Direct containment heating experiments in Zion Nuclear Power Plant Geometry using prototypic core materials, the U2 test

A third Direct Containment Heating (DCH) experiments has been completed which utilizes prototypic core materials. The reactor material tests are a follow on to the Integral Effects Testing (IET) DCH program. The IET series of tests primarily addressed the effect of scale on DCH phenomena. This was accomplished by completing a series of counterpart tests in 1/40 and 1/10th linear scale DCH facilities at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL), respectively. The IET experiments modeled the Zion Nuclear Power Plant Geometry. The scale models included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven at nominally 6.2 MPa. Iron-alumina thermite with chromium was used as a core melt simulant in the IET experiments. While the IET experiments at ANL and SNL provided useful data on the effect of scale on DCH phenomena, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. A prototypic core melt was produced for the experiment by first mixing powders of uranium, zirconium, iron oxide (Fe{sub 2}O{sub 3}), and chromium trioxide (CrO{sub 3}). When ignited the powders react exothermically to produce a molten mixture. The amounts of each powder were selected to produce the anticipated composition for a core melt following a station blackout: 57.8 mass% UO{sub 2} 10.5 mass% ZrO{sub 2} 14.3 mass% Fe, 13.7 mass% Zr, and 3.7 mass% Cr. Development tests measured the initial melt temperature to be in the range of 2600 - 2700 K. The total thermal specific energy content of the melt at 2700 K is 1.2 MJ/kg compared to 2.25 MJ/kg for the iron-alumina simulant at its measured initial temperature of 2500 K

Vapor-explosion experiments with subcooled Freon. [LMFBR]

Vapor-explosion experiments were conducted in a well-wetted Freon-22 and mineral-oil system in which the initial temperature of both the Freon and the mineral oil were varied over a wide range. These experiments were specifically conducted to investigate the importance of interface temperature in determining the explosive behavior of a given system. The results clearly demonstrate that the interface temperature developed upon intimate liquid-liquid contact is a valid characterization of the explosive potential of a given system

Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

Status report on severe accident material property measurements

Measurements of selected material properties of molten reactor core material (corium) were made. The corium used was a mixture of UO{sub 2}, ZrO{sub 2} and Zr, with oxygen content being a parameter to reflect different stages of zirconium oxidation. The mixtures used were representative of typical in-vessel melt sequences. For most measurements, the UO{sub 2}/ZrO{sub 2} mass ratio was 1.51, representative of VVER/440 melt compositions and melt compositions of most US BWRs. Measurements were made of the solidus/liquidus temperatures of corium compositions using a Differential Thermal Analysis technique. Observation of the solubility of unoxidized Zr in the oxide phase was made by metallographic analysis of solidus/liquidus melt samples. The results of laminar flow corium spreading tests in one dimension were used to estimate the viscosity of corium compositions. Measured solidus and liquidus temperatures for compositions representative of Zr oxidation of 30, 50 and 70% were compared with those obtained form a phase diagram provided by Kurchatov Institute. It was found that experimental measurements agreed well with the phase diagram values at 70% oxidation, but the measured solidus temperatures were higher than those on the phase diagram and the measured liquidus temperatures were lower than those on the phase diagram at 30 and 50% oxidation. From a microstructure examination it was determined that there was no global segregation into distinct metal and oxide phases during the cooldown of a sample in which there was initially 70% Zr oxidation. Therefore it is concluded that Zr metal is soluble in the oxide phase under molten conditions. Viscosity estimates were made for compositions representative of Zr oxidation of 30, 50 and 70% by fitting the results of spreading tests to Huppert`s equation. It was found that, at a temperature of 2500 C, the viscosity varied by three orders of magnitude over this range of compositions. 10 refs., 39 figs., 16 tabs

Direct containing heating experiments in Zion Nuclear Power Plant Geometry using prototypic core materials, the U1A and U1B tests

Direct Containment Heating (DCH) experiments have been performed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The IET series of tests primarily addressed the effect of scale on DCH phenomena. This was accomplished by completing a series of counterpart tests in 1/40 and 1/10th linear scale DCH facilities at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL), respectively. The IET experiments modeled the Zion Nuclear Power Plant Geometry. The scale models included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven at nominally 6.2 MPa. Iron-alumina thermite with chromium was used as a core melt simulant in the IET experiments. While the IET experiments at ANL and SNL provided useful data on the effect of scale on DCH phenomena, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. A prototypic core melt was produced for the experiments by first mixing powders of uranium, zirconium, iron oxide (Fe{sub 2}O{sub 3}), and chromium trioxide (CrO{sub 3}). When ignited the powders react exothermically to produce a molten mixture. The amounts of each powder were selected to produce the anticipated composition for a core melt following a station blackout: 57.8 mass% UO{sub 2} 10.5 mass% ZrO{sub 2} 14.3 mass% Fe, 13.7 mass% Zr, and 3.7 mass% Cr. Development tests measured the initial melt temperature to be approximately 2700 K. The total thermal specific energy content of the melt at 2700 K is 1.2 MJ/kg compared to 2.25 MJ/kg for the iron-alumina simulant at its measured initial temperature of 2500 K

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

Corium quench in deep pool mixing experiments

The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO/sub 2/, 16% ZrO/sub 2/, and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m/sup 2/ and 3.7 MW/m/sup 2/, respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

Fragmentation and quench behavior of corium melt streams in water

The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (i) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (ii) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (iii) the quench rate of the molten fuel through the water in the lower plenum, (iv) the steam generation and hydrogen generation during the interaction, (v) the transient pressurization of the primary system, and (vi) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through {minus}6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics

Experimental study of the fragmentation and quench behavior of corium melts in water

The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (1) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (2) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (3) the quench rate of the molten fuel through the water in the lower plasma, (4) the steam generation and hydrogen generation during the interaction, (5) the transient pressurization of the primary system, and (6) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics. 9 refs., 29 figs