Power-operated relief valve stuck-open accident and recovery scenarios in the Institute of Nuclear Energy Research integral system test facility

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.     

秦山核电厂SGTR事故及其处置研究

用ＲＥＬＡＰＳ／ＭＯＤ２程序和ＭＡＲＣＨ３程序对秦山核电厂多种假想ＳＧＴＲ事故及其所致严重事故进行了计算，分析了主要事故序列的事故进程，估算了严重事故下的熔堆时序，探讨了一些有效的事故处置措施及其干预效果。     

Quantification of severe accident source terms of a Westinghouse 3-loop plant

Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment.     

稳压器波动管蠕变破裂失效尺寸敏感性分析

以压水堆严重事故最佳估算程序为计算工具,研究了严重事故中稳压器波动管不同失效尺寸对严重事故进程和结果的影响。计算分析表明,稳压器波动管失效尺寸设为当量直径15cm左右的破口时可获得一个相对保守的计算结果,失效尺寸在12cm以下或18cm以上时,其计算结果没有15cm情况下的严重。研究结果可为深入研究压水堆核电厂严重事故现象提供参考。     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

压水堆核电站完全丧失给水引发的严重事故研究

采用严重事故最佳估算程序RELAP5/SCDAPSIM/MOD3.2，建立美国Surry-2核电站的详细计算模型，对完全丧失给水（TLFW）引发的堆芯熔化事故进行研究分析。为准确预测压力容器内堆芯熔化的进程，为二级概率安全评价提供可信的初始条件，计算中考虑了一回路压力边界的蠕变破裂失效，并评价了人为干预对堆芯熔化进程及事故后果的影响。计算结果表明，由完全丧失给水引发的压水堆核电站严重事故不会出现人们担心的高压熔堆；反应堆压力容器下封头的失效位置不是在其底部，而是在其侧面；通过打开稳压器释放阀对一回路实施主动卸压能够大大推迟事故的进程。     

CPR1000全厂断电叠加蒸汽发生器安全阀误开启事故引起的严重事故分析

利用MELCOR程序对CPR1000全厂断电叠加蒸汽发生器(SG)安全阀误开启事故引发的严重事故进行建模与分析,初步实现了对CPR1000严重事故进程的仿真计算与模拟。文中重点分析了无轴封泄漏和辅助给水、有轴封泄漏和辅助给水、有轴封泄漏但无辅助给水3种不同假设条件下CPR1000全厂断电严重事故的响应进程和结果。计算结果显示,SG安全阀误开启对事故进程有重要影响。在无轴封泄漏和辅助给水的情况下,压力容器在9576 s失效;当存在辅助给水时,压力容器失效延后近30000 s;而当存在轴封泄漏时,压力容器失效延后50 s左右。结果证明了发生全场断电叠加SG安全阀误开启事故情况下辅助给水和轴封泄漏对事故起到有效缓解作用。     

中压安注对RRA连接模式下严重事故进程影响分析

针对中国改进型百万千瓦级压水堆（CPR1000）核电机组在中间停堆反应堆余热排出系统（RRA）连接模式下失去高低压安注和喷淋的冷却剂丧失事故（LOCA）,采用MAAP5程序对参考机组的反应堆堆芯、反应堆冷却剂系统以及安全壳系统进行模拟计算,同时结合计算结果分析中压安注系统对该严重事故序列进程的影响,并研究其对事故的缓解作用。分析结果表明,在RRA连接模式下出现LOCA导致的堆芯裸露和升温过程中,中压安注的及时注入能有效地限制堆芯的升温行为,并可对严重事故进程起到重要的缓解作用,甚至为事故工况下失去高低压安注和喷淋时避免堆芯完整性遭到破坏提供可能。最后,根据分析结果针对现行核电机组的运行规程提出改进建议:对于中压安注箱的行政隔离行为,只对其电气开关做相应的隔离操作,而对安全壳厂房内的阀门就地部分做挂牌警示,不做现场挂锁的操作,这样不仅可避免在正常运行工况下中压安注箱误注入行为的发生,同时能够在RRA连接模式下发生LOCA时有效地保障堆芯的完整性,在保证电厂正常安全运行的同时,提高了机组在该模式下发生严重事故的缓解能力。      

Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis

ABSTRACT

Severe accident codes (e.g. MAAP, RELAP, and MELCORE) model various physical phenomena during severe accidents. Many analyses using these codes for safety margin evaluation are impractical due to large computational costs. Surrogate models have an advantage of quickly reproducing multiple results with a low computational cost. In this study, we apply the singular value decomposition to the time-series results of a severe accident code to develop a reduced order modeling (ROM). Using the ROM, the probabilistic safety margin analysis for the station blackout with a total loss of feedwater capabilities at a boiling water reactor is carried out. The dominant parameters to the accident progression are assumed to be the down-time and the recovery-time of the reactor core isolation cooling system, and decay heat. To reduce the number of RELAP5/SCDAPSIM analyses while maintaining the prediction accuracy of ROM, we develop a data sampling method based on adaptive sampling, which selects the new sampling data based on the dissimilarity with the existing training data for ROM. Our ROM can rapidly reproduce the time-series results and can estimate core conditions. By reproducing multiple results by ROM, a time-dependent core damage probability distribution is calculated instead of the direct use of RELAP5/SCDAPSIM.     

