Exact Solutions of Two Point Time Correlations on Point Reactor Kinetics with Gaussian Reactivity Fluctuations

Station blackout is reported to be a sequence that would likely be a significant contributor to the accident risk at a boiling water reactor (BWR). The occurrence frequency of station blackout is evaluated in probabilistic safety assessment (PSA) to be 6×10^?6 per reactor year at Limerick and less than 10^?7 per reactor year at BWR in Japan.

This report describes an analytical study of thermal-hydraulic and radionuclide behavior during a postulated severe accident of station blackout at a reference BWR plant. The analytical approach was shown in both of hand calculation and the THALES/ART code calculation to better understand wide physical and chemical phenomena in the processes of severe accidents.

We evaluated timing of key events, core cooling and core temperature, reactor vessel failure, debris temperature, containment pressure, and release and deposition of radionuclide in the containment. The THALES and CORCON models on the chemical reactions in the core-concrete interaction lead to great differences in the increasing rate of containment pressure and the release rate of fission products from the core debris.     

Accommodation of potential hydrogen formation in LMFBR accidents

Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus, avoiding the formation of hydrogen. When liner failures are postulated in hypothetical reactor vessel melttrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed.     

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。     

地下核电站安全壳再循环系统设计的初步论证

核电站建造于地下,反应堆厂房洞室外具备天然的裂变产物屏障,在安全壳外洞室内设置安全壳再循环系统,预防并缓解放射性裂变产物释放,维持安全壳的完整性。该系统同时整合了卸压、过滤、排热安全功能,充分发挥地下核电站重力补水和天然屏障的安全优势,可以非能动运行。本文通过简单的计算分析开展初步论证,证明该系统可以有效实现三大安全功能,是适合于地下核电站的安全系统。     

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.     

Basic concept of a near future BWR

This paper shows a basic concept of a near future boiling water reactor (BWR) aiming at evolutional safety and cost savings with minimum change from the current advanced BWR (ABWR). The plant output is uprated to 1500 MWe from 1356 MWe. This power uprate can bring about potential of 11% cost saving per MWe base. Safety improvement as a next generation large reactor is also achieved.

The advanced reinforced concrete containment vessel (ARCCV) is used for the containment vessel to improve safety for severe accidents. The peak pressure of the containment at severe accidents can be kept close to the design pressure. The advanced passive containment cooling system (APCS) is also provided and can accomplish no primary containment vessel (PCV) venting.

The advanced emergency core cooling system (AECCS) consists of four divisions in the front line. The advanced passive cooling system (APCS) is also provided. The combination of the four divisional emergency core cooling system (ECCS) and the passive safety system improves the plant performance in probabilistic safety assessment (PSA).

This plant concept is designed based on the heritage of the current ABWR. No more major research and development (R&D) are necessary. Therefore, construction and operation is possible in the early 2010s.     

AP1000核电厂大量放射性释放源项分析

在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别（安全壳旁通、安全壳早期失效和安全壳隔离失效）,分别选取典型的严重事故序列（蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂）,使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。     

池式钠冷快堆事故源项计算方法研究及其应用

针对传统轻水堆事故源项计算方法不适用池式钠冷快堆的问题,分析可能发生的设计基准事故和超设计基准事故的释放路径,研究建立适用于池式钠冷快堆的堆芯损伤类、泄漏类和钠火类事故源项计算方法。结合示范快堆的6种典型事故：1盒燃料组件瞬时全部堵塞事故、反应堆堆本体覆盖气体边界泄漏事故、一次氩气衰变罐破损事故、主容器泄漏事故、一回路外无保护套管的钠净化管道泄漏事故和一回路无保护套管的外辅助管断裂或泄漏合并隔离阀关不住事故,开展事故源项计算及其剂量后果评价。结果表明：6种事故的放射性后果均低于GB 6249-2011的要求。该方法还可为回路式钠冷快堆、铅铋快堆以及气冷快堆事故源项计算提供参考。     

Study on Calculation Method of Accidental Source Term for Pool-type Sodium-cooled Fast Reactor and Its Application

To deal with the problem that the traditional light water reactor accidental source term calculation method is not suitable for sodium-cooled fast reactor, calculation methods for accidental source term of pool-type sodium-cooled fast reactor, including core damage type, leak type and sodium fire type, were studied and derived on basis of the analysis of release path of potential design basis accidents and beyond design basis accidents. The methods were applied to six typical accidents of the demonstration fast reactor, including the total instantaneous blockage of one fuel assembly, the leakage of cover gas region of reactor main vessel, the damage of primary argon decay tank, the leakage of main vessel, the leakage of sodium purification pipeline without protective sleeve outside the primary circuit, and the leakage of external auxiliary pipeline without protective sleeve outside the primary circuit or the isolation valve tube not be closed. The calculation of accidental source terms and their radiological consequences were carried out. The results show that the radioactive dose consequences of the six accidents are lower than the requirements of GB 6249-2011. The methods proposed can provide reference to the calculations of accidental source term of loop-type sodium-cooled fast reactor, lead-cooled fast reactor and gas-cooled fast reactor.     

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.     

Air ingress phenomena in a depressurization accident of the very-high-temperature reactor

The inherent properties of the very-high-temperature reactor (VHTR) facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However, it is still not clear if the VHTR can maintain a passively safe function during the primary-pipe rupture accident, or what would be a design criterion to guarantee the VHTR with the high degree of passively safe performances during the accident. The primary-pipe rupture accident is one of the most common of accidents related to the basic design regarding the VHTR, which has a potential to cause the destruction of the reactor core by oxidizing in-core graphite structures and to release fission products by oxidizing graphite fuel elements. It is a guillotine type rupture of the double coaxial pipe at the nozzle part connecting to the side or bottom of the reactor pressure vessel, which is a peculiar accident for the VHTR. If a primary pipe ruptures, air will be entered into the reactor if there is air in the reactor containment or confinement vessels. This study is to investigate the air ingress phenomena and to develop the passively safe technology for the prevention of air ingress and of graphite corrosion. The present paper describes the influences of a localized natural circulation in parallel channels onto the air ingress process during the primary-pipe rupture accident of the VHTR.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

AP1000与二代压水堆核电厂的严重事故预防与缓解策略比较

根据压水堆核电厂严重事故发生机理,基于高压堆熔、压力容器失效以及安全壳失效三个关键阶段,针对AP1000和二代核电厂进行比较,在系统结构设计上分析两者在严重事故预防与缓解策略方面的异同,最后对我国在役核电厂的严重事故预防与缓解提出建议。     

AP1000与二代压水堆核电厂的严重事故预防与缓解策略比较

根据压水堆核电厂严重事故发生机理,基于高压堆熔、压力容器失效以及安全壳失效三个关键阶段,针对AP1000和二代核电厂进行比较,在系统结构设计上分析两者在严重事故预防与缓解策略方面的异同,最后对我国在役核电厂的严重事故预防与缓解提出建议。     

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.     

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.     

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.     

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.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.