The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, burnup, control rod position, etc...) anticipated during the transient. The core calculation is then performed using the few group cross sections in a core simulator which uses some type of coarse mesh nodal method. The multi-step approach was identified as inadequate for several applications such as the design of MOX cores and other highly hetereogeneous, high leakage core designs. Because of the considerable advances in computing power over the last several years, there has been interest in high-fidelity solutions of the Boltzmann Transport equation. A practical approach developed for high-fidelity solutions of the 3D transport equation is the 2D-1D methodology in which the method of characteristics (MOC) is applied to the heterogeneous 2D planar problem and a lower order solution is applied to the axial problem which is, generally, more uniform. This approach was implemented in the DeCART code. Recently, there has been interest in extending such approach to the simulations of design basis accidents, such as control rod ejection accidents also known as reactivity initiated accidents (RIA). The current 2D-1D algorithm available in DeCART only provide 1D axial solution based on the diffusion theory whose accuracy deteriorates in case of strong flux gradient that can potentially be observed during RIA simulations.

The primary ojective of the dissertation is to improve the accuracy and range of applicability of the DeCART code and to investigate its ability to perform a full core transient analysis of a realistic RIA.

The specific research accomplishments of this work include:

* The addition of more accurate 2D-1D coupling and transverse leakage splitting options to avoid the occurrence of negative source terms in the 2D MOC equations and the subsequent failure of the DeCART calculation and the improvement of the convergence of the 2D-1D method.

* The implementation of a higher order transport axial solver based on NEM-Sn derivation of the Boltzmann equation.

* Improved handling of thermal hydraulic feedbacks by DeCART during transient calculations.

* A consistent comparison of the DeCART transient methodology with the current multistep approach (PARCS) for a realistic full core RIA.

An efficient direct whole core transport calculation method involving the NEM-Sn formulation for the axial solution and the MOC for the 2-D radial solution was developed. In this solution method, the Sn neutron transport equations were developed within the framework of the Nodal Expansion Method. A RIA analysis was performed and the DeCART results were compared to the current generation of LWR core analysis methods represented by the PARCS code. In general there is good overall agreement in terms of global information from DeCART and PARCS for the RIA considered. However, the higher fidelity solution in DeCART provides a better spatial resolution that is expected to improve the accuracy of fuel performance calculations and to enable reducing the margin in several important reactor safety analysis events such as the RIA.

The primary objective of this study was to estimate the maximum fuel utilization that is achievable using fast reactors that are designed to operate with fuel reconditioning when the fuel reaches its radiation damage constraint. The primary functions of the fuel reconditioning are to relieve the pressure of the gaseous fission products, replace the clad and reduce radiation induced defects in the fuel. That is, the fuel recycling processes considered cannot be used for the separation of actinides from most of the fission products and to extract plutonium or any other actinide from the fuel. Hence, these recycling processes are highly proliferation resistant and, hopefully, less expensive than processes traditionally considered for used fuel recycling. With the fuel reconditioning recycling, the maximum discharge burnup is dictated by the reactivity of the fuel and is not limited by the material damage constraints. A couple of fuel management strategies were examined: the conventional multi-batch fuel management and a breed-and-burn mode of operation. The two recycling processes examined are an AIROX-like process and the melt-refining process.

It is found that using fuel reconditioning it is possible to increase the fuel utilization by up to one order of magnitude with the conventional mode of operation and by two orders of magnitude using the breed and burn (B&B) mode of operation, relative to that achievable in once-trough LWRs. The conventional mode of operation, in which the fast reactor is constantly fed with enriched fuel, enables to achieve a discharge burnup ranging from 52.4% FIMA for a medium size 1200 MWth fast reactor to 65.3% FIMA for a large 3000 MWth fast reactor. With the innovative breed and burn mode of operation only depleted uranium is required for the fuel feed with the exception of the fissile fuel required for establishing the initial criticality. The achievable burnup is up to 57% FIMA in a 3000 MWth B&B reactor and up to 44% FIMA in a 1200 MWth B&B reactor. In order to sustain the breed and burn mode of operation it is necessary to accumulate in the depleted uranium feed an average burnup of at least 20% FIMA, when using metallic uranium fuel alloyed with 10 weight % zirconium in a tight-lattice core.

By discharging the fuel at this minimum required burnup and loading it, after reconditioning, into a new reactor along with fresh depleted uranium, it is possible to spawn additional B&B reactors without need for any additional fissile fuel. With this spawning mode of operation the achievable B&B reactor capacity growth rate is 3.85% per year, without need for uranium enrichment capability or actinides separation capability. The energy value of the depleted uranium currently accumulated in the USA, when used in the proposed breed-and-burn fast reactors, is equivalent to at least seven centuries of the total 2009 USA supply of electricity, all sources included. Relative to LWR operating with the once-through fuel cycle, the fuel discharged from the B&B fast reactors at ~57% FIMA features, per unit of electricity generated: (a) ~40% the amount of TRU and Pu; (b) ~10% the inventory of 237Np and its precursors; (c) ~12% of the decay heat from TRU; (d) ~28% of the radiotoxicity; (e) ~7% the neutron emission rate; the latter three are measured one year following discharge. The fraction of the fissile isotopes in the discharged plutonium is comparable but the decay heat and neutron emission rate per unit mass of discharged plutonium are nearly half as large.

