3 rd LEADER International Workshop Bologna 7 th September 2012 Potential evolution of Severe Accident Codes (ASTEC-MELCOR) for LFR for LFR Mirco Di Giuli Bologna University

For Severe Accidents we intend those accidents that result in a catastrophic fuel failure, partial or extensive core meltdown, and finally possible release of radioactive materials into the environment. SEVERE ACCIDENTS

The numerical simulation of severe accidents general may adopt two approaches:  Integral codes or code systems: that simulate the entire accident, from the initiating event to the possible release of radionuclides outside the containment and taking into account the main safety systems. (ASTEC, MELCOR, MAAP4)  Detailed or mechanistic codes: that provide a finer simulation of single phenomenon or group of phenomena. They are important for interpreting experiments, understanding various phenomena and developing simple models. (SCDAP/RELAP, ATHLET/CD, ICARE/CATHARE,)

ASTEC & MELCOR ASTEC (Accident Source Term Evaluation Code) is an integral source term code developed by IRSN-GRS for the study of severe accidents in Pressurized Water Reactor, especially for French and German reactors. MELCOR is a fully integrated, engineering-level computer code developed at Sandia National Laboratories for the US Nuclear Regulatory Commission, as second- generation plant risk assessment tool and Source Term Code Package. (BWR,PWR) The aim of these codes is to simulate hypothetical severe accident sequences in light water reactors (LWRs), from the initiating event up to the release of radioactive elements out of the Nuclear building. The main applications are: Evaluation of possible releases of radioactivity outside the containment; PSA2 studies; Accident management studies, with emphasis on measures for prevention and mitigation of severe accident consequences; The source term evaluation is the compute of the actual release, vs time of the different radioactive substances to the outside environment

These codes must have the following requirements: sufficient validation to cover the main physical phenomena; take into account safety systems and procedures; fast running code PLANT SPECIFIC,SAMG,SYSTEM PERFORMANCE Very diverse Scientific domains Source term sequence phenomena

MELCOR CODES INTEGRATION

ASTEC STRUCTURE Module simulates phenomena Overall plant response Easier validation Nuclear Safety experiments Parametric models

LFR VS PWR  Fast spectrum, compact core design (Thermal spectrum, most reactive core configuration under operating conditions)  High power density (~400 kW/liter, PWR 100 kw/liter)  Low coolant velocity (~2 m/sec, 6-8 m/s)  High linear pin power (~8 kW/ft, PWR 5 kW/ft), short core height (~ 1.5 m, PWR ~ 4 m)  Fuel pins on a tight triangular lattice (P/D~1, PWR ~ 1.5) grouped in hexagonal ducted fuel assemblies  Oxide fuel with a high fissile content (~20% LFR, PWR ~4%)  Core coolant rise ~80°C (PWR ~18°C), outlet ~480°C (PWR 330°C)  Low pressure coolant (~atmospheric, ~ 150 bar)  Very large temperature margin to boiling ( > 1000°C, PWR °C) Main difference:

General Safety Aspects of LFRs From a safety point of view: Potential Advantages :  High boiling point (T ~ 1745 °C) and low vapor pressure of the coolant  Primary system at atmospheric pressure  Good compatibility with water (secondary coolant), and no fire risk with air  Low pressure drop across core  Natural convection capabilities  Passive decay heat removal (DHR) in natural circulation  High capability of fission product retention and high shielding of gamma radiation Potential Drawbacks :  High corrosivity/erosivity of molten lead (pumps, clad, vessels, etc.),  Overcooling potential & freezing  Oxygen control under all operational and accidental conditions  Difficult in-service inspection  Potential of blockage formation by deposition of oxidation products  Possible difficulties in detection of accidental conditions  Positive coolant feedback and void worth  Potential of SGTR and resulting steam ingress in core, CCI (coolant- coolant interaction)

Fuel relocation and re-criticality issue : Another aspect that to be taken into account id the re-criticality potential. The LWRs have a significant lower probability of a power increase accidents due to reactivity insertion. The reactivity induced accidents (RIA) are not considered as the safety design basis accident. For LFR reactors:  In case of clad failures, fuel dispersion is considered the dominant phenomena respect fuel compaction. This effect should reduce the risk of re-criticality but does not eliminate it.  For a correct severe accident simulation is very important to know the relocation of SM/fuel melted inside the vessel. Possible interaction melted material steam generators.  In case of large core melt, its even more important to know the fuel relocation to determine the decay heat sources locations and if re- criticality could result from local fuel accumulation.

