12:20
        #09-140 Microradon project 20m

        Radon is the most important radioactive noble gas existing in nature. It belongs to the radioactive decay chains of 238U and 232Th. As an inert gas, radon, and in particular the long-lived isotope, 222Rn (T1/2 = 3.82 day), has high mobility and can easily escape from materials containing traces of Uranium. For low energy and low counting rate experiment in particle and astroparticle physics (some neutrino experiments or direct dark matter research), the diffusion of radon, from the detector components, to active part of the experimental setup is very often the origin of the most important background. A radon concentration below a few tens of atoms of 222Rn /m3 is often a prerequisite for future experiments.

        In this context, the MICRORADON project between three IN2P3 laboratories (CPPM-Marseille, CENBG-Bordeaux and IPHC-Strasbourg) has recently been launched. The aim of this project is to study the physics of radon (emanation, transport and capture) in extreme conditions (low temperature, high pressure, liquid gases, etc.) in connection with the requirements of future very low background experiments. Special attention is paid to the development of radon detectors in accordance with these conditions (low background, high sensitivity, high specificity...)

        After the exposure of the basic mechanisms leading to the presence of radon in an experiment, we will present the main results obtained in the MICRORADON project and the perspectives for the future.

        Speaker: LLIDO, Olivier (CPPM / Aix-Marseille Univ)
    • 11:00 12:40
      10 Education, Training and Outreach

      Oral Session

      • 11:00
        #10-268 A complete training program dedicated to nuclear instrumentation at Aix-Marseille University 20m

        Since 2004, the “Filière Instrumentation” unit of the Physics Department of the Faculty of Sciences at Aix-Marseille University has been collaborating with the CEA to develop different education activities on the topic of nuclear instrumentation. These activities involve also several partners such as EDF, IEEE NPSS, international nuclear centers (SCK-CEN in Belgium, NCBJ in Poland, CNESTEN in Morocco) and other key international partners such as the Nuclear Reactor Laboratory of the MIT in USA and the Jozef Stefan Institute in Slovenia. These activities benefit from a strong recognized research activity carried out in the framework of the IM2NP laboratory and the joint laboratory LIMMEX with the CEA.
        The presentation will deal with these different activities and detail the three most recent in particular:
        - A speciality dedicated to test facilities instrumentation. It was developed in collaboration with the CEA in 2004. It is now a diploma co-delivered by AMU and the national nuclear science and technology institute of CEA INSTN. It is part of the master’s degree in Instrumentation, Measurement and Metrology (IMM) and more recently is integrated into the graduate school of a new institute created by AMU in July 2019 called ISFIN (Institute for fusion sciences and Instrumentation in Nuclear environments);
        - A summer school EFMMIN “Ecole Franco-Marocaine de la Mesure et de l’Instrumentation Nucléaires” since 2010. It is a two-year summer school co-organized with the CEA, the CNESTEN, the Sciences Faculty of the Mohammed V University and the AMSSNUR. After editions devoted to research reactors, the fuel cycle, decommissioning, medicine and the environment, and safety, the next edition will focus on innovation in nuclear sensors and detectors;
        - A mobility program called MOBIL-APP since 2018. It is labelled and funded by the Excellence Academy of the A*MIDEX foundation promoting pedagogical innovation, improvement of existing training and international attractiveness. This program concerns the outgoing international mobility of block-release apprentices preparing the IMM master’s degree. It consists in selecting, preparing and sending a group of block-release apprentices for 2 weeks per year to a partner. The two groups realize visits to reactors, laboratories and companies, participate in seminars, short courses, experimental projects and workshops together with local students, visit cultural places and prepare an activity report in English;
        - A new international specialty. It is currently in the process of being set up administratively and pedagogically. This new course will belong to the IMM master’s degree and to the new ISFIN institute. It will dedicated to nuclear instrumentation and measurement for major nuclear research facilities in the fusion and fission fields such as MTRs and tokamaks (for instance the Jules Horowitz Reactor and ITER located in the South of France, 70 km from Aix-Marseille University). This speciality will involve several international partners. For instance the Nuclear Reactor Laboratory of the MIT will carry out training courses, seminar and practice exercises. The students will have to do internships in international facilities after a project realized at Aix-Marseille university labs;
        - A student branch and a chapter dedicated to nuclear instrumentation and associated to IEEE NPSS. It is in the process of creation with about 30 students.

        Speaker: REYNARD-CARETTE, Christelle (Aix-Marseille University)
      • 11:20
        #10-255 The LUCID experiment - releasing the potential of school students in space research 20m

        The LUCID experiment involved groups of students in a school in Kent, UK, working with a huge number of colleagues to put Medipix technology in open space on TechDemoSat 1. Students worked on this outside their formal studies over a period of ten years and benefitted from real engagement with research. LUCID [1, 2] studied the radiation environment in Low Earth Orbit for three years, collecting over 2.1 million frames of radiation data from its five Timepix detectors. Students published their results in Advances in Space Research in 2019. The students who were involved in LUCID are now becoming leaders in science research themselves, working across many areas of physics and engineering.
        Becky Parker [3] the teacher supporting students in this, will outline how this genuine involvement of young people in research can not only inspire students and teachers but also contribute to scientific advance.



        Speaker: Prof. PARKER, Becky (School of Physics and Astronomy, Queen Mary University of London)
      • 11:40
        #10-256 Advanced neutron detection hands-on exercises with MX-10 particle camera educational kit 20m

        MX-10 particle camera [1] as a modified variant of the read-out interface for Timepix pixel detector is dedicated for educational purposes to demonstrate real-time detection and visualization of basic kinds of radiation as alpha, beta, gamma rays and cosmic muons. However, its use is not strictly limited to perform the fundamental exercises that are formulated in the complementary manual book Experiments Using Pixel Detector in Teaching Nuclear and Particle Physics [2], which contains detailed instructions for a set of 50 experiments practicable with Timepix detector. After necessary adaptation, MX-10 can also serve well in more advanced practices as the neutron detection experiments undoubtedly are.
        The contribution presents a set of exercises, methods involved and arrangements of a compact tailor made set-up based on MX-10 edu kit to allow effective performance of neutron experiments in a quite undemanding way easily comprehensible to students. The newly developed experiments are focused on demonstration of basic interactions of neutrons with the matter to provide clear evidence on them. Specifically, the individual exercises cover the following objectives:
        i) The measurement of radiation emitted by a compact a laboratory $^{241}$AmBe radionuclide neutron source and the radiation induced by neutrons interacting in the surrounding materials based on observation of particle tracks resulting from interactions of neutrons and gamma rays in a bare silicon sensor of Timepix detector. The particle track analysis will be applied to recognize the type and energy of every individually interacting radiation quantum. ii) The demonstration of a highly effective way of neutron detection with an adapted sensor is shown applying suitable neutron converters of $^6$LiF (permitting the registration of slow neutron by means of $^6$Li(n,α) reaction) and of polyethylene (the hydrogen rich material for detection of fast neutrons by means of recoil protons). Detection of moderated and fast neutrons (in range 1 - 10 MeV) and their recognition is presented. A possibility of the spectral analysis of a mixed neutron field is also discussed.
        iii) The experiments dedicated to key aspects of neutron shielding and moderation performed with the Timepix detector adapted for detection of slow and fast neutrons realized in various experimental set-ups of different types of materials (high density polyethylene, parafine, Pb, Cd, B, Li and their combination) used for shielding the detector against the radiation from the neutron source. Shielding efficiency of the particular materials is examined considering also the undesirable secondary radiation generated by neutrons interacting with shielding materials.
        The lecture arises from many years of experience gained in using the MX-10 camera to teach the interaction of ionizing radiation at Czech high schools and at a number of international schools/workshops organized in 2018-2020 by the IEEE NPSS focused on representative students of high school and universities. The feedback on the hands-on exercises organized recently as a part of XXI Jorge Andre Swieca Summer School on Experimental Nuclear Physics [3] organized in Sao Paulo 2020 and the virtual IEEE NPSS Workshops on Application of Radiation Instrumentation in Jakarta [4] and Dakar [5] 2020, has proven well how the newly developed exercises help students to understand the physics of neutron interactions with matter. It is worth mentioning here that the practical exercises described in this contribution will be a part of this conference program.

        [1] MX-10 particle camera,
        [2] V. Vicha. Experiments Using Pixel Detectors in Teaching Nuclear and Particle Physics, Czech Technical University, Prague 2017, ISBN 978-80-01-06108-4
        [3] XXI Jorge Andre Swieca Summer School on Experimental Nuclear Physics, IPEN, Sao Paulo, 2020, event web page,
        [4] Virtual IEEE NPSS Workshop on Applications of Radiation Instrumentation ("Jakarta"), event web page,
        [5] Virtual IEEE NPSS Workshop on Applications of Radiation Instrumentation in Dakar, event web page,

        Speaker: Dr HOLIK, Michael (IEAP CTU in Prague, UWB in Pilsen)
      • 12:00
        #10-254 EASY: Educationa Alibava System 20m

        EASY, a plug-and-play educational system, is portable, compact and a complete system for microstrip sensor characterization. Ideal for making basic or complex experiments. It is based on the Alibava System, largely used within the CERN community to test micro-strip detectors for particle experiments. The system can be configured to work with laser light or radioactive sources.
        The aim of this system is to illustrate students in the operation of a silicon strip detectors, in particular:

        • To observe the noise of a silicon strip detector as a function of bias voltage-
        • To observe the signal spectra due to a minimum ionising particle in a silicon detector and demonstrate the Landau distribution shape of collected charge.
        • To determinate the Charge Collection efficiency of a silicon detector and depletion voltage.
        • To observe charge sharing between strip and relate this to the position resolution of the detector.
        • To illustrate the structure of a typical micro-strip detector.
        • To compare the charge deposition of minimum ionization particles and gamma particles.
        • And, to calculate the laser penetration in silicon.

        The components of the EASY systems are the Control Unit and the Sensor Unit. The Control Unit is the heart of the system communicating with the Sensor Unit and the Computer software. It contains the Data Acquisition Control and it is also in charge of processing of the sensor data and trigger inputs. In addition, it contains an adjustable Hight Voltage unit for microstrip sensor bias, with voltage and current display and Includes the laser source. The Control Unit communicates with computer software via USB. The Sensor Unit accommodate a p on n silicon micro-strip sensor segmented in 128 strips.
        EASY comes with an activity book where the students, through 10 exercises, are introduced in the main concepts and functionalities of microstrip silicon detectors, used in the actual particle physic experiments. The book also provides a full description of the EASY device and the data Acquisition system.

