Application of the GIF safety design criteria and safety design guidelines on reactor shutdown system to next generation sodium-cooled fast reactor in Japan

山野 秀将; 二神 敏; 柴田 明裕*

Proceedings of Advanced Reactor Safety (ARS 2024), p.151 - 160, 2024/08

本研究では、動的安全保護系に関して、第4世代国際フォーラムで開発された安全設計クライテリアとガイドラインを我が国で最近に設計されたナトリウム冷却高速炉へ適用した。

シビアアクシデント統合評価解析コードSPECTRAを用いた炉心損傷解析

石田 真也; 内堀 昭寛; 岡野 靖

第28回動力・エネルギー技術シンポジウム講演論文集(インターネット), 4 Pages, 2024/06

本研究では、炉心損傷事故の起因過程から遷移過程までの一貫解析も可能な炉心損傷挙動評価モジュールの開発を行い、ナトリウム冷却高速炉のシビアアクシデント時の原子炉全体の挙動を一貫して評価する解析コードSPECTRAに導入した。本モジュールを含むSPECTRAの統合的な妥当性確認の一環として、混合酸化物(MOX)燃料炉心における炉心流量喪失時原子炉停止機能喪失事象(ULOF)を対象とした解析を実施し、冷却材の沸騰から燃料ピンの破損、損傷領域の拡大に至るまでの高速炉の炉心損傷事故を評価するための機能がSPECTRAに備わっていることを確認した。

Fundamental evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

山本 智彦; 加藤 篤志; 早川 雅人; 下山 一仁; 荒 邦章; 畠山 望*; 山内 和*; 江田 優平*; 由井 正弘*

Nuclear Engineering and Technology, 56(3), p.893 - 899, 2024/03

https://doi.org/10.1016/j.net.2023.10.043

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju." However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. In parallel, we observed experimentally that the fine bubbles of H stably existed in the liquid sodium than expected before. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

部門設立30周年記念出版Vol.3; ナトリウム冷却高速炉の開発; 「常陽」「もんじゅ」から実証炉へ

大野 修司; 前田 誠一郎

第27回動力・エネルギー技術シンポジウム講演論文集(インターネット), 3 Pages, 2023/09

The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published. The book is a collection of the past experience of design, construction, and operation of the experimental reactor "Joyo" and the prototype reactor "Monju", the latest knowledge including related research and development activities and technology for SFR designs, and the future prospects of SFR development in Japan, looking back the history of development of fast reactors started in the early 1960s. The development of sodium-cooled fast reactor in Japan, which contributes to energy security and high-level waste reduction, is reaching to the stage where demonstration reactor will be deployed based on the experience of "Joyo" and "Monju" design, construction and operation. The present report introduces outlines of experiences, results and activities accumulated through these reactors and R&Ds for demonstration reactor.

The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

Li, C.-Y.; Wang, K.*; 内堀 昭寛; 岡野 靖; Pellegrini, M.*; Erkan, N.*; 高田 孝*; 岡本 孝司*

Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07

https://doi.org/10.3390/app13137705

For a sodium-cooled fast reactor, the capability for stable cooling and avoiding re-criticality on the debris bed is essential for achieving in-vessel retention when severe accidents occur. However, an unexploited uncertainty still existed regarding the compound effect of the heterogeneous configuration and dynamic particle redistribution for the debris bed's criticality and cooling safety assessment. Therefore, this research aims to develop a numerical tool for investigating the effects of the different transformations of the heterogeneous configurations on the debris bed's criticality/cooling assessment. Based on the newly proposed methodology in this research, via integrating the Discrete Element Method (DEM) with Computational Fluid Dynamics (CFD) and Monte-Carlo-based Neutronics (MCN), the coupled CFD-DEM-MCN solver was constructed with the originally created interface to integrate two existing codes. The effects of the different bed configurations' transformations on the bed safety assessments were also quantitively confirmed, indicating that the effect of the particle-centralized fissile material had the dominant negative effect on the safety margin of avoiding re-criticality and particle re-melting accidents and had a more evident impact than the net bed-centralized effect. This coupled solver can serve to further assess the debris bed's safety via a multi-physics simulation approach, leading to safer SFR design concepts.

