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CMohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*
Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11
被引用回数：0 パーセンタイル：0(Materials Science, Multidisciplinary)The steam oxidation test on the Cr-coated Zry cladding was studied up to 1400C to understand the oxidation behavior under the accidental conditions. The double-sided oxidation test study showed that Cr coating can protect Zry cladding at 1200C within 5 min. Cr coating has a protective effect on the Zry cladding up to 1200C in a steam environment. However, in the oxidation test up to 1200C/30 min and 1300C/5 min, Cr coating can no longer protect Zry cladding. Furthermore, at 1300C, the intermetallic phase of the Zr(Cr, Fe)
phase that precipitated within the Zry substrate formed as globule microstructures with Fe enrichment. In addition, the transition of the intermetallic phase within the Zry substrate from the solid to the pre-liquid and liquid phases was observed, where it was determined at 1350C/60 min and 1400C/30 min within the ZrO phase (outer side region). The oxidation of the Zr(Cr, Fe) interlayer was also determined in this study, where it resulted in the formation of the oxide phase of Cr, Zr, and Fe. It is worth mentioning that further experiments, such as mechanical testing and modeling, should be considered to support the degradation of the Cr-coated Zry cladding mainly when the liquid phase of the intermetallic phase is obtained for beyond design-basis accident environment.
岡田 裕史; 天谷 政樹
Annals of Nuclear Energy, 145, p.107539_1 - 107539_8, 2020/09
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)In order to evaluate the fracture resistance of fuel rods against a seismic loading which might be applied following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated after the Fukushima-Daiichi Nuclear Power Plant accident. In consideration of previous studies and results, the effect of the amount of oxidation on the maximum bending stress of pre-hydrided cladding tube with a small amount of ballooning was investigated in this study. According to the obtained results, it was suggested that the decrease in the maximum bending stress of the cladding tube experienced LOCA conditions is mainly determined by the hydrogen concentration in the cladding tube after simulated LOCA test, irrespective of pre-hydriding. It was also suggested that the decreasing trend of the maximum bending stress with increasing the hydrogen concentration would be expressed by a form of exponential function, in which the maximum bending stress at a hydrogen concentration of 1500 ppm was estimated to be about a half of that at 0 ppm.
岡田 裕史; 天谷 政樹
Annals of Nuclear Energy, 136, p.107028_1 - 107028_9, 2020/02
被引用回数：5 パーセンタイル：53.85(Nuclear Science & Technology)In order to evaluate the fracture resistance of fuel rods against a seismic loading following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated since the Fukushima-Daiichi Nuclear Power Plant accident. In this study, four-point-bending-tests were performed using Zircaloy-4 cladding tubes which experienced a simulated LOCA sequence in order to investigate the effects of oxidation and secondary hydriding occurring during a LOCA on the bending strength of fuel cladding. According to the obtained results, it was suggested that the maximum bending stress would be affected by the oxygen concentration in the prior-beta layer as well as the thickness of prior-beta layer. It was considered that the decrease in maximum bending stress by secondary hydriding is probably expressed by multiplying a factor of 0.37 by the maximum bending stress which solely takes account of the effect of oxidation.
今井 伸明*; Jeong, S.-C.*; 小柳津 充広*; 新井 重昭*; 渕 好秀*; 平山 賀一*; 石山 博恒*; 宮武 宇也; 田中 雅彦*; 岡田 雅之*; et al.
Review of Scientific Instruments, 79(2), p.02A906_1 - 02A906_3, 2008/02
被引用回数：13 パーセンタイル：51.72(Instruments & Instrumentation)KEKCBは、18GHzの電子共鳴型(ECR)イオン源で、タンデムに設置された短寿命核分離加速実験装置(TRIAC)の一部として、上流で生成され質量分離された短寿命原子核の1+イオンを、ビーム軸上のECRプラズマに入射することで、多価のイオンへその場変換する装置である。これまでの開発研究によって、クリプトン,キセノン等のガス状元素やバリウム,インジウム等の非ガス状元素に対して、質量/電荷比が7以下の多価イオンにまで変換する効率を、それぞれ7%, 2%にまで向上することができた。また、短寿命な同位元素による測定と比較することで、この変換効率は、1秒程度の半減期を持つ同位元素に対しては、変わらないことを確かめた。従来見られていた出力ビーム中のバックグランドは、プラズマ壁及び電極の全アルミ化,高圧純水洗浄等により、10
ppsから600ppsにまで落とすことができた。
松田 誠; 竹内 末広; 月橋 芳廣; 花島 進; 阿部 信市; 長 明彦; 石崎 暢洋; 田山 豪一; 仲野谷 孝充; 株本 裕史; et al.
