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間所 寛; 山下 拓哉; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; 佐藤 一憲; 溝上 伸也*
Nuclear Technology, 209(2), p.144 - 168, 2023/02
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Since the reactor pressure vessel (RPV) lower head failure determines the subsequent ex-vessel accident progression, it is a key issue to understand the accident progression of Fukushima Daiichi Nuclear Power Station (1F). The RPV failure is largely affected by thermal loads on the vessel wall and thus it is inevitable to understand thermal behavior of molten metallic pool with co- existence of solid oxide fuel debris. In the past decades, numerous experiments have been conducted to investigate a homogeneous molten pool behavior. Few experiments, however, addresses the melting and heat transfer process of debris bed consisted of materials with different melting temperatures. LIVE-J2 experiment aimed to provide the experimental data on a solid-liquid mixture pool in a simulated RPV lower head under various conditions. The extensive measurements of the melt temperature indicate the heat transfer regimes in a solid-liquid mixture pool. The test results showed that the conductive heat transfer was dominant during the steady state along the vessel wall boundary and that convective heat transfer takes place inside the mixture pool. Besides the experimental performance, the test case was numerically simulated by using ANSYS Fluent. The simulation results generally agree with the measured experimental data. The flow regime and transient melt evolution were able to be estimated by the calculated velocity field and the crust thickness, respectively.
間所 寛*; 永江 勇二
日本原子力学会誌ATOMO
, 64(9), p.500 - 503, 2022/09
原子炉過酷事故解析などによる福島第一原子力発電所の事故進展解析では、特に原子炉圧力容器(RPV)下部ヘッドの破損から燃料デブリのペデスタル底部への流出に関して不確かさが大きく、モデル高度化が喫緊の課題となっている。また、2号機の内部調査の結果、ペデスタル下部構造物に目立った損傷が見られないことから、RPV下部ヘッドから流出した燃料デブリは溶融金属主体で多くの燃料酸化物は未溶融であった可能性が指摘されている。そこで、廃炉環境国際共同研究センターでは、固液混合状態での圧力容器破損挙動の理解に取り組んできた。本報では、(1)下部ヘッド溶融プールの熱的挙動を把握するための試験(LIVE試験)と、(2)溶融金属による圧力容器下部ヘッド破損挙動を把握するための試験(ELSA試験)の概要を報告する。
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
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)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.
山下 拓哉; 間所 寛; 佐藤 一憲
Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04
Understanding the final distribution of core materials and their characteristics is important for decommissioning the Fukushima Daiichi Nuclear Power Station (1F). Such characteristics depend on the accident progression in each unit. However, boiling water reactor accident progression involves great uncertainty. This uncertainty, which was clarified by MAAP-MELCOR Crosswalk, cannot be resolved with existing knowledge and was thus addressed in this work through core material melting and relocation (CMMR) tests. For the test bundle, ZrO
pellets were installed instead of UO pellets. A plasma heating system was used for the tests. In the CMMR-4 test, useful information was obtained on the core state just before slumping. The presence of macroscopic gas permeability of the core approaching ceramic fuel melting was confirmed, and the fuel columns remained standing, suggesting that the collapse of fuel columns, which is likely in the reactor condition, would not allow effective relocation of the hottest fuel away from the bottom of the core. This information will help us comprehend core degradation in boiling water reactors, similar to those in 1F. In addition, useful information on abrasive water suspension jet (AWSJ) cutting for debris-containing boride was obtained in the process of dismantling the test bundle. When the mixing debris that contains oxide, metal, and boride material is cut, AWSJ may be repelled by the boride in the debris, which may cut unexpected parts, thus generating a large amount of waste in cutting the boride part in the targeted debris. This information will help the decommissioning of 1F.
佐藤 一憲; 山路 哲史*; Li, X.*; 間所 寛
Mechanical Engineering Journal (Internet), 9(2), p.21-00436_1 - 21-00436_17, 2022/04
Interpretation for the two-week long Unit 3 ex-vessel debris cooling behavior was conducted based on the Fukushima-Daiichi Nuclear Power Plant (1F) data and the site data such as pressure, temperature, gamma ray level and live camera pictures. It was estimated that the debris relocated to the pedestal was in partial contact with liquid water for about initial two days. With the reduction of the sea water injection flowrate, the debris, existed mainly in the pedestal region, became "dry", in which the debris was only weakly cooled by vapor and this condition lasted for about four days until the increase of the sea water injection. During this dry period, the pedestal debris was heated up and it took further days to re-flood the heated up debris.
