Risk-informed Applications Research Articles

Improvement activities for domestic NPP’s PRA (Improvement of PRA toward Risk-Informed Applications)

Improvement activities for domestic NPP’s PRA (Improvement of PRA toward Risk-Informed Applications)

REVIEW OF RISK-INFORMED SAFETY DESIGN CONCEPT, METHOD AND APPLICATION

REVIEW OF RISK-INFORMED SAFETY DESIGN CONCEPT, METHOD AND APPLICATION

Research on Time-Dependent Component Importance Measures Considering State Duration and Common Cause Failure

Unlike the current risk monitors, Real-time Online Risk Monitoring and Management Technology is characterized by time-dependent modeling on the state duration of components. Given the real-time plant configuration, it eventually provides the time-dependent risk level and importance measures for operation and maintenance management. This paper focuses on the assessment method of time-dependent importance measures and its risk-informed applications in real-time online risk monitoring and management technology, including Fussell-Vesely (FV), risk achievement worth (RAW), and risk reduction worth (RRW). In this study, the values of component importance have been investigated with a time-dependent risk quantification model, as well as the common cause failure treatment model. Here three options of common cause failure treatment have been developed, assuming that the unavailability of a component could be due to an independent factor (Option 1), a common cause factor (Option 2), or an unconfirmed cause (Option 3). In the special case of “what if a component is out-of-service” of the RAW numerator, a hybrid method for the RAW evaluation is presented resulting in a balanced and reasonable RAW value. A simple case study was demonstrated. The results showed that the absolute values and ranking order of time-dependent importance not only reflected the effect of the cumulative state duration of component on risk, but also comprehensively accounted for all possible situations of component unavailability. Moreover, time-dependent importance measures improved and provided novel insights for online configuration management, 1) ranking SSCs/events/human actions for controlling increased risk and optimizing near–term plans; and 2) exempting or limiting temporary configurations during online operation.

Multi-unit risk aggregation with consideration of uncertainty and bias in risk metrics

The risk significance of multi-unit events has received much interest, especially since the 2011 Fukushima–Daiichi accident. However, there have been limited experiences in performing a multi-unit probabilistic risk assessment (MUPRA), considering the interactions among multiple reactor units and fuel storage facilities on a site. While considerable research and development efforts have been devoted over the past few years to MUPRAs, there is still no consensus on a unified MUPRA methodology. A site-based risk model is of great importance for comprehensive risk-informed applications and safety goals evaluations. Further, a site-based MUPRA model needs to aggregate risks from various reactor units’ internal and external sequences of events and modes of operation. Risk values obtained from various initiating events and modes of operation, however, are usually biased very differently due to the various factors such as conservatism and modeling assumption used during their developments. As such, to appropriately aggregate various sequences of events, they should be least biased. This paper discusses the framework of a probabilistic aggregation approach for the multi-unit risk metrics to make them least biased using expert elicitation. More importantly, the paper, through a sensitivity analysis, shows that a biased risk metric, even when it is not aggregated, could mask important contributors to risk and thereby yield incorrect risk contributors. This is done by comparing the risk insights from importance measures for various modeling scope and assumptions that make a risk metric biased. Through an example, it is shown the masking of important risk contributors and human error events would be an important impediment in the MUPRAs. The example provides an illustration of the aggregation process of multi-unit risk metrics for both external events and internal events. The paper demonstrates the importance of biased correction before approaching to multi-unit risk aggregation, especially for the external events.

Use of Risk-Informed Technique to Evaluate the Balance of Technical Specifications Events

Risk-informed Technique, combining Deterministic analysis with Probabilistic Risk analysis, could evaluate the whole risk of Nuclear Power Plant, which is more and more deeply used in different areas in NPP, such as Technical Specifications, In-service inspection, GQA and so on. In Risk-informed Technical Specifications Optimization, a three-tiered approach for licensees to evaluate risk associated with proposed TS changes: Probabilistic Risk Assessment Capability and insights, Avoidance of Risk-Significant Plant Configuration, and Risk-Informed Configuration risk management. The Avoidance of Risk-Significant Plant Configuration is one of the essential links, which is also the chief content of Risk-informed Technical Specifications Optimization, so identifying the risk-signification configurations is particularly important for evaluating risk impact. This paper takes one proposed Spray isolation valve TS item change of Containment Spray System in some NPP as an example, to illustrate the method and process of Identifying Risk-significant Configuration from importance analysis result, and gives management suggestions based on the Risk-significant Configuration result, which gives a special way to analysis avoidance of risk-significant configuration, and a well practical lesson for Risk-informed application..

