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ASTRID-DYN3D: XS generation and full core simulations

To improve the public acceptance of the future nuclear power in Europe we have to demonstrate that the new reactors have significantly higher safety level compared to traditional reactors. The ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) aims at enhancing further the safety of Generation-IV SFRs and in particular of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the European Sustainable Nuclear Industrial Initiative (ESNII) roadmap and in close cooperation with the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) program. The project aims at 5 specific objectives: Produce new experimental data in order to support calibration and validation of the computational tools for each defence-in-depth level. Test and qualify new instrumentations in order to support their utilization in the reactor protection system. Perform further calibration and validation of the computational tools for each defence-in-depth level in order to support safety assessments of Generation-IV SFRs, using the data produced in the project as well as selected legacy data. Select, implement and assess new safety measures for the commercial-size ESFR, using the GIF methodologies, the FP7 CP-ESFR project legacy, the calibrated and validated codes and being in accordance with the update of the European and international safety frameworks taking into account the Fukushima accident. Strengthen and link together new networks, in particular, the network of the European sodium facilities and the network of the European students working on the SFR technology. By addressing the industry, policy makers and general public, the project is expected to make a meaningful impact on economics, environment, EU policy and society. Selected results and milestones achieved during the first eighteen months of the project will be briefly presented, including proposal of new safety measures for ESFR; evaluation of ESFR core performance; benchmarking of codes; experimental programs; and education and training.

ATHLET is a thermal-hydraulic (TH) system code developed at the GRS for the modeling of Light Water Reactors (LWRs). To extend the applicability of ATHLET to the analyses of Sodium Fast Reactors (SFRs), the code was recently upgraded with the thermal-physical proper-ties of the liquid sodium. The new extension is still under verification and validation phases. The present work contributes to the verification efforts. This study investigated the performance of the extended version of ATHLET as applied to the transient analysis of a set of start-up tests conducted at the Superphénix SFR. The specifications of the corresponding tests such as the simplified SPX reactor core models and the set of reactivity coefficients were adopted primarily from a previous dedicated study performed at PSI and at KIT. The re-activity effects accounted for by ATHLET included fuel Doppler effect and thermal expansion effects of sodium, fuel, diagrid, control rods driveline, strongback, and reactor vessel. The results obtained by ATHLET for main stationary TH parameters, power evolutions, and reactivity feedback components were benchmarked against the reference solutions provided by TRACE. Employing an identical set of reactivity coefficients, either in steady-state or transient calculations, the codes produce consistent and close results.

In the last few years, and within the framework of different European projects, KENO-VI code from SCALE system has been employed to perform detailed continuous-energy Monte Carlo transport calculations for advanced fast reactors. The core characterization of both the sodium-cooled ASTRID and the lead-cooled ALFRED reactors was performed during the FP7 cross-cutting ESNII+ project; more recently, core calculations for the sodium-cooled Superphénix reactor and the improved European Sodium Fast Reactor design were performed within the HORIZON 2020 ESFR-SMART project. In all cases, the effective multiplication factor predicted by KENO-VI was systematically higher (around 400-500 pcm) than the values computed by MCNP and Serpent Monte Carlo codes, using the same nuclear data library. In order to provide insight into the origin of the observed discrepancies, a simplified 2D MOX-fueled SFR pin-cell benchmark has been launched. The multiplication factor, as well as 1-group and VITAMINJ 175-group cross-sections computed by KENO-VI, Serpent and MCNP codes employing ENDF/B-VII.1 data library, have been compared.Significant differences between KENO-VI and the other codes have been found in the unresolved resonance regions of 239Pu and 241Pu capture and production cross sections, while negligible differences appeared outside those energy ranges. On the other hand, calculations without using probability tables have shown very good agreement. Quantitative comparison is presented and analyzed, along with a discussion of the impact of the probability-table treatment in the three codes for MOX-fueled systems with typical SFR spectrum.

Devoted to the Generation-IV European Sodium Fast Reactor safety, the Horizon-2020 ESFR-SMART project was launched in September 2017. Selected results and milestones achieved during the first fifteen months of the project are briefly reviewed in the paper, including 1) proposal of new safety measures for ESFR; 2) evaluation of ESFR core performance; 3) benchmarking of codes; 4) experimental programs; and 5) education and training.

