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論文

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.

論文

Study on heterogeneous minor actinide loading fast reactor core concepts with improved safety

大釜 和也; 大木 繁夫; 北田 孝典*; 竹田 敏一*

Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09

A core concept of minor actinides (MAs) transmutation with improved safety was designed by applying sodium plenum and axially heterogeneous configuration. In this study, heterogeneous MA loading methods were developed for the core concept to explore the potential of further improvement of MA transmutation amount and "effective void reactivity" which was introduced by assuming the axial coolant sodium density change distribution for the unprotected loss of flow accident. By investigating characteristics of heterogeneous cores loading MA in different radial or axial positions, preferable MA loading positions were identified. The core loading MA in the radial position between inner and outer core region attained the largest MA transmutation amount and lowest maximum linear heat rate (MLHR) among heterogeneous cases. The lower region of the core was beneficial to improve the effective void reactivity and MLHR maintaining the nearly same MA transmutation amount as that of the homogeneous core. The radial blanket region was also useful to increased MA transmutation amount without deterioration of the effective void reactivity.

報告書

情報セキュリティ教育教材集; 2013年度$$sim$$2017年度

植野 あすか; 矢城 重夫; 宇野 騎一郎*; 青木 和久

JAEA-Review 2018-006, 115 Pages, 2018/06

JAEA-Review-2018-006.pdf:30.8MB

2015年に発生した日本年金機構における不正アクセスによる情報流出事案、2017年に世界各地で猛威を振るったランサムウェアなど、深刻な被害をもたらすサイバー攻撃が継続している。最近では、実在する企業名を騙り、注文書や発送通知を装い、未知の不正プログラムを添付したファイルの開封、悪意あるリンク先の閲覧を巧妙な日本語で促す不審メールが増加するなど、サイバー空間における脅威が増大している。このような中、日本原子力研究開発機構においても情報セキュリティ対策は重要課題であり、システム計算科学センターでは、(1)情報セキュリティ関連規程類の整備、(2)新たな防御技術を実装した情報セキュリティ製品等による対策、(3)職員等の対応能力の維持・向上を図る情報セキュリティ教育・訓練の実施、を三位一体として、プロアクティブな情報セキュリティ対策の推進に取り組んでいる。本報告書は、情報セキュリティ対策の取り組みの一つである情報セキュリティ教育について、eラーニングにより実施している内容を教材集としてまとめたものである。

論文

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

滝野 一夫; 杉野 和輝; 横山 賢治; 神 智之*; 大木 繁夫

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1214 - 1220, 2018/04

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than that of the conventional ones, nuclear design methods need to improve. In this study, we investigated the effect that the analytical conditions exhibit on the accuracy of estimations of the burn-up nuclear characteristics of next-generation fast reactors. Suitable analytical schemes and conditions that maximize the estimation accuracy, while maintaining a low computational cost, were investigated in this study. We performed core burn-up survey calculations under several analysis conditions. Furthermore, we calculated the 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 cycles. The accuracy of the low-cost calculations was evaluated by measuring the agreements with the referential detailed conditions.

論文

中性子回折ラインプロファイル解析によるフェライト系およびオーステナイト系ステンレス鋼の引張変形中の転位増殖その場観察

佐藤 成男*; 黒田 あす美*; 佐藤 こずえ*; 熊谷 正芳*; Harjo, S.; 友田 陽*; 齋藤 洋一*; 轟 秀和*; 小貫 祐介*; 鈴木 茂*

鉄と鋼, 104(4), p.201 - 207, 2018/00

 被引用回数:1 パーセンタイル:49.77(Metallurgy & Metallurgical Engineering)

To investigate the characteristics of dislocation evolution in ferritic and austenitic stainless steels under tensile deformation, neutron diffraction line-profile analysis was carried out. The austenitic steel exhibited higher work hardening than the ferritic steel. The difference in the work hardening ability between the two steels was explained with the dislocation density estimated by the line-profile analysis. The higher dislocation density of the austenitic steel would originate from its lower stacking fault energy. Dislocation arrangement parameters indicated that the strength of interaction between dislocations in the austenitic steel was stronger than that in the ferritic steel.

論文

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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

Mechanical Engineering Journal (Internet), 4(3), p.16-00592_1 - 16-00592_9, 2017/06

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using the Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions show a good agreement with less than 200 pcm of discrepancy in the neutron multiplication factor, and less than 3% of discrepancy in the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic approaches were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.

