The MAAP4 computer code (Reference 1) is often used to perform thermal hydraulic simulations of severe accident sequences for nuclear power plant Probabilistic Risk Assessments (PRAs). MAAP4 can be used to simulate accidents for both Boiling Water Reactors (BWRs) as well as Pressurized Water Reactors (PWRs). This assessment employs MAAP 4.0.6a for PWRs (References 1 and 5), which incorporates explicit thermal hydraulic modeling of the Reactor Coolant System (RCS) and Steam Generators (SGs), along with a nodalized integrated containment model. In the PRA environment, MAAP4 has been used for applications such as the development of PRA Level 1 and Level 2 success criteria and human action timings. The CENTS computer code (Reference 2) is a simulation tool that is typically used to analyze non-Loss of Coolant Accident (non-LOCA) events postulated to occur in nuclear power plants incorporating Combustion Engineering (CE) and Westinghouse Nuclear Steam Supply System (NSSS) designs. It is licensed by the NRC perform design basis non-LOCA safety analyses. It is a best estimate code which uses detailed thermal hydraulic modeling of the RCS and SGs; however, it does not model the containment performance. It is used to perform a wide spectrum of licensing and best estimate non-LOCA event analysis and has the capability to simulate operator actions. The CENTS models are the basis for several full scope simulators in the industry. The purpose of the analyses described in this paper is to compare MAAP4 and CENTS predictions for the Station Blackout (SBO) and Total Loss of Feedwater (TLOFW) scenarios for a representative PWR in the Westinghouse fleet that employs a CE NSSS design. The results of this comparison are used to highlight postulated MAAP4 user challenges and assist in developing guidance on selecting MAAP4 parameters for use in these scenarios. The results of the analyses presented in this paper indicate several useful insights. Overall, this paper shows that when care is taken to normalize the MAAP4 and CENTS primary side natural circulation flowrate and SG modeling, the trends of the MAAP4 and CENTS predictions of core uncovery agree reasonably well.

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Support Task B, the Fire Probabilistic Risk Assessment (FPRA) Database, is an important organizational task that directly supports nearly all of the NUREG/CR-6850 FPRA development tasks (Reference 1). As a result, the database structure can become quite complex. Westinghouse has created a FPRA Database to support the Wolf Creek Generating Station (WCGS) FPRA development project and has acquired a number of lessons learned and best practices that can be applied to the development of a FPRA for any nuclear power plant. The purpose of this paper is to provide an overview of the WCGS FPRA Database structure and to share the lessons learned and best practices acquired during its development.

Generic guidance for Pressurized Water Reactors (PWRs) has been developed to address the beyond design basis event of coincident loss of all Alternating Current (AC) and Direct Current (DC) power. The generic guidance included a strategy to use a low pressure feed pump to provide adequate secondary side heat removal via the Steam Generators (SGs) to delay or prevent core uncovery following loss of all AC power with battery depletion, loss of all DC power, seismic initiated events, and/or terrorist initiated events. The purpose of the project was to use thermal hydraulic analyses, operating experience, and other engineering analyses to identify and evaluate technical issues associated with the implementation of the low pressure feed pump strategy at Westinghouse and Combustion Engineering (CE) designed plants. The technical issues that were evaluated are those issues typically addressed in the development of a plant’s Emergency Operating Procedures (EOPs) and Off-Normal Operating Procedures (ONOPs). The thermal hydraulic analyses were performed using the computer code MAAP 4.0.5 and a plant model of a 4-loop Westinghouse designed PWR. The results of the analyses are also applicable to 2-Loop and 3-Loop Westinghouse and CE designed PWRs. The results of the evaluation indicated that the key technical issue potentially impacting the prevention of core uncovery for the implementation of the low pressure feed pump strategy is the potential and consequences of injecting nitrogen into the Reactor Coolant System (RCS) from the cold leg accumulators/Safety Injection Tanks (SITs). The results of the evaluation were used to develop sample instructions for implementing the low pressure feed pump strategy for Westinghouse and CE designed PWRs. The sample instructions were developed for two categories of low pressure feed pumps: (1) low pressure feed pumps with shutoff heads greater than the pressure that nitrogen injects into the RCS and (2) low pressure feed pumps with shutoff heads less than the pressure that nitrogen injects into the RCS. The usefulness of the sample instructions is maximized when the low pressure feed pump has local flow indication, throttling capabilities, deadhead protection, and local SG pressure indication is available. The results demonstrated that the design characteristics of the low pressure feed pump are important to prolonging/preventing the time of core uncovery.