Benchmark analyses on control rod worths of TRIGA reactor modeled in the ICSBEP handbook using continuous-energy Monte Carlo code MVP version 3

Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

https://doi.org/10.11484/jaea-conf-2024-001

The benchmark analyses on the TRIGA reactor of the IEU-COMP-THERM-013 (ICT-013) in the ICSBEP handbook using the MVP version 3 code were carried out with the Japanese, US and European nuclear data libraries, including JENDL-5. The analyses on effective neutron multiplication factors (k's) were also performed for another TRIGA reactor defined in the ICT-003 as well as the ICT-013. As a result, it is confirmed that the calculated k's vary in the range of 0.8% with the libraries. The results also suggest that there might be unknown bias in the ICT-013 for the calculated k's. For the analyses on the control rod worths of the ICT-013, it is confirmed that differences in the worths among the libraries become smaller than those in k's. Most of the errors involved in k's are considered to be cancelled in the calculation of the worths since the worths are obtained by subtraction of the reciprocals of two sorts of k's. The differences in the worths by between the benchmark and alternative methods defined in the ICT-013 were tried to be examined from the aspect of the differences in horizontal flux distributions by between those methods. It is confirmed that the differences in the worths for two shim control rods, which are fully withdrawn at a delayed critical state, are well understood by that attempt, but on the other hand the differences are not sufficiently explained for a regulating control rod, which is partially inserted to attain delayed criticality.

Atomic Energy Society of Japan 2017 Annual Meeting, joint session of "sigma advisory committee", "subcommittee on nuclear data" and "subcommittee on reactor physics"; Current status and future perspective of the Verification and Validation (V&V) of JENDL and neutronics calculation codes by use of the benchmark problems and integral experiments, 4, Development of an automatic nuclear data validation system VACANCE

Tada, Kenichi

Kaku Deta Nyusu (Internet), (117), p.23 - 29, 2017/06

http://www.aesj.or.jp/~ndd/ndnews/pdf117/No117-04.pdf

Validation of the nuclear data library using the criticality experiments and nuclear reactor experiments, i.e., integral experiments, is one of the most important process. The analyses of the integral experiments become the more important along with the accuracy improvement of the library. This validation procedure is mainly carried out by the specialists of the reactor physics because it is complicated for the nuclear data evaluators. Furthermore, it takes a long time and much effort even if the specialists carry it out. To realize the efficient nuclear data validation cycle for the next version of JENDL, the automatic nuclear data validation system VACANCE (Validation Environment for Comprehensive and Automatic Neutronics Calculation Execution) is developed. In this presentation, the outline and functions of VACANCE are demonstrated in detail and examples of the new nuclear data evaluation and subsequent integral experiment analyses will be shown.

Integral benchmark test of JENDL-4.0 for U-233 systems with ICSBEP Handbook

Kuwagaki, Kazuki*; Nagaya, Yasunobu

JAEA-Data/Code 2017-007, 27 Pages, 2017/03

JAEA-Data-Code-2017-007.pdf:4.77MB
JAEA-Data-Code-2017-007-appendix(CD-ROM).zip:0.37MB

The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors ('s) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed.

Analysis of the TRIGA MARK-II benchmark IEU-COMP-THERM-003 with Monte Carlo code MVP

Mahmood, M. S.; Nagaya, Yasunobu; Mori, Takamasa

JAERI-Tech 2004-027, 30 Pages, 2004/03

JAERI-Tech-2004-027.pdf:2.26MB

The benchmark experiments of the TRIGA Mark-II reactor in the ICSBEP handbook have been analyzed with the Monte Carlo code MVP using the cross section libraries based on JENDL-3.3, JENDL-3.2 and ENDF/B-VI.8. MCNP calculations have been also performed with the ENDF/B-VI.6 library for comparison between the MVP and MCNP results. For both cores labeled 132 and 133, which have different core configurations, the ratio of the calculated to the experimental results (C/E) for keff obtained by the MVP code is 0.999 for JENDL-3.3, 1.003 for JENDL-3.2, and 0.998 for ENDF/B-VI.8. For the MCNP code, the C/E values are 0.998 for both the Core 132 and 133. All the calculated results agree with the reference values within the experimental uncertainties. The results obtained by MVP with ENDF/B-VI.8 and MCNP with ENDF/B-VI.6 differ only by 0.02% for Core 132, and by 0.01% for Core 133.

Base data for looking-up tables of calculation errors in JACS code system

Murazaki, Minoru*;

JAERI-Data/Code 99-019, 103 Pages, 1999/03

JAERI-Data-Code-99-019.pdf:3.69MB

no abstracts in English

JENDL-5 validation, 1; Benchmark tests with the ICSBEP handbook

Nagaya, Yasunobu; Tada, Kenichi

no journal, , 

The latest version of Japanese Evaluated Nuclear Data Library, JENDL-5, was released in December of 2021. To investigate the prediction accuracy of JENDL-5 for criticality of reactor cores, we performed benchmark tests by using the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. Judging from the overall results, we confirmed that JENDL-5 gives the better results than JENDL-4.0.

Development of nuclear data processing system FRENDY, 4; Implementation of the ACE file generator

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi

no journal, , 

JAEA has been developing the nuclear data processing system FRENDY (FRom Evaluated Nuclear Data librarY to any application). The first goal of the development of FRENDY is generation function of the nuclear data library for the continuous energy Monte Carlo calculation code. To generate the nuclear data library, four procedures, i.e., (1) Generation of point wise cross section, (2) generation of the probability table, (3) processing the thermal scattering law data, and (4) generation of the nuclear data library, are required. In this presentation, generation of the ACE file which is the nuclear data library for the continuous energy Monte Carlo code is described. To verify FRENDY, the K values which is using the ACE file which is processed by FRENDY are compared with that by NJOY99. The difference of the K values are less than 0.05%. This result indicates that FRENDY appropriately generate the ACE file.

Important findings for the validation of nuclear data using the JEFF benchmarks suite

Defrance, V.*; Tada, Kenichi

no journal, , 

For the validation of the nuclear data library, JAEA has prepared the MVP input deck. This presentation shows the comparison of the C/E values of JENDL-4.0 and those of JENDL-5 using this input deck. This presentation also shows the important findings obtained from such comparison results, such as the impact of separating O in HO and O in nitrate ion (NO) on the k-effective value. They will be good information not only for the validation of the JENDL library but also for that of the JEFF library.