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under high temperatures and various water vapor concentrationsMohamad, A. B.; 中島 邦久; 井元 純平*; 高野 公秀
Journal of Nuclear Materials, 631, p.156826_1 - 156826_12, 2026/09
被引用回数：0It is estimated that considerable amounts of Cs remain in the reactors at the Fukushima Daiichi nuclear power plant and may react with the reactor materials such as UO fuel as reported by our previous paper. In order to understand the chemisorption behavior of CsOH vapor on UO fuel, chemisorption tests between CsOH vapor on UO under different water vapor concentrations and temperatures were performed. As predicted by the thermodynamic analysis, the experimental result indicated that, under the poor water vapor condition, chemisorption to form CsUO
occurred even at 1273 K. On the other hand, as predicted by the thermodynamic analysis, the experimental result indicated that, under the rich water vapor condition, chemisorption to form CsUO
occurred at 1273 K and 1473 K. Therefore, this study demonstrates the crucial role of thermodynamics in predicting the conditions under which chemisorption occurs. In addition, it was suggested that the chemisorption rate of CsOH vapor on UO pellet was considerably higher than that onto type 304 stainless steel.
Mohamad, A. B.; Chen, J.*; 井岡 郁夫*; 鈴木 恵理子; 近藤 啓悦; 阿部 陽介; 山下 真一郎; 大久保 成彰; 根本 義之; 岡田 裕史*; et al.
Journal of Nuclear Materials, 625, p.156513_1 - 156513_9, 2026/04
被引用回数：0 パーセンタイル：0.00(Materials Science, Multidisciplinary)Ion irradiation was carried out on Cr coating Zry cladding to investigate their microstructure evolution and mechanical properties. The sample was irradiated at reactor normal operation conditions. Microstructural observation and mechanical testing of non-irradiated samples and irradiated samples were performed to understand irradiation damage to the Cr-coated Zry cladding. Results of High Resolution Transmission Electron Microscopy and chemical analysis revealed Fe enrichment at the Cr coating and Zr substrate interface of irradiated samples due to irradiation enhanced diffusion or irradiation induced mixing. Irradiation led to the formation of Fe enrichment at the Cr Zr interface approximately 15nm. Moreover, hardening of the Cr coating and Zr substrate regions was observed in the irradiated sample.
Mohamad, A. B.; 山下 真一郎; 根本 義之; 阿部 陽介; Pham, V. H.; 井岡 郁夫; 相馬 康孝; 石島 暖大; 佐藤 智徳; Rizaal, M.; et al.
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 8 Pages, 2025/10
In parallel with the Advanced Technology Fuel (ATF) development program in Japan, Japan Atomic Energy Agency has established an R&D program on the ATF to provide more scientific support to the ATF fuel vendors in Japan. Based on the identified phenomenological issues, the R&D program covers the issues mainly on the light water reactors conditions such as normal operation, loss of coolant accident, beyond design basis accident, and severe accident. The R&D program such as irradiation test, corrosion test, LOCA test, and etc. are proposed. By acquiring the data from the experiment, the final goal is to implement the experimental data and model into the simulation code in order to predict the fuel performance behaviour at high burn-up. In 2024 TOPFUEL, the introduction of the R&D has been presented. In 2025, the update on the Cr-coated research to simulate normal operation and accident conditions will be given. For example, tests are currently underway to understand the in-reactor environment, such as ion irradiation, gamma irradiation and proton irradiation tests. In addition, a loss-of-coolant accident is being simulated for accident conditions, and the chemical effects of fission products on the cladding are being studied.
河合 慶人*; 根本 義之; 藤村 由希; 近藤 啓悦; 阿部 陽介; Mohamad, A. B.; Pham, V. H.; 石川 法人; 石島 暖大; 井岡 郁夫; et al.
