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A global initiative on mitigating the use of fossil fuels and transitioning to a more sustainable and cleaner energy source has brought the potential of nuclear energy to the forefront of the conversation. The modular, high temperature, particle fueled reactor platform has become the leading candidate for advanced nuclear reactors because of its robust safety features, enhanced fuel utilization, and the potential of high-quality heat for industrial applications. This particular reactor platform may take on various forms depending on the specific application, but all designs are centered around the use of particle fuel embedded into a solid moderating material and operate at a considerably higher temperature than more conventional nuclear reactors. The particular reactor concepts that are presented in this dissertation are 1) prismatic high temperature gas cooled microreactors, 2) pebble bed fluoride salt cooled high temperature reactors, and 3) pebble bed high temperature gas cooled reactors. The goals and approaches on the various studies for these concepts differ but each study has the underlying theme of evaluating performance and safety to enhance the knowledge of various high temperature reactor technologies/concepts through the use of multiphysics assessment for the enhancement of nuclear power safety, competitiveness, and efficiency. Results from these studies demonstrate the performance and safety potential of various high temperature reactor concepts as well as highlight some areas for modeling development. A preconceptual high temperature gas cooled microreactor was developed to provide a basis for comparing moderators alternative to graphite for the objective of enhancing reactor performance without compromising safety. The results from the microreactor study demonstrated the higher reactor performance of two-phase composite moderators relative to the industry standard, graphite without compromising intrinsic safety features. A coupled reactor physics and thermal-hydraulics model was used to assess the safety response for a generic pebble bed fluoride salt cooled pebble bed reactor under a set of different transient conditions, demonstrating the resilience of the reactor concept to accidents. Finally, the severe accident and consequence modeling code, MELCOR, was used for verification and validation of steady state and transient conditions for a high temperature gas cooled pebble bed reactor.

This thesis started with the goal of developing SPARTACUS, a toolbox to assess the feasibility of a space mission propelled by nuclear electric propulsion (NEP) or nuclear thermal propulsion (NTP), and how they compare with classical chemical rocket propulsion. While NEP uses a very high specific impulse (ISP) electric thruster powered by a nuclear fission reactor to generate low-thrust for long periods of time, NTP uses a fission reactor core to heat up and push a propellant through a nozzle generating a high-thrust impulse. Chemical propulsion is also based on high-thrust impulses, although with about half the ISP of NTP. In the first phase, low-thrust and high-thrust orbital trajectories based on a simple two-body model were generated in order to obtain the propellant and power requirements as a function of the duration of flight. Both low-thrust, obtained with a numerical direct solver and two heuristics, and high-thrust, obtained with a Lambert solver, solutions were validated with literature references. Further, the different types of propulsion and their specifications were discussed. Some important features not implemented due to time constraints are solar electric propulsion (SEP) transfers and gravity assists. In the second phase, the basics of a nuclear fission microreactor, needed to power a NEP system but also usable for fission surface power, were presented. A survey of past space microreactors in the kWe to MWe range was followed by a more detailed analysis of its main components, namely reactor core, shielding, power conversion, and radiator. Since the ultimate goal will be to assess a NEP mission powered by a custom-designed fission reactor, it was decided to first validate the neutronics analysis of the reactor core and shielding with the LANL reactor "KRUSTY" as reference. Fission spectrum, reflector worth, fuel-temperature reactivity, dose behind shielding, and other relations, showed a good level of agreement with literature values. Many pieces are still missing in order to model and design a complete reactor, also due to the different design options existing. Finally, the developed SPARTACUS tool, in the form of a web app with UI, was used to compare the capabilities of NEP, NTP, or chemical propulsion for missions to Mars, Jupiter, and Saturn, as a function of the reactor specific mass and several other parameters. Additional tools such as plotting the optimal ISP or the maximum reactor specific mass for a mission are also implemented. The results highlighted the strengths of NEP, suited for long high-Delta V missions requiring high payload mass ratios, and its weaknesses, such as the inefficiency of low-thrust when first escaping from the strong gravitational pull of a planet during a so-called escape spiral. A possible mitigation is the use of a bimodal system, which can alternate between high-thrust NTP and low-thrust NEP in a single core.

