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Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

滝野 一夫; 大木 繁夫

JAEA-Data/Code 2023-003, 26 Pages, 2023/05





横山 賢治; 丸山 修平; 谷中 裕; 大木 繁夫

JAEA-Data/Code 2021-019, 115 Pages, 2022/03




A Validation study of a neutronics design methodology for fast reactors using reaction rate distribution measurements in the prototype fast reactor Monju

大釜 和也; 竹越 淳; 片桐 寛樹*; 羽様 平

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

In the fast breeder reactor prototype Monju, reaction rate distributions were measured by using activation foils during its system startup test. Reliability and usefulness of the measurements as a validation experiment were investigated through a comparison with a calculation using the latest neutronics design methodology developed in JAEA. As a basic calculation, a three-dimensional diffusion calculation with triangular meshes was performed using effective cross sections generated by a one-dimensional heterogeneous lattice model with the JENDL-4.0 nuclear data library. Best-estimate values of reaction rates were evaluated by considering correction factors such as transport correction factors, fine and ultra-fine energy group correction factors, anisotropic diffusion coefficient correction factors and subassembly heterogeneous factors. Through the comparison, it was confirmed that the both of experimental values and analysis results were agreed well in the core region.


A Design study on a mixed oxide fuel sodium-cooled fast reactor core partially loading highly concentrated MA-containing metal fuel

大釜 和也; 太田 宏一*; 大木 繁夫; 飯塚 政利*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05

A neutronics design study for a mixed oxide (MOX) fuel Sodium-cooled Fast Reactor (SFR) core partially loading highly concentrated Minor Actinide (MA) containing fuel was conducted. To analyze preferable loading positions of highly concentrated MA-containing metal fuel, the characteristics of heterogeneous MA loading cores were evaluated assuming the amount of MA loaded to heterogeneous cores were same as that of a reference homogeneous 3% MAcontaining MOX fuel core. The cores loading MA-containing metal fuel could meet the design limitation of the sodium void reactivity of the SFR except for the one loading MA-containing metal fuel in the core center region. Based on these results, the core design was modified to maximize amount of MA transmutation. The modified core loading 60 subassemblies of 16% MA-containing metal fuel in the outermost region could attain the largest amount of MA transmutation, which was larger by about 60% than that of the reference homogeneous MOX fuel core.


Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

大釜 和也; 池田 一三*; 石川 眞; 菅 太郎*; 丸山 修平; 横山 賢治; 杉野 和輝; 長家 康展; 大木 繁夫

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Detailed model verification & validation (V&V) and uncertainty quantification (UQ) procedure for our deterministic neutronics design methodology including the nuclear library JENDL-4.0 for next generation fast reactors was put into shape based on a guideline for reliability assessment of simulations published in 2016 by the Atomic Energy Society of Japan. The verification process of the methodology was concretized to compare the results predicted by the methodology with those by a continuous-energy Monte Carlo code, MVP with their precise geometry models. Also, the validation process was materialized to compare the results by the methodology with a fast reactor experimental database developed by Japan Atomic Energy Agency. For the UQ of the results by the methodology, the total value of the uncertainty was classified into three factors: (1) Uncertainty due to analysis models, (2) Uncertainty due to nuclear data, and (3) Other uncertainty due to the differences between analysis models and real reactor conditions related to the reactor conditions such as fuel compositions, geometry and temperature. The procedure to evaluate the uncertainty due to analysis models and uncertainty due to nuclear data was established.


Tradeoff analysis of metal-fueled fast reactor design concepts

Stauff, N. E.*; 大釜 和也; Aliberti, G.*; 大木 繁夫; Kim, T. K.*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Within the framework of the U.S.-Japan bilateral, the Civil Nuclear Energy R&D Working Group (CNWG), a core design study was conducted by ANL and JAEA. Its objective was to compare the core performance characteristics of metal-fueled Sodium-cooled Fast Reactors (SFRs) developed with different design preferences: JAEA preferred a loop-type primary system with high coolant temperature, while ANL targeted a pool-type primary system with a conventional coolant temperature. The comparative core design study was conducted based on the 3530 MWth Japan Sodium-cooled Fast Reactor (JSFR) metallic-fuel core concept. This study confirms that both metal fueled SFR core concepts developed by ANL and JAEA based on different design preferences and approaches are viable options.


Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

大釜 和也; 太田 宏一*; 生澤 佳久; 大木 繁夫; 尾形 孝成*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Under the collaborative research of Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Agency (JAEA), the metal fuel core concept has been studied. In this study, a 750 MWe sodium-cooled fast reactor (SFR) with metal fuel designed in a past/precedent study was reevaluated considering the irradiation behaviors of metal fuel such as axial elongation and bond-sodium redistribution, which have significant impacts on the core characteristics such as the multiplication factor and sodium void reactivity worth. The result of reanalysis indicated that the sodium void reactivity worth of the core became higher than that evaluated in the past study, so the redesign of the core was performed to improve the sodium void reactivity worth. To redesign the core, correlations of the sodium void reactivity worth and the dimension of the core and fuel subassemblies was investigated by survey calculations. Based on the results, specifications of the redesigned core were selected. The characteristics of the redesign core were evaluated. To verify the deterministic calculation results, the core characteristics of the redesign core were compared with those by a contentious-energy Monte Carlo simulation with precise geometry modeling, which can provide reference solutions. The both calculations agreed well, and the improvements of core characteristics of the redesign core were verified.



