A Systematic Reliability Improvement (SRI) program has been developed to support the decision-making processes of preventive maintenance (PM) planning for enhancing plant/system reliability in nuclear power plants. The SRI program is based on improved information-handling methods in the Reliability-Centered Maintenance (RCM) methodology. This program enables us to select a proper PM plan using a computerized system that integrates the following three subsystems which have been developed and is implemented on an engineering work station.(1) The maintenance management support subsystem can provide statistical parameters which indicate the characteristics of the failure mode for components based on the statistical analysis of field maintenance data. (2) The FME/CA-database subsystem can manage the system/component failure modes and their characteristics which are estimated by experts at the design stage using 13 types of assessment rankings of the failure mode effects and criticality analysis (FME/CA). (3) The PM planning support subsystem can support decision-making in determining the priority of PM improvement plans using a new method combining the above FME/CA assessment rankings and interactive logic tree analysis (1-L TA). The effectiveness of the SRI program and its support systems has been validated through a feasibility study using simulated data on a primary loop recirculation system in BWR plants.

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In September 2006, the regulatory body of Japan, the Nuclear Industry and Safety Agency (NISA), issued an interim report entitled “The improvement of the inspection system for nuclear power plants” which had been reviewed by the Subcommittee of the Advisory Committee on Nuclear and Industrial Safety. The report addresses the potential use of risk information in order to identify the safety significant inspection scope, to select and evaluate performance indicators, to evaluate the safety significance of inspection findings, and to enhance the maintenance program. NISA has been preparing for the new inspection system in Fiscal Year 2008. Before the implementation, technical bases such as the detailed design of the new inspection system and the pilot application of major issues need to be developed. The technical support of this new inspection program is now in progress by the Japan Nuclear Energy Safety Organization (JNES) to develop methodology and technical bases for improvement of efficiency and transparency of regulatory inspection, application of risk information to develop maintenance program guidelines, selection of performance indicators, identification of the safety significance of inspection findings and comprehensive evaluation of individual plants. This paper shows the development and sample calculation of significance determination process (SDP) which is one part of the new inspection program. The SDP is applied to evaluate the significance of inspection findings. The inspection findings are categorized into four groups such as the safety function facet, the risk facet, the public and occupational radiation exposure facet and the safety importance (SI) of the inspection findings are evaluated with risk information. The sample calculation with this SDP indicated that the level of SI is the same level by the current deterministic evaluation process. At present, the SDP models have been developed into the eight types of typical Japanese nuclear power plants for Boiling Water Reactor (BWR), BWR-3, BWR-4, BWR-5, Advanced BWR, 2-Loop, 3-Loop, 4-Loop Dry Containment Pressurized Water Reactor (PWR) and 4-Loop Ice-Condenser PWR.

The purpose of this paper is to provide the quantitative evaluation method with regard to the Digital Protection systems (DPSs) that were newly adopted in the ABWR plants and to show the results of level1 PSA of the ABWR plants.