Reliability Enhancement of Seismic Risk Assessment of NPP as Risk Management Fundamentals Part I: Uncertainty Analysis with the SECOM2 Code

After the severe accident in Fukushima dai-ichi nuclear power station, safety improvement and enhancement have been installed. In mid term and long term, continuous efforts to improve and enhance safety are required, and technical basis and fundamentals are needed to achieve them. Probabilistic Risk Assessment for seismic event (Seismic PRA) is an effective measure to consider the countermeasures and improvement plans to secure the further safety of nuclear power plants regarding to seismic risk for the earthquake exceeding the reference ground motion. However, the application of the seismic PRA has not been utilized sufficiently so far. One of the reasons is that there is not enough consensus among stakeholders regarding to the evaluation methods and consideration of uncertainty for decision-making. This study proposes the mathematic framework to treat the uncertainty properly related to the evaluation of Core Damage Frequency induced by earthquake, the method to evaluate the fragility utilizing expert knowledge, the probabilistic model to cope with the aleatory uncertainty as well as the development of analyzing code including these considerations for the improvement of the reliability of the method and enhancement of utilization of the products of Seismic PRA. 1. BACKGROUND After the Fukushima dai-ichi accident, the importance of seismic probabilistic risk assessment (Seismic PRA) as a tool to identify potential accident scenarios caused by earthquakes, to estimate their likelihood and consequences and to support assessing the effectiveness of measures to enhance safety against earthquakes has been widely and strongly recognized. As the use of Seismic PRA expands, the quantification and reduction of uncertainties in numerical results of Seismic PRA is becoming more and more important. This study, focusing on uncertainty assessment framework and utilization of expertise, and finally to improve reliability of seismic probabilistic risk assessment (Seismic PRA) by developing relevant computer codes and to promote its further use of the SPRA, develops methodology for quantification of uncertainty associated with final results from Seismic PRA in the framework of risk management of NPP facilities. In this study, a new mathematical framework of Seismic PRA is proposed. Reviewing the current status of assessment procedures of accident sequence analysis in Seismic PRA, this study will develop a new mathematical framework for estimating uncertainty in SPRA results in a more comprehensive way taking into account uncertainties related to correlation effect of components failures which has been difficult to quantify so far. A computer code will be developed to materialize the proposed framework on the basis 23 rd Conference on Structural Mechanics in Reactor Technology Manchester, United Kingdom August 10-14, 2015 Division VII 2 of the SECOM2-DQFM developed by JAEA to estimate the accident sequence occurrence probability and its uncertainty. The proposed mathematical framework is characterized by the following points; • Representation of seismic hazard by a set of time histories of seismic motions using methods currently being developed by Nishida et. al., • Use of probabilistic response analysis by three dimensional building model for determining responses of components to the seismic motions including the correlations among the component responses, • Use of Monte Carlo simulation for quantification of fault trees in accident sequence analysis, and • Use of high performance computing technology for realizing the use of above technologies in Seismic PRA. Current status and some results from scoping calculations will be presented. 2. CURRENT FRAMEWORK AND CHALLENGES OF Seismic PRA METHODOLOGY 2.1 Current Method of Seismic PRA (1) General Procedure of Seismic PRA This study focuses the method of level 1 Seismic PRA that evaluates the frequency of core damage accident. In general, the basic procedures of level 1 Seismic PRA can be shown as followings; a. Collecting the plant information and analyzing brief accident scenarios To investigate the seismic source around the target site, characteristics of soil and structures, and safety system configuration, the brief accident scenarios induced by earthquakes are extracted. b. Seismic hazard analysis Based on the information about faults around the target site and historical earthquake, occurrence frequencies of seismic ground motion exceeding a certain capacity such as maximum ground acceleration. c. Fragility analysis To analyze the response and capacity of structures and components, the failure probabilities of structures and components can be expressed as fragilities i.e. the function of capacity of seismic ground motion. d. Accident sequence analysis To analyze seismic induced core damage accident sequences using event-tree (ET) and fault-tree (FT) techniques, core damage frequencies are evaluated