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Tasaki, Yudai; Udagawa, Yutaka
JAEA-Data/Code 2024-012, 76 Pages, 2024/12
Japan Atomic Energy Agency (JAEA) has been developing a fuel performance code, FEMAXI, to evaluate the behavior of LWR fuels under normal operation and transient conditions. In March 2019, FEMAXI-8, the first systematically validated and performance evaluated code, was released. Since then, the code has undergone various improvements. In parallel, since the 2000s, JAEA has been developing the RANNS module as a branch for design basis accident (DBA) analysis, with a particular emphasis on computational stability, so that fuel behavior can be tracked even for very steep transients, in this case mainly reactivity-initiated accidents (RIAs). The specific models include boiling heat transfer, fission gas release by grain boundary failure, and cladding failure determination based on fracture mechanics parameters, which are essential for predicting such transient behavior. In this report, prior to the release of RANNS, we present a description of the models for accident behavior analysis, the relationship with FEMAXI-8 in terms of the design and structure of the program, and the results of a large-scale validation using the extensive database of RIA experiments conducted and accumulated by JAEA, to evaluate the overall RIA analysis performance. The code will be made available to users as a packaged FEMAXI/RANNS, enabling them to analyze fuel behavior under various conditions. The model parameter sets determined through the above validation analyses are also presented in this report, and by referring to them, the analysis can be easily performed with almost no change in usability from the previously released FEMAXI-8.
Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Kitamura, Akira; Akahori, Kuniaki; Nagata, Masanobu*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(2), p.83 - 93, 2020/12
Direct disposal of spent nuclear fuel (SNF) in deep underground repositories (hereafter "direct disposal") is a concept that disposal canisters stored fuel assemblies dispose without reprocessing. Behavior of radionuclide release from SNF must be different from that from vitrified glass. The present study established a methodology on determination of instant release fraction (IRF) of radionuclides from SNF, which is the one of the parameters on radionuclide release based on the latest safety assessment reports in other countries, especially for IRF values proportional to a fission gas release ratio (FGR). Recommended and maximum values of FGR have been estimated using the fuel performance code FEMAXI-7 after collecting FGR values on Japanese SNFs. Furthermore, recommended and maximum values of IRF for Japanese SNFs used in a pressurized water reactor (PWR) have been estimated using the presently obtained FGR values and experimentally obtained IRF values on foreign SNFs. The recommended and maximum IRF values obtained in the present study have been compared with those of the latest safety assessment reports in other countries.
Uehara, Hiroyuki; Komuro, Kazuma; Usami, Koji; Nihei, Yasuo; Nakata, Masahito
Proceedings of 52nd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2015) (Internet), 5 Pages, 2015/09
For the accurate measurement of the released FP gas during RIA, Gas Chromatograph was improved with the hardware approach and the software one. As the hardware approach, PDD was added to the gas detector of GC and it can detect 50 times higher sensitivity compared with the TCD. As the software approach, ETG function was adopted to deconvolute the overlapped Kr peak from N one. With ETG estimation, Kr volume can be evaluated with the deconvolution from N
peak in neither the long measurement time nor the changing measurement condition.
Suzuki, Motoe; Uetsuka, Hiroshi
JAERI-Data/Code 99-046, p.261 - 0, 1999/11
no abstracts in English
Yanagisawa, Kazuaki; Omichi, Toshihiko; Kanazawa, Hiroyuki; Amano, Hidetoshi; Yamahara, Takeshi
JAERI-Research 97-085, 31 Pages, 1997/11
no abstracts in English
Yanagisawa, Kazuaki; Katanishi, Shoji;
Journal of Nuclear Science and Technology, 31(7), p.671 - 676, 1994/07
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)no abstracts in English
Yanagisawa, Kazuaki
Dai-8-Kai "Kaku Nenryo, Kaki Semina" Kogi Tekisuto, 38 Pages, 1993/00
no abstracts in English
Yanagisawa, Kazuaki; ; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi
JAERI-M 91-202, 33 Pages, 1991/11
no abstracts in English
Yanagisawa, Kazuaki
JAERI-M 90-120, 320 Pages, 1990/08
no abstracts in English
Tasaki, Yudai
no journal, ,
Fuel pellets contained in a fuel rod of a light water reactor release some degree of FP gas, and it raises inner gas pressure of the fuel rod. It is important for fuel design and safety evaluation to estimate fission gas release ratio (FGR), since excessive inner gas pressure may impair fuel integrity. Therefore, JAEA develops fuel performance code FEMAXI for evaluating various fuel behaviors. However, FEMAXI has limited capability for considering FGR in case of evaluating to compare MOX fuels which have different microstructures arised from the difference of production method. Because, FGR model of FEMAXI can apply only one kind of fuel grain. From the above, it is necessary to acquire sufficient understanding of MOX fuel behavior based on experimental data, and FGR modeling has to be improved accordingly. This study modified FGR model so that two kinds of grains with respect to Pu enrichment get applicable and released fission gas values from them are applied weighted average based on volume occupancy ratio of each structures. Upon analyzing two types of MOX fuels, the different homogeneity of microstructure, with and without the modified FGR model, then comparing the experimental data provided by a testing reactor, analytical results of FGR captured tendency of experimental data qualitatively. Hence, it was revealed that modified FGR model is valid for evaluating FGR in consideration of the heterogeneous structure of MOX fuel.