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.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

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。     

PCS表面液膜覆盖率对安全壳完整性影响分析

钢制安全壳是防止严重事故工况下放射性物质向环境释放的最后一道屏障,因此有必要研究分析事故条件下安全壳外液膜覆盖率对安全壳完整性影响,以得到安全壳在事故工况下的失效裕度。应用非能动安全壳分析程序,建立了大功率非能动反应堆非能动安全壳冷却系统(Passive Containment Cooling System,PCS)的热工水力模型,并以冷段双端剪切事故为基准研究对象,分别研究了水分配器单一故障和出水管堵管叠加水分配器故障两种事故工况。分析结果表明,两种事故工况在液膜覆盖率大于35%时,均不会出现短期安全壳超压超温失效;事故后24 h,液膜覆盖率低于45%时,安全壳出现长期冷却失效。此次研究得出结论:在流量大于61.76 m3·h-1、安全壳液膜覆盖率大于45%时,事故发生后24 h安全壳不会失效。     

A study on evaluating a passive autocatalytic recombiner PAR-system in the PWR large-dry containment

A systematic step-by-step framework for analyzing hydrogen behavior and implementing passive autocatalytic recombiners (PARs) to mitigate hydrogen deflagration or detonation risk in severe accidents (SAs) is presented. The procedure can be subdivided into five main steps: (1) modeling the containment based on the plant design characteristics, (2) selecting the typical severe accident sequences, (3) calculating the hydrogen generation including in- and ex-vessel period, (4) modeling the gas distribution in containment atmosphere and estimating the hydrogen combustion modes and (5) evaluating the efficiency of the PAR-system to mitigate the hydrogen risk with and without catalytic recombiners, according to the safety criterion. For the Chinese 600MWe pressurized water reactor (PWR) with a large-dry containment, large break loss-of-coolant accident (LB-LOCA) is screened out as the reference severe accident sequence, considering the nature of hydrogen generation and the probabilistic safety assessment (PSA) result on accident sequences. The results show that a certain number of recombiners could remove effectively hydrogen and oxygen, to protect the containment integrity against hydrogen deflagration or detonation.     

先进压水堆核电厂氢气控制策略分析研究

新建核电厂的设计必须做到“实际消除”早期与大量放射性释放的可能性,氢气燃爆导致的安全壳失效是必须要“实际消除”的严重事故工况之一。因此对各种消氢措施的特点进行分析研究,建立联合消氢策略评价方法,可为先进压水堆核电厂氢气控制策略选择设计评价提供支持手段。根据严重事故管理中对氢气控制策略的考虑,研究安全壳内局部位置的可燃性是相关设计评价的关键问题。根据可燃性准则、火焰加速准则、燃爆转变准则,本文使用三维CFD程序对典型严重事故工况下安全壳蒸汽发生器隔间内的可燃性及氢气风险进行模拟分析。研究结果表明,虽然喷放源项中有大量水蒸气,蒸汽发生器隔间中仍有较大区域处于可燃限值以内,合理布置的点火器能在设计中点燃并消除氢气。本研究建立的分析方法能用于对核电厂氢气控制策略选择设计的评价。     

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。     

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

Simulation of hydrogen deflagration and detonation in a BWR reactor building

A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm² leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis.     

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.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

Pressure release of containments during severe accidents in Switzerland

As required by the Swiss Federal Nuclear Safety Inspectorate (HSK) all Switzerland's five nuclear power plants have to install a containment filtered venting system. The integrity of the containment (the last barrier for radioactive releases to the environment) can be threatened by overpressure due to inadequate heat removal. Design requirements have been provided for a specific class of severe accident scenarios. In general the capacity of the system is considered sufficient if it is able to vent the steam production corresponding to a decay heat level of 1% of the thermal reactor power. The mitigation capacity for the reduction of released radioactive material is specified by a retention factor of 1000 for aerosols to prevent or limit a long term ground contamination and a factor of 100 for elementary iodine for prevention or limiting of thyroid doses and to avoid short term evacuation. Besides existing requirements for design, maintenance and operation, additional claims such as passivity and operability at any pressure conditions inside the containment have to be met. Passivity implies that the system can be initiated after a severe accident without any operator action. The system also has to allow early manual venting. Various filtered venting systems are presently available. The nuclear power plants of Beznau, Gosgen, Leibstadt and Muhleberg have already selected such systems and already implemented them or are going to install them step by step. Beznau selected the Sulzer-EWI system which is using a water pool with nozzles-baffle plates and mixing elements to achieve the required filtration of the aerosols. In both Beznau units, the systems are installed and in standby mode. Gosgen, a pressurized water reactor as well as Beznau, is going to implement a filter system developed by Siemens-KWU, known as sliding pressure venting process, combining a venturi scrubber in a water pool and a mesh filter. The boiling water reactor of Leibstadt also selected the same system as Beznau while Müheberg choose the ABB system but not in the common design. The venturi pipes are thereby integrated in the water pool of the outer torus. The system in all five nuclear power plants is fully operable and in standby mode since December 1993.     

