Predictive and inverse modeling of a radionuclide diffusion experiment in crystalline rock at ONKALO (Finland)

Soler, J. M.*; Keklinen, P.*; Pulkkanen, V.-M.*; Moreno, L.*; Iraola, A.*; Trinchero, P.*; Hokr, M.*; ha, J.*; Havlov, V.*; Trpkoov, D.*; et al.

Nuclear Technology, 209(11), p.1765 - 1784, 2023/11

https://doi.org/10.1080/00295450.2023.2209234

The REPRO-TDE test was performed at a depth of about 400 m in the ONKALO underground research facility in Finland. Synthetic groundwater containing radionuclide tracers (HTO, Cl-36, Na-22, Ba-133, Cs-134) was circulated for about four years in a packed-off interval of the injection borehole. Tracer activities were additionally monitored in two observation boreholes. The test was the subject of a modelling exercise by the SKB GWFTS Task Force. Eleven teams participated in the exercise, using different model concepts and approaches. Predictive model calculations were based on laboratory-based information concerning porosities, diffusion coefficients and sorption partition coefficients. After the experimental results were made available, the teams were able to revise their models to reproduce the observations. General conclusions from these back-analysis calculations include the need for reduced effective diffusion coefficients for Cl-36 compared to those applicable to HTO (anion exclusion), the need to implement weaker sorption for Na-22, compared to results from laboratory batch-sorption experiments, and the observation of large differences between the theoretical initial concentrations for the strongly-sorbing Ba-133 and Cs-134 and the first measured values a few hours after tracer injection. Different teams applied different concepts, concerning mainly the implementation of isotropic vs. anisotropic diffusion, or the possible existence of Borehole Disturbed Zones around the different boreholes. The role of microstructure was also addressed in two of the models.

Development plan for coupling technology between high temperature gas-cooled reactor HTTR and hydrogen production facility, 1; Overview of the HTTR heat application test plan to establish high safety coupling technology

野本 恭信; 水田 直紀; 守田 圭介; 青木 健; 沖田 将一朗; 石井 克典; 倉林 薫; 安田 貴則; 田中 真人; 井坂 和義; et al.

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05

JAEA initiated an HTTR heat application test plan to develop for coupling technology between HTGR and hydrogen production facility. The principal objective of this test plan is to establish the high safety coupling technology for coupling a hydrogen production facility to HTGR through the demonstration of a hydrogen production by the proven technology of methane steam reforming method utilizing the HTTR as a high temperature heat source. The other objective is to develop for coupling equipment such as a high temperature isolation valve, a helium gas circulator and a high temperature insulation pipe. This paper describes the overview of an HTTR heat application test plan such as a draft test schedule and test targets for the demonstration of a hydrogen production. This paper also presents basic specifications of an HTTR heat application test facility such as the HTTR modification strategy, overall system configuration and heat and mass balance at rated test operation for the demonstration of a hydrogen production. Furthermore, the operation plan during the normal start-up and shut-down processes is proposed.

海外炉を用いた中性子照射試験,1; キャプセル温度制御システムを用いた照射試験の検討(共同研究)

高部 湧吾; 大塚 紀彰; 冬島 拓実; 佐谷戸 夏紀; 井上 修一; 森田 寿; Jaroszewicz, J.*; Migdal, M.*; 小沼 勇一; 飛田 正浩*; et al.

JAEA-Technology 2022-040, 45 Pages, 2023/03

JAEA-Technology-2022-040.pdf:6.61MB

中性子照射場として中核を担ってきた材料試験炉(Japan Materials Testing Reactor: JMTR)の廃止に伴い、軽水炉の一層の安全性、信頼性・効率性向上のための技術開発や革新的な原子炉開発に必要な国内照射場が喪失され、照射試験炉の運転技術や照射技術の継承や人材育成も困難な状況となった。こうした課題に対処するため、代替手段として中性子照射場を海外炉に求めた照射試験の実施に係る検討を行った。「ポーランド国立原子力研究センターと日本原子力研究開発機構との間の試験研究炉の研究開発のための共同研究取決め」に基づきポーランド国立原子力研究センター(NCBJ)が所有するMARIA炉(出力30MW)を中性子照射場として、JMTRの有する照射技術の一つである温度制御システムを導入した照射試験の実施可否を検討した。その結果、JMTRの設計・製作基準に則って製作済であったキャプセルに対し改造を施すことで照射試験の実現が可能である見通しが得られた。改造後に浸透探傷検査、絶縁導通試験及びキャプセルの使用温度である室温300Cの範囲における動作試験等を実施し、良好な結果が得られ、MARIA炉への輸送前準備を完了した。

