Comparison of an advanced analytical tool with the simmer code to support astrid severe accident mitigation studies

International audienceThe study presented in this paper deals with the assessment, against SIMMER results, of a physical-probabilistic tool dedicated to molten material core discharge. This 0D tool handles heat transfers from molten, possibly boiling, pools to mitigation tube walls, fuel crust evolution, segregation/mixing of fuel/steel pools, radial thermal erosion of mitigation tube wall, and discharge of molten material with axial thermal erosion of the transverse tube, coupled with neutronic evolution of the fuel power. This tool will be briefly described before presenting the comparison with SIMMER-III results, including a space-and energy-dependent neutron transport kinetics model, on several test cases. This tool, which is very low time consuming, will thus enable large sensitivity studies on different physical and design parameters

A three-dimensional neutronics - thermohydraulics Unprotected Loss Of Flow simulation in Sodium-cooled Fast Reactor with mitigation devices

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ADVANCED STUDIES AND STATISTICAL TREATMENT FOR SODIUM-COOLED FAST REACTOR PIN FAILURES DURING UNPROTECTED TRANSIENT OVERPOWER ACCIDENT

International audienceUsually, simulation tools are validated on experimental data considering a Best Estimate simulation case and there is no quantification of this validation, which remains based on a rough expert judgment. This paper presents an advanced validation treatment of the simulation tool OCARINa, devoted to Unprotected Transient OverPower (UTOP) accidents, on two CABRI tests, considering this time, a Best Estimate Plus Uncertainties (BEPU) approach. The output results of interest are both scalar physical data such as the time and location of the pin failure and associated molten mass and vector data such as temperature axial distribution or temperature evolution versus time. This approach is a first step in quantifying the degree of agreement between the calculation results and the experimental results. It is of great interest for the VV&UQ (Verification, Validation and Uncertainty Quantification) approach, which leads to the qualification of scientific calculation tools

Degraded core relocation in Sodium-cooled Fast Reactor severe accident −- particle-size debris flow

International audienceIn the context of improved safety requirements for generation IV Sodium-cooled Fast Reactors (SFR), an innovative severe accident mitigation scenario is being investigated. The mitigation strategy consists of transfer tubes (DCS-M-TT) and a core catcher. The transfer tubes are dedicated to discharge molten fissile materials from the core center region and guide them towards the core catcher where long-term cooling and sub-critical state are assured. The physical phenomena occurring during the discharge via DCS-M-TT is introduced in this paper. The current demonstration of the mitigation strategy is based on best-estimate calculations with the reference computer code SIMMER. Previous analyses showed that the material discharge via DCS-M-TT can be efficient to avoid re-criticalities and prevent large mechanical energy release. However, uncertainties of SIMMER approach are identified on the relocation process related to a possible particle-size debris accumulation and blockage formation inside the transfer tube. It is believed to originate from the particle treatment in the code. Thus, a review of particulate flow modelling is summarized in this work. The first validation and verification on particle flow treatment in SIMMER is presented. Recommendations for reactor calculations and first orientations for future RandD are highlighted

STUDIES AND CROSS-COMPARISONS OF SEVERE ACCIDENT PREVENTION AND MITIGATION CAPABILITIES OF A SFR AND A GFR

International audienceCEA has developed Generation IV fast neutron reactors with gas and sodium coolants in the two last tenth of years. Namely these reactor projects were the GFR2400 (gas-cooled fast reactor of 2400 thermal megawatts) and ASTRID (sodium-cooled fast reactor demonstrator of 1500 thermal megawatts). The objective of this paper is to provide a cross-comparison of the severe accident prevention and mitigation capability of these reactor concepts based on the work done during a significant time period and taking benefit of the distance since these studies. This comparison can highlight both generic trends resulting from a common study approach used for both concepts and from very detailed results obtained during their conceptual design studies. Despite their power difference and their fuel element design specificities, the main study results and conclusions are quite generical of the two concepts as explained in the paper. Thus they could be extrapolated to other reactor design providing main reactor features are kept (core materials, coolant, neutron spectrum, etc.). The assessment of core melting prevention relies on the study of the natural behavior of the GFR2400 and of ASTRID when facing the various accident sequence families. Then, the efficiency and the features of the systems to be foreseen for core degradation prevention are presented. As far as mitigation is concerned, all the consequences of core melting are investigated (i.e. the induced loadings in terms of nature and of range) by considering various core degraded states. Based on the magnitude of these loadings, the needs of mitigation means are assessed for each concept. Among other trends, the presented work shows the very good prevention capability of the SFR concept but the necessity to mitigate the fast vaporization and expansion of degraded core materials. Conversely, the limited coolant capability of the GFR concept and its low thermal inertia require a pressurized gas circulation into the core, limiting its prevention capability whereas its core melting should not induce substantial mechanical loadings of the reactor structures. However, for this last concept, thermochemical interactions between the core materials are an issue deeply investigated in order to understand and simulate core degradation