日本原子力学会和文論文誌(インターネット), 24(3), p.82 - 98, 2025/08
現行のジルカロイ被覆管よりも事故耐性に優れた事故耐性燃料(ATF)被覆管の開発が進められている。その一つに外面にCrをコーティングしたコーティング被覆管があり、高温水蒸気中での耐酸化性の向上が見込まれている。しかし、冷却水損失事故(LOCA)時には被覆管が破裂し、コーティングされていない被覆管内面も酸化することが予想される。そこで、本研究ではジルカロイ4被覆管とその外表面にCrコーティングを施したコーティング被覆管を用いてLOCA試験及び高温水蒸気中での酸化試験を行い、酸化挙動、水素吸収挙動及び機械特性の違いを検討した。熱天秤を用いた酸化試験の結果においては、コーティング被覆管は非コーティング被覆管と比較して酸化量が半減し、コーティングの酸化抑制効果が示された。しかし、酸化試験後の試料の水素吸収量を測定したところ、同じ酸化試験条件でも、コーティング被覆管での方が非コーティング被覆管よりも、水素吸収量が多いことがわかった。本報では、この水素吸収量の違いの原因や、それに関連した酸化試験後の機械的性質への影響などについて議論した。
廣岡 瞬; 堀井 雄太; 林崎 康平; Mohamad, A. B.
核燃料, (60-1), p.17 - 20, 2025/02
NuMatは2010年に開催されて以来、2年ごとに開催される国際会議である。核燃料・材料に特化した国際会議としては非常に規模が大きく知名度も高く、Elsevierが主催しJournal of Nuclear Materialsと連携して開催されることも本会議の特徴である。本件はNuMat2024(2024.10.14
17、シンガポール)の参加報告を行うものである。
Mohamad, A. B.; 相馬 康孝; 根本 義之; 阿部 陽介; 井岡 郁夫; 佐藤 智徳; 石島 暖大; 三輪 周平; 中島 邦久; 山下 真一郎; et al.
Proceedings of TopFuel 2024 (Internet), 8 Pages, 2024/10
In parallel with the Accident Tolerant Fuel development program in Japan, Japan Atomic Energy Agency has established an R&D program on the ATF to provide more scientific support to the ATF fuel vendors in Japan. Based on the phenomenological issues identified, the R&D program covers the issues mainly on the light water reactors conditions such as normal operation, loss of coolant accident, beyond design basis accident, and severe accident. The R&D program such as irradiation test, corrosion test, LOCA test, and etc. are proposed. By acquiring the data from the experiment, the final goal is to implement the experimental data and model into the simulation code in order to predict the fuel performance behaviour at high burn-up.
Mohamad, A. B.; 宇田川 豊
Nuclear Technology, 210(2), p.245 - 260, 2024/02
被引用回数：2 パーセンタイル：27.37(Nuclear Science & Technology)In the Power to Melt and Maneuverability (P2M) project, a simulation exercise on two past power ramp experiments xM3 on medium burn-up rod and HBC4 on high burn-up rod were performed with the fuel performance code FEMAXI-8 to investigate the fuel behavior under high power and high-temperature conditions toward centerline fuel melting. In order to treat fuel melting, empirical melting temperature models have been incorporated into the FEMAXI-8 code. The present analysis gave reasonable predictions not only on cladding deformation but also on the fuel melting behavior of the HBC4 rod, in which the UO liquidus temperature was reached during the transient. On the other hand, model improvement appeared to be needed for a more accurate treatment of fuel melting behavior of the xM3 rod, in which fuel center temperature reached solidus line, whereas may not reached liquidus line. A reasonable agreement of estimated FGR with the measurement suggested that the high temperature FGR at the given conditions are essentially temperature dependent phenomenon: rate-limited primarily by thermally activated elementary processes such as fission gas diffusion.
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：12 パーセンタイル：62.08(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.