        Speaker: GARCÍA, Carmen (Instituto de Física Corpuscular (CSIC-Universidad de Valencia))
      • 12:20
        Empty 20m
    • 12:40 14:00
      Lunch 1h 20m
    • 14:00 16:00
      01 Fundamental Physics

      Oral Session

      • 14:00
        #01-195 Measuring the delayed neutrons multiplicity and kinetic parameters for the thermal induced fission of $^{235}$U, $^{233}$U and $^{239}$Pu 20m

        Depending on the used library, nuclear data regarding the emission of delayed neutrons present significant discrepancies. Therefore, in the framework of the NEEDS/NACRE collaboration, the ALDEN (Average Lifetime of DElayed Neutrons) program was launched in 2018 by CEA (DES and DRF), CNRS (LPSC, CENBG, LPC Caen), ENSICAEN and Caen University. This program aims at measuring the ($\alpha_i$,$\lambda_i$) parameters describing the emission kinetics of the delayed neutrons as well as their average multiplicity $\nu_d$.
        To this end, a new experimental device, conceived using TRIPOLI4® Monte-Carlo simulation, was designed to measure the thermal induced fission of actinides on the PF1b cold neutron beam-line of the Institut Laue-Langevin (ILL, Grenoble, France). This device consists in a cylindrical matrix of polyethylene, shielded using a thin B4C layer, and at the centre of which a miniature fission chamber containing the actinide to study is placed. This chamber is irradiated for a period of time t$_\textrm{irr}$, then the beam is shut using a motion-controlled thermal neutrons screen. Delayed neutrons, measured over a period of time t, are thermalized in the polyethylene matrix and detected using 16 $^3$He proportional counters. They are positioned into three rings to ensure a detection efficiency as constant as possible between 0.1 and 1 MeV (energy range of the delayed neutrons). Irradiation cycles (t$_\textrm{irr}$+t) are repeated until a good counting statistics is obtained. All events are recorded using the NOMAD digital acquisition system developed by the ILL and the resulting curve is fitted using the CEA-developed CONRAD regression software.
        After detailing the experimental setup, this paper presents the adaptations that were made to it in order to allow studying $^{239}$Pu. Results obtained during the experimental campaign of June 2019, dedicated to the thermal induced fission of $^{235}$U, and first results of the experimental campaign of the beginning of 2021, dedicated to $^{239}$Pu and $^{233}$U, are presented.

      • 14:20
        #01-106 Calculation of spatial response of a collimated segmented HPGe detector for gamma emission tomography by MCNP simulations 20m

        Segmented HPGe (High Purity Germanium) detectors are commonplace in nuclear structure studies, where position and amount of energy deposited by the photon’s interactions inside the detector are fully reconstructed at every event using the pulse shape analysis methods. Recently a similar segmented HPGe detector concept in combination with a multi-slit collimator was proposed for gamma emission tomography. The feasibility of the proposed concept for use in gamma emission tomography and its performance was evaluated by a simulation study. A segmented HPGe can offer high energy resolution and high spatial resolution, and its ability to simultaneously interrogate the fuel object with multiple detection elements can also make it suitable for faster operation requirements in the post-irradiation examinations (PIE). The use of such segmented HPGe detectors can facilitate imaging of nuclear fuel rod internal features like fission product migration, pellet cladding interactions, rod bow and swelling, fuel fragmentation, relocation and dispersal in transient tests etc., as well as for analysis of miniaturized irradiation tested fuel samples.

        It is important to quantify the spatial resolution achievable by using the collimated segmented HPGe before the actual detector is manufactured and operated. The spatial response for a segmented HPGe detector concept was evaluated using simulation with Monte Carlo N-particle transport code MCNP6. Full detector and multi-slit collimator system was modelled and for the quantification of the spatial response, the Modulation Transfer Function (MTF)/Contrast Transfer Function (CTF) was chosen as the performance metric. These performance metrics were obtained through the calculation of the Line Spread Function (LSF) and Edge Spread Function (ESF) by analysing the simulation data. In addition, the proportionality between the activity inside the object and the as-measured profile with the detector collimator system was also examined through simulations of a ramp pattern/stair-step pattern.
        The results from the simulation study will be presented at the ANIMMA-2021 conference.

        Speaker: RATHORE, Vikram (Uppsala University)
      • 14:40
        #01-132 Autonomous Measurements Driven By Machine Learning 20m

        During the last reactor cycle in 2020, a combined team from the Institut Laue-Langevin and Berkeley National Lab has commissioned and tested a self-learning algorithm capable to perform autonomous measurements. For the first time the computer took control of the three axis neutron spectrometer ThALES, without any human intervention. The algorithm was able to explore the reciprocal space and fully reconstruct the signal without any prior knowledge of the physics case under study. Thanks to autonomous learning gpCAM - developed by Marcus Noack of the CAMERA team at Berkeley Lab – estimates the posterior mean and covariance and uses them in a function optimization to calculate the optimal next measurement point. The posterior is based on a prior Gaussian probability density function,
        which is repeatedly retrained on previously measured points. The main advantage of such an approach is clearly the possibility to drastically reduce the number of measurements with respect to a classical grid scan and therefore optimize the beam-time usage. In the present paper, the excellent results obtained will be discussed as well as the opportunities for further improve this technique.

        Speaker: MUTTI, Paolo (Institut Laue-Langevin)
      • 15:00
        #01-244 High-Energy Proton Induced Reactions Of Plutonium 20m

        Plutonium dioxide thin targets were irradiated by 660 MeV proton beam in Joint Institute for Nuclear Research at Dubna, Russia. Irradiated targets had been measured for several months by HPGe detectors, gamma lines of radioactive reaction and subsequent decay products were identified. Based on these data, cross-sections of residual nuclei production were determined. Obtained data are compared with calculation performed by different models using MCNP-6.2 and other codes. Results are very important for heavy actinides spallation models improvement, as these are very rarely validated by experimental data. Also comparison with data measured on thorium, uranium, and other targets is very beneficial, as well as comparison with data obtained by other groups and especially those taken by inverse kinematics method.

        Speaker: TANCHAK, Anhelina (Brno University of Technology)
      • 15:20
        #01-12 Novel neutron detector assembly based on SiPM readout to be coupled with the Active Target for SPES 20m

        The Active Target ATS (Active Target for SPES) is a new time-projection chamber designed for reaction and decay studies with nuclei far from stability. This kind of instrument, initially developed for high-energy physics, has found profitable applications in medium- and low-energy nuclear physics as shown by a successful series of experiments.
        The physics cases for the new-generation active target are related to the ongoing developments of facilities for radioactive ion beams. Thanks to its flexibility, this instrument will be capable of taking advantage of the most exotic beams which will become available at the SPES facility under construction at the Legnaro National Laboratories in Italy. Particular attention will also be paid to coupling it with ancillary detectors for both charged and neutral (gamma and neutrons) particles.
        In particular, in the present study, we will focus the attention on the neutron ancillary detectors.
        The proposed solution takes into account a compact device and the capability of the system to discriminate, by performing pulse shape analysis, neutrons from the high gamma background.
        A potential compact system aimed to discriminate neutrons from gamma ray events, using digital Pulse Shape Discrimination (PSD) techniques, can take advantage of recent improvements in silicon photomultiplier (SiPM) technology and the development of new plastic scintillators exhibiting the PSD phenomena.
        Our presentation will be focused on the first studies of SiPM coupled with CsI(Tl), EJ-276G and EJ-276 scintillators with different configurations, such as: different SiPM models (Advansid and Ketek), different area coverage (from 3 mm x 3 mm to 17 mm x 17 mm) and different preamplifiers.
        Moreover, we will show the comparison of PSD performances of plastic scintillators ranging from 25 mm to 70 mm diameter size, using the best SiPM readout configuration.

        Speaker: PINO ANDRADES, Felix Eduardo (University of Padova)
      • 15:40
        Empty 20m
    • 14:00 16:00
      04 Research Reactors and Particle Accelerators

      Oral Session

      • 14:00
        #04-26 JHR Irradiation Devices. Inspection Methods proposal 20m

        JHR irradiation test devices must undergo a periodic inspection every 40 months (French Regulation rule). The first step of inspection proposal for application of non-destructive techniques (NDT) for these devices is presented, through examining the possible methods and locations that could be utilized in the reactor.

        The selection of NDT methods depends mainly on the device geometry, material and access: main body of the device consists of two vertical nested tubes to be inspected. Once in use, due to deformations of thermal and radiation origins, these tubes are likely impossible to be separated (removal and re-assembly in a hot cell during inspection phase). Material is Zircaloy-4. The length of the tubes is about 3700 mm, with tubes diameter ranging from about 40 mm to 110 mm and thickness ranging from about 4 to 10 mm.
        Some narrow electron beam welds will have to be controlled according to French nuclear pressure equipment regulation rules: thus, the inspection must respect the requirements of RCC-MRx code.
        While ultrasonic volumetric inspection is selected for inspecting the tube bodies for checking no crack type of defects, Eddy current examination should be very efficient in finding surface defects.

        Simulations are performed with CIVA software platform in order to select the best conditions for Ultrasonic testing (UT) and Eddy current (EC) inspection. Suitable sensors are especially selected to optimize the propagation of ultrasonic waves in the tube. Ultrasonic pulse-waves with center frequencies ranging from 0.1-15 MHz are transmitted to detect internal flaws: the beam-defect interaction allows predicting the amplitude and the time of flight of various echoes: direct echo, corner effect, tip diffraction echo, etc. and also echoes scattered back by the geometry (backwall, entry surface and interior specular interfaces echoes) with acoustic mode conversions.

        The initial conclusions of this work are the following:
        - Two inspection methods are likely needed.
        - Eddy current is likely suitable for surface examination while volumetric ultrasonic inspection can be used for tube bodies.
        - Inspection under-water in a storage pool is likely the best option for location in the JHR.
        - Some automation and remote controls will be needed during the inspection process.

        The next step for experimental reference will be a critical step of the work to assess and qualify the inspection methods: a qualification phase will be launched, performing inspection of representative dedicated mock-ups made of Zircaloy-4 material. For ultrasonic reference, a representative mock-up of the cylinder with a weld and artificial defect(s) will be used. The mock-up testing will provide an experimental reference to CIVA software platform simulations.

        This study has been performed in collaboration between VTT Technical Research Centre of Finland Ltd and French CEA Alternative Energies and Atomic Energy Commission.

        Speaker: BAQUE, Francois (CEA)
      • 14:20
        #04-82 Design of an acoustic sensor for fission gas release characterization devoted to JHR environment measurements. 20m

        Among numerous research projects devoted to the improvement of the nuclear fuel behavior knowledge, the development of advanced instrumentation for in-pile experiments in Material Testing Reactor is of great interest. In the frame of JHR reactor, new requirements have arisen creating new constraints. This research is carried out within the framework of a long-standing partnership between CEA, CNRS and the University of Montpellier.