The proposed modes of operation are expected to improve the economics and the proliferation resistance and, hence, may justify sooner deployment of fast reactors. The deployment of the suggested fast reactor system will constitute a significant step forward towards sustainable nuclear energy. However, technologies for fuel reconditioning need be developed and their economic viability need be established.

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the 200 Displacements per Atom (DPA) radiation damage constraint of presently verified cladding materials. The S&B core is designed to have an elongated seed (or “driver”) to maximize the fraction of neutrons that radially leak into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Two high-performance S&B cores were designed to benefit from this unique synergism: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and fabrication capacity per unit of core power.

Nevertheless, these reference cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43.1% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.

Fuel cycle characteristics of the S&B cores were compared with those of the reference ABR, and a Pressurized Water Reactor (PWR). The fuel cycle cost of the S&B reactor with same LWR TRU transmutation rate as the reference (CR=0.5) ABR is 0.53 cents/kWe-h versus 0.73 cents/kWe-h for the ABR – about 27% lower; it is even lower than that of contemporary PWRs. The longer cycle may enable the S&B cores to operate at a ~10% higher capacity factor and thereby further improve their economic viability. The S&B cores can utilize at least 7% of thorium energy value without a need to develop irradiated thorium reprocessing capability. This is ~12 times the amount of energy that the LWRs generate per unit of natural uranium mined. By softening the blanket spectrum the thorium utilization can increase by a factor of at least three when using thorium hydride rather than metallic fuel; Fully Ceramic Encapsulated (FCM) fueled blanket can achieve the discharge burnup of 481.5 MWd/kg if the FCM fuel keeps its integrity up to such burnup – this is over 80 times the energy extracted by present PWR per unit mass of natural uranium.

If reprocessed, the Trans-Th fuel bred in the S&B core can enable to support new fleets of 233U-Th fuel self-sustaining energy systems that use thermal and epithermal reactors such as Molten Salt Reactors (MSR) and Reduced-moderation Boiling Water Reactors (RBWR). Alternatively, the S&B reactors can be used to close the LWR fuel cycle using a 3-tier system: TRU extracted from Tier-1 LWR is used for fueling the seed of Tier-2 S&B cores while the 233U (Trans-Th) extracted from Tier-2 S&B blanket is used as the fissile feed for Tier-3 LWR. It is estimated that in such a 3-tier energy system 1GWe of S&B SFRs can support 3.3 GWe of PWRs versus ~1.7 GWe of PWRs that can be supported by 1GWe of CR=0.5 ABR.

In summary, the Seed-and-Blanket core concept studied in this project is found highly promising as it offers:

• Improvement in the economic viability of Sodium-cooled Fast Reactors due to the significant reduction in the fuel cycle cost and possibly increase in the capacity factor that may be enabled by the longer cycles. The improved economics may justify earlier commercialization of SFRs.

• Significantly smaller investment in the construction of the fuel reprocessing and the remote fuel fabrication infrastructure required for a given capacity of SFRs.

• Several new promising fuel cycle options feature substantial increase of the thorium resource utilization without fuel reprocessing

• Supporting a large number of LWRs by a given capacity of SFRs on the S&B configuration. Thus, it enables to close the nuclear fuel cycle faster and with smaller investment.

In conclusion, the S&B reactor concept we proposed is feasible and potential to significantly improve the economic viability of fast reactors and of LWR TRU transmuting system using existing structural materials. It enables significant utilization of thorium resource without reprocessing.

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions.

A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement.

A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MW_th, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids - flibe (33%7Li2F-67%BeF) - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR - the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR) - as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector.

A novel neutronics and depletion methodology - the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor.

Three parametric studies were performed for exploring the design space of the PB-FHR -- to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends.

A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients.

Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs - the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration.

Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions.

The optimal PB-FHR design - Mark 1 PB-FHR - is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs) – either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile “seed” regions and fertile “blanket” regions in order to enhance breeding and leakage probability upon coolant voiding.

This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues.

A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was then created for each design, using the Serpent/PARCS 3-D core simulator, and the full core performance was assessed.

