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Takino, Kazuo; Oki, Shigeo
JAEA-Data/Code 2023-003, 26 Pages, 2023/05
JAEA-Data-Code-2023-003.pdf:1.66MB
Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.
Collaborative Laboratories for Advanced Decommissioning Science; Ibaraki University*
JAEA-Review 2019-041, 71 Pages, 2020/03
JAEA-Review-2019-041.pdf:3.38MB
JAEA/CLADS, has been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") since FY2018. The Project aims at solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence has been collected from all over the world, and basic research and human resource development have been promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. Among the adopted proposals in FY2018, this report summarizes the research results of the "Contribution to Risk Reduction in Decommissioning Works by the Elucidation of Basic Property of Radioactive Microparticles". In order to establish the decommissioning procedures (recovery of the melted fuels, decontamination inside the reactors, ensuring the safety of the workers, etc.) of the Fukushima Daiichi Nuclear Power Station, radioactive microparticles released by the accident are an important information source for clarifying what had happened inside the reactors in the course of the accident. The purpose of the present study is to obtain detailed knowledge on the basic properties (particle size, composition, electrical/optical properties, etc.) of the radioactive microparticles, as well as to further elucidate the various properties of the radioactive microparticles including the quantitative evaluation of alpha-ray-emitters, through the Japan-UK synergetic research. Thus, we are conducting research and development that will contribute to the comprehensive works towards the risk reduction in the "decommissioning" plan.
Mizukoshi, Seiji; Oshima, Soichiro; Shimada, Taro
JAERI-Tech 2005-011, 122 Pages, 2005/03
JAERI-Tech-2005-011.pdf:13.25MB
The radiological and physical characteristic on decommissioning, such as component and structure weights and radioactivity of the 1.1 MWe class reference nuclear power plants summarized in the NUREG reports of the US NRC were classified,arranged and compared with the domestic commercial nuclear power plants and JPDR from a view point of dismantling plan and waste management for decommissioning. As the results, it was found that the radioactive component and structure weights was about 28,000ton、and non-radioactive structure weights was about 124,000ton less than the domestic commercial BWR. And it was found that this differences has mainly influenced dismantling costs for decommissioning. Farther, it was found that the concrete element composition rates of B, Ni, Nb and so were differerence of one or more figures btween the reference nuclear power plants and the domestic commercial PWR or JPDR.Also,it was found that the this difference became about two or three times by radioactivity concentration and has mainly influenced transport and disposal costs for decommissioning.
Okuno, Hiroshi; Takada, Tomoyuki
Journal of Nuclear Science and Technology, 41(4), p.481 - 492, 2004/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Nuclear characteristic parameters were calculated and subcriticality judgement graphs were drawn for revision purposes of the Data Collection for the Nuclear Criticality Safety Handbook. The nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constants for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-H
O, UF6-HF and Pu(NO
)
-UO(NO) solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese evaluated nuclear data library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgement graphs that depict the region satisfying the inequality relation of the neutron multiplication factor less than 0.98 between the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio were drawn for the same kinds of fuels except UF6-HF in infinite media.
Okuno, Hiroshi; Ryufuku, Susumu*; Suyama, Kenya; Nomura, Yasushi; Tonoike, Kotaro; Miyoshi, Yoshinori
JAERI-Conf 2003-019, p.116 - 121, 2003/10
This paper outlines the data prepared for the 2nd version of Data Collection of the Nuclear Criticality Safety Handbook. These data are discussed in the order of its preliminary table of contents. The nuclear characteristic parameters (k
, M
, D) were derived, and subcriticality judgment graphs were drawn for eleven kinds of fuels which were often encountered in criticality safety evaluation of fuel cycle facilities. For calculation of criticality data, benchmark calculations using the combination of the continuous energy Monte Carlo criticality code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2 were made. The calculation errors were evaluated for this combination. The implementation of the experimental results obtained by using NUCEF facilities into the 2nd version of the Data Collection is under discussion. Therefore, related data were just mentioned. A database is being prepared to retrieve revised data easily.
Nakajima, Ken
Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 8 Pages, 2002/10
The nuclear characteristics of TRACY, such as the criticality, the 
/
ratio, the peak power, the energy of pulse, and the total energy, have been evaluated using the experimental data. TRACY is a supercritical reactor fueled with low-enriched uranyl nitrate aqueous solution to simulate criticality accidents in a fuel processing facility, such as a spent-fuel reprocessing plant. In this evaluation, the availability of criticality calculation and the models to evaluate the power and energy have been studied.
Pu-enriched plutonium*; Yamashita, Kiyonobu; Shindo, Ryuichi; Fujimoto, Nozomu
JAERI-Tech 96-025, 50 Pages, 1996/06
no abstracts in English
Shindo, Ryuichi
Enerugi, Shigen, 14(5), p.419 - 423, 1993/00
no abstracts in English
Journal of Chemical Physics, 84(3), p.1670 - 1676, 1986/00
Times Cited Count：2 Percentile：12.9(Chemistry, Physical)no abstracts in English
Journal of Chemical Physics, 82(2), p.860 - 867, 1985/00
Times Cited Count：2 Percentile：15.03(Chemistry, Physical)no abstracts in English
; ; Aruga, T.;
JAERI-M 8519, 113 Pages, 1979/11
no abstracts in English
; ; *; *; *; *
JAERI-M 8293, 101 Pages, 1979/06
no abstracts in English
S.TERASAWA*; Nozawa, Masao*; T.IWAKI*; J.TAKAHASHI*; H.SHIOZAUKA*; M.FUJITA*
ANL-7520 PART-II, p.99 - 109, 1968/00
no abstracts in English
