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Risk-informed in-service inspection for piping was studied for a BWR plant. Piping segment rupture probabilities were determined by Bayesian transform from piping failure events in the database of the OECD-NEA Piping Failure Data Exchange project. Based on the methodology of the Westinghouse Owners Group, core damage frequency induced by each segment rupture was determined by the use of a surrogate component in the PSA model. Nondestructive examinations were added to leak examinations for segments of the resultant high risk significance. The changes from current examinations gave around 29% reduction of segments subject to both leak and nondestructive examinations within the total segments. Deterministic insights and engineering judgments on top of risk significance should be applied to obtain the final decision of inspection methods. An extent-of-examination was studied by the adoption of the Perdue-Abramson model in the Westinghouse Owners Group methodology. The necessary leak frequency of a crack in a segment was calculated by the probabilistic fracture mechanics code PRAISE. Two segments of high risk significance showed lower or slightly higher extents-of-examination, respectively, than the current extent-of-examination. To contribute to the enhancement of the scientific rationality of piping inspections, technical knowledge was accumulated.

The reliability analysis of the digital reactor protection system (RPS) is one of the essential parts in the probabilistic safety assessment (PSA) of the advanced boiling water reactor (ABWR). In this study, the reliability model and methodology were modified to evaluate the reliability of the digital RPS installed in the Japanese ABWR plant. The hardware failure rates in the foreign data source of digital components were applied, based on the similarity of the function of the digital components. The hardware failure rates of the digital components were estimated to range from 1.0E−5 (/hr) to 1.0E−7 (/hr), according to the types of the components. The software error events and their recovery factors in the design and fabrication stages were evaluated, considering the verification and validation process provided by the Japanese industry guideline on the digital reactor protection system. Then, the software failure probability of the programmable digital component was evaluated, utilizing the probability of software error events and their recovery factors. The software failure probability was estimated to be 3.3E−7 (/demand), which was about one order higher than that of our previous estimation. These models and results were applied to evaluate the reactor trip system (RTS) and the engineered safety feature (ESF) actuation system of the ABWR plant, both of which are the subsystems of the RPS. The unavailability of the digital RTS was estimated to be the mean value of 7.2E−06 (/demand). If both an alternate rod insertion (ARI) and a manual scram were considered, the unavailability was estimated to decrease to 1.6E−09. This value was nearly equal to the mean value of the previous study, 1.1E−09 (/demand), even though the quantification model and data were considerably modified, including the software failure probability. The system unavailability of the emergency core cooling system (ECCS) was also evaluated in conjunction with the ESF actuation system, in order to investigate the effect of the model and data modification. The ECCS unavailability was estimated to be also nearly equal to the same values as the previous estimation, because the system unavailability was dominated by the unavailability of the mechanical components, such as pumps, valves, etc. The sensitivity analyses were conducted systematically, in order to evaluate the effect of the modeling uncertainty on the digital RTS unavailability. The results indicated that the unavailability of the digital RTS only changed within the range of factor 2, even though the various assumptions were used on the hardware and the software failure of the digital components.

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