An Internal Flooding Probabilistic Risk Assessment (IF-PRA) study is required to meet the requirements of the Internal Flooding portions of the ASME Internal Events PRA standard [1, 2, 3] and RG 1.200 R1 [4] for risk-informed applications at Nuclear Power Plants (NPPs). An internal flooding study was performed recently for a Nuclear Power Plant (NPP), Fort Calhoun Station (FCS), with a nuclear steam supply system designed by Combustion Engineering. The study was performed using guidance that is currently being developed for industry use by EPRI [5]. The intent of the EPRI draft guidance is to describe how to perform the various tasks to meet the requirements and expectations of the ASME PRA Standard in order to comply with RG 1.200. Performing this analysis provided an opportunity to gain numerous insights on how to best perform the analysis from a technical standpoint. This paper provides lessons learned and insights gained in the implementation of the EPRI draft guidance on internal flooding for PRA. Insights and lessons learned on the assignment of initiating event frequencies, treatment of dynamic flood volumes and time-based operator response timings, and developing associated peer review materials are provided. Discussions dealing with various sources of pipe failure frequencies highlight key differences between commonly used data. Characterizations of spray scenarios were evaluated to determine their impact on plant risk caused by internal flooding events. Maintenance-induced flooding scenarios that are often neglected were evaluated to gain insights on their impact on plant risk. Insights on performing a timely completion of the analysis are provided. Optimization techniques for plant walkdowns to capture relevant information and identify high priority flood areas are discussed. The impact of the flooding scenarios on PRA-related equipment was integrated into the internal events PRA model to quantify the associated risk. The rules used to perform the integration process and the lessons learned from the integration are also discussed.

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

The U.S. Nuclear Regulatory Commission (NRC) has an ongoing Common Cause Failure (CCF) data analysis program that periodically collects and evaluates information on component failures at U.S. commercial Nuclear Power Plants (NPPs). The primary information sources include the Licensee Event Reports (LER) and records from the Equipment Performance Information Exchange (EPIX) program. Once the information is collected, the failure records are evaluated to identify potential CCF events. CCF events are then coded, reviewed, and loaded into the NRC’s database. Verification of the CCF events is performed with the intended purpose of ensuring that events entered into the CCF database are indeed CCF events and that the event coding is consistent and correct. To ensure technical accuracy and correctness of the events loaded into the CCF database, the NRC requested the Pressurized Water Reactors Owners Group (PWROG) support in reviewing these events. Reviews of multiple data sets of CCF events were conducted on behalf of the PWROG. The data sets included CCF events that have occurred at U.S. commercial nuclear power plants. CCF events that occurred during 2006 through 2007 were included in the most recent data set that was reviewed. The level of information provided for reported CCF events varies from utility-to-utility. Without utility participation or input, the lack of consistency and varying level of detail can lead to incorrect interpretation and classification of a CCF event regarding its Probabilistic Risk Assessment (PRA) impact. This paper offers lessons learned from the reviews that were conducted. Insights for improving the consistency and level of detail related to the PRA information are summarized in this paper. The leading causes of initial misclassification of CCF events and patterns observed in conducting the reviews are discussed. The resolutions of misclassified CCF events are also discussed as part of the evaluation process to enhance the pedigree of the CCF database.

Probabilistic Risk Analyses (PRAs) are increasing]y being used as a tool for supponing the acceptability of design, procurement, construction, operation, and maintenance activities at nuclear power plants. Since the issuance of Generic Letter 88-20 and subsequent lndividualPlant Examinations (IPEs) and Individual PIant Examination of External Events (IPEEEs), the US Nuclear Regulatory Commission has issued several Regulatory Guides such as RG 1,174 and RG 1.177 to describe the use of PRAs in risk-informed regulatory activities. The PRA models developed for the IPEs were typically based on common cause models