Current status of high temperature gas-cooled reactor development in Japan

永塚 健太郎; 野口 弘喜; 長住 達; 野本 恭信; 清水 厚志; 佐藤 博之; 西原 哲夫; 坂場 成昭

Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08

https://doi.org/10.1016/j.nucengdes.2024.113338

高温ガス炉は固有の安全性を有し、二酸化炭素を排出することなく大量の水素や高温の熱供給が可能なことから、産業分野の脱炭素化に貢献できる。本報では、原子力機構で進めるHTTR(高温工学試験研究炉)を利用した炉心強制冷却喪失(LOFC)試験等の研究開発成果に加え、現在設計を進めるHTTRを用いた水素製造実証試験(HTTR-熱利用試験)の計画を紹介する。加えて、2030年代後半の運転開始に向け、基本設計が進められている高温ガス炉実証炉計画を紹介する。

The Development of Petri Net-based continuous Markov Chain Monte Carlo methodology applying to dynamic probability risk assessment for multi-state resilience systems with repairable multi-component interdependency under longtermly thereat

Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖

Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07

https://doi.org/10.1080/00223131.2023.2290067

For all the nuclear reactor systems, quantitative assessment of the accident management (AM) effects against long-term external hazards became one of the essential issues after the lesson learned from the Fukushima Daiichi Nuclear Power Plant accident. However, the influence from the safety systems' stochastic and dynamic shifting between multiple working states, which is related to the interaction with the adjacent components/systems in general, has not been accounted for yet. Therefore, this research aims to develop a dynamic probability risk assessment tool considering repairable multi-component interdependency for investigating the AM influences on the multi-state safety systems under long-term external hazards. Based on the newly proposed methodology in this research via integrating the Petri Net (PN) model with the continuous Markov chain Monte Carlo (CMMC) method, a framework applying PN-CMMC methodology to a severe accident analysis code, SPECTRA, had been originally constructed. Different AM influences on the multi-state decay heat removal systems against long-term volcanic ashfall were also quantitatively confirmed, indicating that halving the repairing time is more influential in suppressing the core damage frequency than doubling the number of adjacent electricity support systems. Therefore, the PN-CMMC-SPECTRA framework can further assess the uncharted dynamic multi-state concerns, leading to a safer AM strategy.

消防自動車を用いたHTTRのBDBA拡大防止対策

島崎 洋祐; 地代所 達也; 石井 俊晃; 猪井 宏幸; 飯垣 和彦

JAEA-Technology 2024-005, 23 Pages, 2024/06

JAEA-Technology-2024-005.pdf:5.53MB

HTTRでは、新規制基準への対応の一環として新たに多量の放射性物質等を放出するおそれのある事故(BDBA)の想定を行うとともに、BDBAの拡大防止対策を定めた。このうち、使用済燃料貯蔵プールに係る冷却水漏洩によって発生するBDBAの拡大防止対策においては、大洗研究所の消防自動車をBDBAの拡大防止対策機器として選定し、揚水性能等の要求性能を定めて検査で確認した。これにより、消防自動車は使用前事業者検査に合格し、HTTRの運転再開に貢献した。

Interaction of solute manganese and nickel atoms with dislocation loops in iron-based alloys irradiated with 2.8 MeV Fe ions at 400 C

Nguyen, B. V. C.*; 村上 健太*; Chena, L.*; Phongsakorn, P. T.*; Chen, X.*; 橋本 貴司; Hwang, T.*; 古澤 彰憲; 鈴木 達也*