Study on Identifying Risk-Significant Configuration in Risk-Informed Technical Specifications Optimization in Nuclear Power Plant

Risk-informed Technique, combining Deterministic analysis with Probabilistic Risk analysis, could evaluate the whole risk of Nuclear Power Plant, which is more and more deeply used in different areas in NPP, such as Technical Specifications, In-service inspection, GQA and so on. In Risk-informed Technical Specifications Optimization, a three-tiered approach for licensees to evaluate risk associated with proposed TS changes: Probabilistic Risk Assessment Capability and insights, Avoidance of Risk-Significant Plant Configuration, and Risk-Informed Configuration risk management。 The Avoidance of Risk-Significant Plant Configuration is one of the essential links, which is also the chief content of Risk-informed Technical Specifications Optimization, so identifying the risk-signification configurations is particularly important for evaluating risk impact. This paper takes one proposed Spray isolation valve TS item change of Containment Spray System in some NPP as an example, to illustrate the method and process of Identifying Risk-significant Configuration from importance analysis result, and gives management suggestions based on the Risk-significant Configuration result, which gives a special way to analysis avoidance of risk-significant configuration, and a well practical lesson for Risk-informed application.

Distribution and Application of Community-based Disaster Risk Information : Lessons from Shiga Prefecture in Japan

Distribution and Application of Community-based Disaster Risk Information : Lessons from Shiga Prefecture in Japan

Methodology to evaluate the monetary benefit of Probabilistic Risk Assessment by modeling the net value of Risk-Informed Applications at nuclear power plants

Probabilistic Risk Assessment (PRA) used in Nuclear Power Plants serves as a pillar of the U.S. Nuclear Regulatory Commission's Risk-Informed Regulatory framework, and is required for new reactor licenses to satisfy regulatory safety compliance. The benefits of PRA are not only experienced in terms of safety, but also from the monetary value derived from Risk-Informed Performance-Based Applications (RIPBAs), where risk estimated from PRA is utilized in decision making to expand the safe operational envelope of plants, leading to either an increase in profits or a reduction in costs. This paper introduces a methodology to evaluate this monetary value by the systematic causal modeling of the net value of RIPBAs and demonstrates the methodology for one of the RIPBAs, called Risk-Managed Technical Specifications (RMTS). The key steps of this methodology are: (i) Cost-Benefit Analysis to formulate the net value of PRA based on the net value of RIPBAs, (ii) Causal modeling to systematically model the operational scenarios leading to costs and benefits associated with RIPBAs, (iii) Uncertainty analysis, and (iv) Sensitivity analysis and validation. The results of this research could help decision makers with evaluating investment strategies in PRA that go ‘beyond-compliance’ to maximize industry profit while maintaining regulatory safety goals.

Qualitative Flood Risk assessment for the Western Province of Sri Lanka

The Western Province of Sri Lanka, is comprised of the 03 most densely populated districts, and is home to the administrative and business capital of the country. The Province contributes the highest share, around 41.2%, to the GDP. The mega scale development projects implemented in this province are often challenged by urban flooding, which is caused by heavy rainfall coupled up with urban expansion, deficiencies in the drainage systems, insufficient retention capacity etc., and there is a prioritized need to reduce the impact of floods. This paper presents the results of a qualitative flood risk assessment at the scale of Grama Niladhari (GN) Division, which is the lowest administrative unit in the country. The essence of the method applied for delineating the combined flood risk levels is a statistical expression of hazard, exposure and vulnerability. Three types of vulnerabilities, namely, social, economic, and physical (housing), have been considered. The results influence to conclude that the flood risk of population is more sensitive to economic vulnerability than to social vulnerability; also that the type of housing units is a reasonable indicator for assessing the risk of housing. The application of risk information in the decision making process reduces the future flood risk as it allows integration of flood mitigation options during development planning, reduce the vulnerability of population and allocate sufficient funding for relief and rehabilitation.

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4in and by only a high pressure safety injection in LOCAs of 0.8–9.4in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety assessment, but also a more realistic core damage frequency of the LOCAs using updated information. The difference between the results of the present study and the conventional knowledge can be used to modify the current emergency operator procedures and to design another nuclear power plant type.

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of material degradation mechanisms and on effective strategies to manage the risks of debris-induced failures during a LOCA.

Risk monitor riskangel for risk-informed applications in nuclear power plants

This paper studied the requirements of risk monitor software and its applications as a plant specific risk monitor, which supports risk-informed configuration risk management for the two CANDU 6units at the Third Qinshan nuclear power plant (TQNPP) in China. It also describes the regulatory prospective on risk-informed Probabilistic Safety Assessment (PSA) applications and the use of risk monitor at operating nuclear power plants, high level technical and functional requirements for the development of CANDU specific risk monitor software, and future development trends.

Emerging I&C Technologies Under the Shifting Regulatory Environment in South Korea

The role of Probabilistic Safety Assessment (PSA) has been supplementary and Risk-Informed Applications (RIAs) based on the insight from PSA has also been utilized limitedly in the licensing process for Nuclear Power Plants (NPPs) in South Korea. However, as the technical significance of PSA is getting increased, PSA has become a mandatory part of Safety Analysis Reports and Periodic Safety Review. It is worthwhile to highlight the role of emerging Instrumentation and Control (I&C) technologies including human-machine interface (HMI) in developing more credible and realistic PSA models. Particularly, it is expected that the information technology (i.e. software) embedded in digital I&C can adjust over- and under conservatism in analyzing risk. In this study, authors proposed the cases which would be able to significantly reduce risk if advanced I&C supported by information technologies is applied. In regard, the several enabling techniques and their effects are proposed. In order to improve the commercial competitiveness of NPPs, the need of collaboration and synergetic outcome of I&C, HMI and PSA should be emphasized.

