Probabilistic Risk Assessments PRAs

Oconee-3 PRA A Probabilistic Risk Assessment of Oconee Unit 3 

Yankee Nuclear Power Station Probabilistic Safety Study 

Reactor Safety Study An Assessment of AccidenI Risk in U.S. Commercial Nuclear Power Plants (WASH-1400) 

Failure and maintenance data using experience to update generic rates 

Generic data set based on plant and industry failure reports over a ten-year span 

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Murphy, J. A., Probabilistic Risk Assessment (PRA) Reference Document. September 1984. 

Section 4.6 Nonprocess Equipment Data Bases Section 4.7 Nonprocess Equipment Data Sources Section 4.8 Nuclear Probabilistic Risk Assessments (PRA)... 

Probabilistic Risk Assessment (PRA) A conunonly used term in the nuclear industry to describe the quantitative evaluation of risk. 

The terminology used varies considerably. Hazard identification and risk assessment are sometimes combined into a general category called hazard evaluation. Risk assessment is sometimes called hazard analysis. A risk assessment procedure that determines probabilities is frequently called probabilistic risk assessment (PRA), whereas a procedure that determines probability and consequences is called quantitative risk analysis (QRA). 

A probabilistic risk assessment (PRA) deals with many types of uncertainties. In addition to the uncertainties associated with the model itself and model input, there is also the meta-uncertainty about whether the entire PRA process has been performed properly. Employment of sophisticated mathematical and statistical methods may easily convey the false impression of accuracy, especially when numerical results are presented with a high number of significant figures. But those who produce PR As, and those who evaluate them, should exert caution there are many possible pitfalls, traps, and potential swindles that can arise. Because of the potential for generating seemingly correct results that are far from the intended model of reality, it is imperative that the PRA practitioner carefully evaluates not only model input data but also the assumptions used in the PRA, the model itself, and the calculations inherent within the model. This chapter presents information on performing PRA in a manner that will minimize the introduction of errors associated with the PRA process. 

HAZAN, on the other hand, is a process to assess the probability of occurrence of such accidents and to evaluate quantitatively the consequences of such happenings, together with value judgments, in order to decide the level of acceptable risk. HAZAN is also sometimes referred to as Probabilistic Risk Assessment (PRA) and its study uses the well-established techniques of Fault Tree Analysis and/or Event Tree Analysis ... 

CPQRA A chemical process quantitative risk analysis is the process of hazard identification followed by numerical evaluation of incident consequences and frequencies, and their combination into an overall measure of risk when apphed to the chemical process industry. It is particularly applicable to episodic events. It differs from, but is related to, a probabilistic risk assessment (PRA), a quantitative tool used in the nuclear industry. 

In performing a probabilistic risk assessment (PRA), initiating events in the chain are usually assumed to be mutually exclusive. While this assumption simplifies the mathematics, it may not match reality. As an example, consider the following description of an accident chain for an offshore oil platform ... 

Selection of licensing basis events (LBEs) is based on the probabilistic risk assessment (PRA) performed as part of the Integrated Approach and constitutes the process which establishes the bridge between the engineering approach and the licensing basis for the Standard MHTGR. The use of the PRA for LBE selection provides a basis for judging, in a quantitative manner, the frequency of the entire event sequence and, therefore, the appropriate dose or risk criteria to be applied. 

The simultaneous occurrence of pipe break and either safe shutdown earthquake (SSE) or operating basis earthquake (QBE) as in Section 2.6, is considered too improbable to be incorporated in the plant design basis. This is reflected in the loading combinations for structures in Section 3.8. It is expected that the probabilistic risk assessment (PRA) will confirm the above, and additional revisions to load factors and load combinations may be made if they are supported by the results of probabilistic analyses. (Ref. 1) The current design is based on the following ... 

Toward the end of the Second World War, systems techniques such as fault tree analysis were introduced in order to predict the reliability and performance of military airplanes and missiles. The use of such techniques led to the formalization of the concept of probabilistic risk assessment (PRA). The publication of the Reactor Safety Study (NRC, 1975)—often referred to as the Rasmussen Report after the name of principal author, or by its subtitle WASH 1400—demonstrated the use of such techniques in the fledgling nuclear power business. Although WASH 1400 has since been supplanted by more advanced analysis techniques, the report was groundbreaking in its approach to system safety. 

