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Probing deformation behavior of a refractory high-entropy alloy using ${it in situ}$ neutron diffraction

Zhou, Y.*; Song, W.*; Zhang, F.*; Wu, Y.*; Lei, Z.*; Jiao, M.*; Zhang, X.*; Dong, J.*; Zhang, Y.*; Yang, M.*; et al.

Journal of Alloys and Compounds, 971, p.172635_1 - 172635_7, 2024/01

The grain orientation-dependent lattice strain evolution of a (TiZrHfNb)$$_{98}$$$$N_2$$ refractory high-entropy alloy (HEA) during tensile loading has been investigated using ${it in situ}$ neutron diffraction. The equivalent strain-hardening rate of each of the primary $$<hkl>$$-oriented grain families was found to be relatively low, manifesting the macroscopically weak work-hardening ability of such a body-centered cubic (BCC)-structured HEA. This finding is indicative of a dislocation planar slip mode that is confined in a few single-slip planes and leads to in-plane softening by high pile-up stresses.


Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

堀井 雄太; 廣岡 瞬; 宇野 弘樹*; 小笠原 誠洋*; 田村 哲也*; 山田 忠久*; 古澤 尚也*; 村上 龍敏; 加藤 正人

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

MOX燃料の照射により生成する主要なFPであるNd$$_{2}$$O$$_{3}$$及びSm$$_{2}$$O$$_{3}$$、模擬FPとして添加したMOXの熱伝導率を評価した。MOX中の模擬FPの均質性の観点から熱伝導率を評価するため、ボールミル法及び溶融法で作製した2種類の粉末を用いて、Nd及びSmの均質性が異なる試料を作製した。模擬FPが均質に固溶した試料では含有量が増加するにしたがってMOXの熱伝導率が低下するが、不均質な模擬FPは影響を及ぼさないことが分かった。熱伝導率に対するNd及びSmの影響を古典的フォノン輸送モデル$$lambda$$=(A+BT)$$^{-1}$$を用いてNd/Sm依存性を定量的に評価した結果、A(mK/W)=1.70$$times$$10$$^{-2}$$ + 0.93C$$_{Nd}$$ + 1.20C$$_{Sm}$$, B(m/W)=2.39$$times$$10$$^{-4}$$と表された。




JAEA-Review 2023-018, 159 Pages, 2023/12



令和3年度福島第一原子力発電所の炉内付着物サンプル等の分析; 令和3年度開始廃炉・汚染水対策事業費補助金に係る補助事業(燃料デブリの性状把握のための分析・推定技術の開発)

池内 宏知; 佐々木 新治; 大西 貴士; 仲吉 彬; 荒井 陽一; 佐藤 拓未; 多木 寛; 関尾 佳弘; 山口 祐加子; 森下 一喜; et al.

JAEA-Data/Code 2023-005, 418 Pages, 2023/12




Oxidation and embrittlement behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12

To evaluate the oxidation and embrittlement behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions, we conducted isothermal oxidation and ring-compression tests on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens. Further, we discussed the loss of coolable geometry of the reactor core loaded with the FeCrAl-ODS cladding tubes under LOCA conditions, using data from the ring-compression tests in this study and the integral thermal shock tests from our previous study. The results reveal that oxidation kinetics of the FeCrAl-ODS cladding tube at 1523 K is four orders of magnitude lower than that of a conventional Zircaloy cladding tube, which highlights the exceptional oxidation resistance of the FeCrAl-ODS cladding tube. The breakaway oxidation of the FeCrAl-ODS cladding tube was observed at 1623 K for durations equal to or exceeding 6 h, and melting was observed at 1723 K. The ring-compression and the integral thermal shock tests indicate that, depending on the oxidation time, the ductile to brittle transition threshold - as determined by the ring-compression test - exists between 1623 K and 1723 K. Meanwhile, the fracture threshold - established through the integral thermal shock test - falls between 1573 K and 1673 K. Therefore, taking a conservative approach based on available data, the fracture and non-fracture results from the integral thermal shock tests can define the lower and upper boundaries of the threshold for the loss of coolable geometry of the reactor core during a LOCA.



