IKE. Modelling of the Late Phase of Core Degradation in Light Water Reactors. Michael Buck. Institut für Kernenergetik und Energiesysteme

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1 IKE Institut für Kernenergetik und Energiesysteme Modelling of the Late Phase of Core Degradation in Light Water Reactors Michael Buck Universität Stuttgart November 7 IKE - 53

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3 IKE Institut für Kernenergetik und Energiesysteme Modelling of the Late Phase of Core Degradation in Light Water Reactors Von der Fakultät Maschinenbau der Universität Stuttgart zur Erlangung der Würde eines Doktor-Ingenieurs (Dr.-Ing.) genehmigte Abhandlung Vorgelegt von Michael Buck geboren in Schwenningen a.n. Hauptberichter: Mitberichter: Prof. G. Lohnert, Ph.D. Prof. Dr.-Ing. E. Hahne Tag der mündlichen Prüfung: 7. November 6 ISSN Universität Stuttgart November 7 IKE - 53

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5 CHAPTER I ABSTRACT iii I Abstract The objective of the present work is to develop a model to describe the late phase of melting and degradation processes in the core of a Light Water Reactor which occur during a severe accident. This model was integrated in the German system code ATHLET-CD. The purpose is to apply the code to severe accident scenarios in order to obtain information on the timing of processes, on possible variants of the evolution and on options of accident management measures during the available time margins. This can also give an improved basis for the development of new LWR concepts with improved safety features, for which the probability and consequences of such accidents are strongly reduced. The model and computational module MESOCO developed in the frame of the present work describes the accident progression under conditions of an already significantly degraded core where melting and relocation of ceramic components dominate. It has been based on a quasicontinuum description for the complex, changing geometrical structure and the melt and steam or gas flows in this structure. Melt accumulation in the core, and, finally modes of melt release to the lower head are considered as the most important effects determining the subsequent processes regarding coolability and retention issues in the lower head and possible failure modes of the RPV as well as the timing of such developments. Emphasis therefore has been laid on the description of melting and solidification behaviour with interactions between different materials and melt pool formation in the core as well as subsequent behaviour of the pool/crust configuration. The time development of these processes, in connection with cooling boundary conditions, given essentially by the water level development, decides on sideways versus downwards progression, on formation of large pools, on temporarily stable crusts or rather on continuous downwards melt flow as well as on subsequent failure and outflow modes. Validation analyses based on PHEBUS experiments FPT4 and FPT are presented, applying separate calculations with MESOCO as well as with the model implemented in ATHLET-CD. Calculations have been performed for reactor scenarios by checking variants of the modelling. From the experiences with these calculations it is concluded that strong trends exist towards pool formation with subsequent lateral melt outflow after continued heatup and subsequent failure. This results in gradual outflow modes with reduced flow rate and favours particulate debris formation in the lower head as potentially coolable configuration and yields also strong arguments against the risk of efficient steam explosions with resulting critical loads. Further analyses for a larger spectrum of conditions and accident scenarios are, however, required to confirm these results and conclusions. The present model and its integration in the system code ATHLET-CD yield a basis for this.

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7 CHAPTER II KURZFASSUNG v II Kurzfassung Das Ziel der vorliegenden Arbeit ist die Entwicklung eines Modells zur Beschreibung der späten Phase des Kernschmelzens und der Kernzerstörung bei schweren Störfällen in Leichtwasserreaktoren. Dieses Modell wurde in den deutschen Systemcode ATHLET-CD integriert. Dieser soll auf Szenarien schwerer Störfälle angewendet werden, um Informationen über die zeitliche Entwicklung, mögliche Varianten im Unfallablauf sowie Eingriffsmöglichkeiten durch Accident-Management-Maßnahmen innerhalb des zur Verfügung stehenden Zeitrahmens zu erhalten. Hierbei wird auch zur Entwicklung neuer Konzepte für Leichtwasserreaktoren beigetragen, bei denen die Eintrittswahrscheinlichkeit und der Folgen solcher schwerer Störfälle stark vermindert sind. Das Modell bzw. Rechenprogramm MESOCO, das im Rahmen der vorliegenden Arbeit entwickelt wurde, beschreibt die Störfallentwicklung unter Bedingungen eines bereits weitgehend zerstörten Reaktorkerns, die vorwiegend durch das Schmelzen und die Verlagerung von keramischen Materialkomponenten geprägt sind. Das Modell basiert auf einer Quasi- Kontinuumsbeschreibung für die komplexe, veränderliche geometrische Struktur und die Strömung von Schmelze und Dampf bzw. Gas durch diese Struktur. Ansammlung von Schmelze im Kern und schließlich die Art des Ausfließens von Schmelze aus dem Kern in die untere Kalotte des Reaktordruckbehälters werden als entscheidende Effekte angesehen, die die nachfolgenden Prozesse hinsichtlich der Kühlbarkeit und der Rückhaltung von Kernschmelze in der unteren Kalotte bzw. die Art und den Zeitpunkt des Versagens des Reaktordruckbehälters bestimmen. Der Schwerpunkt wurde deshalb auf die Beschreibung des Schmelzens und Erstarrens unter dem Einfluss von Material-Wechselwirkung und der Bildung von Schmelzeseen im Kern sowie des nachfolgenden Verhaltens von Schmelzesee und umgebender Kruste gelegt. Die zeitliche Entwicklung dieser Prozesse entscheidet darüber, zusammen mit den Kühlungsbedingungen, die vorwiegend durch die Entwicklung des Wasserstands im Kern bestimmt werden, ob das Schmelzen vorwiegend seitlich oder nach unten fortschreitet, ob sich große Schmelzeseen innerhalb zeitweise stabiler Krusten oder ein eher kontinuierlicher Schmelzefluss nach unten ergeben, sowie letztendlich über die Art des Ausfließen von Schmelze aus dem Kern. Es werden Validierungsrechnungen anhand der PHEBUS Versuche FPT4 und FPT vorgestellt, die sowohl mit MESOCO als auch mit dem in ATHLET-CD integrierten Modell durchgeführt wurden. Anhand von Rechnungen für Reaktorbedingungen wurden verschiedene Varianten der Modellierung überprüft. Aus den Erfahrungen dieser Rechnungen wird geschlossen, dass generell eine starke Tendenz zur Ausbildung von Schmelzeseen besteht, wobei die weitere Aufheizung zum seitlichen Versagen von Krusten und nachfolgendem Ausfließen von Schmelze aus dem Kern führt. Daraus ergibt sich ein eher allmähliches Ausfließen mit begrenzten Massenströmen, wodurch die Bildung von Partikelschüttungen mit hohem Kühlungspotential in der unteren Kalotte begünstigt wird. Ein allmähliches Ausfließen wirkt auch dem Risiko kritischer Druckbelastungen durch starke Dampfexplosionen entgegen. Weitere Untersuchungen für ein breiteres Spektrum von Bedingungen und Unfallszenarien sind jedoch notwendig, um diese Ergebnisse und Schlussfolgerungen zu bestätigen. Hierzu liefert das vorliegende Modell und seine Integration in ATHLET-CD die Grundlage.

