Achievable Burnup Research Articles

Mini-Reactor Proliferation-Resistant Fuel with Burnable Gadolinia in Once-Through Operation Cycle Performance Verification

A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for refueling and fuel shuffling operations. The incorporation of 0.075 wt.% 237Np provides favorable plutonium isotopic vectors throughout an operational lifetime of 5.5 years. providing 35 MWe. Core performance was assessed using a verification-by-comparison approach for core designs with or without 237Np and/or gadolinia burnable absorber. Burnup Monte Carlo calculations were performed via MCOS coupling of MCNP and ORIGEN to an achievable burnup of ~62.5 GWd/t. The results demonstrate a minimal penalty to reactor performance due to the addition of these materials as compared against the reference design. Coupling of a proliferation-resistant fuel concept with a uniform loading of natural gadolinia burnable absorber for LEU+ fuel (7.5 wt.% 235U/U UO2) provides favorable excess reactivity considerations with minimized concerns for additional residual waste and more uniform distribution of un-depleted 235U in discharged fuel assemblies.

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

UO2-fueled microreactors: Near-term solutions to emerging markets

We perform a scoping study of seven classes of microreactor technologies with a diverse set of coolants, that is, heat pipes, sodium, lead–bismuth, helium, FLiBe, organic fluid and light water. A point design is developed for each reactor class under a consistent set of assumptions about road transportability, irradiation cycle length, passive decay heat removal and reactivity control. All reactors have a thermal spectrum and use low-enriched (≤5%) uranium dioxide fuel, which is readily available and inexpensive. Mature moderator materials, in-core structural and cladding alloys were selected in this study including graphite, Zircaloy and stainless steel 316H. We compare the economic potential of these point designs according to a set of figures of merit including the cost of fuel, coolant, major equipment, instrumentation and control (I&C) systems as well as the complexity of operations and maintenance (O&M).The results of our analyses confirm that economies of scale work against microreactors competitiveness, i.e., relatively small absolute costs become high costs per unit energy generated because of the low power output. We find that all designs have very high fuel costs because of the relatively low achievable burnup, and high I&C costs (in particular for the reactor protection system), compared to traditional large light water reactors. Coolant costs are also notably high for the lead–bismuth cooled, organic-cooled and especially FLiBe-cooled systems. The constraints adopted in this study limited the heat pipe reactor’s operating temperature to 650 °C, which resulted in low core power density and low electric power output, thus yielding a poor economic performance. When technology maturity, operating experience, O&M complexity and our estimated costs are holistically taken into account, we judge that the most promising designs examined in this study are (in no particular order) the Na-cooled reactor, the HTGR, the organic-cooled reactor and the water-cooled reactor.

A HELIOS-Based Dynamic Salt Clean-Up Study Analysing the Effects of a Plutonium-Based Initial Core for iMAGINE

Nuclear technologies have strong potential and a unique role to play in delivering reliable low carbon energy to enable a net-zero society for future generations. However, to assure the sustainability required for its long-term success, nuclear will need to deliver innovative solutions as proposed in iMAGINE. One of the most attractive features, but also a key challenge for the envisaged highly integrated nuclear energy system iMAGINE, is the need for a demand driven salt clean-up system based on the principles of reverse reprocessing. The work described provides an insight into the dynamic interplay between a potential salt clean-up system and reactor operation in a plutonium-started core in a dynamic approach. The results presented will help to optimise the parameters for the salt clean-up process as well as to understand the differences which appear between a core started with enriched uranium and plutonium as the fissile material. The integrated model is used to investigate the effects of the initial fissile material on core size, achievable burnup, and long-term operation. Different approaches are tested to achieve a higher burnup in the significantly smaller Pu-driven core. The effects of different clean-up system throughputs on the concentration of fission products in the reactor salt and its consequences are discussed for general molten salt reactor design. Finally, an investigation into how a plutonium loaded core could be used to provide fuel for future reactors through fuel salt splitting is presented, with the outcome that one Pu-started reactor of the same size as a uranium-started core could deliver fuel for 1.5 new cores due to enhanced breeding. The results provide an essential understanding for the progress of iMAGINE as well as the basis for inter-disciplinary work required for optimising iMAGINE.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

MAXIMISING DISCHARGE BURNUP IN AN OPEN CYCLE MOLTEN SALT REACTOR

This paper discusses work done to find an estimate of the maximum achievable discharge burnup in an open cycle molten salt reactor (MSR). An in-development deterministic code (WIMS11) is used to create a model of a simple generic MSR, and the methodology employed is discussed. Some experimentation is done with regards to the internal set-up of the ‘unit cells’ within the core, which shows there is a strong link between this geometry and the achievable burnup. Work is done to quantify the effects of removing volatile fission products and implementing a two-batch refuelling scheme. Finally, an optimization process is carried out whereby the optimal proportion of graphite moderator within the core is found which balances power across the regions while maximising discharge burnup. Two fuels are tested, one which carries only 235U and 238U, and another which also carries 232Th. It is found that the maximum achievable discharge burnup is approximately 155 MWd/kg, which is considerably higher than modern PWRs, despite a lower enrichment and only two batches of fuel being used.

