A BWR control blade degradation observed in situ during a CLADS-MADE-02 test under Fukushima Dai-Ichi Unit 3 postulated conditions

References

• Gauntt RO, Gasser RD. Results of the DF-4 BWR control blade-channel box test. Albuquerque, New Mexico, USA: Sandia National Laboratories; 1990. (SAND90-2716C).
   Google Scholar
• Gauntt RO, Humphries LL. Final results of the XR2-1 BWR metallic melt relocation experiment. Albuquerque, New Mexico, USA: Sandia National Laboratories; 1997. (NUREG/CR-6527, SAND97-1039).
   Google Scholar
• Schanz G, Hagen S, Hofmann P, et al. Information on the evolution of severe LWR fuel element damage obtained in the CORA program. J Nucl Mater. 1992;188:131–145.
   Web of Science ®Google Scholar
• Hagen S, Hofmann P, Noack V, et al. Dry core BWR test CORA-33: test results. Karlsruhe, Germany: Kernforschungszentrum Karlsruhe; 1994. (KfK 5261).
   Google Scholar
• Sepold L, Hagen S, Hofmann P, et al. Behavior of BWR-type fuel elements with B₄C/steel absorber tested under severe fuel damage conditions in the CORA facility. Karlsruhe, Germany: Forschungszentrum Karlsruhe; 2009. (FZKA 7447).
   Google Scholar
• Tokyo Electric Power Company Holdings, Inc. 7.19Unit 3 PCV internal investigation (preliminary report of July 19 investigation) [Internet]. 2017. [cited 2021 Jan 3]. Available from: https://www.tepco.co.jp/en/news/library/archive-e.html?video_uuid=qf64ne97&catid=61785
   Google Scholar
• International Research Institute for Nuclear Decommissioning (IRID). Development of technology for fuel debris analysis and characterization results for FY2018 [internet]. Tokyo, Japan. 2019. cited 2021 Jan 3. Available from: https://irid.or.jp/_pdf/Development%20of%20technology%20for%20fuel%20debris%20characterization%20and%20analysis_FY2018_rev1.pdf
   Google Scholar
• Steinbrück M. Degradation and oxidation of B4C control rod segments at high temperatures. J Nucl Mater. 2010;400(2):138–150.
   Web of Science ®Google Scholar
• Kurata M, Shibata H, Sakamoto K, et al. Experimental study on control blade degradation and its modeling. In Proceedings of Water Reactor Fuel Performance Meeting; 2014 Sep 14–17; Sendai, Japan. (Paper No. 100079).
   Google Scholar
• Shibata H, Tokushima K, Sakamoto K, et al. Control blade degradation test under temperature gradient in steam atmosphere. In Proceedings of Top Fuel; 2016 Sep 11–15; Boise, ID, USA.
   Google Scholar
• Pshenichnikov A, Kurata M, Bottomley D, et al. 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. J Nucl Sci Technol. 2020;57(4):370–379.
   Web of Science ®Google Scholar
• Pshenichnikov A, Yamazaki S, Bottomley D, et al. Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors. J Nucl Sci Technol. 2019;56(5):440–453.
   Web of Science ®Google Scholar
• Pshenichnikov A, Nagae Y, Kurata M. On the degradation progression of a BWR control blade under high-temperature steam-starved conditions. Mech Eng J. 2020;7(3):1:-10.
   Web of Science ®Google Scholar
• Pshenichnikov A, Kurata M, Nagae Y. Characterization of the Fukushima Dai Ichi unit 2 sediments/debris based on the on site video investigations in comparison to the debris obtained after integral CLADS MADE 01 test. In Proceedings of the 24th Symposium on Power and Energy Systems; 2019 Jun 20–21; Tokyo, Japan. (Paper No. B131).
   Google Scholar
• Pshenichnikov A, Nagae Y, Kurata M. Raman characterization of the simulated control blade debris to understand the boric compounds transformations during severe accidents. Mech Eng J. 2020;7(2):1–10.
   Web of Science ®Google Scholar
• Atomic Energy Society of Japan. The Fukushima Daiichi nuclear accident final report of the AESJ investigation committee. Tokyo, Japan: Springer; 2015.
   Google Scholar
• Institute of Nuclear Power Operations. Special Report on the nuclear accident at the Fukushima Daiichi nuclear power station. Atlanta, Georgia, USA: Institute of nuclear power operations; 2011. (INPO 11-005).
   Google Scholar
• International Atomic Energy Agency. The Fukushima Daiichi accident. Technical volume 1. Description and context of the accident. Vienna, Austria: International Atomic Energy Agency; 2015.
   Google Scholar
• Tokyo Electric Power Company Holdings, Ltd. The 5th progress report on the investigation and examination of unconfirmed and unresolved issues on the development mechanism of the Fukushima Daiichi nuclear accident. Tokyo, Japan: Tokyo Electric Power Company Holdings, Ltd; 2017.
   Google Scholar
• Pshenichnikov A, Kurata M, Nagae Y, et al. The behaviour of materials in case of solidified absorber melt - oxidized BWR channel box interaction revealed after CLADS-MADE-01 test. In Proceedings of the International Topical Workshop on Fukushima Decommissioning Research (FDR2019); 2019 May 24–26; Naraha, Fukushima, Japan. (Paper No. FDR2019-1058).
   Google Scholar
• International Atomic Energy Agency. Corrosion of zirconium alloys in nuclear power plants. Vienna, Austria: International Atomic Energy Agency; 1993. (IAEA TECDOC 684).
   Google Scholar
• Hoffmann PB, De Weerd W, Toscano EH. Post-irradiation examination of high burnup BWR fuel rods: techniques and results. In Proceedings of the Annual German Nuclear Society Conference Jahrestagung Kerntechnik; 2005; Nuremberg.
   Google Scholar
• Pshenichnikov A, Yamazaki S, Nagae Y, et al. Heterogeneity of BWR control blade degradation under steam-starved conditions. In Proceedings of ICONE 27 the 27th International Conference on Nuclear Engineering; 2019 May 19–24; Ibaraki, Japan. (Paper No. ICONE27-1077).
   Google Scholar