Koji YAMADA
著者に関連した記事は3件です。
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第13回
原子炉圧力容器の給水ノズルコーナーに対する渦電流探傷技術の開発
著者:
神長 貴幸,Takayuki KAMINAGA,吉川 祐明,Hiroaki KIKKAWA,山田 浩二,Koji YAMADA,平崎 孝幸,Takayuki HIRASAKI,西岡 朋美,Tomomi NISHIOKA,首藤 浩丈,Hirotake SYUTOU,齋藤 康二,Koji SAITO,東海林 一,Hajime SHOHJI,江原 和也,Kazuya EHARA,土橋 健太郎,Kentaro TSUCHIHASHI
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Eddy current testing Fatigue crack Feed water nozzle inside corner Reactor pressure vessel
神長 貴幸,Takayuki KAMINAGA,吉川 祐明,Hiroaki KIKKAWA,山田 浩二,Koji YAMADA,平崎 孝幸,Takayuki HIRASAKI,西岡 朋美,Tomomi NISHIOKA,首藤 浩丈,Hirotake SYUTOU,齋藤 康二,Koji SAITO,東海林 一,Hajime SHOHJI,江原 和也,Kazuya EHARA,土橋 健太郎,Kentaro TSUCHIHASHI
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キーワードタグ:
Eddy current testing Fatigue crack Feed water nozzle inside corner Reactor pressure vessel
It is necessary to confirm whether the feed water nozzle inside corner has crack by eddy current testing, to enable long-term operation of a plant. Therefore, we performed eddy current testing using various types of specimen to collect basic data (probability and length measurement accuracy) of the eddy current testing methods, complying with JEAG4217. As a result, eddy current testing was effective for the inspection of feed water nozzle inside corner. ...
英字タイトル:
Eddy current testing technique for feed water nozzle inside corner of nuclear reactor pressure vessel
英字タイトル:
Eddy current testing technique for feed water nozzle inside corner of nuclear reactor pressure vessel
第10回
実機供用2 相ステンレス鋳鋼の熱時効評価
著者:
山田 卓陽,Takuyo YAMADA,藤井 克彦,Katsuhiko FUJII,青木 政徳,Masanori AOKI,有岡 孝司,Koji ARIOKA
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aging hardening Atom-probe analysis cast duplex stainless steel G-phase spinodal decomposition thermal-aging
山田 卓陽,Takuyo YAMADA,藤井 克彦,Katsuhiko FUJII,青木 政徳,Masanori AOKI,有岡 孝司,Koji ARIOKA
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aging hardening Atom-probe analysis cast duplex stainless steel G-phase spinodal decomposition thermal-aging
In this study, thermal-aging evaluation has been performed using service aged elbow pipe in PWR plant, aged at 320℃ for 196,500h. As a result, micro Vickers hardness of ferrite in service material (SCS14A), HV(0.025) was 616~630. Since micro Vickers hardness of un-aged ferrite phase is about HV(0.025)=300 in commercial SCS14A, the increasing of ferrite hardness during aging was 300. Cr-rich and Fe-rich regions were observed in the ferrite phase using Atom-probe analysis. In addition, Ni, Si and Mo clus...
英字タイトル:
Thermal-aging evaluation of on site aged cast duplex stainless steel
英字タイトル:
Thermal-aging evaluation of on site aged cast duplex stainless steel
第10回
沸騰水型原子炉圧力容器の過渡事象における 加圧熱衝撃の評価
著者:
桝田 祐貴,Yuki MASUDA,高橋 淳介,Junsuke TAKAHASHI,廣川 文仁,Fumihito HIROKAWA,山田 浩二,Koji YAMADA
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Boiling Water Reactor Fracture toughness Pressurized Thermal Shock Reactor Pressure Vessel Stress intensity factor Structural integrity Transient condition
桝田 祐貴,Yuki MASUDA,高橋 淳介,Junsuke TAKAHASHI,廣川 文仁,Fumihito HIROKAWA,山田 浩二,Koji YAMADA
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Boiling Water Reactor Fracture toughness Pressurized Thermal Shock Reactor Pressure Vessel Stress intensity factor Structural integrity Transient condition
The structural integrity for Pressurized Thermal Shock (PTS) was evaluated for the RPVs of Japanese Boiling Water Reactors (BWRs). It has been clarified that the BWR RPVs have the sufficient margin of fracture toughness by calculating the stress intensity factor in transitional condition and the acceptance criteria for RPV shell plate which is assumed to be neutron-irradiated in core region for 60 years. ...
英字タイトル:
Evaluation of Pressurized Thermal Shock in transitional condition for Boiling Water Reactor pressure vessel
英字タイトル:
Evaluation of Pressurized Thermal Shock in transitional condition for Boiling Water Reactor pressure vessel