        An acoustic method was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature during the irradiation test was about 150 °C. In spite of the success of the experiment, it appeared necessary to optimize this type of probe especially the acoustic coupling. To overcome the problems encountered during the REMORA experience, we have developed thick film transducers produced by screen-printing process. They offered a wide range of possible application for the development of acoustic sensors and piezoelectric structure for harsh temperature environment measurements. We proposed a screen-printed modified Bismuth Titanate piezoelectric element on alumina substrate allowing acoustic measurements until high temperature.

        JHR environment imposes a device maximal size and a working temperature of up to 350°C. This drives design choices.In this paper we will focus on the mechanical design of the new sensor. This acoustic sensor is composed of an acoustic element for generation and detection of acoustic waves propagating into a cavity filled with gaz. We will detail the choice of piezoelectric materials, the thickness of the different layers, the cavity shapes, the electrical connections, the means of assembly of the different parts. Theoretical and experimental results will be given. The evolution of the impedance response and the piezoelectric parameters of screen printed piezoelectric structures on alumina will be studied.

        All that points will be discussed in term of acoustic sensor sensitivity versus dimensional constraints, in the case of a high temperature range working.

        Keywords: Material Testing Reactor / Acoustic sensor / Gaz release / In-pile experimentation / High temperature

        Speaker: Dr BAUDRY, Florian (IES, Univ Montpellier / CNRS, UMR 5214, F-34000, Montpellier, France)
      • 14:40
        #04-119 Study review of the CALORRE differential calorimeter: definition of designs for different nuclear environments 20m

        The high neutron and gamma fluxes as well as strong displacements per atom characterize conditions of materials testing reactors (MTRs). These research reactors constitute major support research facilities and allow experimental and real-condition studies of the behavior of fuels and inert materials (vessel, reflector, cladding, etc.) in an extreme radiation environment. These studies are important as they lead to progress on accelerated ageing of materials and/or advanced scenarios up to accidental conditions and consequently they bring data for safety issues, the lifetime of existing nuclear power plants and their advancements with new concepts. Therefore, new instrumentation is needed to measure online key parameters both before the experiments for the device design and during the experiments for the result interpretation.
        The construction of the Jules Horowitz Reactor, a new 100 MWth nominal power MTR with unequalled performance in Europe (high fast neutron flux of 5.5x10^14 n/(cm².s) (from 1 MeV) leading to a high accelerated ageing (up to 16 dpa/year) and a high nuclear absorbed dose rate (up to 20 W/g in aluminum)), initiated new collaborative research work. Since 2009 Aix-Marseille University and the CEA (within the framework of the joint laboratory LIMMEX - Laboratory for Instrumentation and Measurement in Extreme Environments) and its IN-CORE program - Instrumentation for Nuclear Radiations and Calorimetry online in REactor) have been developing a new research topic. More precisely, they have been focusing on innovation in instrumentation and advanced measurement methods for the quantification of key nuclear parameters such as neutron and photon fluxes and nuclear absorbed dose rate, also called nuclear heating rate. The online measurement of this latter quantity requires specific sensors: non-adiabatic calorimeters. With regard to the state of the art, two distinct sensors are used in MTRs: French differential calorimeters (CALMOS, CARMEN or CALORRE type) and European single-cell calorimeters (such as gamma thermometers or KAROLINA-type calorimeters). These two types of calorimeter allow the quantification of the nuclear absorbed dose rate thanks to temperature measurements and preliminary calibration under non-irradiation conditions from steady thermal states in the case of integrated heating elements or from transient thermal states in other cases.
        The paper will present a review of recent work carried out on a new differential calorimeter called CALORRE and patented by AMU and the CEA in 2015. The work allows the design and the characterization of several CALORRE configurations over a wide range of nuclear absorbed dose rate (up to 20 W/g).
        The first part of the paper will describe the first design of CALORRE calorimeter fabricated for a range up to 2 W/g and its first qualification under real conditions during an irradiation campaign inside the MARIA reactor (in November 2015).
        The second part will be dedicated to the characterization of the response of 6 new configurations of CALORRE calorimetric cell by means of a comprehensive approach coupling experimental, theoretical and numerical work from laboratory conditions to nuclear environments. The experimental metrological characteristics of the 6 configurations, in terms of sensitivity, linearity, range, reproducibility and response time, will be detailed. Their responses will be compared and analyzed thanks to a 1-D theoretical thermal model. Moreover, by means of a predictive model (based on a heat balance and calibration curves under laboratory conditions), their response under real conditions will be predicted. Then, thanks to this panel of configurations, the influence of various parameters such as the sensor geometry, its size and its structural material will be given. The advantages of this kind of calorimeter will be discussed and compared to other calorimeters (differential and single-cell calorimeters). Finally, 3-D thermal simulations allowing the optimization of the calorimeter assembly in term of compactness will be summarized for the chosen JHR configuration (for high nuclear absorbed dose rates) already fabricated, tested up to 60 W under laboratory conditions (electrical power range ten times higher than that of previous calibrations for common differential calorimeters).
        The last part will present new results obtained within the framework of a new research program called CALOR-I, funded by Aix-Marseille University, involving the Nuclear Reactor Laboratory of the MIT and the CEA (2020-2022), and focusing on the mapping of an in-core water loop of the MITR reactor in terms of nuclear heating rate by means of a new CALORRE calorimeter in particular. The criteria for the choice of this new configuration of CALORRE will be detailed (mass, size, temperature, sensitivity). Then, the parametrical study for the definition of a new CALORRE differential calorimeter assembly with three calorimetric cells and two kinds of sample material will be presented.

        Speaker: Dr VOLTE, Adrien (Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France)
      • 15:00
        #04-77 Assessment of irradiation performance in the Jules Horowitz Reactor (JHR) using the CARMEN measuring device 20m

        This paper deals with the assessment of irradiation conditions in various experimental positions in the Jules Horowitz Reactor core and reflector. For this purpose, the CARMEN device is foreseen to measure neutron and gamma flux and nuclear heating in the experimental channels of the JHR [1]. CARMEN is an axially mobile measuring system composed of two fission chambers sensitive to both fast and thermal neutrons, an ionisation chamber sensitive to gamma flux and an aluminium-based calorimeter for nuclear heating measurement. This allows for very detailed axial characterization of irradiation conditions in various radial experimental positions. Actually, the measurements performed with this device can serve as a preliminary assessment of “unperturbed irradiation conditions”, before the introduction of the real experimental device at that location. The present study aims at reproducing CARMEN measurements by simulation and connecting them to the irradiation conditions expected in MICA and OCCITANE devices - which are respectively devoted to in-core and out-core experiments for the irradiation of steel materials.
        The OCCITANE device provides spectral tailoring capabilities in order to be representative of the neutron spectrum in Reactor Pressure Vessels, and offers regulated temperature conditions by limiting nuclear heating, thanks to neutron and gamma screens. The present study evaluates the correlation between the measurements obtained by the CARMEN device, in unperturbed conditions, and specific irradiation characteristics occurring locally, inside the experimental loading of the device. OCCITANE capsule must be able to reproduce a subset of irradiation conditions identical to those of OSIRIS to pursue irradiation programs, in connection with already existing experimental data. As a preliminary step before examining JHR experimental positions, the present work first evaluates the transfer function between the measurements of nuclear heating in OSIRIS reactor - performed inside a water channel in the reflector, thanks to a graphite sample calorimeter [2] - and the real irradiation conditions finally obtained in the IRMA capsule.
        Neutron and photon transport calculations were carried out in critical mode by means of Monte-Carlo simulations, using the TRIPOLI-4® code. The advanced modelling of nuclear heating in the calorimeter requires the use of four-particle-type Monte Carlo simulations (involving neutron, photon, electron and positon). Regarding nuclear heating in steel materials, sensitivity studies were conducted to discriminate the respective contribution of gammas originating from the core and gammas resulting from the radiative capture in iron.
        [1] Lyoussi et al., “Advanced methodology and instrumentation for accurate on line measurements of neutron, photon and nuclear heating parameters inside Jules Horowitz MTR reactor”, in Proceedings of RRFM ’12, Prague, Czech Republic, Mar. 2012.
        [2] D. Fourmentel et al., “Nuclear Heating Measurements in Material Testing Reactor: A Comparison between a Differential Calorimeter and a Gamma Thermometer”, IEEE Transactions on Nuclear Science, Vol. 60, N°1, February 2013.

        Speaker: Dr BLANCHET, David (CEA Cadarache, DES/IRESNE/DER, F-13108 Saint-Paul-lez-Durance, France)
      • 15:20
        #04-196 Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor 20m

        The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C – 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C – 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 – July 2020). Special attention is focused on the estimated drift of thermocouples operating in the higher temperature regions. This is particularly important as an input to the final irradiation report for AGR-5/6/7, in which estimates are presented of the temperature distribution of the 570,000 particles irradiated.

        Speaker: PALMER, Joe (Idaho National Laboratory)
      • 15:40
        #04-161 Confocal chromatic sensor for displacement monitoring in research reactor 20m

        Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision [1]. Its operation principle relies on a wavelength encoding of the object position. The light emitted by a polychromatic source is transported by an optical fiber up to an optical system, called optical pen only composed of passive components and presenting a strong chromatism. That means that the various wavelengths will be focused at different positions by the optical system on the optical axis. The surface of the object under test will reflect all the wavelengths but only a small fraction of the reflected spectrum will be focused back on the cleaved end of the optical fiber and then will be able to be guided towards an optical analyzer.
        The change in the spectral properties of the analyzed signal provides an information about the distance of the interface located within the measuring range of the sensor. Historically, this class of sensors has been first developed in the 90’s by the company STIL based in the south of France.

        Of course, this sensor can only operate in a sufficiently transparent medium in the considered spectral domain. But it presents the advantage of being contactless which is a crucial advantage for some applications such as the one of measuring displacement in the core of nuclear research reactor: cladding swelling measurement for instance.

        The extreme environmental conditions encountered in a nuclear reactor -high temperature, high pressure, high fluency irradiations, strong vibrations and surrounding turbulent flow- can affect the performances of those optical system and mitigation techniques need to be put in place to optimize the sensor response for this specific environment. Another constraint concerns the small space available between the rods to be monitored, implying the challenge to conceive a very small sensor able to operate under these constraints.
        This development is made in the framework of a PHD at the Commissariat à l’Energie Atomique (CEA) -in collaboration with the Laboratoire Hubert Curien (LabHC) of the University of Saint-Etienne and also the company STIL.

