Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT

Abstract

The current Russian regulatory documents on the safety of nuclear power plant (NPP) specify the requirements regarding design basis accidents (DBAs) and beyond design basis accidents (BDBAs), including severe accidents (SAs) with core meltdown, in NPP design (NP-001-15, NP-082-07, and others). For a rigorous calculational justification of BDBAs and SAs, it is necessary to develop an integral CC that will be in line with the requirements of regulatory documents on verification and certification (RD-03-33-2008, RD-03-34-2000) and will allow for determining the amount of data required to provide information within the scope stipulated by the requirements for the structure of the safety analysis report (SAR) (NP-006-16). The system of codes for realistic analysis of severe accidents (SOCRAT) (formerly, thermohydraulics (RATEG)/coupled physical and chemical processes (SVECHA)/behavior of core materials relocated into the reactor lower plenum (HEFEST)) was developed in Russia to analyze a wide range of SAs at NPP with water-cooled water-moderated power-generating reactor (WWER) at all stages of the accident. Enhancements to the code and broadening of its applicability are continually being pursued by the code developers (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)) with OKB Gidropress JSC and other organizations. Currently, the SOCRAT/В1 code can be used as a base tool to obtain realistic estimates for all parameters important for computational justification of the reactor plant (RP) safety at the in-vessel stage of SAs with fuel melting. To perform analyses using CC SOCRAT/В1, the experience gained during execution of thermohydraulic codes is applied, which allows for minimizing the uncertainties in the results at the early stage of an accident scenario. This study presents the results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT/В1. Approaches have been considered to develop calculational models and analyze SAs using CC SOCRAT. This process, which is clearly structured in OKB Gidropress JSC, provides a noticeable reduction in human involvement, and reduces the probability of erroneous results.

This study represents the principal results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT, as well as a list of the tasks planned for 2021–2023. CC SOCRAT/B1 is used as the base thermohydraulic SAs code.

Graphical abstract

Keywords

system of codes for realistic analysis of severe accidents (SOCRAT) / design basis accidents (DBAs) / severe accidents (SAs) / computer code (CC) / nuclear power plant (NPP) design / water-cooled water-moderated (WWER) / modeling / model / safety requirements

References

[1] Bolshov L, Strizhov V. SOCRAT: the system of codes for realistic analysis of severe accidents. In: Proceedings of the 2006 International Congress on Advances in Nuclear Power Plants, Reno, USA, 2006

[2] Zaitsev A A, Lukiyanov A A, Strizhov V F, Development of integral code for modeling of beyond design basis accidents at NPP with WWER. In: Proceedings of the 4th International Scientific and Technical Conference “Safety Assurance of NPP with WWER”, OKB Gidropress JSC, 2005

[3] Shchekoldin V V, Pantyushin S I, Sorokin Yu S. Analysis of core melt accidents at WWER RP by codes SOCRAT/B1 and RELAP/SCDAPSIM. In: Proceedings of the 6th ISTC “Safety Assurance of NPP with WWER”, Podolsk, Russia, 2009

[4] Sorokin Yu S, Shchekoldin V V, Tuchnolobov A A, Assessment of codes RATEG/SVECHA/HEFEST and MELCOR for analysis of severe accidents. In: Proceedings of the 3rd ISTC “Safety Assurance of NPP with WWER”, Podolsk, Russia, 2003

[5] Pantyushin S I, Sorokin Yu S, Shchekoldin V V, Sensitivity analysis of the code SOCRAT/B1 and adaptation of calculational model for analysis of a severe accident at RP WWER-1000 (V-320) during operation at 104% power. In: Proceedings of the XIIth International Conference “NPP Safety and Staff Trainin 2011”, Obninsk, Russia, 2011

[6] S.I.Pantyushin, Yu.S. Sorokin, V.V.Shchekoldin, Experience in analyses if severe accidents at WWER RP using the code SOCRAT/B1. Annual Report on Main Scientific and Engineering Work Performed in OKB Gidropress JSC in 2009, 2009

[7] Aulova O V, Gasparov D L, Lityshev A V, Experience in application of the computer code SOCRAT in OKB Gidropress JSC. In: Proceedings of ISPC on Nuclear Power Engineering “Safety, Efficiency, Life of NPP” (ISPC NPE-11), Sevastopol, Russia, 2013

[8] Lityshev A V, Pantyushin S I, Gasparov D L, Development of mathematical model for computer code SOCRAT applied to RP WWER-1200 (NV NPP-2) and cross-verification with thermohydraulic codes TECH-M-97 and KORSAR/GP at the initial stage of severe accidents. In: Proceedings of the 4th STC of Young Scientists and Specialists in Nuclear Industry “TEAM-2012”, Saint-Petersburg, Russia, 2012

[9] Sorokin Yu S, Shchekoldin V V, Tuchnolobov A A, Evaluation of RATEG/SVECHA/HEFEST and MELCOR codes for severe accident analysis. In: Proceedings of “the 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics”, Seoul, Republic of Korea, 2003

[10] Pantyushin S I, Tomashchik D Yu, Dolganov K S. Cross-verification of codes TECH-M-97 and SOCRAT when analyzing the initial stage of severe accident at RP WWER-1000 (V-320) during operation at 104% power. In: Proceedings of International Conference “Scientific Session of NRNU MEPhI”, Moscow, Russia, 2010

[11] LityshevA V, Pantyushin S I, Aulova O V, Experience in severe accident analyzing for RP WWER-1000 using the CC SOCRAT. Collected Scientific, Engineering and Advertising Articles No. 14 Annual Report on Main Scientific and Engineering Work Performed in OKB Gidropress JSC, 2014

