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.     

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。     

严重事故管理导则入口条件研究

严重事故管理导则的入口是从电厂应急运行规程(EOP)向严重事故管理导则(SAMG)转换的条件,也是严重事故缓解行动的重要依据。本文选取失去四级电源导致的典型高压熔堆序列以及大破口失水事故(LLOCA)导致的典型低压熔堆序列,根据严重事故堆芯剧烈氧化机理,得出了燃料温度、氢气产生速率及产氢量、入口集管过冷度以及慢化剂液位的关系。结果表明入口集管过冷度小于0且持续十几分钟,同时慢化剂系统的状态指示慢化剂液位低于6 900mm,可以作为严重事故管理的入口条件。最后,阐述了目前电厂EOP向SAMG转换的机制,并提出了改进的意见。     

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.     

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。     

出口集管LLOCA始发严重事故分析

采用一体化分析程序建立了包括热传输系统、慢化剂系统、端屏蔽系统、蒸汽发生器二次侧系统的重水堆核电厂的严重事故分析模型。并选取出口集管发生双端剪切断裂的大破口失水事故（LLOCA）,同时叠加低压安注失效,辅助给水强制关闭的严重事故序列进行热工水力分析。由于主热传输系统环路隔离阀的关闭,使得两个环路的热工水力响应过程不同。最终由于低压安注的失效,慢化剂系统逐渐被加热,最终导致堆芯熔化、排管容器蠕变失效。在LLOCA事故序列中叠加向排管容器中注水的缓解措施,可以终止事故进程,使堆芯保持安全、稳定的状态。     

Severe accident development modeling and evaluation for CANDU

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents.Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 °C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.     

Severe accident analysis of a 1300 MWe PWR to address the hydrogen issue

The behaviour of the potentially large quantity of hydrogen generated during a severe accident has been recognised as an issue of importance since the accident at Three Mile Island. In this article, we describe a severe accident analysis for the Neckarwestheim 2 1300 MWe PWR “Konvoi” plant, performed primarily to investigate the behaviour of hydrogen in the containment, and draw conclusions regarding the need for hydrogen control systems (igniters). The Modular Accident Analysis Program (MAAP) developed by IDCOR in the United States, and the Westinghouse COMPACT multi-compartment containment code were used. The study investigated the generation, release to containment, distribution within containment and potential combustion of hydrogen produced during two severe accident sequences. Results are summarized which show that hydrogen mixing in containment is generally good and that even without hydrogen control systems, hydrogen combustion, although possible, does not threaten containment integrity.     

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。     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

A simple novel analysis procedure for IVR calculation in core-molten severe accident

In-Vessel Retention (IVR) of core melt is a key severe accident management strategy adopted by operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000, etc. External Reactor Vessel Cooling (ERVC), which involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris relocated to the vessel low head, is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been proposed to evaluate the safety margin of IVR in AP600 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, a simple novel analysis procedure has been developed for modeling the steady-state endpoint of core melt configurations. Furthermore, IVRASA was developed in a more general fashion so that it is applicable to compute various molten configurations such as UCSB Final Bounding State (FIBS). The results by IVRASA were consistent with those of the UCSB and INEEL. Benchmark calculations of UCSB-assumed FIBS indicate the applicability and accuracy of IVRASA and it could be applied to predict the thermal response of various molten configurations.     

大亚湾核电站全厂断电诱发的严重事故过程研究

在大亚湾核电站严重事故计算分析结果的基础上,对全厂断电诱发的典型的严重事故序列及缓解对策进行了分析。结果表明,全厂断电事故发生后,大约1～2h堆芯上部会裸露,压力容器在5～7h后失效。在约100h安全壳超压失效,而堆坑地基在事故后8.7d会被熔蚀5.5m。结果还表明,堆坑注水措施可以防止堆坑地基熔穿并且减少事故中由于堆芯熔融物与混凝土反应产生的氢气。     

An analysis of containment responses during a station blackout accident

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI and water evaporation. It was found that two-thirds of the total mass of steam and gases in the containment is from an MCCI and one-third of the mass is from water evaporation.     

