This dissertation provides a comprehensive exploration of the feasibility and potential benefits of remote operations as a control strategy for advanced nuclear reactors. While remote monitoring and control have been implemented in other industries, the operational feasibility and guidelines for implementing remote-operated control rooms for nuclear facilities remain largely unproven and scarce. Therefore, this research offers valuable insights for the industry as it navigates the future of nuclear power plant operations.

The dissertation is organized into six chapters, with each chapter building on the previous one. It begins with a review of state-of-the-art technology for modern control rooms in various industries, including nuclear. The subsequent chapters discuss remote operations in other industries and ways the nuclear industry can implement remote operations. The research then provides a demonstration of remote operations at an advanced nuclear reactor company, highlighting the technical feasibility of remote operations.

Moreover, this dissertation offers two more original contributions. Firstly, it presents a human factors-based study that sheds light on the challenges of remote operations and emphasizes the significance of appropriate training, communication, and workload management to ensure successful remote operations. Secondly, it provides a comprehensive set of criteria that advanced reactor companies need to consider before implementing a remote control room. These criteria encompass safety, cybersecurity, human factors, training, and licensing, serving as a guideline for advanced reactor companies planning to adopt a remote control operational strategy.

Overall, this dissertation provides valuable contributions to the field of nuclear energy. By identifying the key criteria that must be fulfilled for successful implementation of remote operations, this research offers practical insights that can help industry professionals make informed decisions about the future of nuclear power plant operations.

The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data.

This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study.

A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the transient event and maintains the core and reactor components at acceptable temperatures within allowable safety margins. It is also demonstrated that the transition from steady full-power, forced-cooling mode to steady decay-heat, natural-circulation mode is stable, predictable and well characterized.

Radiative heat transfer (radHT) is the reason we exist. For life, we need energy, and for that energy, we turn to the sun. Thermal radiation, another term for radHT, simply provides the means to transfer the sun's energy to us. On Earth, we interact with radHT as well --- we bask in the heat of a fire from across the room, we see metals glow red hot, we feel the evening warmth from buildings heated by the day as we walk past, and we see sunbeams through water turn blue and eventually peter out. All of these examples, both earthly and stellar, are driven by the transfer of energy via photons.

The amount of thermal radiation emitted by an object scales rapidly (to the fourth power) with temperature --- the reason why the fire across the room feels hotter than the building we pass by, loosely speaking. The proportion of radiation a body interacts with or absorbs varies as well --- the reason why sunlight travels mostly unabated through the atmosphere but attenuates noticeably in water. These two factors, temperature dependence and proportional interaction of various media, are important to consider when analyzing radHT in any system.

Of particular interest to this dissertation, is the consideration of these two factors within heat transfer analysis for nuclear reactors. The system temperatures and participating media interaction of conventional light water reactors are too low to render thermal radiation a significant heat transfer mechanism, less extraordinary circumstances. However, advanced reactors utilize significantly higher temperatures and non-water coolants that have the potential for increased radiative interaction. The fluoride-salt-cooled high-temperature reactor (FHR) is one such advanced reactor concept.

This dissertation addresses the impacts of radHT in FHRs through two means, introduced and contextualized by Chapter 1. On one hand, FHR development requires scaled-down experiments, in which low temperatures and surrogate fluids render radHT insignificant. The proportional impact of thermal radiation will be distorted compared to the prototypical reactor. Chapter 2 presents a scaling methodology to scale the system-level thermal-hydraulic behavior of FHR systems, with particular emphasis on quantifying radHT distortions. On the other hand, thermal radiation will serve as an important heat transfer mechanism in some full-scale FHR scenarios. Thus, radHT modeling must be included in FHR safety analyses to determine if thermal radiation plays a significant heat transfer role, and requires further consideration, or is low enough to be neglected. Chapter 3 details the development process of radHT simulation capabilities for System Analysis Module (SAM), a system-level thermal hydraulics code being developed for advanced reactor modeling. Chapter 4 then ties together the work conducted in Chapters 2 and 3 by laying out a proposed demonstration of the FHR scaling methodology and utilizing system-level modeling for radHT distortion quantification. The FHR scaling methodology and SAM radHT simulation tools presented in this dissertation can be used to address the impact of radHT in FHR systems. It is my hope this work will be utilized to facilitate FHR development and help realize the dream of building FHRs for clean energy production.

