https://hal-cea.archives-ouvertes.fr/cea-02442330Mengelle, S.S.MengelleCEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) - CEA - Commissariat à l'énergie atomique et aux énergies alternativesDouce, S.S.DouceCEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) - CEA - Commissariat à l'énergie atomique et aux énergies alternativesDamian, F.F.DamianCEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) - CEA - Commissariat à l'énergie atomique et aux énergies alternativesVerification of the depletion calculation scheme of an highly heterogeneous PWR core designHAL CCSD2016corePWRdesignVerificationdepletioncalculationschemehighlyheterogeneous[PHYS.NEXP] Physics [physics]/Nuclear Experiment [nucl-ex][PHYS.NUCL] Physics [physics]/Nuclear Theory [nucl-th]amplexor, amplexor2020-01-16 13:42:332020-04-28 11:28:162020-03-12 16:14:12enConference papersapplication/pdf1In order to enhance the fuel utilization in Light Water Reactors and save resources, it is necessary to increase the conversion of fertile material ($^238$U) and ($^{239}$Pu). This can be obtained by designing low moderation PWR fuel assemblies which enhance the conversion of $^{238}$U into $^{239}$Pu. To ensure a high level of performance, the introduction of fertile elements is mandatory. This kind of design would therefore be complex to model. The flux gradient at the boundary between fissile fuel and fertile blankets is difficult to estimate by using a standard neutronic model based on transport-diffusion calculation schemes. The presence of fissile and fertile zones in the core leads to modify the standard calculation scheme based on a transport-diffusion calculation to take into account this heterogeneity. As a consequence, such transport calculations (first step of the calculation) which are performed in fundamental mode on a 3x3 cluster of fuel assemblies representative of the core loading pattern needs to be verified. Up to now, such verification was performed on static configurations by comparisons with the Monte-Carlo code TRIPOLI-4. New capability for depletion calculation has been recently introduced in the reference Monte-Carlo code TRIPOLI-4 by coupling it with the depletion module of the MENDEL code. This allows now to model the burnup depletion of a complex 3D geometry with TRIPOLI-4. This type of calculation is used as a reference to verify deterministic multi-groups calculations. This paper presents the comparisons between these two types of calculations in order to verify the performances and evaluate the safety criteria of highly heterogeneous core designs.