Proceedings of 3rd Annual Meeting of Particle Accelerator Society of Japan and 31st Linear Accelerator Meeting in Japan, p.275 - 277, 2006/00
2005年度のタンデム加速器の運転日数は182日であった。加速管の更新により最高端子電圧は19.1MVに達し18MVでの実験利用が開始された。利用イオン種は21元素(28核種)であり、
Oの利用が全体の約2割で、おもに核化学実験に利用された。p, 
Li, 
Xeの利用はそれぞれ約1割を占め、p, LiはおもにTRIACの一次ビームに利用された。超伝導ブースターの運転日数は34日で、昨年度から始まったTRIACの実験利用は12日であった。開発事項としては、タンデム加速器では加速管を更新し最高電圧が19MVに達した。また高電圧端子内イオン源の14.5GHzECRイオン源への更新計画が進行している。超伝導ブースターは1994年以来高エネルギービームの加速に利用されてきたが、近年になりインジウムガスケットに起因する真空リークが発生している。空洞のQ値も下がってきており、対策として空洞に高圧超純水洗浄を施し性能を復活させる試験を進めている。KEKと共同で進めてきたTRIACは2005年3月に完成し、10月から利用が開始された。TRIACからのビームを超伝導ブースターにて5
8MeV/uのエネルギーまで加速する計画を進めており、TRIACからの1.1MeV/uのビームを効率よく加速するため、low
空洞の開発を行っている。
今井 伸明*; Jeong, S.-C.*; 小柳津 充広*; 新井 重昭*; 渕 好秀*; 平山 賀一*; 石山 博恒*; 宮武 宇也; 田中 雅彦*; 岡田 雅之*; et al.
no journal, ,
KEKCBはTRIACにおける一価のイオンを多価イオンに変換するためのECR型イオン源である。KEKCBを用いることで、一価の気体イオン及び非気体イオンを、A/q
7について、それぞれ7%及び2%の効率で多価イオンに変換することができた。これらの効率は秒オーダーでは半減期によらないことがわかった。また、3つのコリメータをKEKCBの前後に設置してビーム軸を規定することで、KEKCBへのビーム入射時のビームハンドリングが容易となった。さらに、電極及びプラズマチェンバー表面の研磨・洗浄により、KEKCBのECRプラズマからの不純物が劇的に減少した。
長 明彦; 阿部 信市; 遊津 拓洋; 花島 進; 石井 哲朗; 石崎 暢洋; 株本 裕史; 沓掛 健一; 松田 誠; 中村 暢彦; et al.
no journal, ,
東海放射性核種ビーム加速器施設(TRIAC)では、タンデム加速器の陽子や重イオンビームを用いて生成した放射性核種をオンライン同位体分離器で分離し、再加速することができる。2005年の実験共用開始から、ウラン核分裂生成物やLiのビームを実験・研究に提供している。Liの生成には99%濃縮
C同位体焼結標的を用いていた。この標的を装着したイオン源システムからのLiの放出時間は3.2秒と長く、新たに開発する
Li(T
=0.2秒)ビームの生成には適さない。われわれは速い放出時間を持つチッ化ボロン標的の開発を行い、毎秒10
個のLiビームの生成に成功した。
岡田 裕史
no journal, ,
In a case where a pinhole leak occurs in a fuel rod incidentally, it is possible that coolant enters the fuel rod through the pinhole. Since knowledge about the behavior of the fuel rod with a pinhole under LOCA conditions is limited, semi-integral quench tests were performed with non-irradiated zircaloy-4 fuel cladding tubes with a pinhole in order to investigate the difference in the fracture behaviors between normal and leaker fuels under LOCA conditions. Initially injected water affected the oxidation behavior of the inner surface of cladding during the test, and the fracture boundary of the test rod was dependent on not only the axial restrained condition during the test but also the existence of a pinhole and initially injected water. This tendency seemed to be related to the amount of oxidation of cladding inner surface caused by the steam which remained in or entered the test rod during the test.
岡田 裕史
no journal, ,
It is important to evaluate the fracture resistance of fuel rod against a seismic loading following a LOCA since the accident at the Fukushima Daiichi Nuclear Power Plant. In consideration of this, the effects of oxidation and secondary hydriding on the fracture resistance were investigated by four point bending test using the cladding tube with drilled holes. Based on the obtained results, it was suggested that the oxygen concentration in prior-beta layer affects the maximum bending stress. The maximum bending stress related to the secondary hydriding was estimated as about a half compared with that at the rupture opening position.
