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論文

Thermal analysis of the hydrogen release behavior of sodium hydride and kinetic analysis using master plot methods

土井 大輔

International Journal of Hydrogen Energy, 91, p.1245 - 1252, 2024/11

 被引用回数:0

Hydrogen is a major nonmetallic impurity in the coolant of sodium-cooled fast reactors (SFRs) during normal operation. A higher hydrogen concentration than the gas-liquid equilibrium has been transiently detected in the gas space of actual SFR plants. The presence of several sodium compounds can increase hydrogen generation; however, a thorough understanding of the thermal behavior of candidate reactions is lacking. Herein, thermal analysis reveals the hydrogen release behavior of sodium hydride. Mass spectrometry indicates hydrogen generation with decreasing sample mass, indicating thermal decomposition. Detailed kinetic analysis based on master plot methods indicates that the hydrogen release reaction occurred through a mechanism involving random nucleation and growth of nuclei. Furthermore, the reaction rate was newly formulated based on a kinetic model function representing the above mechanism and the Arrhenius-type reaction rate constant comprising an activation energy of 119.0 $$pm$$ 0.8 kJ mol$$^{-1}$$ and a frequency factor of 1.8 $$times$$ 10$$^{7}$$ s$$^{-1}$$. These findings will enable the numerical simulation of the hydrogen release behavior in SFRs.

論文

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

山野 秀将; 二神 敏; 日暮 浩一*

Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08

本論文は、信頼性を向上させた崩壊熱除去系について、第4世代炉国際フォーラムで開発された安全設計クライテリアと安全設計ガイドラインを我が国で最近設計されたナトリウム冷却高速炉へ適用したことを記述する。

論文

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。

論文

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

 被引用回数:2 パーセンタイル:41.04(Nuclear Science & Technology)

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.

論文

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

江村 優軌; 高井 俊秀; 菊地 晋; 神山 健司; 山野 秀将; 横山 博紀*; 坂本 寛*

Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

Boron carbide (B$$_4$$C)- stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B$$_4$$C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B$$_4$$C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B$$_4$$C-liquid SS reaction based on the reduced thickness of B$$_4$$C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B$$_4$$C-solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B$$_4$$C-liquid SS reaction, it was found that similar temperature dependency was identified between solid B$$_4$$C-liquid SS and solid B$$_4$$C-solid SS. Besides, the reaction rate constants of solid B$$_4$$C-liquid SS were smaller than those of solid B$$_4$$C-solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B$$_4$$C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid-solid contact for B$$_4$$C side/SS side.

論文

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。

論文

Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson-Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel

宮澤 健; 丹野 敬嗣; 今川 裕也; 橋立 竜太; 矢野 康英; 皆藤 威二; 大塚 智史; 光原 昌寿*; 外山 健*; 大沼 正人*; et al.

Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05

 被引用回数:0 パーセンタイル:0.00(Materials Science, Multidisciplinary)

This paper discusses the applicability of J.L. Straalsund et al.'s technique for combining the Larson-Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr- oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes in the $$alpha$$-phase, $$alpha$$/$$gamma$$-duplex, $$gamma$$-phase matrices, which are accumulated by the Japan Atomic Energy Agency so far. The technique is adequately applicable to 9Cr-ODS TMS cladding tubes. A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes (creep, tensile, temperature-transient-to-burst) is derived for 9Cr-ODS TMS cladding tubes. This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. The applicable range of the high-temperature strength equation is specified in this study and the upper limit temperature for the equation is found to be 1200$$^{circ}$$C. At temperatures higher than 1200$$^{circ}$$C, the coarsening and aggregation of nanosized oxide particles and the $$gamma$$ to $$delta$$ phase transformation are reported in previous studies. The high-temperature strength equation can be well applied to the creep and tensile strength in the $$alpha$$-phase matrix, the creep strength in the $$gamma$$-phase matrix and the temperature-transient-to-burst strength in both phases except for the low equivalent stress (43 MPa) at temperatures exceeding 1050$$^{circ}$$C. The mechanism of the notable consistency between creep and tensile strength in the $$alpha$$-phase matrix is discussed by analyzing the high-temperature deformation data in the light of existing deformation models.

