Uncovering hidden dispersion patterns of radioactive cesium-rich microparticles from Fukushima Daiichi

宮崎 加奈子*; 笛田 和希*; 門脇 正尚; 寺田 宏明; 香西 直文; 岩田 孟; 堀江 憲路*; 竹原 真美*; 山崎 信哉*; Grambow, B.*; et al.

Journal of Hazardous Materials, 511, p.142180_1 - 142180_17, 2026/06

https://doi.org/10.1016/j.jhazmat.2026.142180

Radioactive cesium-rich microparticles (CsMPs) released from the Fukushima Daiichi Nuclear Power Plant (FDNPP) in 2011 pose a persistent environmental concern, yet their initial atmospheric dispersion has remained poorly constrained. To elucidate their release dynamics, we quantified CsMPs in 100 surface soil samples collected across Fukushima Prefecture in July 2011. CsMP abundance and radioactive fraction (RF) ranged from 0 to 52,300 particles kg; (dry weight) and 0-61.85% of deposited Cs, respectively. Integrating these results with WSPEEDI atmospheric simulations reveals a major formation event of CsMPs around 03:00 JST on 15th March 2011. The plume emitted at this time was heavily enriched in CsMPs, whereas plumes released after 00:00 JST on 16th March 2011 contained no detectable CsMPs, indicating that particle formation had ceased by then. The widespread distribution of CsMPs across Fukushima Prefecture is therefore attributed primarily to this single plume on 15th March 2011. Directional differences in CsMP abundance reflect temporal variations in CsMP density within the plume, with maximum estimated concentrations of 2,070 particles m; toward the southwest and 4,700 particles m; toward the northwest. Additionally, elevated CsMP densities coupled with relatively low RF in the northwest suggest supplementary deposition of water-soluble cesium between 17:00 and 23:00 JST on 15 March. These findings constrain the timing and mechanisms of CsMP formation and dispersion, providing essential insights for environmental reconstruction and future radiological risk assessment.

Transient behavior of a boiling water reactor-type offshore floating nuclear power plant during platform motion

福田 航大; 小原 徹*

Nuclear Technology, 212(6), p.1567 - 1578, 2026/06

https://doi.org/10.1080/00295450.2025.2502266

Offshore floating nuclear power plants (OFNPs) are gaining attention as safe and location-flexible means for nuclear energy utilization. Although platform motion in the marine environment may affect reactor kinetics, particularly in boiling water reactors (BWRs), BWR-type OFNPs are expected to have high economic efficiency. This study investigated the reactor power behavior of a BWR-type OFNP using three-dimensional transient neutronics-thermal hydraulics coupled analysis. Heave and pitch motions were considered under realistic wave conditions using a typical BWR model. The results show that the reactor power and its distribution can vary because of the wave-induced platform motion; however, the amplitude of these variations is sufficiently small to ensure normal operation, even under the extreme wave conditions of a one-in-10,000-year storm. Although the results of the present study demonstrate the ability of BWR-type OFNPs to provide a safe and stable energy supply, they also suggest the need for further research. Further studies are needed to address the complex wave conditions and assess the effects of the platform motion on ancillary systems, such as recirculation systems.

An Estimation method for the total amount of Cs contamination on the bottom of top shield plug in the Fukushima Daiichi Nuclear Power Plant Unit 2

神野 郁夫; 松村 太伊知; 奥村 啓介

Journal of Nuclear Science and Technology, 7 Pages, 2026/00

https://doi.org/10.1080/00223131.2026.2725018

The estimation of the total amount of Cs contamination on the bottom of the top shield plug in the Fukushima Daiichi Nuclear Power Plant Unit 2 is one of the key issues for pushing the decommissioning forward. Direct measurement of the Cs contamination amount is, however, difficult: the contamination is on the other side of a 61 cm thick concrete shield plug, i.e. opposite side of the surface of the shield plug on which ambient dose rate measurements were performed. This paper proposes a method for estimating the total amount of the Cs contamination on the bottom of the top shield plug using a dose rate meter in a lead collimator to define the contamination area from which gamma-rays are emitted. The proposed method is applied to the dose rates measured at 12 points on the shield plug surface. The estimated total amount of Cs contamination is about 30 PBq.

