This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for esbunating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculati~on capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3.A point estimate of overall DEGB pipe break frequency (per Ih-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NI'SH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.
The emergency core cooling system (ECCS) and containment spray system (CSS) in a pressurized water reactor (PWR) are designed to safely shutdown the plant following a loss of coolant accident (LOCA). The assurance of long term core cooling in PWRs following a LOCA has a long history dating back to the NRC studies of the mid 1980s associated with Unresolved Safety Issue (USI) A-43. Results of the NRC research on boiling water reactor (BWR) ECCS suction strainer blockage of the early 1990s identified new phenomena and failure modes that were not considered in the resolution of USI A-43. As a result of these concerns, Generic Safety Issue (GSI) 191 was identified in September 1996 related to debris clogging of the ECCS sump suction strainers at PWRs. Although plants have taken steps to prevent strainer clogging (by increasing the screen area, for example), satisfactory closure of this issue has proved elusive due to long term cooling issues and the effect of chemical precipitates on head loss. Previous investigators have identified bounding scenarios using conservative inputs, methods, and acceptance criteria. The acceptance criteria are applied in a “pass/fail” fashion that ignores risk. That is, if the results are acceptable, the issue has been resolved. Otherwise, it is necessary to either redo the analysis with partial relaxation of analytical conservatisms or perform additional plant modifications to ensure that the acceptance criteria are met. This article describes a new approach to close out the GSI-191 issue by evaluating the risk associated with ECCS performance on post-LOCA core cooling as a basis to change the plant license. The approach includes an assessment of LOCA frequencies as a function of break size at locations along the reactor coolant system, as well as a full quantification of the uncertainties associated with LOCA frequencies and the generation, transport, accumulation, and subsequent impact of debris on ECCS performance. The overall frameworks for the deterministic and risk-informed approaches are summarized with emphasis on the risk-informed method. The differences between the deterministic approach taken in the past and the new risk-informed approach are described. Advantages and disadvantages between the two methods are described and contrasted for the GSI-191 issue. The South Texas Project (STP) GSI-191 risk-informed closure efforts are presented.
NRC ABSTRACTA central question for resolution of GSI-24 is whether or not PWRs that currently rely on a manual system for ECCS switchover to recirculation should be required to install an automatic system. Risk estimates are obtained by reevaluating the contributions to core damage frequencies (CDFs) associated with failures of manual and semiautomatic switchover at a representative PWR. This study considers each separate instruction of the corresponding emergency operating procedures (EOPs), the mechanism for each control, and the relationship of each control to its neighbors. Important contributions to CDF include human errors that result in completely coupled failure of both trains and failure to enter the required EOP.This detailed study finds that changeover to a semiautomatic system is not justified on the basis of cost-benefit analysis: going from a manual to a semiautomatic system reduces the CDF by 1.7~10-~ per reactor-year, but the probability that the net cost associated with the modification being less than $1,000 per person-rem is about 20% without license renewal. Scoping analyses, using optimistic assumptions, were performed for a changeover to a semiautomatic system with automatic actuation and to a fully automatic system; in these cases the probability of having a net cost being less than $l,OOO/person-rem is about 50% without license renewal and over 95% with license renewal.
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