Benchmarks of iron nuclear data for fusion neutron sources

権 セロム*; 今野 力; 本田 祥梧*; 見城 俊介*; 佐藤 聡*

Fusion Engineering and Design, 223, p.115548_1 - 115548_8, 2026/02

https://doi.org/10.1016/j.fusengdes.2025.115548

核融合中性子源設計で使われる最新の核データライブラリ(FENDL-3.2b, JENDL-5, ENDF/B-VIII.0とJEFF-3.3)の鉄データの精度検証のため、QST/TIARAで行われた準単色40と65MeV中性子を用いた鉄実験とJAEA/FNSで行われたDT中性子を用いた鉄実験を使い、最新の核データライブラリのベンチマークテストを行なった。テストの結果、以下のことが判明した。(1)65MeV中性子を用いたTIARA実験で、FENDL-3.2bを用いた計算結果は10-60MeVの連続エネルギー領域の中性子束を40%過小評価、(2)FNS実験で、FENDL-3.2bを用いた計算結果は体系内70cmの深さで10MeV以上の中性子束を20%過小評価し、体系内10cmの深さで10keV以下の中性子束を30%過大評価。これらの問題を詳細に調べ、その原因を特定した。

Separation of Rh(III) and direct electrodeposition in phosphonium-based ionic liquids with electrochemical and spectroscopic analyses for extracted Rh(III) complex

徳満 駿*; 松宮 正彦*; 佐々木 祐二

Separation and Purification Technology, 382(Part 2), p.135631_1 - 135631_9, 2026/02

https://doi.org/10.1016/j.seppur.2025.135631

Solvent extraction (SX) of Rh(III) and direct electrodeposition (ED) of Rh metal were carried out in five cycles using phosphonium-based ionic liquids (ILs). The comparatively high extraction performance (70.1%-86.2%) of Rh(III) was stably retained in each SX process. In the subsequent ED process, the electrodeposits were recovered on the cathode and the reasonable current efficiencies (74.9%-85.6%) were achieved in the IL system. The X-ray photoelectron spectroscopy revealed that the electrodeposits obtained during ED process were found to be in the metallic state.

Development of LASSO based optimized scheme for reconstructing radioactive source distributions using monitoring air dose rates

Shi, W.*; 町田 昌彦; 山田 進; 岡本 孝司*

Measurement, 258(Part D), p.119444_1 - 119444_15, 2026/01

https://doi.org/10.1016/j.measurement.2025.119444

Clarifying the distribution of radioactive sources within nuclear facilities is crucial for ensuring worker safety during decommissioning and for responding to accidents. However, air dose rate measurements in restricted areas are often limited due to complex structures and high radiation levels in contaminated rooms. To address this, we have proposed a machine learning-based approach, the Least Absolute Shrinkage and Selection Operator (LASSO), to reconstruct radioactive source distributions in simplified room models. LASSO method indicates the good performance of reconstructing radioactive source with high accuracy inside simple room model. However, in more complex environments, obstacles can degrade reconstruction accuracy. To overcome these limitations, we developed an optimized scheme based on the LASSO method to improve inverse estimation in complex rooms. In this scheme, the impact of shielding structures is mitigated by normalizing the radioactive contributions from sources. A series of numerical simulations demonstrate that the optimized approach outperforms the non-optimized version in accurately reconstructing source distributions. Furthermore, experiments in a room with complex structures validate the effectiveness of the optimized method. The inverse estimations performed on experimental data confirm that the use of a normalized contribution matrix significantly improves accuracy by reducing the influence of shielding. Conclusively, this paper optimizes LASSO scheme for reconstructing radioactive source distributions in complex building room using air dose rate measurements. It shows significant improvements over existing scheme and is verified to be successfully applied in complicated situations with high accuracy. We confirm that optimized LASSO scheme holds significant promise for future monitoring and decommissioning projects in both operational and damaged nuclear facilities.

