[en] Full text of publication follows: The Gas cooled Fast Reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV initiative. This is the reference Generation 4 concept for the French Commissariat a l'Energie Atomique (CEA). The generation 4 specifications have been declined in a series of design options and, as far as the thermal-hydraulics is concerned, the most significant one is the use of an helium high temperature gas coolant system with a direct Joule-Brayton thermodynamic cycle. Our objective is to demonstrate the viability of the GFR concept in the frame of an international partnership. The helium design option which is very attractive (chemical inertness, neutron transparency, ..) and it leads to very specific thermal-hydraulic issues. In this paper, one will present the ongoing CEA studies about these issues. The first one is about the core design. Using a ceramic fuel concept with a good conductivity, the main requirement is to obtain an average exit core temperature of 850 deg. C with a maximum fuel temperature of 1200 deg. C and with a low core pressure drop (in order to ease the decay heat removal). For different types of fuel element arrangements (pins, plates, blocks), the main core characteristics have been determined. Two reactor unit sizes have been considered: a medium one (600 MWth) and a large one (2400 MWth). A consistent set of core parameters have been determined taking into account the different constraints including the thermal-hydraulics. A second significant issue is the reactor decay heat removal (DHR). Initially, solutions based on passive systems were searched, the most obvious solution being a system using a natural circulation loop. Here, a significant effort has been conducted with pre-design analyses based on the COPERNIC computer tool, verifications with the CATHARE computer code and detailed analyses of some specific points with a CFD code. A reference solution based on natural convection only and alternatives using low power active systems will be presented. (authors)

[en] The Gas-cooled Fast Reactor (GFR) is one of the six reactor concepts selected within the framework of the Generation IV initiative and is the reference concept for the Commissariat a l^,Energie Atomique (CEA). Two reactor unit sizes have been considered; 600 MWth and 2400 MWth. As far as thermal-hydraulics is concerned, reactor Decay Heat Removal (DHR) proves to be a major issue. The CEA has conducted exploratory design studies to address this issue and a reference solution for the 600MWth reactor has been recommended

[en] The SIMMER-III code was originally developed for the safety analysis of core disruptive accident in sodium cooled fast breeder reactor. In parallel to its development, a systematic programme of evaluation was successfully performed. Using the new detailed pin modelling, SIMMER-III is now able to describe a complete sequence of accident for sodium cooled fast reactor. Some tests of the international CABRI programme were successfully recalculated. Applications to a total sodium blockage in PHENIX reactor or to the fuel removal by control guide tube in Super-PHENIX were also investigated. In a recent past, SIMMER-III applications have been extended to gas cooled reactor analysis. These concern not only the 1970's projects but also the current studies in progress on innovative core concepts for gas cooled fast reactors. Transients (depressurisation, loss of flow, etc.) have been simulated on SIMMER-III code and results compared to those of previous studies. Modelling of new fuel core in steady-state and degradation phases, foreseen for the future gas cooled reactor, is being realised in order to be added to SIMMER-III code. SIMMER-III has also been extensively applied for the safety analysis of critical fast burner reactors with sodium cooling within the CAPRA/CADRA program and now for accelerator driven systems (ADS). Due to the potentialities of SIMMER-III code to represent interactions between hot components in multiphase flows, interaction between lithium -lead and water in a fusion reactor blanket, has been analysed. Results of dedicated experiments (BLAST, LIFUS) have been successfully reproduced by SIMMER-III. In conclusion, SIMMER-III code is flexible to be used for a variety of applications in reactor safety in order to study their behaviour under different transients. (author)