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9 CHAPTER III TABLE OF CONTENTS vii III I II Table of Contents Abstract...iii Kurzfassung...v III Table of Contents...vii IV Nomenclature...ix Introduction.... Nuclear reactors and safety research.... Objectives of the present work Approach and outline of the present work...4 Severe accidents with core melting State of the art and experimental database...5. Physical phenomena and processes in severe core melt accidents Severe accident sequences leading to core melting Early phases of core melting Late phases of core melting Processes in the lower head of the RPV.... Experimental database Separate effect tests on debris bed behaviour and molten pool formation Separate effect tests on molten pool behaviour Integral bundle tests TMI- reactor accident Summary of available experimental information and considerations for model verification Modelling approaches for the late phase in severe accident analysis codes Melting and refreezing processes Melt flow in a complex structure of a degrading core Pool formation and continued development of a pool/crust configuration Physical and mathematical model of late phase core degradation Quasi-continuum approach for the description of the degrading core Conservation equations Mass conservation equations for gas, solid and melt Momentum conservation equations for gas and melt Energy conservation equations for gas, solid and melt Constitutive laws Thermal equation of state for solid, melt and gas Interfacial drag, capillary effects Heat transfer between gas and solid or melt...49

10 viii CHAPTER III TABLE OF CONTENTS Effective thermal conductivity, radiation Material interactions, melting and solidification Oxidation of Zircaloy Modelling of molten pool and crust behaviour Numerical solution methods Implementation of the models for late phase core degradation in the code ATHLET- CD The severe accident simulation code ATHLET-CD Coupling concept Realisation of the implementation Model regions and spatial discretisation Initial conditions Boundary conditions Interface with the numerical solution method of ATHLET-CD Model verification PHEBUS experiment FPT Pre-test calculations Post-test calculations PHEBUS experiment FPT Application to reactor conditions Influence of modelling variants on the integral development Effect of descriptions for melting, solidification and material interactions Effect of descriptions for molten pool heat transfer Calculation with ATHLET-CD and integrated MESOCO model... 7 Summary and Conclusions References...35 Appendix A Numerical solution methods in MESOCO...4 A. Spatial discretisation by a Finite Volume method...4 A.. Discretisation of mass and energy conservation equations...4 A.. Discretisation of the momentum equations...44 A. Time integration method...46 A.. Considerations on the choice of the time discretisation method...46 A.. BDF time integration method...48 A..3 Sparse matrix methods for the treatment of the iteration matrix and for solution of linear systems of equations...5

11 CHAPTER IV NOMENCLATURE ix IV Nomenclature Latin symbols a m /s thermal diffusivity, a = λ /( ρ cp) A m area A m /m 3 interfacial area per volume C - constant c, cp J/(kg K) specific heat capacity d m diameter dh m hydraulic diameter, dh = 4 [free flow volume] / [wetted surface] E J/kmol activation energy g m/s gravitational acceleration h J/kg specific enthalpy H m height J - Leverett function K (kg/m ) /s oxidation rate constant l m (characteristic) length L m crust thickness m kg mass m& kg/s mass flow rate M kg/kmol molar mass n /m 3 particle number density N - total number of components p N/m pressure Q W/m 3 volumetric power q W/m heat flux density r m radial coordinate R m radius Rm J/(kmol K) gas constant, Rm = J/(kmol K) s m 3 /m 3 saturation (fraction of liquid in the pore volume) t s time

