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.

Thermodynamic analysis for solidification path of simulated ex-vessel corium

佐藤 拓未; 永江 勇二; 倉田 正輝; Quaini, A.*; Guneau, C.*

CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 79, p.102481_1 - 102481_11, 2022/12

https://doi.org/10.1016/j.calphad.2022.102481

Investigation of the primary containment vessel inside the Fukushima Daiichi Nuclear Power Station showed that a significant amount of the molten corium reached the bottom of the pedestal region. The molten corium and concrete likely caused a complex interaction called Molten Corium Concrete Interaction. The solidification hysteresis of these ex-vessel debris significantly influences its properties. We performed a thermodynamic analysis using the TAF-ID database to infer the solidification path of U-Zr-Al-Ca-Si-O molten corium, which was chosen for a prototypic system of ex-vessel debris. The solidification path for the CaO-rich sim-corium showed that (i) melting as a single liquid phase above 2430 K, (ii) selective solidification of the oxide-rich corium mainly consisted of fuel materials, and (iii) solidification of the remaining materials as a silicate matrix. In contrast, the solidification path for the SiO-rich corium indicated that (i) formation of liquid miscibility gap above 2200 K between U-rich and Zr-rich oxidic melts, (ii) individual precipitation of solid phases in each liquid phase.

Step-by-step challenge of debris characterization for the decommissioning of Fukushima-Daiichi Nuclear Power Station (FDNPS)

倉田 正輝; 奥住 直明*; 仲吉 彬; 池内 宏知; 小山 真一

Journal of Nuclear Science and Technology, 59(7), p.807 - 834, 2022/07

https://doi.org/10.1080/00223131.2022.2040393

東京電力福島第一原子力発電所(1F)の廃炉に向けて、事故直後から、燃料デブリの特性評価について様々な試みが行われてきた。本レビューでは、それらを概説する。事故直後の数年間は、1F現場(特に損傷した1, 2, 3号機の建屋内部)から得られる知見は極めて限定的であった。燃料デブリのおよその所在はミューオントモグラフィーで調査され、その特性は、加圧水型軽水炉である米国スリーマイル原発事故調査の結果等の過去知見に基づいて概略推定された。その後、各種の内部調査ロボットが開発され、2017年より、原子炉格納容器内部の調査が開始された。その結果、3つの号機とも、当初、加圧水型軽水炉の典型的事故シナリオに基づいて予想されたものと異なる炉心破損状態とデブリ堆積状態であることがわかってきた。また、調査ロボットの付着物から微量のウラン含有粒子が回収され、燃料デブリにつながる情報を得るために、その特性分析が継続している。これらと並行して、OECD/NEAなどにおいて、事故シナリオの検討等について国際協力も進んでいる。今後は、典型事故条件に基づくイメージではなく、現場で得られた知見に基づいて、1F事故シナリオの理解を深め、1F燃料デブリの特性を評価していくことが重要である。

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.

Steam oxidation of silicon carbide at high temperatures for the application as accident tolerant fuel cladding, an overview

Pham, V. H.; 倉田 正輝; Steinbrueck, M.*

Thermo (Internet), 1(2), p.151 - 167, 2021/09

https://doi.org/10.3390/thermo1020011

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Features of a BWR neutron absorber melt relocation in an oxidative environment during the CLADS-MADE-02 test

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 7 Pages, 2021/08

https://doi.org/10.1115/ICONE28-65129

The work on the understanding of the accident progression at the Units of the Fukushima Dai-Ichi Nuclear Power Plant (1F) is ongoing. This contribution gives a part of detailed investigations of the control blade melt propagation downwards through the prototypic BWR bundle assembly during the CLADS-MADE-02 test, where the conditions of the 1F Unit 3 was simulated. Interesting features emerged in an oxidative environment.

