TRAC-PF1 choked-flow model

The two-phase, two-component choked-flow model implemented in the latest version of the Transient Reactor analysis Code (TRAC-PF1) was developed from first principles using the characteristic analysis approach. The subcooled choked-flow model in TRAC-PF1 is a modified form of the Burnell model. This paper discusses these choked-flow models and their implementation in TRAC-PF1. comparisons using the TRAC-PF1 choked-flow models are made with the Burnell model for subcooled flow and with the homogeneous-equilibrium model (HEM) for two-phae flow. These comparisons agree well under homogeneous conditions. Generally good agreements have been obtained between the TRAC-PF1 results from models using the choking criteria and those using a fine mesh (natural choking). Code-data comparisons between the separate-effects tests of the Marviken facility and the Edwards' blowdown experiment also are favorable. 10 figures

TRAC-PD2 modeling of a cold-leg 0. 002-M/sup 2/ break in Babcock and Wilcox, Combustion Engineering, and Westinghouse pressurized water reactors

TRAC-PD2 modeling of a cold-leg small break is presented for Babcock and Wilcox (B and W), Combustion Engineering (CE), and Westinghouse (W) pressurized water reactors. Similar break size and safety-systems availability were assumed for all three plants. These calculations are presented and compared to show the differences in plant transient response for similar break and safeguard assumptions. Plant modeling assumptions are discussed and calculational results are compared, such as system depressurization, loop voiding, core cooling, and break and safety-injection flow. The comparison shows that the CE and W transients were similar, but that the B and W transient differed from the other two transients, primarily because of vessel vent valves in the B and W plant design. For all three plants, core cooling was adequate even with availability of only half the safety-injection and auxiliary-feedwater systems. 16 figures

Effect of reactor coolant pumps following a small break in a pressurized water reactor

Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the optimum mode of pump operation is design dependent. In terms of primary system mass depletion, the case with no pump trip was preferable for the W plant, whereas an early pump trip was preferable for the B and W plant. When the pumps were not operating in the W plant, the loop seals plugged with liquid, leading to a pressure buildup in the upper plenum and, consequently, a high liquid flow through the break. The vent valves in the B and W plant mitigated the consequences of the loop seals plugging; the effect was enough to favor an early pump trip

Rapid-response analysis of the Davis-Besse loss-of-feedwater event on June 9, 1985

At the request of the US Nuclear Regulatory Commission (NRC), we performed a rapid-response analysis of the loss-of-feedwater (LOFW) event that occurred at the Toledo Edison Davis-Besse plant on June 9, 1985. The initial 831 s of the plant transient were simulated and, in addition, four postulated transients were calculated to determine how the plant would have responded if feedwater had not been restored, and how it would have responded to a feed-and-bleed (FAB) decay-heat removal procedure initiated at different times. The Transient Reactor Analysis Code was used for this analysis. We completed these calculations within a two-week period and provided a report to the NRC 30 days later. Our analysis showed that FAB was a viable decay-heat-removal procedure for the Davis-Besse plant for the initiation times analyzed. With complete LOFW and no alternate decay-heat removal procedure, we calculated that core uncovery would have occurred at about 9200 s. FAB initiated at 8 min and 13 min after complete LOFW, which in the actual LOFW transient occurred 6 min after the initiating event of a main-feedwater pump trip, was successful in that the primary system remained subcooled and water-solid throughout the transient. FAB initiated at 28 min after complete LOFW was considered successful even though it resulted in a loss of subcooling, a net loss in primary inventory, and a slow voiding of the primary system. The core would have remained covered for the nine hours that we estimated for the primary pressure to decrease to the residual-heat-removal pressure