12 x CHAPTER IV NOMENCLATURE T K temperature v m/s velocity V m 3 volume x m x-coordinate or arc length y m y-coordinate z m z or vertical coordinate Greek symbols α W/(m K) heat transfer coefficient β m 3 /m 3 volume fraction δ m (boundary) layer thickness ε m 3 /m 3 porosity γ /K thermal expansion coefficient Γ kg/(m 3 s) volumetric mass transfer rate η kg/(m s) dynamic viscosity κ m permeability λ W/(m K) thermal conductivity µ m passability ν m /s kinematic viscosity θ rad solid/liquid wetting angle ρ kg/m 3 density σ N/m surface tension ξ kg/kg mass fraction Indices bl c dn eff eut g i int j boundary layer capillary downwards effective eutectic gas interface internal phase

13 CHAPTER IV NOMENCLATURE xi k m max min O p r rel s sd up W component melt maximum value minimum value oxidation particle residual relative solid sideways upward wall Dimensionless numbers Da Gr Nu Pr Ra Ra int Re Damköhler number, Grashof number, Nusselt number, Prandtl number, Rayleigh number, Gr Nu Pr Da = = = ν a internal Rayleigh number, Reynolds number, Re = Ql λ T gγ Tl ν αl λ 3 gγ Tl Ra = Gr Pr = νa Ra = vd ν int 3 = Gr Da = gγql λνa 5

14 xii CHAPTER IV NOMENCLATURE Abbreviations AMM ATHLET BWR CEA CSA DCH EPR FP FZK HPI HTR IRSN LBLOCA LWR MCP MESOCO MP OTSG PORV PWR RCS RPV SA SB TMI- VVER Accident management measure Analysis of Thermal-Hydraulics of Leaks and Transients Boiling Water Reactor Commissariat à l Energie Atomique Core Support Assembly Direct Containment Heating European Pressurised Water Reactor Fission Product Forschungszentrum Karlsruhe High Pressure Injection High Temperature Reactor Institut de Radioprotection et Sûreté Nucléaire Large Break Loss-of-Coolant Accident Light Water Reactor Main Coolant Pump Melting/Solidification Code Melt Progression Once-Through Steam Generator Pilot Operated Relief Valve Pressurised Water Reactor Reactor Cooling System Reactor Pressure Vessel Severe Accident Station Blackout Three Miles Island Unit reactor in Harrisburg, PA, USA Russian type of PWR (water cooled water moderated energy reactor)

15 CHAPTER INTRODUCTION Introduction. Nuclear reactors and safety research Since 954, nuclear fission is used for peaceful purposes and has since then become a resource of energy of continuously increasing importance. Especially, nuclear reactors contribute significantly to electrical energy supply without CO emission world wide. Controversial discussions on the use of nuclear energy are to a large amount based on the potential risk of nuclear power plants for the public and the environment. During nuclear fission in reactors a multitude of highly radioactive nuclides are produced, so called fission products. Even after shut-down of the reactor, heat is further produced due to the radioactive decay of the fission products (decay heat). Without cooling, the decay heat suffices to heat up and finally melt the reactor core within some hours. The largest part (about 93%) of the fission products is bound in the crystal lattice of the fuel. Core melting will lead to the release of fission products from the fuel. Subsequent failure of several safety barriers (fuel cladding, primary circuit, reactor containment) finally results in the release of large amounts of fission products from the reactor building with severe consequences for the environment. Therefore, reactor safety is of crucial importance for the construction and operation of nuclear power plants. High quality standards ensure proper operation and a variety of systems dedicated to safety are designed to control a large spectrum of potential malfunctions, so called design based accidents. However, in spite of high safety standards, failures may occur which exceed the range of design. These can be caused by simultaneous failure of several system components (e.g. double end break of a main coolant pipe and failure of low pressure injection or Station Blackout), human errors or an event for which the plant was not designed. The probability of such initiating events is very low, however they can lead to so called severe accidents with melting of the reactor core and extensive damages to the plant. A well known example for a severe accident is the incidence from March 8, 979 in the Tree Mile Island (TMI-) plant in the United States [], where the coincidence of unfavourable factors and human misjudgement led to melting of parts of the core and to a large release of radioactivity into the containment. In order to avoid the occurrence of such events or at least to reduce the consequences of severe accidents for the environment, accident management measures (AMM) are being provided and constantly improved. The planning and application of appropriate measures requires well established knowledge on the progression of severe accidents. The recognition of the large potential hazard of severe accidents, in spite of their low probability of occurrence, and its impact on the acceptance of nuclear power as a safe source of energy by the public has in many countries resulted in enhanced requirements for the licensing of future nuclear power plants and upgrading of operating plants. The final aim is to include even severe accidents with core melting into the range of design base accidents by ensuring that the consequences are restricted to the interior of the plant. This has motivated the development of so called evolutionary concepts for new light water reactor types, such as the European Pressurised water Reactor (EPR) [], a common development of Framatome and Siemens, the SWR [3] designed by Siemens or the AP6 [4] of Westinghouse. Furthermore, previous and ongoing development focuses on the design of so called innovative reactors concepts, which aim at complete exclusion of severe accidents by inherent safety features, such as the HTR Module [5] developed by Siemens.