One method for analyzing human reUability is a straightforward extension of probabilistic risk assessment (PRA)—in the same way that equipment can fail, so can a human make mistakes and slips. One technique for predicting human error rates is the THERP, which was developed in the 1950s. As with other PRA techniques, THERP models can use either point. 

We arrive at a probabilistic risk assessment (PRA) if the accident consequences are assessed as well and frequency and consequence are combined. This is done in the following steps. 

The goal of human error quantification is to produce error probabilities, building on task analysis and error identification techniques to provide a probabilistic risk assessment (PRA). This provides numerical estimates of error likelihood and of the probability of overall likelihood of system breakdown. Quantification of error is the most difficult aspect of HRA, often heavily reliant on expert judgement, rather than the more rigorous approach of actual observation and recording of error frequencies. Such techniques are little used in healthcare but have been successfully applied to anaesthesia (Pate-Cornell and Bea, 1992). Nevertheless, some hospital tasks, such as blood transfusion, are highly structured and the quantification of errors probabilities would seem to be eminently feasible (Lyons et al, 2004). 

Probabilistic Risk Assessment (PRA) builds on such techniques as FMEA and HAZOP, by adding modelling of fault and event trees and assignment of probabilities to events and outcomes. 

USNRC s policy for implementing risk-informed regulation was expressed in the 1995 policy statement on the use of probabilistic risk assessment (PRA) methods in nuclear regulatory activities. The policy statement says The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC s deterministic approach and supports the NRC s traditional defence-in-depth philosophy. 

Wreathall, J Nemeth, C. (2004). Assessing risk the role of probabilistic risk assessment (PRA) in patient safety improvement.and Safety in Health Care 13, 206-212. 

Mohaghegh, Z, Kazemi, R, Mosleh, A, 2008, Incorporating organizational factors into Probabilistic Risk Assessment (PRA) of complex socio-technical systems A hybrid technique formalization. Reliability engineering and system safety, 94, 5 1000-1018. 

ABSTRACT When a fire Probabilistic Risk Assessment (PRA) is modeled and quantified by using predeveloped internal PRA model, if components are damaged by a fire, the basic event values of the components became True or one (1), which removes the basic events related to the components from the minimal cut sets, and which makes it difficult to calculate accurate component importance measures. Thus, a new method to accurately calculate Fussell-Vesely importance measure in fire PRA is recently introduced. However, the new method has a drawback when the failure probability of the damaged component is small. Thus, another new method could be proposed. Two methods are compared, and the condition in which each method is accurately applicable is derived in this paper. 

For risk informed regulation and applications (RIR A), importance measures (IMs) (Borst et al. 2001, Kim et al. 2003, Vesely et al.l983, Wall et al. 1996), such as Fussell-Vesely (FV), Risk Reduction Worth (RRW) or Risk Achievement Worth (RAW), play a very important role. Especially, IMs in fire probabilistic risk assessment (PRA) as well as internal events PRA are important for the risk informed structures, systems and components (SSCs) categorizations (NRC 2002, NEI 2004). 

The sequences of events that may lead to vessel failure and their frequencies are determined from probabilistic risk assessment (PRA) analyses. The pressure, temperature and heat transfer coefficient time histories at the vessel inner surface are determined from thermal hydraulic analyses for the events identified by the PRA analyses. These time histories are used together with probabilistic fracture mechanics (PFM) analysis to calculate the conditional probability of RPV failure. Discussion of the methodology used to perform the PRA analyses and define the transient events and associated frequencies, and the thermal hydraulic analyses used to define the event pressure and temperature time histories are outside the scope of this chapter. Consequently, the remainder of this chapter focuses on the PFM evaluation assumptions and procedures. 

The NRC used probabilistic risk assessment (PRA) analysis in NUREG/CR-5102 (Reference 2) to evaluate three older operating pressurized water reactors for the effectiveness of proposed requirement changes in reducing the risk of an interfacing system LOCA, and for calculating the contribution of the ISL to the overall core damage frequency estimate. 

Industry, (including plant designers and owner-operators) and the NRC are concerned about designing and operating nuclear power plants safely and reliably. Before the advent of Probabilistic Risk Assessment (PRA) it was difficult to systematically assess plant safety and reliability. Therefore, both industry and regulators consider PRA, as part of a comprehensive reliability program, to be desirable for future plants. The NRC has placed an emphasis on PRA for future plants by including it in the Standardization Rule (10 CFR 52). 

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