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This paper presents an optimization of a thorium–plutonium fuel cycle, through a multi-batch reloading scheme, developed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The I2S-LWR is an advanced 2850 MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. Its baseline fuel and cladding materials are U3Si2 and advanced FeCrAl steel, respectively. The advanced steel cladding can withstand longer exposure periods with significantly lower degradation rates compared to traditional Zr-based alloys. In principle, increasing the number of batches allows higher discharge burnups and thus deeper Pu and transuranic elements incineration to be achieved. Therefore, various refuelling strategies were considered in this study, namely 3-, 5- and 7.56-batch schemes. The Simulated Annealing optimization technique was applied for the different batch schemes to obtain the most favourable loading pattern with respect to cycle length performance. The results confirm that increasing the number of batches allows the discharge burnup to be increased by 20% (above 100 MWd/kg), which improves the Pu incineration performance. In addition, the increased number of batches improves the reactivity coefficients without violating the power peaking factors limits.

Ever-increasing demands of uncertainty-based design, analysis, and optimization in aerospace vehicles motivate the development of Monte Carlo methods with wide adaptability and high accuracy. This paper presents a comprehensive review of typical improved Monte Carlo methods and summarizes their characteristics to aid the uncertainty-based multidisciplinary design optimization (UMDO). Among them, Bayesian inference aims to tackle the problems with the availability of prior information like measurement data. Importance sampling (IS) settles the inconvenient sampling and difficult propagation through the incorporation of an intermediate importance distribution or sequential distributions. Optimized Latin hypercube sampling (OLHS) is a stratified sampling approach to achieving better space-filling and non-collapsing characteristics. Meta-modeling approximation based on Monte Carlo saves the computational cost by using cheap meta-models for the output response. All the reviewed methods are illustrated by corresponding aerospace applications, which are compared to show their techniques and usefulness in UMDO, thus providing a beneficial reference for future theoretical and applied research.

© 2016. This version is made available under the CC-BY-NC-ND 4.0 license http://creativecommons.org/licenses/by-nc-nd/4.0/ In this paper a novel alternative for bulk electromagnetic separation working at high pressures is proposed. It is shown that if a self-induced Hall potential is stimulated in the boundaries, the system will be able to take advantage of the collisions process, boosting the isotopic separation and resulting in a linear-spectrometer with a higher spatial separation per unit length than a traditional calutron. Although originally the concept was devised for the production of medical isotopes where the minority isotope to be separated is produced by neutron capture and is the heavier isotope, if the Hall potential is replaced by an external electrical field, the concept is equally applicable for situations where the minority isotope is the lighter one, as for example in the enrichment of uranium. Additional R&D is required to explore further the possibilities of this concept and to identify optimal values for several of the system design variables.

Graphical abstractAttainment surfaces of the solutions with best (left) and worst (right) IGD values for various algorithms on UF1 problem. Display Omitted HighlightsWe studied the effect of local search on multi-objective memetic algorithm.The base algorithm is single surrogate-based multi-objective memetic algorithm.We investigated four types of local search.Local search based on achievement-type function were more robust than the others.The achievement-type function found better solutions on real-world problem. A comparative study of the impacts of various local search methodologies for the surrogate-assisted multi-objective memetic algorithm (MOMA) is presented in this paper. The base algorithm for the comparative study is the single surrogate-assisted MOMA (SS-MOMA) with the main aim being to solve expensive problems with a limited computational budget. In addition to the standard weighted sum (WS) method used in the original SS-MOMA, we studied the capabilities of other local search methods based on the achievement scalarizing function (ASF), Chebyshev function, and random mutation hill climber (RMHC) in various test problems. Several practical aspects, such as normalization and constraint handling, were also studied and implemented to deal with real-world problems. Results from the test problems showed that, in general, the SS-MOMA with ASF and Chebyshev functions was able to find higher-quality solutions that were more robust than those found with WS or RMHC; although on problems with more complicated Pareto sets SS-MOMA-WS appeared as the best. SS-MOMA-ASF in conjunction with the Chebyshev function was then tested on an airfoil-optimization problem and compared with SS-MOMA-WS and the non-dominated sorting based genetic algorithm-II (NSGA-II). The results from the airfoil problem clearly showed that SS-MOMA with an achievement-type function could find more diverse solutions than SS-MOMA-WS and NSGA-II. This suggested that for real-world applications, higher-quality solutions are more likely to be found when the surrogate-based memetic optimizer is equipped with ASF or a Chebyshev function than with other local search methods.

