Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

反应堆主回路卸压对压力容器外裂变产物释放的影响分析

本工作耦合建立了600 MW压水堆核电厂热工水力、裂变产物行为和放射性后果评价的分析模型,选取SB-LOCA、SGTR、SBO和LOFW等4个高压熔堆事故序列,研究了主回路卸压对压力容器外裂变产物释放的影响,包括主回路卸压对压力容器外裂变产物释放的缓解效应和其他负面影响。分析表明：实施主回路卸压可缓解高压熔堆事故序列下压力容器外的释放,但卸压工况下事故早期安全壳内的气载放射性活度较基准工况下的大。相关分析结论可作为严重事故管理导则制定的技术基础。     

基于氢气风险分析的ADS手动卸压策略研究

先进非能动压水堆设计采用自动卸压系统(ADS)对一回路进行卸压,严重事故下主控室可手动开启ADS,缓解高压熔堆风险。然而ADS的设计特点可能导致氢气在局部隔间积聚,带来局部氢气风险。本文基于氢气负面效应考虑,对利用ADS进行一回路卸压的策略进行研究,为严重事故管理提供技术支持。选取全厂断电始发的典型高压熔堆严重事故序列,利用一体化事故分析程序,评估手动开启第1~4级ADS、手动开启第1~3级ADS、手动开启第4级ADS 3种方案的卸压效果,并分析一回路卸压对安全壳局部隔间的氢气负面影响。研究结果表明,3种卸压方案均能有效降低一回路压力。但在氢气点火器不可用时,开启第1~3级ADS以及开启第1~4级ADS卸压会引起内置换料水箱隔间氢气浓度迅速增加,可能导致局部氢气燃爆。因此,基于氢气风险考虑,建议在实施严重事故管理导则一回路卸压策略时优先考虑采用第4级ADS进行一回路卸压。     

Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences

A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6  min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW.     

Evaluation of the RCS depressurization strategy for the high pressure sequences by using SCDAP/RELAP5

As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough.     

AP1000核电厂丧失正常给水稳压器防满水措施研究

在某AP1000核电厂丧失正常给水事件中，由于一系列的误操作导致稳压器满水，而稳压器安全阀在多次打开后可能无法重新关闭，不满足核电厂Ⅱ类工况验收准则。文章分析了该事件中稳压器满水的原因，即在非能动余热排出热交换器（PRHR HX）冷却能力充足的情况下，系统不适当的降压导致环路中冷却剂闪蒸，进而导致稳压器满水，此时通过开启堆顶放气阀启动应急下泄的方式无法有效降低稳压器液位。最后给出了AP1000核电厂丧失正常给水事故中防止稳压器满水的建议措施，即在RCS降压过程中，应确保RCS压力始终高于热管段温度对应的饱和压力，进而确保冷却剂不发生闪蒸。      

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.     

AP1000核电厂丧失正常给水稳压器防满水措施研究

In an accident of loss of feedwater in an AP1000 plant, the pressurizer was filled with water for a series of improper operations, and the safety valves may not be qualified to re-close following multiple cycles of opening, which is not acceptable in Condition Ⅱ events. The paper analyzes the causes for the filling of water in the pressurizer in this event, that is, the instantaneous evaporation of coolant in the loop during the process of improper depressurization of RCS while the PRHR HX is with sufficient cooling capability. At this time, the water level in the pressurizer level cannot be decreased by opening the reactor vessel head vent valves for emergency letdown. Finally, the recommended measure is provided to prevent the filling of water in the pressurizer during loss of normal feedwater for AP1000 NPP. The RCS pressure should always be higher than the saturation pressure corresponding to the temperature of the hot legs to avoid the coolant evaporation.     

核电厂严重事故下卸压对氢气产生的影响分析

研究了1000MWe压水堆核电厂在典型的高压严重事故序列下卸压对氢气产生的影响。分析结果表明,开启1列、2列和3列卸压阀进行一回路卸压均会在堆芯熔化进程的3个阶段导致氢气产生率的明显增大：1）堆芯温度1500～2100K;2）堆芯温度2500～2800K;3）从形成由硬壳包容的熔融池（2800K）到熔融物向压力容器下封头下落。开启卸压阀的列数越多,氢气产生率的增大越明显。     

压水堆一回路卸压措施影响因素研究

研究压水堆一回路管道小小破口失水事故叠加辅助给水失效导致的高压堆芯熔化严重事故进程，对比验证不同严重事故缓解措施入口温度条件下一回路卸压缓解途径的充分性和有效性，并确认较佳的一回路冷却系统（RCS）降压途径。结果显示，以低于650℃的温度作为降压缓解措施入口条件，可及时恢复可能的堆芯冷却能力。一、二回路卸压效果分析表明，考虑了长期衰变热移出注水流量和堆芯过冷度要求，较佳的卸压配置为初期打开一列稳压器卸压阀，同时迅速恢复辅助给水并开启蒸汽发生器卸压阀。      

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.     

