Yutaka Kukita

Initial free surface instabilities on a high-speed water jet simulating a liquid-metal target

Experiments are conducted on the initial growth of free surface waves on a high-speed (3.5 to 20 m/s) water jet flow that simulates related aspects of the liquidlithium target in the International Fusion Materials Irradiation Facility. The waves are measured by using laser beam refraction at the water surface. The boundary layer at the nozzle exit and the recovery of the free surface velocity along the jet are also measured. The experimental results confirm that the nozzle-exit boundary layer has a significant influence on the initial growth of waves. With a turbulent boundary layer at the exit, the jet is covered by three-dimensional irregular waves from its beginning. With a laminar boundary layer, however, two-dimensional regular waves grow on an initially smooth water surface. For the latter case, the dominant frequency of the two-dimensional waves agrees well with the linear stability theory of Brennen.

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 results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Abstract Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m2K for the full secondary side mass inventory.

Void-fraction distribution under high-pressure boil-off conditions in rod bundle geometry

Abstract Void fractions in a simulated pressurized water reactor (PWR) core rod bundle geometry were measured under boil-off conditions covering pressures from 3 to 12 MPa and mass fluxes from 5 to 100 kg m −2 s −1 , with a particular interest in void fractions at higher pressures and relatively high mass fluxes. Test results showed that the Chexal-Lellouche model predicts best the present (volume-averaged) void-fraction data among correlations and models examined in this study. The volume-averaged void fractions obtained from differential pressure measurements are systematically smaller than the chordally averaged void fractions obtained from gamma densitometer measurements. Local void fractions were measured in the same bundle for non-heated steam-water two-phase flow of 3 MPa by using an optical void probe. It was found that the difference between the volume-averaged and chordally averaged void fractions mentioned above can be explained qualitatively by a local void-fraction distribution in the bundle measured in the latter tests.

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.

Investigation of centrifugal pump performance under two-phase flow conditions

A one-dimensional two-fluid model has been used to study the centrifugal pump head degradation phenomena and to analyze the gas-liquid interaction within the pump impeller under high pressure, steam-water two-phase flow conditions. The analytical model was used to predict the two-phase pump head data for the small-scale and full-scale nuclear reactor pumps and the predictions of the head degradation compared favorably with the test data for different suction void fractions. The physical mechanisms responsible for head degradation were also investigated.

Summary of ROSA-IV LSTF first-phase test program — Integral simulation of PWR small-break LOCAs and transients -

Abstract Significant experimental results obtained at the ROSA-IV Large-Scale Test Facility (LSTF) during the first phase of the test program (1985–1988) are summarized. The LSTF is a 1 48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type four-loop (3423 MW thermal power) pressurized water reactor (PWR). The LSTF first-phase program investigated the fundamental PWR thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and transients. The test matrix included twenty-nine SBLOCA tests, three abnormal transient tests and ten steady-state natural circulation tests.

Core Makeup Tank Behavior Observed During the Rosa-AP600 Experiments

Integral experiments simulating small-break loss-of-coolant accidents in the Westinghouse AP600 reactor are conducted using the ROSA-V large-scale test facility. These experiments show that the core makeup tank (CMT) behavior can be divided into two phases: the natural-circulation and the drain phases. The natural-circulation phase between the CMT and the rest of the primary is established immediately after the opening of the valve in the discharge line. The hot water from the primary, through the pressure balance line (PBL), accumulates in the top of the CMT forming a clear thermal stratification above the cold initial inventory of the CMT. The drain phase is initiated by flashing in the CMT for break diameters ≤1 in. and by a gaseous flow from the primary for break diameters ≥2 in. Interactions between the CMT and the other safety components are observed: The CMT discharge rate is decreased by accumulator injection and is increased by actuation of the automatic depressurization system. When the PBL is empty of liquid, the CMT drain rate is approximately given by the free gravitational drain rate, irrespective of the flow direction in the PBL.

Experimental study of two-phase pump performance using a full size nuclear reactor pump

Abstract The performance of a full-size nuclear reactor primary heat transport pump was investigated experimentally under high pressure, steam–water two-phase flow conditions. A new set of two-phase pump performance test data was obtained with local void fraction and mass flux measurements at the pump suction. The effects of suction temperature and initial flow conditions on the two-phase pump performance characteristics were described.

Implications of the Rosa/AP600 High- and Intermediate-Pressure Test Results

Westinghouse has submitted the new AP600 reactor design for certification under the U.S. Nuclear Regulatory Commission (NRC) regulations for standard designs. The NRC has performed confirmatory testing in the Rig of Safety Assessment (ROSA) - V Large-Scale Test Facility run by the Japan Atomic Energy Research Institute (JAERI). The ROSA/AP600 test results as provided by JAERI are reviewed in terms of their implications for the original technical concerns that were to be addressed by the testing program. Implications for computer code capabilities are also discussed. Since gravity-driven natural circulation flow in a complicated piping network was a key concern, the review concentrates on this aspect of the ROSA/AP600 test results. In particular, it looks at the possible effect of system interactions at high pressure and during early depressurization. It identifies those ROSA/AP600 test occurrences that point to processes that could delay automatic depressurization system initiation or in-containment reactor water storage tank injection. Since most of the tests run were small-break loss-of-coolant accidents (SBLOCAs), the review focusses on this type ofscenario. A comparison of several SBLOCA tests is presented.