        Among various scientific and technological locks, the first to be considered is the Radiation Induced Attenuation (RIA) phenomenon of the fiber and of the bulk optical components constituting the optical system. Indeed, glasses and fibers tends to darken when exposed at high neutron fluencies decreasing the light transmission efficiency. In the state of the art, it has already been shown that some spectral regions show huge RIA levels under high level of mixed gamma/neutron irradiation. In the visible, the absorption band peaking at 630 nm and caused by the Non-Bridging-Oxygen Hole Centers (NBOHCs) can strongly alter the sensing as well as the growing of the vibration band at 1380 nm of hydroxyls that could appear when NBOHCs are passivated by hydrogen. Basic mechanisms at the origin of the RIA are known [1, 2] but still only a few experimental data at the very high fluency associated with the application (neutron fluency about $10^{19} n/cm^{2}$ with gamma doses of more than one GGy, and also high temperature) are available, making difficult to really assess the RIA impact of the sensor performances.

        However, RIA is not the only scientific lock; elevated temperature can also affects the optical system. Indeed, if the sensor exploits the chromatism, related to the refractive index dependence to the wavelength, the temperature is also able to modify the refractive index causing a direct shift of the probing spectral range but also dilatation of the lenses and optical aberrations. Irradiations too can alter the well behaviour of the sensor -because of compaction. We also have to stay cautious about water under pressure with turbulent flow and vibrations.

        We will present the detailed specifications and a preliminary analysis of the constraints, with a special focus on the RIA at high fluency.

        [2] D. L. Griscom, « γ and fission-reactor effects on the visible-range transparency of aluminium-jacked, all-silica optical fibers », J. Appl. Phys., vol. 80, no.4, pp. 2142-2155, Aug. 1996.
        [3] S. Girard et al. “Radiation Effects on Silica-Based Fibers, « Recent Advances and Future Challenges», IEEE Trans. Nucl. Sci., vol 60, no. 3, pp. 2015-2036, June 2013.

        Speaker: AGOYAN, Marion (CEA)
    • 14:00 16:00
      07 Nuclear Fuel Cycle, Safeguards and Homeland Security

      Oral Session

      • 14:00
        #07-28 Characterization of nuclear waste packages by active photon interrogation with mobile or stationary systems based respectively on 7 or 9 MeV linacs 20m

        Decommissioning and dismantling (D&D) operations conducted in former nuclear facilities generate a large amount of nuclear waste packages. Characterization of the latter is an important step in the management process. Active photon interrogation, based on the photofission reaction, is a well-adapted method to characterize nuclear waste packages containing concrete matrices. Indeed, with such matrices, both passive methods and the active neutron interrogation method reach their limits whereas the high-energy photon beam produced thanks to a linear electron accelerator (linac) could enable to interrogate the center of a concrete matrix package. The photofission reaction is based on two steps. First, high-energy photons induce fission reactions on actinides such as uranium and plutonium isotopes. Secondly, prompt and delayed particles are emitted. The energy threshold of the photofission reaction is close to 6 MeV for most actinides (235U, 238U, 239Pu, etc.). In the frame of the MICADO project, which is part of the European Union’s Horizon 2020 research programme, the active photon interrogation method is assessed to characterize packages containing concrete matrices. Two different setups are studied: on the one hand a stationary system based on a 9 MeV linac; on the other hand a mobile system based on a 7 MeV linac. However, characterization of nuclear waste packages using a linac operated at an energy below 10 MeV is not straightforward. Undoubtedly, compared to 9 MeV, the challenge is even higher at 7 MeV, and will require the most effective measurement protocol and sensitive detection systems. Moreover, in both cases, the high-energy photon beam delivered by a linac would enable to gather more information on the package matrices thanks to the high-energy imaging technique. The aim of this paper is to present the recent developments in the field of active photon interrogation for nuclear waste package characterization with mobile or stationary systems based respectively on 7 or 9 MeV linacs. First, performances of the photofission technique with a 7 MeV linac will be assessed by Monte Carlo simulation using both PHITS and MCNP6 codes. Accuracy of cross-sections around the energy threshold of the photofission reaction will be discussed. Secondly, performances of the photofission technique with a 9 MeV linac will be assessed experimentally at the SAPHIR platform (CEA Paris-Saclay, France) using mock-up packages with different matrices including concrete and containing either uranium or plutonium samples. Delayed neutrons from photofission will be detected using 3He-filled gas proportional counters, and high-energy delayed gamma-rays will be detected using large size plastic scintillators. Comparison between simulation and experimental results will enable to validate the MCNP6 simulation model. The latter will then be used to evaluate performances obtained with the photofission technique and a 9 MeV linac on larger packages and more complex package geometries.

        Speaker: Dr SARI, Adrien (CEA List)
      • 14:20
        #07-245 Development of a miniaturized furnace for Scanning Electron Microscopy: Thermal modelling, manufacturing and tests 20m

        In the field of samples observations by scanning electron microscopy (SEM), imaging material samples at elevated temperatures has got an increase of interest during the last decades. Performing in-situ observations at high temperature facilitates the understanding of materials behaviour when they are used under severe environments [1]. It is for instance the case of superalloys used for turbine blade in jet engine which must withstand high mechanical loads and extremely high temperatures. Making high temperature SEM imaging possible requires specific instruments to be developed in order to overcome several constrains (thermoionic emission, detectors thermal sensitivity…). Heating samples at elevated temperature (1000°C and more) can be achieved by the means of miniaturized furnaces directly implemented in the SEM chamber. Most of them are built with a thin electrical resistance that uses Joule effect to warm up the sample. The accurate control and monitoring of the sample temperature needs a precise understanding of the micro-furnace thermal behaviour that is conditioned by the material components properties. In order to predict the micro-furnace capabilities, thermal modelling can be done by using finite element analysis (FEA). This tool is very helpful to design and to make an efficient micro-furnace but it requires a validation by testing a real furnace prototype. It is in this aim that a new micro-furnace prototype called FurnaSEM is manufactured and tested in controlled environments by the means of a specific bench test. The comparison of experimental data and numerical solutions is used to validate the prototype operability at high temperature.

        Numerical model of FurnaSEM
        Thermal modelling of the micro-furnace is ensured by a FEA industrial software: SolidWorks Flow Simulation which allows to take into account and to calculate all the heat transfer mechanisms operating around a solid component (conduction, convection and radiation) in a large variety of situations. Relevant numerical solution requires to consider the material properties such as thermal conductivity, specific heat capacity and radiative emissivity in a large temperature domain. The use of thermophysical properties database is crucial to adjust the parameters for the thermal modelling of solids. These properties are not available in litterature for the temperature range of interest. Consequently we have performed sensibility studies as a function of several parameters (emissivity and thermal contact resistance) to better estimate the micro-furnace thermal behaviour.
        Results obtained by numerical modelling are compared with experimental measurement achieved within a housemade bench test. This bench test is designed to monitor the furnace under vacuum while measuring the temperature at different points of interest using both thermocouples and thermal cameras. The experimental data are used to identify the best fitted parameter values. Finally the more accurate thermal model (i.e. set of parameters adjusted between the numerical model and experimental datasets) is defined to predict the 3D temperature fields. The full thermal map of the micro-furnace gives tips about its weaknesses. This will help us to further optimise the design and efficiency of the micro-furnace and to develop a new generation of micro-furnaces that works at higher temperature (1300-1450°C).

        The manufacturing of the miniaturized furnace was essentially 3D computer-aided. The prototype design is cylindrical and contains few elements that simplify its use. The sample is directly put on a sample holder that is a metallic cylinder piece. This cylinder is directly sustained by the heating system. The heating system contains a heating element that is embedded in a cylindrical hot body. Both pieces are assembled together mechanically. They are associated to the body of the furnace by a thin metallic wire which limits the heat conduction leaks. Thermal shield pieces enclose hot components and are cooled by conduction with a cold casing. The casing temperature is maintained below 50°C by a cooled fluid circulation. The chose of the material compositions were motivated by all these constrains. The heating element is made of a nickel superalloy that withstand high temperature constrains and presents a good resistance to high temperature oxidation. The shields and the sample carrier are made with stainless steel and the casing is made with copper. The micro-furnace temperature is monitored using two K-type thermocouples which are located near the heating element and the sample holder. The diameter of the furnace as well as its height are lower than 30mm. It can be easily placed in the SEM chamber. The compacity of the whole system allows to perform in-situ observations at high temperature using a relatively low working distance between the sample surface and the objective lens of the microscope (10 to 15 mm).

        Tests and SEM applications
        The bench test was specially designed to assess thermal capacities before operating inside the SEM. It contains a vacuum chamber instrumented with a thermal infrared camera which provides surface thermography measurements. The system can operate under a minimum pressure of 3.10-3 Pa and was mainly used for ex-situ thermal characterisation of the micro-furnace. The main advantage of this system is the temperature monitoring by independent measurement devices (i.e. thermocouple sensors, infrared camera and pyrometer) which help to enhance robustness of experiment data. Moreover, continuous thermal cycling of the furnace in repeatable conditions is a good way to monitor the system durability and to identify the degradation mechanisms.
        When it is implemented in the SEM chamber, the miniaturized furnace enables new imaging possibilities, mainly due to the flatness of the sample holder.
        • The furnace can be tilted in the -5° / +5° range while recording images in the SEM chamber. Then, the tilted image series are used to calculate 3D images of the sample surface and thus observe directly the surface rugosity variations as a function of temperature [2].
        • A high temperature backscattered electron (BSE) detector (provided Crytur company) can be used in combination with FurnaSEM. It allows to record BSE images at high temperature, i.e. to characterize continuously the sample chemical modifications during a heat treatment [3].
        • Working with a shorter working distance allows to record images at high temperature with a lower high voltage and with a high resolution.
        This was used in the field of nuclear materials: in-situ observations of uranium oxide microparticles sintering were achieved in order to understand better the physical and chemical mechanisms that operates during the fabrication of the nuclear fuel [4].

        Speaker: MENDONÇA, Jérome (CEA Marcoule; NewTec Scientific Nîmes)
      • 14:40
        #07-14 Boron Coated Straws imaging panel capability for neutron emission tomography for source localization inside radioactive drums 20m

        The evaluation of fissile mass inside radioactive waste drums is essential for radioactive waste management, nuclear safety and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by matrix effects.
        Therefore, an imaging panel proposed by Proportional Technologies Inc and composed of seven Boron Coated Straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated to plutonium or uranium position inside radioactive waste drums. In a previous work, a numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a 252Cf source.
        In the first section, the feasibility of neutron emission tomography by a setup composed of five extended imaging BCS panels is demonstrated by numerical Monte Carlo simulation.
        The second section details the experimental validation of the neutron emission tomography. Measurements are carried out with AmBe and 252Cf located inside an empty 118 L drum by rotating the BCS imaging panel around it. Afterwards, deconvolution algorithms are applied to provide 2D neutron source images for each angle. Finally, the 3D images are reconstructed using the RTK circular projection.
        The results demonstrate the capability of the BCS imaging to provide the 3D location, i.e. axial and radial positions of one and two neutron sources. Furthermore, the first tests with this passive neutron measurement system show a satisfactory 3D reconstruction for 252Cf and AmBe sources separated by 20 cm.
        Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated to plutonium or uranium localization in neutron measurements.
        Work is undergoing to assess the capability of this system for 118 L drums filled with organic and metallic matrices. Additionally, further prospects concern the performance of other deconvolution and reconstruction algorithms.

