Melt Dispersion and Direct Containment Heating (DCH) Experiments in the DISCO-H Test Facility (original) (raw)
Related papers
1991
Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a highpressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure responses in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions.
In case of a core melt accident in a European light water nuclear reactor the pressure vessel may fail, in spite of depressurization of the primary circuit, still at an elevated pressure of 1 to 2 MPa. Then, the molten core debris will be ejected forcefully into the reactor cavity and beyond, depending on the specific reactor design. This may pressurize the reactor containment building beyond its failure pressure. The pressurization of the containment is due to the debris-to-gas heat transfer but also to a large part to hydrogen combustion. Hydrogen combustion contributes to peak containment pressure if the energy release rate is greater than the heat transfer rate to structures and occurs concurrent with the debris-to-gas heat transfer. This paper presents experimental and analytical results of the combustion of hydrogen jets blown into a scaled reactor containment with a prototypic atmosphere of air, steam and preexisting hydrogen. Experimental data are the pressure and temperatur...
1987
The DCH-1 (Direct Containment Heating) test was the first experiment performed in the Surtsey Direct Heating Test Facility. It produced experimental data required to understand the phenomena associated with pressurized melt ejection and direct containment heating. The results will be used to develop phenomenological models for containment response codes. The test involved 2 0 k g of molten core debris simulant ejected into a 1:lO scale model of the Zion reactor cavity. The melt was produced by a metallothermic reaction of iron oxide and aluminum powders to yield molten iron and alumina. The cavity model was placed so that the emerging debris would propagate directly upwards along the vertical centerline of the chamber. Results from the experiment showed that the molten material was ejected from the cavity as a cloud of particles and aerosolg The dispersed debris caused a rapid pressurization of the 103-m chamber atmosphere. Peak pressure from the six transducers ranged from 0.09 to 0. 1 3 MPa (1 3. 4 to 19.4 psig) above the initial value in the chamber. The time interval from the start of debris ejection to pressure peak was on the order of two to three seconds. Posttest debris collection yielded 1 1. 6 kg of material outside the cavity, of which approximately 1. 6 kg was attributed to the uptake of oxygen by the iron particles. Mechanical sieving of the recovered debris showed a lognormal size distribution with a mass mean size of 0. 5 5 mm. Aerosol measurements indicated a substantial portion (-2 to-16%) of the ejected mass was in the size range less than 10 pm aerodynamic equivalent diameter.
1994
The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of-4 MPa, 80 kg of alumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of-6 cm), and a 1/10_ linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon (< 0.25 mol. % 02) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced.
Direct Containment Heating Investigations for European Pressurized Water Reactors
While the issue on Direct Containment Heating (DCH) was resolved for US reactor plants in the 1990s it was found that the consequences of DCH processes are strongly dependent on the reactor cavity configuration. Therefore, an experimental and analytical program was started in 1998 to inves-tigate melt ejection scenarios for typical German and European reactor designs. Six experiments have been performed in a 1:18 scaled reactor geometry, characterized by a narrow pit without exit other than through the annular space between pressure vessel and cavity wall, leading either directly to the upper containment or into the pump and steam generator rooms along the flow path around the main cooling lines. The corium was modeled by an iron-alumina melt that was driven by steam, and a pro-totypic atmosphere in the containment was applied. Since the system pressure at core melt accidents will be low due to compulsory system depressurization, the vessel failure pressures were kept between 0.8 an...
1994
The ContainmentTechnology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratoriesaxe used to perform scaled experiments for the Nuclear Regulatory Commission that simulate high pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experimentsare designed to investigate the effects of direct containment heating (DCH) phenomenaon the containmentload. High-temperature, chemicallyreactive melt is ejected by high-pressuresteaminto a scale model of a reactorcavity. Debris is entrainedby the steam blowdown intoa containmentmodel where specific phenomena, such as the effect of subcompartmentstructures,prototypic atmospheres, and hydrogen generation and combustion, can be studied. Four Integral Effects Tests (IETs) have been performed with scale models of the Surry NPP to investigate DCH phenomena. These experiments were conducted for five primary purposes: (1) to measure the pressureload on the containmentcontainingprototypic subcompartmentstructures, (2) to investigate the amount of hydrogen combustion due to a HPME into a prototypic steam/air/H2 atmosphere, (3) to investigate the effect of an annular gap between the reactor pressurevessel (RPV) and the reactor support skirt, (4) to measureposttest debris distribution in a containmentmodel, and (5) to provide data from prototypic,large-scale experimentsfor validation of DCH models. The 1/6thsc_e Integral Effects Tests (IET-9, IET-10, and IET-11) were conductedin CT1T, which is a 1/6" scale model of the Surryreactorcontainmentbuilding (RCB). The 1/I(Yh scale IET test (IET-12) was performed in the Surtseyvessel, which had been configured as a 1/10_ scale SurryRCB. Scale models were constructedin each of the facilities of the Surry structures, including the RPV, reactor support skirt, control rod drive missile shield, biological shield wall, cavity, instrumenttunnel, residual heatremoval platform and heatexchangers, seal table room and seal table, operating deck, and crane wall. The RPV model had a hemisphericalbottom head with a hole that simulated the ablatedhole in the RPV that would be formed by ejection of an instrumentguide tube in a severe NPP accident. A charge of thermite was used in the RPV model to simulatemolten corium that would accumulateon the bottom head of an actual RPV. This chemically reactive melt was ejected by high-pressuresteam from the melt generator into the scaled reactor catty. Debris was then entrained through the incore instrumenttunnel into the subcompartment structuresand then into the upper dome of the containment models, where the fragmented molten debris particles encountered prototypic air/steam/hydrogen atmospheres. This reportdescribes these experimentsand gives the results.
