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Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*
Transactions of the American Nuclear Society, 124(1), p.544 - 547, 2021/06
Verification calculations for the capability of multi-group cross section generation in FRENDY (FRENDY/MG) are carried out through the comparison of one-group reaction rates using the multi-group cross sections obtained by FRENDY/MG and NJOY2016. Three different neutron spectra (LWR, FR, and 1/E) are used to calculate one-group reaction rates. The discrepancies of one-group reaction rates are small for most cases, showing the validity of FRENDY/MG. The FRENDY/MG will be released as the part of FRENDY nuclear data processing system in the near future.
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi; Endo, Tomohiro*
Transactions of the American Nuclear Society, 124(1), p.556 - 558, 2021/06
The FRENDY nuclear data processing code has been used to generate multi-group cross section libraries for the CBZ reactor physics code system. The newly generated libraries have been applied to neutronics calculations of a fast reactor core MET-1000, and several neutronics parameters are calculated. Calculations with other libraries generated by NJOY2016 have been also conducted, and differences in obtained neutronics parameters between the FRENDY-based library and the NJOY-based library have been quantified. Generally reasonable agreement between them has been obtained, so it has been demonstrated that the multi-group libraries for fast reactor neutronics calculations can be generated successfully by FRENDY. Detailed investigation on the impact of the difference in the processing codes on k-effective has been also carried out with a help of the perturbation theory, and the causes of the differences have been identified.
Endo, Tomohiro*; Noguchi, Akihiro*; Yamamoto, Akio*; Tada, Kenichi
Transactions of the American Nuclear Society, 124(1), p.184 - 187, 2021/06
This study confirmed that the sensitivity analysis of the alpha-eigenvalue can be carried even for non-neutron multiplication systems such as water-only systems. The preliminary results of nuclear data-induced uncertainties of alpha-eigenvalue were smaller than the differences between numerical and experimental results of alpha-eigenvalue. For further investigation, it is necessary to reconsider the experimental bias and the nuclear data-induced uncertainty in alpha-eigenvalue due to the thermal scattering law data of water.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
Transactions of the American Nuclear Society, 124(1), p.824 - 827, 2021/06
The Sodium Chemistry (NAC) package in MELCOR has been developed to enhance application to sodium cooled fast reactors. Based on the recommendations in the previous study through the benchmark analyses of the F7-1 pool fire experiment, this study aims to improve the MELCOR models capturing the oxide layer effect, sodium pool spreading and pool-pan heat transfer, respectively. Each of these models enable a better characterization of the all the processes of relevance to sodium pool fires as observed during the F7-1 test. The MELCOR sodium pool fire enhancement has demonstrated the importance of the improved models.