Experimental study of AESOP code for aerosol removal behavior from a rising gas bubble in water pool and parametric study for application to sodium pool system

宮原 信哉*; 鯉江 竜輔*; 宇埜 正美*; 河口 宗道*; 佐藤 理花; 清野 裕

Nuclear Engineering and Design, 446(Part A), p.114523_1 - 114523_14, 2026/01

https://doi.org/10.1016/j.nucengdes.2025.114523

In a postulated accident of fuel pin failure of a sodium-cooled fast reactor, a fission product of cesium will be released from the failed pin as an aerosol such as cesium iodide and/or cesium oxide together with a fission product noble gas such as xenon and krypton. The xenon and krypton released with the cesium aerosols into the sodium coolant as bubbles have an influence on the removal of cesium aerosols by the sodium pool in a period of bubble rising to the sodium pool surface. Then, the cesium aerosols could transfer into the containment vessel as an initial inventory of a source term. To meet this phenomenon, the computer program AESOP (AErosol scrubbing in SOdium Pool) has been developed to deal with the expansion and the deformation of the bubble together with the aerosol absorption considering the effects of the particle size distribution and the agglomeration in aerosols. In this study, simulation experiments have been conducted using simulant particles under the condition of room temperature in water pool and nitrogen gas bubble systems and the experimental results were compared with the analysis results calculated under the same condition by the AESOP code. Furthermore, to examine the applicability of the AESOP code to the sodium pool system, the sensitivities of the physical parameters on decontamination factor (DF) of fission product aerosols such as the initial bubble diameter, the sodium pool depth and the temperature, the aerosol particle diameter and the density, the initial aerosol concentration in the bubble had been studied and the analysis results were discussed for the sensitivities of the parameter as same as DF of the aerosol.

福島第一原子力発電所2号機原子炉格納容器貫通部X-6内の堆積物の分析

米山 海; 二田 郁子; 田中 康之; 小高 典康; 菊池 里玖; 坂野 琢真; 古瀬 貴広; 佐藤 宗一; 三本木 満; 田中 康介

JAEA-Technology 2025-008, 44 Pages, 2025/12

JAEA-Technology-2025-008.pdf:4.3MB

東京電力ホールディングス株式会社福島第一原子力発電所(1F)の廃炉に向け、原子炉建屋格納容器内部の調査が行われている。燃料デブリの取出しや建屋解体の作業を安全に進めるためには、汚染状況を把握し、作業の計画や作業者の被ばくを管理する必要がある。本件は、2号機原子炉格納容器貫通部X-6(X-6ペネ)内の堆積物について、含まれる元素、放射性核種濃度、核種組成を把握することを目的に分析を実施した。本分析の対象試料は、スミヤろ紙に付着したX-6ペネ内部の堆積物である。堆積物に含まれる核種の把握、また、元素や元素の共存の様子を把握するため、非破壊分析として線スペクトル分析、蛍光X線(XRF)分析、走査型電子顕微鏡-エネルギー分散型X線(SEM-EDX)分析を実施した。さらに、堆積物に含まれる放射性核種やその組成を詳細に明らかにするために、堆積物を硝酸及びフッ化水素酸で溶解し、溶解液中の核種、Sr-90及び核種の放射能分析を実施した。得られた結果を、2020年にX-6ペネ内の異なる場所で採取された堆積物の分析結果と比較した。非破壊での線スペクトル分析では、Co-60、Sb-125、Cs-134、Cs-137、Eu-154、Eu-155及びAm-241が検出された。XRF分析では、格納容器内の構造物由来と考えられるFeが主要な元素として検出され、そのほか燃料や燃料被覆管に由来すると考えられる微量のU及びZrが検出された。SEMEDX分析の結果では、堆積物の主要な元素としてOとFeが検出されたことに加え、Uを含む粒子が観察され、UとともにFe、Si、Cr、Ni、Zrが検出された。これらの結果は2020年採取試料と同様の傾向であった。放射能分析では、非破壊測定で検出された核種(Co-60、Sb-125、Cs-134、Cs-137、Eu-154、Eu-155)に加えて、Sr-90、Pu-238、Pu-239+240、Am-241、Cm-244、U-235、U-238の定量値を得た。これらの結果をもとに、1F事故に由来する汚染の主要な線放出核種であるCs-137を基準とした放射能比を算出した。さらに、U-238に対する放射能比についても算出し、ORIGENによる2号機の燃料組成の計算値と比較した。

