Analysis of the Presence of Vapor in Residual Heat Removal System in Modes 314 Loss-of-Coolant Accident Conditions Using RELAP5

The Westinghouse Nuclear Safety Advisory Letter NSAL-09-8 investigated the possibility of presence of vapor in RHR （residual heat removal） system in modes 3/4 LOCA （loss-of-coolant accident） conditions. This concerns the Westinghouse standard three-loops plant for which the RHR is the low pressure part of the St （safety injection）. In some cases one or both RHR trains may become inoperable for SI function. As a response to this letter, Westinghouse Electric Belgium is providing RELAP5 analyzes for Westinghouse NSSS （nuclear steam supply system） European plants to assess the thermal hydraulic behavior of the RHR suction piping system for ECCS （emergency core cooling system） initiation events postulated to occur during startup/shutdown operations. Several concerns including condensation induced water hammer and voiding at the RHR pump have been investigated. As a conclusion, the analysis allowed to define the bounding hot leg temperature conditions under which both RHR trains remain safely operable. These bounding conditions are then implemented by the customer in their OPs （operating procedures） to achieve safe operations and successful accident management.

A transient fiuid–structure interaction analysis strategy and validation of a pressurized reactor with regard to loss-ofcoolant accidents

A loss-of-coolant accident(LOCA)is one of the basic design considerations for nuclear reactor safety analysis.A LOCA induces propagation of a depressurization wave in the coolant,exerting hydrodynamic forces on structures viafiuid–structure interaction(FSI).The analysis of hydrodynamic forces on the core structures during a LOCA process is indispensable.We describe the implementation of a numerical strategy for prestressed structures.It consists of an initialization and a restarted transient analysis process,all implemented via the ANSYS Workbench by system coupling of ANSYS and Fluent.Our strategy is validated by making extensive comparisons of the pressures,displacements,and strains on various locations between the simulation and reported measurements.The approach is appealing for dynamic analysis of other prestressed structures,owing to the good popularity and acknowledgement of ANSYS and Fluent in both academia and industry.

RELAP5 Code Study of ROSA/LSTF Experiments on PWR Safety System Using Steam Generator Secondary-Side Depressurization

RELAP5 （reactor excursion and leak analysis program, version 5） code analyses were performed on two ROSA/LSTF （rig of safety assessment/large scale test facility） experiments on PWR （pressurized water reactor） safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG （steam generator） secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by core boil-off. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. In the 4-in. break test, on the other hand, there was no core uncovery and heatup due to smaller break flow rate than in the 8-in. break test. Adjustment of Cd （break discharge coefficient） for two-phase discharge flow predicted the break flow rate reasonably well. The code well predicted the overall trend of the major thermal-hydraulic response observed in the two LSTF tests by the Cd adjustment. The code, however, overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.