岡田 裕史
no journal, ,
LOCA後長期冷却期間中の炉心冷却性を維持する観点で、地震時の燃料の耐破損性を把握するためにLOCA模擬試験後4点曲げ試験が実施されている。先行研究では、LOCA時の酸化及び膨れ率がLOCA後被覆管の曲げ強度に及ぼす影響について調べられてきた。本研究では、予め被覆管に開口部を設けた試験燃料棒を用いたLOCA模擬試験後4点曲げ試験を実施し、先行研究で十分に明らかとなっていないLOCA時の二次水素化がLOCA後被覆管の曲げ強度に及ぼす影響について評価した。その結果、二次水素化部の最大曲げ応力はprior-相厚さに依存する傾向が見られ、膨れ破裂部の半分程度と評価された。
岡田 裕史
no journal, ,
In order to evaluate the fracture resistance of cladding tube against a seismic loading which might be applied following LOCA, four-point-bending-test has been conducted. In consideration of previous studies and results, the effect of the amount of oxidation on the maximum bending stress of pre-hydrided test rod with small amount of ballooning was investigated in this study. Based on the obtained results, the maximum bending stress of pre-hydrided test rod decreased compared with that of as-received test rod. In addition, it is considered that the decrease in maximum bending stress is almost determined by hydrogen concentration after simulated LOCA test. According to this trend, it is expected that the maximum bending stress at ballooned-and-ruptured region in pre-hydrided cladding tube is determined by hydrogen concentration after simulated LOCA and its value is similar to the data obtained in this study and literature data.
岡田 裕史
no journal, ,
原子力機構では、LOCA後長期冷却期間中の燃料棒の耐破断性を評価するために被覆管のLOCA模擬試験後4点曲げ試験を実施している。これまで、受取まま被覆管(受取材)の膨れ破裂部や二次水素化部の最大曲げ応力と変態層厚さの関係を把握できたが、通常運転時の腐食に伴う水素吸収がこの関係に及ぼす影響は不明である。そこで本研究では、予め水素吸収させた未照射Zry-4被覆管(水素吸収材)に対してLOCA模擬試験後4点曲げ試験を実施し、その曲げ強度を評価した。その結果、水素吸収材の酸化部は受取材の膨れ破裂部よりも曲げ強度が低い傾向が見られた。また、水素吸収材の酸化部及び二次水素化部と受取材の二次水素化部について、試験後の変態層厚さが同等な条件で曲げ強度を比較した結果、LOCA模擬試験後の被覆管の曲げ強度は、試験前の水素吸収量よりも試験前及び試験中に吸収した全水素吸収量で整理できることが示唆された。
Mohamad, A. B.; 岡田 裕史*; 佐藤 大樹*; 井岡 郁夫; 鈴木 恵理子; 根本 義之
no journal, ,
Chromium (Cr) coated zirconium (Zr) based alloy cladding is the promising material for a near term accident tolerant fuel (ATF). Cr-coated Zr based cladding was fabricated by sputtering technique and simulated LOCA tests were conducted up to different high temperatures (1200C to 1350C). The present work aims to investigate the metallography of Cr-coated Zr cladding after LOCA test. The result showed CrO
layers were formed as a protective oxide layer at the outmost layer for all samples. However, Cr-Zr phases were observed for the samples which tested up to 1300C and 1350C, seems that the Cr-Zr reaction had occurred at this temperature. In addition, Cr-Zr layer in the sample tested up to 1350C was thicker than which in the sample tested up 1300C. In particular, the main phases observed in the cross-sectional area were CrO, Cr, ZrO, Cr-Zr, and Zr situated from outer to inner of the sample after LOCA test. The details of the microstructure on these samples will be discussed in the presentation.
根本 義之; 岡田 裕史*; 佐藤 大樹*; Mohamad, A. B.; 井岡 郁夫; 鈴木 恵理子
no journal, ,
従来のジルコニウム合金製被覆管の外表面にクロム(Cr)等のコーティングを施し、事故時高温水蒸気中での耐酸化性を向上させた事故耐性燃料(ATF)被覆管の開発が進められている。本研究ではCrコーティング被覆管の事故時挙動について評価し、今後の開発に資する知見を得るため、酸化挙動の評価を行った。酸化挙動に関しては、LOCA時の被覆管破裂後を想定した両面酸化条件での酸化試験を行った。その結果、6501150Cの温度域ではいずれの場合もコーティングなしの場合に比較して、コーティングありの場合に、酸化量が低く抑えられて推移する傾向が見られた。
Mohamad, A. B.; 古本 健一郎; 根本 義之; 井岡 郁夫; 佐藤 大樹*; 岡田 裕史*; 山下 真一郎; 逢坂 正彦
no journal, ,
Chromium (Cr) coated zirconium (Zr) based alloy cladding is the promising material for a near term accident tolerant fuel (ATF). Cr-coated Zr based cladding was fabricated by sputtering technique and HT oxidation tests were conducted up to different high temperatures (1100C to 1400C). The present work aims to investigate the metallography of Cr-coated Zr cladding after HT steam test. The result showed CrO layers were formed as a protective oxide layer at the outmost layer for all samples. However, Cr-Zr-Fe phases were observed In particular, the main phases observed in the cross-sectional area were CrO, Cr, ZrO, Cr-Zr, and Zr situated from outer to inner of the sample after HT test. The details of the microstructure and mechanism of these samples will be discussed in the presentation.