The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB).

The capability to validate integral transient response models is a key issue for licensing new reactor designs. This dissertation presents the scaling strategy, design and fabrication aspects, and startup testing results from the Compact Integral Effects Test (CIET) facility at UCB, which reproduces the thermal hydraulic response of an FHR under forced and natural circulation operation. CIET provides validation data to confirm the performance of the direct reactor auxiliary cooling system (DRACS) in an FHR, used for natural-circulation-driven decay heat removal, under a set of reference licensing basis events, as predicted by best-estimate codes such as RELAP5-3D. CIET uses a simulant fluid, Dowtherm A oil, which at relatively low temperatures (50-120°C) matches the Prandtl, Reynolds, Froude and Grashof numbers of the major liquid salts simultaneously, at approximately 50% geometric scale and heater power under 2% of prototypical conditions. The studies reported here include isothermal pressure drop tests performed during startup testing of CIET, with extensive pressure data collection to determine friction losses in the system, as well as subsequent heated tests, from parasitic heat loss tests to more complex feedback control tests and natural circulation experiments. For initial code validation, coupled steady-state single-phase natural circulation loops and simple forced cooling transients were conducted in CIET. For various heat input levels and temperature boundary conditions, fluid mass flow rates and temperatures were compared between RELAP5- 3D results, analytical solutions when available, and experimental data. This study shows that RELAP5-3D provides excellent predictions of steady-state natural circulation and simple transient forced cooling in CIET. The code predicts natural circulation mass flow rates within 8%, and steady-state and transient fluid temperatures, under both natural and forced circulation, within 2°C of experimental data, suggesting that RELAP5-3D is a good EM to use to design and license FHRs.

A key element in design and licensing of new reactor technology lies in the analysis of the plant response to a variety of potential transients. When applicable, this involves understanding of passive safety system behavior. This dissertation develops a framework to assess reliability and propose design optimization and risk mitigation strategies associated with passive decay heat removal systems, applied to the Mk1 PB-FHR DRACS. This investigation builds upon previous detailed design work for Mk1 components and the use of RELAP5-3D models validated for FHR natural circulation phenomenology. For risk assessment, reliability of the point design of the passive safety system for the Mk1 PB-FHR, which depends on the ability of various structures to fulfill their safety functions, is studied. Whereas traditional probabilistic risk assessment (PRA) methods are based on event and fault trees for components of the system that perform in a binary way – operating or not operating –, this study is mostly based on probability distributions of heat load compared to the capacity of the system to remove heat, as recommended by the reliability methods for passive safety functions (RMPS) that are used here. To reduce computational time, the use of response surfaces to describe the system in a simplified manner, in the context of RMPS, is also demonstrated. The design optimization and risk mitigation part proposes a framework to study the elements of the design of the reactor, and more specifically its passive safety cooling system, which can contribute to enhanced reliability of heat removal under accident conditions. Risk mitigation measures based on design, startup testing, in-service inspection and online monitoring are proposed to narrow probability distributions of key parameters of the system and increase reliability and safety.

Another major aspect in the development of novel energy systems is the assessment of their impacts on the environment compared to current technologies. While most existing life cycle assessment (LCA) studies have been applied to conventional nuclear power plants, this dissertation proposes a framework to extend such studies to advanced reactor designs, using the example of the Mk1 PB-FHR. The Mk1 uses a nuclear air-Brayton combined cycle designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and 242 MWe using natural gas co-firing for peaking power. The Mk1 design provides a basis for quantities and costs of major classes of materials involved in building the reactor and fabricating fuel, and operation parameters. Existing data and economic input-output LCA models are used to calculate greenhouse gas emissions per kWh of electricity produced over the life cycle of the reactor. Baseline life cycle emissions from the Mk1 PB-FHR in base-load configuration are 26% lower than average Generation II light water reactors in the U.S., 98% lower than average U.S. coal plants and 96% lower than average U.S. natural gas combined cycle plants using the same turbine technology. In peaking configuration, due to its nuclear component and higher thermal efficiency, the Mk1 plant only produces 32% of the emissions of average U.S. gas turbine simple cycle peaking plants. One key contribution to life cycle emissions results from the amount and type of concrete used for reactor construction. This is an incentive to develop innovative construction methods using optimized steel-concrete composite wall modules and new concrete mixes to reduce life cycle emissions from the Mk1 and other advanced reactor designs.

