Ramiro Freile

This report continues the body of work led by the National Reactor Innovation Center at Idaho National Laboratory following the development of the Road Map for the Development of Commercial Maritime Applications of Advanced Nuclear Technology. Previous efforts examined candidate nuclear-maritime configurations, deployment pathways, feasibility considerations, and the readiness of existing U.S. testing capabilities. The present work addresses the technical research and development needs associated with nuclear systems in maritime environments, with emphasis on the effects of ocean-induced motion. An initial assessment of motion-sensitive physical phenomena is performed to identify the mechanisms through which vessel translation, rotation, inclination, and acceleration may influence reactor-system behavior. A common non-inertial reference-frame formulation is developed in which the reactor geometry remains fixed while transformed gravity and apparent accelerations simulate the effects of ship motion. The formulation is implemented using complementary modeling approaches. RELAP5-3D provides a one-dimensional system-level representation of integrated reactor and balance-of-plant behavior, while the open-source MOOSE capability provides a multidimensional finite-volume representation of local flow and temperature fields. The maritime-motion models are assessed using analytical verification problems, benchmark calculations, and comparisons with available experimental data. The developed capabilities are subsequently applied to selected reactor systems, including light water reactor models and a Generation IV liquid-fueled molten salt reactor. The analyses examine the influence of ocean-induced motion on natural circulation and reactor response during postulated operating and accident conditions. The results are used to identify motion-sensitive parameters, evaluate the applicability of existing modeling approaches, and determine where additional experimental evidence is needed. The report concludes with a discussion of modeling limitations, validation gaps, and data needs for maritime nuclear applications. Simulation and experiment are treated as an iterative process: numerical models guide the selection of test conditions and measurements, while experimental observations support model validation, refinement, and uncertainty reduction. This simulation-to-experiment framework provides a technical basis for future component-level and system-level experimental campaigns supporting the development and deployment of commercial maritime nuclear technologies.

The Lattice Boltzmann Method (LBM) applicability to solidification problems related to nuclear coolants under laminar internal forced flow conditions is demonstrated in this paper. The double distribution Total Enthalpy-Partially Saturated Method is implemented in the open-source code OpenLB and com- pared against Computational Fluid Dynamics (CFD) results, which uses the enthalpy-porosity MultiEulerian method, in finite volume commercial code Star- CCM+. The test cases comprise the solidification of two fluids with significantly different Prandtl numbers, molten salt and lead, in a 3D pipe geometry in lam- inar flow conditions with an initial Reynolds number of 100. The solid interface advances from the cooled wall towards the bulk of the pipe reducing the area of flow passage as the transient evolves. Results of LBM velocity fields, solid frac- tion profiles, temperature fields, and pressure fields are in good agreement with CFD results for both coolants. Furthermore, LBM stability and accuracy limits are independently explored for molten salts and lead using Two-Relaxation Time (TRT) and Bhatnagar-Gross-Krook (BGK) collision operators. Simulation time between CFD and LBM is also contrasted for both coolants. In this work, LBM solvers use explicit time schemes with a regular grid, whereas CFD solutions are able to include implicit time schemes and near-wall mesh refinement. For molten salt flows (high Prandtl number) CFD allows for considerably larger timesteps than LBM, which is limited by the maximum lattice Mach number limitation and stability considerations. For lead flows (low Prandtl number), the implicit scheme in CFD is limited in terms of accuracy for large time steps due to the limiting Fourier-number. On the other hand, LBM does not present notable stability issues for lead and the lattice Mach number limit is not as relevant as for the molten salt case for this Reynolds number, allowing for faster simulation times than CFD.

This report details progress and activities of Idaho National Laboratory (INL) on the NuclearRegulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.”The tasks completed for this report are: Task 1d: The net radiation transfer method is implemented into MOOSE for modeling reactor cavity cooling system geometries. RCCS models for two experiments: (1) Natural Convection Shutdown Heat Removal Test Facility (NSTF) experiment R022 and (2) HTTR VCS mockup, are created. For validation, computed temperature distributions are compared to measured temperatures.•Task 4c: An algorithm for computing the pebble bed reactor equilibrium core isotopic com-position is developed and an initial version is implemented into the reactor multiphysics code Griffin. Initial results for a simplified axisymmetric pebble bed reactor are presented.•Task 7: generation of a reference plant model for molten salt cooled pebble bed reactor.