Minor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.

No abstract

Within the NURESAFE project, a main steam line break benchmark has been defined and solved by codes integrated into the European code platform NURESIM. The paper describes the results of the calculations for this benchmark. Six different solutions using different codes and code systems are provided for the comparison. The quantitative differences in the results are dominated by the differences in the secondary system parameters during the depressurization. The source of these differences comes mainly from the application of different models for the two-phase leak flow available in the system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created. The codes integrated into the NURESIM platform showed their applicability to a challenging transient like a main steam line break.

No abstract

This paper summarizes the nodal level results from the VVER MSLB core simulation in the NURESAFE EU project. The main objective is to implement and verify new developments in the models and couplings of 3D core simulators for cores with hexagonal fuel assemblies. Recent versions of the COBAYA and DYN3D core physics codes, and the FLICA4 and CTF thermal-hydraulic codes were tested standalone and coupled through standardized coupling functions in the Salome platform. The MSLB core transient was analyzed in coupled code simulation of a core boundary condition problem derived from the OECD VVER MSLB benchmark. The impact of node subdivision and different core mixing models, as well as the effects of CFD computed core inlet thermal-hydraulic boundary conditions on the core dynamics were explored.

At UPM, in depth modifications of the core simulator COBAYA, able to perform neutronics diffusion cal culations at both nodal and pin by pin levels, have been accomplished during the 7th Framework EURATOM NURESAFE project.The main goal was to upgrade its integration in the European Platform for Nuclea r Reactor Safety Simulation in order to facilitate the coupling with any other code of the platform for multi physics analysis, focusing also on the code legibility and maintainability. An external and flexible coupling with the thermal hydraulics code COBRA TF was designed. As a result, COBAYM/COBRA TF allows multiscale coupled calculations, enabling both nodal and pin by pin neutronics resolutions using both assembly based channels and pin based subchannels at the thermal hydraulics domain.flexible mapping schemes also in axial direction can be defined. The coupled system was applied to a Main Steam Line Break transient benchmark.Pin by pin 30 sim ulations using one thermal hydraulic channel per assembly were carried out in a reasonable computing time, and results compared to nodal solutions demonstrating the multiscale coupling capability for full core transients. Pin by pin calculations using thermal hydraulics subchannels will be performed in a near future to assess the role that a very detailed mapping can play to predict realistic local parameters. While in asymmetric transients the effect can be important, it is expected that in symmetric transients assembly based thermal hydraulics channels can provide accurate pin by pin solutions in execution times suitable for routine analysis. The performed work will bring the ability to explore in an easy way multiscale effects on safety tran sient evaluations and give recommenda tions for the neutronics/thermal hydraulics mapping depending on the application

Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transport approximation is solved, appropriate correction factors, such as interface discontinuity factors, are required to nearly reproduce the fully heterogeneous transport solution.Calculating exact pin-by-pin discontinuity factors requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration; however, inaccurate correction factors are one major source of error in core analysis when using multi-group diffusion theory.An alternative to generate accurate pin-by-pin interface discontinuity factors is to build a functionalfitting that allows incorporating the environment dependence in the computed values. This paper suggests a methodology to consider the neighborhood effect based on the Analytic Coarse-Mesh Finite Difference method for the multi-group diffusion equation. It has been applied to both definitions of interface discontinuity factors, the one based on the Generalized Equivalence Theory and the one based on Black-Box Homogenization, and for different few energy groups structures.Conclusions are drawn over the optimal functional-fitting and demonstrative results are obtained with the multi-group pin-by-pin diffusion code COBAYA3 for representative PWR configurations.

No abstract