In-vessel Melt Retention Research Articles

Experimental investigation of the permeability of PWR insulation panels for ex-vessel cooling

This work is linked to In-Vessel Melt Retention during severe accidents on existing LWRs. Because of the thermal insulation which has been installed around the reactor vessel, the flow path for the external cooling of the vessel in existing French reactors is not optimal and vapour may accumulate between vessel and insulation, in the present case consisting of screwed panels.The paper describes recent CALO experiments to investigate panel-junction permeability to vapour flow, taking into account:-the scale effect,-the effect of panel-junction clearance,-the effect of the pressure difference between the inner volume (i.e. between vessel and insulation) and the outer volume (between insulation and concrete pit wall),-mechanical deformation of the insulation panels and the related effect on two-phase flow through the junctions between the panels.The CALO experiments demonstrate that there is no significant risk of vapour accumulation with thermal insulation made of screwed panels. Panel-junction clearance of ∼0.5mm is sufficient to release the vapour produced around the lower head. If the clearance is smaller, the panel connection would be deformed due to the pressure difference between the inner and outer volumes. The connection would deform sufficiently, or even screws would break, before vapour accumulation between the vessel and the insulation reached the level of the lower head.

ASTEC application for in-vessel melt retention modelling in VVER plants

The paper is devoted to extended ASTEC analyses of possible in-vessel retention (IVR) phenomena, based on simulation of main accident scenarios for VVER-1000 reactor design with the ASTEC 2.0 rev2 code. To reach the objective of the investigation, an assessment of two hypothetical low-pressure severe accidents scenarios has been carried out, which provided a basis for well-founded insight on key phenomena during IVR in a VVER-1000 plant. This examination allowed to obtain main characteristics of the corium that is relocated to lower plenum, along with necessary understanding of ex-vessel phenomena.Separate simulations of corium behaviour within the vessel lower head in conditions of external cooling show that the efficiency of IVR critically depends on residual nuclear power retained in the corium. Basing on calculation results it may be concluded that the IVR objective can be achieved for VVER-1000 at the level of residual power which corresponds to approximately 24h of accident transient after the initial event. Potentially such conditions can be realized on new Generation III plants or in the course of modernization of the existing ones by increasing water supply from passive safety injection tanks.A code-to-code comparison between ASTEC and the SOCRAT Russian code confirms main results obtained. At the same time it revealed certain deficiencies in existing modelling approaches, which are the result of poorly examined thermodynamical properties of the corium at temperatures higher than 2200K.

Distribution of fission product residual decay heat in stratified core melt of LWR and its influence on sidewall heat flux

The model of spatial distribution of fission product (FP) activity and decay heat in core melt of light water reactor under severe accident conditions is proposed. It is based on thermodynamical consideration of “U–Zr–O–Fe–FP” system as a two-phase (oxide and metallic) stratified melt. The thermochemistry of FP phase distribution, heat-mass transfer and radioactive transformations of ∼100 isotopes of Cs, Mo, Ru, Ba, Sr, Zr, La, Ce, Eu, Nd, Nb and Sb are considered self-consistently. The calculations carried out with specially designed code FPMC (fission products–molten core) showed that up to 15–25% of decay heat may be released in the metallic phase of the melt. Estimation of this value is fundamentally important in simulations of heat and mass transfer in stratified melt, such as in-vessel melt retention and reactor lower head failure. For numerical assessment of the influence of decay heat spatial distribution in stratified molten pool on the lower head failure time the multivariate simulation of the later stage of large break loss-of-coolant accident in VVER-1000 reactor was carried out by SOCRAT and HEFEST codes coupled with FPMC.

The LIVE program – Results of test L1 and joint analyses on transient molten pool thermal hydraulics

The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e.g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO 3-NaNO 3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE activities also provide data for a better understanding of in-core corium pool behaviour. The experimental results are being used for the development and validation of mechanistic models for the description of molten pool behaviour. In the present paper, a range of different models is used for post-test calculations and comparative analyses. This includes simplified, but fast running models implemented in the severe accident codes ASTEC and ATHLET-CD. Further, a computational tool developed at KTH (PECM model implemented in Fluent) is applied. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE (Nuclear Safety Institute of Russian Academy) within the RASPLAV project and was further improved within the ISTC 2936 Project.

Corium phase equilibria based on MASCA, METCOR and CORPHAD results

Experimental data on component partitioning between suboxidized corium melt and steel in the in-vessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCA program and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCOR project under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,…)–O system is reached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has two immiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibility gap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,…)–O phase diagram investigated within the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR tests the equilibrium is established between oxidic liquid and mushy metallic part of the system. Results of comparison are discussed and the implications for IVR noted.

Fracture mechanical evaluation of an in-vessel melt retention scenario

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.

Assessment of reactor vessel integrity (ARVI)

The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

Challenges left in the area of in-vessel melt retention

The in-vessel melt retention becomes an important safety objective for the present or future middle power nuclear plants, so care has to be taken in the evaluation of the various phenomena related to ensuring the feasibility of this objective. Since the prediction of the relevant phenomena has to be performed for the prototypical accident conditions, the applicability of the measured data or of the correlations derived from these measurements have to be established and the uncertainties determined. In this context, most uncertainties are introduced by the non-prototypicalities in the experiments. The paper describes the major findings from the OECD RASPLAV project and discusses the remaining challenges left in the area of in-vessel molten corium coolability.

On heat transfer characteristics of real and simulant melt pool experiments

This paper presents results of analytical studies on natural convection heat transfer in scaled and/or simulant melt pool experiments related to the pressurized water reactor in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing the relevant physical processes in prototypical situations is examined. Experiments and experimental approaches are analyzed by focusing on their ability to represent prototypical situations. Calculations are performed to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. In the present analysis, Rayleigh numbers are limited to 10 12, where uncertainties in turbulence modelling do not override other uncertainties. Calculations are performed to explore limitations of using side wall heating and direct electrical heating. The need for further experimental and analytical efforts is also discussed.

Critical heat flux through curved, downward facing, thick walls

Experimental data are presented that provide a lower envelope on the critical heat flux distribution over the external surface of a hemisphere submerged in water. The experiment was designed and run so as to provide an essentially full-scale simulation of a reactor pressure vessel lower head. Thus the data, and the correlation derived from them, can be applied directly, and they are supportive of the important, severe accident management idea of retaining the core debris in the reactor vessel by external flooding (“in-vessel melt retention”).