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Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of

Despite a hypothetical event, whole core melt accident in a sodium cooled fast reactor is investigated as a part of plant safety analysis. To prevent breach of primary boundary in such accidents, a Core Catcher (CC) is provided at the bottom of the main vessel for retention of corium in subcritical and coolable state. For future Indian FBRs, refractory magnesia-lined CC is planned for protection against melt-attack in case of whole core relocation. To evaluate thermal shock resistance of the magnesia liner, an experimental study has been conducted by subjecting magnesia specimens to melt jets of tin and thermite simulating various conditions of corium relocation on the core catcher. Degradation of the magnesia was assessed by non-destructive and destructive tests before and after the experiments. Tin impingement tests at 1100 °C indicated that magnesia specimens could withstand thermal shock with adequate residual compressive strength indicating insignificant degradation. Magnesia specimens subjected to thermite melt impingement at ∼2500 °C showed cracks with spallation on the melt exposed face, surface cracks in the lateral faces, and fracture at few locations up to a depth of 40 mm. However, no cracks were noticed in the deeper regions and the specimens were found to be intact. As the degradation is limited to the upper half of the specimen even in severe thermal shock conditions leaving most of the specimen intact, magnesia can be considered as suitable choice for sacrificial material of the core catcher of SFR. • Thermal shock resistance of refractory magnesia is studied to assessits suitability as sacrificial layer to SFR core catcher. • Magnesia test specimens subjected to thermal shock using molten tin and thermite melt at various conditions of corium relocation. • Degradation of magnesia was assessed by non-destructive and destructive tests before and after the experiments. • Results indicate that magnesia remains structurally intact with reasonable residual compressive strength after thermal shock. • The magnesia class-98 specimens could withstand thermal shock and hence suitable as sacrificial layer for SFR core catcher. [ABSTRACT FROM AUTHOR]

MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic safety assessments, verification and validation of mitigation actions specified in severe accident management guidelines, and source term quantification. In this study, the uncertainty of in-vessel hydrogen generation predicted by the MAAP5 code was quantified. The surrogate plant that was analyzed is the Lungmen Nuclear Power Station of the Taiwan Power Company. The plant employs an advanced boiling water reactor. We performed sensitivity studies to identify the important model parameters that affect the target output parameters. A range and distribution were assigned to these parameters on the basis of experimental results and expert judgment. The number of input parameters in the analysis was 27. Multiple MAAP5 calculations were performed with an input combination generated from Latin hypercube sampling. The calculation results were analyzed parametrically and nonparametrically to determine the 95th percentile with the 95% confidence level value of the amount of in-vessel hydrogen generation. The Pearson correlation coefficient was used to determine the effect of the model parameters on the target output parameters. The analysis results provide guidance for code applications. The only parameters that pass the threshold of 0.362 for hydrogen generation in the core are FCO and TCLMAX. For hydrogen generation in the lower plenum, FOXBJ is the only input parameter that passes the threshold. [ABSTRACT FROM AUTHOR]

Core catcher is a passive safety device in nuclear reactors for collection and long term retention of corium (mixture of molten fuel and structural material) in subcritical and coolable geometry during hypothetical core melt accidents. Core catchers of earlier Sodium cooled Fast Reactors (SFR) were designed for partial core melt accidents. However in view of enhanced safety, the core catcher design for future SFR considers whole core melt scenario, which results in higher thermal and structural loads. Provision of a sacrificial layer is one of the methods to reduce the thermal load on core catcher and enhance its corium retention capability. An experimental study has been taken up to assess the compatibility of few candidate materials for sacrificial layer, with liquid sodium at high temperature conditions. Based on literature, alumina and Yttria Stabilized Zirconia (YSZ) are chosen due to their favorable properties like high melting point and lower thermal conductivity. Test specimen with Stainless Steel (AISI 316 LN, UNS No. S31653) substrate and candidate ceramic layer coated by plasma spray process, were prepared and exposed to sodium at 400 °C for 500 h. Post exposure investigations indicated that YSZ coating was leached out completely, whereas alumina coating was intact with the substrate. At low pressure, alumina had shown weak interaction with sodium i.e., formation of sodium aluminate within a shallow depth. However, at slightly higher pressure, delamination of coating was observed. The results indicated that plasma sprayed alumina and YSZ coatings are not stable in present test configuration. Alternate methods for sacrificial lining as well as materials need to be explored. [ABSTRACT FROM AUTHOR]

Two steam explosion experiments were performed in the TROI (Test for Real cOrium Interaction with water) facility by using partially oxidized molten corium (core material), which is produced during a postulated core melt accident in a nuclear reactor. A triggered steam explosion occurred in one case, but none occurred in the other case. The dynamic pressure and the dynamic load measured in the former experiment show a stronger explosion than those performed previously with oxidic corium. Meanwhile, a steam explosion is prohibited when the melt temperature is low, because the melt is easily solidified to prevent a liquid-liquid interaction. The partially oxidized corium could enhance the strength of a steam explosion due to the thermal energy from an exothermic chemical reaction between the water and the uranium metal with a sufficient superheat extracted during melting. The melt composition effect on a steam explosion load, which was not included during the nuclear design, needs to be included in it. [ABSTRACT FROM AUTHOR]