Level-1 Probabilistic Safety Assessment Research Articles

Risk reduction by means of FLEX strategies in pressurized water reactors

One of the key lessons learned from the Fukushima Dai-ichi accident was the importance of the challenge presented by the loss of safety-related systems after a Beyond Design-Basis External Event (BDBEE). In particular, Extended Loss of AC Power (ELAP) or Loss of Ultimate Heat Sink (LUHS) can severely compromise key safety functions associated with core cooling and containment, ultimately leading to reactor core damage. Because of that the so-called Diverse and Flexible Coping Strategies (FLEX) were designed.These aforementioned FLEX strategies are also useful to avoid core damage in sequences due to internal events. Therefore, these FLEX strategies should be incorporated in the Level 1 Probabilistic Safety Assessment (L1PSA) to analyse their impact on core damage frequency (CDF). This work quantifies the impact of FLEX strategies by using a generic model (3 loops, PWR-WEC) developed by the Universidad Politécnica de Madrid (UPM) in collaboration with the Spanish Nuclear Regulatory Body (CSN). As a result, a decrease of the CDF at power for internal events is observed and, therefore, indicates that these FLEX strategies have a positive impact on the plant safety not only for BDBEE but also for other accidental conditions.

Overcoming the uncertainty in a research reactor LOCA in level-1 PSA; Fuzzy based fault-tree/event-tree analysis

Probabilistic safety assessment (PSA) which plays a crucial role in risk evaluation is a quantitative approach intended to demonstrate how a nuclear reactor meets the safety margins as part of the licensing process. Despite PSA merits, some shortcomings associated with the final results exist. Conventional PSA uses crisp values to represent the failure probabilities of basic events. This causes a high level of uncertainty due to the inherent imprecision and vagueness of failure input data. In this paper, to tackle this imperfection, a fuzzy approach is employed with fault tree analysis and event tree analysis. Thus, instead of using the crisp values, a set of fuzzy numbers is applied as failure probabilities of basic events. Hence, in the fault tree and event tree analysis, the top events and the end-states frequencies are treated as fuzzy numbers. By introducing some fuzzy importance measures the critical components which contribute maximum to the system failure and total uncertainty are identified. As a practical example, under redesign Iranian heavy water research reactor loss of coolant accident is studied. The results show that the reactor protection system has the largest index in sequences lead to a core meltdown. In addition, the emergency core cooling system has a main role in preventing abnormal conditions.

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty. The probability for steam generator tube creep rupture is estimated at 0.17.

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these parameters upon response of building was investigated. As a result, it was confirmed that nonlinear properties of rubber bearings have influence on response of the building and inner equipment.

Assessment methodology for confidence in safety margin for large break loss of coolant accident sequences

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with calculating the safety margin with 95% confidence and 95% probability in large break loss of coolant accident (LBLOCA). In the present study, five uncertain input parameters were selected. Uniform probability density function was assigned to the uncertain parameters in the selected range and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for five parameters was randomly mixed by LHS to obtain 25 input sets. An event tree for the initiating event, LBLOCA inside containment, has been used from a VVER study on Level-1 PSA and four non-core damage (NCD) accident sequences were identified for this study. In the accident analysis the success and failure of safety systems reflected in event tree was appropriately modeled in the system thermal hydraulics code runs. The PCT was obtained for each of 25 code runs for each accident sequence. A Kolmogorov – Smirnov goodness-of-fit test carried out for PCTs indicated that they followed normal distribution for each of the accident sequences. The probability distribution of safety margin (difference between acceptable value and PCT) in each accident sequence was also obtained. The values of safety margin for the 95% confidence and 95% probability are estimated. The robustness of the system can be judged based on this. This paper describes the methodology. LBLOCA in a VVER type reactor is considered as an example.