水素挙動統合解析システムの開発

寺田 敦彦; Thwe Thwe, A.; 日野 竜太郎*; 原井 康孝*; 佐々木 岳*; 新家谷 英之*; 山下 俊幸*; 米田 次郎*; 岡林 一木*; 坂本 裕之*; et al.

JAEA-Data/Code 2025-012, 151 Pages, 2025/12

JAEA-Data-Code-2025-012.pdf:9.69MB

福島第一原子力発電所事故の経験や、事故から得られた教訓を踏まえ、原子炉のみならず廃止措置、廃棄物管理における水素安全評価・対策に適切に対応するための基盤技術の高度化を図ることを目的として、水素の発生から拡散、燃焼・爆発に至る挙動を予測する解析システムの開発を行った。本システムでは、汎用コード(FLUENT、AUTODYN)を活用し、そこに新規にモジュールやプリ/ポストプロセッサを組み込むことで、一般の実用に堪える解析システムを整備するとともに、より高い汎用性と低コストでの導入が可能なオープンソースコード(OpenFOAM)を活用したシステムの開発を並行して進め、原子力施設の水素防災計画に利用できる形での基盤技術の提供を目指している。これまで、PWR 原子力発電施設を対象に、実用的な観点から考慮すべき現象(火炎伝播加速現象の評価技術、格納容器規模の現象への適用性)に対処するためのシステムの拡充を行った。本報告書は、水素挙動統合解析システムの概要、取り扱い方法及び実機解析事例についてまとめたものである。

A Methodology for the design of non-uniform core configurations in the modified STACY facility

Dechenaux, B.*; Brovchenko, M.*; 荒木 祥平; 郡司 智; 須山 賢也

Annals of Nuclear Energy, 223, p.111555_1 - 111555_11, 2025/12

https://doi.org/10.1016/j.anucene.2025.111555

The safe retrieval of the fuel debris generated during the Fukushima Daiichi nuclear accident poses a number of challenges, among which an important one is to ensure the criticality safety of the recovery operations. At the heart of the problem lies the intrinsically heterogeneous nature of the problem, and the appearance of complex and disordered media whose neutronic properties are difficult to accurately characterize and reproduce in neutron transport simulations. Typically, a similar, simpler, problem is encountered in the modeling of assemblies with missing fuel rods. In both problems, the availability of experimental facilities capable of validating both the nuclear data and the simulation schemes in heterogeneous configurations is crucial. The modified STACY installation has been specially designed to provide and carry out critical experiments, allowing for highly non-uniform configurations, that will directly support fuel debris recovery operations, but can also be used for other experimental programs. The present work describes a method to consistently and orderly sample non-uniform core configurations in the modified STACY facility, and proposes two critical heterogeneous core configurations. They have the advantage to present a high sensitivity to the water thermal scattering law, whose importance was found to be more significant for more heterogeneous configurations. The proposed experiments will contribute to the exploration and validation of heterogeneous critical systems.