NOW THE QUESTIONS ARE:  What are the models can be utilised to describe a severe accident sequence in LFR already present in ASTEC and MELCOR?  What are the models that must be implemented ?

CORE MODELING ASTEC/ICARE-MELCOR/COR  ICARE and COR are respectively ASTEC module and MELCOR package that simulate in-vessel core degradation phenomena. THERMAL THERMAHYDRAULIC MODELS MECHANICS MODELS PROCESS SEVERAL CHEMICAL REACTIONS INCORPORATE FP RELEASE MODELING FR,CR,GR,SH,VS

CORE MODELING ICARE and COR do strong approximations on the geometry of the core and its evolution during the degradation process, and remain very sensitive to the physical-chemical uncertainties, due to the large number of components in interaction and the very high temperature.

AP 1000-ICARE models Early phase AP 1000-ICARE models Later phase During core degradation the phenomena involved change: in a early phase the codes have to consider a set of phenomena with a fixed geometry, fixed material, when the core melt the geometry and the phenomena involved are completely different.

FPT-3 EARLY PHASE CORE MODELFPT-3 LATE PHASE CORE MODEL Change in material properties

EARLY PHASE MODELLING 1. HEAT TRANSFERS 1.1 HEAT EXCHANGES 1.2 HEAT GENERATION 2. CHEMICAL INTERACTIONS 2.1 OXIDATION 2.2 INTERACTION 3.MECHANICS BEHAVIOUR 3.1 THERMAL EXPANSION 3.2 CREEP,BURST,DISLOCATION,RUPTURE 4. MOVEMENTENT IN THE EARLY PHASE THE PH. INVOLVED DESCRIBED BY THE CODES ARE THE FOLLOWING: THE PHYSICAL OBJECTS ANALYSED (FUEL,CLADS,SHROUDS,CONTROL RODS, SPACER GRIDS,PLATES,FLUIDS) ARE LINKED BY MEANS OF VARIOUS PHYSICO-CHEMICAL MODELS. DURING THE EARLY PHASE OF A SA THE CODES CONSIDER THESE INTERACTIONS AMONG THE PHYSICAL OBJECTS :

RADIATIVE EXCHANGE WITH WALL AND FLUID CONVECTION WITH FLUID RADIAL CONDUCTION AXIAL CONDUCTION GAP HEAT TRANSFER MECHANISMS INTERNAL POWER GENERATED These codes can consider six mechanism of heat transfer

CHEMICAL INTERACTIONS DURING THE EARLY PHASE THE CODES CONSIDER THREE OXIDATION PROCESSES: REACTION BETWEEN A METALLIC MATERIAL AND EITHER AIR OR STEAM EXOTERMIC REACTION AND RELEASE OF HYDROGEN Regarding fast reactor To validate the codes also for LFR, it have to be considered lead oxidation. It’s necessary add the oxidation of lead in MELCOR and ASTEC codes, in case of SGTR (steam generator tube rupture), but however the reaction is largely known, however chemical inactivity of lead excludes possibility for fires or other strongly exothermic reactions. THE USER CAN CHOICHE DIFFERENT MODEL TO DESCRIBED THESE REACTIONS :

INTERACTION BETWEEN TWO STRUCTURE MATERIAL DURING THE E.P. THE CODE CONSIDER 5 DIFFERENT INTERACTIONS AMONG SM:  DISSOLUTION OF A MATERIAL BY ANOTHER MATERIAL,  LIQUEFACTION OR DISSOLUTION OF A CERAMIC MATERIALS BY A METALLIC MATERIAL,  PECULIARITY : Reactions often leading to the formation of a special eutectic with a rather low melting point Regarding fast reactor It’s necessary experimental campaigns intended to characterize the interactions (dissolution) taking place at high temperature among lead-iron, lead- boron, lead-uranium and lead-plutonium, the interaction among their oxides, oxides-metal and the possible formation of eutectic.( fusion phenomena at lower temperature)

