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Reactor Integral Test Working Group, JENDL Committee
JAEA-Data/Code 2023-004, 132 Pages, 2023/06
JAEA-Data-Code-2023-004.pdf:11.99MB
JAEA-Data-Code-2023-004-appendix(CD-ROM).zip:566.51MB
The reactor integral test working group organized in the JENDL committee has prepared the benchmark database which is devoted to performance evaluation and validation testing of JENDL-5 regarding simulation of nuclear fission reactors applications, and the report "JAEA-Data/Code 2017-006" for the aforementioned benchmark database was published in 2017. After the publication of the report, the benchmark database has been extended energetically in Japan Atomic Energy Agency under the cooperation of the JENDL Committee. The database contributes improvement of reliability of JENDL-5. This report is a summary of the aforementioned extended database, and the report is expected to be efficiently utilized for not only the developments of nuclear data library in future but also integral tests of released nuclear data libraries.
Reactor Integral Test Working Group, JENDL Committee
JAEA-Data/Code 2017-006, 152 Pages, 2017/05
JAEA-Data-Code-2017-006.pdf:13.46MBJAEA-Data-Code-2017-006(errata).pdf:0.07MBJAEA-Data-Code-2017-006-appendix1(CD-ROM).zip:115.88MBJAEA-Data-Code-2017-006-appendix2(CD-ROM).zip:110.88MB
A benchmark database which is devoted to the evaluation of the future JENDL against the criticality of light water reactors was prepared, where the ICSBEP and IRPhEP handbooks by OECD/NEA were utilized effectively. Specific features of this report can be described as follows: (1) The recommendation for benchmarking is based on careful reviewing for the document and related information. Validity of the original benchmark evaluation is carefully checked, and numerical results obtained with JENDL-4.0 are considered. (2) Heterogeneity effect of PuO
particles dispersed in fuel medium is consistently quantified for the MOX fuel-loaded experimental data. This precise evaluation is realized by the newly developed finite fuel pin bundle model with the Monte Carlo neutron transport code. (3) Sensitivity analysis is conducted in order to specify nuclear data whose difference between recent nuclear data libraries significantly affects the critical parameter calculation.