An approach to the construction of a one top fire event PSA model

An internal fire event probabilistic safety assessment (PSA) model has been generally quantified by modifications of a pre-developed internal events PSA model. New accident sequence logics not covered in the internal events PSA model are separately developed to incorporate them into the fire PSA model. Previous studies on the changes of the one top internal event PSA model for the one top fire event PSA model have been limited to the equipment failures affected by the fire. In addition, they assumed that the probabilities of basic events associated with equipment or cables impacted by the fire are one. However, the probabilities of spurious operation events and human failure events affected by the fire might not be estimated as one. In this study, new modification rules were proposed for the construction of a one top PSA model for fire events by using a one top internal event PSA model. The proposed new modification rules can be applied to all the fire damage events for the fire-induced equipment failure events and spurious operation events, human error events impacted by a fire, regardless of whether they are estimated as one or not. Applications of the proposed modification rules to the compartment and scenario-level fires for the hypothetical plants were performed for demonstrating their appropriateness to the changes of the one top internal event PSA model to the one top fire event PSA model. In addition, quantification procedure with the one top fire event PSA model was presented and discussed.     

Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code_FDS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O₂ depletion, and CO and CO₂ production, etc. Except the amount of O₂ consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H_frac) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H_frac value has a prominent effect on the amount of O₂ consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.     

Insights from fire PSA for enhancing NPP safety

This paper presents the findings of an effort to gain insights from fire probabilistic safety assessment (FPSA) conducted in nuclear power plants. Using probabilistic models, the fire PSA takes into account the possibility of a fire at specific plant locations and its propagation, detection and suppression of the fire; and also helps to assess the effect of the fire on safety-related cables and equipment. The results of FPSA contributed to design modifications in plant to enhance the safety and thereby reduce its contribution to core damage frequency. It also highlights the sources of uncertainty while conducting and suggesting values of risk parameter in FPSA study.     

Study on construction of loss of component cooling water system initiating event fault tree for fire event PSA

ABSTRACT

In this study, the construction of the loss of component cooling water system (LOCCWS) initiating event (IE) fault tree (FT) for an actual fire event probabilistic safety assessment (PSA) model of the Korean reference nuclear power plant considering only IE initiators was validated. The quantification results of the LOCCWS accident sequences obtained using an LOCCWS IE FT model with only initiators are similar to that with initiators and enabling events. This confirmed that the LOCCWS IE FT for an actual fire event PSA model could be constructed by considering only IE initiators. In addition, the same LOCCWS accident sequences were quantified assuming that fire triggering only the LOCCWS IE leads to reactor shutdown. Compared with the quantification result obtained based on the assumption that any fire included in the fire event PSA leads to reactor shutdown, the core damage frequency (CDF) can be reduced. Thus, it can be concluded that there is a possibility of underestimation of CDF when the LOCCWS IE FT model with only initiators is used and the assumption that fire triggering only the LOCCWS IE results in reactor shutdown is employed for the quantification of LOCCWS accident sequences.     

核电厂火灾PSA中火灾场景频率分析方法

核电厂的火灾场景频率分析是火灾概率安全分析的核心内容。本文根据美国NRC和EPRI的《核动力设施火灾概率风险评价方法》,介绍了核电厂内部火灾概率安全评价中,火灾场景频率的分析方法及参数不确定性的处理方法。以福清一期核电厂某房间的电气柜火灾为例,进行了定量分析计算,计算结果表明,在计算中考虑热释放速率参数不确定性的传递可以有效降低计算结果的保守性。     

核电厂运行阶段PSA模型开发的研究

电厂运行阶段的概率安全分析工作通过建立反映电厂实际设计及运行特点的PSA模型,可以定性及定量评价电厂运行阶段的安全性,帮助电厂寻找设计及运行中的薄弱环节,为电厂管理提升及后续技术改造提供技术支持和见解。而且,运行阶段的PSA模型也是电厂开展一系列PSA应用工作的基础。本文首先总结运行电厂的特点及运行阶段PSA模型开发的主要关注事项,并结合秦山第二核电厂运行阶段的PSA模型开发给出电厂运行阶段PSA的技术路线、主要分析结果、分析见解及改进建议,为后续相似工作的开展提供参考和建议。     

