Sensitivity Analysis of the MASLWR Helical Coil Steam Generator Using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and…

Analyses of the TRACE V5 capability for the simulation of natural circulation and primary/containment coupling in BDBA condition typical of the MASLWR

In the short term period the use of advanced Small Modular Reactor (SMR) is one of the most promising options for the deployment of nuclear technology. The validation and assessment of the best estimate thermal hydraulic system code TRACE against SMR thermal hydraulic phenomena is a novel effort. In this framework the use of the natural circulation database developed at the OSU-MASLWR test facility, simulating the MASLWR reactor prototype, is of interest for analyses of the TRACE code capability in predicting natural circulation and primary/containment coupled behavior in SMR. The target of this paper is to analyze the TRACE V5 capability for the simulation of natural circulation phenomena,…

Analysis of the OSU-MASLWR-003A natural circulation test by using the trace code