MECHANICS MODELS Regarding fast reactor It’s necessary investigate the mechanical property of T91 or AISI 316 in the higher temperature range, in presence of lead, to determine the embrittlement criteria and the creep behaviour and develop simple models to implement in the code. The code modeling the mechanical behaviour of the fuel rods and control rods, considering : THIS MECHANICS MODELS ARE REFERRED TO ZIRCALLOY

MOVEMENT OF MATERIALS CANDLING: AXIAL MOVEMENT OF MOLTEN LAYER DECANTING: RADIAL MOVEMENT OF MOLTEN LAYER REGARDING FAST REACTOR Compare a PWR reactor the movement of the molten material inside a LFR reactor is more complex, it’s necessary simulating relative material motion processes in Heavy Liquid Metal. These codes do not consider the possibilities that the fuel accumulation in a fast reactor could introduce local positive reactivity and consequently a power excursion. It is also necessary develop a model that considering the re-criticality phenomena. The codes regard three kinds of relative motion of the structures considered:

LATE PHASE MODELING During the late phase, fuel melts and a molten pool forms in the core. Due to eutectic formation, temperature remains below the UO2 melting point (3100 K) by several hundreds of degrees. As the molten mass increases, the pool expands axially and radially in the core until it reaches either the baffle or the lower core support plate, leading to corium relocation towards the lower plenum. The codes during this phase consider :

CORIUM MODELING MODELLING BASED ON RASPLAV–AW POST TEST EXAMINATION REGARDING FAST REACTORS To validate these codes also for LFR is necessary obtain relevant data on the physical and thermal behavior of fast-reactor corium, to derive thermal-physical property data for various molten core materials and develop new parametric models.  THE CODES SIMULATE THE POSSIBLE STRATIFICATION BY MEAN OF PARAMETRIC MODELS

ASTEC/CESAR and MELCOR/CHV compute the thermal-hydraulics in the PRIMARY (SGs) and SECONDARY circuits, before the accidental transient, they also calculates the thermal-hydraulics in the vessel (with an adapted core modeling) in stationary state. The codes divide the primary circuit in control volume, chosen by the user, utilize simple models. REGARDING FAST REACTOR: To validate these codes for LFR is necessary that this module/package is able to take into account the possible fuel, SM transport outside the core, and also consider that operation of steam generators can be influenced by the interaction with clad, structural materials and fuel melted. He also consider the potential freezing of the coolant that could cause blockage issue. THERMODYDRAULICS PHENOMENA

FP TRANSPORT PHENOMENA SOPHAEROS-ASTEC MELCOR –RN predicts transport in the reactor coolant system (RCS) of vapours and aerosols formed by condensation of material released from the degraded core term. SOPHAEROS module utilize model for chemistry of the FPs more detailed respect RN. More generally these codes computes aerosol and/or vapour behaviour in a piping system. From a safety point of view : ⇒ Calculates retention of radionuclides in the RCS : direct potential of the source term. ⇒ Calculates aerosol size distribution and chemical speciation of aerosol and vapour phases in RCS and the possible release in the containment (natural or engineered mechanisms) REGARDING FAST REACTORS: To validate these codes for LFR,t’s necessary evaluate how the presence of lead can affect the behaviour of FPs ( chemical reaction, kinetic reaction) and the amount of FPs that lead can hold. Experimental test are needed.

 Fast Reactor Core meltdown and determination of corium properties: Investigation by mean of simulation with detailed codes that consider neutronics and TH, and tests are necessary.  Fuel/lead eutectic formation: Test are necessary to investigate the possible eutectic formation and its physical property.  Identified failure mechanism for fuels and clad under lead conditions. (Structural materials, cladding and grids melt before coolant boiling). Experiments and development of models are needed.  Identified the behaviour of fission product in lead and retention factor of the lead. (Aerosol formation and deposition mechanism)  Identified possible relocations of steel and fuel in liquid lead: It’s very important to know the location of decay heats sources.  Identified the possible impact on steam generator of core/structural material melted  Investigate if in some severe accident scenario Re-criticality phenomena is possible, and its more likely locations. CONCLUSION: In order to validate the integral codes also for LFR reactor is necessary investigate regarding these phenomena: DATA IS NEEDED TO DEVELOP EMPIRICAL MODELS TO IMPLEMENTED IN INTEGRAL CODE