Modeling and simulation tools play an important role in the design and licensing of reactors, but models alone are insufficient for licensing reactors. As advanced reactors approach deployment, modeling tools must be validated to ensure their applicability. Code-to-code verification studies can be conducted to assess the performance of simulation tools relative to one another. These studies are particularly important when validation data do not exist. This dissertation uses verification studies to assess modeling capabilities for a fluoride salt-cooled high-temperature reactor (FHR) design called the Generic FHR (gFHR) and code-to-data validation studies to assess modeling capabilities for a high-temperature gas-cooled reactor (HTGR) thermal hydraulics test facility called the High Temperature Test Facility (HTTF). Neutronics and thermal hydraulics studies of the gFHR conducted using SCALE and MELCOR demonstrated a capability to produce results that are comparable to those in the literature. The analysis with SCALE also demonstrated and verified a novel approach for determining the equilibrium state of a pebble-bed reactor. Following verification of the SCALE and MELCOR models, a loss of flow accident (LOFA) was modeled. Sensitivity studies on the LOFA showed that inlet temperature, graphite thermal conductivity, SCRAM time, and friction multiplier had significant control over peak temperatures during the transient. Fuel temperatures remained well within safety limits, but coolant outlet temperatures may slightly exceed safety limits based on the primary loop configuration used in this analysis. The HTTF validation analysis assessed RELAP5-3D. Sensitivity studies revealed that the thermal conductivity of HTTF blocks was the most important parameter controlling their temperature in steady state and transients. Calibration studies aimed at validation demonstrated that the PG-26 depressurized conduction cooldown experiment was challenging to model due to confounding events. The PG-27 pressurized conduction cooldown experiment demonstrated that while the RELAP5-3D model could reproduce steady-state temperatures, it significantly underpredicted the transient temperature rise. This is likely due to the nodalization of the model rather than code deficiencies. Finally, models of HTTF experiment PG-29, a depressurized conduction cooldown demonstrated that the PG-27 calibration was reasonable, but the models could not be validated against PG-29 either.

When designing experiments for full-scale reactor systems, MCNP and Whisper can be used to create neutronic models and compare the similarity of two nuclear systems via correlation coefficients for , effective multiplication factor. This thesis applies this framework to a conceptual heat-pipe, yttrium-hydride moderated microreactor system and experiments. The framework is intended as a supplement to other neutronics/thermal/multiphysics analyses and provides a concrete method to measure the neutronic similarity of two systems. By analyzing the shared nuclear data uncertainty, as well as sensitivity to nuclear data over all neutron energies, highly informative experiments can be designed to aid in the development of microreactor and other advanced reactor technologies and systems.