眞田 幸尚; 御園生 敏治; 尻引 武彦*

海洋理工学会誌, 27(2), p.37 - 44, 2023/12



MAAP code analysis focusing on the fuel debris conditions in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 3

佐藤 一憲; 吉川 信治; 山下 拓哉; 下村 健太; Cibula, M.*; 溝上 伸也*

Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12

Based on the updated knowledge from plant-internal investigations, experiments and computer-model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 3 was analyzed using the MAAP code. In Unit 3, it is considered that ca. 40 percent of UO$$_{2}$$ fuel was molten when core materials relocated to the lower plenum of the reactor pressure vessel. Initially relocated molten materials would have been fragmented by mixing with liquid water, while solid materials would have relocated later on. With this two-step relocation, debris in the lower plenum seems to have been permeable for coolant, thus debris seems to have been once cooled down effectively. Although the present MAAP analysis seems to slightly underestimate core-material oxidation during the relocation period, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Probable debris reheat-up behavior was evaluated based on interpretation of the pressure data. This evaluation predicted that the fuel debris in the lower plenum was basically in solid-phase at the time when it relocated to the pedestal. With this study, basic validity of the former prediction of the Unit 3 accident progression behavior was confirmed, and detailed boundary conditions for future studies addressing the later phases were provided.


Effect of molybdenum release on UO$$_{2}$$/MOX fuel oxidation under severe light water reactor accident conditions

Liu, J.; 三輪 周平; 唐澤 英年; 逢坂 正彦

Nuclear Materials and Energy (Internet), 37, p.101532_1 - 101532_5, 2023/12

To investigate the Mo release behavior and its influence on the fuel oxidation, the oxidation and evaporation behaviors of Mo powders and their influencing mechanism on the oxygen partial pressure around powders were researched by using a thermogravimetric analysis technique. The results revealed that during Mo oxidation and evaporation, the oxygen partial pressure around powders can be dramatically decreased to ensure the mass balance of oxygen. Under guidance of this finding, the oxygen consumption rate by Mo release and the oxidation rate of nuclear fuel in accident conditions were estimated and compared. It is suggested that Mo release can retard the oxidation progress of fuel.


Stress evaluation method by neutron diffraction for HCP-structured magnesium alloy

Harjo, S.; Gong, W.; 川崎 卓郎

Quantum Beam Science (Internet), 7(4), p.32_1 - 32_13, 2023/12

Tensile deformation in situ neutron diffraction of an extruded AZ31 alloy was performed to validate conventional procedures and to develop new procedures for stress evaluation from lattice strains by diffraction measurements of HCP-structured magnesium alloys. Increases in the lattice strains with respect to the applied true stress after yielding largely vary among [${it hk.l}$] grains. The newly proposed procedure of stress evaluation from the lattice strains shows very high accuracy and reliability by weighting the volume fraction of [${it hk.l}$] grains and evaluating them in many [${it hk.l}$] orientations in addition to multiplication by the diffraction elastic constant. When multiple ${it hk.l}$ peaks cannot be obtained simultaneously, we recommend to use the 12.1 peak for stress evaluation.



加藤 友彰; 山岸 功

JAEA-Technology 2023-018, 53 Pages, 2023/11




幌延深地層研究計画; 令和5年度調査研究計画

中山 雅

JAEA-Review 2023-019, 70 Pages, 2023/11





須藤 彩子; M$'e$sz$'a$ros, B.*; 佐藤 拓未; 永江 勇二

JAEA-Research 2023-007, 31 Pages, 2023/11





飛田 実*; 後藤 勝則*; 大森 剛*; 大曽根 理*; 原賀 智子; 青野 竜士; 今田 未来; 土田 大貴; 水飼 秋菜; 石森 健一郎

JAEA-Data/Code 2023-011, 32 Pages, 2023/11


日本原子力研究開発機構の研究施設等から発生する放射性廃棄物は、放射能レベルに応じて将来的にトレンチとピットに分けて浅地中埋設処分される予定であり、埋設処分を開始するまでに、廃棄体の放射能濃度を評価する方法を構築する必要がある。そこで、原子力科学研究所バックエンド技術部では、研究施設等廃棄物に対する放射能濃度評価方法の検討に資するため、JRR-3、JRR-4及び再処理特別研究棟から発生した放射性廃棄物よりコンクリートを試料として採取し、放射化学分析を実施した。本報告書は、令和3年度から令和4年度に取得した23核種($$^{3}$$H、$$^{14}$$C、$$^{36}$$Cl、$$^{41}$$Ca、$$^{60}$$Co、$$^{63}$$Ni、$$^{90}$$Sr、$$^{94}$$Nb、$$^{rm 108m}$$Ag、$$^{137}$$Cs、$$^{133}$$Ba、$$^{152}$$Eu、$$^{154}$$Eu、$$^{rm 166m}$$Ho、$$^{234}$$U、$$^{235}$$U、$$^{238}$$U、$$^{238}$$Pu、$$^{239}$$Pu、$$^{240}$$Pu、$$^{241}$$Am、$$^{243}$$Am、$$^{244}$$Cm)の放射能濃度データについて整理し、放射能濃度評価法検討のための基礎資料としてまとめたものである。


