Speaker
Description
Knowing the neutron fields in irradiation facilities is crucial for experimental research work in nuclear science and technology. Monte Carlo particle transport techniques are becoming the reference in the context of nuclear reactor analysis and are extensively used for detailed computational characteritzation of irraditation facilities, including the determination of neutron energy spectrum. Experimentally, neutron activation dosimetry is one of the main techniques for neutron flux or fluence determination and for neutron spectrum characterization. Main utilized reactions are sensitive either to the thermal and resonance energy ranges (radiative capture reactions – (n,gamma) reactions) and the fast energy range (threshold reactions). In addition, due to development of new reactor technologies, there has been an increasing interest in nuclear reactions which are sensitive specifically to neutrons in the intermediate – epithermal – energy range. In recent years, several experimental campaigns utilizating neutron activation dosimetry have been conducted at the JSI TRIGA Mark II research reactor at the »Jožef Stefan« Institute. In this work the dosimetry reactions, mainly sensitive in the thermal (197Au(n,gamma)198Au, 59Co(n,gamma)60Co) and in the fast region (27Al(n,gamma)24Na) of the neutron spectrum, were analyzed and the effect of neutron spectrum changes was studied. The computational support of dosimetry experiments was performed with Monte Carlo neutron transport code MCNP, the basis of which is a detailed computational model. In the case of JSI TRIGA reactor, such computational model has been developed and upgraded since 2006. The most recent upgrade is the inclusion of nuclide inventory of burned nuclear fuel. Since the reconstruction of the JSI TRIGA in 1991, the same fuel elements has been in use, resulting in accumulation of average burnup of 18 MWd/kg(HM), which is relatively low for typical 19.9 % enriched esearch reactor, however as shown in the results not negligible. Consequently, an investigation of fuel burnup effects on reactor parameter calculations has been performed. The fuel burnup was determined with Serpent-2 Monte Carlo neutron transport and depletion code. From the standpoint of neutron activation dosimentry, the main effects of interest were in the neutron spectrum changes and the integral parameter keff changes due to fuel burnup. The former are important for calculation of reaction rate and the latter for normalization to obtain absolute values. The results indicate a 6 % increase in the thermal peak of the neutron sperctrum in main JSI TRIGA irradiation channels and keff overestimation of 4400 pcm, compared to steady-state calculations using fresh fuel. This paper presents the computational and part of experimental analysis on how the fuel burnup affects the reaction rates of neutron dosimentry reactions sensitive in the thermal and fast part of the spectrum. Fuel burnup effects are evaluated on absolute dosimetry calculations as well as ratios in which reaction rates on a known standard dosimetry reaction 197Au(n,gamma)198Au is used for normalization. The presented work is crucial for the final goal of benchmarking the JSI TRIGA neutron activation dosimetry measurements and for the characterisation of neutron spectra in irradiation facilities.