        Speaker: Dr ELEON, Cyrille (CEA, DES, IRESNE, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance, France)
      • 15:00
        #07-50 The ENEA Neutron Active Interrogation system: from prototype to the industrial implementation 20m

        In 2013, the ENEA Agency (the Italian National Agency for New Technologies, Energy, and Sustainable Economic Development) was involved in the EU FP7 EDEN (End-user DEmo for cbrNe) Project to designing and prototyping a new device for improving CBRNe resilience. The instrument is the Neutron Active Interrogation (NAI) system. Its purpose is the non-destructive inspection of samples suspected of containing fissile or fertile material, and explosives. The operating principle of NAI is an active neutron interrogation technique coupled with the Differential Die-Away time Analysis. Fast neutrons emitted by a neutron source, namely a portable Deuterium-Tritium neutron generator, thermalize in a moderator; if fissile material is present in the inspected sample, induced fission events on fissile material generate fission neutrons that can be detected by the NAI system. The difference between the die-away time of 14 MeV interrogation neutrons and prompt fission neutrons allows the fissile material to be detected. The original experimental setup was tested on the field during the live demo, open to the public, at the EDEN Project Demonstration occurred in September 2015 at ENEA Frascati Research Centre in Rome.

        Over the years, several upgrades have been identified and implemented on the initial prototype to improve its detection performances, both in terms of detection time and detectable amount of fissile. The current prototype is transportable and able to detect 2 g of $^{235}$U contained in a sample with maximum dimensions of about 8 cm x 31 cm x 28 cm in about 2 minutes.

        Currently, the upgrade program for taking NAI to a near-industrial production level, includes the following objectives:

        1. enlarging the neutron interrogation cavity for detecting the presence of fissile material in a sample comparable in size to that of a large luggage;

        2. reducing the minimum detectable quantity of fissile material;

        3. reducing the weight of the device for enhancing transportability;

        4. reducing the time needed for the detection of the dangerous material;

        5. quantitatively relating the recorded signal to the fissile mass present in the sample, independently from the matrix in which the fissile material is incorporated.

        The feasibility study which the improvement of NAI prototype is based on will be presented. Such a study is based on a thorough campaign of Monte Carlo simulations by means of MCNPX (Monte Carlo N-Particle eXtended).

        Speaker: Dr LEPORE, Luigi (ENEA)
      • 15:20
        #07-5 Plastic Scintillators With Tunable Decay Time 20m

        Due to their nature plastic scintillators are widely used in industrial applications as well as in research and scientific fields. In the field of nuclear physics, plastic scintillators show favourable properties such as possibility to be prepared in large volumes and different sizes and shapes; they are cheap, efficient and can be easily modified. Standard plastic scintillators are available with relatively short decay times usualy in the range of 0.7-10 ns and then only with long decay time of 285 ns. For certain applications, however, scintillators with unsual decay times may be required. The work presented proposes a new concept where the scintillation decay time can be finely chosen and tuned in a range starting at 2.5 ns and finishing around 90 ns. Plastic scintillator compositions were modified to include two molecules, both serving as primary fluorophores with strongly different photoluminescence decay times. The relative concentration of the two molecules then led to materials with different decay times. Secondary fluorophore (shifter) had only little impact on the material decay time and it was changed to obtain scintillation material emitting in blue (420 nm), green (500 nm) or red (560 nm) wavelength region. These materials were synthesized, characterized and results are presented.

        Speaker: Dr MATTHIEU, Hamel (CEA LIST)
      • 15:40
        #07-3 Neutron coincidence measurements and Monte Carlo modelling of waste drums containing reference nuclear material 20m

        Within the CHANCE project several non-destructive technique are being considered for the assay of waste bearing drums. Such techniques include calorimetry, gamma-ray spectroscopy and neutron coincidence counting. The aim is to reduce uncertainties on the inventory of radionuclides by combining the signatures from different techniques in the data analysis.
        In this framework, neutron coincidence measurements were carried out with two slab counters based on 3He detectors coupled to shift register electronics. Such a system consists of two identical slabs with 6 detectors each and is transportable, rather compact and flexible in terms of sizes and geometries that can be measured. With this system three 200 l drums containing certified reference nuclear material and different filling materials were measured. The certified nuclear material was in the form of 21 pellets of mixed oxide of U and Pu with a total mass of about 10.5 grams; in addition, a single pellet of about 10.5 grams was also available. The pellets could be placed in predefined positions within the drum in a reproducible way. The geometry and composition of the three drums was well characterized and consisted of Ethafoam, a mixture of Ethafoam, stainless steel and PVC, a mortar with an inner core of extruded polystyrene. The measurement setup was arranged such that the drum was placed between the two slab counters. The positions of the slab counters relative to the drum were accurately measured before each measurement, and a dedicated system was used to minimize the uncertainty on the detector positioning.
        The measurement data were analysed by applying the point model of Hage and the mass of nuclear material in the drum was determined from the totals, reals rate and radionuclide composition. Due to the fact that not all the point model conditions were met, we found that the point model overestimates the mass up to 80%. In addition, a Monte Carlo model of the measurement geometry was developed using the MCNP code. The model was used to determine a calibration factor between the reals rate and the mass of the sample. Measurements with a calibrated 252Cf source were used to verify the model. With a Monte Carlo based approach the mass of the mixed oxide pellets is slightly underestimated up to 10%. The results reveal the importance of an accurate background correction and of accounting for surrounding materials such as walls in the Monte Carlo model.

        Speaker: BORELLA, Alessandro
    • 14:00 16:00
      08 Decommissioning, Dismantling and Remote Handling

      Oral Session

      • 14:00
        #08-137 Advanced Sectorial Gamma Scanning of Radioactive Waste Drums with Spatial Reconstruction of Activity Distributions and Quantification of Uncertainties 20m

        The non-destructive assay based on radiation detection techniques is cost-effective measure to characterize radioactive waste and serves to verify the conformity with safety requirements for waste packages. Gamma scanning is a standard widespread measurement technique for the non-destructive assay of radioactive waste drums to determine the nuclide-specific activity content of the waste product. In the past decades, the pre-dominantly used method is segmented gamma scanning (SGS), which is based on simplifying assumption of a uniformly distributed activity and a homogeneous waste matrix. The simplification leads to large model-error for a non-uniformly distributed activity distribution which reduces the accuracy of the measurement. The deviation from the calibration condition of SGS adds to the measurement uncertainty which in turn increases the upper limit of the confidence interval used to quantify the conservative estimate of the activity content. In some cases, this leads to a significant overestimation of the true activity, the difference constituting ‘virtual activity’ which is not actually present in the waste drum. The virtual activity caused by the large measurement uncertainties results in an excessive and inefficient exhaustion of activity limits for waste packages resulting in higher costs for disposal.
        The companies AiNT and Mirion Technologies (Canberra) collaboratively developed a novel system named Advanced Sectorial Gamma Scanning (ASGS) for the gamma-spectrometric waste assay of waste drums with uniform and non-uniform activity distributions. The ASGS system is designed for commercial use allowing for

        • high throughput,
        • flexibility with respect to size of the waste drum,
        • a high dynamic range with respect to the gamma radiation fields of
          the measured waste drum,
        • a high degree of automation of the measurement and analysis

        The analysis method to reconstruct the spatial activity distribution of nuclides within the waste was implemented in a software module named ECIAD (Efficiency Calculation for Inhomogeneous Activity Distributions). The software undertakes the calculation of the photopeak-efficiencies based on a geometric modeling of the measurement configuration, the waste drum, and the active matrix. The activity distribution within the active matrix is approximated with a discrete model of sources on a grid within the drum. The best estimator for the total activity is determined from the measurement data using an optimization method and considering the Poisson statistics of the counting measurement, uncertainty contributions in the self-absorption of the active matrix as well as the model uncertainty in localization of the activity distribution. The ECIAD software reports the measurement results including the uncertainties and the characteristic limits in accordance with the ISO 11929. This feature is highly relevant for the safety assessment and qualification of radioactive waste packages where a confidence level of the activity inventory needs to be declared.
        An ASGS system was commissioned at the technical center of AiNT for experimental validation of the gamma scanning method and the analysis software. To this end, reference drums with three different matrix materials consisting of medium density fiberboard, polyethylene, and garnet sand with respective densities of $0.7$, $0.9$ and $2.3$ $g/cm^3$ were measured with calibration sources located at various locations within the matrix. In addition, measurements using the method of segmented gamma scanning were performed for a direct systematic comparison of the measurement accuracy. The validation of the ASGS system and the ECIAD software demonstrates the capability to determine the nuclide activities of a ‘hot-spot’ in $1.5$ hours total scanning time with a significantly lower total measurement uncertainty than the standard method of Segmented Gamma Scanning. This paper presents the ASGS system design, the analysis procedure, the treatment of uncertainties and the results from the experimental validation.

        Speaker: Dr DÜRR, Martin (Aachen Institute for Nuclear Training (AiNT))
      • 14:20
        #08-35 Passive isotope specific gamma ray tomography of a nuclear waste drum using a CeBr3 Compton camera 20m

        In various situations such as legacy wastes or optimising waste storage and decontamination costs, it would be very helpful to know the 3D distribution of various contaminants inside an object prior to decision making. Compton camera with their wide filed of view could allow such tomography in a cost effective way, but so far their spatial resolution and sensitivity was too limited for precision 3D imaging.
        We have been imaging a 220 l nuclear waste drum at ANDRA using a temporal imaging CeBr3 camera. Three views of the drum defining an orthogonal trihedral in the referential of the drum center were acquired using a mechanical fixture in order to insure precise relative position and orientation between the views. The drum contained a total activity of 14 MBq with 3 main isotopes detected in the Compton spectrum: 60Co, 110Ag and 137Cs. Each of the 3 acquisitions lasted 20 minutes. As the object imaged is relatively small and low density, one view seems sufficient for each trihedral axis.
        An image reconstruction algorithm has been developed based on cone intersection in 3D space. The image was then treated using LM/MLEM to increase contrast. We then create images of activity distribution for each of the isotopes detected. For this drum, we had also a 2D X-ray image showing structures inside the object. We have thus cut our 3D gamma images in planar slices each 2 cm and project them as a movie with the x-ray image as a background in order to understand the organisation of contamination inside the drum.