Mohamad, A. B.; 中島 邦久; 三輪 周平; 逢坂 正彦
Journal of Nuclear Science and Technology, 60(3), p.215 - 222, 2023/03
被引用回数：1 パーセンタイル：9.64(Nuclear Science & Technology)The distribution of Sr in the reactor would be influenced by a chemical reaction of Sr vapor species with a structural material of internal reactor and fuel cladding materials; stainless steel (SS) or Zircaloy (Zry) cladding during 1F-NPS accident. The chemical interaction between Sr-Zry and Sr-SS has been described. The reaction tests have been performed to investigate the chemical interaction behavior under possible severe accident conditions. The tests have been conducted up to 1523 K under steam atmosphere. It was confirmed that Sr-Zr-O and Sr-Si-O compounds were formed through 2 kinds chemical interactions; gas-solid reaction and liquid-solid reaction. The gas and liquid species of Sr in a good contact with the solid Zry and SS to form Sr-Zr-O and Sr-Si-O compounds, respectively. Sr was deposited onto the Zry and SS surfaces and lead to the formation of reaction product. Thus, this study highlights the possibility that Sr was deposited and retained in the core structure where the temperature was elevated during the accident in the 1F-NPS.
Mohamad, A. B.; 根本 義之; 相馬 康孝; 石島 暖大; 佐藤 智徳; 井岡 郁夫; Pham, V. H.; 三輪 周平; 中島 邦久; 加治 芳行; et al.
no journal, ,
ATF等の新型燃料実用化においては、関連技術開発やそれらの基となる科学的知見の取得及び拡充が不可欠である。原子力機構は、照射試験実施による燃料ふるまい解析技術基盤の構築のための研究開発を行い、長期を要する開発において、開発内容やスケジュールの予見性向上に貢献していくべきと認識している。このため、実装化が最も早いCrコーティング被覆管に関して、燃料ふるまいのメカニズムに立ち返り、「長期照射時の影響」「事故時影響」に関する科学的知見を拡充することを目的とした基礎基盤研究計画を立案し、研究をすすめている。本発表では各研究項目の内容や期待される成果、これまでに得られた結果等を紹介する。
Journeau, C.*; Bechta, S.*; Komlev, A.*; 倉田 正輝; 多木 寛; 松本 俊慶; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Bottomley, D.*; et al.
no journal, ,
The OECD project TCOFF-2 (Thermodynamic Chemistry of Fission Products - Part 2) is an extension of the original project that was part of the near-term projects intended to support the Fukushima Daiichi decommissioning efforts. This second part continues to be mainly financed by Japanese Ministry (MEXT) with JAEA-CLADS support. TCOFF2 started in August 2022 and includes certain new material requirements compared to the first TCOFF project. These are increased emphasis on accident tolerant fuels (ATFs), certain actinides and fission products related to volatility/leachability but also certain combinations of the ceramic oxide systems important for MCCI behaviour. Following a PIRT review of phenomena in TCOFF 1, there was in TCOFF2 a Task 1 to prioritise current needs, particularly for relevance to severe accident phenomena and alternative materials (ATFs). This included a re-evaluation and ranking of the thermodynamic systems to make a Systems Identification and Ranking Table (SIRT) for the improvement of the thermodynamic database foreseen in TCOFF2. The further systems for evaluation were then proposed by the partners according to their particular requirements; this resulted in over 150 thermodynamic systems. In the mid-year meeting (June 2023) discussions were made to rationalise these into the most important 20 systems. The rationale for the reduction of systems to this limit will be explained in the talk, both those systems that were omitted as well as those finally included. These systems will then be used as a priority list of work for the experimental call to members that will be launched by the TCOFF-2 project towards the end of the year 2023 with the intention to initiate the first projects soon afterwards.
