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The installation of passive autocatalytic recombiners (PARs) in the containment of operating nuclear power plants (NPPs) is increasingly based on three-dimensional studies of severe accidents that accurately predict the hydrogen pathways and local accumulation regions in containment and examine the mitigation effects of the PARs on the hydrogen risk. The GOTHIC (Generation Of Thermal-Hydraulic Information for Containments) code is applied in this paper to study the effectiveness of the PARs installed in the Gosgen NPP in Switzerland. A fast release of a mixture of hydrogen and steam from the hot leg during a total station blackout is chosen as the limiting scenario. The PAR modeling approach is qualified simulating two experiments performed in the frame of the OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) THAI (Thermal-hydraulics, Hydrogen, Aerosols and Iodine) project.The results of the plant analyses show that the recombiners cannot prevent the formation...

A detailed experimental database, obtained for a 195-mm inner diameter, 9-m-long pipe was used for the validation of models applied in computational fluid dynamics codes for the simulation ofbubbly flow. Since the bubbles were injected via holes at the pipe wall, very useful information on the bubble migration from the pipe wall toward the pipe's center was obtained by measurements at different distances between gas injection and measuring plane. The bubble migration is determined by the forces acting on the bubbles. The multibubble-size group test solver, introduced earlier but with some new extensions, was used to analyze the data. A comparison of results from a simulation and the experimental findings indicate that the turbulent dispersion force according to the Favre averaged drag model is too strong compared with the drag in the radial direction. No appropriate models for bubble coalescence and breakup, which can be applied for a wide range of gas and liquid volume flow rates, are available as yet. Nevertheless, for selected combinations of volume flow rates, the calculated bubble size distributions and radial gas volume fraction profiles show an acceptable agreement with the experimental data.

An assessment of void-fraction correlations and drift-flux models applied to stationary and transient flashing flows in a vertical pipe has been performed. Experiments have been carried out on a st...

Within the pressurized water reactor (PWR) safety analyses, attention has increasingly focused in recent years on boron dilution events that could potentially lead to reactivity transients. Mixing of the low-boron water with the ambient coolant of higher boron content provides an important mitigation mechanism before the low-boron water enters the core.Experimental support is needed to validate the computational tools to be applied to analyze the mixing of the low-boron water. Experiments were performed in the three test facilities - the Upper Plenum Test Facility (UPTF), the Primaerkreislauf (PKL), and the Rossendorf coolant mixing model (ROCOM) - in Germany.The relevant PKL and UPTF tests were focused on small-break loss-of-coolant accident (SBLOCA) scenarios with reflux-condenser mode and restart of natural circulation. The two test facilities represent a typical western-type PWR and are/were operated by Siemens/KWU now Framatome ANP in Germany. While the restart of natural circulation was investigated in the PKL system test facility (volume 1:145, height 1:1), the UPTF experiments dealt with the mixing of water flows with different boron concentration in the cold legs, reactor pressure vessel (RPV) downcomer, and the lower plenum (all these components were full-scale models).The results from the PKL test facility demonstrate that in case ofmore » a postulated SBLOCA with reflux condensation phase, natural circulation does not start up simultaneously in all loops. This means that slugs of condensate, which might have accumulated in the pump seal during reflux-condenser mode of operation, would reach the RPV at different points in time. The UPTF tests showed an almost ideal mixing of water flows with different boron concentration in the RPV downcomer.The ROCOM test facility has been built in a linear scale of 1:5 for the investigation of coolant mixing phenomena in a wide range of flow conditions in the RPV of the German KONVOI-type PWR. The test results presented are focused on the mixing of a slug of deborated water during the startup of the first reactor coolant pump. Based on experimentally determined pulse responses, a semianalytical model for the description of coolant mixing inside the KONVOI RPV has been developed. Calculations for a presumed boron dilution event during the startup of the first reactor coolant pump have been carried out by means of the semianalytical model and independently by means of the computational fluid dynamics code CFX-4. The semianalytical model is able to describe the time dependent behavior of the deboration front at each fuel element position in a good agreement with the experiment. All main mixing effects, observed in the experiment, are also reproduced by the CFX calculation.« less

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow, natural circulation experiments were performed that were scientifically supported by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility that models VVER-1000 thermal hydraulics. Using the code ATHLET, which is being developed by Gesellschaft fur Anlagen-und Reaktorsicherheit, pre- and posttest calculations were done to determine the thermal-hydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold-leg loop-seal clearing could be confirmed by the tests. The main thermal-hydraulic phenomena were identified and compared with the results gained during similar experiments on VVER-440 test facilities. Besides the thermal-hydraulic standard measuring system, the facility was equipped with needle-shaped conductivity probes for measuring the local void fractions.

The Siemens AG is developing the new innovative boiling water reactor concept SWR1000. New features are the passive safety systems (e.g. emergency condensers, building condensers, passive pressure pulse transmitters, gravity-driven core flooding lines). For the experimental investigation of the emergency condenser effectiveness, the NOKO test facility has been constructed at the Forschungszentrum Jülich in cooperation with Siemens. This test facility has an operating pressure of 10 MPa and a maximum power of 4 MW for steam production. The emergency condenser bundle consists of eight tubes and is fabricated with planned geometry and material of the SWR1000. In more than 200 experiments, the emergency condenser power was determined as a function of pressure, water level and concentration of noncondensables in the pressure vessel as well of pressure, water level and temperature in the condenser. Post test calculations of NOKO experiments were performed with an improved version of ATHLET. To calculate the heat transfer coefficients during condensation in horizontal tubes it was necessary to develop the module KONWAR and to implement it in ATHLET. KONWAR is based on the flow regime map of Tandon and includes several semiempirical correlations for the determination of the heat transfer coefficients. The comparison bet-ween calculations and experiments shows good agreement.

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of the experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.