論文

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

石田 真也; 深野 義隆; 飛田 吉春; 岡野 靖

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

 被引用回数:1 パーセンタイル:41.04(Nuclear Science & Technology)

To improve the safety of future SFRs, the development of SFRs with low void reactivity has been promoted. Small SFRs can have a negative void coefficient of reactivity, so the analysis of the CDA event sequence in small SFRs is valuable for the investigation of the reactor characteristics for the future research and development of SFRs. In this study, the typical initiating events of a CDA in small SFRs were evaluated with the computational code, SAS4A. The event progression of ULOF and UTOP in the low void reactivity reactor is found to be slow due to the effective operation of the negative reactivity feedback and the absence of significant positive reactivity insertion. No power excursion occurs in the initiating phase. In ULOF, the cladding melt and relocation behavior becomes more important for the evaluation of the event progression due to its positive reactivity.

論文

Validation of the hybrid turbulence model in detailed thermal-hydraulic analysis code SPIRAL for fuel assembly using sodium experiments data of 37-pin bundles

吉川 龍志; 今井 康友*; 菊地 紀宏; 田中 正暁; 大島 宏之

Nuclear Technology, 210(5), p.814 - 835, 2024/05

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

ナトリウム冷却高速炉安全性強化研究では、燃料ピンの構造健全性を評価するために各種運転条件下におけるワイヤスペーサ型燃料集合体内熱流動特性の解明が重要である。そこで有限要素法による集合体詳細熱流動解析コードSPIRALが開発されている。本研究では、SPIRALにおける壁近傍低Re数効果を考慮したハイブリッド型乱流モデルの妥当性を確認するために、層流-乱流遷移条件及び乱流条件を含む異なるRe数条件下の37本ピンバンドルナトリウム実験の再現解析を実施した。SPIRALによる予測された温度分布はナトリウム実験で測定され温度と一致した。以上によって、SPIRALにおけるハイブリッド型乱流モデルの広範囲Re数条件下ナトリウム冷却集合体熱流動評価への適用性を確認した。

論文

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

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。

論文

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

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

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$$_{2}$$), 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$$_{2}$$ 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.

論文

Estimation method for residual sodium amount on unloaded dummy fuel assembly

河口 宗道; 平川 康; 杉田 裕亮; 山口 裕

Nuclear Technology, 210(1), p.55 - 71, 2024/01

 被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)

本研究はもんじゅ模擬燃料集合体における残留ナトリウム膜及び塊の評価手法を開発し、実験的にピンの間のギャップを通ってナトリウムが流れる様子を確認した。ピン表面の残留ナトリウムの量は3種類の試験体((a)単ピン,(b)7本ピン集合体,(c)169本ピン集合体)を使って測定した。実験では、ピンの引き抜き速度やナトリウム濡れ性の改善により、残留ナトリウム量が劇的に増加することを明らかにした。さらに、169本ピンの実験により、短尺であるが模擬燃料集合体の実効的な残留ナトリウム量を測定し、模擬燃料集合体を通って流れるナトリウムの振舞いを確認した。開発した予測手法は、4つのモデル(粘性流れモデル、Landau-Levich-Derjaguin(LLD)モデル、Brethertonモデルに関わる実験式、管内の毛細管力モデル)から構築されており、その計算結果は実験の残留ナトリウム量と同程度な結果を与えた。ただし、ナトリウム濡れ性の不確かさはLLDモデルの予測値の約1.8倍である。この予測手法を使って、もんじゅの模擬燃料集合体に残留するナトリウム量を評価することができる。

論文

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.

