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.     

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.     

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...     

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.     

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。     

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。      

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.     

压水堆核电厂高压熔堆严重事故序列分析

压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析（PSA）结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明：5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。     

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.     

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。     

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。     

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。      

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。     

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.     

ACME台架PRHR管线破口位置敏感性试验研究

为研究先进非能动（AP）型核电厂在非能动系统失效条件下的安全性能,利用我国先进堆芯冷却机理整体试验台架（ACME）开展了非能动余热排出（PRHR）管线破口失水试验研究,分析了主要的试验进程和破口位置对事故过程各阶段关键参数的影响。结果表明,ACME PRHR管线破口试验进程与冷管段小破口失水事故（SBLOCA）进程基本一致,再现了非能动核电厂自然循环阶段、自动卸压系统（ADS）喷放阶段和安全壳内置换料水箱（IRWST）安注阶段的安全特性;在不同破口位置的试验中,非能动堆芯冷却系统（PXS）均可保证堆芯得到补水,堆芯活性区始终处于混合液位以下;破口位置对ACME LOCA事故进程、反应堆冷却剂系统（RCS）初期降压速率、PRHR热交换器（HX）流量、喷放流量、堆芯液位、IRWST安注流量等参数具有显著影响,对堆芯补水箱（CMT）和蓄压安注箱（ACC）安注流量的影响较小。      

增强IVR有效性的堆内注水策略研究

熔融物堆内滞留（IVR）是一项核电厂重要的严重事故管理措施,通过将熔融物滞留在压力容器内,以保证压力容器完整性,并防止某些可能危及安全壳完整性的堆外现象。对于高功率和熔池中金属量相对不足的反应堆,若下封头形成3层熔池结构,则其顶部薄金属层导致的聚焦效应可能对压力容器完整性带来更大的威胁。本文考虑通过破口倒灌及其他工程措施实现严重事故下熔池顶部水冷却,建立熔池传热模型,分析顶部注水的带热能力,建立事件树,分析顶部注水措施的成功概率及IVR的有效性。结果表明,通过压力容器内外同时水冷熔融物,能显著增强IVR措施的有效性。     

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.