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 assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSUMASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.
|Numero di pagine||8|
|Rivista||Nuclear Engineering and Design|
|Stato di pubblicazione||Published - 2011|
All Science Journal Classification (ASJC) codes