Analysis of the iaea-benchmark on flow mixingin a 4x4 rod bundle

International audienceThe IAEA is organizing a Coordinated Research Project (CRP) on the application of Computational Fluid Dynamics (CFD) for nuclear reactor design. Within this CRP, an open benchmark is organized on fluid mixing in rod bundles. KAERI has offered precise experimental data of the velocity field and turbulent statistics downstream of a mixed twist/split type mixing grid. The experimental set up consists of a 2000 mm long bundle of 4x4 rods with pitch to diameter ratio of 1.35, placed in a square housing of 142x142 mm. Each rod has a diameter of 25.4 mm. The mixing grid is located at 2/3 height of the channel with 5 spacer grids located upstream and 3 downstream of the mixing grid. The isothermal experiment is analyzed with the TrioCFD code. Turbulence is treated first with the RANS approach by using a non-linear eddy viscosity model. Then LES was applied. Difficulties were found to correctly prediction inlet boundary conditions upstream of the mixing grid. It is shown that the flow is not fully developed when reaching at the mixing grid. The comparison of measured and calculated mean velocity profiles in the central sub-channel is shown. The best accordance is achieved for LES on a fine mesh. Measured profiles of velocity mean values were calculated correctly for the central sub-channel close and far from the mixing grid. The RANS approach on coarse mesh did not reproduce correctly fine flow featured close to the mixing grid

Large Eddy Simulation of the Injection of Cold ECC Water into the Cold Leg of a Pressurized Water Reactor

International audienceThe aging of the reactor pressure vessel (RPV) can be a limiting factor in life time extension of pressurized water reactors. Important thermal stresses can be generated in the RPV in the case of loss of coolant accidents. Cold emergency core cooling (ECC) water will be injected under high pressure conditions into the cold leg (horizontal pipe) and can come in contact with the hot RPV wall. This process is called pressurized thermal shock (PTS). The efficiency of the PTS is dependent on the mixing between cold ECC water and the hot coolant inventory in the cold leg and the upper downcomer (annular space). Several PTS scenarios have been analyzed experimentally in the TOPFLOW-PTS facility, situated at Helmholtz Zentrum Dresden Rossendorf (HZDR), Germany. The experimental objective was reproducing whole scenarios of PTS events under realistic small break loss of coolant accident thermal hydraulic conditions. A single phase TOPFLOW experiment is analyzed numerically by two different numerical approaches in order to get access to detailed information on the temporal development of both thermal stratification and flow behavior in the cold leg and the upper downcomer Large Eddy Simulations with two approaches to account for thermal effects incompressible fluid hypothesis with Boussinesq approximation and dilatable fluid hypothesis. The CEA in-house CFD code TrioCFD is used. Calculated temperature profiles in the cold leg are compared to detailed temperature measurements for a 400 s ECC injection transient. The dilatable approach reproduces well the experimental results

Entwicklung und Verifizierung eines mikrometeorologischen Modells zur Bestimmung der Spurenstoffkonzentration in Taufluessigkeit

Copy held by FIZ Karlsruhe; available from UB/TIB Hannover / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekSIGLEDEGerman

CFD analysis of non-axial flow in fuel assemblies

International audienceCFD studies with the Trio_U code were performed to evaluate the flow field and pressure drop in rod bundles with a slope angle of 22° between main flow and the axis of the rods. Comparison to the CEA experimental program EOLE is discussed which was performed with inclined rod bundles in in a square channel. Finally, the inter-assembly flow in two adjusted parts of fuel assemblies, separated by the assembly bypass flow, is analyzed. The large eddy simulation (LES) method is used to account for nonisotropic turbulence. A mesh of 1.1 billion control volumes was used for this analysis. The simulation has been performed on 10,000 processor cores of the HPC computer CURIE of the TGCC. The correlation of the EOLE program under estimates the pressure drop of the assembly calculation by a factor of about six. In contrary to the flow in the open, unbounded core region modeled by the assembly calculation, the EOLE experiments have been performed in a square channel. Thus, questions rise concerning the applicability of correlations derived from a channel configuration to the unbounded core region. Simulating the flow in the inclined rod bundle of the EOLE experiment by using the open boundary conditions of the assembly calculation confirms the under prediction of the correlation

Analysis of the IAEA benchmark on heat transfer in a 4 4 rod bundlewith mixing grid

International audienceThe IAEA is organizing a Coordinated Research Project (CRP) on the application of Computational Fluid Dynamics (CFD) for nuclear reactor design. Within this CRP, an open benchmark is organized on fluid mixing in rod bundles. KAERI has offered precise experimental data of the velocity field downstream of a mixed twist-split type mixing grid. Additionally, temperature measurements in various axial and azimuthal locations of a partly heated rod were made available by KAERI. The experimental set up consists of a 2000 mm long bundle of 4x4 rods with pitch to diameter ratio of 1.35, placed in a square housing of 142x142 mm. Each rod has a diameter of 25.4 mm. The mixing grid is located at 2-3 height of the channel with 5 spacer grids located upstream and 3 downstream of the mixing grid. These experiments are analysed with the TrioCFD code. Turbulence is treated with the RANS approach by using a non-linear eddy viscosity model. The comparison of measured and calculated mean velocity profiles is presented. Measured profiles of velocity mean values were calculated correctly for the central sub-channel and the rod to rod gap both close and far from the mixing grid. The measured temperature is underestimated by the calculations, most probably due to the localizations of the thermocouples inside of the wall of the rod. A correction based on the estimation of the temperature difference between surface and wall leads to satisfying comparison between measurement and calculation