In this paper we present a multi-fidelity (MF) extension of non-intrusive polynomial chaos based on regression (point collocation) for uncertainty quantification purposes. The proposed method uses the principle of a global correction function from a previous similar method that uses spectral projection to estimate the coefficients. Due to its usage of regression to estimate the coefficients, the present method offers high flexibility in the sampling and generation of the polynomial basis. The method takes advantage of a nested sampling plan to create the samples for the low-fidelity (LF) and correction expansions where the high-fidelity (HF) samples are a subset of the LF ones. To build the polynomial basis, a total order or hyperbolic truncation strategy is used with a highly flexible combination of the LF and correction polynomial expansions. The method is demonstrated on some artificial test problems and aerodynamic problems of the Euler flow around an airfoil and common three-dimensional research models. In order to derive the strategies for successful MF approximation, the effect of the correlation and the errors between the LF and HF functions is also studied. The results show that high correlation and moderately low errors are important to improve the MF approximation’s accuracy. On a common research model problem, the MF approach with partially-converged simulations as the LF samples can successfully reduce the computational cost to about 40% for similar accuracy compared to an approach using a single HF expansion.

Reliability-based robust design optimization of modern small satellites has recently been receiving much attention. The role of multidisciplinary design optimization considering system uncertainty is increasingly being recognized in improving satellite performance, safety, and reliability. However, high-dimensional uncertainty hampers the efficiency and accuracy of reliability-based robust design optimization. In this study, a novel multidisciplinary optimization methodology suitable for high-dimensional uncertainty is established and verified. The in-loop uncertainty quantification employing active subspace identification makes the uncertainty propagation and management much easier. Based on the discovered active subspace and a system decoupling strategy, a multiobjective alliance algorithm is presented for the complex constrained design optimization. Typical uncertain factors in small satellites are characterized and interdisciplinary relations are analyzed. A standard test example of speed reducer desi...

Many quantities of interest (qois) arising from differential-equation-centric models can be resolved into functions of scalar fields. Examples of such qois include the lift over an airfoil or the displacement of a loaded structure; examples of corresponding fields are the static pressure field in a computational fluid dynamics solution, and the strain field in the finite element elasticity analysis. These scalar fields are evaluated at each node within a discretised computational domain. In certain scenarios, the field at a certain node is only weakly influenced by far-field perturbations; it is likely to be strongly governed by local perturbations, which in turn can be caused by uncertainties in the geometry. One can interpret this as a strong anisotropy of the field with respect to uncertainties in prescribed inputs. We exploit this notion of localised scalar-field influence for approximating global qois, which often are integrals of certain field quantities. We formalise our ideas by assigning ridge approximations for the field at select nodes. This embedded ridge approximation has favourable theoretical properties for approximating a global qoi in terms of the reduced number of computational evaluations required. Parallels are drawn between our proposed approach, active subspaces and vector-valued dimension reduction. Additionally, we study the ridge directions of adjacent nodes and devise algorithms that can recover field quantities at selected nodes, when storing the ridge profiles at a subset of nodes—paving the way for novel reduced order modelling strategies. Our paper offers analytical and simulation-based examples that expose different facets of embedded ridge approximations.

Numerical instability resulting from coupling isotopic depletion and neutronics is widely reported, but little work has been done to interrogate the phenomenon theoretically. This work extends the pioneering analysis developed by Densmore et al., who examined the stability of a monoenergetic neutron diffusion problem coupled with a simple depletion system through the explicit Euler scheme. Their Fourier stability analysis is applied to the predictor-corrector scheme and its iterated and implicit variants. The results reproduce behaviour described in previous reports of numerical instability and its solution, and develop a future avenue for diagnosing the stability of burn-up problems.

Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO2 and micro-heterogeneous ThO2-UO2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% 235U for UO2 and duplex fuels, respectively, loaded into 13×13 pin arrays. High thickness (150 μm) ZrB2 integral fuel burnable absorber (IFBA) pins and boron carbide (B4C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty. In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the ∼15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve ∼4% longer core life, has a ∼3% lower initial reactivity and ∼30% lower reactivity swing over life than the final UO2 core design. The duplex core is therefore the more successful design, giving a core life of ∼16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and ‘open the option’ of designing the functional cores with both the duplex and UO2 fuel cores.

The stochastic implicit Euler scheme and its variants can be used to prevent non-physical behaviour that may emerge when coupling Monte Carlo neutron transport and isotopic depletion solvers for spatially-decoupled reactor problems. However, stochastic implicit methods tend to require many iterations to obtain a stable solution. This paper demonstrates that this is due to using a sub-optimal relaxation scheme: rather than using a variable relaxation factor, a fixed relaxation factor can give a more stable solution in fewer iterations. Furthermore, like stochastic implicit schemes, even though multiple transport solutions are required, using a fixed relaxation factor allows computational effort to be reduced by lowering the number of particles simulated during the corrector step while still providing stable results. This shows that using a fixed relaxation factor to stabilise Monte Carlo burn-up calculations can be more effective than applying the stochastic approximation.

© 2018 Elsevier Ltd Multi-objective optimization of nuclear engineering fuel assembly design problems is particularly difficult due to the highly non-linear interactions of a large number of possible variables. In addition, effective optimization algorithms are often highly problem-dependent and require extensive tuning, which reduces their applicability to the real world. To address this issue, Differential Evolution (DE) algorithms have been proposed as a new and effective method for heterogeneous fuel assembly optimization design problems. This paper presents the first complete study to investigate their applicability and performance. Firstly, two multi-objective DE algorithms have their performance compared against an Evolutionary Algorithm (EA) from the literature in optimizing a CORAIL mixed-oxide (MOX) fuel assembly for maximum plutonium content and minimum power peaking factor. Statistical analysis of the results shows the DE algorithms exhibit superior performance to the EA. The DE algorithms are then used to optimize a MOX fuel assembly with gadolinia poison, with results showing DE produces assembly designs comparable in performance to those in the literature. Finally, a sensitivity study is conducted on the control parameters of the better performing of the DE algorithms. Results indicate DE performance remains consistent for a wide range of values of both control parameters, suggesting the algorithm is able to perform effectively without requiring user expertise or effort to find the ‘optimal’ control parameter settings.

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings – austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials – as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and, for purposes of comparison, 15% 235U enriched homogeneously mixed all-UO2 fuel, loaded into 13 × 13 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level. The results show that candidate fuels with 310SS cladding exhibit a ∼ 13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼ 10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼ 1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239Pu (for UO2 fuel) and 233U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239Pu and 233U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