RISK INFORMED DECISION MAKING USING MULTIOBJECTIVE OPTIMIZATION TECHNIQUE

Nuclear Power Plant operations are guided by Limiting conditions of operations (LCO's) laid out in the document referred to as the Technical Specifications (TS). This Technical Specification is a legitimate framework approved by the Regulatory Bodies for the Safe operations of the Nuclear Power plants. In the past, the regulatory bodies used a deterministic approach as the basis for making decisions on safety issues and organizing the activities that they carry out. This was done by applying high level criteria such as the need to provide defence in depth and adequate safety margins. However with the availability of detailed Plant Specific Level-1 Probabilistic Safety Assessment (PSA), these limiting conditions need to be reviewed/revised based on the analysis results. This review of the LCO's is not a trivial exercise if the entire solution space of the variables defining the variables has to be investigated. This paper reviews the case for revision of Surveillance test frequencies of the Emergency Core Cooling System injection valves, using the multiobjective optimization technique.

Uncertainty Analysis Based on Probability Bounds (P‐Box) Approach in Probabilistic Safety Assessment

A wide range of uncertainties will be introduced inevitably during the process of performing a safety assessment of engineering systems. The impact of all these uncertainties must be addressed if the analysis is to serve as a tool in the decision-making process. Uncertainties present in the components (input parameters of model or basic events) of model output are propagated to quantify its impact in the final results. There are several methods available in the literature, namely, method of moments, discrete probability analysis, Monte Carlo simulation, fuzzy arithmetic, and Dempster-Shafer theory. All the methods are different in terms of characterizing at the component level and also in propagating to the system level. All these methods have different desirable and undesirable features, making them more or less useful in different situations. In the probabilistic framework, which is most widely used, probability distribution is used to characterize uncertainty. However, in situations in which one cannot specify (1) parameter values for input distributions, (2) precise probability distributions (shape), and (3) dependencies between input parameters, these methods have limitations and are found to be not effective. In order to address some of these limitations, the article presents uncertainty analysis in the context of level-1 probabilistic safety assessment (PSA) based on a probability bounds (PB) approach. PB analysis combines probability theory and interval arithmetic to produce probability boxes (p-boxes), structures that allow the comprehensive propagation through calculation in a rigorous way. A practical case study is also carried out with the developed code based on the PB approach and compared with the two-phase Monte Carlo simulation results.

The Impact of Severe Accident Management Guidance on the Frequency of Containment Failure Modes of a Boiling Water Reactor

Probabilistic safety assessment (PSA) employs a systematic approach to estimate the risk associated with the operation of nuclear power plants (NPPs). Severe accident management guidance (SAMG), which delineates the mitigation actions of core meltdown accidents of NPPs, is developed to support operators and staffs in the technical support centers during the emergency responses of core melt accidents. Proper execution of SAMG could lower the failure probability of containment and reduce the amount of radionuclides released to the environment during the accident. It can be expected that the implementation of SAMG will reduce the risk of NPPs. However, SAMG is not available when most of the conventional level-2 PSA analyses are performed. In the present study, the mitigation actions of SAMG are incorporated into the level-2 PSA model of the ChinShan Nuclear Power Station of the Taiwan Power Company. The NPP analyzed employs a General Electric-designed boiling water reactor-4 with Mark I containment.The effectiveness of the mitigation actions specified in SAMG to terminate the progression of the accident is verified and validated using the MAAP4 code. The containment system event trees and containment phenomenological event trees of the level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The Human Cognitive Reliability (HCR) and Technique for Human Error Rate Prediction (THERP) models are used to quantify the human error probability (HEP) of all the actions in the level-2 PSA model. The MAAP4 code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the total frequency of accident progression beyond vessel failure is reduced by 41% and the change in the probability of containment staying intact is not very significant because of the implementation of SAMG. After SAMG implementation, the frequency of containment early failure is reduced by 69.9%. The frequency of suppression pool venting is increased by 77.9%. The changes in the frequency of other containment failure modes are relatively insignificant. The most important human action is specified in Guideline RC/F of Severe Accident Guideline-1, i.e., In-Vessel Injection to Arrest Core Damage.

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed.