Space exploration enables humans to not only explore unknown worlds but also to satisfy the curiosity about the contents and origins of the Universe. The last and only extraterrestrial body to be explored by humans is the Moon. The next natural step is human exploration of Mars and beyond. The long trip times associated with space exploration beyond low Earth orbit pose many risks to humans such as psychological effects associated with confinement and isolation, cancer and disease associated with increased radiation dose, and deleterious effects of extended exposure to microgravity which affects a range of physiological systems such as ocular and cardiac. One approach to overcoming these challenges is limiting exposure time via high-speed transit. This requires advanced propulsion beyond conventional systems that feature high thrust and high fuel efficiency. A propulsion system that can enable such quick trip times to destinations throughout the solar system is the gas core nuclear rocket concept. The gas core nuclear rocket offers substantial advantages over chemical or electric or even solid core nuclear propulsion systems. Operating at high temperatures, the gas core nuclear rocket achieves high specific impulse and high thrust, essentially eclipsing conventional solid-core nuclear thermal rockets. Indeed the core itself is in the gaseous state and thus conventionally can operate at temperatures so high that the core itself is in the plasma state – thereby enabling the prospect of heating propellant to many times higher than the melting point of the nuclear fuel. Challenges to the realization of this technology include 1) stably confining the fissioning gas core, 2) preventing plasma erosion due to mixing, 3) optimizing the heat transfer from uranium plasma to the hydrogen propellant, 4) protecting the nozzle from the high-temperature exhaust, and 5) obtaining a self-sustaining critical nuclear rocket engine. In this research, the criticality of an open cycle gas core nuclear rocket engine and the heat transfer from uranium to hydrogen was studied. Material properties at high temperature and high pressure were calculated or obtained from the literature. Criticality analysis was carried out in the MCNP transport code for spherical non-homogenous gas core nuclear rocket geometry. These criticality analyses proved that a self-sustaining engine can be achieved by optimizing the geometry and the heat transfer in this engine. The heat transfer analyzes were carried out from the basic heat transfer equations. The hydrogen propellant temperatures varied from 3000 K to 40,000 K depending on the uranium core temperature which varied from 10,000 K to 55,000 K. For these temperatures, a critical engine could be obtained by adjusting the amount of uranium in the system. In this modeling activity, confinement of the uranium core is assumed to be achieved hydrodynamically to reduce the loss of uranium from the system. An optimized nozzle was designed for the gas core nuclear rocket commensurate with the high chamber pressure required for this engine to operate. Using this nozzle design and energy transferred to the hydrogen propellant, the gas core nuclear rocket’s performance was calculated. The calculated specific impulse ranged from 1000 s to 6200 s and the thrust ranged from 50,000 N to 300,000 N with a mass of approximately 100,000 kg. This analysis confirms that high-performance is achievable from a gas core nuclear rocket. Indeed, criticality is possible provided technical challenges of confinement and nozzle survivability are addressed.

Neutronics study to determine the validity and real world extendibility through Artificial Intelligence (AI) optimization of a transmutation concept with the aim of significantly reducing the radiotoxicity of Spent Nuclear Fuel (SNF) together with its required storage duration and volume. In this transmutation concept (the transmutator) utilizes a novel source providing laser-generated neutrons to transmute transuranic elements separated from SNF and dissolved in a molten salt within a subcritical core. The source neutrons are generated via beam-target fusion whereas the beam is created by laser irradiation of nanometric foils through the Coherent Acceleration of Ions by Laser (CAIL) process. This relatively low deuteron energy is catapulted by fusion and eventually by secondary fission processes. A significant factor in this study is that this can be accomplished using relatively cheap fiber lasers terminating onto small scale targets. Consequently, this makes possible the use of multiple tunable and distributable neutron sources. Such a source has not previously been considered and encourages an investigation with the aid of AI into new spatial arrangement and temporal control operation strategies as done here. This source is combined with a molten salt core whose liquid state allows and facilitates homogeneity by mixing, safety, laser irradiation, in-situ processing, and monitoring of chemical and physical properties. The combined use of molten salt and laser also allows for the introduction of rapid feedback or feedforward control of the system’s operation. This further extends the transmutator concept as this work demonstrates here by neutronics simulations showing: Efficient and continuous Minor Actinide (MA) waste only burning through AI optimized and balanced processing scheduling without the use of isotopic separation. As well as shaped thermal reactivity insertion for increased tank usage efficiency and active thermal management.

