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論文

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

浜瀬 枝里菜; 大釜 和也; 河村 拓己*; 堂田 哲広; 田中 正暁; 山野 秀将

Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01

高速炉プラント動特性解析コードSuper-COPDのスクラム不作動流量喪失事象に対する妥当性確認のため、FFTFの受動的安全性試験LOFWOS No.13試験を対象としたIAEAベンチマークに参加した。ブラインドフェーズで課題として抽出された燃料集合体出口温度及び全反応度の評価精度向上のため、集合体間熱移行及び集合体間ギャップ部流れを考慮した全炉心モデル及び炉心湾曲反応度簡易評価モデルを導入した。最終フェーズ解析の結果、2次ピーク時の集合体出口温度を良好に再現するとともに、全反応度の実測値の挙動を概ね評価できたことから、LOFWOSに対するSuper-COPDの妥当性を確認した。

論文

Validation practices of multi-physics core performance analysis in an advanced reactor design study

堂田 哲広; 加藤 慎也; 浜瀬 枝里菜; 桑垣 一紀; 菊地 紀宏; 大釜 和也; 吉村 一夫; 吉川 龍志; 横山 賢治; 上羽 智之; et al.

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08

安全かつ経済的で持続可能な先進的原子炉を実現するために革新的設計システム(ARKADIA)を開発している。本論文では、ARKADIAの一部である設計研究のためのARKADIA-Designに着目し、炉心設計の数値解析手法の妥当性確認について紹介する。ARKADIA-Designでは、炉物理、熱流動、炉心構造、燃料ピン挙動の解析コードを組み合わせたマルチフィジックス解析により、ナトリウム冷却高速炉の炉心性能を解析する。これらの解析の妥当性を確認するため、実験データ及び信頼できる数値解析結果を選定し、検証マトリックスを作成する。解析コードのモデル及び検証マトリクスの代表的な確認解析について説明する。

論文

Evaluation of sodium radioactivity in the primary system of the prototype fast reactor Monju

毛利 哲也; 大釜 和也; 羽様 平

Nuclear Technology, 209(7), p.1008 - 1023, 2023/07

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

高速増殖原型炉「もんじゅ」で測定された1次系放射化ナトリウムである$$^{24}$$Na及び$$^{22}$$Naの放射化量を評価し、測定値及び計算値の信頼性を検討した。JENDL-4.0による計算値と測定値との比(C/E)及びその不確かさは、$$^{24}$$Naで0.97$$sim$$1.07及び8.1$$sim$$11.0%、$$^{22}$$Naで1.03$$sim$$1.16及び23.3$$sim$$24.1%となった。$$^{23}$$Na(n,2n)断面積の違いにより、ENDF/B-VIII.0による$$^{22}$$Na放射化量計算値は、JENDL-4.0及びJEFF-3.3による計算値より40%大きくなることが明らかになった。また、$$^{22}$$Na放射化量を正確に評価するためには、$$^{22}$$Na自身の中性子捕獲効果を考慮することが重要であることが確認された。本実験データは、将来的なナトリウム冷却高速炉設計の信頼性向上のための計算手法の検証に活用であると判断できる。

論文

炉心変形反応度評価のための燃料集合体湾曲解析モデルの検証

堂田 哲広; 上羽 智之; 大釜 和也; 吉村 一夫; 根本 俊行*; 田中 正暁; 山野 秀将

日本機械学会関東支部第29期総会・講演会講演論文集(インターネット), 5 Pages, 2023/03

ナトリウム冷却高速炉の炉心変形による反応度をより現実的に評価するため、炉物理、熱流動、構造力学の連成解析による炉心変形反応度評価手法を開発した。本評価手法では、燃料集合体の湾曲を有限要素法のビーム要素でモデル化し、集合体ラッパ管のパッド部での隣接集合体間の接触をパッド部専用の要素でモデル化した解析手法を採用した。その検証として、過去に実施されたベンチマーク問題の集合体単体の自由熱湾曲及び炉心体系での集合体熱湾曲による隣接集合体間接触について計算し、本解析モデルによる解析結果が理論解またはベンチマークに参加した他機関の解析結果とよく一致することを確認した。この結果から、本解析モデルが集合体の熱湾曲を適切に計算できることを確認した。

論文

Evaluation of fuel reactivity worth measurement in the prototype fast reactor Monju

大釜 和也; 竹越 淳*; 片桐 寛樹; 羽様 平

Nuclear Technology, 208(10), p.1619 - 1633, 2022/10

 被引用回数:3 パーセンタイル:71.05(Nuclear Science & Technology)

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies are investigated through a comparison with calculations by using the latest methodology developed in Japan Atomic Energy Agency. Calculated-to-experiment values (C/Es) and their uncertainties of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the measurements and calculations were found consistent and reliable.