間所 寛; 山下 拓哉; 佐藤 一憲; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; St
ngle, R.*; Wenz, T.*; Vervoortz, M.*; 溝上 伸也
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Debris and molten pool behavior in the reactor pressure vessel (RPV) lower plenum is a key factor to determine its failure mode, which affects the initial condition of ex-vessel accident progression and the debris characteristics. These are necessary information to accomplish safe decommissioning of the Fukushima Daiichi Nuclear Power Station. After dryout of the solidified debris in the lower plenum, metallic debris is expected to melt prior to the oxide debris due to its lower melting temperature. The lower head failure is likely be originated by the local thermal load attack of a melting debris bed. Numerous experiments have been conducted in the past decades to investigate the homogeneous molten pool behavior with external cooling. However, few experiments address the transient heat transfer of solid-liquid mixture without external cooling. In order to enrich the experimental database of melting and heat transfer process of debris bed consisted of materials with different melting temperatures, LIVE-J1 experiment was conducted using ceramic and nitrate particles as high melting and low melting temperature simulant materials, respectively. The test results showed that debris height decreased gradually as the nitrate particles melt, and molten zone and thermal load on vessel wall were shifted from bottom upwards. Both conductive and convective heat transfer could take place in a solid-liquid mixture pool. These results can support the information from the internal investigations of the primary containment vessel and deepen the understanding of the accident progression.
間所 寛; 佐藤 一憲
Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05
被引用回数：6 パーセンタイル：69.63(Nuclear Science & Technology)Estimation of the final debris distribution at the Fukushima Daiichi Nuclear Power Plant (1F) is inevitable for a safe and effective decommissioning. It is necessary to clarify possible failure modes of the reactor pressure vessel (RPV), which is influenced by the thermal status of slumped debris that highly depends on the in-vessel accident progression. The accident analysis of 1F Unit 2 (1F2) was conducted using the RELAP/SCDAPSIM code. One of the unsolved issues of 1F2 is the mechanism of three pressure peaks measured through late Mar. 14 to early March 15, 2011. Comparing the results of previous boiling water reactor (BWR) core degradation experiments and that of 1F2 numerical analysis, it can be estimated that most relocated metallic materials had solidified at the core bottom at the onset of first pressure peak. It is likely that the pressure increase occurred due to the evaporation of injected water reaching the heated core plate structures. Between the first and second pressure peaks, the water is assumed to have been injected continuously and the water level was likely to have recovered to BAF at the initiation of the second pressure peak. Probable slumping of a certain amount of molten materials initiated the second pressure peak and the subsequent gradual pressure increase continued possibly due to massive reaction between coolant and remaining Zircaloy in the core. Assuming the closure of the safety relief valve (SRV) at 0:00 on Mar. 15, the third pressure peak was well reproduced in the analysis.
Li, X.*; 山路 哲史*; Duan, G.*; 古谷 正祐*; 深井 尋史*; 佐藤 一憲; 間所 寛; 大石 佑治*
no journal, ,
The Moving Particle Semi-implicit (MPS) method is being developed for simulation of multi-component liquid/solid relocation with solid-liquid phase changes. Main model developments and validation of the developed code against the simulated spreading and ablation experiments are summarized in the current paper.
間所 寛; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; Stngle, R.*; Wenz, T.*; Vervoortz, M.*; 山下 拓哉; 佐藤 一憲; 溝上 伸也
no journal, ,
福島第一原子力発電所(1F)2号機では、ペデスタル内部の構造物が比較的健全であることから、原子炉圧力容器(RPV)からペデスタル内部に移行した燃料デブリは比較的低温であり、下部ヘッド内部では、燃料デブリ中の酸化物成分が溶融する温度に至っておらず、金属成分が中心に溶融していたと推定される。RPV破損を推定するには下部ヘッド溶融プールの熱的挙動の把握が必要であるが、固液混合溶融プールに着目した試験は少なく、実験データの拡充が不可欠となっている。本研究では、独・カールスルーエ工科大学におけるLIVE試験装置を用いて、溶融プール形成過程及び伝熱挙動に関する試験を実施した。固液混合状態においても対流がある程度発達し、RPV側部に最も熱的負荷がかかることが分かった。