Pebble-bed fuelled, molten salt-cooled nuclear reactors (PB-FHRs) involve interactions between a fluid coolant phase, solid particle fuel phase and neutron interactions enabling fission of nuclear fuel for heat production. As these three aspects are highly interdependent, this work presents a three-way coupled methodology for simulation of PB-FHRs to provide a robust analysis tool for detailed analysis of such reactor designs. Computational fluid dynamics (CFD) techniques are implemented for simulation of the coolant fluid’s flow and temperature fields. Discrete element methods (DEM) are used to model individual pebble fuel elements present within the core, focusing on particle-particle interactions and structures formed within packed beds of mono-sized spheres. Modelling of the interaction between these two phases is realised through a two-way coupled methodology and open-source codes. Additional to the interaction of the fluid and pebble phases, consideration of the neutronic interactions resulting in nuclear heating of the fuel pebbles is included. The Monte Carlo neutron transport method is employed for determination of the fission rate distribution for PB-FHR geometries. Inclusion of the neutronics results within the CFD and DEM solvers, and vice versa, results in three-way coupled multiphysics simulations for PB-FHRs. This work has been built up over successively more complex simulations, starting with CFD-DEM replication of an isothermal pebble re-circulation experiment and culminating in the three-way coupled simulation of a true PB-FHR design. Additional analysis focusing on the neutronic behaviour of the PB-FHR geometry for various geometric and material property configurations is also presented. It was found that the multiplication factor of PB-FHRs with a free surface of pebbles within the core region are highly influenced on the shape of this free surface. The three-way coupling methodology was shown to provide detailed representations of PB-FHR geometries in a timely manner without the need for simplification of the geometry or material properties. This allowed for detailed analysis of coolant, fuel pebble and neutronic properties throughout the core region. The results obtained demonstrate the suitability for use as an accurate design tool.

The neutronic behavior of accident tolerant fuel (ATF) concepts was simulated in light water reactors (LWRs) to establish design parameters to match reactivity lifetime requirements of standard UO2 [uranium dioxide]/Zircaloy fuel. The two concepts discussed in this dissertation are fully ceramic micro-encapsulated (FCM) fuel and alternate cladding concepts. To compare the required fuel alterations against standard UO2/Zircaloy fuel, a 2D lattice-physics based reactivity equivalence method was established to estimate excess reactivity at the completion of each weighted batch cycle. In the case of FCM fuel, the uranium-based tristructural isotropic (TRISO) kernel and the surrounding particle layers/matrix material were altered to increase fissile loading. To match the lifetime of an 18-month pressurized water reactor (PWR) cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life (BOL) compared with that of a standard UO2 rod. When investigating alternate cladding concepts, cladding walls were thinned with the outer diameter unchanged, so the pellet volume and enrichment of UO2 fuel were increased. In the PWR study, a cladding thickness of 350 μm [micrometer] was simulated. Austenitic stainless steels required an increase of about 0.5 wt % enrichment to match fuel cycle requirements, while the required increase in enrichment for FeCrAl was about 0.1%. Due to the presence of the channel box, the boiling water reactor (BWR) ATF designs required additional fissile material. With the FeCrAl cladding and channel box thicknesses halved, it was estimated that an average enrichment increase of 0.6% would be required. Verification of the 2D reactivity results was performed with a 3D full-core parametric study of a representative BWR demonstrating the applicability of the 2D reactivity equivalence method for the cases herein studied. A LWR optimization code (LWROpt) was used to determine loading (LP) and control blade (CB) patterns for the ATF BWR concepts, so to help regain thermal and reactivity margins. Fuel performance was investigated with the BISON-CASL code using linear heat rate data from the optimized full-core results. The analysis demonstrated that varying power histories between FeCrAl and Zircaloy cladding greatly affect thermal expansion and centerline temperatures of the fuel rods.

