New insight on the thermal impact on cementitious materials due to high-temperature with water supply; Continuous expansive spalling in water

三浦 泰人*; 宮本 慎太郎*; 丸山 一平*; Aili, A.*; 佐藤 拓未; 永江 勇二; 五十嵐 豪*

Case Studies in Construction Materials, 21, p.e03571_1 - e03571_14, 2024/12

https://doi.org/10.1016/j.cscm.2024.e03571

In this study, the expansion behavior of cement materials after high-temperature heating and water immersion was observed experimentally. Two experiments were conducted using mortar specimens with different sand-to-cement ratios subjected to different high-temperature histories up to 1000C. In Case 1, the specimens were immersed in water after high-temperature heating and then cooled naturally; in Case 2, the specimens were immersed in water at high temperatures without the cooling process. Based on the results, it was confirmed that lime expansion due to the rehydration of CaO by heating occurred in Case 1. In contrast, dynamic continuous explosive spalling occurred in Case 2 because of water penetration into the specimen at a high temperature. The explosive spalling in water observed in Case 2 is a phenomenon that has not been reported to date. Possible failure mechanisms for lime expansion and continuous expansive spalling in water are suggested.

Numerical analysis of melt penetration behavior in the control rod drive housing of Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; 山路 哲史*; 佐藤 一憲*; 山下 拓哉; 永江 勇二

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05

For Fukushima Daiichi Nuclear Power Station (1F) Unit-2, the muon radiography investigation results indicate that the fuel debris are largely retained inside the RPV. The current study focuses on the analysis of metallic melt penetration behavior in the CRD housing with Moving Particle Semi-implicit (MPS) method. A three-dimensional CRD housing model with simplified inner structures was established. The injection of SS-Zircaloy eutectic melt into the CRD housing was simulated and its downstream penetration and freezing behavior under vertically varying temperature boundary conditions was analyzed. It is found that the melt would start to freeze and form channel blockages soon after it enters the region with a relatively cold boundary in the downstream.

The Results of the CLADS-MADE-03 BWR bundle degradation test in steam under 1F Unit 1 postulated conditions

Pshenichnikov, A.; 永江 勇二

Nuclear Engineering and Design, 415, p.112729_1 - 112729_16, 2023/12

https://doi.org/10.1016/j.nucengdes.2023.112729

The paper presents the results of a CLADS-MADE-03 BWR mock-up assembly degradation in high-temperature steam under a transient heating rate of 1 K/s, which are the assumed accident conditions for the beginning of the accident at the Unit 1 of the Fukushima Dai-Ichi Nuclear Power Station. In situ video investigations of the control blade provided visual evidences for the features of the control blade melt relocation. Based on the results, factors like an increase in the amount of core liquefied materials due to change of the melt composition, subsequent change in melt relocation features because of liquid/solid material ratio within the melt and B release and transport by aerosol formation were specially emphasized.

Axial variations of oxide layer growth and hydrogen uptake of BWR fuel claddings under steam starvation conditions

坂本 寛*; Adachi, Mika*; 徳島 二之*; 青見 雅樹*; 柴田 裕樹; 永江 勇二; 倉田 正輝

Zirconium in the Nuclear Industry; 20th International Symposium (ASTM STP 1645), p.411 - 432, 2023/11

https://doi.org/10.1520/STP164520220043

Steam oxidation tests under steam-starved and non-steam-starved conditions were conducted up to 1573 K using prototypic BWR fuel assembly (four fuel pins and fuel channel box) with approximately 750 mm length. Significant suppression of oxide layer growth and enhancement of hydrogen uptake were found at the downstream positions under the steam-starved conditions. To understand the results obtained in the tests using the prototypic BWR fuel assembly, three separate-effects tests were conducted to obtain a fundamental understanding of mechanism of oxygen and hydrogen uptake and its axial variations and evaluation of hydrogen solubility in oxygen-dissolved Zircaloy-2. It is retrieved that the fuel channel box contributes to the axial variations of oxide layer growth and hydrogen uptake of the fuel pins by acting as a source of hydrogen and a sink of oxygen. The evaluation of hydrogen uptake and release requires a detailed estimation of steam oxidation with time at each elevation.

Numerical simulation method using a Cartesian grid for oxidation of core materials under steam-starved conditions

山下 晋; 佐藤 拓未; 永江 勇二; 倉田 正輝; 吉田 啓之

Journal of Nuclear Science and Technology, 60(9), p.1029 - 1045, 2023/09

https://doi.org/10.1080/00223131.2022.2162996

We newly developed a detailed simulation method for the oxide layer growth/recession under steam-starved conditions using computational fluid dynamics (CFD) methodologies to elaborate the understanding of failure conditions of fuel assemblies during severe accidents. The new method uses the concept of the distance function in a Cartesian grid and is implemented in the original multiphase/multicomponent CFD code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). A distance calculation of the normal direction from the interface is generally difficult in a Cartesian grid. However, the distance function can give a distance normal to the surface of materials by referring to the value of the function. Thus, the growth/recession calculations, which require the distance normal to the interface, become very easy. We checked the availability of JUPITER, considering these models against the verification and validation problems. As a result, we confirmed that JUPITER gives good results, which may contribute to understanding the progress of core degradation under steam-starved conditions.

