Uncertainty quantification of Np, Am, and Am reaction rates in highly enriched uranium fuel cores at Kyoto University Critical Assembly

Pyeon, C. H.*; 大泉 昭人; 方野 量太; 福島 昌宏

Nuclear Science and Engineering, 199(3), p.429 - 444, 2025/03

https://doi.org/10.1080/00295639.2024.2380624

京都大学KUCAの高濃縮ウラン燃料による複数の炉心で取得したNp、Am、Amの核分裂率とNp捕獲反応率に関する積分実験データに対して、ENDF/B-VIII.0及びJENDL-5とSerpent2コードを用いた実験解析を実施した。Np/U、Am/U、Am/Uの核分裂率比の実験値と解析値の比較では、ENDF/B-VIII.0及びJENDL-5のいずれにおいて、それぞれ約5、15、および10%の精度が示された。Np/Auの捕獲反応率比については、熱中性子スペクトルの炉心において高い精度が得られることを明らかにした。また、Np/U、Am/U、Am/Uの核データに起因した不確かさは、概ね4%以内であることを示した。

First freezing experiments with a molten mixture of boron carbide and stainless steel in core disruptive accidents of sodium-cooled fast reactors

江村 優軌; 松場 賢一; 菊地 晋; 山野 秀将

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11

Assuming the CDA of SFRs, the eutectic melting between BC as a control rod material and stainless steel (SS) as a structural material could occur below their melting points. After that, the mixture produced by eutectic melting between BC and SS (BC-SS mixture) would relocate inside or outside of the original core region. From the viewpoint of core reactivity changes, the relocation behavior of BC-SS mixture induced by its melting/freezing behavior, is one of the key elements to evaluate the CDA consequences. Many experimental studies on freezing behavior using core materials and its simulants, including molten UO, SS, tin, wood's metal have been reported in the past. Based on these experimental findings, the freezing/blockage model for the severe accident simulation code was established and discussed through analyses of freezing process. Specifically, it has been considered that the experimental correlation of melt-penetration length was a key indicator to quantitatively describe freezing behavior. However, there was no experimental data for the freezing behavior of actual BC-SS mixture. Therefore, the freezing experiments of BC-SS mixture were conducted to investigate the freezing and blockage behavior inside a flow path such as fuel pin bundle. In the freezing experiments, BC powder and SS block were heated up to around 1,750 K using a graphite heating furnace, then BC-SS mixture flowed down into an SS pipe for cooling below 750 K. The experimental results showed that the BC-SS mixture solidified and resulted in the blockage in the SS pipe with 4 mm or 6.7 mm in inner diameter, respectively. Furthermore, the observations for cross section of SS pipe suggested that the BC-SS mixture penetrated deeper than molten SS. This difference is considered to be influenced by decrease of the melting point.

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

江村 優軌; 高井 俊秀; 菊地 晋; 神山 健司; 山野 秀将; 横山 博紀*; 坂本 寛*

Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07

https://doi.org/10.1080/00223131.2023.2287109

Boron carbide (BC)- stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using BC pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid BC with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid BC-liquid SS reaction based on the reduced thickness of BC pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid BC-solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid BC-liquid SS reaction, it was found that similar temperature dependency was identified between solid BC-liquid SS and solid BC-solid SS. Besides, the reaction rate constants of solid BC-liquid SS were smaller than those of solid BC-solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for BC side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid-solid contact for BC side/SS side.