This dissertation discusses the research and development for a coupled neutron trans- port/isotopic depletion capability for use in high-preformance computing applications. Accurate neutronics modeling and simulation for \real" reactor problems has been a long sought after goal in the computational community. A complementary \stretch goal" to this is the ability to perform full-core depletion analysis and spent fuel isotopic characterization. This dissertation thus presents the research and development of a coupled Monte Carlo transport/isotopic depletion implementation with the Exnihilo framework geared for high-performance computing architectures to enable neutronics analysis for full-core reactor problems. An in-depth case study of the current state of Monte Carlo neutron transport with respect to source sampling, source convergence, uncertainty underprediction and biases associated with localized tallies in Monte Carlo eigenvalue calculations was performed using MCNPand KENO. This analysis is utilized in the design and development of the statistical algorithms for Exnihilo's Monte Carlo framework, Shift. To this end, a methodology has been developed in order to perform tally statistics in domain decomposed environments. This methodology has been shown to produce accurate tally uncertainty estimates in domain-decomposed environments without a significant increase in the memory requirements, processor-to-processor communications, or computational biases. With the addition of parallel, domain-decomposed tally uncertainty estimation processes, a depletion package was developed for the Exnihilo code suite to utilize the depletion capabilities of the Oak Ridge Isotope GENeration code. This interface was designed to be transport agnostic, meaning that it can be used by any of the reactor analysis packages within Exnihilo such as Denovo or Shift. Extensive validation and testing of the ORIGEN interface and coupling with the Shift Monte Carlo transport code is performed within this dissertation, and results are presented for the calculated eigenvalues, material powers, and nuclide concentrations for the depleted materials. These results are then compared to ORIGEN and TRITON depletion calculations, and analysis shows that the Exnihilo transport-depletion capability is in good agreement with these codes.

The objective of this research is to, based on the original design for the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), develop an MCNPX model of the reactor core with the objective to attain criticality and to breed new fuel. A brief but complete description of a first approach to the PB-AHTR will be provided and a MCNPX model will be run in order to ascertain the difficulties of that configuration. On the second part, a modification of the original model will be evaluated and compared in order to resolve the difficulties encountered in the original design. Finally, in an effort to optimize the design, an evolutionary approach will be analyzed, based on the previous model, and conclusions will be attained

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (

The purpose of this research has been to develop an advanced Super-Nodal method to reduce the run time of 3-D core neutronics models, such as in the NESTLE reactor core simulator and FORMOSA nuclear fuel management optimization codes. Computational performance of the neutronics model is increased by reducing the number of spatial nodes used in the core modeling. However, as the number of spatial nodes decreases, the error in the solution increases. The Super-Nodal method reduces the error associated with the use of coarse nodes in the analyses by providing a new set of cross sections and ADFs (Assembly Discontinuity Factors) for the new nodalization. These so called homogenization parameters are obtained by employing consistent collapsing technique. During this research a new type of singularity, namely "fundamental mode singularity", is addressed in the ANM (Analytical Nodal Method) solution. The "Coordinate Shifting" approach is developed as a method to address this singularity. Also, the "Buckling Shifting" approach is developed as an alternative and more accurate method to address the zero buckling singularity, which is a more common and well known singularity problem in the ANM solution. In the course of addressing the treatment of these singularities, an effort was made to provide better and more robust results from the Super-Nodal method by developing several new methods for determining the transverse leakage and collapsed diffusion coefficient, which generally are the two main approximations in the ANM methodology. Unfortunately, the proposed new transverse leakage and diffusion coefficient approximations failed to provide a consistent improvement to the current methodology. However, improvement in the Super-Nodal solution is achieved by updating the homogenization parameters at several time points during a transient. The update is achieved by employing a refinement technique similar to pin-power reconstruction. A simple error analysis based on the relative residual in the 3-D few group diffusion equation at the fine mesh level is also introduced in this work. The proposed Super-Nodal method was tested for PWR cores with large heterogeneities and under severe reactor accident conditions. Two nodalizations, SN1 and SN2, have been used with three types of collapsing schemes, namely SN1, SN2, and SN1-r. The core is axially collapsed from 18 nodes to 12 and 8 nodes in the SN1 and SN2 nodalizations, respectively. While the SN1 and SN2 schemes refer to SN1 and SN2 nodalizations, SN1-r refers to SN1 nodalization with multiple refining steps. The first test case is the steady state analysis of a core. The general core geometry is modified to introduce a strong heterogeneity by introducing partial length BP rods in the four central fuel assemblies. Even though the Super-Nodal method agrees well with the fine i.e. reference, solution for the axial and radial power distributions, it fails to represent the intra-nodal flux shape for the BP loaded assemblies. For the second test case the core is depleted and a control rod bank is inserted and withdrawn in 40 seconds. A computer CPU time speedup of 1.53 and 1.64 are achieved for the SN1 and SN2 schemes, respectively. Both schemes represent the core power level with a small error; however, both schemes fail to accurately represent the peak node power during the transient. The SN1-r scheme with two additional refining steps removes this discrepancy with a decreased speedup of 1.24. Test Case 3 is a rod ejection accident. Similar to Test Case 2, all three schemes predict accurately the core power level; however, only the SN1-r scheme shows good accuracies for the peak node power. The SN1, SN2 and SN1-r schemes give speedups of 1.53, 2.08 and 1.26, respectively. Test Case 4 is a large steam line break accident. The SN2 and SN1 schemes fail to converge beyond the point that the prompt criticality is reached. The SN1-r scheme agrees well with the reference solution?s axial and radial power distributions. It also provides an accurate prediction of peak node power except at the time of prompt criticality where the peak node power error reaches 10% of full rated power. The speedup is also small for this case, 1.15, due to the need for frequent refine-collapse steps in order to achieve adequate accuracy.

