Application Of Probabilistic Safety Assessment Research Articles

A Bayesian two-stage approach to integrate simulator data and expert judgment in human error probability estimation

With the ongoing efforts to collect new data for Human Reliability Analysis (HRA) (in particular, from nuclear power plant control room simulators), it becomes important that the coming data will be processed traceably, addressing its underlying variability, eventually in combination with expert judgment. In this direction, this work presents a two-stage Bayesian model to integrate expert-elicited probability estimates and empirical evidence from simulator data in the quantification of HEP values and of the associated variability distributions. The general aim is to provide a data aggregation framework able to mathematically combine diverse information sources throughout the HEP estimation process, in a systematic and reproducible way, contributing to strengthening the empirical basis of future HRA methods. The Bayesian model can be used to produce reference values and bounds for HRA methods as well as to improve the quality of plant-specific HEP estimates for use in Probabilistic Safety Assessment applications. The model is first verified with artificial data and then applied to quantify the HEP of human failure events from literature. Model sensitivity to biases in expert estimates is also investigated.

Application of probabilistic safety assessment (PSA) to the power reactor innovative small module (PRISM)

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

Probabilistic safety assessment and human reliability analysis for the Wendelstein 7-X fusion device

Probabilistic Safety Assessment (PSA) allows an integrated identification of safety “bottlenecks” including a focus on specific equipment, human errors, failure modes in complex systems of fission, fusion and other devices. The objective of this paper is to demonstrate the application of probabilistic safety assessment for an increase of safety and reliability of thermonuclear fusion installations taking into account both equipment and operators’ reliability. Analysis of possible failure of the cooling circuit in one of the most recently constructed fusion devices, namely Wendelstein 7-X (W7-X) stellarator, is used as an example.In the paper, PSA goals for fusion installations were proposed, namely calculation of unavailability of the systems and analysis of accidents leading to equipment damage. For safety analysis of the considered cooling circuit of W7-X, a more precise model, reflecting the effect of human errors was developed. On a system level for the fusion device, the PSA was performed as an alternative to the so-called RAMI approach which focused on the operational functions rather than on physical components. The calculated risk decrease and cost efficiency of risk decrease investments are an easy and clear indicator which supports decision making regarding risk decrease alternatives implementation.

Fault-tree-based instantaneous risk computing core in nuclear power plant risk monitor

Risk monitor is an application of Probabilistic Safety Assessment (PSA) methodology, which has been widely used to determine the instantaneous risk based on the actual configuration of the nuclear power plant (NPP). The instantaneous risk model which has been converted from the baseline risk model of PSA is the fundament of instantaneous risk analysis in a risk monitor. However, the instantaneous risk model is usually very huge and complex, which is hard to be calculated and analyzed by hand. A fault-tree-based instantaneous risk computing core was developed by FDS Team. This code adopts the fast fault tree analysis method based on modified Zero-suppressed Binary Decision Diagram (ZBDD), which includes a set of algorithms such as fault tree importing, fault tree reduction, ZBDD variable ordering, ZBDD parallel manipulation and cut-sets-based ZBDD reconstruction, etc. Benchmarks of actual NPP instantaneous risk models compared with similar codes showed that this code was correct and efficient. This computing core was applied in the Third Qinshan NPP risk monitor TQRM, and some cases based on the actual NPP instantaneous risk models are given.

Markov analysis of redundant standby safety systems under periodic surveillance testing

In modern applications of probabilistic safety assessment (PSA), maintenance planning and changes to technical specifications play an important role, not least due to regulatory requirements. In particular, standby safety systems under periodic surveillance testing are at the center of this issue. Since traditional PSA techniques impose limitations when complex maintenance and repair strategies are to be taken explicitly into account, we introduce continuous time Markov models to discuss various strategies for organizing repair and testing of two-train standby safety systems, which have the potential to replace traditional system models based on fault tree techniques in PSA. Besides a conventional steady state analysis of these Markov models, we provide a general numerical method which allows the calculation of the probability of exceeding allowed outage times of equipment in Markov models of safety systems, and we apply it to the models introduced in the present paper.