UO dissolution in bicarbonate solution with HO; The Effect of temperature

McGrady, J.; 熊谷 友多; 北辻 章浩; 桐島 陽*; 秋山 大輔*; 渡邉 雅之

RSC Advances (Internet), 13(40), p.28021 - 28029, 2023/09

https://doi.org/10.1039/D2RA08131H

使用済核燃料の埋設処分において、キャニスタが破損して核燃燃料が地下水と接触すると、水の放射線分解により発生した酸化性物質とUOが反応する。炭酸水素イオンが存在する地下水中での、HOによるUOの酸化溶解がこれまでに研究されてきた。埋設されたキャニスタ周辺の温度は、時間や配置場所により変化するが、これまでに酸化溶解の温度効果は明らかにされていない。種々の炭酸水素イオン濃度と温度条件下でのUO表面におけるHOの反応とU溶解を調べた。炭酸水素イオン濃度や温度によりUO表面でのHOの反応機構が変化し、これによりUOの溶解速度が変化することが分かった。これは、UO表面に生成される化学種の違いに起因することが示唆された。

The Effects of unburned-gas temperature and pressure on the unstable behavior of cellular-flame fronts generated by intrinsic instability in hydrogen-air lean premixed flames under adiabatic and non-adiabatic conditions; Numerical simulation based on the detailed chemical reaction model

Thwe Thwe, A.; 門脇 敏; 永石 隆二

Journal of Nuclear Science and Technology, 60(6), p.731 - 742, 2023/06

https://doi.org/10.1080/00223131.2022.2143448

本研究では、詳細な化学反応を考慮した非定常反応流れの数値計算を遂行し、固有不安定性による水素-空気希薄予混合火炎の不安定挙動を調べ、未燃ガス温度と圧力の影響を明らかにした。広い空間における火炎の不安定挙動をシミュレートし、セル状火炎の燃焼速度を求めた。そして、火炎不安定挙動に及ぼす熱損失および火炎スケールの効果を精査した。平面火炎の燃焼速度は、未燃ガスの温度が上昇すると増加し、未燃ガスの圧力と熱損失が上昇すると減少する。平面火炎の燃焼速度で標準化したセル状火炎の燃焼速度は、圧力(温度)の上昇と共に増大(減少)する。熱損失が存在する場合、標準化したセル状火炎の燃焼速度は、断熱の場合より大きくなる。これは、未燃ガスの高圧力と熱損失は、火炎の不安定挙動と不安定性をプロモートするからである。

Status report of the n_TOF facility after the 2nd CERN long shutdown period

Patronis, N.*; 木村 敦; 他130名*

EPJ Techniques and Instrumentation (Internet), 10, p.13_1 - 13_10, 2023/05

https://doi.org/10.1140/epjti/s40485-023-00100-w

During the second long shutdown period of the CERN accelerator complex (LS2, 2019-2021), several upgrade activities took place at the n_TOF facility. The most important have been the replacement of the spallation target with a next generation nitrogen-cooled lead target. Additionally, a new experimental area, at a very short distance from the target assembly (the NEAR Station) was established. In this paper, the core commissioning actions of the new installations are described. The improvement in the n_TOF infrastructure was accompanied by several detector development projects. All these upgrade actions are discussed, focusing mostly on the future perspectives of the n_TOF facility. Furthermore, some indicative current and future measurements are briefly reported.

Evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

山本 智彦; 加藤 篤志; 早川 雅人; 下山 一仁; 荒 邦章; 畠山 望*; 山内 和*; 江田 優平*; 由井 正弘*

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 6 Pages, 2023/04

In the secondary cooling system of sodium-cooled fast reactor (SFR), a rapid detection of hydrogen explosion due to sodium-water reaction by water leakage from heat exchanger tube is steam generator (SG) is important in terms of safety and property protection. For the hydrogen detection, Ni-membrane hydrogen detectors using atomic transmission phenomenon were used in Japanese proto-type sodium-cooled fast reactor "Monju". However, during the plant operation, many alarms of water leakage were occurred without sodium-water reaction in relation to temperature up and down. The authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal operation and the hydrogen generated by the sodium-water reaction and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). As the results of theoretical estimation, dissolved H or NaH, rather than H, is the predominant form of the background hydrogen in liquid sodium, and hydrogen produced in large amounts by sodium-water reaction can exist stably as fine bubbles with a NaH layer on their surface. Currently, the authors study the new hydrogen detector system focusing on the difference between the background hydrogen (dissolved H) and the hydrogen by sodium-water reaction (fine bubbles H). This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium.

