Probabilistic Safety Assessment Methodology Research Articles

Methodology for Estimating the Contribution of Forest Fires in Loss of Offsite Power Events

Forest fires are disasters caused by natural or human factors and have environmental as well as economic impacts such as loss of biodiversity, agricultural damage, and property damage. According to statistics, the number of forest fires and the area damaged are gradually increasing in Korea. In addition, there are cases where the units 1, 2, and 6 of the Hanul Nuclear Power Plant (NPP) have already been affected by forest fires. Korea has recently initiated research on forest fires under multi-unit probabilistic safety assessment (PSA) project. This paper deals with the impact on the loss of offsite power (LOOP), which is one of the NPP incidents and accidents affected by forest fires. To estimate the frequency of LOOP induced by forest fires, the PSA methodology for earthquakes, a representative external event, was referred.

Methodology for estimating the contribution of forest fires in loss of offsite power events

Forest fires are disasters caused by natural or human factors and have environmental as well as economic impacts such as loss of biodiversity, agricultural damage, and property damage. According to statistics, the number of forest fires and the area damaged are gradually increasing in Korea. In addition, there are cases where the units 1, 2, and 6 of the Hanul Nuclear Power Plant (NPP) have already been affected by forest fires. Korea has recently initiated research on forest fires under multi-unit probabilistic safety assessment (PSA) project. This paper deals with the impact on the loss of offsite power (LOOP), which is one of the NPP incidents and accidents affected by forest fires. To estimate the frequency of LOOP induced by forest fires, the PSA methodology for earthquakes, a representative external event, was referred.

Probabilistic safety and reliability studies toward licensing and deploying HTGR technology in the Polish chemical industry

The pivotal aspects of the Probabilistic Safety Assessment (PSA) and reliability studies for the foreseen HTGR-based nuclear-chemical (Nuc-Chem) facilities are discussed. The traditional PSA methodology has been extended to include the mutual dependence of nuclear and chemical systems of the hybrid Nuc-Chem installations, where HTGR serves as a heat source for various chemical processes. The Dynamic Bayesian Network (DBN) approach has been utilized for modeling time sequences of failures, including interactions between nuclear and non-nuclear systems, as well as external hazards. The Global Variance Method (GVM), based on the Sobol indices, has been proposed for uncertainty and sensitivity studies concerning numerous sources of variability in a consistent manner. The paper presents the outcomes and achievements from research oriented toward licensing and deploying High Temperature Gas-cooled Reactors (HTGR) within the Polish chemical industry. The general framework developed in this work for PSA of the hybrid Nuc-Chem facilities can be applied to support the licensing process and safe operation of the Polish chemical plants to be equipped with HTGRs.

Analyses of the unavailability dynamics of emergency core cooling system

Abstract The Probabilistic Safety Assessment (PSA) studies can be used to evaluate the time-dependence of the safety systems availability. The purpose of this work is to analyze the unavailability dynamics of the Emergency Core Cooling System (ECCS). Using the PSA methodology, the ECCS was modeled. In order to show how the unavailability of the system varies as function of time (mission time, test interval and component ageing), sensitivity analyses are carried out. Following these analyses, polynomial functions of ECCS unavailability as function of time have been obtained. These functions allow to estimate the unavailability without modelling other ECCS fault trees being a simplified estimation of the system unavailability. In the same time, taking into account that the ECCS unavailability must be less than 10−2, the best time variations can be established to have the unavailability complied with this criterion. Also, the best time variations could be similarly established for any safety systems having the unavailability less than 10−3.

Development of dynamic integrated consequence evaluation (DICE) for dynamic event tree approaches: Numerical validation for a loss of coolant accident

Although conventional probabilistic safety assessment (PSA) makes a sufficient contribution to evaluating the safety of nuclear facilities, it suffers some limitations in performing optimal assessments close to the actual behavior by reflecting time dependencies in accident management. However, integrated deterministic and probabilistic safety assessment (IDPSA) literally combines two methodologies to enable synchronized evaluation based on physical behavior, system availability of facilities, and operator's response. Dynamic integrated consequence evaluation (DICE) is one of the IDPSA tools developed for this purpose, which is the first achievement in South Korea and supported by a regulatory authority. Based on dynamic discrete event tree (DDET) methodology, DICE consists of a scheduler that supports the exchange of information between modules, including a physical module that computes thermal-hydraulic simulations, a diagnosis module that specifies branching points for safety systems, and a reliability module that provides system availability. This paper deals with numerical validation comparing the conventional deterministic and probabilistic results for a sample scenario with those of DICE, and discusses an application of DICE to support and extend the scope of conventional PSA through a case study. It is expected that DICE emphasizes the compliance with conventional PSA methodologies while providing magnified perception in investigating unforeseen scenarios.

