AP1000核电厂直接注射管线双端断裂小破口失水事故计算

基于压水堆最佳估算程序RELAP5/MOD3.3,对AP1000核电厂冷却剂系统和非能动堆芯冷却系统进行建模分析,得到在直接注入管线发生双端断裂事故下,系统压力、破口流量、系统水装量等关键参数的瞬态变化,计算结果与西屋公司采用NOTRUMP程序的计算结果基本一致。分析表明:AP1000的非能动专设安全设施能有效对一回路进行冷却和降压,防止堆芯过热,验证了AP1000发生DVI双端断裂事故后的安全性。     

RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究

AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的RELAP5/MOD3.3的适用性,以AP1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的RELAP5/MOD3.3的计算结果与试验数据符合较好。     

AP1000核电厂主给水管道断裂事故瞬态特性分析

AP1000是目前国际上典型的“三代”非能动核电厂,基于最佳估算程序RELAP5/MOD3.3,对AP1000核电厂系统进行了详细的建模分析,获得了主给水管道断裂事故下AP1000核电厂关键参数的瞬态特性和非能动系统响应特性。结果表明,事故过程中一、二回路的压力和温度呈现波动变化,一回路压力最大值为17.13 MPa,低于设计压力的91%,主蒸汽系统的压力也低于设计值的91%,满足验收准则的要求。     

福岛条件下AP1000核电厂事故前期研究

假设AP1000核电厂发生类似福岛核事故的初因事件,利用RELAP5/MOD3.3程序对事故早期的一、二回路系统和非能动安全系统进行模拟计算,得到了反应堆冷却剂系统压力、堆芯冷却剂温度、非能动安全系统流量等重要参数的瞬态变化。分析表明：在非能动余热排出系统完好的情况下,反应堆系统能顺利进入热停堆状态;如果非能动余热排出系统1根换热管发生双端断裂,则反应堆系统将会在5 h内发生严重事故。     

AP1000冷却剂强迫流动全部丧失事故瞬态特性分析

AP1000反应堆是目前国际上典型的"三代"核电厂,利用RELAP5/MOD 3.3程序对AP1000核电厂一回路系统进行整体建模,分析冷却剂强迫流动全部丧失事故下一回路主要热工水力参数的瞬态特性,并与COAST和LOFTRAN程序的计算结果进行了对比,发现两者具有相同的分布规律,表明利用RELAP5程序建立的计算模型可以准确模拟AP1000冷却剂强迫流动全部丧失事故下的热工水力特性。     

AP1000核电厂丧失主给水ATWS事故CMT注射分析

堆芯补水箱（CMT）是AP1000核电厂非能动堆芯冷却系统（PXS）的重要组成部分。在通常情况下,当主泵开启时,CMT即使被触发,也不能注入堆芯。然而在某些事故工况下,即使主泵开启,CMT也有可能注入,它将直接影响事故进程及分析结果。应用压水堆核电厂通用系统程序RELAP5MOD3.1对AP1000核电厂丧失主给水ATWS事故进行了计算分析,验证了美国西屋公司LOFT4AP2.0.1程序计算结果的正确性,并分析找出了CMT成功注入的根本原因。     

基于RELAP5的船用核动力装置二回路数字模型

根据核动力装置二回路系统一种新的结构设计方案,结合RELAP5/MOD3.4程序,建立了二回路汽轮机、冷凝器、给水泵及预热器等主要部件的物理模型。对二回路主要部件进行了单一部件模型适应性验证分析,并探讨了系统分析程序的局部计算能力。结果表明,RELAP5/MOD3.4程序模型的稳态计算结果与设计值基本吻合,动态计算也能够满足二回路主要部件的计算精度要求。     

基于RELAP5程序的AP1000典型事故瞬态特性研究

基于最佳估算程序RELAP5/MOD3.3,对AP1000系统进行了详细的建模分析,选取冷却剂泵卡轴事故、蒸汽发生器(SG)传热管破裂事故和直接注射管线双端断裂事故作为典型事故,获得了典型事故工况下关键参数的瞬态特性和非能动系统响应特性。结果表明:对于冷却剂泵卡轴事故,一回路最大压力为16.82 MPa,燃料包壳表面温度最大值为1 299K,满足验收准则的要求;对于SG传热管破裂事故,破损SG的水体积为231.54m3,小于AP1000蒸汽发生器255.563m3的总容积;对于直接注射管线双端断裂事故,AP1000的非能动堆芯冷却系统能对一回路进行冷却和降压,并防止堆芯裸露和燃料包壳过热。     

AP1000核电厂自动泄压系统误启动事故计算

基于最佳估算程序RELAP5/MOD3.3,对AP1000核电厂冷却剂系统和非能动堆芯冷却系统进行了建模分析,得到了自动泄压系统(ADS)阀门误启动事故下,系统压力、破口流量、系统水装量等参数的瞬态变化,计算结果与西屋公司采用NOTRUMP程序的计算结果进行了比较与分析。结果表明:AP1000核电厂的非能动专设安全设施能有效对一回路进行冷却和降压,防止堆芯过热,验证了AP1000发生ADS阀门误启动事故后的安全性。     

