Corium Spreading Over Concrete: The Vulcano VE-U7 and VE-U8 Tests

Two experiments have been performed in the VULCANO facility in which prototypic corium has been spread over concrete. In the VE-U7 test, a mixture representative of what can be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. In the VE-U8 test, a UO2-ZrO2 mixture, prototypic of in-vessel corium, has been spread over a lime-siliceous concrete. Although residual power was not simulated in this experiment, up to 2 cm of concrete have been eroded during the test. Results in terms of spreading behaviour, effects of gases, concrete erosion and thermal attack are presented and discussed.

Experimental Study on Fluid Mixing for Evaluation of Thermal Striping in T-Pipe Junction

A water experiment is performed to investigate thermal striping phenomena in a T-pipe junction which is a typical geometry of fluid mixing. The flow velocity ratio and temperature difference were experimental parameters. The jet form was classified into four patterns; (1) impinging jet, (2) deflecting jet, (3) re-attachment jet and (4) wall jet according to the inflow condition. The parameter experiments showed that the jet form could be predicted by a momentum ratio between the two pipes. The thermochromic liquid crystal sheet showed that a cold spot was formed at the wall surface in the main pipe in the cases of the impinging jet and the wall jet. From the temperature measurement in the fluid, temperature fluctuation intensity was high along the edge of the jet exiting from branch piping. A database of temperature fluctuation and frequency characteristics was established for an evaluation rule of thermal striping in a T-pipe junction.

Dynamic Loads Due to Blowdown Through a Sparger

An experiment has been performed using a facility, which simulates the safety depressurization system (SDS) and in-containment refueling water storage tank (IRWST) of APR1400, an advanced PWR being developed in Korea, to investigate the dynamic load resulting from the blowdown of steam from a steam generator through a sparger. The influence of the key parameters, such as air mass, steam pressure, submergence, valve opening time, and pool temperature, on frequency and peak loads was investigated. The blowdown phenomenon was analyzed to find out the real cause of the initiation of bubble oscillation and discrepancy in frequencies between the experiment and calculation by conventional equation for bubble oscillation. The cause of significant damping was discussed and is presumed to be the highly tortuous flow path around bubble. The Rayleigh-Plesset equation modified by the introduction of method of image reasonably reproduce the bubble oscillation in a confined tank. Right after the completion of air discharge the steam discharge immediately follows and it condenses abruptly to provide low-pressure pocket. It may contribute to the negative maximum being greater than positive maximum. The subsequently discharging steam does not play as the driving force anymore. The effect of various parameters on the peak pressure and frequency was analyzed. The air mass, valve opening time, and submergence give significant effect on both of frequency and peak pressure while steam mass flux gives questionable effect and pool temperature gives almost none.

Prediction of Parameters Distribution of Upward Boiling Two-Phase Flow With Two-Fluid Models

In this paper, a multidimensional two-fluid model with additional turbulence k–ε equations is used to predict the two-phase parameters distribution in freon R12 boiling flow. The 3D module of the CATHARE code is used for numerical calculation. The DEBORA experiment has been chosen to evaluate our models. The radial profiles of the outlet parameters were measured by means of an optical probe. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and volumetric interfacial area at the end of the heated section shows that the multidimensional two-fluid model with proper constitutive relations can yield reasonably predicted results in boiling conditions. Sensitivity tests show that the turbulent dispersion force, which involves the void fraction gradient, plays an important role in determining the void fraction distribution; and the turbulence eddy viscosity is a significant factor to influence the liquid temperature distribution.

Experimental Study on Heat Transfer of Single-Phase Flow and Boiling Two-Phase Flow in Vertical Narrow Annuli

Water single-phase and nucleate boiling heat transfer were experimentally investigated in vertical annuli with narrow gaps. The experimental data about water single-phase flow and boiling two-phase flow heat transfer in narrow annular channel were accumulated by two test sections with the narrow gaps of 1.0mm and 1.5mm. Empirical correlations to predict the heat transfer of the single-phase flow and boiling two-phase flow in the narrow annular channel were obtained, which were arranged in the forms of the Dittus-Boelter for heat transfer coefficients in a single-phase flow and the Jens-Lottes formula for a boiling two-phase flow in normal tubes, respectively. The mechanism of the difference between the normal channel and narrow annular channel were also explored. From experimental results, it was found that the turbulent heat transfer coefficients in narrow gaps are nearly the same to the normal channel in the experimental range, and the transition Reynolds number from a laminar flow to a turbulent flow in narrow annuli was much lower than that in normal channel, whereas the boiling heat transfer in narrow annular gap was greatly enhanced compared with the normal channel.