先进压水堆大破口始发严重事故下安全壳内氢气风险分析

本文采用集总参数法,在先进非能动压水堆核电厂严重事故一体化分析模型基础上,考虑先进压水堆非能动安全特性以及严重事故下采取熔融物堆内滞留(IVR)措施等特性对氢气风险的影响,开展了典型严重事故下安全壳内氢气风险分析。分别选取了冷段双端剪切断裂大破口、冷段大破口叠加IRWST重力注水有效以及ADS-4误启动三个典型大破口失水事故序列,对事故进程中的氧化温度、产氢速率以及产氢质量等特性进行了研究。选取产氢量最大的冷段大破口叠加IRWST重力注水有效事故序列,分析了氢气点火器系统的消氢效果。结果表明,堆芯再淹没过程产生大量氢气,采用点火器可有效去除安全壳内的氢气,从而降低氢气燃爆风险。     

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.     

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.     

Analysis of steam and hydrogen distributions with PAR mitigation in NPP containments

The 3-D-field code, GASFLOW is a joint development of Forschungszentrum Karlsruhe and Los Alamos National Laboratory for the simulation of steam/hydrogen distribution and combustion in complex nuclear reactor containment geometries. GASFLOW gives a solution of the compressible 3-D Navier–Stokes equations and has been validated by analysing experiments that simulate the relevant aspects and integral sequences of such accidents. The 3-D GASFLOW simulations cover significant problem times and define a new state-of-the art in containment simulations that goes beyond the current simulation technique with lumped-parameter models. The newly released and validated version, GASFLOW 2.1 has been applied in mechanistic 3-D analyzes of steam/hydrogen distributions under severe accident conditions with mitigation involving a large number of catalytic recombiners at various locations in two types of PWR containments of German design. This contribution describes the developed 3-D containment models, the applied concept of recombiner positioning, and it discusses the calculated results in relation to the applied source term, which was the same in both containments. The investigated scenario was a hypothetical core melt accident beyond the design limit from a large-break loss of coolant accident (LOCA) at a low release location for steam and hydrogen from a rupture of the surge line to the pressurizer (surge-line LOCA). It covers the in-vessel phase only with 7000 s problem time. The contribution identifies the principal mechanisms that determine the hydrogen mixing in these two containments, and it shows generic differences to similar simulations performed with lumped-parameter codes that represent the containment by control volumes interconnected through 1-D flow paths. The analyzed mitigation concept with catalytic recombiners of the Siemens and NIS type is an effective measure to prevent the formation of burnable mixtures during the ongoing slow deinertization process after the hydrogen release and has recently been applied in backfitting the operational German Konvoi-type PWR plants with passive autocatalytic recombiners (PAR).     

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

严重事故下正常余热排出系统向堆芯注水策略分析

选取导致堆芯熔化频率最高的始发严重事故--直接注入（DVI）管线断裂事故,以及典型高压熔堆事故--丧失主给水始发事故(LOFW),利用MAAP4程序,分析反应堆堆芯热工水力行为,并对正常余热排出系统（RNS）堆芯注水策略的有效性与负面效应进行评估。分析结果表明,在DVI管线断裂事故和LOFW严重事故序列中,利用RNS进行堆芯注水可有效终止堆芯熔化进程,维持堆芯长期冷却。但堆芯再淹没会产生更多的氢气,存在增加安全壳氢气燃烧风险的可能性。此外通过分析利用严重事故管理导则中辅助计算文件给出的堆芯最小流量实施堆芯注水策略,讨论注水流量对堆芯冷却的影响,结果表明,在实施堆芯注水策略时,建议在系统允许的情况下采用更高的流速进行堆芯冷却。     

Formation, characterisation and cooling of debris: Scenario discussion with emphasis on TMI-2

Regarding safety improvements for existing nuclear power plants, the TMI-2 accident is interesting because of the present commercial dominance of light water reactors (LWR). This accident demonstrated that the nuclear safety philosophy evolved over the years has to cover accident sequences involving massive core melt progression in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although the TMI-2 core was reflooded, the results also appear applicable to the general melt progression phenomenology of most unrecovered (unreflooded) blocked core accident scenarios. Nevertheless, a large range in the initial conditions of core melt progression provides significant uncertainties in assessing the integrity of the lower head, the containment in severe reactor accidents, and the consequences of recovery actions in accident management, as well as core reflooding in particular. The probability of success of reflooding as an accident management strategy – in-vessel reflooding to terminate the accident and ex-vessel flooding to prevent reactor vessel melt-through – has to be assessed and discussed in detail.