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Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration.
Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration.

RELAP-7 is a nuclear reactor system safety analysis code. Development started in October 2011, and during the past quarter the initial capabilities of RELAP-7 were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels (Fig. 1). The PWR configuration matched that of the Three Mile Island 1 LWR, which is a benchmark problem from the Organisation for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA). Although the simulation used a simplified system model, the resolution was consistent with the level of detail in current system safety analyses. [INL]

The simulation begins as a transient problem and progresses toward a converged steady state problem. The simulation results match the benchmark data very closely, within about 0.1 K of the benchmark reactor inlet and outlet temperatures. [INL]

Parallel to RELAP-7 development for LWR analyses, component models are being developed for advanced sodium-cooled fast reactor (SFR) concepts. Several test SFR models have been developed based on the Advanced Burner Test Reactor (ABTR) conceptual design. Models for a five-channel core, primary loop, intermediate loop, and entire plant have been built. Steady-state parameters for the test problems agree well with expected ABTR operating conditions. [ANL]

During the development of the ABTR test problems, some issues in the RELAP-7 component models were identified due to the very different operating conditions of an SFR and LWR. Most notably, when a sinusoidal power distribution was applied the predictions of the core channel inlet temperatures were not consistent with the upstream temperatures of other components, and the flow area for the secondary side of the heat exchanger component was incorrect. These issues are being corrected. [ANL, INL]