1992
The fourth experiment of the Integral Effects Test (IET-4) series was conducted to investigate the effects of high pressure melt ejection on direct containment heating. Scale models (I:I0) of the Zion reactor pressure vessel (RPV), cavity, instrument tunnel, and subcompartment structures were constructed in the Surtsey Test Facility at Sandia National Laboratories. The RPV was modeled with a melt generator that consisted of a steel pressure barrier, a cast MgO crucible, and a thin steel inner liner. The melt generator/crucible had a hemispherical bottom, head containing a graphite limitor plate with a 3.5-cm exit hole to simulate the ablated hole in the RPV bottom head that would Le formed by tube ejection in a severe nuclear power plant accident. The reactor cavity model contained 3.48 kg of water with a depth of 0.9 cm that corresponded to conder_sate levels in the Zion plant. A 43-kg initial charge of iron oxide/aluminum/chromium thermite was used to simulate corium debris on the bottom head of the RPV. Molten thermite was ejected into the scaled reactor cavity by 6.7 MPa steam.
Summary The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to per- form experiments to investigate direct containment heating (DCH) effects during a severe ac- cident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to contain- ment pressurization. The main experimental findings are presented and critical parameters are identified. In the framework of SARNET part of this experim ental data base is subject for a com- mon interpretation work performed by the participating organizations, including application and assessment of the DCH in MAAP, ASTEC/RUPUICUV, COCOSYS and CONTAIN. Even the more mechanistic approaches in the models require the use and adoption of empiri- cal p...
Corium Dispersion and Direct Containment Heating Experiments at Low System Pressure
Experiments in a reduced scale were performed with an iron-alumina melt, steam and a prototypic atmosphere in the containment, to investigate the fluid-dynamic, thermal and chemical processes dur-ing melt ejection out of a breach in the lower head of a PWR pressure vessel at pressures below 2 MPa. A cavity geometry with a direct path into the containment and one with a closed reactor pit, where the only flow path out of the pit is along the main cooling lines leading into reactor rooms, were investigated. Also, an experiment with nitrogen driven melt is compared to one with steam driven melt. With a closed reactor pit, there will be a considerable melt ejection into the pump and steam generator rooms, but almost nothing into the open space of the containment. The pressure increase will stay moderate and well below the design pressure of most containments. The comparison of two tests with and without steam, showed the strong effect of hydrogen production and combustion on both, the m...
Containment behaviour in the event of core melt with gaseous and aerosol releases (CONGA
Nuclear Engineering and Design, 2001
The CONGA project concentrated on theoretical and experimental studies investigating the behaviour of advanced light water reactor containments containing passive containment heat removal systems and catalytic recombiners expected to be effectively operational during a hypothetical severe accident involving large quantities of aerosol particles and noncondensable gases. The central point of interest was the investigation of the effect of aerosol deposition on the condensation heat transfer of specially designed finned-type heat exchangers (HX) as well as the recombination efficiency of catalytic recombiners. A conceptual double-wall Italian PWR design and a SWR1000 design from Siemens were considered specifically as the reference Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) designs. An assessment of selected accident scenarios was performed in order to define the range of boundary conditions necessary to perform the experimental studies of the other work packages. Experimental investigations indicated that aerosol deposition accounted for up to 37% loss in the heat removal capacity of the two-tube-layer BWR HX units. However, no significant heat transfer degradation could be observed for the PWR HX units. These results can be attributed to the important differences in the designs and operating conditions of the two units. The tests to study the effect of hydrogen (simulated by helium) on the heat transfer rate for heat exchanger units designed for BWR and PWR applications indicated a degradation less than 30% under various conditions. This was found to be acceptable within the over capacity designed for the heat exchangers or containment characteristics.