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Frontiers in Energy

ISSN 2095-1701

ISSN 2095-1698(Online)

CN 11-6017/TK

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Front. Energy    2021, Vol. 15 Issue (4) : 872-886    https://doi.org/10.1007/s11708-021-0796-2
REVIEW ARTICLE
Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT
S.I. PANTYUSHIN, А.V. LITYSHEV, A.V. NIKOLAEVA, O.V. AULOVA, D.L. GASPAROV, V.V. ASTAKHOV, M.A. BYKOV()
Non-stationary Thermohydraulics Department, OKB Gidropress JSC 142103, Podolsk, Russia
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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.
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     
Corresponding Author(s): M.A. BYKOV   
Online First Date: 16 December 2021    Issue Date: 04 January 2022
 Cite this article:   
S.I. PANTYUSHIN,А.V. LITYSHEV,A.V. NIKOLAEVA, et al. Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT[J]. Front. Energy, 2021, 15(4): 872-886.
 URL:  
https://academic.hep.com.cn/fie/EN/10.1007/s11708-021-0796-2
https://academic.hep.com.cn/fie/EN/Y2021/V15/I4/872
Thermohydraulic parameter/signal of control and protection system Value of the parameter in the RELAP5 code Value of the parameter in the SOCRAT code
Reactor outlet temperature reached its maximum value 4 s/603 K 2 s/600 K
Secondary-side pressure reached a setpoint for turbine stop control valves (SCV-T) closing 24 s/4.90 MPa 23 s/4.90 MPa
Pressure in the pressurizer reached its local minimum value 40 s/13.50 MPa 45 s/13.50 MPa
Reactor inlet temperature reached its local minimum value 46 s/547 K 34 s/546 K
Reactor outlet temperature reached its local minimum value 52 s/558 K 53 s/560 K
Secondary-side pressure reached a setpoint for steam dump valve to the condenser (BRU-K) actuation 163 s/6.67 MPa 162 s/6.67 MPa
Pressure in the pressurizer reached its local maximum value 166 s/15.10 MPa 171 s/15.40 MPa
Reactor inlet temperature reached its set value 110 s/553 K 110 s/550 K
Reactor outlet temperature reached its local maximum 130 s/581 K 140 s/584 K
Secondary-side pressure reached a setpoint for BRU-K closing 189 s/5.28 MPa 193 s/5.28 MPa
Pressure in the pressurizer reached its maximum value 200 s/14.20 MPa 200 s/14.10 MPa
Tab.1  Comparison of the calculation results obtained using SOCRAT/B1 and RELAP5/MOD3.2
Fig.1  Scheme of interaction of the RATEG, SVECHA and HEFEST modules of СС SOCRAT.
Fig.2  Simplified diagram of interaction between modules RATEG and ANGAR/KUPOL.
Fig.3  Nodalization of a loop with PRZ.
Fig.4  Nodalization of the secondary side steam generator.
Fig.5  Nodalization of the reactor pressure vessel.
Fig.6  Nodalization of the reactor pressure chamber (reactor downcomer and lower plenum).
Fig.7  Sequence of events at the in-vessel stage of SAs.
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