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.

To promote the practical fire Probability Risk Assessment (PRA) among Japanese industries based on the domestic fire experience, it is important to estimate generic fire ignition frequency distributions for Japanese nuclear power plants (NPPs), which are applicable to the fire PRA during full power operation. Firstly, the Central Research Institute of Electric Power Industry (CRIEPI) proposed the estimation process of the distributions, which consists of the four steps including sub steps as follows: Step 1: Fire event data collection including development of criteria Step 2: Fire event data analyses including development of criteria: Step 2.1: Screening of fire events to identify events to be classified Step 2.2: Classification of fire events from the viewpoint of fire severity of each event. Step 2.3: Categorization of fire events from the viewpoint of fire ignition source types Step 3: Estimation of generic fire ignition frequency distributions Step 4: Comparison of the results of Step 3 with fire ignition frequency estimated in other countries In this paper, CRIEPI introduced the detail of Step 1 and 2 above and discussed the study results for them, which are essential to estimate generic fire ignition frequency distributions properly. Through Step 1, CRIEPI established a preliminary fire event database based on fire records at Japanese NPPs provided by domestic utilities. In step 2.1, CRIEPI identified the 180 fire events for future steps. Finally, CRIEPI classified the fire events and defined the fire ignition source categories in Step 2.2 and 2.3. As a result, CRIEPI found the major three risk contributors among fire ignition source categories: transient related fires, electrical cabinet fires and pump fires.

No abstract

After the severe accident at the Fukushima-Daiichi nuclear power station, the regulation of nuclear plant safety in Japan was upgraded, and continuous effort to enhance risk management in the medium and long term is required. Seismic risk is important in Japan, and one of the major issues of current probabilistic risk assessment （PRA） methodology is the enhancement of seismic PRA including the evaluation of initiating events induced by simultaneous multiple failures of components and of subsequent accident sequences. The Nuclear Regulation Authority （NRA） addressed the issue as "…multiple-failure-initiating events should be considered." In addition, the standard for the procedure of seismic PRA for nuclear power plants in 2015 identifies this issue in the same manner. This study proposes an enhanced seismic PRA methodology considering multiple-failure-initiating events that is expected to contribute to improving the safety of nuclear power plants. This study also discusses the issues that need more research for their resolution.

No abstract