16 CHAPTER INTRODUCTION For Light Water Reactors, a possible concept for retention of molten core material in case of a severe accident is the external cooling of the reactor pressure vessel (RPV) by flooding of the reactor cavity with water. This is e.g. applied in the Finnish Loovisa plant [6] and is foreseen for the SWR [3] reactor type. The external cooling concept, however, might only feasible for low power density reactors. The EPR concept therefore relies on a device outside the reactor vessel (ex-vessel), the so called core catcher. This is designed to collect the core melt (corium) released from the RPV after thermal failure into the reactor pit and to spread it subsequently on a large surface in a dedicated compartment, in order to stabilise and cool the corium []. An alternative to the EPR core catcher design is the COMET concept [7] developed at FZK, which aims at fragmenting and cooling an ex-vessel corium layer by water injection from below. In a further concept, applied as mitigation ex-vessel measure in Swedish and Finnish BWR plants, a several meters deep water pool below the RPV shall be established by flooding the below drywell compartment. In the case of RPV failure, the released melt shall be fragmented in the deep water pool and lead to formation of coolable particulate debris. A common feature of the above concepts (at least the former ones) is that they concentrate on bounding configurations (e.g. large molten pool in the lower head of the RPV) in order to merge a wide variety of potential scenarios and to become largely independent of the accident sequence. The consequence of this, at first an appealing approach, is that the bounding configurations have to be ultimately mastered. However, problems with such absolute concepts are the extreme difficulties to finally make them waterproof. This is e.g. highlighted by the problem of metal layers leading to high local heat fluxes in the external vessel cooling concept or the questionability of proper spreading in the EPR concept if melt is released from the pressure vessel in several batches distributed over a longer time period. These problems bring into play again the necessity to consider the whole accident sequences. The concentration on bounding cases also neglects the potential of reaching coolable configurations and stopping accident progression at intermediate stages, before reaching the ultimate states. In contrast to this, almost any accident management guidelines at present require to inject water into the primary cooling system (PCS) whenever possible and at any stage of the accident. In order to evaluate the chances (and also the potentially adverse effects such as hydrogen production and steam explosion risks) of such interventions, it is essential to have sufficient knowledge on the evolution and on the physical processes of core melting and degradation even in the late phase of an accident. This has promoted large experimental programmes as LOFT [8], ACRR [9], CORA [] and PHEBUS [], [3], and numerous small scale and separate effect tests. The results of these investigations have entered into the development of analytical simulation tools such as the MELCOR [5], MAAP [6], SCDAP/RELAP [7], ICARE/CATHARE [8], ATHLET-CD [9] and KESS [] codes. The KESS (Kernschmelz-Simulations-System) code is being developed at the Institut für Kernenergetik und Energiesysteme (IKE) and concentrates on the description of the processes inside the RPV. Models from KESS are implemented in the system code ATHLET-CD, developed by the Gesellschaft für Anlagen und Reaktorsicherheit (GRS), which allows also the simulation of the primary and secondary reactor cooling system. These codes aim at a largely mechanistic, physically based description of the development under severe accidents until failure of the RPV. For each phase of the accident, the influences of AM measures shall be ascertainable additionally, i.e. the effects on the coolability and retention of the melt in the RPV.

17 CHAPTER INTRODUCTION 3. Objectives of the present work As basis for the present work, simulation models for the earlier phases of severe accidents are already available in KESS and ATHLET-CD. These models treat processes within a still largely intact geometry of the core. In contrast to this, the late phases of core melting are characterised by proceeding core damage, loss of geometry, massive melt formation, relocation and accumulation and finally release of melt to the lower head of the RPV. In order to describe also these processes, a different modelling approach is required. Due to the complexity of the involved phenomena, it is essential not to describe every effect in detail, but rather to work out and concentrate on the decisive processes, i. e. the modelling has to be oriented only at the problems to be answered with respect to the safety relevant issues of the late phase. With respect to the earlier phases of boil-off, heat-up and melting of the reactor core, detailed knowledge of conditions and processes of the initiation of core melting as well as their consequences are of particular importance (onset of rapid cladding oxidation, early melt formation due to material interactions, first formation of blockages and their influence on the subsequent melt-down scenario and the corresponding chances of accident management measures). With further progression of core melting, the issue rather becomes to narrow down time margins and possible evolution variations also under conditions of severe core damage and effluence of melt from the core into the lower head of the reactor pressure vessel. Based thereon, chances of interventions even with such advanced progression of failure are to be investigated. This is considered to support back-up measures and to reveal their general effectiveness even without claiming absolute solutions, but before reaching bounding conditions. The formation of a melt pool in the core region can be decisive for the potentiality of cooling during this phase as well as for the possible formation of coolable structures in the lower head. The melt can in principle be relocated from the core without or with limited prior accumulation in a pool as a relatively continuous, slight flow, as effluence from a melt pool in the form of melt jets or in a larger pour. This gives direction to the constellation forming in the lower head. E.g., potentially coolable debris beds with sufficiently solidified particles can be obtained or a melt pool which hardly can be cooled by water injection into the RPV. The latter situation may be deteriorated by overlying metallic melt layers and resulting effects of focussing heat fluxes on local regions, caused by the relatively high thermal conductivity of the metal. Then, the time development of the thickness of such layers during buildup of the melt pool in the lower head becomes important. This depends on the melting and relocation of metallic parts. Such conditions determine the feasibility of cooling concepts by external vessel flooding. Finally, time margins for RPV failure, the kind of RPV failure and the mode of outflow of melt from the RPV determine concepts of corium retention in the reactor containment. Thus, major effects to be considered within the late phase behaviour are the tendency towards melt accumulation in the core and the resulting outflow modes of melt from the core as determining the formation of coolable structures. Releases starting from the upper edge regions of melt pools in the core leading to rather gradual melt outflow as well as slow melt progression without major accumulation occurring in the core further reduce the potential threat of the RPV by steam explosions. Strong explosions may rather occur in scenarios with collapse-like failure of the structures supporting a molten pool and the resulting large pours of melt into the residual water of the lower head. On the other hand, early melt flows into the lower head could induce earlier heat-up of the vessel wall and thus favour premature failure of the RPV. Ex-vessel melt retention concepts could be affected by this and by the integral duration of subsequent batches of melt release to the lower head. Depending on the concept, the presence of water can be beneficial or counterproductive.