Flow regimes and heat transfer for opposing flow mixed convection in the thermal entry region of a vertical tube

茂木 孝介; 柴本 泰照; 日引 俊詞*

International Journal of Heat and Mass Transfer, 252, p.127451_1 - 127451_16, 2025/12

https://doi.org/10.1016/j.ijheatmasstransfer.2025.127451

This study focused on the thermal entry length problem for turbulent opposing flow mixed convection in a vertical tube. A Reynolds-averaged Navier-Stokes (RANS) simulation was performed using the turbulence model, and its results were compared with previous experimental data. The simulation results revealed several flow structures and heat transfer characteristics in the entry region, which varied depending on the competing strength of forced and natural convection. Flow regimes were classified based on their flow structures as follows: (a) Non-separating flow regime: When natural convection minimally influences the flow field, the Nusselt number in the entry region is higher than that in the fully developed region. As the influence of natural convection strengthens, the entry length decreases. (b) Separation bubble regime: As the influence of natural convection on the flow field increases, the velocity boundary layer along the heated wall in the entry region separates, leading to the formation of a recirculation region known as a separation bubble. Here, the depression in the Nusselt number distribution was observed owing to the thickening of the thermal boundary layer caused by the separation bubble. (c) Reverse flow regime: Further strengthening of the influence of natural convection leads to the formation of reverse flow along the entire heated wall. Here, the Nusselt number exhibits a nearly flat distribution in the entry region because the reverse flow thickens the thermal boundary layer. A flow regime map based on the results of the RANS simulation was created and proposed in this paper.

Plastics as vectors of radiocesium in river environments of Fukushima, Japan

Battulga, B.*; 中西 貴宏; 池之上 翼; 安藤 麻里子; 小嵐 淳

Journal of Hazardous Materials, 500, p.140593_1 - 140593_11, 2025/12

https://doi.org/10.1016/j.jhazmat.2025.140593

Plastic debris is widely recognized as a major environmental challenge, yet its interactions with radionuclides remain poorly understood. Here, we investigated the behavior and fate of plastics and associated radiocesium (Cs) in the coastal river environments of Fukushima, Japan, to elucidate their interactions and the role of plastics as vectors for Cs transport into the coastal ocean. At four river mouth sites, we characterized the properties of plastics, including microplastics, and elucidated the activity concentration of Cs associated with them. Our findings reveal that Cs was concentrated in biofilms on plastics, but the radioactivity of biofilm-associated plastics was lower than that of the surrounding natural media. Furthermore, we estimated the daily fluxes of plastics (1.3-301.4 kg/day) and associated Cs (5-6,752 Bq/day) in the studied rivers across different seasons. This study provides the first quantitative assessment of plastic-associated Cs flux in the riverine environments.

バイオアッセイ迅速分析法の確立に向けた検討

松尾 一樹; 井上 和美; 國分 祐司; 藤田 博喜

KEK Proceedings 2025-2(インターネット), p.124 - 129, 2025/12

https://lib-extopc.kek.jp/preprints/PDF/2025/2525/2525002.pdf

福島第一原子力発電所の廃炉作業において、作業者による放射性物質の内部取り込みが発生した場合には、内部被ばくによる線量評価を行うために様々な線放出核種や純線放出核種を対象としたバイオアッセイが必要となる。このバイオアッセイの手順に関して、放射能測定法シリーズのような標準マニュアル等はこれまでに制定されていない。そこで、本研究では、複数の放射性元素(Sr、U、Pu、Am、Cm)を対象としたより迅速な分析法の確立を目指し、文献を調査して分析手順を作成し、その手順が利用できる可能性を評価するために、化学分析回収率を求めた。なお、この検討には、模擬便と模擬尿を使用した。作成した分析法を用いて回収率を求めたところ、模擬尿(n=3)では、Puについては55%-75%程度、Uについては50%-80%程度、Amについては90%-100%程度であった。一方、模擬便(n=3)については、回収率がPuについては50%-70%程度、Uについては90%程度、Amについては100%前後であった。なお、模擬便においては、図の分析フローではUの回収率が低下したため、UTEVAレジンをTEVAとDGAの間に加えた。なお、Srについては回収率測定中である。作成したバイオアッセイ系統分析法の分析フローに対して、回収率取得実験を実施した。その結果、本研究で検討した線放出核種については50%以上の回収率を得られたため、本分析手順はバイオアッセイに利用できる可能性が示された。