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.     

Severe accident mitigation strategy for the generation II PWRs in France – Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

The 3rd Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured.Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications.Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define.Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others.This paper presents the analysis conducted by IRSN during the 3rd periodic safety review of the French 1300 MWe PWRs. Future NPP upgrades to limit radioactive releases in case of containment filtered venting, to prevent containment venting and basemat melt-through are analysed in another framework (post-Fukushima and long-term operation projects).     

Accident management strategy in Sweden — implementation and verification

A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management.     

The role of BWR secondary containments in severe accident mitigation: Issues and insights from recent analyses

All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by secondary containments. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident.The fission product retention capability of an intact secondary containment will depend on several factors. Recent analyses indicate that the major factors influencing secondary containment effectiveness include: the mode and location of the primary containment failure, the internal architectural design of the secondary containment, the design of the standby gas treatment system, and the ability of fire protection system sprays to remove suspended aerosols from the the secondary containment atmosphere. Each of these factors interact in a very complex manner to determine secondary containment severe accident mitigation performance.This paper presents a brief overview of US BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented.     

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it.Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.     

严重事故氢气燃爆缓解措施的初步研究

轻水堆核电站发生严重事故时,氢气的大体积氢燃爆可能会严重威胁安全壳的完整性.氢气点火器与氢气复合器是2种严重事故下的氢气燃爆缓解设备.本文分别研究了3种氢气燃爆缓解措施,包括仅采用氢气点火器、仅采用氢气复合器和采用氢气复合器结合点火器.结果表明,采用氢气复合器结合点火器的方式可以安全、持续、有效地降低大体积氢燃爆带来的风险.     

大型干式安全壳消氢系统的初步设计

以岭澳核电站为分析对象,利用MELCOR和TONUS(CEA)程序进行分析计算,给出了初步的消氢系统设计方案,对不同核电站的消氢系统设计方案进行了对比和讨论.结果表明:安全壳内安装33个FR750型或者17个左右的FR1500型氢气复合器可以满足氢气控制要求.     

双堆布置核电厂公用设施对双堆超设计基准事故缓解的影响和改进

福岛事故后,同一厂址多台机组同时发生超设计基准事故(包括严重事故)的后果开始受到关注,为此需要从设计上保证核电厂事故应对措施的独立性。我国运行和在建的大部分核电厂为双堆布置的二代改进型核电厂。分析表明,水压试验泵和安全壳过滤排放系统(EUF)为双堆公用,对双堆超设计基准事故的应对能力存在影响。进而研究了这两个公用设施在现有电厂中的潜在改进选项,从尽量减少硬件改动的目的出发提出了最可能的改进方案。其中EUF交替排放仅仅通过操作规程的变化,凭借一套公用系统即可实现双堆的卸压目的。进一步计算也证明,合理选取交替排放的时间窗口,EUF交替排放在最保守及最现实的事故情况下均能确保双堆安全壳的安全。     

Strategy evaluation for fire spray system on Advanced Passive PWR Severe Accident Management Guideline

The fire spray system (FSS) of the Advanced Passive PWR, as a part of the fire protection system, can provide a non-safety related containment spraying function for severe accident mitigation which is included in the Severe Accident Management Guidelines (SAMG) of the Advanced Passive PWR when dealing with severe accidents. The effectiveness of the FSS is investigated on three effects for severe accident mitigation which are controlling the containment condition, washing out fission product and injecting into the containment through three representative severe accident scenarios analysis with integral accident analysis code since there is no sufficient data support, besides the negative impact is also discussed. Results show that the FSS can be effective for controlling the containment condition, washing out fission product and injecting into the containment, however the effect is limited due to system limitation: the FSS can only cool the containment atmosphere for a short term; the flow rate of FSS cannot fulfill the success criteria given in the PRA report of the Advanced Passive PWR. Meanwhile, the hydrogen concentration and the containment water level should be the long-term monitored because actuating the FSS may cause hydrogen risk in the containment and containment flooding. Despite its limitation and negative impact, the FSS can be effective as an alternative severe accident mitigation measurement for postponing the process of accidents for safety system recovery.     

核电厂严重事故下安全壳通风导致放射性后果的快速评价

提出一种严重事故下安全壳通风导致放射性后果的快速评价方法。通过预先计算通风后安全壳的释放份额和1%初始堆芯总量释入安全壳时的公众个人终身剂量,以及通过事故下安全壳的辐射监测仪表间接得到堆芯向安全壳的释放份额,能够快速评价厂外不同距离处公众的个人终身剂量,它可为严重事故的管理和厂外应急策略的实施提供强有力的支持。     

Severe accident mitigation programme for the forsmark and ringhals plants

In February 1986 licensing requirements regarding severe accidents in nuclear power plants were given by the Swedish Government. This regulation constitutes conditions for operation of the plants beyond 1988. The requirements are based on the conditions previously given for the Barsebäck plant including construction of the filtered venting system, which was completed at Barsebäck in 1985.For the Forsmark and Ringhals plants a strategy is being implemented to meet the new requirements. A strong emphasis is put on both hardware and procedural measures to bring the reactor core back to stable cooling - even if it is severely damaged - and maintain the containment integrity during an accident. The hardware modifications include measures to prevent temperature or pressure induced early containment failure for the BWRs, reliable back-up water sources for containment spray and means for filtered venting of all plants to prevent late containment failure by overpressure. The ultimate aim is to minimize the environmental impact of a severe accident and meet a release limit set at 0.1% of the core fission product inventory excluding noble gases.     