事故時の原子炉圧力容器及び炉内構造物の解析評価に用いる強度特性データ集

下村 健太; 山下 拓哉; 永江 勇二

JAEA-Data/Code 2022-012, 270 Pages, 2023/03

JAEA-Data-Code-2022-012.pdf:38.25MB

発電用原子炉である軽水炉において、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と同様な全交流電源喪失が発生した場合には、原子炉圧力容器(RPV: Reactor Pressure Vessel)内の冷却機能の喪失、炉内の水位低下による燃料棒の露出、炉心溶融に伴うRPVの破損やRPV破損に伴う炉内の放射線物質の漏えいが発生することが考えられる。事故進展におけるRPVの損傷、溶融した燃料デブリの流出・拡大等の過程を検証、推定することは、廃炉作業を進める上で重要な情報となる。RPVの破損要因については、RPV下部構造部に加えられる荷重・拘束に起因する破損(力学的破損)、低融点金属や高融点酸化物とRPV底部の構造部材との共晶現象による破損(材料間反応による破損)、RPV底部の構造部材の融点近傍での破損が考えられる。破損要因の内、力学的破損については、数値解析(熱流動解析及び構造解析)により検証を行う。このような数値解析を実施する際には、RPV及び炉内構造物を構成する材料(ジルコニウム、炭化ホウ素、ステンレス鋼、ニッケル合金、低合金鋼等)の融点近傍までの伝熱特性(熱伝導率、比熱、密度)や材料特性(熱膨張係数、ヤング率、ポアソン比、引張、クリープ)が必要となる。本資料においては、公開文献を基に、RPV及び炉内構造物を構成する各材料の融点近傍までの母材の特性データをデータ集として取りまとめた。なお、RPV及び炉内構造物を構成する構造物の中には溶接部も存在するため、今回限られたデータであるが、溶接部に関する特性データも併せて示した。

X ray spectroscopy on atoms (J-PARC E03, E07 and future)

山本 剛史; 藤田 真奈美; 後神 利志*; 原田 健志*; 早川 修平*; 細見 健二; 市川 裕大; 石川 勇二*; 鎌田 健人*; 叶内 萌香*; et al.

EPJ Web of Conferences, 271, p.03001_1 - 03001_5, 2022/11

https://doi.org/10.1051/epjconf/202227103001

X-ray spectroscopy of hadronic atoms give us various information on the strong interaction between hadrons and nuclei. At J-PARC, we are aiming for the world-first detection of the X rays from atoms with a doubly strange hyperon, . Recently, two experiments, J-PARC E07 and J-PARC E03, have been performed for the detection of X rays from atoms. Overview and status of these measurements will be presented. We will also discuss future plan of X-ray spectroscopy on -atoms, which can be performed together with high resolution hyper nuclear spectroscopy using newly constructed S-2S spectrometer. Preparation status will be shown in this contribution.

An Investigation of the microstructure and phase composition of the Zr bearing metallic debris in a bypass channel of a BWR fuel after the exothermic reaction in the CLADS-MADE-04 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

CLADS-MADE-04は、下部コア領域での溶融伝播挙動の理解を目的としたシリーズの次のテストである。この寄稿では、電子プローブ マイクロアナライザー(EPMA)によって調査された金属破片の微細構造を含む試験後分析の最近の結果について説明する。テスト中、制御棒ブレードの溶融は、比較的ゆっくりと(数cm/分)急激な強い熱放出の波で発生し、最も高温の領域から、劣化しつつある制御棒ブレードとチャネルボックスに沿って下方に広がり、ジルカロイ-4で作られた壁を消費した。サンプル支持板にも大きな損傷が発生した。このような金属リッチ破片の微細構造の調査により、強化された局所コア劣化のメカニズムを理解できるようになる。EPMAによる相同定を徹底した上で、放熱性が高く周囲への拡散の可能性があることを確認する必要がある。Fe-B共晶デブリとZr-Fe共晶デブリの違いについて概説する。これは、下部炉心プレートのメルトスルーと、下部プレナムへのZr-Fe溶融材料の進行の可能性を理解するために特に重要である。

BWR lower head penetration failure test focusing on eutectic melting

山下 拓哉; 佐藤 拓未; 間所 寛; 永江 勇二

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

AA2022-0018.pdf:8.64MB

https://doi.org/10.1016/j.anucene.2022.109129

Decommissioning work occasioned by the Fukushima Daiichi Nuclear Power Station (1F) accident of March 2011 is in progress. Severe accident (SA) analysis, testing, and internal investigation are being used to grasp the 1F internal state. A PWR system that refers to the TMI-2 accident is typical for SA codes and testing, on the other hand, a BWR system like 1F is uncommon, understanding the 1F internal state is challenging. The present study conducted the ELSA-1 test, a test that focused on damage from eutectic melting of the liquid metal pool and control rod drive (CRD), to elucidate the lower head (LH) failure mechanism in the 1F accident. The results demonstrated that depending on the condition of the melt pool formed in the lower plenum, a factor of LH boundary failure was due to eutectic melting. In addition, the state related to the CRD structure of 1F unit 2 were estimated.

Raman investigation of the CLADS-MADE-02 test debris to confirm the mechanism of the volatile and non-volatile boron compounds formation

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Proceedings of TopFuel 2021 (Internet), 12 Pages, 2021/10

The results of the several recent tests performed in JAEA/CLADS are outlined in this paper. However, particular point of this work is focused on the interesting effect that was found on the debris, that contained partially reacted BC (control blade debris). A Raman investigation of the control blade metallic debris helped to refine the governing mechanism of the B-compounds formation and transport, which is probably specific mostly for BWRs due to unique bundle configuration and materials morphology. All these factors may directly influence the accident progression in BWR and influence the final debris properties.