Analysis of the thermal-mechanical behavior of SFR fuel pins during fast unprotected transient overpower accidents using the GERMINAL fuel performance code

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume

Multiphysics analysis of power transients based on power system and nuclear dynamics software chaining

International audienceIn the framework of energy transition, the massive insertion of Variable Renewable Energies leads to think to new ways of controlling and stabilizing electrical networks. Small Modular Reactors (SMRs) could be a sustainable solution if they are proven to be flexible enough. Previous studies have shown the positive influence of SMRs on grids at short-time scales due to inertia and frequency regulation phenomena. In addition, this article aims at studying the influence of constrained grid events on nuclear systems safety and operation. The models of a power system dynamics (PowerFactory) and a nuclear dynamics software (CATHARE) are compared and chained. Two application cases are carried out to quantify the impact and the relevance of this chaining: a short-circuit and a load loss. This article finally concludes that this chaining is relevant to accurately simulate nuclear reactor behavior following grid events. Moreover, a chaining could be insufficient for electrical simulations after severe events such as shortcircuits or for high nuclear insertion's rate in an energy mix, a coupling, i.e co-simulation, could be considered

Multiphysics analysis of power transients based on power system and nuclear dynamics software chaining

International audienceIn the framework of energy transition, the massive insertion of Variable Renewable Energies leads to think to new ways of controlling and stabilizing electrical networks. Small Modular Reactors (SMRs) could be a sustainable solution if they are proven to be flexible enough. Previous studies have shown the positive influence of SMRs on grids at short-time scales due to inertia and frequency regulation phenomena. In addition, this article aims at studying the influence of constrained grid events on nuclear systems safety and operation. The models of a power system dynamics (PowerFactory) and a nuclear dynamics software (CATHARE) are compared and chained. Two application cases are carried out to quantify the impact and the relevance of this chaining: a short-circuit and a load loss. This article finally concludes that this chaining is relevant to accurately simulate nuclear reactor behavior following grid events. Moreover, a chaining could be insufficient for electrical simulations after severe events such as shortcircuits or for high nuclear insertion's rate in an energy mix, a coupling, i.e co-simulation, could be considered

Small Modular Reactor-based solutions to enhance grid reliability: impact of modularization of large power plants on frequency stability

International audienceIn the current renewable energies’ expansion framework, the increasing part of intermittent electricity production sources (solar or wind farms) in the energy mix and the reducing part of thermal power stations that are nowadays useful to ensure grid stability will lead to a complete paradigm shift concerning the means to ensure grid stability. Nuclear energy, which is carbon-free and dispatchable, may be a sustainable solution to this grid reliability issue if it is adequately designed and implemented on the grid. Several solutions aiming at improving the future nuclear power flexibility are currently under investigation in the literature, among them are those based on Small Modular Reactor (SMR) plants. In order to demonstrate their potential ability to stabilize electric grids, it is necessary to perform electrical dynamic simulations taking into account a spatial and temporal discretization of the grid. In this paper, such calculations are performed using the PowerFactory software. This tool can reproduce electrical grids thanks to models of turbo generators, lines, transformers, loads, I&C systems, etc. The objective is to assess to what extent the innovative SMR features may enhance the frequency control of a grid. For this purpose, a short-circuit event and three frequency stability criteria are firstly defined. Then, a verification of the correct behaviour of the IEEE 39-bus (or New England) grid with regulations is carried out. The relevance of implementing Small Modular Reactors (SMR) instead of large power plants on such frequency stability criteria on this grid is finally assessed, in order to conclude in a preliminary way the possible contribution of small reactors to the future grid’s sustainability

Multi-physics calculations of FFTF loss of flow without scram transient: impact of neutronics and thermomechanics on CATHARE3 models

International audienceThermal hydraulics computational codes used for the design studies and safety assessment of sodium fast reactors need to be able to model a large scope of scales and physical phenomena. IAEA’s Coordinated Research Project on the Fast Flux Test Facility (FFTF) sets a landmark for the validation of such codes by providing precious experimental data on a reactor scale. The transient under consideration is the Loss Of Flow WithOut Scram (LOFWOS) in which the primary circuit undergoes a forced to natural convection transition after the primary pumps trip. In order to predict this transient correctly, thermos-hydraulic codes need to take into account various time and dimension scales as well as interaction with other physics (such as neutronics and thermo-mechanics).Neutronic feedbacks, power distribution, Pellet Cladding Mechanical Interaction (PCMI) and fuel thermal behavior have been calculated based on FFTF’s fuel composition using CEA in-house codes, namely PARIS/ERANOS for neutronics and GERMINAL V2 for thermo-mechanics. Results from these codes are given to CATHARE3 thermal-hydraulics system code. This paper presents the impact of neutronics and themo-mechanics inputs on a system scale calculation

CADOR “Core with Adding DOppleR effect” concept application to sodium fast reactors

Generation-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters

Approach for the adaptations of a nuclear reactor model towards more flexibility in a context of high insertion of renewable energies

International audienceThe massive penetration of renewable energy sources (RES) that are variable and not “dispatchable”, may weaken the power system supply-demand balance. Nuclear power plants (NPP) contribute in part to this daily and seasonal balance thanks to the “load-following” mode in France for example, but there are still limits to their use. These limits prevent a nuclear power modulation as efficient and quickly as the conventional thermal power plants. The need in terms of power ramps for nuclear in a constrained power system has been quantified in previous studies. Nuclear may compensate for the removal of thermal power plants, in order to fulfill energetic strategies of CO 2 reduction. The possibility that nuclear reactors can achieve power ramps of significant values (>5%Pn/min) is put forward and could make possible to replace the services currently provided by thermal power plants. The objective of the study is then to use these power system requirements as the main input parameter for the modelling of a current simplified nuclear reactor capable of responding to frequency control within a specific hypothesis framework. In this paper, a French 1300 MW pressurized water reactor is modelled. Parametric studies are carried out in order to reveal technical and technological constraints when increasing electric power ramp. The study explores ways of design, which may influence reactor flexibility, such as the neutron parameter, Doppler coefficient, or the thermohydraulic parameter, delay in the primary loop

Simplified approach to determine the requirements of a “flexible nuclear reactor” in power system with high insertion of variable renewable energy sources

International audienceThe objective of the paper is to study the potential behaviour of a power system with high share of nuclear and less thermal plants, in which variable RES insertion increases − for example the French case −, in order to determine the specifications for the design of a potential nuclear reactor with high “manoeuvrability”. Moreover, the flexible reactor may participate more in the supply − demand balance and in particular during large frequency fluctuations caused by the high variability of RES. The studies are carried out with the PowerFactory software, which make it possible to highlight specific needs regarding the power ramp for an “ideal” flexible nuclear reactor. Using a benchmark network, the Kundur “2 areas-4 machines”, the flexibility requirements are obtained as a function of the grid disturbances. For this purpose, the penetration of variable RES is progressively increased, while nuclear power is reduced and thermal power plants are totally suppressed. The study shows that a drop in RES production directly impacts the minimal frequency. A faster response speed of nuclear power makes it possible to restore this stability and return to normal operating conditions imposed by the grid operator. This paper describes therefore the process of obtaining the flexibility criterion for different cases of insertion of variable RES