[en] In the framework of Gas Cooled Reactor design assessment, an important point to calculate is the temperature fields on the main structures in nominal and accidental situations in order to determine the consequences on the reactor lifetime. This document presents such thermal-hydraulic and thermal-mechanical studies for the HTGR (High Temperature Gas-cooled Reactor) vessel system in normal operation and pressurized LOFC (Loss Of Forced Coolant) accidents. Thermal- hydraulic calculations address the key issues for pressurized LOFC transients and evaluate the contribution of main design and modeling parameters. These calculations are performed using the CFD (Computational Fluid Dynamics) code STAR-CD. For these transients where the primary system remains pressurized, it is necessary to carry out mechanical analyses on the structures to assess the damage levels reached. Sensitivity studies are conducted taking into account different irradiation levels and types of graphite, different assumptions of mass exchanging rate between the stagnant helium beside the vessel and the coolant in the annular channel between core barrel and vessel, and different assumptions regarding the natural convection of helium. To determine the structure temperatures, the thermal-hydraulic studies show that the conductivity value of graphite reflectors is the main factor for both the normal operation and the accidental situation considered (pressurized LOFC). The thermal-mechanical analyses allow evaluating the consequences of these loading situations for the lifetime assessment of the main metallic structures, namely the core barrel and the pressure vessel. The results obtained show that, for both structures, the damage levels remain below design limitations. (authors)

[en] The revision of the GFR (Gas cooled Fast Reactor) core design (plate type with a honeycomb SiC structure containing separated uranium/plutonium fuel pellets) is underway. The self-breeding searched for is achieved with an optimized core design called 12/06-E. The S/A fuel is divided into 3 zones: the upstream zone, the fissile zone and the downstream zone. The fissile zone is made of 24 fuel plates separated by helium channels. The core coolability in natural circulation for pressurized conditions has been studied. The design and safety criteria are met but no margin remains. Fluid dynamics calculations confirms the maximum pressure drop in the central S/A. Finally, the new core design is considered as acceptable from the thermal-hydraulics point of view. The project is now addressing the feasibility and the design of the S/A fuel. (authors)

[en] In the frame of the IP-EUROTRANS Project, an experimental program, focused on studying the LBE/water interaction has been performed using the LIFUS 5 facility available at ENEA-Brasimone. The physical effects and the possible consequences of this interaction have been evaluated over a wide range of different conditions. Besides the experimental activities, a numerical simulation activity has been performed with SIMMER code in order to better investigate the thermo-hydraulic phenomena involved in the interaction and to confirm the capabilities of the code to simulate this kind of phenomena. The experimental and the calculated results in terms of pressure and temperature evolutions in the system show a good agreement.

[en] SIMMER-III is a general two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-time and energy-dependent neutron transport kinetics model. The philosophy behind the SIMMER development was to generate a versatile and flexible tool, applicable for the safety analysis of various reactor types with different neutron spectra and coolants, up the new accelerator driven systems (ADS) for waste transmutation. Its flexibility also allows the application to non-reactor safety problems as e.g. criticality accidents (JCO). Currently, a three-dimensional version is also available, coined SIMMER-IV. The main backbone for analyses is however still SIMMER-III. SIMMER-III has proven especially well suited for fast spectrum systems as the LMFR, where it is one of the key codes for safety analysis, including its application within licensing procedures. To serve especially the last purpose, the code must be made sufficiently robust and reliable, and be tested and validated extensively. A comprehensive and systematic assessment program of the code has been conducted. This paper gives the major achievement of this assessment program. The SIMMER-III code handles by default LMFR core materials, fuel, steel, coolant, control and fission gas, in solid, liquid and vapor states. The total of 27 density and 16 energy components are modeled in three velocity fields and one structure field, in order that important fluid motions in a degraded core are simulated adequately. The spatial differencing method is based on Eulerian staggered mesh, with a higher-order-differencing scheme to mitigate numerical diffusion. An improved analytic EOS model provides good accuracy especially at high temperature and pressure. Multiple flow-regime treatment is available over the entire void fraction range. An interfacial area convection model improves the flexibility of the code by tracing transport and history of interfaces, and thereby better represents physical phenomena. A generalized and flexible code framework, along with improved numerical stability and accuracy, allows us to apply it to a variety of simple and complex multiphase flow problems. The code assessment program is an ongoing effort. Two major milestones have been achieved in the past by completing two assessment campaigns, Phase 1 and Phase 2. Phase 1 for fundamental code assessment of individual models; and Phase 2 for integral code assessment for key phenomena relevant to LMFR safety. Through this systematic code assessment program, comprehensive validation of the physical models has been conducted step-by-step. The assessment program has demonstrated that SIMMER-III is a state-of-the-art code with advanced models sufficiently flexible for simulating transient multiphase phenomena occurring during CDAs. The various applications of SIMMER will be discussed in a separate paper at this conference. In this paper we will concentrate on the specifics of the code, mainly reflected at its application to core melt accidents in the LMFR. (author)