The RBWR-SS benefited from a harder spectrum than the RBWR-TR; a hard spectrum promotes breeding and increases the discharge burnup, but reduces the TRU transmutation rate. This led the RBWR-SS to have a very tight lattice, which has a lot of experimental uncertainty in the thermal hydraulic correlations. Two different RBWR-SS designs were created assuming different thermal hydraulic assumptions: the RBWR-SSH used the same assumptions as Hitachi used for the RBWR-AC, while the RBWR-SSM used more conservative correlations recommended by collaborators at MIT. However, the void feedback of the pure Th-fed system was too strongly negative, even with a single elongated seed. Therefore, instead of using just thorium, the self-sustaining designs were fed with a mix of between 30% and 50% DU and the rest thorium in order to keep the void feedback as close to zero as possible. This was not necessary for the RBWR-TR, as the external TRU feed fulfilled a similar role.

Unfortunately, it was found that the RBWR-SSM could not sustain a critical cycle without either significantly downgrading the power or supplying an external feed of fissile material. While the RBWR-SSH and the RBWR-TR could reach similar burnups and transmutation rates to their DU-fueled counterparts as designed by Hitachi, the thorium designs were unable to simultaneously have negative void feedback and sufficient shutdown margin to shut down the core. The multi-seed approach of the Hitachi designs allowed their reactors to have much lower magnitudes of Doppler feedback than the single-seed designs, which helps them to have sufficient shutdown margin. It is expected that thorium-fueled RBWRs designed to have multiple seeds would permit adequate shutdown margin, although care would need to be taken in order to avoid running into the same issues as the DU fueled RBWRs. Alternatively, it may be possible to increase the amount of boron in the control blades by changing the assembly and core design.

Nonetheless, the uncertainties in the multiplication factor due to nuclear data and void fraction uncertainty were assessed for the RBWR-SSH and the RBWR-TR, as well as for the RBWR-TB2. In addition, the uncertainty associated with the change in reactor states (such as the reactivity insertion in flooding the core) due to nuclear data uncertainties was quantified. The thorium RBWRs have much larger uncertainty of their DU-fueled counterparts as designed by Hitachi, as the fission cross section of 233U has very large uncertainty in the epithermal energy range. The uncertainty in the multiplication factor at reference conditions was about 1350 pcm for the RBWR-SSH, while it was about 900 pcm for the RBWR-TR. The uncertainty in the void coefficient of reactivity for both reactors is between 8 and 10 pcm/% void, which is on the same order of magnitude as the full core value.

Finally, since sharp linear heat rate spikes were observed in the RBWR-TB2 simulation, the RBWR-TB2 unit cell was simulated using a much finer mesh than is possible using deterministic codes. It was found that the thermal neutrons reflecting back from the reflectors and the blankets were causing extreme spikes in the power density near the axial boundaries of the seeds, which were artificially smoothed out when using coarser meshes. It is anticipated that these spikes will cause melting in both seeds in the RBWR-TB2, unless design changes – such as reducing the enrichment level near the axial boundaries of the seeds – are made.

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come.

Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed.

In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built.

The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Breed and Burn (B&B) reactors are a special class of fast reactor that promises enormous sustainability and proliferation-resistance benefits compared to current commercial nuclear power systems. Utilizing a very tight neutron economy in the fast spectrum, B&B reactors are able to sustain criticality in the equilibrium state without either fuel enrichment or reprocessing. As with many advanced reactor systems, however, B&B reactors currently suffer from a number of uncertainties and infeasibilities hampering their practical deployment. This dissertation aims to better understand and propose solutions to some of these primary uncertainties. Utilizing state-of-the-art computational tools and techniques, the properties of a prototypical B&B core are investigated and quantified in high fidelity, enabling deeper understanding of the physical phenomenon driving their unique characteristics. Subsequently, three of the most pressing current uncertainties in B&B cores are investigated in detail: (1) reactor orificing, (2) transient performance and inherent safety, and (3) the tradeoff between material integrity and resource utilization.

To understand the feasibility of adequately cooling fuel assemblies over their in-core lifetimes, new mathematical optimization techniques are developed and applied to devise optimal orificing strategies subject to a variety of important operational constraints, the results of which demonstrate how certain methods of orificing are better suited to B&B operation than others. To understand B&B transient performance, the first open-literature transient models are developed and analyzed for a wide variety of hypothetical accident scenarios, finding challenges associated with the unfavorable reactivity feedback mechanisms inherent to B&B core designs. In order to improve transient performance, a novel passive safety system is incorporated into the core and its behavior simulated using a newly-coupled code system. Finally, the topic of material damage is explored by first using newly-proposed methods for uncertainty quantification on the material damage levels. Once the uncertainties have been understood, the study moves to exploring the tradeoff between material damage and resource utilization when low-enriched uranium is used as a feed fuel. In order to improve this tradeoff, new two-tiered reactor systems are designed and analyzed which enable significantly higher resource utilization than current light water reactor systems without requiring fuel reprocessing or physical/chemical separations. Each of these explorations allows for some of the uncertainties around B&B reactors to be clarified, and this thesis therefore enables improved viability of implementing the breed-and-burn mode of operation as part of future nuclear power systems.