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

F Bertrand (1) , N. Marie (1) , A. Bachrata (1) , J.-B. Droin (1) , X. Manchon (1)

1 CEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire)

A 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.

Mots clés: Simplified criteria; Reactor accident behavior; SFR; GFR; VHTR; PWR; prevention; mitigation

Domaines: Physique Nucléaire Expérimentale [nucl-ex]; Physique Nucléaire Théorique [nucl-th]

F Bertrand, N. Marie, A. Bachrata, J.-B. Droin, X. Manchon. Simplified thermohydraulic criteria for a comparison of the accidental behaviour of GEN IV nuclear reactors and of PWRs. NURETH 18 - The 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulic, Aug 2019, Portland, United States. ⟨cea-02614127⟩