intermetallic phase up to the eutectic temperatureMohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*
no journal, ,
The development of Accident Tolerant Fuel (ATF) had been started by conducting the investigation on new concepts to improve the safety of Light Water Reactors (LWRs). It is well known that the Cr coating on Zry cladding has shown an improvement in behavior under accident conditions and normal operation. In the Cr-Zr system, the eutectic phase of ZrCr is present at 1332C and forms as intermetallic compounds. There is still lack of data on the evolution of the intermetallic phase when the oxidation temperature reaches the eutectic temperature of Cr-Zr. Therefore, the purpose of this study will be to understand the solid-to-liquid phase transition of Zr(Cr, Fe). High temperature oxidation tests were performed in a steam atmosphere to the target temperature (i.e., 1100C, 1200C, 1300C, 1350C, and 1400C) for different exposure times of 5, 30, and 60 min. From the tests, the transition of Zr(Cr, Fe) that formed at the Cr-Zr interface and also that precipitated in the Zry cladding were studied with varied oxidation time and temperatures. The microstructural evolution of the intermetallic phase was observed in the Zr substrate within the progress of the oxidation of Cr-coated Zry. A dendritic structure was observed at 1400C, indicating the formation of the Zr(Cr, Fe) liquid phase when the oxidation temperature is above the eutectic temperature.
Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*
no journal, ,
The development of Accident Tolerant Fuel (ATF) started with the investigation of new concepts to improve the safety of Light Water Reactors (LWR). It is well known that the Cr coating on Zry cladding has shown improved behaviour under accident conditions and in normal operation. However, many questions remain about the oxidation behaviour of Cr-coated Zry cladding as it approaches the Cr-Zr eutectic temperature. In the present study, the steam oxidation tests were carried out under different oxidation conditions in order to understand the oxidation behaviour of the Cr-coated material mainly above the eutectic temperature. The results obtained showed that the Cr coating can protect the Zry substrate at 1100C to 1200C/5min. However, at 1200C/30min, the Cr coating no longer protected the Zry substrate. This is due to the formation of Zr at the Cr grain boundary where it becomes a short path for O diffusion and reacts with the Zry substrate.
Mohamad, A. B.; Chien, J.; 井岡 郁夫; 鈴木 恵理子; 近藤 啓悦; 根本 義之; 大久保 成彰; 山下 真一郎; 岡田 裕史*; 佐藤 大樹*
no journal, ,
As a candidate for accident tolerent fuel (ATF) cladding tubes, chromium (Cr) coated Zry cladding tubes are being developed. To realize the Cr-coated Zry cladding for future cladding application, the integrity of this material needs to be confirmed with the reactor environment conditions. In order to understand an effect of irradiation on the Cr-coated Zry cladding, ion irradiation test is carried out on the cross-sectional specimens. In this study, the following content mainly focuses on the microstructural evolution and mechanical behavior induced by ion irradiation in Cr-coated Zry cladding. The 10 MeV-Fe
irradiation was chosen to induce the damage on the cross section of the Cr-coated Zry cladding. The sample was irradiated at 350 
C and the peak irradiation damage was approximately 30 dpa. TEM-EDS shows that the Fe-enrichment peaks are observed around 15 nm at the interface regions between the coating and the Zry substrate for the sample irradiated up to 30 dpa. In addition, the hardness of irradiated sample is higher compared to that un-irradiated sample as result of irradiation-induced hardening. The details of the irradiation effect on the un- and irradiated Cr-coated Zry cladding will be discussed in detail during the presentation.
Mohamad, A. B.; 根本 義之; 古本 健一郎*; 岡田 裕史*; 佐藤 大樹*
no journal, ,
It is widely recognized that the Cr coating on Zry cladding has shown an improvement in the behavior under accident conditions and normal operation. Many research groups around the world have conducted the high-temperature oxidation and LOCA tests on Cr-coated Zry under accident conditions and explained the degradation phenomena from these tests. Although many literatures have revealed the mechanism and phenomena of the degradation of the Cr-coated, there is still a lack of data on the Zr-Cr-Fe phase or intermetallic phase behavior when the temperature reaches and exceeds the eutectic temperature of Zr-Cr (1332C). In the present study, a high temperature steam oxidation test is carried out from 1100 to 1400C in order to understand the behavior of Cr-coated Zry as it approaches the eutectic temperature. Fromelectron probe microanalysis, the Fe enrichment of the Zr(Cr,Fe) phase is identified for the sample tested at 1300C. In addition, the liquid formation of the Zr(Cr,Fe) phase is observed at 1300C.