based on these accident sequences, results of hazard analysis and fragility analysis. (2) Mathematical Framework of Current Method In this study, focusing on above item c. and d., mathematical model considering uncertainties of components and system failures will be studied. The mathematical framework for evaluating frequencies of accident sequences of Seismic PRA are as followings; • The results of hazard analysis will be expressed as exceeding probabilities that is occurrence frequencies of seismic ground motions depending on the capacity on the target site. The levels of seismic ground motions are expressed as maximum accelerations of the surface. • The wave used for response analysis is one of the time histories of waves such as design basis seismic ground motion. The impacts of variability of ground motion spectra are considered as variability of response factors explained later. • The fragilities of components can be expressed as the probability that response exceeds capacity of the components, based on the assumption that probability distributions of response and capacity depending on the levels of seismic ground motion are the log-normal distribution respectively. • The median values of response depending on the seismic level are evaluated by linear extrapolation for the component response results associated with design basis seismic motion or interpolation for the results associated with several seismic ground motion. 23 rd Conference on Structural Mechanics in Reactor Technology Manchester, United Kingdom August 10-14, 2015 Division VII 3 • Standard deviations on the log scale for the response can be evaluated by expert opinion based on the results of the similar response analysis or comparison among observation points. Usually response can be analysed by the Sway-rocking model. • Since responses are usually analysed based on the design basis framework, response factors are introduced to consider impacts included in the assumption to secure conservatives of the design and to describe impacts of the uncertainty of model or data. • Component capacities are expressed by median value and standard deviation, these parameters are set based on the results of structural analysis or verification test, and, if necessary, expert opinion. • Occurrence conditions of accident sequences are expressed as groups of minimal cutsets (MCS) equivalent to logical expression of accident conditions expressed by ET and FT. To calculate occurrence probabilities of these MCSs, the probability of certain accident sequence can be evaluated associated with the certain level of seismic ground motion. • Core damage frequencies can be evaluated by the integration of the product of the probability of accident sequence associated with the certain ground motion level and seismic frequencies all over seismic ground motion levels. 2.2 Studies about Uncertainty Analysis Framework (1) Mathematical Framework of Current Method Current method was proposed to evaluate component failure probabilities by Kennedy et. Al. in 1980. The characteristics are as followings; • Uncertainty of seismic hazard is expressed by the fractile curves that are composed of multiple curves corresponding to the percentage of the confidence level. • Main causes of variability of model and data expressing response and capacity can be categorized to “aleatory uncertainty” (or “uncertainty due to randomness”) and “epistemic uncertainty” (or “uncertainty due to lack of knowledge”). The first one can not be reduced by the insights of experiments or theoretical study because this variability actually exists and means that natural phenomena are essentially varies with randomly. The second one can be reduced by the insights of expansion of experimental data and enhancement of analysis model because this variability comes from lack of knowledge or simplification of analysis model. • Usually uncertainty of analysis model of accident sequence are considered by sensitivity studies. (2) Issues of Current Mathematical Framework Seismic PRA is expected that it can provide insights and information for the quantitative evaluation of the safety level comparing core damage frequency and safety goal, and extraction of important accident sequences in a viewpoint of contribution to the total risk to enhance the safety features and accident countermeasures. So the followings are desirable and these are enhanced after Fukushima dai-ichi accident. • To evaluate core damage frequency as far as precisely and its uncertainty. • Plant damage states should be analysed in detail. For example, how many systems failed simultaneously, how many structures such as buildings or pipes failed or how they failed? In the multiple units site, what are the impacts of simultaneous occurrence of accidents in the multiple units? However, current Seismic PRA method has several issues described later, and many of them are tightly related to the simplification in the mathematical framework described above, and are shown as followings. a. Issues Mainly Related to the Hazard Analysis