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Lu, K.; 高見澤 悠; Li, Y.; 眞崎 浩一*; 高越 大輝*; 永井 政貴*; 南日 卓*; 村上 健太*; 関東 康祐*; 八代醍 健志*; et al.
Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08
A probabilistic fracture mechanics (PFM) analysis code, PASCAL, has been developed by Japan Atomic Energy Agency for failure probability and failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. To strengthen the applicability of PASCAL, considerable efforts on verifications of the PASCAL code have been made in the past years. As a part of the verification activities, a working group consisted of different organizations from industry, universities and institutes, was established in Japan. In the early phase, the working group focused on verifying the PFM analysis functions for RPVs in pressurized water reactors (PWRs) subjected to pressurized thermal shock (PTS) events. Recently, the PASCAL code has been improved in order to run PFM analyses for both RPVs in PWRs and boiling water reactors (BWRs) subjected to a broad range of transients. Simultaneously, the working group initiated a verification plan for the improved PASCAL through independent PFM analyses by different organizations. Concretely, verification analyses for a PWR-type RPV subjected to PTS transients and a BWR-type RPV subjected to a low-temperature over pressure transient were performed using PASCAL. This paper summarizes those verification activities, including the verification plan, analysis conditions and results. Based on the verification studies, the reliability of PASCAL for probabilistic integrity assessments of Japanese RPVs was confirmed with confidence.
長谷 竹晃; 相樂 洋*; 小菅 義広*; 能見 貴佳; 奥村 啓介
Journal of Nuclear Science and Technology, 60(4), p.460 - 472, 2023/04
被引用回数：1 パーセンタイル：29.26(Nuclear Science & Technology)This paper provides an overview of the applicability of the Differential Die-Away Self-Interrogation (DDSI) technique for quantification of spontaneous fissile nuclides in fuel debris at the Fukushima Daiichi Nuclear Power Plants. In this research, massive fuel debris stored in a canister was evaluated, and the void space of the canister was assumed to be filled with water for wet storage and air for dry storage. The composition of fuel debris was estimated based on elements such as the inventory in the reactor core and operation history. The simulation results show that for wet storage, the DDSI technique can properly evaluate the neutron leakage multiplication and quantify spontaneous fissile nuclides with a total measurement uncertainty (TMU) of approximately 8%. For dry storage, the known-alpha technique, which was previously established, can be applied to quantify spontaneous fissile nuclides with a TMU of approximately 4%. In both cases, the largest uncertainty factor is the variation in water content in the canister. In the case of wet storage, the uncertainty could be significantly increased in cases where the fuel debris is extremely unevenly distributed in the canister.
中村 秀夫; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
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
The WGAMA activity achievements have been published as technical reports, becoming reference materials to discuss innovative methods, materials and technologies in the fields of thermal-hydraulics, computational fluid dynamics (CFD) and severe accidents (SAs). The International Standard Problems (ISPs) and Benchmarks of computer codes have been supported by a huge amount of the databases for the code validation necessary for the reactor safety assessment with accuracy. The paper aims to review and summarize the recent WGAMA outcomes with focus on new advanced reactor applications including small modular reactors (SMRs). Particularly, discussed are applicability of major outcomes in the relevant subjects of passive system, modelling innovation in CFD, severe accident management (SAM) countermeasures, advanced measurement methods and instrumentation, and modelling robustness of safety analysis codes. Although large portions of the outcomes are considered applicable, design-specific subjects may need careful considerations when applied. The WGAMA efforts, experiences and achievements for the safety assessment of operating nuclear power plants including SA will be of great help for the continuous safety improvements required for the advanced reactors including SMRs.