Feasibility study on burnable poison credit concept to HTGR fuel fabrication from core specification perspective

Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also degrades the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept, and the 14 wt% fuel can be fabricated even with severe Light Water Reactor (LWR) class safety measure corresponding to 5 wt%.

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

Small modular reactors (SMRs) may offer potential benefits relative to large light water reactors, such as enhanced flexibility in deployment and operation. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary neutronics and fuel cycle analyses conducted in support of the US Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a “cartridge type” one-batch core with slightly less than 5.0% enrichment.The high-level issues identified and preliminary scoping calculations in this paper are intended to inform decision makers regarding potential fuel cycle impacts of one-batch thermal-spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and a reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for the simplified example SMR analyzed herein are compared with those for a conventional three-batch light water reactor (LWR) in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of per energy generated, mass of depleted uranium disposed of per energy generated, land use per energy generated, and carbon emissions per energy generated.Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores, which mean high leakage (there is a radial and an axial component); (2) a heterogeneous core and extensive use of control rods and burnable poisons; and (3) single-batch cores. But not all SMR designs have all of these traits. The approach used in this study is an example bounding case, and not all SMRs may be impacted to the same extent.

POWER LEVEL EFFECTS ON THORIUM-BASED FUELS IN PRESSURE-TUBE HEAVY WATER REACTORS

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.

Effects and Modeling of Power History in Thorium-Based Fuels in Pressure-Tube Heavy Water Reactors

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels and power history on the reactivity and achievable burnup for 35-element fuel bundles made with thorium-based fuels, such as (Pu,Th)O2 and (233U,Th)O2. These bundles are designed to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700-MW(electric)–class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Methods have been developed to model time-dependent power histories in lattice physics calculations that are more consistent with core physics analysis results. Results demonstrate that the impact of power/flux level and the modeling of time-dependent power histories on the core power distributions and achievable fuel burnup are modest for Pu/Th fuels but are more significant for 233U/Th fuels. Thus, to reduce the neutron capture rate in 233Pa and to increase fuel burnup and fissile utilization, there may be an incentive to develop solutions to reduce the time-average specific power in the fuel.

Two Methods for Converting a Heavy-Water Research Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup

AbstractThis article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically reconfigured. The analysis was performed for Iran’s IR-40 reactor at Arak in support of negotiations with Iran, but the methods have application to future reactors that present similar nonproliferation challenges. Two methods are considered, and both retain identical power, thermal-hydraulic, and safety profiles as the original reactor design. The conversion options can be implemented at any time during the reactor’s life. The two methods have competing effects on achievable burnup, and they can be combined to produce an optimized core that matches both the fresh-core reactivity and maximum burnup of the original reactor. For the IR-40 example, the optimized design produces weapon-grade plutonium at only about 19% of the rate of the unmodified reactor for the same power level. Ad...

In-core fuel management for AHWR

The primary aim for the in-core fuel management in a nuclear reactor is to achieve higher fuel utilization without compromising the safety during operation. Loading pattern optimization during each refueling stage is the main challenge in light water reactors like pressurized water reactors (PWRs) and boiling water reactors (BWRs). Whereas, in case of heavy water reactors like pressurized heavy water reactors (PHWRs), the development of refueling scheme is relatively simple mainly due to on-line refueling, use of small length bundle and natural uranium as fuel. The use of small length bundle and flexibility of multi-bundle shift scheme helps in controlling the ripples in power due to refueling. Even though the PHWRs possess best neutron economy, the use of natural uranium fuel in PHWRs limits the achievable burn-up to 7500MWd/Te and as a result large amount of waste management is a serious concern. In Indian advance heavy water reactor (AHWR), both the features of PWRs (high discharge burn-up fuel) and PHWRs (on-power refueling) are being included along with advanced safety features. The AHWR shall possess good neutron economy and owing to higher discharge burn-up, lower waste is expected. However, during on-power refueling with full length channel and high enrichment fuel, the challenge is to control the refueling ripples and maintaining the operational parameters under their design limits. This necessitates the implementation of a special refueling strategy to control the power peaking due to on-power refueling. This paper describes about the special refueling strategy for this kind of problems. A computer code based on the heuristic approach has been developed for facilitation of on-power refueling in automatic manner for high discharge burn-up fuel in AHWR core. Further, parallel processing on shared memory architecture (personal computer) has been employed to enhance the efficiency of this code.