Radioactive particles from a range of past nuclear events; Challenges posed by highly varied structure and composition

Johansen, M. P.*; Child, D. P.*; Collins, R.*; Cook, M.*; Davis, J.*; Hotchkis, M. A. C.*; Howard, D. L.*; Howell, N.*; 池田 篤史; Young, E.*

Science of the Total Environment, 842, p.156755_1 - 156755_11, 2022/10

https://doi.org/10.1016/j.scitotenv.2022.156755

While they have appeared only recently in earth's history, radioactive particles from anthropogenic sources are widespread in global environments and present radiological harm potentials to living organisms. Here we compare a varied set of particles from past nuclear fission and non-fission sources in Australia of highly diverse magnitudes, release modes, and environments. Numerous radioactive particles persist in soils 60 + years after their release events. Particles can be distinguished by their Ca/Fe and Si/Fe elemental ratios, which in this study range over orders of magnitude and reflect the materials available during their individual formation events. The particles from nuclear testing have dominant Pu activity concentrations, relative to Sr and Cs, which increases long-term radiological hazard from alpha emissions if inhaled or ingested, and contrasts with particles from nuclear power accidents (e.g., Fukushima). Internal fracturing is more prevalent than previously reported, and fracturing is greater in Ca-rich vs. Si-dominated matrices.

Study on the effect of long-term high temperature irradiation on TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; 植田 祥平; 相原 純; 柴田 大受; 坂場 成昭

Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08

https://doi.org/10.1016/j.net.2022.02.026

In the core of the WWR-K reactor, a long-term irradiation of tri-structural isotopic (TRISO)-coated fuel particles (CFPs) with a UO kernel was carried out under normal operating conditions of the high-temperature gas-cooled reactor (HTGR). This TRISO fuel was attained at the temperature of 950 to 1,100 C, and the uranium burnup of 9.9% FIMA (fission per initial metal atom) during the irradiation. The release of the gaseous fission product from the fuel was measured in-pile, and its release-to-birth (R/B) ratio was evaluated using the model developed in the High-Temperature Engineering Test Reactor (HTTR) project. After the irradiation test, fuel compacts were subjected to electric dissociation and nondestructive inspections such as X-ray radiography and gamma spectrometry. Finally, it was concluded that integrity of the TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated, and a low fuel failure fraction and a low R/B measured with krypton-88 indicated good performance and reliability of the high burnup TRISO fuel.

鉄筋コンクリート耐震壁に対する等価線形解析の適用性検討; 原子炉建屋耐震壁終局応答試験の三次元有限要素法シミュレーション解析

市原 義孝*; 中村 尚弘*; 森谷 寛*; 堀口 朋裕*; 崔 炳賢

日本原子力学会和文論文誌, 21(1), p.1 - 14, 2022/03

https://doi.org/10.3327/taesj.J20.038

本研究は、鉄筋コンクリート構造物の非線形性の影響を近似的に等価線形解析手法による地震応答解析で評価することを目的に、1996年にOECD/NEAによる国際解析コンペで使用された原子炉建屋耐震壁終局応答試験の三次元有限要素法によるシミュレーション解析を実施した。耐震壁の等価剛性及び等価減衰は、日本電気協会が提案するトリリニア型スケルトンカーブ、Cheng et al.が提案する履歴曲線より求め、せん断ひずみ調整ファクターは感度解析より0.70に決定した。その結果、せん断ひずみ=2.010程度までの試験体上部の卓越振動数,最大応答加速度,最大応答変位,慣性力-変位関係,床応答スペクトルを良く再現できることを明らかにした。本報における等価線形解析は、=4.010程度の終局破壊時の最大応答変位を過小評価している。このため、破壊直前の急激な変位の増大を含む試験結果の評価に本手法を適用する場合は、その適用性に十分留意する必要がある。

Characterizing the reactivation mechanisms of coseismic surface ruptures associated with the 2011 Mw 6.7 Fukushima-ken Hamadori earthquake in Japan through borehole hydromechanical testing