        Speaker: Dr HMISSI, Mohamed-Zied (Damavan Imaging)
      • 14:40
        #08-165 Statistical evaluation of radioactive waste packages 20m

        For the characterisation of radioactive waste packages, like 200l barrels, preferably non-destructive methods like segmented gamma scanning (SGS), transmission measurement or computer tomography are used. Segmented gamma scanning allows to identify gamma-emitting isotopes and their distribution in the vessel. To determine the corresponding activities additional information on the absorption of the waste matrix is needed. In addition, supplemental information on the waste like its chemical composition, materials and nuclide vectors are frequently available.
        A unified method allowing to combine data from measurements and a-priory knowledge on each waste package giving a precise result as well as robust error estimates is currently not available. A method allowing to achieve this combination is Bayesian statistics.
        The application of the Bayesian approach typically requires solve integrals (marginalize), whose dimensions are determined by the number of parameters. Depending on the size of the problem, various methods to accomplish this are described in literature. Common to all of them is the need of a model to predict the outcome of the measurement given the parameters of interest like the radioactive inventory. In case of large number of parameters, the Marcov-Chain-Monte-Carlo (MCMC) algorithm is the method of choice.
        The talk will focus on the modelling of the SGS measurement process using information of the matrix derived from computer tomography. In a first step the CT-data is transferred into a 3d-voxel model. The loss of photons due to absorption and scattering on their path to the detector is calculated using an adapted Box-Intersect algorithm to get the individual contribution of each voxel. In addition, the solid angle and attenuation due to the collimator are incorporated using geometric considerations. The approach allows to assign a weight to each voxel allowing to predict the photon count rate given the activity in the voxel.
        The validity of the procedure has been tested using various vessels with well-known matrix and calibration standards. The results of these comparisons will be presented.
        The model predicts the count rate in each segment of the scan given the radioactive inventory in each voxel, one of the key elements to run the Marcov-Chain-Monte-Carlo algorithm. Besides the model connecting the observed quantities with the parameters of interest. Further considerations are necessary to model the activity. In a straightforward manner each voxel is allowed to carry an activity determined by prior knowledge. We demonstrate that this assumption delivers good performance for a subset of problems. However once the measurement don’t constrain the spatial distribution of the radioactive inventory properly convergence problems arise. To mitigate them the activity distribution is modeled using a compositional distribution consisting of point sources and various forms of homogeneous sources. This effectively reduces the dimensionality of the posterior distribution and improves the convergence of the algorithm.

        Speaker: RUMMEL, Stefan (TUM / RCM)
      • 15:00
        #08-53 The Euratom project MICADO and its innovative characterization process of the Nuclear Waste Packages 20m

        In 2019 the MICADO (Measurement and Instrumentation for Cleaning And Decommissioning Operations) project started under the H2020 Euratom call aiming to become a reference in the nuclear waste characterisation field. It can be possible by the system under development and to the partners involved representing some of the most important actors of the sector. They are covering the roles of academia, industry, and the end-users.
        All over the world the nuclear waste management sector is always considered by the population a scaring problem and it is always part of the public debate, mainly due to complexity and costs coming from the storage and management of the waste produced. It is important to underlying that this waste is not only produced by the nuclear power industry, but also by hospitals, universities and non-nuclear industries like oil and gas companies with the production of NORM (Naturally Occurring Radioactive Materials) and TENORM (Technologically Enhanced Naturally Occurring Radioactive Materials) waste. Independently from their origin, the main concern is the radiation emission, which makes it a particular hazard for human health and the environment. It must therefore be managed with special care, from production to final disposal.
        The situation gets even more complicated when considering country dependent legislation, storage and final disposals sites. This means dealing with different definition of waste categories and activities (i.e. free release), the necessity to use multiple radiological sensors with not compatible outputs inducing the need to reprocess the characterization at each site, to analyse a large amount of off-line data, process manual reports of the operators, etc.
        MICADO wants to show a way to improve the characterization of nuclear waste packages and change current manual operations applying a digital analysis procedure, waste-type dependent, and combining information from different detectors to better qualify the waste package under investigation.
        MICADO has established a characterization process, data analysis and information storage able to cope with different types of waste activities (VLLW, LLW, ILW, legacy waste), types (Metallic & concrete filling) and drum dimensions.
        This is done with a toolbox with up to date and novel gamma and neutron detection technologies (two other abstracts submitted on the monitoring grid and on the photofission system), working as modular elements, and a digital software platform used as a base for the digitalization of waste information and the off-line analysis for the uncertainties assessment. The procedure was defined to reduce the measurement time in each step and being able to select the required detection technology avoiding multiple identical measurements of the same waste package. The combined data analysis, like other big-data studies, fuses different measurements results to extract information not available by the individual system and reduces the individual uncertainty. This aspect is extremely important as a possible solution to the problem of having a satisfying and reliable categorization of the waste package activity of complex cases as high density waste drums or the request for the free release. The software platform also aims at reducing operator costs and improving the ALARA principle, decreasing the time spent on field by the operators and promise a simple and easy data control on historical basis of all the already characterized waste packages.
        The presentation will start with the project overview but will focus on the status of the overall technologies after one year, from its start and tests performed. It will also be given a look to the future steps toward the end of the project and more till the organization of the final demonstration.

        Speaker: MORICHI, Massimo (CAEN S.p.A.,Via Vetraia 11, Viarregio 55049, Italy)
      • 15:20
        Empty 40m
    • 16:00 16:20
      Coffee Break 20m
    • 16:20 16:45
      03 Fusion Diagnostics and Technology: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #03-186 Towards the Evaluation of Radiation Tolerance of Scientific Cameras for ITER Diagnostics 5m

        ITER Diagnostic systems will use about 200 scientific-grade UV, visible and IR cameras for real time monitoring of the plasma and of the in-vessel components during operation. These cameras will be exposed to nuclear radiation during the deuterium-tritium operation phase. Impacted systems rely currently on the use of commercial-off-the-shelf cameras not designed to withstand ITER harsh environment. Among the different mitigation strategies to cope with radiation effects, radiation mitigation plans are based on a combination of countermeasures including camera relocation, local shielding, periodical replacement, assessment of radiation tolerance and adapted system operational procedures. In this paper, the different options are reviewed with an emphasis on the methodology for the development of radiation test plan for the evaluation of radiation-tolerance of semiconductor imagers. Potential camera standardization are also evaluated towards the need of radiation assured electronics.

        Speakers: Dr MARTIN, Vincent (Bertin Technologies) , Dr REICHLE, Roger (ITER Organization)
      • 16:25
        #03-162 Data acquisition system prototype of the ITER diagnostic Divertor Neutron Flux Monitor testing at research nuclear facilities 5m

        The Divertor Neutron Flux Monitor is the one of the neutron diagnostics of the International Thermonuclear Experimental Reactor ITER. This diagnostic consists of three subsystems. Each subsystem consists of the detector module includes six fission chambers and the data acquisition system. Detector modules are placed on the inner shell of ITER Vacuum Vessel under the Divertor Cassette. Expected neutron flux at the Divertor Neutron Flux Monitor detector modules position is from 1E3 to 1E13 n sm-2 s-1. The multidetector module is used to solve the task of the neutron flux measurements at ten orders of magnitude with 1 ms of time resolution. Divertor Neutron Flux Monitor module consists of two detector units. The first detector unit concludes three-section fission chamber with 235U as the fissile material and the second concludes three-section fission chambers with 238U. To demonstrate the possibility of the neutron flux measuring in a wide dynamic range using such a detector module and to evaluate the characteristics of the data acquisition system prototype a number of tests were carried out under conditions of the intense neutron radiation. Detector units, which are similar to those planned for use on the ITER, and the mobile version of the prototype were used for the tests. The tests were carried out at the plasma neutron diagnostic stand based on the NG-24M neutron generator with yield of 1E11 n s-1, and at the IBR-2 pulsed reactor of the Joint Institute for Nuclear Research with neutron flux of 1E9 - 1E12 n cm-2 s-1 depends on monitor position at peak power 1.6-1.85 MW. During tests at the plasma neutron diagnostic stand the mobile version of the data acquisition system prototype functionality was checked. The averaged pulse shapes and pulse-height spectra of all fission chambers were recorded. The data for the calibration of the data acquisition and processing system measuring channels were obtained. During the tests at the IBR-2 pulsed reactor the signals from the measuring channels of the subsystem were recorded while the neutron flux was changed. This paper shows the test results and discusses techniques and methods of the DNFM neutron diagnostic subsystem calibration.

        Speaker: MARTAZOV, Evgeny (National Research Nuclear University MEPhI)
      • 16:30
        #03-139 Design of Real-Time Electronics System for Two-Dimensional Hard X-Ray Diagnostics with Intensity Imaging and Energy Spectrum in EAST 5m

        Hard X-ray diagnosis is the most important diagnostic in Experimental Advanced Superconducting Tokamak (EAST) for studying low-hybrid wave physics and fast electrons, which helps to improve the operating performance and guarantee the operation safety of EAST Tokamak. The Hard X-Ray diagnostics used 16×16-channels pixel array Cadmium Zinc Telluride (CZT) detector which has the features of high resolution, large absorption coefficient, high detection efficiency, small size, and can work under normal temperature conditions. A charge-sensitive pre-amplifier card is designed which is magnetic shielded and easy to maintain, with low noise and low signal crosstalk. The card has 32 channels and the actual conversion gain can reach up to 1mv/fc. Then, a compacted pre-amplifiers module is built which is composed of 8 preamplifier cards and one main-board (transform signal from detector to pre-amplifier). Moreover, the multi-channel high precision data acquisition and signal process system based on FPGA (made in China) and PXIe bus is designed, which solves the problem of time synchronization during the acquisition of multiple boards. The direct memory access(DMA) mode is used for hardware programming to achieve high-speed data transmission. Meanwhile, a software driver has been developed based on the Linux platform, and through the QT framework. The real-time display of the two-dimensional intensity distribution of hard X-ray radiation over time and the three-dimensional curve of the pixel signal spectrum over time have been realized, and the data is stored quickly to facilitate further offline analysis and processing. The electronics system has been tested and the results show that it has reached the design performance.