Application of external flood probabilistic safety assessment methodology to Prototype Fast Breeder Reactor

ABSTRACT This paper presents the methodology adopted, accident sequence analysis and salient results from the external flood probabilistic safety analysis exercise for Prototype Fast Breeder Reactor. The major steps performed are the probabilistic hazard estimation, fragility computation, system reliability modelling and risk quantification. The hazard analysis has been performed for rainfall, storm surge and tsunami as typically associated with a coastal site, from which correlation function relating flood level to the annual exceedance frequency is estimated. The fragility analysis of different safety systems is estimated by considering submergence mode of failure using step fragility functions. The secondary effects of flood such as debris impact, load falling and flood-induced sodium fires are not considered. The human reliability analysis of manual operations necessary for plant safety is done for the control room as well as onsite actions. The challenges encountered during the analysis, salient results and uncertainties involved are presented with inferences. Abbreviations: AHX: air heat exchanger; CDF: core damage frequency; EFPSA: external flood probabilistic safety assessment; EL: elevation levels; EOP: emergency operation procedure; ET: event tree; FT: fault tree; HRA: human reliability analysis; IE: initiating events; NPP: nuclear power plant; OGDHRS: operation grade decay heat removal system; PFBR: prototype fast breeder reactor; POT: peak over threshold; PSA: probabilistic safety assessment; PTHA: probabilistic tsunami hazard assessment; RY: reactor year; RCB: reactor containment building; RP: return period; SGDHRS: safety grade decay heat removal system; SSC: structures, systems and components

Risk-informed approach for safety improvement of domestic research reactor

This paper describes the effort to improve the safety level of a domestic research reactor and optimize its operation with regard to safety by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety for an operating research reactor in Korea, evaluate whether it is probabilistically safe to operate and reliable to use, and obtain insights about the procedure and design improvements for achieving increased safe operation. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with core damage. The technical objectives of this study were to identify the accident sequences causing core damage and to derive insights, such as design and procedure improvements, from the dominant accident sequences by conducting sensitivity analyses.The event tree (ET)/fault tree (FT) linking approach was used as the basic method for accident sequence quantification for the PSA. The accident sequence models followed the small ET/large FT approach. The total core damage frequency (CDF) was estimated as the sum of the frequencies of the individual ET sequences causing core damage. A truncation value of 1.0E−15/yr was applied for all the FTs or accident sequences. The PSA results indicated a point estimate of 4.55E−06/yr for the overall CDF, attributable to the internal initiating events for the Korea research reactor.Based on the dominant accident sequences from the PSA, eight types of sensitivity analyses were performed, and relevant insights for improved safe operation were derived. When the insights from these sensitivity analyses were applied to the reactor design and the operating procedure, the risk was reduced by approximately ten times compared to that before, and its safety was significantly improved. The PSA methodology is very effective in improving reactor safety in the full-power operating phase, and, in particular, it is a highly suitable approach to determine the deficiencies of a reactor at full power and improve the reactor safety by overcoming them.