間所 寛; 永江 勇二
no journal, ,
東京電力福島第一原子力発電所(1F)事故から11年が経過し、内部調査や数値解析によって事故時の各号機の挙動が徐々に明らかになってきている。しかしながら、OECD/NEAの枠組みで実施された原子炉過酷事故解析コード(SAコード)による1F事故進展解析ベンチマークプロジェクトBSAFでは、解析コードやユーザーによる不確かさが大きい結果となり、SAコードの精緻化が必要である。特に、原子炉圧力容器(RPV)下部ヘッドの破損から燃料デブリのペデスタル底部への流出に関しては、不確かさが大きく、モデルの高度化は喫緊の課題となっている。1F2号機の格納容器内部調査結果によると、ペデスタル下部構造物に目立った損傷が見られないことから、RPV下部ヘッドから流出した燃料デブリは溶融金属主体であり、燃料酸化物の多くが未溶融であった可能性が指摘されている。下部ヘッド内部に堆積した燃料デブリのうち、酸化物成分は固体のまま、主に融点の低い金属成分が溶融し、金属主体の固液混合溶融プールが形成され、材料間の反応によって下部ヘッドが破損した可能性がある。本報では、下部ヘッド溶融プールの熱的挙動に着目したLIVE試験及び、下部ヘッド構造材と燃料デブリの材料間反応による下部ヘッド破損挙動に着目したELSA試験について報告する。
倉田 正輝; 間所 寛; 奥村 啓介; 佐藤 一憲; 溝上 暢人*; 伊東 賢一*; 溝上 伸也*
no journal, ,
福島第一原子力発電所(1F)からの燃料デブリ取出し工程の設計や取出し作業の安全な進捗に向けて、燃料デブリやその他の堆積物や破損物等の分析にニーズを有している課題を抽出し、課題解決に必要となる特性や事象に分解しとりまとめた。抽出した特性や事象について、それらの評価のために必要となる分析手法と取得できる知見について検討を行った。さらに、サンプル分析で得られる限定的なデータを用いて、燃料デブリの広い領域の評価につなげられるかについて予備的に検討した。
佐藤 一憲; 山路 哲史*; 古谷 正祐*; 大石 佑治*; Li, X.; 間所 寛; 深井 尋史*
no journal, ,
CLADS英知事業において福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定に取り組んでおり、その一環として2・3号機における燃料デブリ状態に関するいくつかの重要な論点を整理している。すなわち、3号機でのペデスタル移行デブリによるMCCIの程度、あるいは2号機におけるペデスタル移行デブリの性状を考えると、プラント内部調査などの個々の要素間には現状知見による統一的な理解が困難な要素がいくつかある。これらは事故進展解明に係る中長期的課題であり、本英知事業はその解明に向けた一部と位置付けられる。本報告ではこれらの論点について報告する。
佐藤 拓未; 山下 拓哉; 間所 寛; 永江 勇二
no journal, ,
本研究では、沸騰水型軽水炉の圧力容器下部構造物である制御棒駆動機構の金属系デブリとの反応・溶融試験を実施し、その破損挙動を観察した。その結果、構造材/金属デブリの反応により、単体での融点よりも低い低温でCRD構造物の破損が進行することが明らかになった。
佐藤 一憲; 山路 哲史*; 古谷 正祐*; 大石 佑治*; Li, X.*; 間所 寛; 深井 尋史*
no journal, ,
3号機ペデスタル移行デブリの熱が、ペデスタル液相水の蒸発とそれにより発生した蒸気の圧力抑制室(S/C)での凝縮によってS/C水に伝えられていた可能性をドライウェル(D/W)とS/Cの圧力履歴をもとに評価した。この結果、デブリは強く冷却されていたと推定された。
Li, X.*; 山路 哲史*; 佐藤 一憲; 古谷 正祐*; 間所 寛; 大石 佑治*
no journal, ,
The Moving Particle Semi-implicit (MPS) method is developed to simulate debris relocation and interaction with pedestal structures in Fukushima Daiichi (1F) Unit-3. Different debris distributions and structure damages are evaluated with different debris relocation amount / intervals and convective vapor cooling from the debris surface.
佐藤 一憲; 山路 哲史*; 古谷 正祐*; 大石 佑治*; Li, X.*; 間所 寛
no journal, ,
事故時のプラントデータ分析、内部調査結果の分析、及び多様な解析評価に基づき、2号機,3号機における燃料デブリのペデスタルへの移行履歴を検討した。
間所 寛; 山下 拓哉; 佐藤 一憲
no journal, ,
The test bundle of the latest test CMMR-4, Core-Material Melting and Relocation experiment, consists of 48 fuel rods filled with ZrO simulant pellets with Zircaloy claddings, a control blade with B
C particles in SS tube and sheath, two Zircaloy channel box walls, and lower support structures. The height of the test bundle was 80 cm and the heating system of the test was the plasma heating, which enabled melting of the oxide simulant fuel pellets. The test confirmed that macroscopic gas permeability existed until the ceramic-fuel melted and that the hot fuel rods tended to remain as columns in the core region, which suggests the heating of the support structure in earlier phase is unlikely. This information is useful not only for 1F decommissioning but also for further understanding of a BWR severe accident progression. The test bundle was cut by using the abrasive waterjet (AWJ) technique that uses abrasive garnet of 150-300 micro m with feed rate of approximately 1.5 kg/min. In order to cut off about 30 mm of ZrB spot contained in the relocated melts, 750 liters of water, 84 kg of garnet and one nozzle replacement were necessary. The EPMA and XRD analyses of the cross-section showed that the place where repelled the garnet-contained waterjet contained ZrB. Since the cutting by AWJ technique has the property of selectively abrading the soft spots of the material, it must be noted that, in case of utilizing the technique in 1F decommissioning, garnet might be repelled by a hard boride and abrades places which were not expected.
佐藤 一憲; 山路 哲史*; 古谷 正祐*; 大石 佑治*; Li, X.*; 間所 寛; 深井 尋史*
no journal, ,
福島第一原子力発電所2号機におけるRPVバウンダリー破損モードについて、これまでの解析評価や内部調査の結果などに基づいて検討し、考えられる3つのRPVバウンダリー破損モードを提示した。
山路 哲史*; 古谷 正祐*; 大石 佑治*; 佐藤 一憲; Li, X.*; 深井 尋史*; 間所 寛
no journal, ,
MPS法による溶融物挙動解析、模擬溶融物流下実験、浮遊法による高温融体物性評価と、実機プラントデータ・事故進展解析等の分析から、福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定に取り組んでいる。一年目は基盤技術の整備等に取り組んだ。