論文

Progress of design and related researches of sodium-cooled fast reactor in Japan

上出 英樹; 阪本 善彦; 久保 重信; 大木 繁夫; 大島 宏之; 神山 健司

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

東日本大震災以降、日本におけるナトリウム冷却高速炉の開発において安全性の強化、特にシビアアクシデント対策が重要な視点となっている。本論文ではこれらの点での設計ならびに研究開発の進捗を報告する。崩壊熱除去系の強化では炉心損傷事故時の対応を含む多様性、信頼性の向上、熱流動評価手法にかかる研究が行われている。炉心損傷事故時の溶融燃料の挙動について、国際協力を含む炉内試験、炉外試験、基盤的研究が行われ、シビアアクシデントの発生防止の観点での炉心設計改良が進んでいる。

論文

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

前田 誠一郎; 大木 繁夫; 大塚 智史; 森本 恭一; 小澤 隆之; 上出 英樹

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

安全性、環境負荷低減、経済競争力等の幾つかの目標を狙って、日本において次世代高速炉の研究が行われている。安全面では炉心損傷事故での再臨界を防止するため、FAIDUS(内部ダクト付燃料集合体)概念が採用されている。放射性廃棄物の量及び潜在的放射性毒性を低減するために、マイナーアクチニド元素を含むウラン・プルトニウム混合酸化物(MOX)燃料が適用される。燃料サイクルコストを低減するために、高燃焼度燃料が追及される。設計上の工夫によって様々な設計基準を満足する炉心・燃料設計の候補概念が確立された。また、原子力機構においてMA-MOX燃料の物性、照射挙動が研究されている。原子力機構では特にMA含有した場合を含む中空ペレットを用いた燃料ピンの設計コードの開発を進めている。その上、原子力機構では高燃焼度燃料のために酸化物分散強化型フェライト鋼製被覆管の開発を進めている。

論文

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.

論文

Comparative study on burnup characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using the Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, the atomic number densities and core characteristics at the end of cycle were evaluated by the best estimate deterministic methodologies of ANL and JAEA. The atomic number densities of plutonium isotopes calculated by both institutions showed a good agreement with less than 0.5% of discrepancy, except for the atomic number density of Pu-241. The atomic number densities of americium and curium isotopes showed less than 6% of discrepancy. The results of core characteristics at the end of cycle obtained by both institutions showed a reasonably good agreement with less than 400 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. A burnup sensitivity analysis was employed to identify the major factors of the difference in the calculated atomic number densities at the end of cycle.

論文

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.

論文

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.

論文

Core concept of minor actinides transmutation fast reactor with improved safety

藤村 幸治*; 糸岡 聡*; 大木 繁夫; 竹田 敏一*

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

A core concept of minor actinides (MAs)transmutation with improved safety was developed. Coresafety was improved by reducing sodium void reactivitywith sodium plenum and optimized configuration ofaxially heterogeneous core. "The effective void reactivity"by assuming the axial coolant sodium density changedistribution for ULOF (unprotected loss of flow) accidentwas introduced. MA content in the core fuel wasincreased up to 11 wt% for the condition that negativeeffective void reactivity was attained. Therefore, the largeMA transmutation amount which is almost two times tothe conventional Japanese fast reactor was obtained. Wealso conducted thermal hydraulics and fuel integrityevaluation of the core concept and it was confirmed thattheir results meet to the Japanese fast reactor designconditions. Finally we evaluated the transient behavior ofthe core concept and it was confirmed that this core hassluggish response during ULOF accident due to thenegative coolant reactivity effect of the upper sodiumplenum.

論文

Core performance requirements and design conditions for next-generation sodium-cooled fast reactor in Japan

大木 繁夫; 丸山 修平; 近澤 佳隆; 大滝 明; 久保 重信; 日比 宏基*; 菅 太郎*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

A conceptual design study on a next-generation sodium-cooled fast reactor was conducted in Japan. This paper describes a recent review and modification of core performance requirements and design conditions for the demonstration and the commercial phases. We have highlighted the fuel composition (i.e., heavy metal nuclide composition). The fuel composition for next-generation fast reactors has a wide range depending on a variety of spent fuels used in light water reactors and the methods of recycling them in a fast reactor fuel cycle. The design envelopes of fuel composition were determined by using a remarkable correlation between fuel composition and core characteristics. The consistency of those design envelopes was checked by comparing them with the results of representative fast reactor deployment scenario simulations. Moreover, reflecting the realistic situation that a fast reactor core accepts various fuel compositions in the design envelope simultaneously, the design procedure of multiple fuel-composition loading was introduced. This paper describes the fundamental consideration of its effects, and the accompanying paper describes its practical application to core design. The design conditions and procedures concerning fuel composition variety facilitate sophisticated core design for next-generation sodium-cooled fast reactors.

論文

Core design of the next-generation sodium-cooled fast reactor in Japan

菅 太郎*; 小倉 理志*; 日比 宏基*; 大木 繁夫; 前田 誠一郎; 丸山 修平; 大釜 和也

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

In Japan, a 1500MWe-scale sodium-cooled fastreactor (FR) has been designed as a commercial phaseFR for utilizing in an equilibrium FR operation era, and a 750MWe-scale FR has been as a demonstration phase FRfor realizing the commercial phase FR. Thedemonstration phase core adopts a core and a blanketfuel subassembly with the same specifications of thecommercial phase core, and is designed to satisfy designrequirements, especially to accept a broad range of fuelcompositions, which arises in a transition period from anLWR are to an FR era. By optimizing an arrangement offuel subassemblies and control rods, and employing a fluxadjuster, the demonstration phase core gets flat powerdistribution giving high core performances. And its coreand fuel specifications are materialized to satisfy thedesign requirements desired for the next-generation FR.