Co radiation source under an intense 
Cs radiation field using an application-specific CeBr
spectrometer suited for use in intense radiation fields冠城 雅晃; 島添 健次*; 加藤 昌弘*; 黒澤 忠弘*; 高橋 浩之*
Journal of Nuclear Science and Technology, 59(8), p.983 - 992, 2022/08
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Passive 
-ray spectroscopy is a useful technique for surveying the radioactive wastes and spent nuclear fuels under nuclear decommissioning. However, this method depends on material properties such as the activity, density, element, scale, and (especially) low-energy rays from 
U and 
Pu. The -decay lines of 
Cs, Cs, Co, and 
Eu occur at greater energies (than those of U and Pu), and these nuclides provide significant information on spent nuclear fuel and radioactive wastes. A CeBr
spectrometer with a small-volume crystal has been previously developed for use in intense radiation measurements. We exposed the spectrometer to radiation dose rates of 0.025, 0.151, 0.342, 0.700, and 0.954 Sv/h under a standard Cs radiation field. A 6.38 MBq Co calibration source was placed in front of the detector surface. Identification of the full energy peak at 1173 keV was impossible at dose rates higher than 0.700 Sv/h. However, subtraction of the Cs radiation spectra from the -ray spectra enabled the identification of the full energy peaks at 1173 and 1333 keV at dose rates of up to 0.954 Sv/h; the relative energy resolution at 1173 and 1333 keV was only slightly degraded at this dose rate.
Lu, K.; 勝山 仁哉; Li, Y.
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 10 Pages, 2020/08
In Japan, Japan Atomic Energy Agency has developed a probabilistic fracture mechanics (PFM) analysis code, PASCAL4, for probabilistic evaluation of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. Besides severe PTS events, however, transients associated with normal operations, such as the cooldown and heatup transients associated with reactor shutdown and startup, respectively, should also be considered in the integrity assessment of RPVs in both PWRs and boiling water reactors (BWRs). With regard to a heatup transient, because temperature is at its minimum, and tensile stress at its maximum on the RPV outer surface, outer surface crack and embedded crack near the RPV outer surface should be taken into account. To extend the applicability of PASCAL4, we improved the code to include analysis functions for these cracks. The improved PASCAL4 can be used to run PFM analyses of RPVs subjected to both cooldown (including PTS) and heatup transients. In this paper, improvements made to PASCAL4 are firstly described, including the incorporated stress intensity factor solutions and the corresponding calculation methods for vessel outer surface crack and embedded crack near the outer surface. Using the improved PASCAL4, PFM analysis examples for a Japanese BWR-type model RPV subjected to thermal transients including a low temperature overpressure event and a heatup transient are presented.
Lu, K.; 勝山 仁哉; Li, Y.; 宮本 裕平*; 廣田 貴俊*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*
Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06
Probabilistic fracture mechanics (PFM) is considered a promising methodology in assessing the integrity of structural components in nuclear power plants because it can rationally represent the influence parameters in their probabilistic distributions without over-conservativeness. In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) which enables the probabilistic integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Several efforts have been made to verify PASCAL4 to ensure that this code can provide reliable analysis results. In particular, a Japanese working group, which consists of different participants from the industry and from universities and institutes, has been established to conduct the verification studies. This paper summarizes verification activities of the working group in the past two years. Based on those verification activities, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs have been confirmed with great confidence.