Guglielmi, Y.*; 青木 和弘; Cook, P.*; Soom, F.*; Cappa, F.*; 田中 遊雲

Tectonophysics, 819, p.229084_1 - 229084_22, 2021/11

https://doi.org/10.1016/j.tecto.2021.229084

2011年Mw6.7の福島県浜通り地震に伴う地表断層の再活性化メカニズムについて、ボーリング孔への注水試験を実施し、検討した。本研究では、断層のリアクティベーションのために、2つの断層区間を選択した。1つは地震時に地表変位が現れた塩ノ平サイト、もう1つは地震時に変位しなかった区間である水上北サイトである。断層の上下を密閉したボーリング孔に注水を行うことで、断層を固定している断層面の有効応力を段階的に減少させ、断層を破壊させる。この断層が破壊している間に、ボーリング孔の変位,流体圧,注入流量を連続的にモニターした。取得したデータを水理-力学連成モデルを用いて解析し、断層の再活性化モードを検討した。その結果、塩ノ平サイトでは、純粋なダイラタントスリップ応答を示したが、水上北サイトでは、断層の高い摩擦強度、粘着力に関連した複雑な応答を示した。また、塩ノ平サイトのすべりの解析から、塩ノ平断層の摩擦がすべり速度に基づくクーロン破壊として合理的にモデル化され、室内試験で求めた摩擦係数と整合的であった。本研究では、同一断層系の2つのセグメント間で、断層の物理的性質や破断メカニズムに大きな違いがあることを明らかにし、大規模地震時の地表破断の影響等の地震ハザードをより適切に評価するための展望を開くことができた。

Quantifying the porosity of crystalline rocks by in situ and laboratory injection methods

Mri, A.*; Mazurek, M.*; 太田 久仁雄; Siitari-Kauppi, M.*; Eichinger, F.*; Leuenberger, M.*

Minerals (Internet), 11(10), p.1072_1 - 1072_17, 2021/10

https://doi.org/10.3390/min11101072

The porosity and pore geometry of rock samples from a coherent granodioritic rock body at the Grimsel Test Site in Switzerland was characterised by different methods using injection techniques. Results from in situ and laboratory techniques are compared by applying in situ resin impregnation techniques as well as rock impregnation and mercury injection under laboratory conditions. In situ resin impregnation of the rock matrix shows an interconnected pore network throughout the rock body, providing an important reservoir for pore water and solutes, accessible by diffusion. Porosity and pore connectivity do not vary as a function of distance to brittle shear zones. In situ porosity is about half the porosity value that was determined on rock samples in the laboratory. This could be due to that such samples were affected by artefacts created since core recovery. The extrapolation of laboratory measurements to in situ conditions requires great care and may not be feasible in all cases.

Evaluation of brittle crack arrest toughness for highly-irradiated reactor pressure vessel steels

岩田 景子; 端 邦樹; 飛田 徹; 廣田 貴俊*; 高見澤 悠; 知見 康弘; 西山 裕孝

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07

The crack arrest fracture toughness, K, values for highly-irradiated reactor pressure vessel (RPV) steels are estimated according to a linear relationship between crack arrest toughness reference temperature, T, and the temperature corresponding to a fixed arrest load, equal to 4 kN, T, obtained by instrumented Charpy impact test. The relationship between T derived from the instrumented Chrapy impact test and fracture toughness reference temperature, T, was expressed as an equation proposed in a previous report. The coefficients in the equation could be fine-tuned to obtain a better fitting curve using the present experimental data and previous K data. The K curve for RPV;A533B class1 steels irradiated up to 1.310 n/cm (E 1 MeV) was compared with a K curve defined in JEAC4206-2016. It was shown that the K curve was always lower than the 1%ile curve of K for these irradiated RPV steels. This result indicates that the conservativeness of the method defined in JEAC4206-2016 to evaluate K using K curve is confirmed for highly-irradiated RPV steels.

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; 石塚 悦男; Ho, H. Q.; 坂本 直樹*; 奥村 啓介; 竹本 紀之; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

https://doi.org/10.1016/j.fusengdes.2020.112181

JMTRとJRR-3Mの運転中に一次冷却水へ放出されるトリチウムについて研究してきた結果、ベリリウム中性子反射体の二段核反応によるLi(n,)Hで生成する反跳トリチウムが主要因であることが明らかになった。この結果から、一次冷却水へ放出するトリチウムを少なくするためには、ベリリウム中性子反射体の表面積を小さくするか、他の材料で反跳トリチウムを遮蔽する必要がある。本報告では、ベリリウム中性子反射体のトリチウム反跳防止材の概念検討として、Al, Ti, V, Ni, Zr等の多様な材料を候補材として、障壁厚み、長期運転後の放射能、反応度への影響を評価した。この結果、Alがベリリウム中性子反射体のトリチウム反跳防止材として適した候補材になり得るとの結果を得た。