論文

Development of Lagrangian particle method for temperature distribution formed by sodium-water reaction in a tube bundle system

小坂 亘; 内堀 昭寛; 岡野 靖; 柳沢 秀樹*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08

ナトリウム冷却型高速炉における蒸気発生器(SG)の安全性評価及び設計について、SG内伝熱管からの加圧水のリーク及びその後の事象進展の評価は重要である。解析コードLEAP-IIIは半経験式や1次元保存式などの低計算コストなモデルで構成されるために短い計算時間で水リーク率等を評価でき、革新炉開発における多様なSG設計の探求を加速させることが期待される。しかし、現在の温度分布評価モデルには、過度な保守性を示す場合があること、及びチューニングのために予備的な実験又は詳細な数値解析が必要とされて準備に時間がかかることに課題がある。これらを改善するため、より単純な計算原理に従い、機構論的な側面を持ちつつも高速計算可能なラグランジュ粒子法コードの開発に取り組んでいる。今回は、本粒子法コードに実装されている粒子ペア探索手法の効率化、及び粒子ペア探索を用いずに同等の結果を得るためのモデルの開発を行った。テスト解析を通して、これらのモデル改良による計算時間短縮効果を確認し、また、伝熱管破損判定に重要な伝熱管周囲の代表温度について、詳細な機構論的解析コード(SERAPHIM)による評価結果とよい一致を示すことを確認した。

論文

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

 被引用回数:1 パーセンタイル:33.61(Chemistry, Multidisciplinary)

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.

論文

Effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake

栗坂 健一; 西野 裕之; 山野 秀将

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05

本研究の目的は破損拡大抑制技術によって過大地震時の原子炉構造レジリエンス向上策の有効性評価手法を開発することである。安全上重要な機器・構造物のレジリエンス向上策によって耐震裕度が増すとみなす。同向上策の有効性を評価するため、炉心損傷頻度CDFを指標に選び、CDFの低減を 地震PRAによって定量化する。崩壊熱除去機能喪失に至る事故シーケンスがナトリウム冷却高速炉SFRの地震時CDFに有意な寄与を示す。また、同事故シーケンスは超高温を経て炉心損傷に至る。本研究では過大地震時の振動への対策のみならず超高温での対策も評価するよう手法を考案した。手法の適用性を検討するため、ループ型SFRを想定して試計算を実施した。仮定した範囲内では、レジリエンス向上策は設計地震動の数倍の地震までCDFを有意に低減する効果を示した。適用性検討を通じて、有効性評価手法が開発された。

論文

Analysis by hazard plotting on steam generator tube leak in sodium-cooled fast reactors Phenix and BN600

栗坂 健一

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

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トレンド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BN600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トレンドをハザードプロット法により定量化した。結果、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。長期トレンドとしては、Phenix及びBN600両者の管漏えい発生率は減少傾向を示した。この傾向は溶接及び運転条件の改善並びに初期故障の除去によるものと考えられる。

論文

Development of dynamic PRA methodology for external hazards in sodium-cooled fast reactor via applying Markov chain Monte Carlo method to severe accident analysis code; Assessment of accident management of assigning independent emergency diesel generators to each air cooler

Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Quantitative assessment of the effect of accident management on the various external hazards is essential in the nuclear safety analysis. This study aims to establish the dynamic probabilistic risk assessment methodology for sodium-cooled fast reactors that can consider the transient plant status under continuous external hazards with corresponding countermeasures operating stochastically. Specifically, the Continuous Markov chain Monte Carlo (CMMC) and Deterministic and Stochastic Petri Nets (DSPN) methods are newly applied to the severe accident analysis code, SPECTRA, which can conduct dynamic plant evaluation in the different severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. In the DSPN-CMMC-SPECTRA coupled frame, the latest safety functions of the plant components/systems can be stochastically determined by the DSPN-CMMC grounded on the current plant states under continuous hazard and the interaction between the multi-state components/systems; then, SPECTRA can evaluate the following plant state determined by the latest safety function of the components/systems. Therefore, the advantage of this newly developed DSPN-CMMC-SPECTRA frame is having the capability to quantitatively and stochastically evaluate the transient accident progressions that potentially lead to the core damage under the continuous external hazard scenario. As for the preliminary exam on the DSPN-CMMC-SPECTRA frame, one of the typical external hazards of continuous volcanic ashfall is selected in this research. In addition, the numerical investigation of alternative accident management' effects has also been carried out and quantitatively confirmed in this research.

論文

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$$_{2}$$, 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$$_{2}$$). This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium.

論文

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。

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