The objectives for the ASP large early release frequency (LERF) model development work is to build a Level 2 containment response model that would capture all of the events necessary to define LERF as outlined in Regulatory Guide 1.174, can be directly interfaced with the existing Level 1 models, is technically correct, can be readily modified to incorporate new information or to represent another plant, and can be executed in SAPHIRE. The ASP LERF models being developed will meet these objectives while providing the NRC with the capability to independently assess the risk impact of plant-specific changes proposed by the utilities that change the nuclear power plants' licensing basis. Together with the ASP Level 1 models, the ASP LERF models provide the NRC with the capability of performing equipment and event assessments to determine their impact on a plant's LERF for internal events during power operation. In addition, the ASP LERF models are capable of being updated to reflect changes in information regarding the system operations and phenomenological events, and of being updated to assess the potential for early fatalities for each LERF sequence. As the ASP Level 1 models evolve to include more analysis capabilities, the LERF models will also be refined to reflect the appropriate level of detail needed to demonstrate the new capabilities. An approach was formulated for the development of detailed LERF models using the NUREG-1150 APET models as a guide. The modifications to the SAPHIRE computer code have allowed the development of these detailed models and the ability to analyze these models in a reasonable time. Ten reference LERF plant models, including six PWR models and four BWR models, which cover a wide variety of containment and nuclear steam supply systems designs, will be complete in 1999. These reference models will be used as the starting point for developing the LERF models for the remaining nuclear power plants.

Severe accident natural circulation flows have been investigated at the Idaho National Engineering Laboratory to better understand these flows and their potential impacts on the progression of a pressurized water reactor severe accident. Parameters affecting natural circulation in the reactor vessel and hot legs were identified and ranked based on their perceived importance. Reviews of the scaling of the 1/7-scale experiments performed by Westinghouse were undertaken. RELAP5/MOD3 calculations of two of the experiments showed generally good agreement between the calculated and observed behavior. Analyses of hydrogen behavior in the reactor vessel showed that hydrogen stratification is not likely to occur, and that an initially stratified layer of hydrogen would quickly mix with a recirculating steam flow. An analysis of the upper plenum behavior in the Three Mile Island, Unit 2 reactor concluded that vapor temperatures could have been significantly higher than the temperatures seen by the control rod drive lead screws, supporting the premise that a strong natural circulation flow was likely present during the accident. SCDAP/RELAP5 calculations of a commercial pressurized water reactor severe accident without operator actions showed that the natural circulation flows enhance the likelihood of ex-vessel piping failures long before failure of the reactor vessel lower head.

Following the Three Mile Island accident, a number of suggestions were offered for improving the safety of nuclear power plants. Many of these approaches involve increased dependence on complex and sophisticated safety systems and control technologies. An alternative approach to increased safety is explored; this approach involves the maximization of inherent safety features and minimization of safety and control system action. Specifically, the effects of changing pressurizer size and surge line configuration are examined to determine if the operation of the power-operated relief valve (PORV) could be eliminated during a turbine trip transient in a pressurized water reactor. The study is accomplished using the RETRAN-02 thermal-hydraulic analysis code. The results show that the opening of the PORV could be eliminated if the pressurizer size was doubled and the PORV setpoint increased slightly.

Since the Three Mile Island Accident of March 1979, various approaches have been undertaken to reduce the probability of such occurrences. In general, these involve the addition of more complex and sophisticated engineered safety systems and instrumentation and control systems. Such improvements will undoubtedly improve overall plant safety. This project explores an alternative approach - the optimization of the transient behavior of the reactor and associated systems. Specifically, the ability of B and W and Westinghouse reactor to withstand a turbine trip was investigated. The project examined whether the plant's transient behavior with respect to peak reactor coolant system (RCS) behavior could be optimized. Several key parameters including pressurizer size, surge line impedence, time to SCRAM, and fuel and moderator temperature coefficients were varied. It was found that increasing pressurizer size decreased peak RCS pressure.