三輪 一爾; 武田 聖司; 飯本 武志*
Radiation Protection Dosimetry, 184(3-4), p.372 - 375, 2019/10
被引用回数：0 パーセンタイル：0.01(Environmental Sciences)福島事故後の除染作業によって発生した除去土壌を再生資材として再利用する方針が環境省により示されている。有効な再利用用途の1つである海面埋立地では、施工時に溶存した放射性Csの他に土粒子に付着した放射性Csの海洋への流出が予想されるため、安全評価上、両形態の核種移行を評価できるモデルが必要となる。そこで本研究では、施工時および供用時の放射性Csの流出をモデル化し、海洋に流出した核種についてはOECDにより示されたSediment modelにより移行評価を行った。沿岸域における核種移行評価にSediment modelを用いることの妥当性を、福島沿岸域の実測値の再現計算により確認した。施工時および供用時の核種流出を評価するモデルおよびSediment modelをクリアランスレベル評価コードPASCLR2に組み込むことで、海洋へ流出した核種からの被ばく線量評価を行えるようにした。
岩崎 幸生; 岡 眞; 鈴木 渓*; 吉田 哲也*
International Journal of Modern Physics; Conference Series (Internet), 49, p.1960002_1 - 1960002_6, 2019/07
被引用回数：0 パーセンタイル：0.09(Astronomy & Astrophysics)P波のチャーモニウムメソンに強い磁場がかかったときにスペクトルおよび波動関数がどのように変化するかを磁場の強さの関数として解析した。
Lu, K.; 勝山 仁哉; Li, Y.; 吉村 忍*
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 9 Pages, 2019/07
Probabilistic fracture mechanics (PFM) methodology, which represents the influence parameters in their inherent probabilistic distributions, is deemed to be promising in the structural integrity assessment of pressure-boundary components in nuclear power plants. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) which can be used to evaluate the failure frequency of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analyses are performed for a Japanese model RPV using PASCAL4, and the effects of non-destructive examination and neutron fluence mitigation on failure frequency of RPV are quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for the structural integrity assessment of RPVs and can enhance the applicability of PFM methodology.
Lu, K.; 勝山 仁哉; Li, Y.; 宮本 裕平*; 廣田 貴俊*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
Probabilistic fracture mechanics (PFM) is considered as a promising methodology in the integrity assessment of structural components in a nuclear power plant since it can rationally represent the influence parameters in their inherent probabilistic distributions without over-conservativeness. In Japan, a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) has been developed by Japan Atomic Energy Agency, which can be used for structural integrity assessments of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Up till now, many efforts have been made on verifying the PASCAL4 code. Among them, a Japanese working group which is consisted of seven participants from industries, universities and institutes was established to conduct the verification studies. Based on verification activities during the past two years, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs were confirmed with great confidence. This paper summarizes the verification activities in this working group including the verification plan, analysis conditions and results.
岩崎 幸生; 岡 眞; 鈴木 渓*; 吉田 哲也*
Physics Letters B, 790, p.71 - 76, 2019/03
被引用回数：11 パーセンタイル：71.24(Astronomy & Astrophysics)ヘビークォークを含むメソンに強い磁場がかかったときにスペクトルおよび波動関数がどのように変化するかを磁場の強さの関数として解析した。
Lu, K.; 眞崎 浩一; 勝山 仁哉; Li, Y.
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07
In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessment of reactor pressure vessels (RPVs). The most recent release is PASCAL Version 4 (hereafter, PSACAL4) which can be used to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and pressurized thermal shock events. For the integrity assessment of RPVs, development of crack evaluation models is important. In this study, finite element analyses are performed firstly to verify the stress intensity factor calculations of cracks in PASCAL4. In addition, the applicability of the crack evaluation models in PASCAL4 such as the location of embedded cracks, crack shape and depth of surface cracks, and the increment of crack propagation is investigated. Based on sensitivity analyses of crack evaluation models for Japanese RPVs using PASCAL4, the effects of these evaluation models on failure frequency are clarified. From the analysis results, crack evaluation models recommended to the failure frequency evaluation for a Japanese model RPV are discussed.
Lu, K.; 眞崎 浩一; 勝山 仁哉; Li, Y.; 宇野 隼平*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWRs) for structural integrity assessment of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. By reflecting the latest knowledge and findings, the PASCAL code has been continuously improved. In this paper, the development of PASCAL Version 4 (hereafter, PASCAL4) is described. Several analysis functions incorporated into PASCAL4 for evaluating the failure frequency of RPVs are introduced, for example, the evaluation function of confidence level of failure frequency considering epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions and KI calculation methods considering complicated stress distributions, and the recent Japanese irradiation embrittlement prediction method. Finally, using PASCAL4, a PFM analysis example for a Japanese model RPV is presented.