On the degradation progression of a BWR control blade under high-temperature steam-starved conditions

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Mechanical Engineering Journal (Internet), 7(3), p.19-00503_1 - 19-00503_10, 2020/06

https://doi.org/10.1299/mej.19-00503

High-temperature control blade degradation tests simulating a beginning phase of a severe accident in BWRs has been comprehensively performed in Japan Atomic Energy Agency (JAEA). In the latest test, a mock-up of BWR bundle material has been investigated under postulated Fukushima Dai-Ichi (1F) Unit 2 accident conditions in a complex heating transient scenario including a phase of lack of available steam. The progress in control blade degradation was monitored with help of an in situ video and the detailed analysis of the solidified metallic melt, so-called metallic debris, was carried out by conventional SEM and XRD methods. These results indicated that the composition of the metallic debris at different elevations has been significantly changed due to the redistribution and relocation of steel elements under the influence of B and C, sometimes accompanied by a formation of high-melting-point layers. The results of this paper significantly contribute to the physical understanding of control blade degradation phenomenology during beginning phase of a core degradation for a special case of steam-starved conditions at 1F Unit 2.

New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1; Introduction of boiling water reactor mock-up assembly degradation test programme

Pshenichnikov, A.; 倉田 正輝; Bottomley, D.; 佐藤 一憲; 永江 勇二; 山崎 宰春

Journal of Nuclear Science and Technology, 57(4), p.370 - 379, 2020/04

https://doi.org/10.1080/00223131.2019.1691070

The new research and development programme of JAEA/CLADS tests complement the previous investigations related to BWR severe accidents. A series of tests aiming at closing the gaps in understanding of the Fukushima Daiichi degradation sequence at each unit. The paper emphasises the problem of control blade degradation, which influences the accident progression at an early stage and shows the approach for thorough investigation of this problem.

Oxidation kinetics of silicon carbide in steam at temperature range of 1400 to 1800C studied by laser heating

Pham, V. H.; 永江 勇二; 倉田 正輝; Bottomley, D.; 古本 健一郎*

Journal of Nuclear Materials, 529, p.151939_1 - 151939_8, 2020/02

AA2019-0197.pdf:1.61MB

https://doi.org/10.1016/j.jnucmat.2019.151939

As expected for accident tolerant fuels, investigation of steam oxidation for silicon carbide under the conditions beyond design basis accident scenarios is needed. Many studies focused on steam oxidation of SiC at temperatures up 1600C have been conducted and reported in the literature. However, behavior of SiC in steam at temperatures above 1600C still remains unclear. To complete this task, we have designed and manufactured a laser heating facility for steam oxidation at extreme temperatures. With the facility, we report the first results on the steam oxidation behavior of SiC at temperatures range of 1400-1800C for short term exposure of 1-7 h under atmospheric pressure. Based on the mass change of SiC samples, parabolic oxidation rate and linear volatilization rate were calculated. The oxidation layer appears to be maintained at 1800C in steam, but the bubble formation phenomenon suggests other volatilization reactions may limit its life.

Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; 永江 勇二; 倉田 正輝; 古本 健一郎*; 佐藤 寿樹*; 石橋 良*; 山下 真一郎

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Silicon carbide (SiC) has recently attracted much attention as a potential material for accident tolerant fuel cladding. To investigate the performance of SiC in severe accident conditions, study of steam oxidation at high temperatures is necessary. However, the study focusing on steam oxidation of SiC at temperatures above 1600C is still certainly limited due to lack of test facilities. With the extreme oxidation/corrosion environment in steam at high temperatures, current refractory materials such as alumina and zirconia would not survive during the tests. Application of laser heating technique could be a great solution for this problem. Using laser heating technique, we can localize the heat and focus them on the test sample only. In this study, we developed a laser heating facility to investigate high-temperature oxidation of SiC in steam at temperature range of 1400-1800C for 1-7 h. The oxidation kinetics is then being studied based on the weight gain and observation on cross-sectioned surface of tested sample using field emission scanning electron microscope. Off-gas measurement of hydrogen (H) and carbon monoxide (CO) generated during the test is also being conducted via a sensor gas chromatography. Current results showed that the SiC sample experienced a mass loss process which obeyed paralinear laws. Parabolic oxidation rate constant and linear volatilization rate constant of the process were calculated from the mass change of the samples. The apparent activation energy of the parabolic oxidation process was calculated to be 85 kJ.mol. The data of the study also indicated that the mass change of SiC under the investigated conditions reached to its steady stage where hydrogen generation became stable. Above 1800C, a unique bubble formation on sample surface was recorded.