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.

Study on chemical interaction between UO and Zr at precisely controlled high temperatures

白数 訓子; 佐藤 拓未; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06

https://doi.org/10.1080/00223131.2022.2138602

ジルカロイ被覆管とUO燃料の溶融反応のメカニズム解明に資するため、温度誤差が可能な限り最小となるよう検討を行い、1840Cから2000Cの範囲でZrとUOの高温反応試験を実施した。UOるつぼにZr試料を装荷し、アルゴン雰囲気中加熱を行い、生成した反応相の成長状況や溶融状態、組織変化の観察を行った。1890 Cから1930 Cで加熱した試料は、丸く変形しており、-Zr(O)相と、少量のU-Zr-O溶体相で形成されていた。1940C以上で加熱した試料は大きく変形し、急激に溶体形成反応が進行する様子が観測された。U-Zr-O溶体相の形成反応はZr(O)中の酸素濃度に依存し、酸素濃度の低いZr(O)へ反応はどんどん進展する。そして酸素含有量が高いZr(O)中では、U-Zr-O溶体相の生成が抑制されることが確認された。

Thermodynamic evaluation on solidification path for U-Zr-Fe-O corium

多木 寛; 永江 勇二; 倉田 正輝

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

The analysis of small samples retrieved from the inside of Primary Containment Vessel (PCV) of Units 1, 2, and 3 at Fukushima Daiichi Nuclear Power Station detected various types of U-bearing particles. Elucidation of the formation mechanism of these particles is expected to be good practice for real debris characterization. In this study, we attempted to analyze the solidification path of U-bearing particles by the thermodynamic approach. From the thermodynamic solidification path patterns analysis, pattern II (at low oxidation condition) was identified as the available one for Units 1 & 2 particles, whereas pattern IV (at high oxidation condition) would be additionally possibly for Unit 3. From these thermodynamic analyzes, the following characteristic are speculated for the debris in PCV: 1) The debris are likely to have been solidified by gradual cooling from high temperatures to intermediate temperatures and solid-solid transition at lower temperatures may be limited. 2) Units 1 & 2 debris might be exposed to slightly hypo-stoichiometric conditions than Unit 3, and whereas Unit 3 debris might have a wider variation in the oxidation degree.

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.