用語解説「放射性廃棄物(Radioactive wastes)」

松枝 誠

知能と情報, 35(4), P. 88, 2023/11



Microstructural evolution of intermetallic phase precipitates in Cr-coated zirconium alloy cladding in high-temperature steam oxidation up to 1400$$^{circ}$$C

Mohamad, 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

The steam oxidation test on the Cr-coated Zry cladding was studied up to 1400$$^{circ}$$C to understand the oxidation behavior under the accidental conditions. The double-sided oxidation test study showed that Cr coating can protect Zry cladding at 1200$$^{circ}$$C within 5 min. Cr coating has a protective effect on the Zry cladding up to 1200$$^{circ}$$C in a steam environment. However, in the oxidation test up to 1200$$^{circ}$$C/30 min and 1300$$^{circ}$$C/5 min, Cr coating can no longer protect Zry cladding. Furthermore, at 1300$$^{circ}$$C, the intermetallic phase of the Zr(Cr, Fe)$$_{2}$$ 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 1350$$^{circ}$$C/60 min and 1400$$^{circ}$$C/30 min within the ZrO$$_{2}$$ phase (outer side region). The oxidation of the Zr(Cr, Fe)$$_{2}$$ 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.



三星 夏海; 長谷 竹晃; 小菅 義広*; 鈴木 梨沙; 岡田 豊史

第44回日本核物質管理学会年次大会会議論文集(インターネット), 4 Pages, 2023/11

中性子計測による燃料デブリ中の核燃料物質定量において、性状によって変化する中性子漏れ増倍率の評価が課題の一つである。本試験では、中性子吸収材等をMOX試料の周囲に配置し、燃料デブリを模擬した試料を中性子測定装置にて測定した結果、DDSI(Differential Die-away Self-Interrogation)法は、中性子漏れ増倍率の評価に有効であることを明らかにした。


Estimation of continuous distribution of iterated fission probability using an artificial neural network with Monte Carlo-based training data

Tuya, D.; 長家 康展

Journal of Nuclear Engineering (Internet), 4(4), p.691 - 710, 2023/11



An Estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator

丸山 修平; 遠藤 知弘*; 山本 章夫*

Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11

A new estimation method of an unknown covariance, which is defined by the difference between the true covariance (the population covariance) and a prior covariance assumed by an analyst, is proposed. The unknown covariance is estimated using an empirical covariance consistent with the observed data. To estimate the unknown covariance, an unbiased and consistent estimator in regression analysis has been incorporated into the conventional cross-section adjustment. This estimator does not require assumptions for the probability distribution of the observation data. The statistical properties of this estimator were numerically verified. In addition, the effectiveness of the proposed method was confirmed by another numerical test using actual integral experimental data. In the second numerical test, the modeling uncertainty (covariance) due to the deterministic analysis method was assumed to be unknown. The results showed that the proposed method could practically estimate the unknown covariance and adjusted cross-sections using only prior information on covariances.


Mutual separation of Am and Cm using ADAAM (alkyl DiAmide AMine) and reduction of volumes for liquid waste generated via batch-wise multistage extractions

佐々木 祐二; 金子 政志; 伴 康俊; 鈴木 英哉*

Journal of Nuclear Science and Technology, 11 Pages, 2023/11

アルキルジアミドアミン(ADAAM)を使ったAm/Cm相互分離を行った。ADAAMは硝酸-ドデカン系で非常に高いAm/Cm分離比5.9を示した。1.5M硝酸-0.2M ADAAM条件を用いる多段抽出で抽出後の有機相中にAm 96.5%、Cm 1.06%回収できることを確認した。Am/Cm相互分離後に発生した水相、有機相体積削減のための付加的な多段抽出を行い、Am, Cmを2, 3段の水相に濃縮できることを確認した。



大澤 崇人

研究開発リーダー, 20(8), p.7 - 11, 2023/11


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