The major motivation of the Work Package 5 (WP5) "Corium and Debris Coolability" is to reduce or possibly solve the remaining uncertainties on the possibility of cooling corium and structural materials during a severe accident, either in the reactor pressure vessel (RPV) or in the reactor cavity, so as to limit the progression of the accident. These issues are covered within severe accident management measures for current reactors, and also within the scope of the design and safety evaluation of future reactors. The current Probabilistic Safety Assessment (PSA) level 2 studies still show very large uncertainties in the results. For the core reflooding phase, in principle two different aspects have to be considered, the probability for reflooding systems to begin operation in due time, and the status of core damage, i.e. whether the fuel rods are still intact or a debris bed has already formed.

 

One of the ultimate objectives is to improve the corresponding physical models in the simulation scodes, in particular on the ASTEC integral code.

 

The following three key situations and processes for the investigation of corium and debris coolability are considered.

 

The focus is on the phase after boil-off during the accident. Heating and melting may produce a severely damaged, partly molten core with relocated material and partly broken parts. Quenching of such a hot and partly degraded core is the main issue here.

 

The experimental database on degraded core reflood was analysed to derive the crucial information about success of reflood. The behaviour of fuel rod bundles can be outlined in a preliminary reflood map with respect to the reflood mass flow rate and the core damage state to deduce the limits up to which final bundle cooling can be expected to be successful and hydrogen production may be tolerated. The present analyses show that even at the onset of severe core degradation, i.e. temperatures up to app. 2200 K, the accident progression can be stopped with a sufficiently high flowrate for core reflood of ~1 g/s*rod.

 

The reflood map on core degradation and hydrogen release is still under development and is considered as a tool to summarize the existing knowledge and to identify blank areas for efficient future experimental work. Several experimental facilities in Europe contribute to solving the still open issues, such as debris particle configurations and core size with 3d effects. A still open issue is the transfer of the rather detailed research findings to existing power plants. As a first step to this goal, ASTEC is under validation using this reflood database.

 

If core cooling fails, a melt pool will form in the core and melt might flow down into remaining water in the lower head. The TMI accident indicated that even though coolability of the core is not attained, a coolable configuration may result from break-up of the melt in the water of the lower head. If cooling in the core and in the lower head is not possible, the development of a melt pool in the lower head has to be analysed and whether a melt pool can be kept in-vessel due to external vessel cooling or the timing and modes of vessel failure have to be considered.

 

This is the general objective of the LIVE programme at KIT. These phenomena resulting from core melting are studied experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. One of the experimental results is that melt pouring near the vessel wall at the beginning of the test results in considerable asymmetric heat flux distribution even during the steady state, i.e. 3 d effects generally have to be taken into account. The experiments are complemented by analyses with Computational Fluid Dynamics (CFD) codes providing valuable support for understanding and improvement of modelling in severe accident codes such as ASTEC.

 

The second important aspect of achieving in-vessel coolability, the external cooling conditions, is investigated by an experimental programme at CEA Cadarache. Two facilities have been built up to reproduce the thermal-hydraulics conditions existing during external cooling of a PWR Reactor Pressure Vessel (RPV): CNU-M1 as a small scale mock-up and CNU-M2, which is a unique experimental set-up, large scale, dedicated to the study of two-phase flow with steam production around a heated RPV geometry. During SARNET2 valuable information on the external cooling will be achieved.

 

If all the attempts to cool the vessel fail, the modes of vessel failure concern location and size of the failure. Up to now the following main conclusions can be drawn for large PWRs: When the vessel fails, the liquid corium is mainly oxidic with potentially some metal. The mass of corium that can be ejected in the reactor pit at vessel failure is estimated between 2 and 20 tonnes. The breach is most probably located on the lateral surface of the vessel. Only local breaches are expected and not vessel unzipping.

 

The related coolability questions of the released melt masses are considered in the ex-vessel case.

 

A porous debris bed can be formed in a water pool of the reactor cavity due to the fragmentation of the molten corium jet ejected from the lower head of the vessel. The water pool is available through cavity flooding (e.g. Severe Accident Management (SAM) in Swedish and Finnish BWRs) or water accumulation in the sump of a PWR due to Loss of Coolant conditions or containment spray. This is a similar process to the in-vessel situation, when melt relocates from the core to a water filled lower head. Deep water pools in BWRs yield additional effects.

 

In order to investigate experimentally the long-term coolability of debris beds, several test facilities are established in Europe. The major objectives of the experimental investigations at this test facility are the determination of local pressure drops for steady state boiling to check friction laws, the determination of dryout heat fluxes under various conditions for validation of numerical models, and the analysis of quenching processes of dry hot debris beds.

 

The newly performed studies focus on investigations with volumetrically heated particle beds in configurations comparing the dryout heat fluxes for top- and bottom-flooding flow situations. It has to be pointed out that with the change from top- to bottom-flooding flow conditions a strong increase of dryout heat fluxes can be observed which improves distinctly the bed’s coolability. These data are used for validation of numerical models for calculations of coolability limits of debris beds and will be implemented in codes such as ASTEC.

 

 

As one example, in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising SAM strategy for VVER-440/V213 reactors. The most important design features of these reactors, favourable for adoption of the IVR concept, are low thermal power, reactor pressure vessel (RPV) without penetration in lower head, massive stainless steel vessel internals, large volume of residual water in lower head and high driving head for natural circulation in ERVC loop.

 

Recent activities devoted to IVR concept via ERVC for standard VVER-440/V213 reactors are performed in the frame of SARNET 2 as well as within national programmes performed in the countries operating this type of reactors. From the results obtained so far it follows, that there should be sufficient coolant flow in natural circulation regime to adequately cool the outer surface of the RPV wall. Further research should be focused on confirmation of the estimated low maximum heat flux values. Here the outcomes within the SARNET 2 project and results of ASTEC V2 analysis will be of high importance.

 

The involved experiments in the three different corium and debris coolability situations are shown in the Table I below. For interpretation of these experiments or understanding of the phenomenology, the following computer codes will be used: ASTEC, MAAP, MELCOR, SCDAP/RELAP, ICARE/CATHARE, ATHLET-CD.

TABLE I

Main experimental programmes on corium and debris coolability

 

Physical process	Programme	Organisation (country)
Reflooding and coolability of a degraded core	QUENCH	KIT (Germany)
	CODEX	AEKI (Hungary)
	POMECO	KTH (Univ. of Stockholm, Sweden)
	DEBRIS	IKE (Univ. of Stuttgart, Germany)
	PEARL	IRSN (France)
Remelting of debris, melt pool formation and coolability	LIVE	KIT (Germany)
	CNU	CEA (France)
Ex-vessel debris formation and coolability:	DEFOR	KTH (Sweden)
	STYX	VTT (Finland)
	+ POMECO, DEBRIS, LIVE