5 research outputs found
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Analysis of reactor material experiments investigating corium crust stability and heat transfer in jet impingement flow
Presented is an analysis of the results of the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten corium initially at 3080/sup 0/K was directed downward upon a stainless steel plate. The experiments are a continuation of a program of reactor material tests investigating LWR severe accident phenomena. Objective of the present analysis is to determine the existence or nonexistence of a corium crust during impingement from comparison of the measured heatup of the plate (as measured by thermocouples imbedded immediately beneath the steel surface) with model calculations assuming alternately the presence and absence of a stable crust during impingement
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Heatup of the TMI-2 lower head during core relocation
An analysis has been carried out to assess the potential of a melting attack upon the reactor vessel lower head and incore instrument nozzle penetration weldments during the TMI core relocation event at 224 minutes. Calculations were performed to determine the potential for molten corium to undergo breakup into droplets which freeze and form a debris bed versus impinging upon the lower head as one or more coherent streams. The effects of thermal-hydraulic interactions between corium streams and water inside the lower plenum, the effects of the core support assembly structure upon the corium, and the consequences of corium relocation by way of the core former region were examined. 19 refs., 24 figs
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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
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Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors
Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO/sub 2//sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included