Effects of obstacle position and number on the overpressure of hydrogen combustion in a semi-confined compartment

During a severe accident in a pressurized water reactor, large quantities of hydrogen can be produced and released into the containment. The hydrogen mixed with air may be ignited. Rapid pressure rise is likely to occur during hydrogen combustion in a semi-confined compartment because the length–diameter ratio of the compartment is relatively large and there generally are some devices and components in the compartment. Obstacle position and number have a significant and complex effect on combustion pressure. In this paper, a numerical analysis of the effects of obstacle position and number on the overpressure of hydrogen combustion in a semi-confined compartment was carried out to identify the arrangement of obstacles that can decrease the pressure loads generated by hydrogen combustion. The overpressure in the compartment with a single obstacle was found first increases to a maximum value and then decreases as the obstacle moving from the closed end to the open end. When there are multiple obstacles existing in the compartment, minimize the obstacle spacing and number is likely to reduce the maximum overpressure of hydrogen combustion. In addition, intentionally adjust the length of obstacle array may help for reducing the maximum overpressure.     

Simulation of containment thermal-hydraulics in the Marviken Blowdown 16 experiment with ASTEC and CONTAIN codes

The boiling heavy water reactor Blowdown 16 experiment, which was performed in the Marviken experimental facility, was simulated with the ASTEC and CONTAIN codes. The main purpose of the work was the assessment of the codes for simulating thermal-hydraulic phenomena in a BWR containment at accident conditions. Simulated pressures, atmosphere temperatures and wetwell pool masses are compared to experimental measurements. The results show that both codes satisfactorily reproduced the overall containment thermal-hydraulic behaviour. The simulations also allow a more detailed understanding of the governing mechanisms during the performed experiment.     

Evaluation of uncertainties in risk emanating from a number of severe accident issues

A study has been performed to estimate, for a particular pressurized water reactor, the uncertainty in risk associated with a number of key phenomenological issues. A second objective was to distinguish the individual importance of the various issues as contributors to the overall uncertainty in risk. The issues considered touched upon the areas of system behavior, containment loading, containment performance, and fission product source term behavior. It was found that the most important source of uncertainty for the plant in question (Surry) was direct containment heating (i.e., the transfer of heat from the core debris to the containment atmosphere when the debris is ejected at high pressure from the reactor vessel and dispersed throughout the atmosphere). Other significant issues included hydrogen burning, containment failure pressure, aerosol agglomeration uncertainties, the frequency of check valve failures leading to a loss-of-coolant accident (LOCA) outside containment, and the potential for having a LOCA induced by high temperatures in the reactor coolant system.     

严重事故管理导则中的氢气风险判断及事故模拟分析

建立严重事故管理导则中用于判断氢气燃烧、超压风险以及安全壳降压时氢气风险的判断工具。用一体化事故分析程序对全厂断电事故进行模拟计算,用该氢气风险判断工具对不同事故阶段的氢气风险进行分析。结果表明:在全厂断电始发的严重事故下,没有氢气复合器且没有安全壳喷淋时,安全壳大气在一段时间内会被水蒸气惰化,不会发生燃烧,但如果应急电源恢复,重新启动安全壳喷淋时,有可能引起氢气燃烧甚至造成安全壳超压;在增加氢气复合器后,没有造成安全壳超压的风险,并且判断结果是保守的。     

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.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

严重事故下安全壳排气对乏燃料厂房氢气风险影响研究

以我国某三代压水堆核电厂为例,选取了2个典型严重事故工况,采用严重事故一体化程序MAAP开展建模与计算,对安全壳排气的过程及对乏燃料厂房造成的氢气风险进行了分析。结果表明,如果不考虑乏燃料厂房的通风系统,从安全壳内释放的混合气体由于水蒸气的冷凝,会对乏燃料厂房造成一定的氢气风险;如果考虑乏燃料厂房通风系统的作用,乏燃料厂房的氢气风险将会消除。