Application of Probabilistic Safety Assessment to Rokkasho Reprocessing Plant (III)

In-cell solvent fire has been treated globally as a typical accident of a reprocessing plant. In order for the fire to break out, leakage of a significant amount of solvent, a temperature rise above its flash point, and the presence of ignition source have to occur simultaneously. In this paper, such an accident was analyzed from a probabilistic viewpoint. The occurrence frequency was shown to be 10−7/y because a frequency or probability of each event is so small. Nevertheless, we assumed the fire occurring for 1 h. One hour should be enough for the assessment of the consequence because an operator can implement measures to extinguish the fire by inducing oxygen deficiency in the cell, such as closing a fire damper or stopping the exhausting fan. Since a certain amount of soot produced by fire does not deteriorate the function of a HEPA filter, the effective dose for the general public is less than 0.1 mSv. The results will be useful for judging the safety importance level of the safety measures against fire.

The new IAEA safety standards on probabilistic safety assessment

Abstract Increasingly wide application of Probabilistic Safety Assessment (PSA) to support a variety of issues dealing with operation and design of Nuclear Power Plants (NPP) imposes certain requirements on PSA quality in terms of its scope, level of detail, approaches, assumptions, and data used. If a unified set of balanced requirements is developed and consistently implemented, it would give confidence in using PSA results and insights in different applications and Risk-Informed Decision Making (RIDM); in addition, it would be possible to consistently compare PSA results for different plants. Such requirements are currently developed or being developed in several countries. The International Atomic Energy Agency (IAEA) is in the process of development of several publications in the IAEA Safety Standards Series that would provide guidance on the standard technical content of PSA. The paper discusses the latest developments at the IAEA in the area of PSA and provides a brief overview of the publications.

Overview on the different applications of probabilistic safety assessment for nuclear power plants

Abstract Worldwide it can be recognised that the use of probabilistic safety assessment (PSA) in regulatory as well as operational decision-making is state of the art and seen as a successful development. Therefore, in most cases the regulator encourages the performance of PSAs to provide information to complement and support the defence in depth philosophy as well as operational configuration decisions. The main application of the PSA is still as part of integrated safety reviews, in particular in the frame of comprehensive (periodic) safety reviews. Other more specific applications areas of PSA are, among others, design evaluation, event analysis with aid of PSA, evaluation of technical specifications; risk-informed in-service inspection, risk monitoring and accident management. The extent of these applications vary from country to country but has been increasing during the last years.

Scoping-level Probabilistic Safety Assessment of a complex experimental facility: Challenges and first results from the application to a neutron source facility (MEGAPIE)

This paper presents a scoping-level application of Probabilistic Safety Assessment (PSA) to selected systems of a complex experimental facility. In performing a PSA for this type of facility, a number of challenges arise, mainly due to the extensive use of electronic and programmable components and of one-of-a-kind components. The experimental facility is the Megawatt Pilot Target Experiment (MEGAPIE), which was hosted at the Paul Scherrer Institut (PSI). MEGAPIE demonstrated the feasibility of a liquid lead-bismuth target for spallation facilities at a proton beam power level of 1 MW. Given the challenges to estimate initiating event frequencies and failure event probabilities, emphasis is placed on the qualitative results obtainable from the PSA. Even though this does not allow a complete and appropriate characterization of the risk profile, some level of importance/significance evaluation was feasible, and practical and detailed recommendations on potential system improvements were derived. The second part of the work reports on a preliminary quantification of the facility risk. This provides more information on risk significance, which allows prioritizing the insights and recommendations obtained from the PSA. At the present stage, the limited knowledge on initiating and failure events is reflected in the uncertainties in their probabilities as well as in inputs quantified with bounding values. Detailed analyses to improve the quantification of these inputs, many of which turn out to be important contributors, were out of the scope of this study. Consequently, the reported results should be primarily considered as a demonstration of how quantification of the facility risk by a PSA can support risk-informed decisions, rather than precise figures of the facility risk.

Application of probabilistic safety assessment to parameters of operational limits and conditions of the opal research reactor

This paper gives a brief explanation of Probabilistic Safety Assessment (PSA), the meaning and application of Operational Limits and Conditions (OL&Cs) for facilities and describes the application of PSA to the derivation of OL&C parameters such as Completion Times. ANSTO’s Research Reactor OPAL is used as a case study but the authors conclude that the methodology, based on NUREG/CR-6141, could equally be applied to major hazard facilities.The PSA indicates that the OPAL reactor meets the regulatory safety limits and safety objectives. The very low risk is achieved by important design characteristics such as the two independent and diverse shutdown systems, minimal dependence on support systems for the safety functions, the absence of in-core experiments, and the redundancy in cooling modes.Completion times are derived for safety systems and some sub-systems by balancing the incremental risk of single outages of those systems, and also by balancing the incremental risks associated with the expected number of outages of those systems and subsystems while ensuring that the incremental risks are tolerable.This methodology can give an estimate of the total incremental risk of allowable outages of safety systems and subsystems taking into account the expected number of outages of those systems each year. An assessment of the tolerability of that additional risk can be made.

