Analysis of a CRDM Nozzle Break LOCA Without Scram Using the U.S. NRC Coupled Code TRAC-M/PARCS
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Recently, through-wall circumferential cracks in several control rod drive mechanism (CRDM) nozzle penetrations were detected at the Oconee-3 nuclear power plant. The presence of these cracks was seen as a potential precursor to a small break loss of coolant accident. In order to assess the impact of a postulated failure of a CRDM housing, analyses were performed using the U.S. NRC coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. Although deemed highly unlikely, it was assumed that no control rods inserted in order to bound any possible reactivity transient associated with the break. The thermal-hydraulic model used to perform the study is based on an existing model of Oconee, built for PTS analysis, which models the whole plant and some of its control systems. A refined vessel model based on a TMI model was used to increase the resolution of the results and facilitate coupling to PARCS. All relevant ECCS systems were modeled and the control system allowed for, in addition to automatic actions, some assumed operator intervention. In particular, HPI throttling was modeled to maintain the hot leg subcooled at 42 +/- 7 K, and prevent an excessive amount of cold water from being injected into the system. A spatial kinetics analysis of this event was necessary because of the wide range of core conditions which occurred during the transient, from hot full power operation conditions to the cold zero power shutdown state. Analysis of the event with point kinetics and best estimate reactivity coefficients resulted in significant miss-prediction of the core power response. Conversely, the three-dimensional kinetics solution with cross section data generated over the entire range of the event led to a more accurate calculation of the power response and the overall analysis of the system transient response. This paper will describe the analysis of the control rod drive nozzle break event without scram using TRAC-M/PARCS.
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10th International Conference on Nuclear Engineering, Volume 2