The combination of an increased demand for electricity for economic development in parallel with the widespread push for adoption of renewable energy sources and the trend toward liberalized markets has placed a tremendous amount of stress on generators, system operators, and consumers. Non-guaranteed cost recovery, intermittent capacity, and highly volatile market prices are all part of new electricity grids.

In order to try and remediate some of these effects, this dissertation proposes and studies the design and performance, both physical and economic, of a novel power conversion system, the Nuclear Air-Brayton Combined Cycle (NACC). The NACC is a power conversion system that takes a conventional industrial frame type gas turbine, modifies it to accept external nuclear heat at 670°C, while also maintaining its ability to co-fire with natural gas to increase temperature and power output at a very quick ramp rate. The NACC addresses the above issues by allowing the generator to gain extra revenue through the provision of ancillary services in addition to energy payments, the grid operator to have a highly flexible source of capacity to back up intermittent renewable energy sources, and the consumer to possibly see less volatile electricity prices and a reduced probability of black/brown outs.

This dissertation is split into six sections that delve into specific design and economic issues related to the NACC. The first section describes the basic design and modifications necessary to create a functional externally heated gas turbine, sets a baseline design based upon the GE 7FB, and estimates its physical performance under nominal conditions.

The second section explores the off-nominal performance of the NACC and characterizes its startup and shutdown sequences, along with some of its safety measures. The third section deals with the power ramp rate estimation of the NACC, a key performance parameter in a renewable-heavy grid that needs flexible capacity. The fourth section lays out the cost structure of the Mk1 Pebble-Bed Fluoride-salt-cooled High-temperature Reactor (FHR) with the NACC, since the NACC cannot be treated separately from its heat source.

The fifth section evaluates the cost structure of a twelve-unit Mk1 FHR and NACC, including capital construction costs, operating costs, fuel and decommissioning costs in bottom up methodology. The sixth section proposes alternative NACC configurations and scales (mobile, remote NACC) or alternative power cycles to the NACC that can be coupled to the FHR (supercritical carbon dioxide Brayton cycle).

Advances in computer abilities, intense competition on the energy market and stringent regulatory

requirements during the last decade have spurred the development of robust numerical

models to support nuclear reactor safety analysis and design optimization. This dissertation

aims to develop methodologies for numerical modeling of Pebble-Bed Fluoride Salt Cooled Reactors

(PB-FHR), a novel reactor design that combines TRISOparticle fuel and flibe salt coolant.

The use of a large number of fuel pebbles in PB-FHR cores poses great challenge in computational

cost and violates the assumptions in most of the traditional deterministic codes developed

for Light Water Reactors (LWRs). This project developed dedicated models with different

levels of spatial and energy resolution for the broad need in FHR design and analysis, including

coupled heat diffusion and point kinetics unit cell models,Monte Carlo neutronicsmodels, and

coupled multi-group neutron diffusion and multi-scale porous media model. Documentation,

version control, testing, verification are all indispensable parts of numerical modeling software

development. These steps are followed meticulously to ensure high quality open source codes

that promote open science and reproducible research.

Unit cell models can compute representative fuel and coolant temperatures and full core power

within a short amount of time and thus is adequate for scoping analysis, or uncertainty and

sensitivity analysis, where a large number of runs is required to cover the input space. FHRs

have substantial graphite reflectors that slows down the neutron generation time by hosting

neutrons while they get thermalized. A ’multi-point’ correction is derived from perturbation

theory to take the reflector effects on reactivity and neutron generation time into account.