核电厂火灾PSA方法浅析

作为PSA工作中不可缺少的一部分,核电厂火灾PSA正在发挥着越来越重要的作用。本文对核电厂火灾PSA的发展、应用和研究的基本情况进行了论述。     

Safety Margins of Spent Fuel Transport and Storage Casks Considering Aircraft Crash Impacts

Abstract

The safety of spent fuel transport casks in severe accident conditions is always a matter of concern. This paper surveys German missile impact tests that have been carried out in the past to demonstrate that German cask designs for transport and interim storage are safe even under conditions of an aircraft crash impact. A fire test with a cask beside an exploding propane vessel and temperature calculations concerning prolonged fires also show that the casks have reasonably good safety margins in thermal accidents beyond regulatory fire test conditions.     

HFETR一级PSA始发事件分析

在高通量工程试验堆（HFETR）一级概率安全分析（PSA）中,始发事件分析是首要任务。首先综合应用了工程评价、参考以往的始发事件清单、演绎分析和运行经验总结等方法,确定了HFETR运行阶段一级PSA始发事件清单,然后对始发事件进行适当的归并分组,最后结合故障树分析、HFETR运行事件统计及参照国内外相同类型研究堆等方法,给出了各始发事件组的频率,为后续开展HFETR一级PSA奠定了基础。      

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。     

Formal qualification of the various sources of an uncertainty employed in the level 2 PSA and the underlying issues

In the Level 2 PSA, a probabilistic treatment of complex phenomenological accident pathways inevitably leads to two sources of an uncertainty: (a) an incomplete modeling of these accident pathways and their subsequent impacts on the Level 2 risk, and (b) an expert-to-expert variation in the their probabilistic estimates. The impacts of the foregoing sources of an uncertainty on the Level 2 risk measures are different from each other, thus leading to a different conclusion in the decision-making process. An important aspect of the foregoing distinction of an uncertainty is that it plays an essential role in identifying what sources of an uncertainty impact more on the Level 2 risk and what sources among them should be focused on for a greater reduction of the overall Level 2 uncertainty. Another aspect is closely related to its importance in combining the Levels 1 and 2 uncertainties. A primary objective of this paper is to qualify formally the typical sources of an uncertainty that would often be employed in the Level 2 PSA and the underlying issues for a further clarification.     

PSA as a tool for decision making

The question on “How safe is safe enough?” is being responded presently by deterministic criteria. Probabilistic criteria in support to more rational and less emotional decisions in regulatory and licensing issues, rationalization of resource allocation and research prioritization, among others, have a potential which is only marginally being explored.This paper discussed PSA limitations and proposes three areas for the use of PSA in decision making, namely:

1. (a) preventing accidents,

2. (b) mitigating accidents, and

3. (c) defining regulatory requirements.

Current activities of the International Atomic Energy Agency in these areas are mentioned.PSA studies depict clearly the uncertainties and this is viewed as a positive aspect, which is unique to the use of probabilistic methods.     

Development of probabilistic models to estimate fire-induced cable damage at nuclear power plants