Development of Efficient External Multi-Hazard Risk Quantification Methodology for Nuclear Facilities

Probabilistic safety assessment (PSA) of nuclear facilities on external multi-hazards has become a major issue after the Fukushima accident in 2011. However, the existing external hazard PSA methodology is for single hazard events and cannot cover the impact of multi-hazards. Therefore, this study proposes a methodology for quantifying multi-hazard risks for nuclear energy plants. Specifically, we developed an efficient multi-hazard PSA methodology based on the probability distribution-based Boolean algebraic approach and sampling-based method, which are currently single-hazard PSA methodologies. The limitations of the probability distribution-based Boolean algebraic approach not being able to handle partial dependencies between the components are solved through this sampling-based method. In addition, we devised an algorithm that was more efficient than the existing algorithm for improving the limits of the current sampling-based method, as it required a significant computational time. The proposed methodology was applied from simple examples to single- and multi-hazard PSA examples of actual nuclear power plants. The results showed that the proposed methodology was verified in terms of accuracy and efficiency perspectives. Regarding the sampling-based method, it was confirmed that the proposed algorithm yielded fragility and risk results that have similar degrees of accuracy, even though it extracted a smaller number of samples than the existing algorithm.

Development of logical structure for multi-unit probabilistic safety assessment

Abstract Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

A Holistic Methodology for Probabilistic Safety Assessment of Timber Elements Combining Onsite and Laboratory Data

ABSTRACTThe assessment of existing timber structures comprises several fields of knowledge, each one providing its own piece of information aiming at an effective answer. Even if general guidelines are available, there is still a need for a holistic approach that is able to combine information from different sources and infer upon that information allowing also, in that process, for an updating scheme aiming at a more substantiated decision process. In that scope, this article proposes a methodology for a holistic assessment of timber elements. The methodology combines information gathered in different scales and follows a probabilistic framework that allows for the structural assessment of existing timber elements with possibility of inference and updating of the mechanical properties of the elements, through Bayesian methods. The described methodology is based in four main steps: (i) scale of information; (ii) local and global data; (iii) probability assignment; and (iv) long-term structural analysis. After the description of the methodology, its limitations and possible future modifications are discussed.

Probabilistic safety analysis of external floods – method and application

AbstractThe events of Fukushima amplified the scientific interest in the improvement of methods for probabilistic safety assessment (PSA) of extreme external events. The assessment of consequences of external floods belongs to this group of events. The paper presents the key steps of methodology for probabilistic safety assessment of external floods and a recent application for a nuclear power plant in Switzerland. The presented methodology is an extension of earlier activities and provides more focus on the PSA methodology part that may be applicable also for other studies.

A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

Dynamic methodologies for probabilistic safety assessment (PSA) are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30years, numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented, as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

Assessment of the Standard EPRTMReactor against the European Utility Requirements (EUR)

The EUR Steering Committee selected the EPRTM design as one of the first designs to be reviewed and assessed against the EUR. This work was completed in 1999 within the framework of a joint working group involving the designer and EUR group representatives taking into account the design work available at that time. The Volume 3 EPR subset revA was issued in December 1999.At the end of 2003, TVO selected the EPRTM reactor for construction on the Olkiluoto 3 and in 2006 EDF decided to launch the construction of an EPRTM reactor on the Flamanville site.At the end of 2006, AREVA proposed to the EUR Steering Committee to undertake a revision of the EPR Volume 3 to take into account the present status of the work but also the evolution of the EUR document (the assessment was done against Volume 2 Rev B, whereas Volume 2 Rev C is now applicable). This proposal was accepted by the EUR Steering Committee in 2007.By early 2009, the EUR organization completed the compliance assessment of AREVA’s EPRTM reactor standard design against the EUR. The EUR organization formally presented with the certificate during a special ceremony held in Brussels on July 15, 2009. This award confirms that the EPRTM reactor can be built throughout Europe and meets international regulatory requirements.The EUR requirements cover an exhaustive range of requirements for a nuclear plant to operate efficiently and safely and include such areas as layout, systems, material, components, probabilistic safety assessment methodology, availability assessment and many more.The EUR assessment was extremely detailed, covering 4,678 requirements. The EPRTM reactor complied with over 99% of these requirements (including the non-applicable ones). The assessment also showed only a 0.1% level of non-compliance, all of which were substantiated by AREVA.Furthermore, in some cases, the EPRTM reactor exceeded the EUR requirements; these include aircraft crash protection (large commercial aircraft are deterministically covered), waste storage capacity, and a design lifetime of 60 years compared to the EUR requirement of 40.This paper presents the highlights of this review.

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.