18 4 CHAPTER INTRODUCTION At large, melt retention, its associated cooling and the extent of fission product release are the central problems and points of orientation also for the late phases of an accident. The involved major issues and tasks of the modelling can be summarised as follows: time development of the core states: evaluation of time margins, effect of dedicated measures (coolant injection), depending on the state of the core, mode of melt release to the lower head, mass flow rates, state of melt (temperature, composition, i. e. ceramic and metallic parts), hydrogen and fission product release into the primary circuit to determine the source terms for the containment..3 Approach and outline of the present work At first, an overview on the state of the art is given. The major processes and phenomena for the evaluation of severe accidents in LWR are outlined, with emphasis on the late phases of core degradation. A short compilation of the present knowledge available on these phases from experiments and the TMI accident is presented, as well as the modelling approaches followed in computer codes for the analysis of core melt accidents. A two-dimensional, quasi-continuum model for description of the late phase core degradation behaviour has been developed as part of the KESS code system. The reasoning for the major modelling assumptions also in comparison with alternative approaches in other codes is presented. The resulting governing equations are given, together with a description of the numerical solution methods. The model has been implemented in the system code ATHLET-CD. This allows the integral simulation of the behaviour of the reactor cooling system, taking into account realistic boundary conditions for the core, especially the cooling conditions governed by the development of the residual water level in the core and the downcomer. The coupling concept with other components of ATHLET-CD as well as its realisation is described. In order to verify the modelling and to evaluate the capability to extrapolate to real reactor conditions calculations have been performed. On the one hand, specific model aspects have been tested on the basis of simplified test cases with examinations of the sensitivity to main modelling variants and major parameters. On the other hand, the model has been applied in post-test calculations to the PHEBUS experiments FPT and FPT4. The results of these calculations are presented and assessed with respect to the objectives given in the previous section. Furthermore, application calculations are presented for the integral simulation of the late phase core melting in a PWR up to the release of melt from the core, oriented at the accident in the TMI- plant. These calculations concentrate on the possible range of melt release modes depending on the cooling conditions. Finally, in the conclusions, the major results are discussed with respect to their impact on further accident progression and resulting coolability options. Further perspectives for improvement and completion of the modelling are outlined.

19 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART 5 Severe accidents with core melting State of the art and experimental database. Physical phenomena and processes in severe core melt accidents The ultimate aim of safety considerations in nuclear power plants is to ensure the retention of radioactive nuclides. The so called defence in depth safety concept foresees several barriers for the safe inclusion of fission products and safety systems and measures to protect the barriers. These barriers are the lattice of the fuel, in which the major part of the fission products are bound, the leak-retention of the fuel rod claddings, the reactor pressure vessel, together with the connected reactor cooling circuit, the leak-proof and pressure-resistant safety containment, which encloses the pressure vessel and the cooling system. Within the safety concept, staged measures dedicated to protect these barriers are foreseen: Strong requirements to the quality of design, construction and operation shall, as far as ever possible, prevent events which could endanger the barriers. In case of operational faults automated control and limiting facilities help to keep the plant within permissible design bounds during normal operation. A variety of additional safety systems are foreseen (e.g. high and low pressure safety injection, emergency backup generators), which protect the barriers against a spectrum of possible accidents, so called design base accidents. These systems are implemented redundantly and diversitarily. They are activated and controlled automatically, thus no manual interventions are possible approximately into the first 3 minutes of an accident. Even in case of the failure of these dedicated safety systems, the function of the barriers will be maintained over a longer period of time. This period can be used to activate measures, to mitigate the consequences of the accident, supported by additional safety equipment as e.g. hydrogen recombiners and containment spray systems. Altogether, the core melt frequency in a LWR is very low ( 5-6 per year). It is only possible, if the reactor core is partly or totally uncovered and not cooled by water and the heat produced in the core (radioactive decay heat of the fission products and heat generated by exothermal chemical reactions) can not be discharged by the cooling systems. Sequences leading to core melting are quite complex, and in general require simultaneous failure of several safety systems. They will be exemplified in the following by two representative scenarios of a PWR... Severe accident sequences leading to core melting Among the considered potential accident sequences the Station Blackout (SB) scenario has the highest probability to lead to core melting. It can be initiated by total loss of offsite power and non-availability of emergency cooling systems due to failure of emergency backup generators. Upon initiation of the accident, reactor scram will be activated by insertion of the control rods to stop the nuclear chain reaction. Tripping of the turbine and of the coolant pumps will cause the secondary side of the steam generators to dry out within a short period of time. Consequently, the heat generated in the core can not be discharged from the primary coolant system via the

