John T. Groome

Testing of Passive Safety System Performance for Higher Power Advanced Reactors

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

Aircraft Impact Considerations for NuScale SMR Plant Design

All new nuclear power plants to be constructed and operated in the United States must meet regulatory requirements for aircraft impact from a large commercial aircraft under 10CFR50.150. Under the regulation, the applicant, using realistic analyses, must identify and incorporate into the design those design features and functional capabilities to show that, with reduced use of operator actions, 1) either the primary containment system remains intact or the reactor core remains cooled, and 2) either spent fuel cooling or spent fuel pool integrity is maintained. Small modular reactors have both advantages and disadvantages over conventional large plant designs in this regard. Small modular reactors generally have smaller footprints and can be configured where the reactor vessels and containment systems are entirely below grade. This minimizes the exposed structure that houses the reactors and spent fuel pool, which generally means that the structural configuration can be more efficiently hardened to resist the impact forces without excessive costs. However, the smaller footprint also means that transmission of shock through the structure can affect more safety equipment than in the larger conventional plant where the safety related equipment for the divisions are physically farther apart. Modular designs by nature tend to have all associated safety equipment together for each reactor module. Larger plants may be more tolerant for allowing internal damage and controlling ensuing fire due to perforation of aircraft wreckage at some strike locations, whereas the smaller footprints for small modular reactors could mean more systems are at risk if the reactor building is not hardened to prevent perforation. This paper presents design considerations employed for the NuScale 12 Module Power Plant in regards to aircraft impact requirements.Copyright

ICONE15-10026 Status of the INL Gas Reactor Test System Experiment Facility

The Gas Reactor Test System (GRTS) is an experiment facility for examining the thermal hydraulic performance of the Generation IV, Very High Temperature Reactor (VHTR) during a LargeBreak Loss of Coolant Accident (LB-LOCA). The LB-LOCA is defined as the double guillotine break of the VHTR coaxial inlet and outlet cross duct. Two system safety codes, MELCOR and RELAP5-3D were used to calculate core temperatures and flow rates during the LB-LOCA transient. Computational fluid dynamics modeling of the transient produced flow vectors and gas species distribution. The most important phenomenon during the transient is the lock-exchange process, which suppresses the onset of natural circulation until considerable molecular diffusion has occurred. The GRTS was designed based upon a hierarchical two tier scaling analysis whose primary objective was replicating the lockexchange and natural circulation characteristics of the VHTR. The GRTS uses a scaled graphite core to represent the VHTRs graphite core. An in-depth scaling analysis was performed for the GRTS in order to ensure that it accurately simulated the VHTR thermal responses. RELAP5-3D thermal analyses, ProEngineer stress analyses, and combined FLUENT – STARCD CFD analyses have provided a system design that fulfills the GRTS mission statement. This paper discusses the design analyses and their implications on the GRTS capabilities. A discussion is also presented on the preliminary instrumentation plan. The GRTS will provide an extensive temperature map of the VHTR core outlet plenum and its core support, oxygen transport rates during the lock-exchange phenomenon, and thermal conduction rates from the core to the vessel. As a result of the GRTS using helium coolant at 950 C, the resulting experiment data is expected to considerably extend the U.S. database for hightemperature gas reactor operations. Finally, the discussion will present conclusions from the GRTS manufacturing and quality control processes that may benefit the VHTR design.