Numerous Probabilistic Risk Assessments (PRAs) have shown that fire is a major contributor to Nuclear Power Plant (NPP) risks. However, prediction and estimation of the likelihood of fire-induced damage to electrical cables and circuits and their potential effects on the safety of the NPPs are still a practical challenge, particularly because of the lack of physics-based models with which to perform consistent and objective assessments.This paper contains a discussion of two models - the heat transfer and the IR “K-factor” models - to estimate the likelihood of fire-induced cable damage given a specified fire profile. The results of this research will help to (1) develop a consistent framework to estimate the likelihood of fire-induced cable failure modes, and (2) develop some guidance to evaluate and/or reduce the risks associated with these failure modes in existing and new NPPs.The models are developed (i.e., their parameters are estimated) based on the test data from various fire damage tests sponsored by the nuclear industry and the U.S. Nuclear Regulatory Commission (NRC). Among the models evaluated, the physics-based heat transfer model is promising because it takes into account the properties and characteristics of the cables and cable materials and the characteristics of the thermal insult. This model can be used to estimate the probability of cable damage (P_CD) under different thermal conditions.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Modular HTR confinement/containment and the protection against aircraft crash

Does an HTR need a containment – pressure resistant – or is it possible – licensable – to have only a so-called confinement.The answer depends on both the results of the safety analysis of the accidents considered in the design and the acceptance by the licensing authorities and the public of a safety approach only based on severe core damage exclusion.The safety approach to be developed for modular HTRs must describe the application of the defence in depth principle for such reactors. Whatever the requirements on the last confinement barrier could be, a convincing demonstration of the exclusion of any severe core damage is needed, relying on exhaustive and bounding considerations of severe core damage initiators and the use of non-questionable arguments.The paper presents the containment issues for HTRs based on German experience background and considerations for modern modular HTR safety approach including beyond design situations.

• For the German HTRs (designed in the 80s), it could be shown in the licensing procedures in Germany that there was no need for a pressure retaining and gas tight containment to enclose radioactive nuclides released from the nuclear heat source. Instead, the confinement envelope acted in conjunction with other barriers to minimize the release of radioactive nuclides and the radiological impact on the environment.

• The confinement envelope consisted of the reactor building, a sub-atmospheric pressure system, a building pressure relief system, an HVAC systems isolation and a filtration system.

• During a major depressurization accident, unfiltered releases were discharged to the environment. The analyses results show that the environmental impact was far below the dose limits according to the German Radiological Protection Ordinance, even when the effect of filters was not taken into account.

• The demonstration strongly relied on the assumptions made for the source term definition, e.g. the fuel particles failure rates (under irradiation and during accidental conditions), the diffusion data, the dust data and the deposition/lift-off mechanisms.

• For modern modular HTRs, the last confinement barrier performances will have to be determined in accordance with the set of accidents to be considered in the design including internal and external hazards and the limits targeted for the public and the environment protection.

Further more the paper presents an analysis of effects of a deliberate crash of a large commercial airliner on a former German HTR design.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

液态金属反应堆湿式燃料贮存桶内部事件概率安全分析

液态金属冷却剂在给反应堆带来运行安全与热效率优势的同时,也给反应堆带来了复杂的换料系统,其中大型液态金属反应堆采用的湿式乏燃料贮存桶是乏燃料卸料过程的核心设备,临时装载了大量的乏燃料组件,具备一定的安全风险。本文采用概率安全分析（PSA）方法对乏燃料贮存桶进行风险评价,通过运行状态分析、始发事件分析、事故序列分析以及简单的定量化,初步获得其导致乏燃料组件发生损伤的事故序列和最小割集,识别了关键系统与设备。结果表明,相对于反应堆本身的风险,乏燃料贮存桶本身风险虽低但依然不可忽略,且风险评价结果对反应堆的运行方式以及清洗系统的可靠性较为敏感。此外还对该系统的设计改进与安全优化进行了讨论。     

Methodology and results of the seismic probabilistic safety assessment of Krško Nuclear Power Plant

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Kr

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ko plant. The methodology adopted is the seismic PSA (probabilistic safety assessment). The Kr

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ko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of site hazard, calculation of plant structures response including soil–structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Relay chatter analysis and soil stability studies were also performed. The seismic PSA described here is limited to the analysis of CDF (level 1 PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Kr

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ko NPP but are not further described in this paper. The results of the seismic PSA study indicates that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable with most US and Western Europe NPPs located in high seismic areas.