Superprism-Sized Breed-and-Burn Sodium-Cooled Core Performance

A preliminary feasibility study is performed for a sodium-cooled breed-and-burn (B&B) fast reactor core for achieving high uranium utilization without solid fission product separation that could fit within a reactor vessel of the dimensions of SuperPRISM (S-PRISM). This 1000-MW(thermal) B&B core is to be fueled with depleted uranium with the exception of the fissile loading required for achieving initial criticality. When the fuel reaches its radiation damage limit, it is reconditioned using the melt-refining process and reloaded into the core until it runs out of reactivity.It is found that the maximum burnup at which the S-PRISM-sized B&B core can be designed to discharge its fuel is 43% fissions per initial metal atom. The corresponding uranium utilization is nearly 90 times higher than that of a light water reactor. The achievable burnup strongly depends on the fuel volume fraction but is almost insensitive to the core power density, fuel-reconditioning frequency, and duration of the fuel-reconditioning process.

Nuclear fission sustainability with subcritical reactors driven by external neutron sources

Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

Revisiting the concept of HTR wallpaper fuel

As early as the 1970s, attempts were made to reduce the peak fuel temperature by means of so-called “wallpaper fuel”, in which the fuel is arranged in a spherical shell within a pebble: By raising particle packing fraction, fuel kernels are condensed to the outer diameter of the fuel zone, leaving a central part of the pebble free of fuel. This modification prevents power generation in this central fuel-free zone and decreases temperature gradient across the pebble. Besides particle temperature reduction, the wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. To assess such improvements, calculations were performed using Monte Carlo neutron transport and depletion codes MCNP/MCB. Among others, investigations of conversion ratio, temperature coefficient of reactivity, spent fuel composition and neutron multiplication (for which a method to determine the six-factor formula was developed), were conducted. It is demonstrated that this fuel type impacts positively on the fuel cycle, reduces production of minor actinides (MA) and improves the safety-relevant parameters of the reactor. A comparison of these characteristics with PBMR-type fuel is presented: By comparison with PBMR fuel, the “wallpaper design” results in an effective neutron multiplication coefficient increase (by about 1750 pcm), which is combined with a decrease of between 4.6 and 17.5% in MA production. An improved neutron economy of the heterogeneous design enables enrichment of the “wallpaper type” of fuel to be reduced by more than 6%. The fuel changes suggested in this paper offer more versatility to the HTR concept: Conversion ratio can be decreased (leading to lower MA build-up and fuel reprocessing cost) or raised (leading to lower fuel consumption and fuel cost). Variations around this concept also enable higher reactivity, thus higher achievable burn-up, improving sustainability of HTRs.

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U–ZrH 1.6 fuelled PWR cores are investigated and compared against those of the currently used UO 2 fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k ∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U–ZrH 1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 v/ o of the ZrH 1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.

Neutronics calculations on the impact of burnable poisons to safety and non-proliferation aspects of inert matrix fuel

Inert matrix fuels (IMF) with plutonium may play a significant role to dispose of stockpiles of separated plutonium from military or civilian origin. For reasons of reactivity control of such fuels, burnable poisons (BP) will have to be used. The impact of different possible BP candidates (B, Eu, Er and Gd) on the achievable burnup as well as on safety and non-proliferation aspects of IMF are analyzed. To this end, cell burnup calculations have been performed and burnup dependent reactivity coefficients (boron worth, fuel temperature and moderator void coefficient) were calculated. All BP candidates were analyzed for one initial BP concentration and a range of different initial plutonium-concentrations (0.4–1.0gcm−3) for reactor-grade plutonium isotopic composition as well as for weapon-grade plutonium. For the two most promising BP candidates (Er and Gd), a range of different BP concentrations was investigated to study the impact of BP concentration on fuel burnup. A set of reference fuels was identified to compare the performance of uranium-fuels, MOX and IMF with respect to (1) the fraction of initial plutonium being burned, (2) the remaining absolute plutonium concentration in the spent fuel and (3) the shift in the isotopic composition of the remaining plutonium leading to differences in the heat and neutron rate produced. In the case of IMF, the remaining Pu in spent fuel is unattractive for a would be proliferator. This underlines the attractiveness of an IMF approach for disposal of Pu from a non-proliferation perspective.

Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235U is necessary, and the 235U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO2-UO2) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233U from the 232Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.