CPR1000全厂断电事故瞬态特性分析

用RELAP5/MOD3.4程序对CPR1000压水堆一回路系统进行整体建模,分析全厂断电事故下一回路主要参数的瞬态热工水力特性,并将RELAP5模型计算结果与THEMIS程序的计算结果进行对比,二者符合得较好。计算结果表明：该模型可较准确地模拟CPR1000在事故下的热工水力特性。     

AP1000冷管段小破口失水事故分析

基于压水堆最佳估算程序RELAP5/MOD3.4,对AP1000的冷却剂系统和非能动堆芯冷却系统进行建模分析,得到了系统压力、破口流量、燃料包壳温度等关键参数的瞬态变化,计算结果与西屋公司采用NOTRUMP程序计算的结果基本一致。分析表明：AP1000的非能动专设安全设施能有效地对一回路进行冷却和降压,防止堆芯过热,验证了AP1000发生冷管段小破口失水事故后的安全性。     

主给水管道破口尺寸对CPR1000二次侧非能动应急热阱事故缓解能力影响研究

在主给水管道破裂事故下，针对不同破口面积，利用RELAP5/MOD3.4程序对CPR1000压水堆一回路和二次侧非能动应急热阱的主要热工水力参数瞬态特性进行分析计算，验证采用CPR1000二次侧非能动应急热阱对事故的缓解能力和不同破口面积对主要参数的影响。结果表明：CPR1000在发生主给水管道破裂事故后，二次侧非能动应急热阱可及时向蒸汽发生器补水，同时导出堆芯余热，保证反应堆处于安全状态，随着破口面积的增大，初始时刻一回路压力和温度升高更快，随着二次侧非能动应急热阱的投入，压力和温度又迅速降低，说明CPR1000二次侧非能动应急热阱在文中所研究的破口面积范围内可非常有效地缓解事故。     

基于二回路的非能动排热系统的设计与MFLB条件下的验证

以中国改进型压水堆核电站CPR1000为研究对象,在其蒸汽发生器（SG）二次侧设计了1套非能动排热系统。为验证该系统在主给水管道破裂（MFLB）事故下的热量排出能力,采用RELAP5/MOD3.2程序对系统进行合理的简化并建模。结果表明：MFLB事故发生后,系统内可迅速建立起自然循环流动;该系统的及时投入可使一回路温度和压力的上升得到有效缓解,在隔离受影响的SG之前,一回路未出现整体沸腾,稳压器未满溢,保证了堆芯和一回路冷却剂系统的完整性。     

CPR1000二次侧非能动应急热阱设计与事故缓解能力分析

在主给水管道破裂事故下,利用RELAP5/MOD34程序对CPR1000压水堆一回路热工水力参数瞬态特性进行分析计算,验证采用空冷换热器的CPR1000二次侧非能动应急热阱对事故的缓解能力。计算结果表明：CPR1000在发生主给水管道破裂事故后,二次侧非能动应急热阱完全可及时向蒸汽发生器补水,同时导出堆芯余热,保证反应堆处于安全状态,从而验证了CPR1000二次侧非能动应急热阱的设计是成功的。     

Analysis of large break LOCA experiment at ROSA-III

Analysis of the ROSA-III test RUN 704 was performed by using the computer codes RELAP4J, RELAP4/MOD6 and RELAP5/MOD0 to verify the predictive capability of the codes for a BWR LOCA. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core, designed for in tegral LOCA/ECCS tests. The RUN 704 experiment at the ROSA-III test facility simulated a 200% double-ended offset shear break on the inlet side of the pump in the recirculation loop. From present analyses, key parameters which are important to predict major behavior during a BWR large break LOCA have been clarified and the promising predictive capability of the advanced code RELAP5 has been verified.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

Design and transient analyses of emergency passive residual heat removal system of CPR1000. Part Ⅰ: Air cooling condition

The steam generator secondary emergency passive residual heat removal system (EPRHRS) is a novel design for the conventional generation Ⅱ+ reactor CPR1000. The EPRHRS is designed to improve the safety and reliability of CPR1000 by completely or partially replacing the traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The EPRHRS consists of a steam generator (SG), a heat exchanger (HX), an air cooling tower, an emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, a model of the primary loop system and the EPRHRS was developed using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system affected by the EPRHRS. The transient characteristics of the primary loop system and the EPRHRS were calculated in the event of the feed line break accident. Sensitivity studies were also conducted to investigate effects of the main parameters of the EPRHRS on the transient characteristics of the primary loop and the EPRHRS. The EPRHRS could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRS could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and the EPRHRS loop, respectively. The present work is instructive for engineering design of the EPRHRS for Chinese NPPs.