Monte Carlo based codes, on the other extreme, can provide accurate results with minimal assumptions,

but they are typically only used for generating cross-sections for deterministic models,

for example point kinetics models andmulti-group neutron diffusion models in this project,

or to provide reference results for benchmarking lower resolution codes due to high computational

cost. PB-FHR has not yet test reactors to provide experimental data for tool validation.

High fidelity models based on direct coupling between neutron transport and Computational

Fluid-Dynamics (CFD) was used in this project as reference for code-to-code verification.

Multi-group neutron diffusion models were developed for design optimization and safety analysis,

because they are compatible with current computation resources in nuclear industry, with

which simulations can be carried out on stand-alone workstations or small computation clusters

within a reasonable time. For areas where the diffusion assumption is limited, e.g., in vicinity

of control rods, the simplified spherical harmonics equations were implemented to improve

the accuracy of diffusion equation by relaxing the isotropic assumption in neutron transport.

The neutronics model is coupled to a porous media CFD module with multi-scale treatment

that computes conductive heat transfer inside TRISO particles and fuel pebbles, as well as convective

heat transfer between fuel pebbles and coolant. Radiative heat transfer may be significant

in the high temperature reactors, but is not modeled in this project due to lack of material

propertiesmeasurement.

Results 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,

or a referenceMonte Carlo model. Although the absolute value of the multiplication factor can

not be accurately matched between the Monte Carlo statistics and the eigenvalue found from

the coupled model, this can be corrected and more important for transient modeling is the

change in the multiplication factor during a transient. In fact, the important parameters for a

transient study, such as flibe and fuel temperature feedback coefficients, time scale, and control

rod worth matches well between the coupledmodel and the referencemodel.

The multiphysics models are capable of simulating both steady state and a broad spectrum

of transient scenarios, involving either coolant inlet condition or reactivity induced transients,

especially Anticipated Transient Without Scram (ATWS). The methodology was applied to the

TMSR SF-1 and Mk1 design, which are both FHRs that uses TRISO fuel particles in spherical

fuel elements and Fluoride salt coolant. For both designs, steady state power, neutron flux, and

temperature distributions were computed, and reactivity insertion and overcooling transients

were simulated. The study shows that FHR cores are extremely resilient to the investigated

transient scenarios, for the following reasons:

• The graphite based fuel elements in FHRs can withstand up to 1600 ±C without risk of

radioactive release. And the flibe coolant used in FHRs also has high thermal resistance,

with a boiling point at 1430 ±C. Due to the large thermal margin to the failure of TRISO

particles, the thermal constraintswill be more likely in the metallic structures. These temperature

limits should be defined based on both the temperature and the time of exposure

for creep limits or reduced yield stress limits at elevated temperatures.

• In FHRs, the role of coolant and moderator are separated. The Doppler feedback, the

mainmechanism to stabilize the reactor during an ATWS, is more negative than in LWRs.

• Online refueling fuel management regime allows the reactor to operate at a low excess of

reactivity because the reserve for compensating fuel burnup is small in this case.

The radioactive waste classification system currently used in the United States primarily relies on a source-based framework. This has lead to numerous issues, such as wastes that are not categorized by their intrinsic risk, or wastes that do not fall under a category within the framework and therefore are without a legal imperative for responsible management. Furthermore, in the possible case that advanced fuel cycles were to be deployed in the United States, the shortcomings of the source-based classification system would be exacerbated: advanced fuel cycles implement processes such as the separation of used nuclear fuel, which introduce new waste streams of varying characteristics. To be able to manage and dispose of these potential new wastes properly, development of a classification system that would assign appropriate level of management to each type of waste based on its physical properties is imperative. This dissertation explores how characteristics from wastes generated from potential future nuclear fuel cycles could be coupled with a characteristics-based classification framework. A static mass flow model developed under the Department of Energy's Fuel Cycle Research & Development program, called the Fuel-cycle Integration and Tradeoffs (FIT) model, was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices: two modified open fuel cycle cases (recycle in MOX reactor) and two different continuous-recycle fast reactor recycle cases (oxide and metal fuel fast reactors). This analysis focuses on the impact of waste heat load on waste classification practices, although future work could involve coupling waste heat load with metrics of radiotoxicity and longevity. The value of separation of heat-generating fission products and actinides in different fuel cycles and how it could inform long- and short-term disposal management is discussed. It is shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system, and that it is useful to classify waste streams based on how favorable the impact of interim storage is on increasing repository capacity. The need for a more diverse set of waste classes is discussed, and it is shown that the characteristics-based IAEA classification guidelines could accommodate wastes created from advanced fuel cycles more comprehensively than the U.S. classification framework.