Application of CFD Method for Risk Assessment In Road Tunnels

The definition of the deterministic approach in safety analyses arises from the need to understand the conditions that emerge during a fire accident in a road tunnel. The key factor of the tunnel operations during a fire is ventilation, which during the initial phases of the fire has a strong impact on the evacuation of people and later on the access of the intervention units to the tunnel. The paper presents the use of the CFD (Computational Fluid Dynamics) model in tunnel safety assessment process. The set-up of the initial and boundary conditions and the requirement for grid density found from validation tests of an FDS (Fire Dynamics Simulator) are used to prepare three kinds of fire scenarios—20 MW, 50 MW and 100 MW, with different ventilation conditions: natural, semi-transverse, transverse and longitudinal ventilation. The observed variables, soot density and temperature, are presented in minutes time steps through the entire tunnel length. Comparing the obtained data in a table allows the analyses of the ventilation conditions for different heat releases from fires. The second step is to add additional criteria of human behaviour inside the tunnel (evacuation) and human endurance to the elevated gas concentrations and temperature. What comes out is a fully deterministic risk matrix that is based on the calculated data where the risk is ranged on five levels, from the least to a very dangerous level. The deterministic risk matrix represents the alternative to a probabilistic safety assessment methodology, wherein the fire risk is represented in detail and the CFD model results are physically correct.

Methodology of using CFD-based risk assessment in road tunnels

The definition of the deterministic approach in the safety analyses comes from the need to understand the conditions that come out during the fire accident in a road tunnel. The key factor of the tunnel operations during the fire is the ventilation, which during the initial fazes of the fire, impact strongly on the evacuation of people and latter on the access of the intervention units in the tunnel. The paper presents the use of the computational fluid dynamics model in the tunnel safety assessment process. The model is validated by comparing data with experimental and quantifying the differences. The set-up of the initial and boundary conditions and the requirement for grid density found during the validation tests is used to prepare three kind of fire scenarios 20 MW, 50 MW, and 100 MW, with different ventilation conditions; natural, semi transverse, full transverse, and longitudinal ventilation. The observed variables, soot density and temperature, are presented in minutes time steps trough the entire tunnel length. Comparing the obtained data in a table, allows the analyses of the ventilation conditions for different heat releases from fires. The second step is to add additional criteria of human behaviour inside the tunnel (evacuation) and human resistance to the elevated gas concentrations and temperature. What comes out is a fully deterministic risk matrix that is based on the calculated data where the risk is ranged on five levels, from the lowest to a very danger level. The deterministic risk matrix represents the alternative to a probabilistic safety assessment methodology, where the fire risk is represented in detail as well as the computational fluid dynamics model results are physically correct. .

An Introductory Course of the Probabilistic Safety Assessment (PSA) for LWR ― (1) The Progress and Fundamental Philosophy of the Application of PSA Methodology

An Introductory Course of the Probabilistic Safety Assessment (PSA) for LWR ― (1) The Progress and Fundamental Philosophy of the Application of PSA Methodology

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.

Probabilistic safety assessment past, present and future An IAEA perspective

Despite the high level of development that probabilistic safety assessment (PSA) methods have reached, a number of issues place constraints on its use in supporting decision making on safety matters. A recent publication of the International Nuclear Safety Advisory Group (INSAG) represents an important step in reaching international consensus on the use of PSA. PSA is “strongly encouraged” by INSAG; however, it is noted that “PSA methodology is not sufficiently mature for its present status to be frozen”. The main aspects of the report are discussed in this paper. The paper next discusses three main categories of PSA application, namely the adequacy of design and procedures, optimization of operational activities and regulatory applications. For each of the applications, the objectives, specific modelling requirements and the prospects for implementation are presented. Consistent with its statutory functions, an important aspect of the work of the IAEA is to reach international consensus on the possibilities of and limitations on the use of PSA methods. Whereas past efforts have been concentrated on promotion and assistance to perform Level 1 PSAs, work is now extending with emphasis on PSA applications, Level 2 and Level 3 analysis, external events and shutdown risks. The main elements of IAEA's PSA Programme are discussed. Finally some challenges related to the use of PSA in the backfitting of nuclear power plants in Eastern Europe and countries of the former USSR are addressed.

Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor(JPSR).

A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10−7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study i...