論文

Rapid measurement scheme for texture in cubic metallic materials using time-of-flight neutron diffraction at iMATERIA

小貫 祐介*; 星川 晃範*; 佐藤 成男*; 徐 平光; 石垣 徹*; 齋藤 洋一*; 轟 秀和*; 林 眞琴*

Journal of Applied Crystallography, 49(5), p.1579 - 1584, 2016/10

AA2016-0186.pdf:1.13MB

 被引用回数:7 パーセンタイル:25.12(Chemistry, Multidisciplinary)

The authors have developed texture measurement system on a time-of flight neutron diffractometer, iMATERIA (BL20, MLF/J-PARC, Japan). Quantitative Rietveld texture analysis for a duplex stainless steel was possible with one neutron beam exposure for several minutes without sample rotation. The minimum number of diffraction spectra required for the Rietveld texture analysis was experimentally determined as $$sim$$100. The suggested rapid measurement scheme used 132 spectra and determined volume fractions of texture components in both ferrite and austenite phases, quantitatively. This quantitative and rapid measurement scheme was established by utilizing the features of iMATERIA as a powder diffractometer, i.e. fairly high resolution in d-spacing and numerous detectors equipped in a wide range of scattering angle.

論文

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

大釜 和也; 中野 佳洋; 大木 繁夫

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

 被引用回数:1 パーセンタイル:75.52(Nuclear Science & Technology)

JSFR(Japan Sodium-cooled Fast Reactor)では、炉心崩壊事故(CDA)対策として、内部ダクト付燃料集合体を採用している。炉心核計算において、この内部ダクト構造を直接取扱い、全内部ダクトが炉心中心に対して外側を向くように集合体を配列した場合(外向)、全内部ダクトが内側を向くように集合体を配列した場合(内向)に比較して、炉心中心付近の出力分布が高くなることが報告されている。この要因を分析するため、本研究では、モンテカルロ法に基づく輸送計算および燃焼計算コードを使用し、種々の内部ダクト配列において炉心の出力分布および炉心特性を評価した。この結果、外向および内向配置における炉心中心の出力分布の違いの主要因は、内部ダクト配列の違いに起因する核物質の空間分布の違いであることがわかった。同じメカニズムで、炉心中心以外においても内部ダクト配置の違いにより出力分布に影響が生じることがわかった。また、内部ダクト配置の違いによる制御棒価値への影響を確認した。

論文

量子ビーム回折法を用いた鉄鋼組織解析

佐藤 茂男*; 小貫 祐介*; 菖蒲 敬久; 城 鮎美*; 田代 均*; 轟 秀和*; 鈴木 茂*

金属, 86(8), p.654 - 660, 2016/08

鉄鋼材料に代表される金属のミクロ組織に対し、放射光、中性子線の量子ビームを用いた新たなアプローチが可能になる。放射光の高輝度特性を活かしたマイクロ領域測定は高温その場測定により、構造内部の転位の分布、転位の回復現象の追跡が可能となる。また、中性子線はバルク平均情報の特徴を活かすことで、複相鋼の強化機構メカニズムが詳細に考察できる。これら解析事例を原理とともに紹介している。

論文

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 6 Pages, 2016/06

日米二国間協力枠組による民生原子力研究開発ワーキンググループにおける国際協力の下、アルゴンヌ国立研究所(ANL)および原子力機構は、JSFR金属燃料炉心のベンチマーク研究を実施してきている。このベンチマーク研究では、ANLおよび原子力機構の決定論最確評価手法およびモンテカルロ法により、平衡サイクル初期の炉心核特性を評価した。ANLおよび原子力機構の解析結果は、中性子増倍率において200pcm以下の差で、ナトリウムボイド反応度、ドップラ係数および制御棒価値において3%以下でよく一致した。決定論による近似の影響を分析するとともに、解析手法の違いによる結果への影響を把握するため、決定論およびモンテカルロ法の計算結果を比較した。また、核データライブラリの違いによる影響を感度解析法により分析した。

論文

Actinide management with commercial fast reactors

大木 繁夫

AIP Conference Proceedings 1702, p.040008_1 - 040008_4, 2015/12

 パーセンタイル:100

高速炉はプルトニウム(Pu)やマイナーアクチニド(MA)をシステム内で柔軟にリサイクルでき、Puの増殖・持続的利用とMAの核変換要求に対応可能であり、エネルギー問題の重要な解決策の一つであるとともに、システム外に排出する放射性廃棄物の有害度を合理的な範囲で最小化できる。本報告では、日本が国家プロジェクトして研究開発を進めてきた第4世代のナトリウム冷却高速炉(JSFR:75万kWe実証炉、150万kWe実用炉)のレファレンス炉心である高内部転換型炉心(MOX燃料使用)におけるPu増殖性能とMA核変換特性について述べる。増殖性能としては、Pu需給バランスに応じて増殖比で1.0$$sim$$1.2までの変更が比較的軽微な炉心仕様の変更で可能であることを示す。MA核変換特性としては、炉心燃料におけるMA含有率(MA/HM)が3$$sim$$5wt%の場合、MA核変換量は50$$sim$$100kg/GWe y、また、30$$sim$$40%という高い燃料取出し時のMA核変換率を達成可能である。

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