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The system thermal-hydraulic (STH) code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. GRS carried out pre-test simulations with ATHLET–ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET–ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared.

One of the most interesting technologies recognized as promising for the future of nuclear energy is the Accelerator Driven System (ADS). The purpose of the ADS is to function as an element of an integrated nuclear power enterprise comprising of conventional and advanced power reactors for energy production, and for reducing the radiotoxicity of the nuclear waste produced by these power reactors before entombment in a geologic repository ( Stanculescu, 2000 ). The use of new types of coolants such as molten lead or lead–bismuth eutectic alloy (LBE) represents another particular feature in the design concept of these reactors, since it permits one to take advantage of higher Heavy Liquid Metals (HLM) boiling temperature compared to water, leading to an improvement in thermal efficiency. Furthermore, it allows the reactor to be operated at a lower pressure, thus reducing the probability of a Loss-Of-Coolant Accident (LOCA) and, consequently, increasing reactor safety. The proposed use of relatively new types of coolants, especially in combination with new fuel types and cladding materials, demands specific attention to the thermal–hydraulics and core mechanics in normal and abnormal conditions. For that reason, a detailed analysis using system codes is necessary. In this paper, a new version of the ATHLET system code is tested, in which the physical properties of liquid metals like sodium, lead and LBE are implemented. The object of this study is the spallation loop of the MYRRHA facility, in which LBE is circulated by forced convection to remove the heat deposited by a proton beam. A detailed nodalization is set up for performing thermal–hydraulic calculations for both nominal conditions and accident scenarios in order to have a good characterization of the entire loop. The start-up transient has verified the correct removal of the heat generated in the target by the foreseen heat exchanger. During accidental transients, it was noted that the level of LBE in the beam line changes in agreement with preliminary studies on the target device. The pump failure test represents the most dangerous scenario, as the temperature in the target area will reach a very high value within a few seconds after the blockage of the pump impeller. The results obtained have subsequently been compared to the ones achieved with previous numerical simulations which were performed using a version of RELAP5/mod 3.3 system code that was modified at the University of Pisa to account for LBE properties. The comparison has confirmed the capability of the new version of the ATHLET code for the analysis of hydraulic circuits cooled by liquid metals, although both codes use different heat transfer correlations.

A new data-driven coupling method between the system code ATHLET and ANSYS CFX was developed. The new approach allows for coupled simulations of single and two-phase flow with heat transfer and transport of tracer concentration. Any number of coupling interfaces can be defined with the aid of a graphical user interface. Two coupling schemes have been implemented: an explicit scheme where the coupling variables are exchanged at the end of each CFD time step and a semi-implicit scheme in which the variable exchange is performed for each iteration within each time step. The coupling method was validated with measured data from a single-phase double T-junction mixing experiment. The results of the coupled simulation for the double T-junction case, where 3-D effects are very important, were found in excellent agreement with the experimental data. The semi-implicit scheme was found numerically more accurate and stable than the explicit scheme.Copyright © 2014 by ASME

The thermal-hydraulic (TH) system code ATHLET is developed by the Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH in the last decades for the analysis of the whole spectrum of design basis and beyond design basis accidents for PWRs and BWRs. Recently, ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve the system code simulation capabilities for flows with pronounced 3D phenomena such as flow mixing, thermal stratification, etc. The aim of the current development of the coupled code ATHLET – ANSYS CFX is focused on the extension of the physical models of the computer program for the application to innovative reactor concepts. Within the European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are carried out. The Swedish experimental facility TALL, operated by KTH, Stockholm is currently being refurbished for thermal-hydraulic experiments with lead-bismuth eutectic alloy (LBE) coolant at natural and forced circulation conditions. Simulations with ATHLET – ANSYS CFX will be carried out and the results will be compared with experimental data. The paper gives an overview on models’ extensions and the implementation of different physical properties of liquid metals in ATHLET and ANSYS CFX. First coupled 1D-3D calculations for simplified configurations are presented. Then, preliminary results of the supporting 3D simulations for the design process of the 3D TALL test section are discussed.Copyright © 2012 by ASME

The paper is focused on the detailed thermo-hydraulic (TH) modelling of the Reactor Pressure Vessel (RPV) only with the help of a TH system code including the full scope of Nuclear Power Plant (NPP) modelling (primary and secondary sides, RPV, active core and all operational and safety systems). The idea of the pseudo-3D TH methodology of the GRS system code ATHLET [1] is based on the application of a set of parallel TH (PTH) channels with cross flow connections between them. The aim of the activities described in the paper is to validate the recent developments of the methodology based on a fine TH nodalization of ATHLET on different experiments and thus to meet the challenge of detailed modelling of neutron-kinetics (NK) in the active core and the TH processes in the whole RPV with a good accuracy applying only a system TH code. This methodology enables to boost the performance of coupled code safety analysis for production purposes and it also allows the efficient application of uncertainty and sensitivity (U&S) analyses. Primary concern of the validation is related to the methodology itself that includes the correct modelling of the mixing flow phenomena mainly in the coolant legs, downcomer, bottom and upper plenum of RPV, and active core. Another aim of the study is to find the limits of application and accuracy of the system code ATHLET by the simulation and prediction of local (core) safety related parameters with the enhanced nodalization and PTH channel application.The paper describes and discusses the comparisons done with the ATHLET system code using the new enhanced nodalization scheme with 3 types of measured data: data from two experiments (C1 and C3) on fluid mixing at Upper Plenum Test Facility (UPTF) [2] and a set of detailed neutron-physical/thermal-hydraulic measurements performed at Kalinin-3 NPP during an asymmetric transient.The performed studies are of great interest because they have shown acceptable accuracy and also some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the PTH channel methodology within a 1-D TH system code.Copyright © 2012 by ASME