Reliability data update using condition monitoring and prognostics in probabilistic safety assessment

Probabilistic safety assessment (PSA) has had a significant role in quantitative decisionmaking by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still “conservative” aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide timeand condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

Risk‐Informed Regulation and Safety Management of Nuclear Power Plants—On the Prevention of Severe Accidents

There are four operating nuclear power plant (NPP) units in Finland. The Teollisuuden Voima (TVO) power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213 units at the Loviisa site. In addition, a 1600 MWe European Pressurized Water Reactor supplied by AREVA NP (formerly the Framatome ANP--Siemens AG Consortium) is under construction at the Olkiluoto site. Recently, the Finnish Parliament ratified the government Decision in Principle that the utilities' applications to build two new NPP units are in line with the total good of the society. The Finnish utilities, Fenno power company, and TVO company are in progress of qualifying the type of the new nuclear builds. In Finland, risk-informed applications are formally integrated in the regulatory process of NPPs that are already in the early design phase and these are to run through the construction and operation phases all through the entire plant service time. A plant-specific full-scope probabilistic risk assessment (PRA) is required for each NPP. PRAs shall cover internal events, area events (fires, floods), and external events such as harsh weather conditions and seismic events in all operating modes. Special attention is devoted to the use of various risk-informed PRA applications in the licensing of Olkiluoto 3 NPP.

Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited from PSA is indeed the systematic way which PSA follows. This is more important especially for the new reactors.

ICONE19-43321 Development of Integrated Parameter Database for Risk Assessment at the Rokkasho Reprocessing Plant

A study to develop a parameter database for Probabilistic Safety Assessment (PSA) for the application of risk information on plant operation and maintenance activity is important because the transparency, consistency, and traceability of parameters are needed to explanation adequacy of the evaluation to third parties. Application of risk information for the plant operation and maintenance activity, equipment reliability data, human error rate, and 5 factors of "five-factor formula" for estimation of the amount of radioactive material discharge (source term) are key inputs. As a part of the infrastructure development for the risk information application, we developed the integrated parameter database, "R-POD" (Rokkasho reprocessing Plant Omnibus parameter Database) on the trial basis for the PSA of the Rokkasho Reprocessing Plant. This database consists primarily of the following 3 parts, 1) an equipment reliability database, 2) a five-factor formula database, and 3) a human reliability database. The underpinning for explaining the validity of the risk assessment can be improved by developing this database. Furthermore, this database is an important tool for the application of risk information, because it provides updated data by incorporating the accumulated operation experiences of the Rokkasho reprocessing plant.

Analysis of an Operators' Performance Time and Its Application to a Human Reliability Analysis in Nuclear Power Plants

With a scarcity of empirical data, much of the input regarding a human reliability analysis (HRA) is dependent on an analysts' subjective judgment, which is one of the major causes of a high uncertainty in the results of an HRA. When considering that risk-informed activities require the probabilistic safety assessment (PSA) quality to be as high as possible, an HRA should be performed by using a systematic method with realistic plant specific data to meet the established requirements for risk-informed applications. In order to obtain more objective HRA results, data extracted from real experiences or simulators is essential. This paper presents a human performance analysis of the emergency tasks for nuclear power plants (NPPs) to support an HRA. Using the full-scope simulator of a reference plant, more than 110 simulator records with six emergency scenarios were gathered and analyzed to generate an operators' performance time such as the time taken to diagnose what event has occurred, and the time to execute a procedural step or a task. A sensitivity analysis for an HRA has also been performed by using an operators' performance time.

Key attributes of the SAPHIRE risk and reliability analysis software for risk-informed probabilistic applications

The Idaho National Laboratory is a primary developer of probabilistic risk and reliability analysis (PRRA) tools, dating back over 35 years. Evolving from mainframe-based software, the current state-of-the-practice has led to the creation of the SAPHIRE software. Currently, agencies such as the Nuclear Regulatory Commission, the National Aeronautics and Aerospace Agency, the Department of Energy, and the Department of Defense use version 7 of the SAPHIRE software for many of their risk-informed activities. In order to better understand and appreciate the power of software as part of risk-informed applications, we need to recall that our current analysis methods and solution methods have built upon pioneering work done 30–40 years ago. We contrast this work with the current capabilities in the SAPHIRE analysis package. As part of this discussion, we provide information for both the typical features and special analysis capabilities, which are available. We also present the application and results typically found with state-of-the-practice PRRA models. By providing both a high-level and detailed look at the SAPHIRE software, we give a snapshot in time for the current use of software tools in a risk-informed decision arena.