Nuclear Materials and Energy (Internet), 39, p.101639_1 - 101639_9, 2024/06

https://doi.org/10.1016/j.nme.2024.101639

In reactor pressure vessel materials, the formation of Mn- and Ni-rich nanoclusters is a major cause of neutron irradiation embrittlement. The segregation of these solute atoms into dislocation loops has attracted attention as a mechanism to accelerate solute clustering. In this study, the behaviors of solute Mn and Ni atoms in Fe-0.6wt.%Ni, Fe-1.4wt.%Mn, and Fe-1.4wt.%Mn-0.6wt.%Ni alloys irradiated at 400 C up to 3 dpa were analyzed using three-dimensional atom probe tomography. Solute atom clusters were observed in all materials, and their shapes were spherical, flat, and torus in FeNi, FeMn, and FeMnNi, respectively. In ternary alloy FeMnNi, Mn and Ni atoms were concentrated in the sample in the form of arcs, and the orientation of the plane containing the arcs was estimated by comparing field desorption images. The size, number density, and orientation of this structure were found to be in good agreement with those of both types of dislocation loops, namely, b = 1/2 111 and b = 100, identified in a previous study using the same material. The positions of Ni and Mn enrichment did not fully overlap. Ni atoms tended to be concentrated more in the inner part of the loop than the Mn atoms. Mn atoms were enriched only in the vicinity of the dislocation loops, whereas Ni atoms showed a higher concentration inside the dislocation loops than in the bulk.

Time-dependent change in occurrence rate of steam generator tube leak in sodium-cooled fast reactors; Phenix and BN-600

栗坂 健一

Mechanical Engineering Journal (Internet), 11(2), p.23-00377_1 - 23-00377_14, 2024/04

https://doi.org/10.1299/mej.23-00377

本研究は、既存のナトリウム冷却高速炉SFRにおける観測データに基づき蒸気発生器SG伝熱管漏えいの発生率の時間変化を把握することを目的とする。対象とするSFRは仏国のPhenix及び露国のBN-600である。公開文献を基に、管-管板溶接数、管-管溶接数、母材の伝熱面積、SG運転時間、SG伝熱管漏えい発生日、漏えい位置、漏えいモジュールの交換などの漏えい後の是正措置を調べた。これらのデータを踏まえ、漏えい発生までの運転時間を推定し、上記部位毎に伝熱管漏えい発生率の時間変化をハザードプロット法により定量化した。結果、Phenix及びBN-600両者の管漏えい発生率は減少傾向を示した。Phenixの傾向は溶接及び運転条件の改善によるものと考えられる。BN-600については運転初期に破損に拡大した初期欠陥が原因と考えられ、漏えい後特別な対策が講じられていないことから単純に発生率が時間とともに減少したと考えられる。またPhenixの管-管溶接部の漏えい発生率は繰り返し熱応力によって短期に増大する傾向が示された。

An Analytical model to decompose mass transfer and chemical process contributions to molecular iodine release from aqueous phase under severe accident conditions

Zablackaite, G.; 塩津 弘之; 城戸 健太朗; 杉山 智之

Nuclear Engineering and Technology, 56(2), p.536 - 545, 2024/02

https://doi.org/10.1016/j.net.2023.10.030

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

Measurements of neutron capture cross-sections for nuclides of interest in decommissioning; Sc, Cu, Zn, Ag, and In

中村 詔司; 芝原 雄司*; 遠藤 駿典; Rovira Leveroni, G.; 木村 敦

Journal of Nuclear Science and Technology, 16 Pages, 2024/00

https://doi.org/10.1080/00223131.2024.2346347

廃止措置においては、生成放射能の評価に資する核データとして、中性子捕獲断面積を整備する必要がある。本研究では、整備すべき対象核種のうちSc, Cu, Zn, Ag及びInを選定し、KURのTC-Pnを用いてそれらの熱中性子捕獲断面積測定を行った。その結果、熱中性子捕獲断面積の結果が、以下のとおり得られた:Sc(n, )Sc反応は27.180.28 barn、Cu(n, )Cu反応は4.340.06 barn、Zn(n, )Zn反応は0.7190.011 barn、Ag(n, )Ag反応は4.050.05 barn、そしてIn(n, ) In反応は8.530.27 barn。ScとZnの結果は、これまでに報告されている評価値を誤差範囲で支持するが、一方、他の核種については評価値と異なる結果となった。得られた結果は、生成量評価にはもちろん、これらの核種を中性束モニタとして利用する場合に用いることが考えられる。

Measurements of capture cross-section of Nb by activation method and half-life of Nb by mass analysis