The ESFR-SMART core description has been established in several steps on the basis of ESFR-WH core design proposed for the EURATOM CP-ESFR project and experiences gained in EURATOM ESNII+ project. The axial arrangement above the core (Na plenum and absorber) of CP-ESFR was adopted. For further Na void effect reduction, it was proposed to reduce the inner core fissile height: following late CP-ESFR and ESNII+ studies. Unlike ESNII+, an option was considered to keep the upper fissile boundaries at similar axial locations in the inner and outer cores. The ESFR-SMART core includes extra fuel subassemblies at the outer core periphery in order to compensate the inner fissile height reduction. It was also aimed to use the same fuel enrichments in the inner and outer cores, if possible. The corium discharge tubes were included: at the central position, between inner and outer cores, and at the core periphery. The number of DSD locations, including those for passive safety devices, was increased. The axial part between the fissile region and lower gas plenum was proposed to be a combination of the fertile lower blanket and steel reflector below: to reduce the sodium void effect, but prevent breeding. The fissile and fertile heights and single fissile enrichment were finally chosen on the basis of fine optimization studies. A 6-batch fuel reloading scheme was proposed, instead of a 5-batch one in CP-ESFR. The core is surrounded by 2 rings of steel reflector and 1 ring of absorber subassemblies. Outside of absorber there are locations for spent fuel subassemblies, including 3 inner and 3 outer core batches. In the new ESFRSMART core, the calculated void effect is significantly reduced: to a value well below 1$ at the end of cycle. In the following, core specifications are given, which can be most easily used for deterministic neutronics codes such as ERANOS. These specifications contain dimensions and nuclear densities at the room temperature, temperatures related to operating conditions, and tables for thermal expansion. To facilitate model preparations and calculations with Monte-Carlo codes, an additional dataset for a simplified core description at operating conditions is also provided. The appendix contains the EDF report on details of the core design optimization with the SDDS multi-physics and multi-objective method.

AZNHEX is a neutron diffusion code for hexagonal-z geometry currently under development as part of the AZTLAN project in which a Mexican platform for nuclear core simulations is being developed. The diffusion solver is based on the RTN0 (Raviart-Thomas-Nédélec of index 0) nodal finite element method together with the Gordon-Hall transfinite interpolation which is used to convert, in the radial plane, each one of the four trapezoids in a hexagon to squares. The main objective of this work is to test the AZNHEX code capabilities against two well-known diffusion codes DYN3D and PARCS. In a previous work, the Serpent Monte Carlo code was used as a tool for preparation of homogenized group constants for the nodal diffusion analysis of a large U-Pu MOX fueled Sodium-cooled Fast Reactor (SFR) core specified in the OECD/WPRS neutronic SFR benchmark. The group constants generated by Serpent were employed by DYN3D and PARCS nodal diffusion codes in 3D full core calculations. A good agreement between the reference Monte Carlo and nodal diffusion results was reported demonstrating the feasibility of using Serpent as a group constant generator for the deterministic SFR analysis. In order to verify the under development solver inside AZNHEX, the same Serpent generated cross sections sets for each material were exported to AZNHEX format for four different states (as in DYN3D and PARCS): a) a reference case in which the multiplication factor (keff) is the compared value, b) the Doppler constant (KD), c) the sodium void worth, and d) the total control rod worth. Additionally, the radial power distribution was also calculated. The results calculated with AZNHEX showed also a quite good agreement in the direct comparison with DYN3D (-66 pcm in keff) and PARCS (-109 pcm in keff) and therefore against the Serpent reference solution (-194 pcm in keff). As AZNHEX is still under development further improvements will be implemented and new tests will be carried out, but so far the results presented here give confidence in the development.

To improve the public acceptance of the future nuclear power in Europe we have to demonstrate that the new reactors have significantly higher safety level compared to traditional reactors. The ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) aims at enhancing further the safety of Generation-IV SFRs and in particular of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the European Sustainable Nuclear Industrial Initiative (ESNII) roadmap and in close cooperation with the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) program. The project aims at 5 specific objectives: Produce new experimental data in order to support calibration and validation of the computational tools for each defence-in-depth level. Test and qualify new instrumentations in order to support their utilization in the reactor protection system. Perform further calibration and validation of the computational tools for each defence-in-depth level in order to support safety assessments of Generation-IV SFRs, using the data produced in the project as well as selected legacy data. Select, implement and assess new safety measures for the commercial-size ESFR, using the GIF methodologies, the FP7 CP-ESFR project legacy, the calibrated and validated codes and being in accordance with the update of the European and international safety frameworks taking into account the Fukushima accident. Strengthen and link together new networks, in particular, the network of the European sodium facilities and the network of the European students working on the SFR technology. By addressing the industry, policy makers and general public, the project is expected to make a meaningful impact on economics, environment, EU policy and society. Selected results and milestones achieved during the first eighteen months of the project will be briefly presented, including proposal of new safety measures for ESFR; evaluation of ESFR core performance; benchmarking of codes; experimental programs; and education and training.