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Analysis of Fukushima-Daiichi Nuclear Power Plant Unit 3 pressure data and obtained insights on accident progression behavior

佐藤 一憲

Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11

The D/W (Drywell) and S/C (Suppression Chamber) pressure data of Fukushima-Daiichi Nuclear Power Plant Unit 3 was analyzed in depth. This analysis provided valuable information related to the accident progression behavior on one hand, and gave a hint for understanding of the debris-to-coolant heat transfer when fuel debris relocated to the pedestal on the other hand. In this unit, the D/W and S/C pressure increased and decreased cyclically with a relationship, which seems to have been dependent on the composition of vapor and non-condensable gases in the S/C cover gas region. Based on this characteristic, the vapor pressure in the S/C cover gas region was evaluated for two pressure decrease cycles during and after the expected debris relocation to the pedestal respectively. This evaluation allowed an understanding that the S/C vapor pressure increased due to the heat transfer from the debris relocated to the pedestal.




JAEA-Review 2021-011, 86 Pages, 2021/08


工務技術部は、原子力科学研究所及びJ-PARC の水、電気、蒸気、排水等のユーティリティ施設、原子炉施設及び核燃料物質取扱施設内の特定施設(受変電設備、非常用電源設備、気体・液体廃棄設備、圧縮空気設備)並びに一般施設内の機械室設備の運転、保守管理を担っている。さらに、建物・設備の補修・改修工事及び点検・整備業務、電子装置及び機械装置の工作業務を行ってきた。本報告書は、令和元年度の工務技術部の業務実績の概況、主な管理データ及び技術開発の概要を記録したものであり、今後の業務の推進に役立てられることを期待する。


熱流動とリスク評価,1; リスク評価における熱流動解析の役割

丸山 結; 吉田 一雄

日本原子力学会誌ATOMO$$Sigma$$, 63(7), p.517 - 522, 2021/07



Transient response of LWR fuels (RIA)

宇田川 豊; 更田 豊志*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

This article aims at providing a general outline of fuel behavior during a reactivity-initiated accident (RIA) postulated in light water reactors (LWRs) and at showing experimental data providing technical basis for the current RIA-related regulatory criteria in Japan.


Extension of PASCAL4 code for probabilistic fracture mechanics analysis of reactor pressure vessel in boiling water reactor

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.


Numerical simulation of two-phase flow in 4$$times$$4 simulated bundle

小野 綾子; 山下 晋; 鈴木 貴行*; 吉田 啓之

Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06



Advancement of elemental analytical model in LEAP-III code for tube failure propagation

内堀 昭寛; 柳沢 秀樹*; 高田 孝; Li, J.*; Jang, S.*

Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06



Boron chemistry during transportation in the high temperature region of a boiling water reactor under severe accident conditions

三輪 周平; 高瀬 学; 井元 純平; 西岡 俊一郎; 宮原 直哉; 逢坂 正彦

Journal of Nuclear Science and Technology, 57(3), p.291 - 300, 2020/03

 被引用回数:2 パーセンタイル:42.44(Nuclear Science & Technology)



Irradiation growth behavior of improved Zr-based alloys for fuel cladding

天谷 政樹; 垣内 一雄; 三原 武

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

New fuel cladding alloys of which composition was changed from conventional ones have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rod and/or fuel assembly during normal operation, the irradiation growth behavior of the improved Zr-based alloys for light-water reactor fuel cladding was investigated. The coupon specimens were prepared from fuel cladding tubes with various kinds of improved Zr-based alloys. The specimens were loaded into test rigs and had been irradiated in the Halden reactor in Norway under several coolant temperature conditions up to a fast-neutron fluence of $$sim$$7.8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$ 1 MeV). Irradiation conditions such as specimen temperatures had been continuously monitored during the irradiation. During and after the irradiation, the amount of irradiation growth of each specimen was evaluated as a part of the interim and final inspections. The effect of the difference in alloy composition on the amount of irradiation growth seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication and the irradiation temperature were the same.


Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 宇田川 豊; 三原 武; 天谷 政樹; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

A pulse-irradiation test CN-1 on a high-burnup MOX fuel with M5$$^{TM}$$ cladding was conducted at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Although the transient signals obtained during the pulse-irradiation test did not show any signs of the occurrence of PCMI failure, the failure of the test fuel rod was confirmed from the visual inspection carried out after test CN-1. Analyses using fuel performance codes FEMAXI-8 and RANNS were also performed in order to investigate the fuel behavior during normal operation and pulse-irradiation regarding the test fuel rod of CN-1, and the results were consistent with this observation result. These experimental and calculation results suggested that the failure of test fuel rod of CN-1 was not caused by hydride-assisted PCMI but high-temperature rupture following the increase in rod internal pressure. The occurrence of this failure mode might be related to the ductility remained in the M5$$^{TM}$$ cladding owing to its low content of the hydrogen absorbed during normal operation.


Study on the two-phase flow in simulated LWR fuel bundle by CFD code

小野 綾子; 山下 晋; 鈴木 貴行*; 吉田 啓之

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08



Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

宮原 直哉; 三輪 周平; 堀口 直樹; 佐藤 勇*; 逢坂 正彦

Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02

 被引用回数:4 パーセンタイル:68.11(Nuclear Science & Technology)




安全研究・防災支援部門 安全研究センター

JAEA-Review 2018-022, 201 Pages, 2019/01




燃料挙動解析コードFEMAXI-8の開発; 軽水炉燃料挙動モデルの改良と総合性能の検証

宇田川 豊; 山内 紹裕*; 北野 剛司*; 天谷 政樹

JAEA-Data/Code 2018-016, 79 Pages, 2019/01




Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon

内堀 昭寛; 渡部 晃*; 高田 孝; 大島 宏之

Nuclear Technology, 205(1-2), p.119 - 127, 2019/01

 被引用回数:1 パーセンタイル:23.13(Nuclear Science & Technology)



Development of numerical analysis method for tube failure propagation under sodium-water reaction accident

内堀 昭寛; 柳沢 秀樹*; 高田 孝; 大島 宏之

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11



Advancement of numerical analysis method for tube failure propagation

内堀 昭寛; 高田 孝; 柳沢 秀樹*; Li, J.*; Jang, S.*

Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), p.1289 - 1294, 2018/11



Numerical study on effect of nucleation site density on behavior of bubble coalescence by using CMFD simulation code TPFIT

小野 綾子; 鈴木 貴行*; 吉田 啓之

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10



Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

 被引用回数:6 パーセンタイル:72.04(Nuclear Science & Technology)

An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.


Visualization study on droplet-entrainment in a high-speed gas jet into a liquid pool

杉本 太郎*; 齋藤 慎平*; 金子 暁子*; 阿部 豊*; 内堀 昭寛; 大島 宏之

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07


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