Lu, K.; 勝山 仁哉; 宇野 隼平; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessments of reactor pressure vessels (RPVs) by considering the inherent probabilistic distributions of various influence factors. For practical applications, several evaluation models are improved, and have been implemented into the current PASCAL code. In this paper, the improvements of PASCAL are introduced firstly, such as the evaluation method for underclad cracks, treatments of the complicated welding residual stress distribution, and evaluation models for the warm pre-stressing effect. In addition, the effects of these improvements on failure probability or failure frequency of RPVs are investigated by performing PFM analyses for domestic RPVs using PASCAL. From the analysis results, the effects of the improved evaluation models are discussed.
生島 一樹*; 木谷 悠二*; 柴原 正和*; 西川 聡*; 古川 敬*; 秋田 貢一; 鈴木 裕士; 諸岡 聡
溶接学会論文集(インターネット), 35(2), p.75s - 79s, 2017/06
In this research, to investigate the effect of shot peening on operation, an analysis method to predict the behavior of stress distribution on shot peening was proposed. In the proposed system, the load distribution on the collision of shots was modeled, and it was integrated with the dynamic analysis method based on the Idealized explicit FEM (IEFEM). The thermal elastic plastic analysis method using IEFEM was applied to the analysis of residual stress distribution of multi-pass welded pipe joint. The computed residual stress distribution was compared with the measured residual stress distribution using X-ray diffraction (XRD). As a result, it was shown that the both welding residual stress distribution agree well with each other. Considering the computed welding residual stress distribution, the modification of stress distribution due to shot peening was predicted by the proposed analysis system.
高松 邦吉; 松元 達也*; 守田 幸路*
Annals of Nuclear Energy, 96, p.137 - 147, 2016/10
被引用回数：5 パーセンタイル：43.12(Nuclear Science & Technology)東京電力の福島第一原子力発電所事故(以下、福島事故)後、深層防護の観点から炉心損傷の防止対策が重要になった。そこで、動的機器および非常用電源等を必要とせず、福島事故のようにヒートシンクを喪失することのない、受動的安全性を持つ原子炉圧力容器の冷却設備を提案する。本冷却設備は安定して冷却できるため、定格運転時の一部の放出熱、および炉停止後の一部の崩壊熱を、常に安定的に受動的に除去できる。特に事故時において、本冷却設備が持つ冷却能力の範囲まで崩壊熱が減少した際、それ以降は非常用電源等が必要なくなり、長期間(無限時間)に渡って受動的な除熱が可能となる。一方、本冷却設備の優れた除熱性能を示すために、等倍縮小した除熱試験装置を製作し、ふく射および自然対流に関する実験条件をグラスホフ数を用いて決定することもできた。