一体化多用途的非能动小型压水反应堆发生假定事故的初步分析

采用RELAP5/MOD3.2系统程序建立一体化小型反应堆的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施。一体化多用途的非能动小型压水反应堆(SIMPLE)热功率为660 MWt(电功率大于200 MWe)。针对SIMPLE的直接安注管线(DVI)双端断裂事故和DVI2英寸(50.8mm)小破口失水事故(SBLOCA)进行分析。计算结果表明:对于直接安注管线双端断裂事故,破口和自动降压系统(ADS)能有效地使反应堆冷却系统降压,安注箱(ACC)和安全壳内置换料水箱(IRWST)能实现堆芯补水,确保堆芯冷却;对于DVI的SBLOCA,非能动专设安全设施能有效对RCS进行冷却和降压,防止堆芯过热。     

堆芯晚期注水对一回路压力的影响分析

分析典型的1000 MW级压水堆核电厂在高压严重事故序列下,堆芯晚期注水对压力容器失效时一回路压力的影响.分析结果表明,在开启1列稳压器卸压阀的情况下,稳压器波动管可能会在压力容器失效之前发生蠕变失效使一回路被动卸压,堆芯晚期注水不会造成一回路压力大幅增大,但波动管失效的时间和尺寸存在较大的不确定性.在开启2列或3列卸...     

大型先进压水堆IVR策略评估和裂变产物分布计算

大型先进压水堆通过堆内熔融物滞留（IVR）策略来缓解严重事故后果以降低安全壳失效风险。其中堆腔注水系统（CIS）被引入来实现IVR。本文使用严重事故分析软件计算大型先进压水堆在冷管段双端断裂事故下的事故进程、热工水力行为、堆芯退化过程和下封头熔融池传热行为,评估能动CIS的事故缓解能力。计算结果表明,事故后72 h,下封头外表面热流密度始终低于临界热流密度（CHF）,表明IVR策略有效。此外,计算分析了惰性气体、非挥发性和挥发性裂变产物的释放和迁移行为。计算发现,IVR下更多的放射性裂变产物分布在主系统内,壁面核素再悬浮形成气溶胶的行为被消除,安全壳壁面上沉积的核素被大量冷凝水冲刷进入底部水池。总体来说,IVR策略能更好地管理放射性核素分布,减小放射性泄漏威胁。     

船用堆大破口失水叠加全船断电严重事故源项分析

以某船用压水堆为研究对象,采用MELCOR程序建立事故分析模型,研究大破口失水事故叠加全船断电严重事故下放射性裂变产物的行为,着重分析了惰性气体和CsI的释放、迁移、滞留特点及在堆舱内的分布。结果表明,83.12%惰性气体从堆芯释放出来,并主要存在于堆舱的气空间;83.08%的CsI从堆芯释放出来,其中,72.66%滞留在堆坑熔融物与一回路内,27.34%释放到堆舱内,并主要溶解于舱底水池中。本文分析结果可为舱室剂量评估、核应急管理提供依据。     

AP1000核电厂应对全厂断电事故的稳压器防满溢对策研究

AP1000核电厂若在全厂断电事故下丧失正常给水,会引起稳压器满溢,将通过稳压器安全阀排放液体冷却剂,引起反应堆冷却剂水装量流失,增大反应堆堆芯裸露的风险。与此同时,安全壳内的放射性水平因稳压器满溢可能会增大,增大向环境排放大量放射物质的可能。为防止稳压器满溢,本工作进行了解决或缓解稳压器满溢的对策研究。结果表明,增大非能动余热排出系统（PRHRS）热交换器的传热面积,可防止稳压器满溢;合理降低安全壳内置换料水箱（IRWST）的背压,可增大达到稳压器满溢的裕度,有效地缓解稳压器满溢;增大稳压器的自由容积,可防止稳压器满溢。此结论对AP1000核电厂的设计和事故分析有一定的参考作用。     

核电厂大破口失水事故始发严重事故的源项研究

本工作以900MW核电厂为研究对象,利用一体化安全分析程序研究大破口失水事故始发严重事故下惰性气体类、挥发类和非挥发类裂变产物释放、迁移特性及分布状况,在此基础上,计算释入环境的源项。结果表明,几乎所有的惰性气体类放射性核素均释入环境,挥发类放射性核素释入环境的份额为10-3数量级,非挥发类放射性核素释入环境的份额为10-6～10-8数量级。计算所得源项可应用于厂外后果评价。     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.     

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.