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One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method. In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.

Part I of this study introduced a novel methodology for the optimization of energy group structures to be used in diffusion calculations of Sodium cooled Fast Reactors. The goal of the optimization is to speed up calculations, particularly in transient analyses, while maintaining an acceptable accuracy of the results. The proposed methodology is based by either direct or simulated annealing search techniques. The capabilities of the method were preliminarily demonstrated on a single core state of the Superphenix reactor by using the DYN3D nodal diffusion code. The scope of Part II is the further demonstration of the methodology efficiency through its application to more challenging “real-life” cases. In this respect, a static Superphenix neutronic benchmark comprising 13 different core states and a transient test initiated by an increase of the core inlet temperature at the Phenix reactor are considered. For both Superphenix and Phenix reactor cores, 2- to 12-group optimal condensed energy group structures are identified using the 24-group structure as a starting point. The obtained optimal structures are thus employed in DYN3D and compared to the reference 24-group DYN3D solutions. The outcomes show that also for a broader range of core configurations and under transient conditions the optimal energy group structures allow for significant acceleration of the DYN3D performance with practically negligible degradation of the solution accuracy.

Low-frequency neutronic noise with magnitudes of up to ten percent of the reactor power have been observed in Konvoi-type PWR in Germany in the last years. Several attempts were made to identify the reasons for increased magnitudes of the neutronic fluctuations in comparison with pre-Konvoi reactors, and various hypotheses have been created to explain this effect. In this paper, results of noise simulations performed with use of the reactor dynamics code DYN3D are presented. Fluctuations of the coolant inlet temperature and mass flow rate were considered. Besides of un-correlated fluctuations, correlated temperature fluctuations were simulated. In a preliminary analysis with harmonic temperature oscillations, the importance of existing correlations between temperature fluctuations in the individual fuel assemblies was pointed out. In our analyses, the correlations between fluctuations in the individual fuel assemblies were obtained based on an experimentally validated coolant mixing model. However, the features of the neutronic noise found in the simulations do not correspond to the measurements. Obviously, more complex mechanisms than only temperature and/or mass flow fluctuations have to be considered. Simulated generic fluctuations of the local moderator density, that were introduced independently from thermal hydraulics, indicate that potentially deformations or vibrations of the fuel rod lattice or turbulent fluctuations of the coolant temperature and density inside the fuel rod bundles leading to variations of the local moderator content might be responsible for the observed neutronic noise. Therefore, advanced models coupling neutronics, thermal hydraulics, turbulence and mechanical modelling have to be developed.

In the framework of the European project NURESAFE, the reactor dynamics code DYN3D, developed at Helmholtz-Zentrum Dresden-Rossendorf (HZDR), was coupled with the Computational Fluid Dynamics (CFD) solver Trio_U, developed at CEA France, in order to replace DYN3D’s one-dimensional hydraulic part with a full three-dimensional description of the coolant flow in the reactor core at higher spatial resolution. The present document gives an introduction into the coupling method and shows results of its application to the simulation of a Main Steamline Break (MSLB) accident of a Pressurised Water Reactor (PWR).

This work deals with the \Full-Core" VVER-1000 calculation benchmark which was proposed on the 26th Symposium of AER [1]. Recently, the calculation benchmarks \Full-Core" VVER-440 [2] and its extension [3] have been introduced in the AER community with positive response [4, 5]. Therefore we have decided to prepare a similar benchmark for VVER- 1000. This benchmark is also a 2D calculation benchmark based on the VVER-1000 reactor core cold state geometry, explicitly taking into account the geometry of the radial reflector. The loading pattern for this core is very similar to the fresh fuel loading of cycle 9 at Unit 1 of the Temelin NPP (Czech Republic). This core is filled with six types of fuel assemblies with enrichment from 1.3%w 235U to 4.0%w 235U with up to 9 fuel pins with Gd burnable absorber per FA. The main task of this benchmark is to test the pin-by-pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron- physics calculations especially for VVER reactors. In this contribution we present the overview of available macro-codes results.

The focus of this work is the validation of Serpent-SUBCHANFLOW-TRANSURANUS, a high-fidelity multiphysics system coupling Monte Carlo neutron transport, subchannel thermalhydraulics and fuel-performance analysis. A full-core pin-by-pin burnup calculation for the first operating cycle of a Pre-Konvoi PWR plant is presented and the results are assessed using experimental data. The critical boron concentration and a set of pin-level neutron flux profiles are compared against measurements, with very good agreement. The impact of using fuel-performance analysis is discussed comparing the results obtained with the three-code coupling with the ones using the traditional neutronic-thermalhydraulic approach. The studies presented here are part of the final stage of the EU Horizon 2020 McSAFE project, closing the development cycle of the Serpent-SUBCHANFLOW-TRANSURANUS system, from implementation to validation.

The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research. The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantified using a cross-section library that is generated from the output of corresponding CASMO5 calculations. The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes.