Estimation of the in-depth debris status of Fukushima Unit-2 and Unit-3 with Multi-physics modeling (Contract research); FY2021 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Waseda University*

JAEA-Review 2022-054, 150 Pages, 2023/02

JAEA-Review-2022-054.pdf:7.26MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (hereafter referred to "1F"). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Estimation of the in-depth debris status of Fukushima Unit-2 and Unit-3 with multi-physics modeling" conducted from FY2019 to FY2021. Since the final year of this proposal was FY2021, the results for three fiscal years were summarized. Continuous update on understanding of the damaged 1F reactors is important for safe and efficient decommissioning of the reactors. This study aimed to estimate the in-depth debris status of the damaged 1F Unit-2 and Unit-3 through multi-physics modeling, which comprises of MPS method, simulated molten debris relocation experiment and high-temperature melt property data acquisition in the three-year project from FY2019.

3D FEM soil-structure interaction analysis for Kashiwazaki-Kariwa Nuclear Power Plant considering soil separation and sliding

Ichihara, Yoshitaka*; Nakamura, Naohiro*; Moritani, Hiroshi*; Choi, B.; Nishida, Akemi

Frontiers in Built Environment (Internet), 7, p.676408_1 - 676408_14, 2021/06

https://doi.org/10.3389/fbuil.2021.676408

The objective of this study is the improvement of response evaluations of structures, facilities and equipment in evaluation of three-dimensional seismic behavior of nuclear power plant facilities, by three-dimensional finite element method model, including separation and sliding between the soil and the basement walls. To achieve this, simulation analyses of Kashiwazaki Kariwa nuclear power plant unit 7 reactor building under the 2007 Niigataken-chuetsu-oki earthquake event were carried out. These simulation analyses consider soil-structure interaction using a three-dimensional finite element method model in which the soil and building are three-dimensionally modeled by the finite element method. It is found that basemat uplift is generated on east side of the basemat edge, and this has an important influence on the results. The importance is evidenced by the difference of local response in soil pressure characteristics beneath the edge of basemat, the soil pressure characteristics along the east side of basement wall and the maximum acceleration response at the west end of the embedded surface. Although, in this particular study, basemat uplift, separation and sliding have only a relatively small influence on the maximum acceleration response of embedded surface and the soil pressure characteristics along the basement walls and beneath the basemat, under strong earthquake motion, these influences can be significant, therefore appropriate evaluation of this effect should be considered.

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for Unit 2

Sonnenkalb, M.*; Pellegrini, M.*; Herranz, L. E.*; Lind, T.*; Morreale, A. C.*; Kanda, Kenichi*; Tamaki, Hitoshi; Kim, S. I.*; Cousin, F.*; Fernandez Moguel, L.*; et al.

Nuclear Engineering and Design, 369, p.110840_1 - 110840_10, 2020/12

https://doi.org/10.1016/j.nucengdes.2020.110840

This is the second paper in a series of 3 in which results of severe accident analyses for Unit 2 of Fukushima Daiichi are presented, gained in Phase 2 of the OECD/NEA project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)". Nine organizations of six countries submitted results of their calculated severe accident scenarios for Unit 2 of Fukushima Daiichi using different severe accident codes. The present paper describes the findings of the comparison of the participants' results for Unit 2 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on reactor pressure vessel status, melt release and fission product behavior and release. Unit 2 specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized.

Analysis for the accident at unit 2 of the Fukushima Daiichi NPS with THALES2/KICHE code in BSAF2 project

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.100 - 111, 2019/08

JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 2 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. One of focusing points in the BSAF2 project was the trend of measured data of reactor vessel from 20:00 March 14 to 02:00 March 15. An assumption was made that the lower part of the suppression chamber failed to form a water leakage path. The released iodine and cesium within three weeks after the earthquake were predicted to be approximately 3% and 0.1% of the initial inventory, respectively.

Sensitivity analysis of source term in the accident of Fukushima Dai-ichi Nuclear Power Station Unit 1 using THALES2/KICHE

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2018 (ASRAM 2018) (USB Flash Drive), 6 Pages, 2018/10

In the accidents at Fukushima Dai-ichi Nuclear Power Station, Tsunami caused loss of electric power supply and this event led to core melt and failure of Containment vessel. Finally, fission products were released to the environment. Currently, the activities for understanding of accident progressions are carried out based on the measured data during the accident, accident progression analysis using integrated severe accident analysis codes and investigation of inside of reactor buildings and containment vessels. On the other hand, there are some research activities with combination of accident progression analysis and accident consequence analysis. In Japan Atomic Energy Agency (JAEA), the research project of combination of these analyses using the computational simulation codes has been started. The results obtained from the combination analysis are expected to have broad width of uncertainty because of many uncertainty factors in this combined analysis. In order to perform the analysis efficiently, sensitivity analysis for failure location on containment vessel and its failure size were carried out by THALES2/KICHE developed by JAEA at first. This analysis was performed on unit 1, since it was the first plant to release radioactive materials to the environment during the accident and its consequence had no effect from other plants. The authors focused on the failure of containment vessel head flange, penetration seal and vacuum breaker pipe, and possibility of partial open of vent valve based on the investigations of reactor building inside performed by TEPCO. This paper presents the results obtained from this sensitivity analysis.

Report of Examination of the Sample from Core Shroud (2F2-H3) at Fukushima Dai-ni Power Station Unit-2 (Contract research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi; Nakajima, Hajime*; Shibata, Katsuyuki; Tsukada, Takashi; Suzuki, Masahide; Kiuchi, Kiyoshi; Kaji, Yoshiyuki; Kikuchi, Masahiko; Ueno, Fumiyoshi; Nakano, Junichi; et al.

JAERI-Tech 2004-015, 114 Pages, 2004/03

JAERI-Tech-2004-015.pdf:38.06MB

The Tokyo Electric Power Company (TEPCO) visually inspected the weld joint of core shroud at Fukushima Dai-ni Nuclear Power Station Unit-2 by a direction of the Nuclear and Industrial Agency, cracks were observed at outer side of the ring weld joint (H3) between a core shroud middle trunk and a middle ring. TEPCO has conducted a material examination with Nippon Nuclear Fuel Development Co. Ltd. (NFD) on the specimen including cracks sampled from the core shroud. The present examination has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage. Based on results of the present examination, the probable presence of tensile residual stress by welding process and dissolved oxygen contents in the cooling water, it was shown that the cracks were considered to be stress corrosion cracking (SCC). However, the cause of the cracks needs more consideration on the way of shroud construction.

Criticality control of fuel debris / TMI-2 review and Fukushima expectation

Tonoike, Kotaro; Yamane, Yuichi

no journal, , 

It is conceivable that a large amount of fuel debris has been produced in the reactors of Fukushima Daiichi Nuclear Power Station (1FNPS) due to the severe core damage and melting. The amount is far beyond minimum critical mass, and essentially subject to criticality control. This presentation will outline the criticality control of fuel debris in the Three Mile Island Nuclear Station, Unit 2 reactor (TMI-2). Then, criticality control problems of fuel debris during the 1FNPS decommissioning will be described contrasting with the TMI-2 case. Finally, technical expectations and research activities of the fuel debris criticality control for the 1FNPS will be presented. It is possible that the fuel debris will be retrieved under nonborated water at the risk of criticality. In this scenario, the risk control, mitigation measures in case of criticality, would be a key factor in safety of the decommissioning.