再処理施設の高レベル廃液蒸発乾固事故時の化学挙動解析コードSCHERNのRuO気液間移行モデルの高度化

吉田 一雄; 桧山 美奈*; 玉置 等史

JAEA-Research 2025-011, 25 Pages, 2025/11

JAEA-Research-2025-011.pdf:2.15MB

再処理施設の過酷事故の一つである高レベル放射性廃液貯槽の冷却機能喪失による蒸発乾固事故では、沸騰により廃液貯槽から発生する硝酸-水混合蒸気とともにルテニウムの揮発性の化学種(RuO)が放出される。このためリスク評価の観点からは、Ruの定量的な放出量の評価が重要な課題である。RuOは施設内を移行する過程で床面に停留すると想定されるプール水中の亜硝酸(HNO)によって化学吸収が促進されることが想定され、この挙動は実験的に確認されており、Ruの施設内での移行に重要な役割を担う。HNOを含む硝酸水溶液へのRuOの移行に係る実験から得られた成果をもとに、新たな化学吸収及び物理吸収モデルが提案されている。本報では、SCHERNの解析性能の向上の一環として、これらの吸収モデルを組込み、施設内を移行するRuOの気液各相での挙動の解析を試行した。その結果、RuOの放出が急激に増加する沸騰晩期では、液相中のHNOも急増する傾向が見られ、その濃度変化がその後のRuOの移行挙動に大きく影響することを確認した。この結果から硝酸-水混合蒸気の凝縮に伴う気液各相のHNOの化学的挙動の解析精度の向上が不可欠である。

Design policy of pilot plant for accelerator-driven system

西原 健司; 菅原 隆徳; 福島 昌宏; 岩元 大樹; 方野 量太; 阿部 拓海

Progress in Nuclear Science and Technology (Internet), 8, p.185 - 190, 2025/09

https://doi.org/10.15669/pnst.8.185

マイナーアクチノイド核変換用の熱出力800MWの鉛ビスマス冷却ADSのスケールダウンとして、加速器駆動システムのパイロットプラントを提案する。本論文では、安全性、未臨界管理、未臨界炉心、加速器、ターゲット及び炉内機器の各設計領域ごとにパイロットプラントの設計方針を詳述する。また、以下のようなパイロットADSの概念設計も示す:熱出力200MW、keff0.95以下、制御棒の導入、1.0GeV-10mA加速器、および炉内機器の概算寸法

Integrity evaluation of boundary function of main components in nuclear plants during severe accidents

月森 和之; 矢田 浩基

Journal of Pressure Vessel Technology, 147(3), p.031901_1 - 031901_9, 2025/06

https://doi.org/10.1115/1.4067760

日本では、福島第一原子力発電所の事故以来、原子力プラントに対して厳しい安全対策が取られている。シビアアクシデント時において、放射性物質を内包する危機がそのバウンダリ機能を維持できるか否かが重要な関心事となる。本研究では、高速炉の1時冷却材を内包する容器等のバウンダリを構成する構造部材である鏡板とベローズに着目し、設計を超える過大な圧力を受けた場合の座屈、座屈後変形、さらにバウンダリの貫通破損(バウンダリ機能喪失)までの挙動を検討した。研究は、2013年度に始まり、段階的に進められたが、最終段階として、新たに提案する破損クライテリアの鏡板、ベローズへの適用結果を示すものである。

Neutronics-thermal-hydraulics-coupled transient analysis for reactor power change in an inclined offshore floating boiling water reactor

