An improved CFD model for a MYRRHA based primary coolant loop

Vincent Moreau, Manuela Profir, Steven Keijers, Katrien Van Tichelen

Research outputpeer-review


Within the SESAME framework, the Computational Fluid-Dynamics (CFD) modeling of Heavy Liquid Metal (HLM) flows in pool configuration is (mainly) aimed at giving a technical support to the design of innovative GEN-IV nuclear reactors, such as ALFRED, for electricity production, or MYRRHA, as a multi-purpose research installation. This paper presents the CFD model of the primary coolant loop of MYRRHA version 1.4, built and operated with STAR-CCM+ versions 11–13. CFD is the only tool able to capture the 3D flow effects in the plena of a pool reactor. It is however still not feasible to simulate the detailed geometry of an entire reactor core by CFD. Fortunately, for many applications (safety studies, design) the exact geometrical representation of the core, heat exchangers, pumps and bypasses can be replaced by simplified models without loss of relevant information. These parts must be modeled such that they are functionally consistent with their role. Inspiration is taken from the system code techniques like homogenization procedures. A combination of solid and porous volumes is used within the CFD model. As with system code coupling, the homogenization can lead to loss of relevant information. In this case, the flow in the plena suffers from inadequate boundary conditions provided by the over-simplified modeled parts. To reach a sufficient level of maturity, the simplification procedure must be applied only where it is strictly necessary and at the right geometrical scale. First, an overview of the general features of the design and of the most important parameters that must be preserved by the CFD model is given. Then, we concentrate on the numerical model. Its topological structure is illustrated, based on a two-level geometrical and functional decomposition. Next, we focus on the core representation. Each sub-assembly (fuel assembly, dummy assembly, control rod and scram rod) has its own position. The core horizontal cross section is vertically propagated to represent the Above Core Structure (ACS). To obtain the right flow mixing the porosity of the various interfaces is calibrated. Due to the mixture of large and small length scales detailed meshing of the core and the ACS would result in very high mesh numbers. A specific procedure has been employed, to compensate cross sectional variations by physical property changes, maintaining the most important global characteristic. The ACS parameters have then been tuned in order to reduce the hot plenum surface temperature. Bypass flows have been implemented. Their role is to connect the fluid volumes and to avoid stagnation regions, which could cause potential corrosion issues. With the bypass flows and the simplified core/ACS structure models, a versatile CFD model for nominal operation is available. The Volume of Fluids (VoF) method used in the simulation allows the simulation of scenarios with variation of the free surface levels. Transient scenarios can be implemented to simulate operational incidents or accidents. To evaluate the safety of the reactor design, combinations of extreme parameter values are used in the simulations to demonstrate all safety criteria are met under worst case conditions. The simulations can also be used to define technical requirements of the system components (pumps and heat exchangers). Some examples are illustrated and partial conclusions are drawn.
Original languageEnglish
Article number110221
Pages (from-to)1-13
Number of pages13
JournalNuclear Engineering and Design
StatePublished - 22 Jul 2019

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