With the growing demand for clean, carbon free electricity around the globe due to industrialization and increasing populations, nuclear reactors will become a necessity to supplement power from intermittent renewable sources. Advanced reactor designs such as the Fluoride Salt-cooled High Temperature Reactor, or FHR, are especially desirable due to their small, modular design, their passive safety systems, and their smaller capital costs. The reactor’s small size, high temperatures, and single-phase molten salt coolant meant that conventional heat exchanger designs could not be used as the primary form of heat removal. This lead to the development of the Coiled Tube Gas Heater, or CTGH, so that the reactor could be coupled with an air Brayton reheat cycle. The CTGH is a shell-and-tube heat exchanger that uses an annular tube geometry to reduce the overall volume of the heat exchanger. In the FHR design, the air flows up the center of the annular bundle and flows out radially through the coiled tubes. The molten salt coolant is distributed vertically to multiple tubes within multiple sub-bundles in the CTGH. Starting at the outer radius of the bundle, the salt tubes coil around the bundle multiple times before reaching the inner manifolds and flowing out the bottom of the heat exchanger. This creates a heat exchanger with high effectiveness due to a large heat transfer surface area density and with a design that is essentially a counterflow heat exchanger. By using seamless tubes that are in compression rather than tension, the design minimizes the points of stress concentration and the risk of tubes bursting. Due to its high effectiveness, relatively low pressure drops, compact design, structural integrity, and resistance to damage from thermal shocks, the CTGH is an optimal choice for the primary heat exchanger in the FHR design.

In order to design the CTGH specifically for the FHR, it was necessary to simulate the conditions in the FHR. Conventional modeling codes would take too long to model the complex geometry of the CTGH to be an effective design tool, so this dissertation developed an effectiveness modeling code specifically for CTGHs called Transverse Heat Exchange Effectiveness Model, or THEEM. THEEM is an unconventional finite volume method code that uses empirical heat transfer and pressure drop correlations instead of governing equations for its calculations. It was used to model the temperature and pressure distributions across the CTGH bundle and, consequently, to calculate the effectiveness, fluid outlet temperatures, and fluid pressure drops across the bundle. This program was then used to model the CTGH for the FHR.

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It found that this design, compact enough to fit on a rail car, would transfer the desired heat between the salt and the air, and it would have relatively low pressure drops for the salt and air.

In order to use THEEM as a design tool, it was necessary to validate the code experimentally. Two different experiments that both used water and air as heat transfer fluids were performed for this purpose. The first experiment primarily served as a proof of concept for fabrication of a CTGH tube bundle, but due to poor construction, it did not provide useful experimental data. The second experiment was constructed specifically for comparison with the THEEM code. Its measurements provided useful data to validate THEEM for larger CTGH designs. The second experimental setup also provided opportunities for other experiments. First, airflow measurements were taken around the bundle in order to measure the airflow distribution around the bundle, which can be used to model flow distribution in CTGH bundles. Next, using the Wilson plot method, the setup was used to derive empirical Nusselt number correlations for both the tube-side and shell-side of the heat exchanger. These can be used in system modeling codes to calculate the heat transfer in the CTGH for different reactors. Finally, the setup was used to perform binary impulse measurements for the CTGH. This measured the response of the heat exchanger to an immediate loss of heating power. With further additions to the setup, these tests could be modified to simulate the CTGH’s response to various reactor accidents and transients in both the FHR and other reactors.

