Ken Muramatsu

Development of the DQFM method to consider the effect of correlation of component failures in seismic PSA of nuclear power plant

Abstract This paper presents a new calculation method for considering the effect of correlation of component failures in seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs) by direct quantification of Fault Tree (FT) using the Monte Carlo simulation (DQFM) and discusses the effect of correlation on core damage frequency (CDF). In the DQFM method, occurrence probability of a top event is calculated as follows: (1) Response and capacity of each component are generated according to their probability distribution. In this step, the response and capacity can be made correlated according to a set of arbitrarily given correlation data. (2) For each component whether the component is failed or not is judged by comparing the response and the capacity. (3) The status of each component, failure or success, is assigned as either TRUE or FALSE in a Truth Table, which represents the logical structure of the FT to judge the occurrence of the top event. After this trial is iterated sufficient times, the occurrence probability of the top event is obtained as the ratio of the occurrence number of the top event to the number of total iterations. The DQFM method has the following features compared with the minimal cut set (MCS) method used in the well known Seismic Safety Margins Research Program (SSMRP). While the MCS method gives the upper bound approximation for occurrence probability of an union of MCSs, the DQFM method gives more exact results than the upper bound approximation. Further, the DQFM method considers the effect of correlation on the union and intersection of component failures while the MCS method considers only the effect on the latter. The importance of these features in seismic PSA of NPPs are demonstrated by an example calculation and a calculation of CDF in a seismic PSA. The effect of correlation on CDF was evaluated by the DQFM method and was compared with that evaluated in the application study of the SSMRP methodology. In the application study, Bohn et al. showed that correlation had a significant effect on CDF and may vary it by up to an order. However, in the results calculated by the DQFM method correlation varied CDF by at most 2 or 3 times compared with CDF for a case where no correlation was assumed. Although some factors should further be examined, this implied that the MCS method may have overestimated the effect of correlation on CDF and the effect of correlation on CDF may not be so significant as that evaluated in the SSMRP.

Systematic Source Term Analyses for Level 3 PSA of a BWR With Mark-II Type Containment With THALES-2 Code

The THALES-2 code is an integrated severe accident analysis code developed at the Japan Atomic Energy Research Institute in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant. As part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment, a series of calculations were performed by THALES-2 to evaluate the source terms for extensive accident scenarios. For some of the containment failure modes not modeled in THALES-2, such as steam explosion, simple models were coupled with the analysis results of THALES-2 to estimate the source terms. This paper presents the methods and insights from the analyses. An insight from the analyses was that the source terms depend more strongly on the differences in the containment function failure scenarios, such as overpressure failure, controlled containment venting, and small leakage to the reactor building, than those core damage sequences.© 2002 ASME

Development of systems reliability analysis code SECOM-2 for seismic PSA

Abstract An integrated code system SECOM-2, developed at the Japan Atomic Energy Research Institute (JAERI), has the following functions for systems reliability analysis in seismic probabilistic safety assessments (PSAs): (1) calculation of component failure probability, (2) extraction of minimal cut sets (MCSs) from a given fault tree (FT), (3) calculation of frequencies of accident sequences and core damage, (4) importance analysis with several measures with consideration of unique parameters of seismic PSAs, (5) sensitivity analysis, and (6) uncertainty analysis. This paper summarizes the special features of SECOM-2 to perform the analyses mentioned above. At JAERI, using an integrated FT which represents seismically induced core damage due to all initiating events as a system model to calculate core damage frequency of a nuclear power plant, SECOM-2 can calculate conditional point estimate probabilities of system failures, losses of safety functions, and core damage as a function of earthquake motions. The point estimate is computed by a method which gives an exact numerical solution using the Boolean arithmetic model method. As for consideration of correlation of component failure, which has been an important issue in seismic PSAs, a new technique based on direct FT quantification by a Monte Carlo simulation is being added to SECOM-2. Adding this technique, the core damage frequency can be calculated not only with the upper bound approximation based on MCSs but also with a near exact solution taking into account the correlation among all components. This paper also presents the preliminary results of a seismic PSA of a generic BWR plant in Japan performed at JAERI to demonstrate the functions of the SECOM-2 code.

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

Abstract This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I): Development of Models

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include: 1. Froth level (or dry-out level) calculation 2. Transition and mixing between convection flow regimes in convective heat transfer 3. Radiant heat transfer between solid walls and flowing gas 4. Heat generation by zirconium-water reaction 5. Crucibilization effect of zirconium-oxide layer 6. Steam starvation effect on zirconium-water reaction. This code does not calculate motion of fuel rod material but predicts the beginning of relocation. The major affecting models, froth level calculation model, heat transfer model and crucibilization model, are verified through analyses of experiments. This code can be used for thermal hydraulic analysis of a severe accident and fuel damage experiment until significant material relocation occurs.

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III): Analysis of Power Burst Facility Severe Fuel Damage 1-1 Test with SEFDAN Code

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test. The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO2 during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO2 layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction o...