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Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association "Atomic Energy Research" (AER), the second exercise has been organised under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the differences in the thermohydraulics were assumed in the difficult modelling of the vertical once-through steam generator with steam superheating. Sensitivity analyses which considered the influence of the nodalisation and the impact of the coolant mixing model were performed for the DYN3D-ATHLET solution of the OECD benchmark. The solution of the benchmarks essentially contributed to the qualification of the code complex DYN3D-ATHLET as an advanced tool for the accident analysis for both VVER type reactors and Western PWRs.

Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association "Atomic Energy Research" (AER), the second exercise has been organised under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the differences in the thermohydraulics were assumed in the difficult modelling of the vertical once-through steam generator with steam superheating. Sensitivity analyses which considered the influence of the nodalisation and the impact of the coolant mixing model were performed for the DYN3D-ATHLET solution of the OECD benchmark. The solution of the benchmarks essentially contributed to the qualification of the code complex DYN3D-ATHLET as an advanced tool for the accident analysis for both VVER type reactors and Western PWRs.

The test facility ROCOM (Rossendorf Coolant Mixing Model) has been built for the investigation of coolant mixing processes in the reactor pressure vessel of pressurised water reactors (PWR). ROCOM is a 1:5 model of the German PWR KONVOI and has been designed for a wide range of different mixing scenarios. ROCOM disposes of four loops with fully controllable coolant pumps. The test facility is operated with demineralised water. For the investigation of mixing, tracer solution (water labelled with salt) is injected into the facility. The transient distribution of the electrical conductivity is is measured at different positions of the flow path by means of wire-mesh sensor technique with high resolution in space and time. The measured conductivity is transformed into a dimensionless mixing scalar. The mixing at quasi-stationary conditions (constant loop mass flow rates) has been investigated in the presented experiments. That concerned nominal operation conditions, the operation with a reduced number of loops and the investigation of cold-water transients with running pumps and conditions of developed natural circulation. In special experimental series, the reproducibility of the results at identicla boundary conditions within the confidence intervalls has been shown. Further, the influence of various factors on the mixing has been investigated. This included the pressure losses at the core bottom plate, the global coolant flow level and the influence of the loop flow rate on the perturbed sector at the core inlet. An analysis of the measurement error of the used measurement technique completes the report.

The test facility ROCOM (Rossendorf Coolant Mixing Model) has been built for the investigation of coolant mixing processes in the reactor pressure vessel of pressurised water reactors (PWR). ROCOM is a 1:5 model of the German PWR KONVOI and has been designed for a wide range of different mixing scenarios. ROCOM disposes of four loops with fully controllable coolant pumps. The test facility is operated with demineralised water. For the investigation of mixing, tracer solution (water labelled with salt) is injected into the facility. The transient distribution of the electrical conductivity is is measured at different positions of the flow path by means of wire-mesh sensor technique with high resolution in space and time. The measured conductivity is transformed into a dimensionless mixing scalar. The mixing at quasi-stationary conditions (constant loop mass flow rates) has been investigated in the presented experiments. That concerned nominal operation conditions, the operation with a reduced number of loops and the investigation of cold-water transients with running pumps and conditions of developed natural circulation. In special experimental series, the reproducibility of the results at identicla boundary conditions within the confidence intervalls has been shown. Further, the influence of various factors on the mixing has been investigated. This included the pressure losses at the core bottom plate, the global coolant flow level and the influence of the loop flow rate on the perturbed sector at the core inlet. An analysis of the measurement error of the used measurement technique completes the report.