Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than tomore » the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in EREC calculation is also more uniform than in the experiment. Therefore, both teams concluded that there was only minimal agreement between the calculated and measured mixture level and entrainment in the core. Some changes in input modeling which can improve the prediction of the core void distribution were determined by the INEEL analyst. Application of these findings to the full size plant need to consider that PSBVVER has only one simulated fuel bundle, and as a result there may be scaling issues that need to be addressed. (authors)« less

Severe accident natural circulation flows have been investigated at the Idaho National Engineering Laboratory to better understand these flows and their potential impacts on the progression of a pressurized water reactor severe accident. Parameters affecting natural circulation in the reactor vessel and hot legs were identified and ranked based on their perceived importance. Reviews of the scaling of the 1/7-scale experiments performed by Westinghouse were undertaken. RELAP5/MOD3 calculations of two of the experiments showed generally good agreement between the calculated and observed behavior. Analyses of hydrogen behavior in the reactor vessel showed that hydrogen stratification is not likely to occur, and that an initially stratified layer of hydrogen would quickly mix with a recirculating steam flow. An analysis of the upper plenum behavior in the Three Mile Island, Unit 2 reactor concluded that vapor temperatures could have been significantly higher than the temperatures seen by the control rod drive lead screws, supporting the premise that a strong natural circulation flow was likely present during the accident. SCDAP/RELAP5 calculations of a commercial pressurized water reactor severe accident without operator actions showed that the natural circulation flows enhance the likelihood of ex-vessel piping failures long before failure of the reactor vessel lower head.

The Advanced Test Reactor (ATR) is a 250-MW material test reactor operated at the Idaho National Engineering Laboratory for the Department of Energy. Mechanistic severe-accident analyses are being performed to improve understanding of the conservatisms in the safety basis of the ATR. The computer codes available for severe-accident analyses were developed for commercial light water reactors. New models were developed to describe some of the unique features of this reactor. The best-estimate codes were then used to analyze a hypothetical transient from initiation through off-site consequences in order to demonstrate and assess their capability to simulate ATR severe accidents. The transient investigated was a loss-of-coolant accident (LOCA) in which the inlet header tee fails, resulting in two 0.61-m-diam breaks on the reactor vessel inlet side and one 0.91-m-diam break on the primary coolant pump discharge side. A comprehensive analysis was performed, including the primary coolant system thermal-hydraulic and core damage behavior, in-core recriticality potential, fission product release and transport, confinement response, and atmospheric dispersion and off-site dose calculations.

How much energy is removed from the core and where it is deposited are important considerations in severe accidents. The core heatup rate will affect the timing of the damage progression and the nature of the core debris. Heat transferred from the core to other structures in the reactor coolant system can affect the structural integrity of the piping and the fission product transport and retention. The SCDAP/RELAP5 computer code has been used to simulate a hypothetical station blackout transient in the Surry nuclear power plant. The natural circulation flows present during this high pressure sequence transferred energy from the core to other structures in the reactor coolant system. Heating of the piping in the coolant loops could lead to the creep rupture failure of the pressurizer surge line prior to melt-through of the reactor vessel lower head. Studies of insulation performance indicate that the piping temperatures are not very sensitive to the boundary conditions on the outer surface.

The statement that RCP parameters can be used to measure the local fluid density and to identify transient type, have been verified and extended to commercial PWRs using extensive LOFT and test pump data for void fractions up to approx. 0.3. It has been further demonstrated that the local density is a conservative (low) measure of the density of the primary coolant system. An exception to this may exist as indicated by the calculation of commercial PWR operations wherein the average primary system void fraction started to increase prior to that at the RCP inlet. Thus, RCPs can be used to identify the transient type and to monitor the recovery from the transient.

Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)) commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected.