• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.608-619
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• 2010

A multi-scale analysis of water-cooled reactor thermalhydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermalhydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.969-978
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• 2009

The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced $CANDU^{(R)1}$ reactor fuel bundle. Based primarily on the $CANFLEX^{(R)2}$ fuel bundle, several geometry changes (such as element sizes and pitch-circle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• v.38 no.8
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• pp.733-752
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• 2006

Severe accidents in PWRs are characterized by a continuously changing geometry of the core due to chemical reactions, melting and mechanical failure of the rods and other structures. These local variations of the porosity and other parameters lead to multi-dimensionnal flows and heat transfers. In this paper, a comprehensive set of multi-dimensionnal models describing heat transfers, thermal-hydraulics and melt relocation in a reactor vessel is presented. Those models are suitable for the core description during a severe accident transient. A series of applications at the reactor scale shows the benefits of using such models.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.365-376
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• 2010

This paper summarizes the results of a Writing Group on the Extension of CFD codes to two-phase flow safety problems, which was created by the Group for Analysis and Management of Accidents of the Nuclear Energy Agency' Committee on the Safety of Nuclear Installations (NEA-CSNI). Two-phase CFD used for safety investigations may predict small scale flow processes, which are not seen by system thermalhydraulic codes. However, the two-phase CFD models are not as mature as those in the single phase CFD and potential users need some guidance for proper application. In this paper, a classification of various modelling approaches is proposed. Then, a general multi-step methodology for using two-phase-CFD is explained, including a preliminary identification of flow processes, a model selection, and a verification and validation process. A list of 26 nuclear reactor safety issues that could benefit from investigations at the CFD scale is identified. Then, a few issues are analyzed in more detail, and a preliminary state-of-the-art is proposed and the remaining gaps in the existing approaches are identified. Finally, guidelines for users are proposed.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.103-112
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• 2017

Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.