Core designing of a new type of TVS-2M FAs:neutronics and thermal-hydraulics design basis limits

One of the most important aims of this study is to improve the core of the current VVER reactors to achieve more burn-up(or more cycle length)and more intrinsic safety.It is an independent study on the Russian new proposed FAs,called TVS-2M,which would be applied for the future advanced VVERs.Some important aspects of neutronics as well as thermal hydraulics investigations(and analysis)of the new type of Fas are conducted,and results are compared with the standards PWR CDBL.The TVS-2M FA contains gadolinium-oxide which is mixed with UO_(2)(for different Gd densities and U-235 enrichments which are given herein),but the core does not contain BARs.The new type TVS-2M Fas are modeled by the SARCS software package to find the PMAXS format for three states of CZP and HZP as well as HFP,and then the whole core is simulated by the PARCS code to investigate transient conditions.In addition,the WIMS-D5 code is suggested for steady core modeling including TVS-2M FAs and/or TVS FAs.Many neutronics aspects such as the first cycle length(first cycle burn up in terms of MWthd/kgU),the critical concentration of boric acid at the BOC as well as the cycle length,the axial,and radial power peaking factors,differential and integral worthy of the most reactive CPS-CRs,reactivity coefficients of the fuel,moderator,boric acid,and the under-moderation estimation of the core are conducted and benchmarked with the PWR CDBL.Specifically,the burn-up calculations indicate that the 45.6 d increase of the first cycle length(which corresponds to 1.18 MWthd/kgU increase of burnup)is the best improving aim of the new FA type called TVS-2M.Moreover,thermal-hydraulics core design criteria such as MDNBR(based on W3 correlation)and the maximum of fuel and clad temperatures(radially and axially),are investigated,and discussed based on the CDBL.

Seismic performance of steel MRF building with nonlinear viscous dampers

This paper presents an experimental study of the seismic response of a 0.6-scale three-story seismicresistant building structure consisting of a moment resisting frame （MRF） with reduced beam sections （RBS）, and a frame with nonlinear viscous dampers and associated bracing （called the DBF）. The emphasis is on assessing the seismic performance for the design basis earthquake （DBE） and maximum considered earthquake （MCE）. Three MRF designs were studied, with the MRF designed for 100%, 75%, and 60%, respectively, of the required base shear design strength determined according to ASCE 7-10. The DBF with nonlinear viscous dampers was designed to control the lateral drift demands. Earthquake simulations using ensembles of DBE and MCE ground motions were conducted using the real-time hybrid simulation method. The results show the drift demand and damage that occurs in the MRF under seismic loading. Overall, the results show that a high level of seismic performance can be achieved under DBE and MCE ground motions, even for a building structure designed for as little as 60% of the base shear design strength required by ASCE 7-10 for a structure without dampers.

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

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.

An old issue and a new challenge for nuclear reactor safety

Nuclear reactor safety(NRS)and the branch accident analysis(AA)constitute proven technologies:these are based on,among the other things,long lasting research and operational experience in the area of water cooled nuclear reactors(WCNR).Large break loss of coolant accident(LBLOCA)has been,so far,the orienting scenario within AA and a basis for the design of reactors.An incomplete vision for those technologies during the last few years is as follows:Progress in fundamentals was stagnant,namely in those countries where the WCNR were designed.Weaknesses became evident,noticeably in relation to nuclear fuel under high burn-up.Best estimate plus uncertainty(BEPU)techniques were perfected and available for application.Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked(however,quite irrelevant in case of LBLOCA).The time delay between technological discoveries and applications was becoming longer.The present paper deals with the LBLOCA that is inserted into the above context.Key conclusion is that regulations need suitable modification,rather than lowering the importance and the role of LBLOCA.Moreover,strengths of emergency core cooling system(ECCS)and containment need a tight link.