Simplified thermohydraulic criteria for a comparison of the accidental behaviour of GEN IV nuclear reactors and of PWRs

International audienceA comparison of 3 Generation IV reactor concepts between them and with a PWR of 2nd generation is presented in this paper. The 3 Gen IV reactor concepts considered have been studied at CEA and are briefly presented in the first part of the paper SFR of 1500 MWth, GFR of 2400 MWth and VHTR of 600 MWth. In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria are aimed at assessing the main physical thresholds to exceed in order to have a core degradation phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is assessed. A third set of criteria deals with core features and are aimed at assessing the possible reactivity insertion that withstands each concept up to core melting and the associated expected power peaks in case of coolant voiding/depressurization and in case of core materials relocation. Finally, a last criterion set deals with the assessment of the possibility to challenge physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concept investigated in terms of severe accident prevention and mitigation

Severe accident studies on the efficiency of mitigation devices in a SFR core with SIMMER code

International audienceSodium-cooled Fast Reactors (SFR) are investigated as future Generation IV reactor concepts. In SFR, the core configuration is not the most reactive during the nominal reactor operation, its geometry change or coolant voiding can induce a reactivity insertion. Within these considerations, mitigation devices should be implemented into the core in order to limit the thermal energy released into the fuel during a severe accident and thereby the possibility to induce mechanical loadings of the reactor structure when vaporized materials expand. Complementary safety devices called Transfer Tubes (TT) are studied in a French concept as an effective mitigation measure of severe accidents to reach a final reactor safe state, even in case of a failure of all shutdown systems including the passive shutdown safety rods. More precisely the TT, with their upper part located in and above the core, cross the core support structure and are linked to the core catcher zone situated in the main vessel bottom. Their purpose is to extract fissile materials from the core zone and secondly to favour the transfer of the molten fuel from the core region to the core catcher in case of severe core damage. The fuel discharge out of the core zone is necessary in order to lower the core reactivity, the relocation into the core catcher is necessary to enable the stabilization, the sub-criticality and the fuel cooling. An extensive work on severe accident calculations and phenomena identification related to fuel discharge through TT was performed. These studies represent the outcome of the work performed in 2014-2019, performed with collaboration of Japanese JAEA and MFBR and Framatome in the framework of the SFR project carried out by CEA. Calculations of fuel discharge were carried out with the mechanistic calculation code SIMMER and the main results are presented in this paper. Firstly, the whole unprotected loss of flow (ULOF) accident sequence was calculated with the SIMMER code. From these calculations, the power and flow evolution were set as boundary conditions for a more detailed model with six fuel assemblies around TT. This model was used to focus on key phenomena and to check impact of design parameter as local bottom restriction inside TT on fuel discharge way

A physical tool for severe accident mitigation studies

International audienceWithin the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R and D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic model, on several test cases. Then some sensitivity studies on design parameters are presented providing preliminary information for this reactor fuel oxide core design

Mitigation of severe accidents for SFR and associated event sequence assessment

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Simplified criteria for a comparison of the accidental behaviour of Gen IV nuclear reactors and of PWRS

International audienceA cross-comparison of four Generation IV reactor concepts and a PWR of 2nd generation is presented in this paper. 4 Gen IV reactor concepts are considered and are briefly presented in the first part of the paper: SFR of 1500 MWth, GFR of 2400 MWth, MSR of 3000 MWth and VHTR of 600 MWth. In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria aims at assessing the main physical thresholds to exceed in order to have a core degradation: phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assessed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is evaluated. A third set of criteria deals with core features and aims at assessing the possible reactivity insertion that withstands each concept up to core melting (or boiling for the MSR) and the associated expected power peaks in case of coolant voiding/depressurization and in case of fissile material compaction. Finally, a last criterion set deals with the assessment of the possibility to jeopardize physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant/fuel and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concepts investigated in terms of severe accident prevention and mitigatio