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Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09
A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
Wakaida, Ikuo; Oba, Hironori; Miyabe, Masabumi; Akaoka, Katsuaki; Oba, Masaki; Tamura, Koji; Saeki, Morihisa
Kogaku, 48(1), p.13 - 20, 2019/01
By Laser Induced Breakdown Spectroscopy and by related resonance spectroscopy, elemental and isotope analysis of Uranium and Plutonium for nuclear fuel materials and in-situ remote analysis under strong radiation condition for melt downed nuclear fuel debris at damaged core in "Fukushima Daiichi Nuclear Power Station", are introduced and performed as one of the application in atomic energy research field.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*
Nuclear Engineering and Design, 331, p.186 - 193, 2018/05
Times Cited Count：4 Percentile：42.98(Nuclear Science & Technology)A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.
Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki
Procedia Chemistry, 21, p.279 - 284, 2016/12
Times Cited Count：4 Percentile：92.94Sakai, Mikio; Yamamoto, Toshihiro; Murazaki, Minoru; Miyoshi, Yoshinori
Nuclear Technology, 149(2), p.141 - 149, 2005/02
Times Cited Count：3 Percentile：17.99(Nuclear Science & Technology)In the conventional criticality evaluation of the nuclear powder system, the effects of particulate behavior have not been considered. In other words, it is difficult to reflect the particle behavior into the conventional criticality evaluation. We have developed a novel criticality evaluation code to resolve this issue. The criticality evaluation code, coupling a Discrete Element Method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effect of the particulate behavior on a criticality evaluation. The criticality evaluation code has been applied to the powder system of the MOX fuel powder agitation process. The criticality evaluations have been performed under mixing the MOX fuel powder in a stirred vessel to investigate the effects of the powder boundary deformation and particulate mixture conditions on the criticality evaluation. The evaluation results revealed that the powder uniformity mixture condition and the boundary deformation could make the neutron effective multiplication factor decrease.
Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki*
Annals of Nuclear Energy, 30(7), p.811 - 830, 2003/05
Times Cited Count：1 Percentile：11.05(Nuclear Science & Technology)no abstracts in English
Fu, X.*; Takahashi, Masashi; Ueta, Shohei; Sawa, Kazuhiro
JAERI-Tech 2002-049, 35 Pages, 2002/05
JAERI-Tech-2002-049.pdf:3.01MB
no abstracts in English
Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nagatomi, Hideki; Kaminaga, Masanori; Funayama, Yoshiro
JAERI-Tech 2002-034, 40 Pages, 2002/03
JAERI-Tech-2002-034.pdf:1.97MB
JRR-4, a swimming-pool type research reactor with a thermal power of 3.5MW, attained criticality in July 1998, after replacing its 90% enrichment fuel with a 20% enrichment fuel under the Reduced Enrichment Program. As a part of the program, safety analysis on thermo-hydraulics of the reactor core was conducted on cases including single channel blockage accident. With the conclusion that a certain margin on thermo-hydraulics was necessary, investigation and experiments were carried out with an aim to increase the core flow rate. To increase the core flow, it was carried out to reduce the bypass flow in the core and to increase the primary coolant flow rate from 7m
/min to 8m/min. After flow measurements using a mock-up fuel element, flow velocity of the fuel channel was determined as 1.45m/s as opposed to the designed value of 1.44m/s, and the ratio of core flow to total flow was 0.88, exceeding the value 0.86 used for the safety analysis.This report describes the JRR-4 core flow increase plan as well as the results of the channel flow rate measurement
Tachimori, Shoichi
JAERI-Research 2001-048, 23 Pages, 2001/10
JAERI-Research-2001-048.pdf:1.88MB
A new chemical process, ARTIST process, is proposed for the treatment of spent nuclear fuel. The main concept of the ARTIST process is to recover and stock all actinides (Ans) in two groups, uranium (U) and a mixture of transuranics (TRU), to preserve their resource value and to dispose solely fission products (FPs). The process composed of two main steps, an U exclusive isolation and a total recovery of TRU; which copes with the nuclear non-proliferation measures, and additionally Pu separation process and soft N-donor process if requested, and optionally processes for separation of long-lived FPs. These An products: U-product and TRU-product, are to be solidified by calcination and allowed to the interim stockpile for future utilization. These separations are achieved by use of amidic extractants in accord with the CHON principle. The technical feasibility of the ARTIST process was explained by the performance of both the branched-alkyl monoamides the diglycolic amide (TODGA) in thorough extraction of all TRU by tridentate fashon.
Yamane, Yuichi; Miyoshi, Yoshinori
Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 1, p.180 - 185, 1999/00
no abstracts in English
Ogawa, Toru; Akabori, Mitsuo; R.G.Haire*; Kobayashi, Fumiaki
Journal of Nuclear Materials, 247, p.215 - 221, 1997/00
Times Cited Count：11 Percentile：65.93(Materials Science, Multidisciplinary)no abstracts in English
Ogawa, Toru; Mukaiyama, Takehiko; Takano, Hideki; Takizuka, Takakazu; ; *
JAERI-M 89-123, 38 Pages, 1989/09
no abstracts in English
;
JAERI-M 85-215, 83 Pages, 1986/01
no abstracts in English
Yamamoto, Katsumune; ; ; Yokouchi, Iichiro; ;
Nihon Genshiryoku Gakkai-Shi, 28(5), p.425 - 427, 1986/00
Times Cited Count：0 Percentile：0.02(Nuclear Science & Technology)no abstracts in English
; Ichikawa, Michio; *; *; *; *; *; *
JAERI 1298, 90 Pages, 1985/12
no abstracts in English
; ; *
JAERI-M 85-186, 58 Pages, 1985/11
no abstracts in English
Kaminaga, Masanori; ; ; Sudo, Yukio
JAERI-M 85-071, 65 Pages, 1985/06
no abstracts in English
; ; ; ;
JAERI-M 85-047, 81 Pages, 1985/04
no abstracts in English
; *;
JAERI-M 85-002, 62 Pages, 1985/02
no abstracts in English
; E.Groos*; J.Rau*
Nuclear Technology, 69, p.368 - 379, 1985/00
Times Cited Count：2 Percentile：37.74(Nuclear Science & Technology)no abstracts in English