The conceptual design of an Inertial Confinement Fusion (ICF) Hybrid Breeder reactor is presented. The primary design objectives were as follows, (1) identify a workable ICF hybrid system in which radiation damage to structural components is reduced to allow for extended operation between component replacement and (2) enhance fissile fuel production from the hybrid to provide as large a light water reactor (LWR) support ratio as possible. Radiation damage to structural components is reduced by filling the fusion chamber with flowing lithium jets. These jets form a protective curtain which shields the first wall from x-ray and charged particle debris and reduces radiation damage to structural material by softening the neutron spectrum. This flowing lithium first wall protection scheme is adapted from that proposed for the High Yield Lithium Injected Fusion Energy (HYLIFE) fusion reactor design. Hybrid reactor configurations are presented in which radiation damage is reduced significantly. In some situations no component replacement may be necessary during the entire hybrid reactor operating life time. In the designs presented here, fertile thorium is circulated through the hybrid blanket zone dissolved in a LiF-BeF(,2)-ThF(,4) molten salt. Considerable experience with the handling and processing of salts of this type has been gained as a result of the Molten Salt Breeder Reactor (MSBR) Program. Salts of these constituents have low parasitic neutron absorption cross sections, are capable of dissolving large amounts of fertile and fissile material and are chemically stable at very high temperatures and in intense radiation environments. The fluidized blanket leads to unique operational characteristics for this hybrid. Fissile fuel production is enhanced since bred ('233)U can be extracted from the blanket zone on a continuous basis. Fissioning of bred fuel is minimized since the fissile concentration in the blanket zone is always low. This leads to a very low blanket zone power density (7.3 W/cm('3)). Fission products that do accumulate in the blanket fuel salt can also be removed as an on-line process so that radiological problems in accident situation are minimized. For hybrid reactors investigated in this study, blanket energy multiplications (total blanket power/incident neutron power) are as low as 1.4 but large values of specific fuel production, 1.3-1.55 kg/MW(,th)-yr (0.75 capacity factor included), result. While the low power producing hybrids have large recirculating power fractions, large LWR support ratios are obtained. In this situation symbiotic system electricity costs are found to be much less sensitive to increases in hybrid reactor capital cost than they are for high energy multiplication hybrids with lower support ratios. This leads to more favorable economics for the suppressed fission, enhanced fuel production hybrid reactor designs presented here.