NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE

Alexandre Ezzidi Nakata, Masanori Naitoh, Chris Allison

DOI: http://dx.doi.org/10.17146/tdm.2019.21.3.5630

Abstract


Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.

Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation


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References


  1. J. R. Wolf, et al., “TMI-2 Vessel Investigation Project Integration Report”, NUREG/CR-6197, March 1994.
  2. Gauntt, R.O., Cash, J.E., Cole, R.K. et al., MELCOR Computer Code Manuals, Sandia National Laboratories.
  3. EPRI, Modular Accident Analysis Program 5 (MAAP5) Applications Guidance, 2015.
  4. M. Naitoh et al., “Development of severe accident analysis code SAMPSON in IMPACT project”, J. Nucl. Sci. Technol., 36, 11, 1999, 1076.
  5. SCDAP/RELAP5, NUREG/CR-6150, Vol. 1, Rev. 2 INEL-96/0422.
  6. World Nuclear Association, “Fukushima Accident 2011,” September 2014, http://www.world-nuclear.org/info/Safety-and-Security/Safety-of-Plants/Fukushima-Accident-2011. Radioactivity units converted from 940 petabecquerels of Iodine-131 equivalent.
  7. OECD/NEA, “Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF Project)”, Nuclear Regulation, NEA/CSNI/R(2015)18, February 2016.
  8. Marco SANGIORGI, Alexandre Ezzidi et al.,” In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 NPP”, European Commission, JRC Technical Reports, 2016.
  9. A. Ezzidi Nakata et al., “Sampson Analysis of PWR Induced Steam Generator Tube Rupture”, NTHAS11: The Eleventh Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety, Busan, Korea, Paper No. N11O0173, November 18-21, 2018.
  10. Takashi Ikeda et al., “Analysis of International Standard Problem No. 45, QUENCH06 Test at FZK by Detailed Severe Accident Code, IMPACT/SAMPSON”, Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 40, No. 4, p. 246-255 (April 2003).
  11. Takashi Ikeda et al., “Analysis of Core Degradation and Fission Products Release in Phebus FPT1 Test at IRSN by Detailed Severe Accident Analysis Code, IMPACT/SAMPSON”, Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 40, No. 8, p. 591-603 (August 2003).
  12. C.M. Allison, R.J. Wagner, L.J. Siefken, J.K. Hohorst, “The Development of RELAP/SCDAPSIM/MOD4.0 for Reactor System Analysis and Simulation”, Proceedings of the 7th International Conference on Nuclear Option in countries with Small and Medium Electricity Grids, 25-29 May 2008, Dubrovnik (Croatia), Paper No. S-05.01.
  13. US-NRC RELAP5/MOD3.3 Code Manual, Volumes I to VIII. NUREG/CR-5535/Rev 1 2001.
  14. C.M. Allison, J.K. Hohorst, A. Ezzidi, and M. Naitoh, “Preliminary Assessment of the new ASYST - ISA Integral Analysis Code Using Selected Integral Thermal Hydraulic - Fuel Assembly Experiments”, 25th International QUENCH-Workshop, KIT, Germany, Oct. 22-24, 2019.
  15. Title 10 CFR, Part 810 – Assistance to Foreign Atomic Energy Activities, US-NRC.


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