[12] Gasparov D L, Pantyushin S I, Aulova O V, Analysis of the core behavior under severe accidents using CC SOCRAT. In: Annual Conference of Young Specialists “Innovations in Nuclear Power Engineering”, Moscow, Russia, 2014

[13] Pantyushin S I, Sorokin Yu S, Sotskov E V, Comparison of codes TECH-M-97, KORSAR/GP and SOCRAT/B1 when analy-zing the initial stage of severe accident at RP WWER-1200. In: Proceedings of the 6th ISTC “Safety Assurance of NPP with WWER”, Podolsk, Russia, 2009

[14] Computer code SOCRAT/B1. Qualification certificate No. 275 of 13.05.2010 for the software. 2010

[15] Zaitsev A A, Lukiyanov A A, Strizhov V F, Development of integral code for modeling of beyond design basis accidents at NPP with WWER. In: Proceedings of the 4th ISTC “Safety assurance of NPP with RP WWER”, Podolsk, Russia, 2005

[16] Sorokin Yu S, Pantyushin S I. Adaptation and application of the code SOCRAT for analysis of severe accidents at RP WWER. In: Proceedings: Seminar on “Analysis of Severe Accidents and Accident Management”, Yerevan, Armenia, 2010

[17] Lityshev A V, Pantyushin S I, Gasparov D L, Results of adaptation and pilot usage of the computer code SOCRAT using the supercomputer technologies. In: Proceedings of the 8th ISTC “Safety assurance of NPP with RP WWER”, Podolsk, Russia, 2013

[18] Dolganov K S, Tomashchik D Yu, Kiselev A E, Experience gained in implementing the elements of parallel computing to SOCRAT severe accident code. In: Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermalhydraulics, Pisa, Italy, 2013

[19] Pantyushin S I, Bukin N V, Bykov M A, Adaptation of the severe-accident code SOCRAT for the usage of supercomputer technologies. Collected Scientific, Engineering and Advertising Articles No. 12 Annual Report on Main Scientific and Engineering Work Performed in OKB Gidropress JSC in 2011, 2011

[20] Efanov A D, Zaitsev A A, Lukiyanov A A, Verification of the containment code KUPOL-M by the data of large-scale integral experiments. In: Proceedings of the 2nd ISTC “Safety assurance of NPP with WWER”, Podolsk, Russia, 2001

[21] Berkovich V M, Taranov G S, Peresadko V G, Computational analysis of processes in the containment and reactor plant AES-92 under LB LOCA and loss of A.C. power supply sources at NPP of a new generation. In: Proceeding of the 2nd ISTC “Safety assurance of NPP with WWER”, Podolsk, Russia, 2001

[22] Del Nevo A, Adorni M, D’Auria F, Further development of estimation database for advanced computer codes with respect to WWER technologies: comparative data on transients with LB-LOCA at a test bench PSB-WWER. In: Proceedings of the 6th ISTC “Safety assurance of NPP with WWER”, Podolsk, Russia, 2009

[23] Berkovich V Ya. Results of stress-tests and planning of the measures on NPP safety improvement on their basis. In: Proceedings of Meeting of the Board of Governors and the Board of Directors of WANO NPP-M, Sofia, Bulgaria, 2012

[24] Pantyushin S I, Gasparov D L, Asadskiy S I, Findings of assessment of a feasibility to apply the external reactor pressure vessel cooling system at existing NPP with WWER-440. In: Proceedings of the 8th ISTC “Safety assurance of NPP with WWER”, Podolsk, Russia, 2013

[25] Pantyushin S I, Gasparov D L, Asadskiy S I. Collected scientific, engineering and advertising articles No. 14 Annual Report on Main Scientific and Engineering Work Performed in OKB Gidropress JSC in 2013. 2014

[26] Pantyushin S I. The results of evaluation of technical possibilities of implementing external cooling of a reactor vessel (IVR) at operating NPPs with WWER-1000. In: Meeting on “In-vessel retention (IVR) Strategy to Develop the Severe Accident Management Guidelines of VVER-1000 Reactors”, Vienne, France, 2013

[27] Ryzhov S B, Mokhov V A, Nikitenko M P. New designs of reactor plants with medium-power WWER. In: Proceedings of Conference “Nuclear Power Engineering and Safety Assurance” (NPES-2009), Yerevan, Armenia, 2009

[28] Pantyushin S I, Sorokin Yu S, Bukin N V, Evaluation of thermal loads and conditions for melt retention in the reactor pressure vessel under severe accidents at WWER-600 RP and WWER RP TOI. The XIIth International Conference “NPP safety and staff training”, Obninsk, Russia, 2011

[29] Pantyushin S I, Bykov M A, Mokhov V A, Development of the reactor vessel cooling and melt retention system in case of severe beyond design-basis accidents for NPP with WWER-600 RP and WWER-TOI RP. In: Proceedings of the 7th ISTC “Safety assurance of NPP with WWER”, Podolsk, Russia, 2011

[30] Pantyushin S I, Frizen E A, Semishkin V P, Consideration of a possibility for corium retention (reactor internals and core melt) in the vessel of WWER reactor with power from 600 to 1300 MW. In: Proceedings of the “5th European Review Meeting on Severe Accident Research (ERMSAR)”, Cologne, Germany, 2012

[31] Gasparov D L, Pantyushin S I, Frizen E A, The results of computational analyses In-vessel corium retention for NPP with WWER-6a00 using realistic approaches. In: Proceedings of the “7th European Review Meeting on Severe Accident Research (ERMSAR-2015)”, Marseille, France, 2015

[32] Moiseenko E V, Filippov A S, Drobyshevsky N I, Varia: an application for mass computing and statistical analysis of the simulation results in BEPU safety assessment. In: 19th International Conference on Nuclear Engineering, Chiba, Japan, 2013

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