CFD Analysis of a Steam Generator Separation Test in the Kozloduy VVER-1000 Reactor

International audienceComputational fluid dynamics (CFD) research for nuclear reactor safety dedicates to real scale reactor circuits under realistic thermal hydraulic conditions. In the framework of an OECD/NEA benchmark, CEA has attempted 10 years ago with the code TrioCFD to study the temperature distribution at the core inlet in a main steam line break (MSLB) accident scenario in a Bulgarian VVER1000 reactor. This work is resumed here by completing the geometry of the reactor pressure vessel (RPV) and by capitalizing both code development and high performance computing (HPC) resources. Before modelling the full scale RPV thermal-hydraulics, a PIRT (Phenomena Identification and Ranking Table) was performed to classify the existing physical phenomena in a ranking table. Three single effect validation test cases were defined in a test matrix. The CFD approach was validated single effect by single effect by reproducing the defined well suited test cases. The core outlet temperature distribution was measured during a commissioning steam generator separation test at Kozloduy nuclear power plant. This temperature distribution is compared to the CFD calculations and helps to validate integrally the full scale reactor calculation. Tetrahedral meshes of 50 to 400 million velocity control volumes were generated for the complete RPV; self-evidently the mesh refinement reflects the restrictions of the former defined test matrix. In the OECD benchmark, the core inlet temperature was calculated from the measured core outlet temperature by simple energy conservation. With the integral calculation we were able to review this process with the calculated core inlet and outlet temperature

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).11This work has been partly financed by IRSN. The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result. © 2007 Elsevier B.V. All rights reserved

LES analysis of jets in cross flow

International audienceJets in cross flow are of fundamental industrial importance and play an important role in the validation of turbulence models. Two jet configurations are investigated with the TrioCFD code, the CFD reference code of the Nuclear Energy Division of CEA:• A tee junction of circular tubes where a hot jet discharges into a cold main flow,• A rectangular jet marked by a scalar discharging into a main flow in a rectangular channel.The $tee-junction$ configuration is very important for the phenomena of thermal fatigue. The OECD/NEA benchmark on the Vattenfall tee junction flow is analyzed. This test case is selected because, beside the experimental results, various calculation results with several turbulence modelling approaches have been published. A LES modelling and calculation strategy is developed and validated on this data for jets in crossflow under thermal fatigue conditions.The rectangular jet configuration is important for basic physical understanding and modelling and has been analyzed experimentally at CEA. The experimental work was focused on the turbulent mixing between a rectangular channel flow with grid turbulence exiting a heated jet into a confined grid turbulent crossflow, with both kinematic and passive scalar high quality measurements in order to characterize its statistical properties (energy spectra, Reynolds stresses (anisotropy), PdF). These experiments are analyzed for the first time with LES by applying the strategy developed for the Vattenfall configuration. The turbulent inlet boundary conditions are well controlled in both experiment and calculation(grid turbulence). Structured and unstructured tetrahedral grids are used to predict the measured mean values and turbulent fluctuations of the velocity and the temperature

Numerical analysis of two experiments related to thermal fatigue

International audienceJets in cross flow are of fundamental industrial importance and play an important role in validating turbulence models. Two jet configurations related to thermal fatigue phenomena are investigated:•T-junction of circular tubes where a heated jet discharges into a cold main flow and•Rectangular jet marked by a scalar discharging into a main flow in a rectangular channel.The T-junction configuration is a classical test case for thermal fatigue phenomena. The Vattenfall T-junction experiment was already subject of an OECD/NEA benchmark. A LES modelling and calculation strategy is developed and validated on this data.The rectangular-jet configuration is important for basic physical understanding and modelling and has been analyzed experimentally at CEA. The experimental work was focused on turbulent mixing between a slightly heated rectangular jet which is injected perpendicularly into the cold main flow of a rectangular channel. These experiments are analyzed for the first time with LES.The overall results show a good agreement between the experimental data and the CFD calculation. Mean values of velocity and temperature are well captured by both RANS calculation and LES. The range of critical frequencies and their amplitudes, however, are only captured by LES

temporal variation of the thermal stratification in the cold leg during emergency core cooling injection

International audienceThe aging of the reactor pressure vessel (RPV) can be a limiting factor in life time extension of pressurized water reactors. Important thermal stresses can be generated in the RPV in the case of loss of coolant accidents. Cold emergency core cooling (ECC) water will be injected under high pressure conditions into the cold leg (horizontal pipe) and can come in contact with the hot RPV wall. This process is called pressurized thermal shock (PTS). The efficiency of the PTS is dependent on the mixing between cold ECC water and the hot coolant inventory in the cold leg and the upper downcomer (annular space). Several PTS scenarios have been analyzed experimentally in the TOPFLOW-PTS facility, situated at Helmholtz Zentrum Dresden Rossendorf (HZDR), Germany. The experimental objective was reproducing whole scenarios of PTS events under realistic small break loss of coolant accident thermal hydraulic conditions. A single phase TOPFLOW experiment is analyzed numerically by two different numerical approaches in order to get access to detailed information on the temporal development of both thermal stratification and flow behavior in the cold leg and the upper downcomer: Large Eddy Simulations with two approaches to account for thermal effects: incompressible fluid hypothesis with Boussinesq approximation and dilatable fluid hypothesis. The CEA in-house CFD code TrioCFD is used. Calculated temperature profiles in the cold leg are compared to detailed temperature measurements for a 400 s ECC injection transient. The dilatable approach reproduces well the experimental results