        Speakers: Mr CAO, Hongrui (Institute of Plasma Physics, Hefei Institutes of Physical Science) , Mr WU, Guo Bin (Hefei University of Technology)
      • 16:35
        #03-110 Development of a real-time signal processing unit for diamond detectors of ITER Vertical Neutron Camera 5m

        Diamond detectors are going to be used in ITER neutron diagnostics, including the Vertical Neutron Camera (VNC). They are meant for neutron flux measurements and measurements of the energy spectrum of fast neutrons.
        One of the main functions of VNC is measurement of the neutron source profile. In accordance with ITER requirements VNC has to calculate an updated neutron source profile each millisecond (at a frequency of 1kHz), as well as provide neutron spectrum updates each 100 ms. The required measurement accuracy is 10%. This necessitates development of a high-speed real-time signal processing unit.
        Nuclear reactions between fast neutrons and carbon atom nuclei lead to ionization in the diamond detector’s volume. If voltage is applied to the crystal, a pulse of current is generated on each interaction between a neutron and a carbon atom nucleus. The amplitude of the resulting signal depends on the amount of energy $E_p$ absorbed on each interaction. Typical width at base for pulses registered at the ADC input is around 30 ns. Due to this we have to use an ADC with a sample rate of at least 500 MHz.
        This study describes the development process of a diamond detector signal processing unit, based on a heterogeneous computing device, consisting of an industrial computer and a fast ADC coupled to a high-speed I/O board with data processing capabilities provided by an on-board reconfigurable FPGA. In it we describe techniques used for signal filtration and pulse detection.
        An algorithm has been developed to identify pulses in real-time and measure their parameters: amplitude, width at base. It is able to reject pulses based on their duration to avoid pile-ups, which has an unpredictable influence on the resulting statistics. This algorithm is able to mitigate a constant non-zero shift of the data baseline (from which amplitudes of pulses are measured), as well as baseline fluctuations.
        This algorithm has been successfully applied in an experiment on the NG24M neutron generator. Diamond detector measurements have been performed in a neutron field with a flux of 14 MeV neutrons of up to $10^9$ $\frac{n}{cm^2 \cdot s}$. In this experiment the pulse count-rate of the diamond detector exceeded 300 kHz. Results of this experiment show that this algorithm is suitable for use in the VNC diagnostics.
        This work is being carried out in accordance with the state contract dated April 21, 2020 No. H.4a. “Development, pilot production, testing and preparation for the supply of special equipment to ensure the fulfillment of Russia’s obligations under the ITER project in 2020”.

        Speaker: Mr ZHURAVLEV, Michael (Institution "Project Center ITER")
      • 16:40
        #03-96 Implementation, enhancement and positioning performance of the Very High spatial Resolution IR thermography diagnostic in WEST Tokamak 5m

        The Very High spatial Resolution (VHR) infrared thermography diagnostic developed by CEA-IRFM is installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocks components as foreseen for the International Thermonuclear Experimental Reactor (ITER) Divertor. Purposes of such apparatus are to study the evolution of components introduced in the tokamak, the influence of the shaping of these components submitted to high heat load deposition, the behavior of the leading edges regarding the assembling tolerances between adjacent monoblocks and finally contribute to the specification assessment of the ITER divertor units. It is a major diagnostic to investigate the behavior of Plasma Facing Units (PFU) during experimentation thanks to its very high spatial resolution (0.1 mm / pixel), movable field of view (FoV) and wide-range temperature measurement. This diagnostic was operational during the 2019 WEST experimentation and its design has been enhanced for the 2020 WEST experimentation campaign.
        In order to monitor 28x12mm monoblocks temperature, the VHR IR diagnostic is able to move its narrow FoV (64x51mm) to specific targets on the observable divertor (360×420mm). Position and focus steering is achieved through motor-controlled mirrors and lens operated by a dedicated Labview application on a remote unit. The distance separating the first mirror with the target area induces large ratio distance over angle that must be taken into account to avoid detrimental misplacements. Positioning mirror maneuver between pulses under a 10-6Pa vacuum, a 1T magnetic field, at 70°C, and can withstand thermal radiation and disruptions during pulses. Motors are outside of the VV, operate mirrors through stainless steel cables and pulleys inside the VV, inducing a hysteresis that has to be dealt with. Repeatability, hysteresis, mechanical components reliability are critical parameters to provide accurate positioning (<1mm).The tuning and calibration of these actuators in accordance with WEST environment is crucial to guarantee the best position and optical performances. It is calibrated on a test bench at CEA-IRFM and inside WEST tokamak to insure valid positioning throughout experimentations.
        This paper gives the implementation details and FoV positioning performance comparison between the two mechanical designs and software implementations for the 2019 and 2020 campaign.

        Speaker: Mr DUBUS, Léo (CEA)
    • 16:20 16:55
      04 Research Reactors and Particle Accelerators: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #04-39 Performance assessment of compact calorimeters in the forward region of future linear colliders 5m

        The FCAL collaboration is preparing large-scale prototypes of special calorimeters to be used in the very forward region at future electron positron colliders. The forward region sets challenging requirements on several detector parameters, such as detector compactness, radiation hardness, or readout ASICs parameters. Prototype detector planes assembled with dedicated FE and ADC ASICs for LumiCal and BeamCal have been built. We present R&D activities focused on the development of prototype detectors able to cope with these requirements.
        The results of their performance following tests in an electron beam are reported. High statistics data were used to perform sensor alignment, and to measure the longitudinal and transversal shower development in the compact calorimeter prototype. Besides, Geant4 MC simulations were done and compared to the data.

        Speaker: GHENESCU, Veta (Institute of Space Science)
      • 16:25
        #04-29 Development of an LVDT Conditioning unit for use on LVDTs in a research reactor 5m

        Research into materials and fuels for nuclear power reactors is carried out in so-called research reactors where various types of fuels and materials can be monitored online by in-core instruments. At the Halden reactor in Norway, online measurements such as cladding elongation, inner fuel rod pressure, fuel swelling, material creep and stress relaxation were based on Linear Variable Displacement transducers (LVDT). These LVDTs require custom made conditioning units, with some particular requirements such as low driving frequency, ac constant current through the primary coil and high stability over a long period of time. In relation to the closure of the Halden reactor (June 2018) and the ensuing reorganizations at the Institute for Energy technology (IFE), the required LVDT conditioning units are no longer manufactured by IFE in Halden and the production of the LVDTs is almost terminated. Therefore, a project was started at SCK CEN to develop both LVDTs and the required conditioning units, having similar characteristics as those used at Halden. While the development of the LVDTs is still ongoing, the development and testing of the conditioning unit for a 4-wire LVDT has been completely finished and tested. In contrast to the conditioning unit developed at Halden, the system developed by SCK CEN is based on a digital system, using a commercially available chip by TEXAS Instruments. The device implements a digital demodulation algorithm that not only extracts the amplitude but also the phase of the signal with respect to the primary excitation. Additional circuitry was needed to reverse the polarity of the DC output signal at the zero crossing of the LVDT, based on the phase of the signal.
        The system is set to drive the primary coil of the LVDT in constant current mode, at an operating frequency of 1 kHz. This frequency is a factor 2.5 higher than the frequency which was used at Halden while still being compatible with long signal cables and in-core use. As a result, for the same LVDT, the sensitivity of the LVDT is increased by a factor 2.5 as well. The external circuitry allows obtaining 50 mA current over a load resistance of 400 Ohm. This high resistance is required because of the combined resistance of coils and signal cables at high temperature. The overall performance was tested on existing LVDTs produced at Halden and was found to be excellent.
        The present system is made to fit in a 19-inch rack, containing 6 independent LVDT conditioning units, each with their own frequency and current generator. The unit has been CE and EMC approved.

        Speaker: Dr VAN NIEUWENHOVE, Rudi (SCK CEN)
      • 16:30
        #04-25 Characterization of neutron emission during pulse mode of low output electronic neutron generator 5m

        Portable electronic neutron generators are an established technology that can be used for many lab and field applications. Offering significant regulatory and operational benefits over radionuclide neutron sources, they start to be more widely used at universities and research centres. Beside the established industrial applications (e.g., in logging or mining industry), applications are being developed for fields including security, safeguards, or waste characterization. They are becoming more used for E&T in nuclear engineering, nuclear analytical techniques, or nuclear related fields.
        Electronic neutron generators can be utilized for many analytical techniques (including NAA, PGNAA, DNC, DDA). In many cases, the pulsed operation of the generator is used to improve the efficiency of the applied methods. Neutron generators can provide information on neutron production via a TTL signal that can be used for triggering the data acquisition system. However, to improve the interpretation of the measured data, the knowledge on the time profile of neutron production within the pulse duration is important. This can be achieved experimentally by a proper neutron detection system. The response of the neutron detection system is formed by neutrons of different energies for detection systems designed to measure thermal, epithermal, and fast neutrons. For various applications, specific frequencies and pulse durations can be needed. Thus, various frequencies may require or enable various approaches to efficiently relate the neutron detector response to the neutron emission by the generator within the required time resolution. Monte-Carlo particle transport codes are efficient tools to simulate it.
        The paper deals with the characterisation of the neutron output profile during neutron pulsing of a neutron generator. First, it provides a validation of time-dependent calculations in MCNP Monte Carlo code showing the role of several model simplifications on the achieved results. Second, it compares the applicability of thermal, epithermal and fast neutron detection systems for pulse characterisation. The study is performed for two cases: first, the generator and detector are placed in a moderated environment; second, the equipment is placed in a free space. Calculations are performed in MCNP6 code. The study includes experimental data measured with the D-D P-385 Thermo Fisher Scientific neutron generator with a maximal neutron output of 7 x 10^6 n/s for several frequencies.

        Speaker: Dr BILY, Tomas (Czech Technical University in Prague)
      • 16:35
        #04-21 Characterization of the X-ray spectrum of a linear electron accelerator 5m

        The Nuclear Measurement Laboratory (LMN) at CEA Cadarache in France uses high-energy electron linear accelerators, LINAC (9-20 MeV), to characterize nuclear waste drums or study corium-water interaction. These high-energy accelerators allow to explore new examination modalities, such as active photon interrogation or dual-energy CT to scan large concrete objects with diameters up to 140 cm. These techniques require precise awareness of the photon spectrum emitted by the LINAC. However, direct measure of this photon energy spectrum (for example, with a hyper-pure germanium spectrometer) can not be achieve because of the accelerator pulses causing detector saturation. During the last few years, a large number of indirect methods have been developed. Some employ Monte-Carlo simulation to determine precisely without experimental protocol the accelerator photon energy spectrum. Nevertheless, these methods require a precise knowledge of the accelerator characteristics including electron beam energy spectrum, difficult to assess accurately experimentally or numerically. Other indirect methods are based on an experimental protocol using energy spectrum from the spectroscopy of Compton-scattered photons, or transmission measurements through different thicknesses of a well known material. The latter is the simplest indirect method from an experimental point of view because it can be set up easily and accurately using an ionization chamber as well as an appropriate screen. The obtained transmission curve depends on the photon energy spectrum which can be estimated using inverse models.