[en] SIMMER-III is a general two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian,fluid dynamics code coupled with a space-time and energy-dependent neutron transport kinetics model. The philosophy behind the SIMMER-III development was to generate a versatile and flexible tool, applicable for the safety analysis of various reactor types with different neutron spectra and coolants including the new accelerator-driven systems for waste transmutation. Currently, a three-dimensional version is also available, coined SIMMER-IV The main backbone for analyses, however, is still SIMMER-III. SIMMER-III has proven especially well suited for fast spectrum systems such as the liquid-metal-cooled fast reactor where it is one of the key codes for safety analysis, including its application within licensing procedures. To serve especially the last purpose, the code must be made sufficiently robust and reliable and be tested and validated extensively. A comprehensive and systematic assessment program of the code has been conducted. This paper gives the major achievements of this assessment program. The SIMMER-III code handles by default liquid-metal-cooled fast reactor core materials-fuel, steel, coolant, control rod, and fission gas, in solid, liquid, and vapor states. The total of 27 density and 16 energy components are modeled in three velocity fields and one structure field in order that important fluid motions in a degraded core are simulated adequately. The spatial differencing method is based on Eulerian staggered mesh with a higher-order differencing scheme to mitigate numerical diffusion. An improved analytic equation-of-state model provides good accuracy especially at high temperature and pressure. Multiple flow-regime treatment is available over the entire void fraction range. An interfacial area convection model improves the flexibility of the code by tracing transport and history of interfaces and thereby better represents physical phenomena. A generalized and flexible code framework, along with improved numerical stability and accuracy, allows us to apply it to a variety of simple and complex multiphase flow problems. The code assessment program is an ongoing effort. Two major milestones have been achieved in the past by completing two assessment campaigns, Phase 1 and Phase 2: Phase 1 for fundamental code assessment of individual models and Phase 2 for integral code assessment for key phenomena relevant to liquid-metal-cooled fast reactor safety. Through this systematic code assessment program, comprehensive validation of the physical models has been conducted step-by-step. The assessment program has demonstrated that SIMMER-III is a state-of-the-art code with advanced models sufficiently flexible for simulating transient multiphase phenomena occurring during core disruptive accidents. This paper concentrates on the specifics of the code, mainly reflected in its application to sodium experiments related to the safety of liquid-metal-cooled fast reactors. (authors)

[en] CABRI is an experimental pulse reactor at the Cadarache research center. CABRI is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (³He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the ³He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The ³He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/³He capture reactions that induce a gas heating. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Two methods were identified to evaluate the gas density evolution: CFD calculations and reverse point kinetics. The first one consists in adding a heat source in transient rods based on the experimental power conversion. The second one consists in using the measured power by boron ionization chambers to evaluate the net reactivity by a reverse point kinetics (PK) method and to subtract the reactivity feedbacks calculated with the DULCINEE multi-physics code. (authors)

[en] SIMMER-III and SIMMER-IV are general two-dimensional and three-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics codes coupled with space-, time- and energy-dependent neutron transport models. The philosophy behind the SIMMER development was to generate a versatile and flexible tool, applicable for the safety analysis of various reactor types with different neutron spectra and coolants. Recently the code has been extended and applied for new designs as accelerator driven systems (ADS) for waste transmutation and molten salt reactors. Its flexibility also allows the application to non-reactor safety problems as e.g. criticality accidents (JCO) or fire safety of fuel casks. SIMMER-III has been originally developed for solid fuel fast spectrum systems as the LMFR, where it is one of the key codes for safety analysis, including its application within licensing procedures. To serve especially the last purpose, the code must be made sufficiently robust and reliable, and be tested and validated extensively. Originally, SIMMER has been mainly applied within the framework of severe Core Disruptive Accidents (CDA). The application has however been extended to cover transients within DBC (Design Basis Conditions) and DEC (Design Extension Conditions) where no core degradation takes place. Examples of this application are provided, namely analyses for an accelerator driven subcritical Pb/Bi cooled reactor, a critical fast transmuter cooled with supercritical water, and finally a molten salt reactor. The extension of the application domain requires cooperative use of SIMMER with other code systems as described in this paper. (author)