20 6 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART secondary cooling system. The pressure will rise in the still intact primary circuit due to heat-up and evaporation of the coolant. As long as the system is not depressurised, the water in the reactor coolant system will boil-off under high pressure through the safety valves. However, the potential failure of the pressure vessel due to core melting under high pressure carries particular hazards, which can lead to early failure of the containment (formation of projectiles with large momentum, fast pressure buildup direct containment heating, DCH). Therefore, present accident management strategies foresee and aim at exclusion of the high pressure path by depressurisation of the primary circuit through the Pilot Operated Relief Valve (PORV) usually situated on top of the pressuriser ( exclusion by design ).The loss of coolant will finally lead to dry-out of the core. The time period until core uncovering clearly depends on the specific plant design, but usually ranges between and 3 hours. A different scenario leading also to core melting may occur in a large break loss of coolant accident (LBLOCA). Here, e.g. a double ended break of a main coolant pipe or of the surge line between primary circuit and the pressuriser and failure of the active components of the emergency core cooling system (high and low pressure safety injection) is assumed. The pressure in the reactor cooling system is relieved through the break. Since the water discharged through the leak is not available for cooling of the core, the time period until core uncovering is much shorter in the LBLOCA scenario than in the SB scenario. Again, it strongly depends on the plant design, but it is typically in the range of 3 to 6 s. Also, due to the dynamics of the thermal-hydraulic transient, the water level at the beginning of core heat-up may be significantly below the top of the core. The probability of occurrence of an LBLOCA sequence is significantly lower than that of the SB sequence. However, due to the faster course the time span for initiation of accident management measures before failure of the pressure vessel is considerably shorter... Early phases of core melting If the core is not sufficiently cooled by water, the decay power of the fission products in the fuel heats up the reactor core and leads to evaporation of the water. If the water supply is less than the evaporation rate, the water level in the RPV decreases. The fuel and control rods in the dry region of the core heat up. With increasing temperature, several chemical interactions between the materials in the core and formation of liquid phases will occur. An overview is given in Figure.. During the early phase, the following major processes take place, which have been observed e.g. in the CORA and PHEBUS experiments: With increasing temperature, the pressure in the fuel and control rods rises and can lead to creep and rupture of the cladding. This initiates the release of highly volatile fission products (I, Cs) through the cracks. Above K, absorber material (silver-indium-cadmium alloy, AIC) starts to melt and interacts with the iron (Fe) in the stainless steel cladding and subsequently dissolves the Zircaloy (Zry) of the guide tube. Above 5 K local interactions, e.g. between Inconel (from spacer grids) and Zircaloy (from fuel rod claddings or absorber rod guide tubes) lead to formation of small amounts of melt and can cause local penetrations through which fission products are released. The heat-up of fuel and control rods is strongly accelerated at temperatures beyond approximately 5 K by the exothermal reaction between the superheated steam and the zirconium of the rod claddings or the guide tubes (Zr-oxidation). The oxidised surfaces are eroded and ZrO layers form. By this reaction, also large amounts of hydrogen are

21 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART 7 3 Melting of UO 96 Melting of ZrO 9 Melting of UO +x 8 Formation of liquid phase of oxidic UO -ZrO -eutectica start of ZrO /UO interaction (formation of oxidic melt) 75 Melting of UO -x 67 Formation of α-zr(o)/uo und U/UO monotectica Temperature / K 45 Melting of α-zr(o) 7 Formation of α-zr(o)/uo eutectica 3 Melting of "as-received" zircaloy-4 start of Zry/UO interaction (formation of metallic melt) 7 Melting of stainless steel 65 Melting of inconel 5 Liquefaction of inconel/zircaloy start of rapid Zr oxidation and cladding embrittlement 4 Formation of liquid U due to UO /zircaloy interaction Formation of Fe/Zr and Ni/Zr eutectica Melting of absorber material Figure.: Chemical interactions and formation of liquid phases in the core with increasing temperature produced, which arrive in the reactor containment, either through a leak or through the safety valves of the reactor cooling system. High concentrations of hydrogen in the containment can there lead to deflagration or worse detonation of the gas mixture with potential premature failure of the containment. At first, metallic liquid is produced when reaching the melting point of Zircaloy. This melting point depends on the oxygen content of the Zircaloy cladding and thus the previous oxidation history and lies between K and 5 K. The molten Zircaloy interacts with

