This work presents a whole-core benchmark problem based on a 2-loop pressurized water reactor with both UO₂and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and core level. The geometry and material compositions are fully described and multi-group cross section libraries are provided in 2, 4, and 8 group formats. Simplifications made to the benchmark specification include a Cartesian boundary, to facilitate the use of transport codes that may have trouble with cylindrical boundaries, and control rod homogenization, to reduce the geometric complexity of the problem. These modifications were carefully chosen to preserve the physics of the problem and a justification of these modifications is given. Detailed Monte Carlo reference solutions including core eigenvalue, assembly averaged fission densities and selected fuel pin fission densities are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in the paper namely all-rods-out, all-rods-in, and some-rods-in.

The project was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. Measurement data from a set of mixing experiments, gained by using advanced measurement techniques with enhanced resolution in time and space help to improve the basic understanding of turbulent mixing and to provide data for Computational Fluid Dynamics (CFD) code validation. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: The Rossendorf coolant mixing model ROCOM and the VATTENFALL test facility, modelling a German Konvoi type and a Westinghouse type three-loop PWR, respectively. Additional data on slug mixing in a VVER-1000 type reactor gained at a 1:5 scaled metal mock-up at EDO Gidropress are provided. Experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility are made available. Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests (Sizewell-B for PWRs, Loviisa and Paks for VVERs) are used together with the data from the ROCOM facility as a basis for the flow distribution studies. The test matrix on flow distribution and steady state mixing performed at ROCOM comprises experiments with various combinations of running pumps and various mass flow rates in the working loops. Computational fluid dynamics calculations are accomplished for selected experiments with two different CFD codes (CFX-5, FLUENT). Best practice guidelines (BPG) are applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The BPG contain a set of systematic procedures for quantifying and reducing numerical errors. The knowledge of these numerical errors is a prerequisite for the proper judgement of model errors. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. Besides of the benchmark cases, additional experiments were calculated by new partners and observers, joining the project later. Based on the "best practice solutions", conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The high importance of proper grid generation was outlined. In general, second order discretization schemes should be used to minimise numerical diffusion. First order schemes can provide physically wrong results. With optimised "production meshes" reasonable results were obtained, but due to the complex geometry of the flow domains, no fully grid independent solutions were achieved. Therefore, with respect to turbulence models, no final conclusions can be given. However, first order turbulence models like K-e or SST K-w are suitable for momentum driven slug mixing. For buoyancy driven mixing (PTS scenarios), Reynolds stress models provide better results.

The Institute of Safety Research (ISR) is one of the six Research Institutes of Forschungszentrum Dresden-Rossendorf e.V. (FZD), which is a member institution of the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz (Leibnizgemeinschaft). Together with the Institutes of Radiochemistry and Radiation Physics, ISR implements the research programme „Nuclear Safety Research“, which is one of the three scientific programmes of FZD. The programme includes two main topics, i. e. “Safety Research for Radioactive Waste Disposal” and “Safety Research for Nuclear Reactors”.

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A new whole-core transport method is described for 3-D hexagonal geometry. This is an extension of a stochastic-deterministic hybrid method which has previously been shown highly accurate and efficient for eigenvalue problems. Via Monte Carlo, it determines the solution to the transport equation in sub-regions of reactor cores, such as individual fuel elements or sections thereof, and uses those solutions to compose a library of response expansion coefficients. The information acquired allows the deterministic solution procedure to arrive at the whole core solution for the eigenvalue and the explicit fuel pin fission density distribution more quickly than other transport methods. Because it solves the transport equation stochastically, complicated geometry may be modeled exactly and therefore heterogeneity even at the most detailed level does not challenge the method. In this dissertation, the method is evaluated using comparisons with full core Monte Carlo reference solutions of benchmark problems based on gas-cooled, graphite-moderated reactor core designs. Solutions are given for core eigenvalue problems, the calculation of fuel pin fission densities throughout the core, and the determination of incremental control rod worth. Using a single processor, results are found in minutes for small cores, and in no more than a few hours for a realistically large core. Typical eigenvalues calculated by the method differ from reference solutions by less than 0.1%, and pin fission density calculations have average accuracy of well within 1%, even for unrealistically challenging core configuration problems. This new method enables the accurate determination of core eigenvalues and flux shapes in hexagonal cores with efficiency far exceeding that of other transport methods.

A significant factor in the overall efficiency of the balance of plant for a future magnetic fusion energy (MFE) reactor is the thermal performance of the divertor. A significant fraction of the reactor power is delivered to the divertor as plasma impurities and fusion products are deposited on its surface. For an advanced MFE device, an average divertor heat load of 10 MW/m² is expected at steady-state operating conditions. Helium cooling of the divertors is one of the most effective ways to accommodate such a heat load. Several helium-cooled divertor designs have been proposed and/or studied during the past decade including the T-Tube divertor, the helium-cooled flat plate (HCFP) divertor, the helium-cooled multi-jet (HEMJ) divertor, the helium-cooled modular divertor with integral fin array (HEMP), and the helium-cooled modular divertor with slot array (HEMS). All of these designs rely on some form of heat transfer enhancement via impinging jets or cooling fins to help improve the heat removal capability of the divertor. For all of these designs very large heat transfer coefficients on the order of 50-60 kW/m²-K have been predicted. As the conditions of a fusion reactor and associated helium flow conditions (600 °C and 10 MPa) are difficult to achieve safely in a controlled laboratory environment, the study of these divertors often relies on computer simulations and experimental modeling at non-prototypical, albeit dynamically similar, conditions. Earlier studies were based on the assumption that, for geometrically similar divertor test modules, dynamic similarity can be achieved by matching only the Reynolds number. Experiments conducted in this investigation using different coolants and test module materials have shown this assumption to be false. Modified correlations for the Nusselt number and loss coefficients for the HEMJ and HEMP-like divertor modules have been developed. These have been used to develop generalized performance curves to predict the divertor performance, i.e. the maximum allowable heat flux and corresponding pumping power fraction, at prototypical conditions. Additionally, a numerical study has been performed to optimize the fin array geometry of the HEMP-like divertor module. The generalized correlations and performance curves developed in this investigation can be incorporated into system design codes, thereby allowing system designers to optimize the divertor geometry and operating conditions.