Hideaki Asaka

Results of 0.5% cold-leg small-break LOCA experiments at ROSA-IV/LSTF: Effect of break orientation

Abstract Three 0.5% cold-leg small-break loss-of-coolant accident (SBLOCA) experiment were conducted at the ROSA-IV Large Scale Test Facility (LSTF) to investigate the effects of break orientation on system thermal-hydraulic responses. In these three experiments, the break hole was located at the side, bottom, and top of the horizontal cold leg, respectively. Although the key phenomena observed in the three experiments were basically the same, the break flow rate was affected by the break orientation when phase stratification occured in the cold leg; the break flow rate was largest for the side break and smallest for the top break. The RELAP5/MOD2 code failed to predict the difference in the break flow rate observed in the experiments. Modification to the break flow calculation models, for both subcooled and two-phase flow discharge conditions, resulted in good agreement between data and predictions.

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

Abstract This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.

RELAP5/MOD3 Code Analyses of LSTF Experiments on Intentional Primary-Side Depressurization Following SBLOCAs with Totally Failed HPI

The effectiveness of intentional depressurization ofa pressurized water reactor primary system as a means to maintain core cooling during a small-break loss-of-coolant accident (SBLOCA) was studied. The investigation was based on experiments conducted at the Rig of Safety Assessment-V (ROSA- V) Large Scale Test Facility (LSTF) and RELAP5/MOD3 code calculations performed for LSTF geometry, together with single lumped-volume model calculations-all simulating hypothetical total failure of the high-pressure-injection system. For cold-leg breaks ≥2.5% of the leg cross-sectional area, experimental and analytical results have shown that the break discharge depressurizes the primary system to the accumulator (ACC) and low-pressure-injection (LPI) system injection pressures, and thus the core cladding temperature would be maintained below ∼1000 K. For break areas ≤1.0%, on the other hand, additional depressurization means are needed to initiate the ACC injection before the core is overheated. RELAP5/ MOD3 calculations have shown that steam venting through the pressurizer power-operated relief valves would be effective in depressurizing the primary system to the ACC and LPI pressures. However, for break areas <0.5%, the peak cladding temperature would finally reach the safety criterion of 1473 K.

RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW) transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW) was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG) secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR) power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

Secondary-Side Depressurization during PWR Cold-Leg Small Break LOCAs Based on ROSA-V/LSTF Experiments and Analyses

The effectiveness of an operator-initiated steam generator (SG) secondary-side depressurization on the core cooling performance during small-break loss of coolant accidents (SBLOCAs) in a pressurized water reactor (PWR) with total failure of the high pressure injection (HPI) systems is studied. The study is based on experiments conducted in the ROSA-V Large Scale Test Facility (LSTF) and analyses with the RELAP5/Mod3 code. The sensitivity of the core minimum liquid level and peak cladding temperature (PCT) to the secondary-side depressurization rate and the initiation time of the depressurization is evaluated analytically for various break sizes. It is shown that the PCT takes a maximum value for break areas between 1.0% and 1.5% of the cold leg cross-sectional area. The conditions which the depressurization rate and the initiation time should satisfy to limit the maximum PCT are derived.

Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-of-Coolant Accident

The consequence of intentional depressurization of the steam generator (SG) secondary side during a small break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR) was studied using the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1:48 volumetrically scaled full-height model of a Westinghouse-type PWR. The experiment simulated a 0.5% cold leg small-break LOCA with total failure of high pressure injection (HPI) system. The SG secondary-side atmospheric relief valve (ARV) was latched open after the core top region was uncovered. Then, the primary pressure closely followed the secondary pressure and dropped to the accumulator (ACC) injection pressure of 4.51MPa before the core became severely overheated. A post test analysis, which was performed using the RELAP51 MOD2 code with modifications made by the authors, predicted well the primary and secondary side responses during this experiment.

New Non-Equilibrium Thermal-Hydraulic Model, (II)

A through analysis of LOFT L2-3 is made with the THYDE-P code, which is based on the new non-equilibrium thermal-hydraulic model. The various assumptions and correlations in use for the present analysis are explained. In order to investigate DNB (departure from nucleate boiling), rewetting and quenching at the pool condition, several parameters are defined. The interfacial heat transfer coefficient between the gas and the liquid is assumed to have the same pressure dependence as of nucleate boiling. The overall trends of the experiment are well reproduced by the calculation. Much effort was found necessary, however, to better understand DNB, rewetting and quenching at the pool condition, by taking into account the stored thermal energy and the stationariness of the temperature distribution in the fuel rod.

Effects of Secondary Depressurization on Core Cooling in PWR Vessel Bottom Small Break LOCA Experiments with HPI Failure and Gas Inflow

The effects of steam generator(SG) secondary depressurization, as one of accident management measures during a beyond-design-basis loss-of-coolant accident(LOCA) at pressurized water reactor(PWR), are experimentally investigated at the Large Scale Test Facility(LSTF) which simulates a 4-loop PWR by full-height and 1/48 volumetric scaling. Multiple instrument-tube break at the reactor vessel bottom, equivalent to 0.2% cold leg break, was simulated in the experiments with high pressure injection(HPI) system failure and non-condensable gas inflow from the accumulator injection system(AIS). The experiments showed that secondary depressurization to achieve primary system cooling at a rate of −55 K/h was sufficient to achieve core cooling by the low pressure injection(LPI) system when the gas inflow was prevented, while it was insufficient for core cooling in case of the gas inflow. The rapid secondary depres- surization, however, was successful to achieve core cooling by the LPI actuation irrespective of the AIS gas inflow because it contributed to lower the primary pressure and conserve primary coolant mass as shown in a Pressure-Mass map. The RELAP5/MOD3 code analyses well predicted these thermohydraulic phenomena including SG heat transfer coefficient which was significantly degraded by the gas accumulation.

Pressurized water reactor core instrument tube ruptures: experimental simulation at the ROSA-IV lstf

Abstract A small-break loss-of-coolant accident (SBLOCA) initiated by the rupture of pressurized water reactor (PWR) core instrument tubes was simulated using the ROSA-IV Large Scale Test Facility (LSTF). The experimental results were characterized by a single-phase liquid discharge continuing for a long period until the pressure vessel coolant inventory was significantly depleted. This experiment was analyzed using three advanced LOCA analysis codes, RELAP5/MOD2, TRACPF1/MOD1 and CATHARE-1, to assess the predictive abilities of these codes. Although all these codes simulated the experimental results qualitatively, discrepancies were found between the predictions and data regarding the break flowrate, spatial distribution of coolant in the primary system and core rod temperature responses.

Analysis of system thermal hydraulic responses for passive safety injection experiment at ROSA-IV/large scale test facility : using JAERI modified version of RELAP5/MOD2 code

An experiment was conducted at the ROSA-IV/large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. She experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top