Once THEEM had gone through initial experimental validation, it could be expanded as a design tool for other applications besides the FHR. First, a parametric study was performed to measure the effect of changing each aspect of the CTGH geometry on the outlet parameters of the bundle. These results were then used to develop an optimization tool that could design CTGHs for various applications outside of the FHR. This optimization tool used a Monte Carlo method algorithm as well as physical constraints set by the user to design the optimal CTGH for a given application. This tool was then used to design CTGHs for different applications coupling a single-phase coolant with a Brayton cycle. These examples included a CTGH coupling a sodium fast reactor with a supercritical carbon dioxide Brayton cycle, a CTGH coupling a different molten salt reactor with an air cooling system, and CTGHs used in different electrically-heated salt loops that thermally modeled nuclear reactors. This showed that the CTGH could be used in multiple nuclear applications and that THEEM would be an effective design tool for those CTGHs.

The work performed in this dissertation will be essential to deploying large scale CTGHs for the FHR and other reactors. The THEEM code and the results of the experiments in this dissertation can be used for performing a structural analysis of the heat exchanger, studying flow-induced vibration through the tube bundle, studying the effects of tube fouling over the lifetime of the heat exchanger, and performing a cost analysis on fabrication of the heat exchanger. With these future studies and the work presented in this dissertation, the CTGH design will become a more efficient and cost-effective heat exchanger for the FHR and other nuclear Brayton cycles.

This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs.

Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core.

Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time distributions for each fluid flow configuration.

The axial flow configurations in PREX 3.1 showed small changes in pebble motion compared to a reference case with no fluid flow and showed similar overall behavior to PREX 3.0. This suggests that dry experiments can be used for core designs with uniform one-dimensional coolant flow early in the design process at greatly reduced cost. Significant differences in pebble residence times were observed in the cross fluid flow configurations, but these were not accompanied by an overall horizontal diffusion bias. Radial zones showed only a small shift in position due to mixing in the diverging region and remained stable in the active core. The results from this study support the overall viability of the annular PB-FHR core by demonstrating consistent granular flow behavior in the presence of complex reflector geometries and multidimensional fluid flow fields.

GRECO simulations were performed for each of the experiments in this study in order to develop a preliminary validation basis and to understand for which applications the code can provide useful analysis. Overall, the GRECO simulation results showed excellent agreement with the gravity-dominated PREX experiments. Local velocity errors were found to be generally within 10-15% of the experimental data. Average radial zone interface positions were predicted within two pebble diameters. GRECO simulations over predicted the amount of mixing around the average radial zone interface position and therefore can be treated as a conservative upper bound when used in neutronics analysis. Residence time distributions from the GRECO velocity data based on the Monte Carlo algorithm closely matched those derived from the experiment velocity statistics. GRECO simulation results for PREX 3.1 with coupled drag forces showed larger errors compared to the experimental data, particularly in the cases with cross fluid flow. The large discrepancies suggest that GRECO results in systems with coupled fluid drag forces cannot be used with high confidence at this point and future development work on coupled pebble and fluid dynamics with multidimensional fluid flow fields is required.

The decay heat from radioactive waste that is to be disposed in the once proposed geologic repository at Yucca Mountain (YM) will significantly influence the mois- ture conditions in the fractured rock near emplacement tunnels (drifts). Additionally, large-scale convective cells will form in the open-air drifts and will serve as an impor- tant mechanism for the transport of vaporized pore water from the fractured rock, from the hot drift center to the cool drift end. Such convective processes would also impact drift seepage, as evaporation could reduce the build up of liquid water at the tunnel wall. Characterizing and understanding these liquid water and vapor transport processes is critical for evaluating the performance of the repository, in terms of water- induced canister corrosion and subsequent radionuclide containment. To study such processes, we previously developed and applied an enhanced version of TOUGH2 that solves for natural convection in the drift. We then used the results from this previous study as a time-dependent boundary condition in a high-resolution seepage model, allowing for a computationally efficient means for simulating these processes. The results from the seepage model show that cases with strong natural convection ef- fects are expected to improve the performance of the repository, since smaller relative humidity values, with reduced local seepage, form a more desirable waste package environment.