Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy...Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy water moderator and the utilization of liquid fuel,HWMSRs can achieve a high neutron economy.In this study,a large-scale small modular HWMSR with a thermal power of 500 MWth was proposed and studied.The criticality of the core was evaluated using an in-house critical search calculation code(CSCC),which was developed based on Standardized Computer Analyses for Licensing Evaluation,version 6.1.The preliminary fuel cycle performances(initial conversion ratio(CR),initialfissile fuel loading mass,and temperature coefficient)were investigated by varying the lattice pitch(P)and the molten salt volume fraction(VF).The results demonstrate that the temperature coefficient can be negative over the range of investigated Ps and VFs for both 233U-Th and LEU-Th fuels.A core with a P of 20 cm and a VF of 20%is recommended for 233U-Th and LEU-Th fuels to achieve a high performance of initial CR and fuel loading.Regarding TRU-Th fuel,a core with a smaller P(~5 cm)and larger VF(~24%)is recommended to obtain a negative temperature coefficient.展开更多
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)...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.展开更多
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.11905285)+1 种基金the National Natural Science Foundation of China(No.11790321)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)。
文摘Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy water moderator and the utilization of liquid fuel,HWMSRs can achieve a high neutron economy.In this study,a large-scale small modular HWMSR with a thermal power of 500 MWth was proposed and studied.The criticality of the core was evaluated using an in-house critical search calculation code(CSCC),which was developed based on Standardized Computer Analyses for Licensing Evaluation,version 6.1.The preliminary fuel cycle performances(initial conversion ratio(CR),initialfissile fuel loading mass,and temperature coefficient)were investigated by varying the lattice pitch(P)and the molten salt volume fraction(VF).The results demonstrate that the temperature coefficient can be negative over the range of investigated Ps and VFs for both 233U-Th and LEU-Th fuels.A core with a P of 20 cm and a VF of 20%is recommended for 233U-Th and LEU-Th fuels to achieve a high performance of initial CR and fuel loading.Regarding TRU-Th fuel,a core with a smaller P(~5 cm)and larger VF(~24%)is recommended to obtain a negative temperature coefficient.
文摘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.