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。     

百万千瓦级压水堆严重事故下局部隔间氢气风险分析

核安全法规要求控制严重事故下核电厂安全壳内的氢气浓度。除安全壳整体外,局部隔间的氢气浓度同样是关注的重点。本文采用一体化严重事故分析程序对百万千瓦级压水堆核电厂安全壳局部隔间进行建模,分析了不同事故下的氢气风险。结果表明,严重事故下部分隔间短时间内可能存在燃烧风险。本文对降低燃烧风险的方法进行分析计算和筛选,得出的结论可以为安全壳隔间的设计优化提供参考依据。     

Analysis of strategies for containment venting in case of severe accidents

During a core melt accident, a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. This failure mode is expected to be the most likely one for large dry containments under accident conditions. Also during a core melt accident, a large quantity of hydrogen may be generated, giving the potential of a loss of containment integrity due to violent hydrogen combustion. Timely venting of the containment atmosphere can prevent overpressurization and may perhaps make the hydrogen situation in the containment less severe. This paper discusses the thermodynamic consequences of different vent strategies for a large German PWR during core melt accidents.     

先进非能动压水堆防火喷淋对严重事故的缓解作用研究

依据先进非能动压水堆的严重事故管理导则(SAMG),消防系统中的防火喷淋系统,尽管属于非安全相关的系统,仍可以作为严重事故缓解策略,在以下三个方面起到严重事故缓解的作用:减少放射性气溶胶的质量;安全壳降温降压;安全壳注水。因此本文利用一体化严重事故分析程序,选取典型事故序列,评估防火喷淋系统在严重事故中的三种缓解作用的有效性为防火喷淋在严重事故管理导则中的应用提供技术支持。分析结果表明,防火喷淋系统能够实现堆腔淹没,在一定时间内进行安全壳降压,以及减少安全壳中放射性气溶胶的含量的作用,但由于系统限制,防火喷淋进行堆腔淹没的流量不能满足安全限值,并且只能推迟而不能够避免安全壳的失效。防火喷淋系统对严重事故的缓解作用虽然是有限的,但可为其他相关系统或设备的修复提供一定时间。     

Risk reduction by filtered venting in PWR large dry-containments

The potential risk reduction associated with a Filtered-Vented Containment System is evaluated. A low-volume venting strategy has been considered and data referring to the Zion power plant, along with the results of the Zion Probabilistic Safety Study, have been used. An estimate of the reduction factor is first made for a single core melt accident sequence whose containment failure mode is late overpressure. The result, interpreted as a reduction factor applicable to the release category associated with containment late overpressure is then used for the estimation of the overall risk reduction factor. In particular, the case of internal and external risk for the Zion power plant are considered. Because the contribution from seismic events dominates the overall risk, the importance of different assumptions for seismic fragility is also assessed.Finally an uncertainty analysis of the risk reduction factor for a single accident sequence is performed. An estimate is also obtained on the level of confidence with which certain required values of risk reduction can be achieved.     

EPR accident scenarios and provisions

For the EPR an improved defence-in-depth concept is applied. In an evolutionary way, accident control is developed from existing French and German PWR designs, thereby achieving a high safety level quantified by probabilistic safety assessment. Independent of that, severe accidents are considered in the design. By a robust containment and severe accident mitigation measures, the need for offsite emergency response actions (population evacuation or relocation) is restricted to the immediate plant vicinity. This paper gives an overview of the features introduced for, and the analyses correlated to, the dedicated primary depressurization, melt–coolant interaction, melt stabilization, hydrogen control, and containment heat removal.     

AP1000小破口叠加重力注射失效严重事故分析

应用新版MELCOR程序,建立了AP1000一二回路、非能动安全系统及安全壳隔室的热工水力模型,并以热段小破口叠加重力注射系统失效事故为例,对该严重事故进程在压力容器内阶段进行模拟计算,对缓解措施的功能进行了分析和评价。结果表明：自动卸压系统（ADS1～4）的成功实施,可使来自堆芯补水箱和安注箱的冷却水快速有效地注入堆芯,在冷却水完全耗尽前,堆芯始终处于淹没的状态。ADS4爆破阀开启后,使回路压力快速与安全壳压力平衡;非能动安全壳冷却系统对抵御严重事故下由于衰变热和非冷凝气体带来的缓慢升温升压是行之有效的措施;点火器在氢气浓度较低时点火,缓解了安全壳大空间发生全局燃爆而引发安全壳超压失效的风险,但连续点火燃烧会引起局部隔室温升远超出设计温度而危及后备缓解设施的存活。     

An evaluation of a general venting strategy in a CANDU 6 reactor building

If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the containment filtered vent system is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D₂O vapour recovery system which has a 0.76 m (30 in.) diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a Small LOCA and a Large LOCA. The reactor building pressure behavior was analyzed with the ISAAC computer code for four different cases: without venting, 379 kPa(g)/345 kPa(g) (55 psig/50 psig), 345 kPa(g)/276 kPa(g) (50 psig/40 psig) and 345 kPa(g)/207 kPa(g) (50 psig/30 psig) valve open/close pressures. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D₂O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 379 kPa(g)/345 kPa(g) (55 psig/50 psig) is safe and effective. Based on the current study, the strategy of reactor building venting is involved in severe accident management guidance-5.     

Containment venting as an accident management strategy for BWRS with Mark I containments

A quantitative evaluation of primary containment venting was performed to assess its risk reduction potential. A boiling water reactor with a Mark I containment was evaluated by developing simplified containment event trees for its risk dominant sequences. Risk results were benchmarked with those from the NUREG-1150 risk rebaselining effort, and sensitivity studies then were performed. It was found that for station blackout sequences, containment venting by itself does not significantly reduce overall risk. For sequences involving loss of long-term decay heat removal or failure to scram, however, venting is potentially an important mechanism in preventing or delaying core melting. Subsequent studies show that when venting is combined with other potential containment improvements, there is a large potential for risk reduction.