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Journal Articles

Development methodology on determination of instant release fractions for generic safety assessment for direct disposal of spent nuclear fuel

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.

JAEA Reports

Light water reactor fuel analysis code FEMAXI-V (Ver.1); Detailed structure and user's manual

Suzuki, Motoe; Uetsuka, Hiroshi

JAERI-Data/Code 99-046, p.261 - 0, 1999/11

JAERI-Data-Code-99-046.pdf:10.25MB

no abstracts in English

JAEA Reports

Fission gas release from rock-like fuels, PuO$$_{2}$$-ZrO$$_{2}$$(Y){or ThO$$_{2}$$}-Al$$_{2}$$O$$_{3}$$-MgO at burn-up of 20 MWd/kg

Yanagisawa, Kazuaki; Omichi, Toshihiko; Kanazawa, Hiroyuki; Amano, Hidetoshi; Yamahara, Takeshi

JAERI-Research 97-085, 31 Pages, 1997/11

JAERI-Research-97-085.pdf:2.38MB

no abstracts in English

Journal Articles

Pellet-cladding mechanical interaction of PWR fuel rod under rapid power transient

Yanagisawa, Kazuaki; Katanishi, Shoji;

Journal of Nuclear Science and Technology, 31(7), p.671 - 676, 1994/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior of light water reactor fuel during transient

Yanagisawa, Kazuaki

Dai-8-Kai "Kaku Nenryo, Kaki Semina" Kogi Tekisuto, 38 Pages, 1993/00

no abstracts in English

JAEA Reports

Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

Yanagisawa, Kazuaki; ; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi

JAERI-M 91-202, 33 Pages, 1991/11

JAERI-M-91-202.pdf:1.27MB

no abstracts in English

JAEA Reports

Behavior of water reacter fuel rod

Yanagisawa, Kazuaki

JAERI-M 90-120, 320 Pages, 1990/08

JAERI-M-90-120.pdf:12.75MB

no abstracts in English

Oral presentation

Preliminary analysis on fission gas release of MOX fuel in consideration of the heterogeneous structure

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.

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