Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR)

Advances in computer abilities, intense competition on the energy market and stringent regulatory/nrequirements during the last decade have spurred the development of robust numerical/nmodels to support nuclear reactor safety analysis and design optimization. This dissertation/naims to develop methodologies for numerical modeling of Pebble-Bed Fluoride Salt Cooled Reactors/n(PB-FHR), a novel reactor design that combines TRISOparticle fuel and flibe salt coolant./nThe use of a large number of fuel pebbles in PB-FHR cores poses great challenge in computational/ncost and violates the assumptions in most of the traditional deterministic codes developed/nfor Light Water Reactors (LWRs). This project developed dedicated models with different/nlevels of spatial and energy resolution for the broad need in FHR design and analysis, including/ncoupled heat diffusion and point kinetics unit cell models,Monte Carlo neutronicsmodels, and/ncoupled multi-group neutron diffusion and multi-scale porous media model. Documentation,/nversion control, testing, verification are all indispensable parts of numerical modeling software/ndevelopment. These steps are followed meticulously to ensure high quality open source codes/nthat promote open science and reproducible research./nUnit cell models can compute representative fuel and coolant temperatures and full core power/nwithin a short amount of time and thus is adequate for scoping analysis, or uncertainty and/nsensitivity analysis, where a large number of runs is required to cover the input space. FHRs/nhave substantial graphite reflectors that slows down the neutron generation time by hosting/nneutrons while they get thermalized. A ’multi-point’ correction is derived from perturbation/ntheory to take the reflector effects on reactivity and neutron generation time into account./nMonte Carlo based codes, on the other extreme, can provide accurate results with minimal assumptions,/nbut they are typically only used for generating cross-sections for deterministic models,/nfor example point kinetics models andmulti-group neutron diffusion models in this project,/nor to provide reference results for benchmarking lower resolution codes due to high computational/ncost. PB-FHR has not yet test reactors to provide experimental data for tool validation./nHigh fidelity models based on direct coupling between neutron transport and Computational/nFluid-Dynamics (CFD) was used in this project as reference for code-to-code verification./nMulti-group neutron diffusion models were developed for design optimization and safety analysis,/nbecause they are compatible with current computation resources in nuclear industry, with/nwhich simulations can be carried out on stand-alone workstations or small computation clusters/nwithin a reasonable time. For areas where the diffusion assumption is limited, e.g., in vicinity/nof control rods, the simplified spherical harmonics equations were implemented to improve/nthe accuracy of diffusion equation by relaxing the isotropic assumption in neutron transport./nThe neutronics model is coupled to a porous media CFD module with multi-scale treatment/nthat computes conductive heat transfer inside TRISO particles and fuel pebbles, as well as convective/nheat transfer between fuel pebbles and coolant. Radiative heat transfer may be significant/nin the high temperature reactors, but is not modeled in this project due to lack of material/npropertiesmeasurement./nResults from the coupled full core modelwere verified with analytical solutionswhen they exist, for example the steady state outlet bulk average temperature, computed from energy balance,/nor a referenceMonte Carlo model. Although the absolute value of the multiplication factor can/nnot be accurately matched between the Monte Carlo statistics and the eigenvalue found from/nthe coupled model, this can be corrected and more important for transient modeling is the/nchange in the multiplication factor during a transient. In fact, the important parameters for a/ntransient study, such as flibe and fuel temperature feedback coefficients, time scale, and control/nrod worth matches well between the coupledmodel and the referencemodel./nThe multiphysics models are capable of simulating both steady state and a broad spectrum/nof transient scenarios, involving either coolant inlet condition or reactivity induced transients,/nespecially Anticipated Transient Without Scram (ATWS). The methodology was applied to the/nTMSR SF-1 and Mk1 design, which are both FHRs that uses TRISO fuel particles in spherical/nfuel elements and Fluoride salt coolant. For both designs, steady state power, neutron flux, and/ntemperature distributions were computed, and reactivity insertion and overcooling transients/nwere simulated. The study shows that FHR cores are extremely resilient to the investigated/ntransient scenarios, for the following reasons:/n• The graphite based fuel elements in FHRs can withstand up to 1600 ±C without risk of/nradioactive release. And the flibe coolant used in FHRs also has high thermal resistance,/nwith a boiling point at 1430 ±C. Due to the large thermal margin to the failure of TRISO/nparticles, the thermal constraintswill be more likely in the metallic structures. These temperature/nlimits should be defined based on both the temperature and the time of exposure/nfor creep limits or reduced yield stress limits at elevated temperatures./n• In FHRs, the role of coolant and moderator are separated. The Doppler feedback, the/nmainmechanism to stabilize the reactor during an ATWS, is more negative than in LWRs./n• Online refueling fuel management regime allows the reactor to operate at a low excess of/nreactivity because the reserve for compensating fuel burnup is small in this case.

Advisor: Peterson, Per F