The Consideration of Entrainment in the Drift-Flux Correlation Package MDS

A comprehensive drift-flux correlation package (MDS) based on the SONNENBURG drift-flux correlation has been established. Its aim is to support thermal-hydraulic mixture-fluid models needed for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters for single- or two-phase fluids flowing up- and downwards along coolant channels of different types (e.g., channel elements of NPP-s, steam generators etc.). Hence, the resulting package MDS had to give special consideration to the two-phase behaviour at co- and counter-current flow situations, its inverse solutions needed for steady state simulations and its behaviour when approaching the lower or upper boundary of a two-phase region. Its characteristic properties, its verification and behaviour with respect to other correlations have (together with an adequate driver code MDSDRI) already be at the ICONE-9 conference at Nice (April 2000). The extension of the code package to situations where droplet entrainment can be expected is subject of this paper. It will be demonstrated that entrainment can, according to the criteria by ISHII-GROLMES (inception) and Ishii-MISHIMA (entrainment fraction), only take place if the given total mass flux exceeds a certain lower (mass flux) limit, a limit being only dependent on system pressure and geometry data. The same is the case for the appearance of counter-current flow. It can only be expected if the given total mass flux is situated within a certain window, a window lying in a low flux range. It is thus not overlapping with the entrainment region, proving that entrainment is not a pre-stadium of counter-current flow. Test calculations will demonstrate the ability of the code package MDS to calculate two-phase flow behaviour along a coolant channel within a wide range of upwards and downwards flow conditions and give an insight over the influence of entrainment at high void fractions and, at lower mass flux regions, of counter-current flow to the overall two-phase behaviour.

Effect of Non-Condensible Gas on the Subcooled Water Critical Flow in a Safety Valve

The effect of non-condensable gas on the subcooled water critical flow in a safety valve is investigated experimentally at various subcoolings with 3 different disk lifts. To evaluate its effect on the critical pressure ratio and critical flow rate, three parameters are considered: the ratios of outlet pressure to inlet pressure, the subcooling to inlet temperature, and the gas volumetric flow to water volumetric flow are 0.15–0.23, 0.07–0.12, and 0–0.8, respectively. It turns out that the critical pressure ratio is mainly dependent on the subcooling, and its dependency on the gas fraction and the pressure drop is relatively small. When the ratio of nitrogen gas volumetric flow to water volumetric flow becomes lower than 20%, the subcooled water critical flow rate is decreased about 10% compare to the water flow rate of without non-condensable gas. However, it maintains a constant value after the ratio of gas volumetric flow to water volumetric flow becomes higher than 20%. The subcooled water critical flow correlation, which considers subcooling, disc lift, backpressure, and non-condensable gas, shows good agreement with the total present experimental data with the root mean square error 8.17%.

Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

Evaluation of Debris Cooling Models in RPV Lower Head Based on Analysis for JAERI-ALPHA Test

For severe accident assessment in a light water reactor (LWR), heat transfer models in a narrow annular gap between the overheated core debris and the reactor pressure vessel (RPV) are important for evaluating RPV integrity and emergency procedures. Some heat transfer models have been proposed as gap cooling CHF (critical heat flux) but local heat fluxes on the hot surface were not taken into account. Therefore, using the existing data, the authors developed heat transfer models on the average CHF restricted by CCFL (counter-current flow limitation) and local boiling heat fluxes, and showed that the average CHF depended on the steam-water flow pattern in the narrow gap and that the local heat fluxes were similar to the pool boiling curve. We evaluated the validity of heat transfer models by simple calculations for an ALPHA/IDC001 experiment performed by JAERI (Japan Atomic Energy Research Institute). Results showed heat fluxes on the crust surface were restricted mainly by its thermal resistance after the crust formation, emissivity on its surface did not have much effect on the heat fluxes, and the calculated vessel temperature during the heat-up process agreed well with the measurements. However, the vessel cooling rate was underestimated mainly due to underestimation of the gap size. The heat fluxes on the vessel inner surface were much higher than the pool film boiling therefore local boiling heat transfer should be studied to improve the heat transfer models.

Description of Typical Flow Instabilities in the Natural Circulation System

The experiments were performed on the test loop HRTL-5, which simulates geometry and system design of the 5-MW Nuclear Heating Reactor developed by the Institute of Nuclear Energy Technology, Tsinghua University. Because of the difference of the geometry design and operating conditions between the heating reactor and the boiling water reactor, the flow behavior presents great differences too, some of which haven’t been deeply studied so far. Results show that in heating reactor, sub-cooled boiling, condensation and flashing play an important role on the flow instabilities of the natural circulation system. Correspondingly, geysering instability, flashing instability, and flow excursion are the very typical instabilities occurring in the primary loop of HRTL-5, which are different from those in boiling water reactor conditions. The compressibility of the steam space on the top of the primary loop has also great influence on the instability of the natural circulation system.