中村 詔司; 芝原 雄司*; 遠藤 駿典; 木村 敦

Journal of Nuclear Science and Technology, 60(11), p.1361 - 1371, 2023/11

https://doi.org/10.1080/00223131.2023.2198526

廃止措置の対象核種の中からNbについて、熱中性子捕獲断面積()および共鳴積分(I)を放射化法により測定した。また、Nbの半減期を質量分析により測定した。ニオブ93試料は、京都大学複合原子力科学研究所の研究炉に整備されている水圧輸送艦を用いて照射した。金-アルミ合金線、およびコバルト-アルミ合金線を用いて、照射位置における中性子束成分を測定した。厚さ25mのガドリニウム箔を用いた照射も行って、熱および熱外中性子による反応率の寄与をより分けた。ガドリニウムの厚さから、カットオフ・エネルギーは0.133eVに設定した。ニオブ試料中に含まれている不純物により生成されるTaの放射能を減衰させるために、約2年間ニオブ試料を冷却した。線スペクトロスコピーにより照射されたニオブ試料およびモニタの生成放射能を測定した。Westcottコンベンションに基づいて解析を行い、およびIを、それぞれ1.110.04barnおよび10.50.6barnと導出することができた。線測定の後に、ニオブ試料の質量分析を行い、反応率を導出した。線スペクトロスコピーと質量分析で得られたデータを組み合わせることにより、Nbの半減期を、(2.000.15)10年と導出することができた。

Molecular dynamics analysis of reactor graphite for preparing thermal neutron scattering law

沖田 将一朗; 後藤 実

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

The recently released JENDL-5 and ENDF/B-VIII.0 have adopted porosity-dependent thermal neutron scattering law (TSL) data for reactor graphite, and they improve neutronic calculation accuracy of criticality for graphite-moderated cores. Currently, we can only handle neutronic calculations for three graphite porosities of 0%, 10%, and 30%. The uncertainties associated with the difference between the porosity of actual reactor graphite (20%) and the porosity remains. Toward the future update of JENDL-5, we are planning to preparing new TSL data of reactor graphite. As a first step, it is essential to evalute phonon density state distribution of reactor graphite. In this study, in order to evalute it, molecular dynamic (MD) analysis is performed for three MD models: ideal crystalline graphite (Ideal model), 20%-porous reactor graphite with monoatomic random pore (Monoatomic random model), and 20%-porous reactor graphite with atomic cluster random pore (Cluster random model). The ideal crystalline graphite is modeled without any pores for reference. The 20%-porous reactor graphite with monoatomic random pore is modeled by randomly removing atoms from the ideal crystalline graphite. The 20%-porous reactor graphite with cluster random pore is modeled by randomly removing atomic clusters of approximately 2 nm in diameter from the ideal crystalline graphite. Their interatomic interactions are on the basis of Reactive Empirical Bond Order (REBO) potential. Velocity autocorrelation functions and phonon density of states distributions are calculated for these models. For validation, specific heat for each model is evaluated, and they are compred with experimental values.

Main outputs from the OECD/NEA ARC-F Project

丸山 結; 杉山 智之*; 島田 亜佐子; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08

http://dx.doi.org/10.13182/NURETH20-40302

The Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi Nuclear Power Station (FDNPS) (ARC-F) project was initiated in January 2019 for three years with 22 signatories from 12 countries. Three main tasks were implemented in the ARC-F project, which were relevant to 1) refinement of analysis for accident scenarios and associated fission product (FP) transport and dispersion, 2) compilation and management of data and information, and 3) discussion for the next-phase project. Various activities were performed in Task 1, covering improvement of analysis for accident scenarios, and in-depth analyses for specific phenomena such as in-vessel melt progression, molten core/concrete interaction, FP transport and source term, hydrogen combustion and atmospheric dispersion of FPs. Through these studies, analyses for accident scenarios with severe accident codes were refined and important phenomena with large uncertainties were clarified. In order to share well selected and organized information from the FDNPS with the project partners, two databases, information source database and sample database, were built under Task 2. The analysis techniques including the separation of iodine species were developed also in Task 2 and applied to the analysis of FPs in several samples taken from the FDNPS. The next-phase project was discussed in Task 3, resulting in launching the Fukushima Daiichi Nuclear Power Station Information Collection and Evaluation (FACE) project. The FACE project officially started in July 2022 with the participation of 23 organizations from 12 countries and the European Commission.