高松 邦吉; 松元 達也*; 守田 幸路*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1250 - 1257, 2016/04
東京電力の福島第一原子力発電所事故(以下、福島事故)後、深層防護の観点から炉心損傷の防止対策が重要になった。そこで、動的機器および非常用電源等を必要とせず、福島事故のようにヒートシンクを喪失することのない、受動的安全性を持つ原子炉圧力容器の冷却設備を提案する。本冷却設備は安定して冷却できるため、定格運転時の一部の放出熱、および炉停止後の一部の崩壊熱を、常に安定的に受動的に除去できる。特に事故時において、本冷却設備が持つ冷却能力の範囲まで崩壊熱が減少した際、それ以降は非常用電源等が必要なくなり、長期間(無限時間)に渡って受動的な除熱が可能となる。一方、本冷却設備の優れた除熱性能を示すために、等倍縮小した除熱試験装置を製作し、ふく射および自然対流に関する実験条件をグラスホフ数を用いて決定することもできた。
高松 邦吉; Hu, R.*
Annals of Nuclear Energy, 77, p.165 - 171, 2015/03
被引用回数：14 パーセンタイル：72.94(Nuclear Science & Technology)東京電力の福島第一原子力発電所事故(以下、福島事故)後、深層防護の観点から炉心損傷の防止対策が重要になった。安全上優れた特性を有する冷却設備に関する研究は、極めて重要なテーマである。そこで、動的機器および非常用電源等を必要とせず、福島事故のようにヒートシンクを喪失することのない、受動的安全性を持つ原子炉圧力容器の冷却設備を提案する。本冷却設備は変動がなく、安定して冷却できるため、定格運転時の一部の放出熱、および炉停止後の一部の崩壊熱を、常に安定的に受動的に除去できることがわかった。特に事故時において、本冷却設備が持つ冷却能力の範囲まで崩壊熱が減少した際、それ以降は非常用電源等が必要なくなり、長期間(無限時間)に渡って受動的な除熱が可能となる。
波多江 仰紀; 内藤 磨; 北村 繁; 佐久間 猛*; 濱野 隆*; 中塚 正大*; 吉田 英次*
Journal of the Korean Physical Society, 49, p.S160 - S164, 2006/12
誘導ブリルアン散乱位相共役鏡を応用し、トムソン散乱計測の測定性能改善を図った。液体フロン化合物を用いた位相共役鏡はレーザー平均出力145W(50Hz)の入力で95%以上の反射率を示した。トムソン散乱への直接的な応用としては、位相共役鏡によりレーザービームを往復させ、迷光を著しく増加させることなく散乱光を倍増させる手法(ダブルパス散乱)を開発した。初期実験ではJT-60に位相共役鏡を取り付けダブルパス散乱させた結果、散乱光を1.6倍に増加させることができた。ダブルパス散乱を発展させ、一対の位相共役鏡間にレーザー光を閉じ込め、数倍以上の散乱光を発生することができるマルチパス散乱も考案した。また、散乱光のS/N改善のため、位相共役鏡を既存のレーザー装置に組み込み、レーザー装置の高出力化を行った。位相共役鏡は高出力増幅器で誘起される波面歪みが効果的に補正し、レーザー出力が当初の8倍を超える368W(7.4J
50Hz)に到達した。この結果からITERの周辺トムソン散乱用レーザーで必要とされる5J, 100Hzの出力を得る見通しがついた。これらを踏まえ位相共役鏡を搭載した、最適化されたITER用レーザーシステム、さらに高空間分解型LIDARトムソン散乱用レーザーシステムの設計・検討を行った。
波多江 仰紀; 近藤 貴; 内藤 磨; 中塚 正大*; 吉田 英次*
Proceedings of 12th International Symposium on Laser-Aided Plasma Diagnostics (LAPD-12) (CD-ROM), 6 Pages, 2005/09
JT-60の近年のトムソン散乱計測の開発研究について報告を行う。非協同トムソン散乱計測では、誘導ブリルアン散乱位相共役鏡を応用し、トムソン散乱計測の測定性能改善を図った。トムソン散乱への直接的な応用としては、位相共役鏡によりレーザービームを往復させ、迷光を著しく増加させることなく散乱光を倍増させる手法(ダブルパス散乱)を開発した。ダブルパス散乱を発展させ、一対の位相共役鏡間にレーザー光を閉じ込め、数倍以上の散乱光を発生することができるマルチパス散乱も考案した。散乱光のS/N改善のため、位相共役鏡を既存のYAGレーザー装置に組み込み、レーザー装置の高出力化も行った。位相共役鏡は高出力増幅器で誘起される波面歪みを効果的に補正し、レーザー出力が当初の8倍を超える368W(7.4J50Hz)に到達した。これを踏まえ位相共役鏡を搭載した、最適化されたITER用レーザーシステムの設計・検討を行った。また、JT-60では核燃焼プラズマのイオン温度や高速
粒子の振る舞いを測定する炭酸ガスレーザを用いた協同トムソン散乱計測の開発を行っている。初期実験結果に基づきレーザーの縦モードの質,ヘテロダイン受信機の電気ノイズの低減,光軸調整の改善を進め、2005年冬から始まるJT-60実験では改善された計測装置で測定を行う予定である。