22 8 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART the fuel and dissolves parts of it. The resulting U-Zr-O molten alloy is rich in zirconium (U content < 3 w %) and is therefore usually denoted as metallic melt. Relocation of melt is usually initiated by mechanical failure of the residual confinement (cladding/guide tube or ZrO layer). The melt flows out through the cracks in the form of streaks and relocates along the rods into colder regions of the core. There, the melt refreezes and forms a crust. This type of melt relocation usually is denoted as candling process. Extended crust formation can lead to blockage of the free flow area. This hampers the access of steam to the region above the blockage and can cause melt to pile up, spread laterally and form a melt pool. The fuel rods and their cladding amount to approximately 9 % of the mass inventory in the core. Material interactions with spacer grids and control rod behaviour mainly have only local effects. The core behaviour in the early phases and its impact on the further progress of the accident therefore is essentially determined by oxidation, metallic melt formation and relocation and the cooling boundary conditions as well as the interaction between these processes: Initially, the heat-up of the core is determined by the axial and radial profile of the decay power distribution and the cooling conditions imposed by the residual water level in the core, leading to maximum temperatures in the central part of the upper third of the core. Radiation heat transfer tends to flatten the temperature distribution. Decay power, heat conduction along the rods and radiation to the region below the water level cause evaporation and further decrease of the water level. The core parts still covered with water are sufficiently cooled, while the heatup in the upper regions is partly delayed by cooling by the steam flow. The exothermal zirconium oxidation strongly accelerates the heat up of the core. Besides the temperature dependent reaction kinetics, the reaction rate is determined also by the amount of steam available for oxidation. The oxidation process can be limited in upper parts of the core due to the consumption of steam in lower parts (steam starvation). The metallic melt relocation and crust formation has an important influence on the temperature evolution. Heat is transported with the relocating melt to colder regions, leading there to faster heat-up and possibly further temperature escalation due to initiation of fast Zr-oxidation. With the fuel content of the metallic melt also fission products are relocated, which leads to a change of the initial decay power profile. These processes will accelerate evaporation and will further decrease the residual water level. The formation of blockages causes deviation of the steam flow to the sides. This influences the cooling and availability of steam for oxidation in the region above the blockage. The amount of zirconium oxidised to ZrO as well as the amount and composition of metallic melt formed is the result of competing processes. Zircaloy oxidation by steam and fuel dissolution by Zircaloy melt counteract. Further, these processes are only effective, when these components are kept in place and are in contact at elevated temperature. These interactions are stopped by cladding failure and subsequent relocation of the melt (mainly Zircaloy) to colder regions. Early melt relocation will result in less hydrogen and ZrO formation and in Zr rich metallic melt. These effects are further influenced by potential steam starvation. If water injection can be re-activated during the early phase of core melting, experiments like LOFT [8], CORA [] and QUENCH [4] indicate that the potential to stop melting and to cool the core is still high. However, large amounts of hydrogen may be produced during reflooding, caused by the formation of fresh surfaces for oxidation through shattering of oxide layers as a result of thermal stresses as well as the high availability of steam replacing possible previous steam starved conditions. This was also observed in the TMI- accident, where a significant

23 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART 9 amount of hydrogen was produced during an attempt to reflood the core (estimated to 6 kg out of 7 kg produced during the whole accident)...3 Late phases of core melting The late phases of core melting and degradation are characterised by the gradual loss of the original bundle geometry of the core and massive formation of melt. With increasing temperature and melting/relocation of the cladding the fuel pellets loose their support and collapse. This leads to the formation of an extended particulate porous debris bed mainly composed of fuel fragments and broken pieces of oxidised cladding piled up upon a previously formed metallic crust or supporting core structures (e.g. lower grid plate). This debris bed may also incorporate stubs of rods. Analysis of the debris generated during the TMI- accident showed that the fuel was fragmented to particles in a size range between 3 µm (length scale of fuel grains) to 4 mm (order of magnitude of fuel pellet size). This range of particle sizes can be explained by different phenomena. Some fragmentation already occurs during the normal operation of the reactor caused by the thermal stresses. During the accident, additional fragmentation is caused by stresses due to the growth of fission gas bubbles within the fuel (fuel swelling), dissolution of some fragments by liquid Zircaloy as well as thermal stresses during possible reflooding with water. The low permeability due to dense packing (estimated porosity of 36 5 %) and small particle sizes of the porous debris allows only limited access of fluid which impedes cooling. The main steam flow is deviated into still intact regions of the core. Heat transport from inside of the debris bed to its boundaries is also very small due to the low thermal conductivity of the mainly oxidic materials (UO and ZrO ), numerous contact resistances and the short range of radiation heat transfer. This favours further heat-up and melting of the oxidic components (UO and ZrO ), which constitute the major part of the mass inventory of the core. Massive formation of oxidic melt is expected to start with the formation of ZrO melt at temperatures above 96 K which dissolves UO, although some melt may already be produced by eutectic interaction at lower temperatures of around 8 K. The velocity of the oxidic melt flowing through the pores of the debris is quite low (~ mm/s cm/s) due to the large interfacial friction. Also, capillary effects have to be considered. Initially, downward melt flow will be stopped by crusts forming in contact with colder parts (debris, previously formed metallic crust or lower core support structures). This causes lateral spreading of melt, lateral crust formation and accumulation of melt inside the crust enclosure. A typical configuration of the core during late phase core melting is given in Figure.. The further development is the result of the competing processes of continuous heat-up, remelting of crusts, refreezing close to still covered boundaries, downwards versus lateral progression of the melt and possible melt accumulation in a molten pool. These processes are essentially governed by thermal boundary conditions depending on: cooling of the surroundings of the crust/melt system by steam flow and heat conduction or radiation to water. Thereby, the development of the residual water level is crucial since evaporation provides a major heat sink to stop downwards melt progression. For sideways melt progression enhanced cooling of lateral regions due to deviation of steam flow by blockages has to be considered, as well as cooling of the radial core surroundings (core baffle, reflector) by steam flow and the water level in the downcomer, which can be higher than in the core, thermal inertia depending on the heat-up history. Especially previous metallic melting plays an important role. On the one hand, remelting of metallic crusts acts as a heat sink. On the

