PRA analysis

Hannaman, G.W., Spuigin, A.J., Lukic, Y.D. 1984. Human cognitive reliability model for PRA analysis. In draft report NUS-4531, EPRI Project RP2170-3. 

First, the results of the PRA analysis of TS changes must he presented in such a way that uncertainty is well shown, which depends on the type of imcertain-ties being considered. T3rpically, the ways in which the imcertainty in PRA results is presented include the following ... 

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. 

The main tasks of Level-I PRA analysis include the assessment of... 

The PRA analysis provides the following insights into the APIOOO design (Section 19.59.4.2 of... 

Results of the PRA analysis are described above (core fuel melting probability lower than lx 10 ). 

The PRA model is adopted for risk modeling. RG 1.174 establishes that PRA used should be performed in a manner that is consistent with accepted practices. RG 1.174 states that the quality of a PRA analysis can be measured in terms of its appropriateness with respect to scope, level of detail and technical acceptability. In addition, RG 1.177 requires that the quality of the PRA must be compatible with the safety implications of the TS change being requested and the role the PRA plays in justifying the change. 

The US Nuclear Regulatory Commission (NRC) Safety Evaluation Report (U.S. NRC, 2014) gives the criteria for risk assessment based on core damage frequency and the time scales for the use of safety and nonsafety systems, as derived from a full-scope Probabilistic Risk Assessment (PRA) analysis (Bhatt and Wachowiak, 2006). This NRC approach states the safety guidelines as follows ... 

Risk-Based Inspection. Inspection programs developed using risk analysis methods are becoming increasingly popular (15,16) (see Hazard ANALYSIS AND RISK ASSESSMENT). In this approach, the frequency and type of in-service inspection (IS I) is determined by the probabiUstic risk assessment (PRA) of the inspection results. Here, the results might be a false acceptance of a part that will fail as well as the false rejection of a part that will not fail. Whether a plant or a consumer product, false acceptance of a defective part could lead to catastrophic failure and considerable cost. Also, the false rejection of parts may lead to unjustified, and sometimes exorbitant, costs of operation (2). Risk is defined as follows ... 

The acronym for chemical process quantitative risk analysis. It 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 applied to the chemical process industry. It is particularly applied to episodic events. It differs from, but is related to, a probabilistic risk analysis (PRA), a quantitative tool used in the nuclear industry... 

Hazard analysis (HAZAN) is a quantitative way of assessing the likelihood of failure. Other names associated with this technique are risk analysis, quantitative risk assessment (QRA), and probability risk assessment (PRA). Keltz [44] expressed the view that HAZAN is a selective technique while HAZOP can be readily applied to new design and major modification. Some limitations of HAZOP are its inability to detect every weakness in design such as in plant layout, or miss hazards due to leaks on lines that pass through or close to a unit but cany material that is not used on that unit. In any case, hazards should... 

The PRA procedures guide, NUREG/ CR-23(X), partitions human reliability analysis (HRA) into four phases (Figure 4.5-1). The familiarization phase, evaluates a sequence of events to identify human actions that directly affect critical process components. From plant visits and review, this part of HRA identifies plant-specific factors that affect human performance such as good or bad procedures used in the. sequence under consideration. The familiarization phase notes items overlooked during systems evaluation. 

In 1988, an internal events PRA was published (Johnson, 1988) this was followed by an external events analysis. The results were reported by Johnson (1991), and Flanagan (1990). The basic approach to risk is that of Kaplan (1981) that asks the questions "What can happen, and How likciy is it and what are the consequences " These are organized as triplets to characterize the risk... 

Several initiating events are unique 1) large LOCA (6), 2) flow blockage events (4a. 4b, and 4d). and 3) fuel defects (4e). The analysis of these required special tools not found in coinmcrcial power plant PRAs. 

Because of these major differences, an LWRs PR A cannot be adapted to make an SRS PRA which must be made from basics. All phases of the analysis are as unique to SRS. 

Joksimovich, V, 1984, "A Review of Plant Specific PRA s," Risk Analysis 4 p 255, December. 

Risk study (CPQRA or PRA) Specific study performed on a particular facility to determine the areas of weakness and strength in equipment and plant performance reliability may include consequence analysis and usually implies some judgment of the risk. 

The main objective of the In-Plant Reliability Data System (IPRDS) was to develop a comprehensive and component-specific data base for PRA and other component reliability-related statistical analysis. Data base personnel visited selected plants and copied all the plant maintenance wor)c requests. They also gathered plant equipment lists and plant drawings and in some cases interviewed plant personnel for Information on component populations and duty cycles. Subsequently, the maintenance records were screened to separate out the cases of corrective maintenance applying to particular components these were reviewed to determine such things as failure modes, severity, and, if possible, failure cause. The data from these reports were encoded into a computerized data base. 

The report presents the findings from the analysis of the RCP failures. Estimates of the annual frequency for the spectrum of leak rates induced by RCP seal failures and their impact on plant safety (contribution to coremelt frequency) are made. The safety impact of smaller RCP seal leaks was assessed qualitatively, whereas for leaks above the normal makeup capacity, formal PRA methodologies were applied. Also included are the life distribution of RCP seals and the conditional leak rate distributions, given a RCP seal failure the contribution of various root causes and estimates for the dependency factors and the failure intensity for the different combinations of pump designers and plant vendors. 

PRA Documents generally have a limited distribution and comprise several volumes of reports. The data may be available in government or national laboratory libiaries but access is usually restricted to the utility and contractors who performed the analysis. 

A systematic approach was undertaken for the BRP PRA to identify all potential sources of common mode failure. The first step in the treatment of common mode failures was a compilation of a detailed list of common mode initiators. To achieve this, available literature on common mode failure analysis was reviewed. The next step was to qualitatively assess the potential effects of these initiators on BRP systems. The initiator categories and the systems selected for examination are presented in Table VI.1 of the BRP PRA. 

Opening segments of the IP2 PRA data analysis section describe the definitions of terms and concepts employed, the assumptions made, and limitations recognized during the data base construction. A set of 39 plant-specific component failure mode summaries established the basis for component service hour determinations, the number of failures, and the test data source for each failure mode given for each component. Generic data from WASH-1400, IEEE Std 500, and the LER data summaries on valves, pumps, and diesels were combined with plant-specific failure data to produce "updated" failure information. All the IP2 specialized component hardware failure data, both generic and updated, are contained in Table 1.5.1-4 (IP3 1.6.1-4). This table contains (by system, component, and failure mode) plant-specific data on the number of failures and service hours or demands. For some components, it was determined that specifications of the system was warranted because of its impact on the data values. 

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