High-temperature gaseous reaction of cesium with siliceous thermal insulation; The Potential implication to the provenance of enigmatic Fukushima cesium-bearing material

Rizaal, M.; 中島 邦久; 斉藤 拓巳*; 逢坂 正彦; 岡本 孝司*

ACS Omega (Internet), 7(33), p.29326 - 29336, 2022/08

https://doi.org/10.1021/acsomega.2c03525

Here we report an investigation of the gas-solid reaction between cesium hydroxide (CsOH) and siliceous (calcium silicate) thermal insulation at high temperature, which was postulated as the origin for the formation mechanism of cesium-bearing material emitted from the Fukushima Daiichi Nuclear Power Plant. A developed reaction furnace consisting of two heating compartments was used to study the reaction at temperatures of 873, 973, and 1073 K. Under the influence of hydrogen-steam atmospheric conditions (H/HO = 0.2), the reaction between cesium hydroxide vapor and solid thermal insulation was confirmed to occur at temperatures of 973 and 1073 K with the formation of dicalcium silicate (CaSiO) and cesium aluminum silicate (CsAlSiO). Water-dissolution analyses of the reaction products have demonstrated their stability, in particular, the CsAlSiO. Constituents similarity of the field-observed cesium-bearing materials near the Fukushima Daiichi Nuclear Power Plants with CsAlSiO suggests for the first time that gaseous reaction between CsOH with calcium silicate thermal insulation could be one of the original formation mechanisms of the cesium-bearing materials.

J-PARCにおける高エネルギー陽子入射核破砕反応による核種生成断面積の測定

中野 敬太; 松田 洋樹*; 明午 伸一郎; 岩元 大樹; 竹下 隼人*; 前川 藤夫

JAEA-Research 2021-014, 25 Pages, 2022/03

JAEA-Research-2021-014.pdf:2.1MB

加速器駆動核変換システム(ADS: Accelerator-Driven transmutation System)の開発に資するデータとして、Be, C, Al, Sc, V標的に対する高エネルギー陽子入射反応による核種生成断面積の測定を行った。得られた実験値は最新の核反応モデルによる計算値や評価済み核データライブラリの値と比較を行い、その再現性について議論を行った。

工学的近似を用いたナトリウム-水反応ジェット挙動評価用粒子法コードの研究

小坂 亘; 内堀 昭寛; 柳沢 秀樹*; 高田 孝; Jang, S.*

日本機械学会論文集(インターネット), 88(905), p.21-00310_1 - 21-00310_9, 2022/01

https://doi.org/10.1299/transjsme.21-00310

仮にナトリウム(Na)冷却高速炉の蒸気発生器(SG: Steam generator)内部の伝熱管から加圧された水/水蒸気が漏えいした場合、管内外の圧力差のため高流速かつ高温の腐食性ジェットを形成される。このジェットが隣接する管に新たな破損を生じさせ、管破損が伝播する可能性がある。このような伝熱管破損伝播事象の評価はNa冷却高速炉の安全性評価において重要である。数値解析コードLEAP-IIIは、実験相関式や1次元保存式などで構成され、事象進展に伴う水リーク率変化を低計算コストで評価できるが、構成モデルのひとつである現在の反応ジェット温度分布評価モデルでは、いくつかの条件において高温領域を過度に広く評価する。低計算コストという利点を生かしつつ、この温度分布評価を精緻化するため、工学的近似に基づいたLagrange粒子法コードを開発してきた。本論文では、管群内の流体挙動評価に注目し、管近傍の粒子挙動に関する新しいモデルを構築した。本粒子コードと既往の計算流体力学コードを用いてテスト解析を行い、構築したモデルにより、ターゲット管内部に粒子が侵入しない、管周りに粒子が広がる、最終的に浮力方向への運動に帰結する、という基本的な挙動が示されることを確認した。