論文

Benchmark analysis of FFTF Loss of Flow Without Scram Test No.13 using fast reactor plant dynamics analysis code Super-COPD

浜瀬 枝里菜; 大釜 和也; 河村 拓己*; 堂田 哲広; 山野 秀将; 田中 正暁

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10

プラント動特性解析コードSuper-COPDの予測精度向上のため、米国中性子試験炉FFTFスクラム不作動流量喪失事象を対象としたIAEAベンチマークに参加している。ブラインド解析で課題として抽出された燃料集合体出口温度の再現性向上のため、自然循環時における集合体間熱移行及び集合体間流量再配分を精度よく評価可能な全炉心モデルを用いてプラント動特性解析を実施した。また、全炉心モデルと一点炉動特性モデルを連成した過渡解析の妥当性を確認するため、主要な反応度フィードバックであるGEM、炉心湾曲等を考慮した解析を実施した。その結果、2次ピーク時の温度を良好に再現するとともに、実測値の過渡挙動を概ね評価できることを確認した。

論文

A Design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

大釜 和也; 原 俊治*; 太田 宏一*; 永沼 正行; 大木 繁夫; 飯塚 政利*

Journal of Nuclear Science and Technology, 59(6), p.735 - 747, 2022/06

 被引用回数:0 パーセンタイル:33.72(Nuclear Science & Technology)

A metal fuel fast reactor core for high efficiency minor actinide (MA) transmutation was designed by loading silicon carbide composite material (SiC/SiC) which can improve sodium cooled fast reactor (SFR) core safety characteristics such as sodium void reactivity worth and Doppler coefficient due to neutron moderation. Based on a 750 MWe metal fueled SFR core concept designed in a prior work, the reactor core loading fuel subassemblies with SiC/SiC wrapper tubes and moderator subassemblies was designed. To improve the reactor core safety characteristics efficiently, three layers of SiC/SiC moderator subassemblies were loaded in the core by replacing 108 out of 393 fuel subassemblies with the moderators. The reactor core with approximately 20 wt% MA-containing metal fuel satisfied all safety design criteria and achieved the MA transmutation amount as high as 420 kg/GWe-y which is twice as high as that of the axially heterogeneous core with inner blanket and upper sodium plenum, and two-thirds of that of the accelerator-driven system.

論文

Evaluation of fixed absorber reactivity measurement in the prototype fast reactor Monju

大釜 和也; 片桐 寛樹; 竹越 淳*; 羽様 平

Nuclear Technology, 207(12), p.1810 - 1820, 2021/12

 被引用回数:3 パーセンタイル:47.54(Nuclear Science & Technology)

In the prototype fast breeder reactor Monju, fixed absorber worth was measured as a difference of core reactivity measured by control rod worth between cores with and without a single or three fixed absorbers. In the present paper, the measurements were evaluated in detail and its reliability and usefulness as a validation data were investigated through a comparison with calculations using the latest neutronics design methodology developed in Japan Atomic Energy Agency. Calculated-to-experiment values (C/Es) and their uncertainties of fixed absorber worth were 1.00$$pm$$0.05 and 1.02$$pm$$0.04, respectively. Through this study, the measurements and calculations were found consistent and reliable.

論文

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

太田 宏一*; 大釜 和也; 山野 秀将

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

A detailed core-bowing analysis code, ARKAS_cellule, has been developed. The detailed shell model applied to ARKAS_cellule was verified with a conventional beam model for the IAEA benchmark problems. As a result, ARKAS_cellule was properly verified for the thermal bowing analysis of the core. In addition, it was confirmed that ARKAS_cellule simulates the change in duct stiffness with the contact conditions.

論文

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.

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.

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

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 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

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:10.78(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

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。

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