24 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART void dry or partially saturated debris accumulated melt / pool steam flow Largely intact rods or rod stubs saturated debris crust residual water Figure.: Typical configuration in the core during late phase core melting. other hand, metallic melt acts as a precursor by pre-heating lower structures and causing faster water level decrease during downwards relocation, distribution of heat sources from decay power and chemical reactions. The decay power with an initial axial profile of an approximate cosine shape supports steep axial temperature gradients in the lower part of the core, which favour formation of stable crusts. The power distribution is however modified due to relocation of fission products with the fuel melt and fission products release, specific design features of the reactor. Melting of structure components with large mass (e.g. massive fuel rod end plugs, heavy reflector) provides a temporary heat sink which delays melt progression. Depending on the downwards versus sideways progression of the melt/crust relocation, in principle two bounding configurations can be identified: without the formation of a stable bottom crust, larger amounts of melt cannot accumulate in the core. Melt would then be gradually released from the core to the lower head with mass flow rates corresponding approximately to the melting rate, if stable crusts stop downward and sideways progression, at least when reaching the core boundaries (lower core support plate and core barrel or side reflector), melt will accumulate in the core central region and form a large molten pool. In the latter case, the thermal behaviour of the melt pool is determined by a natural circulation flow that commences in the pool after melting of residual solid parts. The natural circulation pattern generally favours heat transfer to the top and to upper lateral boundaries over heat transfer to the bottom. The stability of the crust depends on the heat flux distribution at the pool inner boundaries and the external heat removal. The thermal loading on the pool confinement is modified by the formation of a metallic layer composed of Zircaloy rich melt and steel, stemming e.g. from melting of upper plenum structures. Due to the higher thermal conductivity of the metallic components heat fluxes will be focused towards the lateral boundary of this layer.

25 CHAPTER SEVERE ACCIDENTS WITH CORE MELTING STATE OF THE ART The thermal loads will finally lead to a failure of the support of the melt pool and an effluence of melt. Depending on the type and location of the failure, three major modes of melt release to the lower head of the RPV are possible: the pool may fall down together with the supporting crust and the lower grid plate or, more likely, it will tilt due to failure of the lower core support structures, resulting in a massive splash of melt into the water still remaining in the lower head, if crust failure occurs in the bottom central part of the pool, the accumulated melt will be released in a continuous stream through the lower grid plate. The melt flow will then be distributed and enter the lower head in the form of several jets, crust failure at the lateral boundary of the pool will most likely appear at the upper rim. Non-symmetric effects will lead to opening of one or several breaches of limited size. Although the breach can widen due to ablation by the hot melt, melt outflow rates will be relatively low due to the limited size of the holes and the limited height of the fluid above. Melt relocation from the core to the lower head through this mode will most likely occur in several batches, spread over a rather long period of time. The progression of core degradation (geometry loss and debris bed formation), melting and melt accumulation during the late phase is decisive for the cooling potential during this phase. This may in fact be quite low for a largely destroyed core. Evidence for this is given by analyses of the TMI- accident []. They revealed that a debris bed formed in the lower central part of the core either prior to or during reflooding. Although this debris bed later was surrounded by water, it could not be cooled and finally led to formation of a molten pool encapsulated in the debris bed, lateral failure of the supporting crust and effluence of melt into the lower head of the RPV. After relocation to the lower head, the corium formed a particulate debris bed, which finally was cooled, although still including a low porosity cake. Thus, the TMI- accident also indicates that a configuration in the core which is not accessible for cooling can be transformed in a coolable configuration through the processes in the lower head. Although the conclusions from the TMI- accident cannot be generalised, they indicate that the results of the late phase of core melting processes dominate the subsequent progression of the accident and the cooling potential...4 Processes in the lower head of the RPV When melt is released from the core, it interacts with the residual water in the lower head. The intensity of the interaction and the resulting configuration of corium debris depends on several parameters which are the results of the previous development. These are mainly the height of the residual water level, the condition of the melt (superheat, composition) and especially the mode of melt relocation from the core to the lower head. If melt enters the lower head in the form of a large, single jet (diameter of ~ 5 cm) after complete failure of the lower core support, the interaction with the water will be limited. The relocated melt will remain largely liquid and will immediately form a molten pool surrounded by crusts in the lower head, possibly with an overlying water layer on top. Such a melt pool in the lower head is almost impossible to cool, even if water injection into the vessel is available. Due to the low thermal conductivity of the oxidic corium and the limited access of water to the crust (only to the top crust and possibly parts of the lateral crust through small lateral gaps), cooling will only be possible if the involved corium masses are quite small (~ 5 t []). Otherwise, if the melt is released from a molten pool in the core through breaches in the supporting crust, it will enter the lower head in the form of one or several melt jets (typically in the range of 5 cm diameter). Outflow from the upper edge of the pool and relocating along the bypass or the downcomer will result in a single or few melt jets, while outflow through a