Conference Agenda

The Water-Cooled Lithium-Lead Test Blanket Module (WCLL-TBM) is an essential component in ITER that will provide crucial information for the development of the DEMO driver blanket. Our research aims to build a multi-scale system code for the thermal-hydraulic analysis of the WCLL-TBM. The OpenModelica software is used to develop a robust and modular object-oriented library for the components of the WCLL-TBM and the Water Cooling System (WCS) in this work. The various objects contain modelled thermal flow loops with 0D/1D interconnected components such as pipes, heat ports, orifices and valves. The objects describe the different multi-scales and can be nested and combined to form new objects. Such objects include the Double-Walled Tubes (DWTs), First Wall (FW), Breeding Units (BU), Breeding Module (BM), and larger outer circuits - all of which are designed to have replaceable modules with different levels of fidelity. The code aims at fast and reliable thermal-hydraulic predictions of the WCLL-TBM components and WCS during nominal operating conditions (gauge pressure 15.5 [MPa], inlet temperature 295 [textdegree C], outlet temperature 328 [textdegree C]), as well as the transient response of the system to off normal scenarios by varying certain parameters and loading conditions.

This research utilizes the system-level modeling language Modelica and its open-source libraries, Transform and Hybrid, to develop a natural circulation model for the primary loop of a Low-Temperature Nuclear Heating Reactor(NHR200-II). This model includes components such as the reactor core, coolant channels, heat exchangers , control system model and so on. Based on these models, the steady-state behavior of the primary loop under 100% nominal reactor power conditions was simulated. Also, transient simulation analyses were performed for step and ramp changes at 90% nominal power. The simulation results, when compared with RELAP5 data, demonstrated excellent agreement, confirming the validity and accuracy of using Modelica for simulation modeling. Furthermore, the primary control system model established in this study can regulate the core outlet temperature by controlling the reactivity of the core, and the results show that the reactivity control scheme is feasible. The research work in this paper lays a foundation for using Modelica language to carry out nuclear energy system simulation application.

Different types of medical isotopes are needed for kinds of procedures where many options for production are possible. One particular isotope, Astatine-211 or At-211, can be produced using cyclotron based irradiation where a Bismuth target is converted to At-211. During irradiation, significant heat is generated within the target and appropriate cooling is needed to prevent target melting and increase isotope yield. In support of higher At-211, University of Washington (UW), Oak Ridge National Laboratory (ORNL) and Virginia Commonwealth University (VCU) are collaboratively developing better At-211 target and target holder designs to enable higher isotope production yields. In this study, VCU is focused on the thermal-hydraulics modeling of different target designs including material and geometric parameters to enable UW and ORNL collaborators to hit desired yields. Initially VCU is focused on 0-D modeling using lumped parameter analysis approaches to enable a design space to be developed using multi-objective optimization. This enables the ability to explore both differential holder materials (e.g. aluminum or stainless steel varieties) and coolant channel geometries rapidly to reduce the total number of high-fidelity CFD simulations and experiments. The 0-D model was created in Python to include two energy balance ordinary differential equations (ODEs) to predict the Bismuth and sample holder temperatures during irradiation. The heat generation within the Bismuth target and the coolant conditions were acquired from the UW team and implemented in the model. Using the 0-D model,, we’ve identified several potential improvements to both the sample holder design and coolant channels for follow-on CFD and experimental studies.

Fluid-structure interactions play a critical role in numerous engineering applications, such as jet flows around fuel rods in nuclear reactors. Under specific flow conditions, these interactions can give rise to vortex-induced vibrations (VIV), a phenomenon where large-amplitude oscillations occur due to vortex shedding. VIV poses a significant threat to system stability and can lead to operational failure. Therefore, understanding and controlling VIV is essential to mitigate its detrimental effects.

This study explores the passive control of VIV in a circular cylinder that oscillates freely, using a non-linear energy sink (NES). The NES is designed as a secondary system incorporating linear damping and a key non-linear cubic stiffness component. Simulations are conducted using the Reynolds-averaged Navier–Stokes (RANS) turbulence model with strongly coupled fluid-structure interaction model, utilizing the dynamic response of both the cylinder and the NES, as well as the surrounding fluid flow. By systematically adjusting the sink parameters—mass, spring, and damping—this research investigates their influence on the behavior of the coupled system. The system's response is analyzed at reduced velocity within the lock-in range, where the cylinder's motion synchronizes with vortex shedding. The model is validated against existing data from literature which indicate the optimum values for these parameters to achieve the best performance. Key results are presented in terms of vibration amplitude, drag and lift coefficients, Strouhal number analysis, and vortex visualization, providing insight into the effectiveness of the NES in controlling VIV.

Solid-solid contact thermal conduction is a basic heat transfer problem in thermal power engineering, which is significant for thermal design and safe operation of system equipment, such as high temperature thermal protection of aircrafts, efficient thermal management of space orbits, and superior heat transfer chain of nuclear engineering. Objective to the solid-solid contact thermal conductivity of typical structure based on tubes and holes in heat-pipe nuclear reactor systems, theoretical models of thermal conductivity, mechanics and thermodynamic coupling at the microscopic contact interface were established in the paper, obtaining the interface contact thermal conductivity characteristics under the filling of helium in micro gaps, as well as the influence on the contact thermal resistance for interface temperature and external loads. According to the CFD simulation results under different interface temperature and external loads, the Levenberg-Marquardt algorithm was used to fit a high temperature contact thermal resistance correlation on the cylindrical interface. By selecting appropriate fitting parameters, the R-squared corresponding to the fitting results was greater than 0.95, indicating that the calculation model had a good predictive ability for the contact thermal resistance. It was applicable for the rapid evaluation of contact thermal resistances for the solid-solid interface in engineering design and heat transfer analysis of heat pipe reactors.

The fuel irradiation device serves as a critical platform for conducting nuclear fuel irradiation experiments in research reactors. Its structural design, thermal characteristics, and the arrangement of measurement points at the outlet significantly influence experimental results, thereby affecting the thermal power determination of the device and the evaluation of fuel performance. This study focuses on the HFETR irradiation device, employing a CFD-based three-dimensional high-resolution modeling method to investigate the impacts of outlet sensor placement and structural thermo-mechanical properties on power measurement accuracy. Computational results demonstrate that positioning temperature sensors 130 mm upstream of the physical outlet plane effectively represents the outlet temperature field. From a thermal-hydraulic perspective, an annular gap thickness of 1 mm achieves a coolant flow partitioning of 22% through the bypass channel, with parasitic heat losses limited to 4.7% of the total generated power. This configuration ensures adequate cooling of the samples while avoiding excessive heat leakage.