Analysis of pressurizer level control system using MAAP4 code

The code initialization effort has been troubling code users for decades for system transient and severe accident analyses using codes such as RETRAN, MAAP4, MAAP5 and MELCOR. The purpose of this work is to demonstrate an approach that could be considered a generic method to address the code initialization problem. This was demonstrated by developing a pressurizer level control model and temperature dependent level control logic in MAAP4 without re-compiling with the source code. The method would enhance the simulation capability and accuracy of a severe accident analysis by transient and severe accident analyses codes. The demonstration case used MAAP4 to show that the adopted proportional-integral controller with the temperature dependent level control logic would reduce its code steady state errors to zero. The subsequent transient response would become more realistic. The proposed method provides a convenient and exemplified approach for code initialization which is applicable to the next generation of codes that couple with the balance of plant models. These codes include the MAAP5 code and others future codes that could simulate the whole plant by a single and elaborate plant model with exhausting component and phenomenological models.     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

大型非能动压水堆核电厂安全注射系统旋启式止回阀阻力变化对长期堆芯冷却影响分析

大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

Optimum RCS depressurization strategy for effective severe accident management of station black out accident

Plant specific severe accident management guidelines (SAMG) for operating plants are developed and implemented in Korea as was required by government policy on severe accident. Korea Institute of Nuclear Safety (KINS) has recently reviewed feasibility of the developed SAMG for Ulchin unit 1 plant. Among the strategies referred in SAMG, we have intensively analyzed the reactor coolant system (RCS) depressurization strategy during station black out (SBO) accident scenario, which has a high probability of occurrence according to Ulchin unit 1 Probabilistic Safety Analysis (PSA). In depressurization strategy of the current SAMG, operators need to depressurize rapidly RCS pressure below 2.75 MPa using pressurizer (PZR) pilot operated safety relief valves (POSRVs) for high pressure accident like SBO. The rapid depressurization is effective in allowing the water of safety injection tank (SIT) to be injected into the core, but an excessive discharge of the SIT water is not desirable for an economical use of SIT inventory. Lack of SIT water accelerates the core damage in case the failed electric power do not recover in due to time. The SIT inventory economy means here that we should not waste the water inventory of SIT and use it in the most efficient way to cool the core. In case we do not use it in an economical way, the SIT might be depleted too rapidly, thus leaving an insufficient reservoir for post-depressurization cooling. The quantification of this SIT inventory economy for plant specific situation is of interest to develop an optimum depressurization strategy. In this study we have analyzed an effectiveness of current depressurization strategy for SBO accident with the severe accident analysis code MELCOR 1.8.5 which has been used for regulatory purpose in KINS. The entry time of severe accident management, a grace time gained by the current strategy, and the economy of the discharge mass flow rate for Ulchin plant were evaluated. Moreover, through a simple energy balance equation we could find an optimum strategy for RCS depressurization. The proposed strategy is based on finding an optimum discharge rate for an efficient use of the SIT inventory and it allows us to handle an SBO accident with higher confidence. The proposed strategy is yet a theoretical one, but possibilities of how to incorporate this strategy into engineered safety features are also discussed.     

稳压器卸压阀卸压效果研究

基于国际上模拟严重事故瞬态过程最详细的机理性程序SCDAP/RELAP5/MOD3.1,主要分析研究了核电站未紧急停堆的预期瞬变(ATWS)初因(失去主给水、失去厂外电和控制棒失控提升)叠加辅助给水失效导致的堆芯熔化严重事故进程,并验证阻止ATWS导致堆芯熔化进程的一次侧卸压缓解措施的充分性和有效性.计算分析结果显示,一列稳压器卸压阀不足以充分降低一回路压力,压力仍然停留在10MPa以上,存在很大高压熔堆的风险.增加一列卸压阀可把一回路压力降低到3MPa左右,安注系统得以投入,及时有效地阻止堆芯熔化进程,降低了高压熔堆风险.分析结果还显示高压安注系统的投入对一回路卸压具有重要影响.     

小型堆包壳破损状态下利用稳压器去除放射性研究

利用MELCOR程序对小型船用堆稳压器喷雾除气过程及停堆过程进行建模,进而模拟核动力装置从功率运行至降功率除气,以及除气结束后停堆消除稳压器气腔的全部物理过程。通过对反应堆关键运行参数变化趋势的仿真分析,验证了模拟的物理过程的合理性。结合建立的除气及停堆仿真模型,计算分析了包壳破损状态下,稳压器喷雾除气、停堆过程对稳压器内惰性气体含量的影响,评估了稳压器高点放气和喷雾除气对放射性物质的去除作用。研究结果能为小型堆包壳破损状态下放射性安全管理策略提供指导和帮助。     

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4–CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4–CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4–CANDU code predictions.