Quantification of epistemic and aleatory uncertainties in level-1 probabilistic safety assessment studies

There will be simplifying assumptions and idealizations in the availability models of complex processes and phenomena. These simplifications and idealizations generate uncertainties which can be classified as aleatory (arising due to randomness) and/or epistemic (due to lack of knowledge). The problem of acknowledging and treating uncertainty is vital for practical usability of reliability analysis results. The distinction of uncertainties is useful for taking the reliability/risk informed decisions with confidence and also for effective management of uncertainty. In level-1 probabilistic safety assessment (PSA) of nuclear power plants (NPP), the current practice is carrying out epistemic uncertainty analysis on the basis of a simple Monte-Carlo simulation by sampling the epistemic variables in the model. However, the aleatory uncertainty is neglected and point estimates of aleatory variables, viz., time to failure and time to repair are considered. Treatment of both types of uncertainties would require a two-phase Monte-Carlo simulation, outer loop samples epistemic variables and inner loop samples aleatory variables. A methodology based on two-phase Monte-Carlo simulation is presented for distinguishing both the kinds of uncertainty in the context of availability/reliability evaluation in level-1 PSA studies of NPP.

Evaluation of the Impact of SAMG on the Level-2 PSA Results of a Pressurized Water Reactor

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will lower the containment failure frequency and reduce the amount of radionuclides released to the environment during the accident. The plant studied is the Maanshan NPP of Taiwan Power Company, which employs a Westinghouse-designed three-loop pressurized water reactor (PWR) with large dry containment.The containment system event trees and containment phenomenological event trees of the Level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The HCR (Human Cognitive Reliability) and THERP (Technique for Human Error Rate Prediction) models are used to quantify the human error probability (HEP) of all the actions in the Level-2 PSA model. The MAAP4 (Module Accident Analysis Program version 4) code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the frequency of most of the source term categories is reduced except the one in which both the reactor pressure vessel and containment are intact. The containment failure frequency is reduced by 14.8% after the implementation of SAMG. The frequency of containment early failure is reduced by 16.2%. Most of the reduction in the containment early failure frequency comes from the reduction in the induced steam generator tube rupture (STGR). The frequency of induced SGTR was reduced from 2.3 × 10–7/reactor·yr to 1.0 × 10–8/reactor·yr.

An assessment methodology for in-vessel corium retention by external reactor vessel cooling during severe accidents in PWRs

A consistent probabilistic approach is proposed to evaluate the feasibility of in-vessel retention of the molten corium through external reactor vessel cooling (IVR-ERVC) during severe accidents of pressurized water reactors (PWRs). By combining the results of Level-1 probabilistic safety assessment, a critical heat flux correlation, and wall heat flux distributions calculated by a severe accident code with appropriate adjustment, we can reasonably predict the overall success probability of the IVR-ERVC from the viewpoint of thermal failure. The practicability of the proposed approach is illustrated with a preliminary application to the Korean Standard Nuclear Power Plant. This paper also discusses future developmental needs for more reliable assessment.

A concept paper on dynamic reliability via Monte Carlo simulation

The methodologies employed in the probabilistic safety assessment (PSA) of hazardous engineering systems have reached a high level of maturity. However, a few issues are still worth of further investigation in order to increase the confidence in the results obtained. In this view, in the past 5–10 years, researchers in the field of reliability have proposed a more `dynamic' approach to PSA with the aim of addressing issues concerning the possible mutual interactions between the hardware system states and the plant physical evolution. Nonetheless, some objections have been raised against such a dynamic approach, especially against its practical complexity and the lack of a clear definition of the domain of its applicability. In this paper, an attempt to define more precisely the field of application for a dynamic approach is propounded on the basis of the concept of accident duration. The qualitative discussion is supported with examples of postulated severe accidents in nuclear power plants like those investigated in level-2 PSA. The application of Monte Carlo simulation as a tool capable, in principle, of handling all the features of dynamic PSA is illustrated. Monte Carlo algorithms are illustrated aiming at improving the stochastic part of the analysis and the deterministic integration as well, so as to allow for a considerable reduction in the computation times.