Development and validation of a system thermal-hydraulic/CFD codes coupling methodology for multi-scale transient simulations of pool-type reactors
Author: Toti, A.
Subject: Development and validation of a system thermal-hydraulic/CFD codes coupling methodology for multi-scale transient simulations of pool-type reactors
Promotor: Vierendeels, J.
SCK CEN Mentor: Belloni, F.
Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is currently developing and designing a flexible irradiation facility, configured as an Accelerator Driven System (ADS) able to operate in critical and sub-critical modes. In addition to material testing and fuel research, the objectives of the reactor are to prove the feasibility of the ADS technology for the transmutation of long-lived nuclear waste as well as to represent a demonstration plant for Generation IV heavy liquid metal-cooled reactors. The current system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). The innovative primary system configuration brings along the need of new modeling and simulation capabilities to support the design and safety analyses. The pool-type primary cooling system, typical of liquid metal-cooled fast reactor designs, implies the presence of complex coolant flow fields and three-dimensional effects which might have an impact on the integral system behavior during accidental transients such as loss of flow events, dissymmetric conditions among others, and therefore on safety-relevant parameters. In particular, local flow mixing and a three-dimensional temperature profile distribution can affect the evolution of the coolant mass flow during the transition from forced to natural convection, and the development of thermal stratification may worsen the effectiveness of the passive emergency cooling systems. The aforementioned phenomena are of difficult prediction for industry standard System Thermal-Hydraulic (STH) codes, reference tools for nuclear power plants safety assessments. These codes, based on one dimensional lumped parameters formulation, were originally developed for loop-type systems analysis, and are not validated to correctly simulate the physics of the phenomena occurring in a pool-type reactor. Modern Computational Fluid Dynamic (CFD) codes, on the other side, are of particular interest for more realistic representation of complex fluid flow and heat transfer phenomena, therefore their use in nuclear reactor thermal-hydraulic and safety analyses is constantly increasing. However, the use of CFD codes for integral system analyses is computational intensive and often not practical for industrial applications. The PhD research project presented in this dissertation is focused on the development, verification and preliminary validation of novel code infrastructure based on the integrated use of STH and CFD codes, specifically conceived for high-fidelity transient simulations of advanced nuclear reactors. The proposed multi-scale methodology couples the 1D system thermal-hydraulic code RELAP5-3D to the CFD code FLUENT, and is based on the domain decomposition technique and dynamic exchange of boundary conditions at coupling interfaces. An extensive investigation of coupling numerical algorithms was performed. As expected, the use of explicit schemes led to numerical stability issues, especially in the computation of fast transients in incompressible fluid systems, due to imbalance of pressure-velocity fields between the domains. The implementation of implicit schemes led to numerical stability and a significant improvement of the results. To accelerate convergence rates and reduce computational costs, dynamic relaxation algorithms have been investigated; among them, the implementation of a Quasi-Newton coupling algorithm has shown significant improvements of the performance of the tool. A novel numerical technique for the application of this coupling algorithm to multi-domain coupled problems has been developed and successfully applied on a number of cases. A further extension of the multi-scale modeling capabilities has been achieved through the implementation of thermal coupling interfaces for the computation of conjugate heat transfer phenomena. A first validation of the numerical method against experimental data has been carried out on the basis of the experimental campaign at the TALL-3D facility, a LBE loop in operation at the Royal Institute of Technology (KTH) in Sweden. The design of the experimental facility was specifically conceived to induce mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in a pool-type test section, in order to support the verification and validation of multi-scale codes coupling approaches. The analysis discussed in this dissertation focused on a loss of flow experimental test, characterized by an oscillating transition from forced to natural circulation. Compared to full system code models, the multi-scale approach showed higher accuracy in the prediction of the dynamic behavior of the system, in terms of characteristic frequency and amplitude of mass flow and temperature oscillations, and of local parameters affected by 3D flows. The discrepancies observed were mostly attributed to deficiencies in the single codes, confirming that the use of well validated STH and CFD codes is fundamental pre-requisite for accurate coupled analyses. In view of the application of the tool on the safety analysis and licensing of future nuclear reactors, the validation on a more representative configuration which allows for the investigation of relevant phenomena and transient conditions is needed. To this purpose, the pool-type mock-up E-SCAPE (European SCAled Pool Experiment), a scale model of the MYRRHA primary cooling system, has been designed and built at SCK•CEN and is currently in the phase of commissioning tests. A coupled model for the facility has been developed and applied to a number of pretest loss of flow transient simulations, and assessed against full 1D models. The results of the simulations confirmed the validity of the implemented numerical method, and highlighted potential inaccuracy of 1D models when particular flow conditions in natural circulation establish, confirming that 3D modeling might be required for certain fast reactor transient simulations. The experimental data-set to be generated will be used first for the validation of the stand-alone codes in relation to the regions and phenomena relevant to them, and integral test data will be successively used for the validation of the coupling tool. The final part of the research was centered on the application of the tool to the analysis of the MYRRHA reactor. The purpose of this activity was twofold: from one side, it aimed at the assessment of the tool performance on a first reactor-scale application, and on the other side, it allowed investigating relevant thermal-hydraulic transient phenomena in the reactor and identifying limitations and potential model improvements of STH codes. A reference protected loss of flow transient was analyzed and the results, similarly to the work on E-SCAPE, were found in good overall agreement. This suggests that no significant impact of 3D effects on the integral behavior in loss of flow conditions is expected in the current MYRRHA design, although some discrepancy on certain parameters was observed due to 3D effects.