Conference Agenda

A new theory referred as Apparent Reynolds Number (ARN) has been developed to better explain the physics and mechanisms of flow laminarisation of isothermal turbulent flows caused by non-uniform body force, and then that of heated flows with strong influence of buoyancy, e.g., an upward pipe flow of air and supercritical CO₂. This concept has been extended to describe predict heat transfer deterioration in fully developed pipe air flow and now addresses mixed convection in developing air flows. In particular, the inertial terms in the momentum equations have been found to have a similar effect as the buoyancy in terms of strengthening or attenuating turbulence, leading to enhancing or deteriorating heat transfer. This understanding has prompted treating the inertia as a pseudo-body force. The ARN concept is then used to make predictions of heat transfer of developing air flow by linking turbulence mixing in complex flows such as this to that in a simple unheated shear flow based on a new equal-pressure-gradient reference framework. This has led to the development of ARN-based mixing length model. The full paper will demonstrate that this simple ARN-mixing length model can predict mixed convection heat transfer in a developing flow of air, validated against DNS data. This new physics-based modelling approach significantly simplifies the complexity of traditional turbulence models while reliably predicting complex heat transfer phenomena. It hence provides a route for modelling large energy systems with affordable computing resources. Additionally, the ARN theory enhances understanding of heat transfer behaviour in mixed convection flows.

The UK's Advanced Gas-cooled Reactor (AGR) fleet use graphite blocks for the core structure and as a moderator. The graphite undergoes dimensional change with irradiation and loses weight due to oxidation from the carbon dioxide coolant. These effects change the flow behaviour in the core and can challenge the structural integrity of the graphite bricks. Providing accurate understanding and prediction of the condition of the graphite is essential for ongoing extensions to the operational life of these reactors.

The rate of oxidation is reduced by the presence of low concentrations of other gases in the coolant. These gases need to be continuously provided to the interior of the bricks by transport of the coolant through the porous graphite material. This requires a pressure difference to be imposed across the bricks. The oxidation effects also depend on the temperature of the graphite.

To provide increased predictive insight into the flows, pressures and temperatures that influence these processes, CFD models have been built of AGR fuel channels, including all flow paths and porous flow predictions inside the bricks. The dimensions of the channel and properties of the graphite vary with the irradiation and weight loss of the bricks, which evolves and accumulates through operational life. The CFD models are able to integrate with this information coming from other parts of the analysis toolchain, and provide pressure and temperature predictions in return.

The High-Temperature Gas-Cooled Reactor (HTGR), a proposed Generation IV nuclear reactor, is gaining increasing attention for its inherent safety, high thermal efficiency, and ability to produce high-temperature process heat. The successful deployment of the HTGR technology depends on an in-depth understanding of reactor physics, particularly coolant flow, heat transfer within fuel assemblies, and their impact on reactor structural integrity. While Computational Fluid Dynamics (CFD) can provide detailed 3-D predictions of the thermal-hydraulic behaviour in the reactor core, the large computational resources required make it impractical for real-world nuclear engineering applications.

This work presents a coarse-grid CFD approach, initially developed for light water reactors, which has now been extended to prismatic HTGR fuel assemblies. This method, known as Subchannel CFD (SubChCFD), combines the strengths of traditional subchannel codes and modern CFD. It offers CFD-like 3-D predictions for a large range of scenarios, and meanwhile, the results produced are consistent with well-calibrated empirical correlations. By using a coarse mesh, SubChCFD reduces the computing costs by up to 1 to 3 orders of magnitude compared to conventional Reynolds Averaged Navier Stokes (RANS) CFD, depending on the complexity of the problem. This potentially makes the full reactor core simulations more feasible and cost-effective. To demonstrate the versatility of SubChCFD, the General Atomics modular HTGR fuel assembly is investigated. The results show that SubChCFD simulations of a full-length prismatic HTGR fuel assembly closely align with conventional RANS simulations for the same problem, but the computational cost is significantly lower than the latter.

The Single Primary Heat Extraction and Rejection Emulator (SPHERE) facility at Idaho National Laboratory was recently utilized to generate data for the startup and steady operation of a high-performance, sodium heat pipe over the course of 1000 hours, as a test of detrimental, long-term effects of heat pipe operation. The setup consists of a single, sodium heat pipe enclosed in a stainless-steel vacuum chamber, heated radiatively via a cylindrical ceramic fiber heater configuration and cooled via a water-cooled calorimeter. Measurements include temperatures at several axial locations along the outer surface of the heat pipe, the power provided to the heaters, and the heat removal rate of the calorimeter. In this work, this data is utilized to validate heat pipe models in the heat pipe application Sockeye, which is based upon the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Sockeye provides various heat pipe models at an engineering scale appropriate for the multiphysics simulation of microreactors, which may feature several hundred heat pipes. This work details models of this experiment in SPHERE using various heat pipe models with Sockeye, including heat conduction-based models and compressible flow models of the heat pipe interior. These models are compared to the experimental data to assess the accuracy of several aspects of heat pipe modeling, including frozen startup, the effect of non-condensable gases, and the coupling of the heat pipe to its environment.

Due to their compact structure and strong mobility, solid-fuel heat pipe reactors have gradually become a research focus for small reactors. Current research mainly concentrates on nuclear-thermal-structural multi-physics coupling, considering thermal expansion. However, during the long-term operation of heat pipe reactors, the reactivity feedback caused by fuel irradiation-induced swelling must be considered. Therefore, based on RMC and OpenFOAM, this paper develops an analysis process for nuclear-thermal-structural multi-physics coupling in solid-state heat pipe reactors, taking irradiation effects into account, and conducts a study on the KRUSTY heat pipe reactor. The results show that for KRUSTY, due to the low burnup, the negative feedback from irradiation effects is not as significant as that caused by thermal expansion. However, the overall calculation indicates that irradiation effects must be considered for reactors with high burnup.

System-level modeling of the 200 MW General Pebble Bed Reactor (GPBR200) is performed with SAM. Using SAM’s component-based system, a core channel approach is developed and used to model the core of the GPBR200. For comparison, a SAM porous-media multi-D model is also developed for the same reactor. Good comparisons are obtained for the two model during steady-state normal operating condition. Furthermore, transient simulations are performed for the de-pressurized and pressurized loss-of-forced cooling accidents (DLOFC and PLOFC). The core channel model compares well with the porous media model during DLOFC but overpredicts the overall temperature of the reactor during PLOFC. The good comparison during DLOFC indicates that the core channel model is able to capture radial conduction well. On the other hand, the overprediction of temperature by the core channel model during PLOFC suggests that the model underestimates the effects of in- core natural circulation during the transient.