横山 賢治; 神 智之; 平井 康志*; 羽様 平

JAEA-Data/Code 2015-009, 120 Pages, 2015/07




Neutronics design of the low aspect ratio tokamak reactor, VECTOR

西谷 健夫; 山内 通則*; 西尾 敏; 和田 政行*

Fusion Engineering and Design, 81(8-14), p.1245 - 1249, 2006/02

 被引用回数:13 パーセンタイル:66.51(Nuclear Science & Technology)



Proceedings of the 11th International Workshop on Ceramic Breeder Blanket Interactions; December 15 - 17, 2003, Tokyo, Japan

榎枝 幹男

JAERI-Conf 2004-012, 237 Pages, 2004/07


本報文集は、「IEA核融合炉工学に関する実施取り決め」に基づくセラミック増殖材ワークショップ及び日米核融合共同研究の一環として開催された「第11回セラミック増殖材ブランケット相互作用国際ワークショップ」の報文をまとめたものである。本ワークショップでは、欧州連合,ロシア,日本のセラミック増殖ブランケットの設計,HICU, EXOTIC-8, IVV-2Mによる照射試験の最新の成果,Li$$_{2}$$TiO$$_{3}$$等のトリチウム放出挙動のモデリング,Li$$_{2}$$TiO$$_{3}$$とLi$$_{4}$$SiO$$_{4}$$微小球の製造技術開発と物性値研究,Li$$_{2}$$TiO$$_{3}$$とLi$$_{4}$$SiO$$_{4}$$微小球充填層の熱機械挙動測定とモデリングに関する研究,境界テーマとして、ブランケット筐体製作技術開発,核融合中性子によるブランケットモックアップの中性子工学実験,トリチウム回収システム開発、などについての研究開発の現状と今後の課題についての情報交換が行われた。



代谷 誠治*; 三澤 毅*; 宇根崎 博信*; 市原 千尋*; 小林 圭二*; 中村 博*; 秦 和夫*; 今西 信嗣*; 金澤 哲*; 森 貴正

JAERI-Tech 2004-025, 93 Pages, 2004/03




Current status of the AGS spallation target experiment

中島 宏; 高田 弘; 春日井 好己; 明午 伸一郎; 前川 藤夫; 甲斐 哲也; 今野 力; 池田 裕二郎; 大山 幸夫; 渡辺 昇; et al.

Proceedings of 6th Meeting of the Task Force on Shielding Aspects of Accelerators, Targets and Irradiation Facilities (SATIF-6), (OECD/NEA No.3828), p.27 - 36, 2004/00

米国ブルックヘブン国立研究所AGS(Alternating Gradient Synchrotron)加速器を用いて行われている一連の核破砕ターゲット実験及びその解析の概要について報告する。本実験では、中性子発生特性,遮蔽設計パラメータに関する情報を得ることを目的として、AGS加速器から得られる数GeV,数百kJの陽子ビームを水銀核破砕ターゲットに入射し、そこで発生する二次粒子を用いて、中性子工学及び遮蔽に関する実験を過去4年間にわたって行ってきた。昨年、遮蔽実験を行うとともに、これまでの実験結果の解析を通して大強度陽子加速器施設の設計コードの精度検証が精力的に行われている。本報告では、昨年行った遮蔽実験の最新結果及びこれまで行ってきた実験解析の結果について紹介する。


Development of supercritical pressure water cooled solid breeder blanket in JAERI

秋場 真人; 石塚 悦男; 榎枝 幹男; 西谷 健夫; 小西 哲之

プラズマ・核融合学会誌, 79(9), p.929 - 934, 2003/09




久語 輝彦; 土橋 敬一郎*; 中川 正幸; 井戸 勝*

JAERI-Data/Code 2000-011, p.138 - 0, 2000/02





久語 輝彦; 中川 正幸

JAERI-Data/Code 2000-004, p.97 - 0, 2000/02




Application of neural network to multi-dimensional design window search in reactor core design

久語 輝彦; 中川 正幸

Journal of Nuclear Science and Technology, 36(4), p.332 - 343, 1999/04

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



Applicability of design window search procedure using neural network to neutronics

久語 輝彦; 中川 正幸

Proc. of Int. Conf. on the Phys. of Nucl. Sci. and Technol., 1, p.704 - 711, 1998/00



Development of intelligent code system to support conceptual design of nuclear reactor core

久語 輝彦; 中川 正幸; 土橋 敬一郎

Journal of Nuclear Science and Technology, 34(8), p.760 - 770, 1997/08

 被引用回数:2 パーセンタイル:23.16(Nuclear Science & Technology)



SRAC95; 汎用核計算コードシステム

奥村 啓介; 金子 邦男*; 土橋 敬一郎

JAERI-Data/Code 96-015, 445 Pages, 1996/03




Design and techniques for fusion blanket neutronics experiments using an accelerator-based deuterium-tritium neutron source

大山 幸夫; 今野 力; 池田 裕二郎; 前川 藤夫; 前川 洋; 山口 誠哉; 津田 孝一; 中村 知夫; M.A.Abdou*; Bennett, E. F.*; et al.

Fusion Technology, 28(1), p.56 - 73, 1995/08


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