福田 航大; 小原 徹*; 須山 賢也

Nuclear Technology, 211(5), p.963 - 973, 2025/05

https://doi.org/10.1080/00295450.2024.2368966

An application of the boiling water reactor (BWR) to an offshore floating nuclear power plant (OFNP) is discussed in Japan. The BWR-type OFNP has some challenges for practical use, although it has high economic efficiency because of downsizing and simplification. One challenge is understanding reactor kinetics under conditions specific to the marine environment. This study quantitatively clarifies the total and spatial changes in power when the BWR is inclined during regular operation. Therefore, the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes were used to perform a three-dimensional neutronics-thermal-hydraulics-coupled transient analysis. The calculation model is based on Peach Bottom II. This study clarifies the changing trend in total and local BWR power by inclination with simplified modeling and conditions. Reasons for such changes are discussed based on changes in several thermal-hydraulic parameters. The difference in BWR power against the inclinations is small. Thus, it was implied that the BWR-type OFNP is expected to have a stable power supply capability during natural disasters. Finally, requires further studies to support the obtained conclusions are discussed.

Design policy of pilot plant for accelerator-driven system

西原 健司; 菅原 隆徳; 福島 昌宏; 岩元 大樹; 方野 量太; 阿部 拓海

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

マイナーアクチニド核変換用の熱出力800MWの鉛ビスマス冷却ADSのスケールダウン版として、加速器駆動システムのパイロットプラントを提案する。本発表では、パイロットプラントの設計方針について述べる。

Fukushima Daiichi Nuclear Power Plant accident analysis considering the thermal stratification and containment leakage

中村 勇気*; 小島 良洋*; 山下 拓哉; 下村 健太; 溝上 伸也

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 12 Pages, 2024/08

At the Fukushima Daiichi Nuclear Power Plant accident, it has been reported that several units of containment vessel had failed, and large quantity of radionuclides had been released into the environment. However, the detailed accident progression of such a containment failure, which includes core melt, reactor vessel failure and following containment vessel behavior, has still large uncertainties. Especially for the unit 2 and 3, they had succeeded in the initial core cooling, but at last lost their cooling system and fell into severe accident to release the fission product into the environment. Nowadays, several information has been obtained by the internal inspection into the containment of the Fukushima Daiichi Nuclear Power Plants. To clarify the uncertainties in the accident scenario, considering the information and several insights already accustomed by previous research, the latest accident scenario in unit 2 and unit 3 of the Fukushima Daiichi Nuclear Power Plants accident are suggested and tested by the severe accident analysis code, MAAP in this study. It is shown that unit 2 and 3 both accident scenario would have resulted in the thermal stratification in suppression pool which encouraged the containment pressure response in the early phase of the accident. In addition, containment vessel leakage would have occurred and affected the containment depressurization.

Quantification of risk dilution induced by correlation parameters in dynamic probabilistic risk assessment of nuclear power plants

久保 光太郎; 田中 洋一*; 石川 淳

Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 238(3), p.464 - 474, 2024/06

https://doi.org/10.1177/1748006X231167201

Nuclear power plants are critical infrastructures that provide electricity. However, accidents in nuclear power plants can cause considerable consequences such as the release of radioactive materials. Therefore, appropriately managing their risk is necessary. Various nuclear regulatory agencies employ probabilistic risk assessment (PRA) to effectively evaluate risks in nuclear power plants. Dynamic PRA has gained popularity because it allows for more realistic assessment by reducing the assumptions and engineering judgments related to time-dependent failure probability and/or human-action reliability in the conventional PRA methodology. However, removing all assumptions and engineering judgments is difficult; thus, the risk analyst, e.g., the regulator, must understand their effects on the assessment results. This study focuses on "risk dilution," which emerges from the assumptions about uncertainty. Dynamic PRA of a station blackout sequence in a boiling-water reactor was performed using the dynamic PRA tool, namely, Risk Assessment with Plant Interactive Dynamics (RAPID) and the severe-accident code Thermal-Hydraulic Analysis of Loss of Coolant, Emergency Core Cooling, and Severe Core Damage version 2 (THALES2), which altered the correlation parameters among the uncertainties of the events that occurred in sequence. The results demonstrated that the conditional core-damage probability and mean value of the core-damage time varied from 0.27 to 0.47 and from 7.1 to 8.7 h, respectively. When the dynamic PRA results are used for risk-informed decision making, the decision maker should adequately consider the effect of risk dilution.