        In this paper, we present the development of a numerical model to determine energy spectrum from attenuation curve via transmission measurements which combines two types of inverse models: a continue model and a discrete model. The continue model consist in estimating spectrum shape via a continue mathematical function, in our model we use a log-normal function. Photon energy spectrum is first estimated from discrete log-normal equation which gives us the initial spectrum shape and the associated transmission curve. Then, the discrete model consist in tweaking spectrum values channel by channel with an iterative process to minimize cost function between analytical transmission curve of the reconstructed spectrum and the measured curve.

        We validate this tool using a test spectrum and its transmission curve obtained via Monte-Carlo simulation (MCNP6). This qualification allowed us to determine its sensitivity (signal-to-noise ratio, SNR) in order to have a good convergence. We show that if the SNR is less than 4%, we have a good estimation of the photon energy spectrum. Then, it was experimentally tested with a transmission curve obtained at the laboratory.

        Speaker: MAULIN, Maëva (CEA, DES, IRESNE, Nuclear Measurement Laboratory)
      • 16:40
        #04-18 C/O logging by using the associated alpha particle method: Proof of principle 5m

        The possibility of using the fast neutron beam with the associated alpha particle technique (APT) by the specially designed tool for Carbon-to-Oxygen (C/O) nuclear well logging was investigated. Measurements were done to show the influence of borehole fluid, iron casing, and the cement around the casing on the C/O analysis. Tests were conducted by using fast neutrons, which is the conventional approach for C/O analysis in well logging, and by using fast neutrons with APT. Analysed samples were made from quartz sand and graphite powder with the different C/O ratio characteristics for the oil-containing formation around the borehole. Diesel fuel was used as a simulant of the borehole fluid. Monte Carlo (MC) simulations were conducted and the MC results were compared to the experimental data. In both tests, the linear relation between the measured and real C/O ratios was obtained, but in the case in which the APT technique was not used, the linear relation was significantly influenced by the presence of the borehole fluid. The use of the APT technique resulted in a significant reduction of a background signal (coming mainly from the borehole fluid) in comparison to the conventional technique.

        Speaker: Mr BATUR, Josip (Rudjer Boskovic Institute)
      • 16:45
        #04-13 Experimental study of ISHTAR thermostatic irradiation device for the MARIA research reactor 5m

        Materials and core components for the next generation power reactors technologies require testing that can be performed in existing research reactors. Such experiments employ devices dedicated to reflect the relevant thermal and neutron parameters simulating conditions present in, for example, but not limited to, HTR reactors. A novel thermostatic irradiation device named ISHTAR (Irradiation System for High-Temperature Reactors) has been designed and constructed in the MARIA research reactor. Its mission is to enable irradiation of samples in controlled, homogeneous temperature field reaching 1000 °C and inert gas atmosphere. The high temperature is achieved by a combination of electric and gamma heating, together with carefully designed thermal insulation. Additionally samples holder made of graphite with high thermal conductivity enables the temperature homogenization in all directions. Device will be placed inside the Beryllium matrix of MARIA core and cooled with forced circulation of water from the reactor pool loop. This paper presents the outcome of experiments conducted with the rig prototype in external hydraulic mock-up of the MARIA reactor irradiation channel. The results have proved that the desired conditions for irradiation of the samples were achieved and their comparison against computational data has shown the adequacy of the design process. Finally, the loss of flow scenario was tested in protected and unprotected conditions (meaning with and without the safety system based on temperature feedback), proving the operational safety of the ISHTAR design. Experimental results will be used in the future to validate the numerical models (two and three dimensional) of the irradiation rig, providing an improved understanding of free convection and radiation phenomena modeling.

        Speaker: LIPKA, Maciej (National Centre for Nuclear Research)
      • 16:50
        #04-8 Investigation of pulsed fields generated by pulsed ionizing radiation generating equipment, testing of radiation measuring detectors, detection systems and simulations supporting testing 5m

        Devices that generate ionizing radiation that operate with a short pulse time are increasingly used in industry, healthcare, and scientific research. In the case of pulsed operation, these devices create pulsed ionizing fields, which have different properties compared to those experienced at stationary sources. Correct measurement of pulsed spaces is a challenge for manufacturers and users, as well as for the organizations involved in licensing, and should be reviewed.
        In pulsed space, currently used radiation measuring detectors typically measure lower-than-true values ​​because they typically do not detect a pulse lasting less than one second or are unable to properly follow sudden, magnitude changes over a short period of time. In the case of short-pulse radiation, it is advisable to use special detectors with a sufficiently good time resolution and a wide measuring range, since in extreme cases the characteristics of the pulsed fields reach or even exceed the operating limits of the detectors.
        In recent years, the EK SBL has set up a test track where various detectors could be tested first for stationary spaces and then for dynamic spaces. In 2019, EK BL started testing pulsed fields with equipment that generates ionizing radiation but does not contain radioactive material that produces a pulsed field.
        The availability and testing period of the equipment was limited, so the SBL plans to generate a pulsed-field using a constant gamma beam source, in which the shielded-collimated beam is generated using a gamma chopper. With the help of the device to be created in this way, the available time required for the examination of the SBL pulsed space will increase, with the help of which we will be able to perform several examinations and tests in the future. The pulsed-field generated by the device is to be measured optically using a visible light beam and a brightness sensing device. Thus, we can measure the beam characteristic by measurement (frequency, brightness change over time, pulse length, pulse time). In addition to the measurements, we can theoretically determine and then compare the measured and calculated results based on the data of the rotating mechanics. As a result of these tests, the EC will be able to produce pulsed spaces known for their SBL characteristics.
        With the help of the new device, there are more testing possibilities than before, in which we want to test several detectors for several characteristics, and then we formulate a recommendation based on the results, which detectors follow well on a given characteristic and are close to the expected value. measures, or if it follows the trend but does not show real value, we try to apply a correction function, with the help of which the device shows the expected value using the characteristic function later when measuring the pulsed space with a given characteristic.

        Speakers: BODOR , Károly, GULYÁS, Attila
    • 16:20 17:25
      09 Environmental and Medical Sciences: Parallel Intensive Oral Sessions

      Oral Session

      • 16:20
        #09-258 A prototype of pCT scanner: first tests 5m

        Proton therapy is a cancer treatment technique that allows for a more selective application of dose to tumors in comparison with conventional radiotherapy with X- or $\gamma$-rays. This is due to the fact that higher dose is concentrated in the region where the protons stop, whereas far less dose is deposited in the traversed tissue. In this context, a system of imaging and dose verification to define effective and accurate treatment plans and to guarantee the correct location of the applied dose is mandatory.
        However, the treatment planning in proton therapy facilities is guided via X-ray computed tomography (X-ray CT) images. This requires the conversion “a posteriori” of the map of Hounsfield Units (HU) obtained from X-ray CT to Relative Stopping Powers (RSP) useful for proton therapy treatment plans [1]. This conversion induces a large uncertainty in the range of the protons (up to 5% in the abdomen and up to 11% in the head) [2-4]. Meanwhile, treatment plans made via proton-CT (pCT) images will offer more accurate estimations of proton ranges with an uncertainty below 1% and a better control of the treatment [1].
        With this purpose, we are building a prototype for pCT scanner using particle detectors extensively used in experimental nuclear physics. Those are the Double-Sided-Silicon-Strip-Detectors (DSSDs) and the LaBr$_3$(Ce) scintillation detectors. The former, being segmented horizontal and vertically, are to be used as tracking detectors and the latter, an array of 2x2 modern scintillators of LaBr$_3$(Ce) offering fast response and good energy resolution, is to be used as residual energy detector. With these detectors, an image of the sample is taken by mapping the energy losses with respect to horizontal and vertical positions. In this way, we obtain a 3D distribution of Relative Stopping Powers (RSP) needed to design proton therapy treatment plans.
        A first test using low-energy protons (10 MeV) was carried out at the CMAM tandem (Madrid, Spain) in June 2019 to test the proton tracker. Monte Carlo simulations were used to optimise the setup and obtain predicted images of the scanned 2D objects. A second experiment at proton energies relevant in proton therapy (100-200 MeV) will be performed in Cyclotron Centre Bronowice (CCB) in Krakow (Poland) in May 2021. This second measurement will allow to test and optimise the full setup, including the residual energy detector by scanning 3D objects.
        In this contribution we report on the results of the first test at CMAM (Madrid) which demonstrated the viability of the pCT scanner prototype. Additionally, preliminary results from the study at CCB Krakow will be shown.

        [1] Robert P Johnson, 2018, Rep. Prog. Phys. 81 016701
        [2] A. A. Mustafa and D. F. Jackson, Physics in Medicine & Biology, vol. 28, no. 2, 169, 1983.
        [3] U. Schneider, E. Pedroni, and A. J. Lomax, Physics in Medicine & Biology, vol. 41, no. 1, 111, 1996.
        [4] N. Kanematsu, N. Matsufuji, and R. Kohno, Physics in Medicine & Biology, vol. 48, 1053, 2003.

        Speaker: BRIZ MONAGO, Jose Antonio (Instituto de Estructura de la Materia, CSIC, Madrid)
      • 16:25
        #09-232 Effects of Low Doses Ionizing Radiation 5m

        Ionizing radiation is part of the human environment has many sources; some of them natural (earth, cosmos...) and athers artificials (medical, industrial and military). They generate effects that include a wide range of reactions that are different from each other in their dose-effect relation. These effects are often subdivided into two main categories: deterministic and stochastic.
        The carcinogenicity of ionizing radiation, which described early at the beginning of the 20th century, has been since then widely documented. other wise, the hereditary effects of irradiation have never been demonstrated in humans. Many studies have shown the existence of a linear relation between radiation exposure and tumors excess from an exposure to 200 mSv.
        The problem of the risks exposure to low doses of ionizing radiation is a concern of the scientific community. No health effects have been observed in the majority of epidemiological studies, and the absence of this effects do not rule out the risk existence.
        Research on low doses shows that the nature of the biological responses of low doses of radiation differs from those of by high doses of radiation. They also show the diversity of relation ship betwin dose response of the mechanism and the genetic predisposition in individual susceptibility to low doses of radiation.
        The current regulatory texts on low-dose and low-dose rate exposures to ionizing radiation make the implicit assumption of the persistence of a residual risk whatever the level of exposure, with tables for the repair of occupational diseases ranging from simple anemia to bone sarcoma.
        An immense amount of research remains to be done, particularly in radiobiology and in the modeling of radon-induced carcinogenesis, as well as the follow-up of cohorts of exposed subjects, even after the cessation of occupational exposure, to better understand and quantify the effect of low and very low doses of ionizing radiation.
        The objective of our work is to explain the scientific foundations on which the regulatory texts are based and to draw up the current state of knowledge on processes involved in the biological effects of ionizing radiation.

        Speaker: LAILA, Nahime
      • 16:30
        #09-214 Development of a non-intrusive, real-time radio-nuclide and toxic heavy metals monitoring system 5m