山下 真一郎; Mohamad, A. B.; 根本 義之; 相馬 康孝; 石島 暖大; 佐藤 智徳; 井岡 郁夫; Pham, V. H.; 三輪 周平; 中島 邦久; et al.
no journal, ,
原子力機構(JAEA)では事故耐性の向上を目指した燃料被覆管の各種コーティング技術の研究を行っている。講演では全体概要の他、それらの研究に用いることを目的としたJAEAでの新規の装置開発について紹介する。
山下 真一郎; Mohamad, A. B.; 相馬 康孝; 根本 義之; 井岡 郁夫; 加治 芳行; 逢坂 正彦
no journal, ,
2015年以降、日本の事故耐性燃料(ATF)研究開発プログラムは、国内商用炉の燃料や炉心の設計・評価、研究開発における経験を最大限に活用するため、発電プラントメーカ、燃料メーカや大学と連携し進められてきている。本講演では、同プログラムにおける原子力機構の役割を含めて、現在の研究開発について進捗状況の概要を紹介する。
三輪 周平; 中島 邦久; 唐澤 英年; Rizaal, M.; Luu, V. N.; Mohamad, A. B.
no journal, ,
東京電力福島第一原子力発電所内のセシウム等のFPの分布や性状を把握することが廃炉に向けた重要な課題であり、日本原子力研究開発機構ではそれらに大きな影響を与えるFP化学に着目した基礎研究を実施し、事故時の原子力発電所内のFP化学を評価するためのデータベースECUMEを開発している。ECUMEは福島第一原子力発電所の事故により明らかになった重要な現象、例えば、セシウムの制御材ホウ素との反応や、構造材との反応に関するデータやモデルを収納している。近年の内部調査により明らかとなったシールドプラグでの高線量等の原因を明らかにするため、コンクリートや他の炉内物質との化学反応を調べ、モデル化を行い、ECUMEの更新を進めている。
河合 慶人*; 根本 義之; 藤村 由希; 近藤 啓悦; 阿部 陽介; Mohamad, A. B.; Pham, V. H.; 石川 法人; 石島 暖大; 井岡 郁夫; et al.
no journal, ,
現行のジルカロイ被覆管よりも事故耐性に優れた事故耐性燃料(ATF)被覆管の開発が進められている。その一つに被覆管外面にCrをコーティングしたコーティング被覆管があり、高温水蒸気中での耐酸化性の向上が見込まれている。しかし、冷却水損失事故(LOCA)時には被覆管が破裂し、コーティングされていない被覆管内面も酸化することが予想される。そこで、本研究ではジルカロイ4被覆管及びその外面にCrコーティングを施したコーティング被覆管を用いて高温水蒸気中での酸化試験を行い、酸化挙動、水素吸収挙動及び機械特性の違いを検討した。
鵜飼 重治; 阿部 陽介; Geng, D.*; 久保 淳; 佐々木 泰祐*; Mohamad, A. B.; 藤田 洋平*; 大久保 成彰; 山下 真一郎; Yu, H.*; et al.
no journal, ,
本研究では、CrとAlの相互拡散により引き起こされる連続的な組成傾斜を有するコンビナトリアルFeCrAl合金を作製し、イオン照射とナノインデンテーションの連続剛性法による硬さ測定を組み合わせることで、連続的な(Cr,Al)組成に対応した大量のハイスループットNI硬さデータを一度に取得した。その結果、濃度が約5at.%で
'析出物生成によるNI硬さはピークを取ること、Cr濃度が30at.%以上でスピノーダル分解による'生成の促進など、2次元(Cr-Al)組成マップ上で'析出物生成の照射下ふるまいを明瞭に評価することができた。また、1.44dpa(0.64dpa)の低照射量において'析出物はすでに飽和傾向にあることを明らかにした。
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。
手塚 健一*; 木野 千晶*; 山下 晋; Mohamad, A. B.; 根本 義之
no journal, ,
原子力発電所の過酷事故発生時に水素発生・炉心溶融の進展を抑制することを目的とした事故耐性燃料の開発が進んでおり、PWR用には、CrコーティングしたZr被覆管(Crコーティング被覆管)が検討されている。本研究では、Crコーティング被覆管を用いたATFの事故時挙動を評価するための解析手法を開発した。解析ツールとして、国産のSA解析コードであり、ソースコードに容易にアクセス可能なSAMPSONを用いた。解析の結果、現行被覆管に比べて、Crコーティング被覆管を用いることで、事故模擬条件において有意な水素発生抑制効果を確認することができた。
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 350C 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.