Application of Probabilistic Safety Assessment to Rokkasho Reprocessing Plant, (II) The Occurrence Frequency of Boiling Accident of Highly Active Liquid Waste

A probabilistic safety assessment (PSA) is made on the boiling accident of a highly active liquid waste tank, which may result in significant consequences, in accordance with the procedure for PSA developed for nuclear power plants. Obtained as results are the frequency of boiling accident of a certain tank of 2.0×10−8/y (frequency of boiling accident of any tank of 4.1×10−8/y), its error factor of approx. 6, and information on the relative risk importance based on the FV index and RAW for various components, systems and activities of personnel and on the sensitivity of key parameters. Furthermore, the effect of the time required for repairing failed instruments on the frequency of accident, how to deal with the common cause of failure of the duplicated dynamic components, one of which is at least in operation, and conservative exposure dose in the event of an accident are examined. The database for the Rokkasho reprocessing plant has not been established yet, but the PSA results utilizing available failure rate databases of existing nuclear power plants and reprocessing plants in Japan and abroad can be used effectively to optimize operations and maintenance, if they are interpreted properly and some uncertainties are taken into account.

六ヶ所再処理工場の確率論的安全評価（５） _-_高レベル廃液貯槽の冷却機能喪失／感度解析_-_

六ヶ所再処理工場の確率論的安全評価（５） _-_高レベル廃液貯槽の冷却機能喪失／感度解析_-_

六ヶ所再処理工場の確率論的安全評価（４） - 高レベル廃液貯槽の冷却機能喪失／冷却系統構成及び発生頻度評価 -

六ヶ所再処理工場の確率論的安全評価（４） - 高レベル廃液貯槽の冷却機能喪失／冷却系統構成及び発生頻度評価 -

六ヶ所再処理工場の確率論的安全評価,(I)

六ヶ所再処理工場の確率論的安全評価,(I)

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Methods for probabilistic safety assessments of ships

At the 62nd MSC conference (MSC62) in 1993, the UK proposed a new methodology for the consideration of safety regulations. This method is called formal safety assessment (FSA). FSA is an application of probabilistic safety assessment (PSA). Risk is used as an index of safety. One of the most important parts of FSA is to evaluate the risk to a ship when it is equipped with the safety measures recommended by the proposed safety regulations. The National Maritime Research Institute (NMRI, formerly the Ship Research Institute) has been developing a method which allows the risk to be obtained holistically by utilizing a scientific method. To obtain the risk, the probability and consequences of every accident must be evaluated. This paper examines the following points: (a) a holistic methodology for risk evaluation; (b) a method used in the process of estimating the probability of collision; (c) a method to reduce the numbers of fire escalation scenarios; (d) a trial risk evaluation of cabin fire.

95/03592 Application of probabilistic safety assessment in the regulatory process in Germany

95/03592 Application of probabilistic safety assessment in the regulatory process in Germany

Probabilistic safety assessment for non-full- power states of nuclear power plants / Probabilistische Sicherheitsanalyse für Kernkraftwerke außerhalb des Leistungsbetriebs

It has become obvious in the last few years that non-full-power states may have a significant impact on the safety of a nuclear power plant. In a logical extension of the application of probabilistic safety assessment (PSA), plant states other than power operation states are analysed now in nearly all countries operating nuclear plants. The application to events during low-power and shutdown states requires further development of PSA methodology in many areas. This paper describes the international state of the art and presents some details of studies performed for a pressurised water reactor and a boiling water reactor

Determination of input data for probabilistic safety assessment / Die Gewinnung von Eingabedaten für probabilistische Sicherheitsanalysen

The application of probabilistic safety assessment in decision processes depends to a great extent on the quality of the input data. The plant specific acquisition of these data must be performed in a way which allows a reliable prediction of system failures in the future. Comfortable computer tools are available to facilitate data processing and to make use of data that are already available via the computerized plant information system. The correct classification of failure occurrences with respect to the PSA relevant failure modes requires thorough evaluation of all information available in maintenance records, test reports, and licensing event reports. The flow of information relevant to future data acquisition must be properly organized. The application of generic common cause failure data requires careful evaluation of the underlying failure causes and of the level of defence achieved in the plant under consideration to prevent too pessimistic analysis results