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

JAEA-Data-Code-2023-003.pdf:1.66MB

次世代高速炉は、従来炉よりも高い炉心取出燃焼度を目指しているため、炉心核設計の高度化が求められる。そのため、燃焼核特性解析では、計算コストを抑えつつ十分な計算精度が得られる適切な解析条件が必要とされる。そこで、次世代高速炉の燃焼核特性の計算精度に及ぼす解析条件の影響を、中性子エネルギー群、中性子輸送理論、空間メッシュに着目して調査した。本検討では燃焼核特性として、平衡サイクルにおける臨界性、燃焼反応度、制御棒価値、増殖比、集合体単位の出力分布、最大線出力、ナトリウムボイド反応度、ドップラー係数を取り扱った。検討の結果、エネルギー群を18群とし、拡散近似を用いて1集合体あたり6メッシュ分割して、エネルギー群、空間メッシュ、輸送効果の補正係数を適用することが最適であることが分かった。

Evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

山本 智彦; 加藤 篤志; 早川 雅人; 下山 一仁; 荒 邦章; 畠山 望*; 山内 和*; 江田 優平*; 由井 正弘*

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 6 Pages, 2023/04

In the secondary cooling system of sodium-cooled fast reactor (SFR), a rapid detection of hydrogen explosion due to sodium-water reaction by water leakage from heat exchanger tube is steam generator (SG) is important in terms of safety and property protection. For the hydrogen detection, Ni-membrane hydrogen detectors using atomic transmission phenomenon were used in Japanese proto-type sodium-cooled fast reactor "Monju". However, during the plant operation, many alarms of water leakage were occurred without sodium-water reaction in relation to temperature up and down. The authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal operation and the hydrogen generated by the sodium-water reaction and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). As the results of theoretical estimation, dissolved H or NaH, rather than H, is the predominant form of the background hydrogen in liquid sodium, and hydrogen produced in large amounts by sodium-water reaction can exist stably as fine bubbles with a NaH layer on their surface. Currently, the authors study the new hydrogen detector system focusing on the difference between the background hydrogen (dissolved H) and the hydrogen by sodium-water reaction (fine bubbles H). This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium.

Chapter 5, Sodium-cooled Fast Reactor (SFRs)/ Chapter 12, Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan

久保 重信; 近澤 佳隆; 大島 宏之; 上出 英樹

Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03

第4世代原子炉の最近の開発進捗を網羅するよう取りまとめ、2016年発行の第1版から第2版として更新したもの。著者らは、本ハンドブックの第5章ナトリウム冷却高速炉ならびに第12章日本における第4世代ナトリム冷却高速炉概念の章を担当し、それぞれナトリウム炉の特徴と安全性を含む新しい技術展開、日本におけるナトリウム炉開発の成果と革新技術、東京電力福島第一原子力発電所事故を受けての安全性強化の取組を示した。

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

https://doi.org/10.1080/00295450.2022.2045532

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.

Development of safety design criteria and safety design guidelines for Generation IV sodium-cooled fast reactors

二神 敏; 久保 重信; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10

In the framework of the GIF, an effort to develop Safety Design Criteria (SDC) for SFR systems was initiated in 2011. For this purpose, an SDC task force (SDC-TF) was formulated in July 2011. The SDC-TF members consist of representatives of CIAE (China), CEA (France), JAEA (Japan), KAERI, KINS (Republic of Korea), IPPE (Russia), ANL, INL, ORNL (United States of America), EC and IAEA. This paper describes the outline of the SDC and SDGs contents and its development background as shown above. These SDC and SDGs refer related IAEA safety standards, such as SSR-2/1 Safety of Nuclear Power Plants: Design, SSG-52 Design of the Reactor Core for Nuclear Power Plants. This paper focuses on both technology neutral aspects, which are common parts between the SDC/SDG and IAEA standards, and SFR specific aspects.