In an effort to decarbonize the marine sector, there are growing interests in replacing the contemporary, traditional propulsion systems with nuclear propulsion systems. The latter system allows freight ships to have longer intervals before refueling; subsequently, lower fuel costs, and minimal carbon emissions. Nonetheless, nuclear propulsion systems have remained largely confined to military vessels. It is highly desirable that a civil marine core not to use highly enriched uranium, but it is then a challenge to achieve long core lifetime while maintaining reactivity control and acceptable power distributions in the core. The objective of this study is to design a civil marine core type of single batch small modular reactor (SMR) with low enriched uranium (LEU) ( 20% 235U enrichment), a soluble-boron-free (SBF) and using mixed D 2 O + H 2 O coolant for operation period over a 20 years life at 333 MWth. Changing the coolant properties is the way to alter the neutron energy spectrum in order to achieve a self-sustaining core design of higher burnup. Two types of LEU fuels were used in this study: micro-heterogeneous ThO2-UO2 duplex fuel (18% 235U enriched) and all-UO2 fuel (15% 235U enriched). 2D Assembly designs are developed using WIMS and 3D whole-core model is developed using PANTHER code. The duplex option shows greater promise in the final burnable poison design with high thickness ZrB2 integral fuel burnable absorber (IFBA) while maintaining low, stable reactivity with minimal burnup penalty. For the final poison design with ZrB2, the duplex contributes ∼ 2.5% more initial reactivity suppression, although the all-UO2 design exhibits lower reactivity swing. Three types of candidate control rod materials: hafnium, boron carbide (B4C) and 80% silver + 15% indium + 5% cadmium (Ag-In-Cd) are examined and duplex fuel exhibits higher control rod worth with the candidate materials. B4C shows the greatest control reactivity worth for both the candidate fuels, providing ∼ 3% higher control rod worth for duplex fuel than all-UO2. Finally, 3D whole-core results from PANTHER show that the use of the mixed coolant contributes to ∼ 21.5 years core life, which is a ∼ 40% increase in core life compared to H 2 O coolant ( ∼ 15.5 years) while using the same fuel candidates and fissile enrichment. The mixed coolant provides excellent core lifetimes comparable to those of HEU military naval vessels ( ∼ 25 years vs. ∼ 21.5 years) while utilizing LEU candidate fuels.

Copyright © 2018 ASME. The UK Department of Business, Energy and Industrial Strategy (BEIS) recently launched an R&D programme in Digital Reactor Design, incorporating the development of a Nuclear Virtual Engineering Capability with an integrated Modelling and Simulation programme. A key challenge of nuclear reactor design and analysis is the system complexity, which arises from a wide range of multi-physics phenomena being important across multiple length scales. This project constitutes the first step towards developing an integrated nuclear digital environment (INDE) linking together models across physical domains and incorporating real world data across all stages of the nuclear lifecycle. Simulation case studies will be developed within the INDE framework, delivering an enhanced modelling capability while ensuring the framework has immediate application. For these case studies have been specified that are relevant to design and operation phases for AGR and PWR type reactors. The AGR case considers the through-life structural performance of graphite bricks. This involves modelling of multi-scale, multi-physics phenomena in the support of reactor operations. The PWR case study is based on core multiphysics modelling, with potential relevance to operating and future PWRs, and in particular in the design of SMRs.

This article presents an optimisation framework for the compliance minimisation of structures subjected to design-dependent pressure loads. A finite element solver coupled to a Lattice Boltzmann method is employed, such that the effect of the fluid-structure interactions on the optimised design can be considered. It is noted that the main computational expense of the algorithm is the Lattice Boltzmann method. Therefore, to improve the computational efficiency and to assess the effect of the fluid-structure interactions on the final optimised design, the degree of coupling is changed. Several successful topology optimisation algorithms exist with thousands of associated publications in the literature. However, only a small portion of these are applied to real-world problems, with even fewer offering a comparison of methodologies. This is especially important for problems involving fluid-structure interactions, where discrete and continuous methods can provide different advantages. The goal of this research is to couple two key disciplines, fluids and structures, into a topology optimisation framework, which shows fast convergence for multi-physics optimisation problems. This is achieved by offering a comparison of three popular, but competing, optimisation methodologies. The needs for the exploration of larger design spaces and to produce innovative designs make meta-heuristic algorithms less efficient for this task. A coupled analysis, where the fluid and structural mechanics are updated, provides superior results compared with an uncoupled analysis approach, however at some computational expense. The results in this article show that the method is sensitive to whether fluid-structure coupling is included, i.e. if the fluid mechanics are updated with design changes, but not to the degree of the coupling, i.e. how regularly the fluid mechanics are updated, up to a certain limit. Therefore, the computational efficiency of the algorithm can be considerably increased with small penalties in the quality of the objective by relaxing the coupling.

Uncertainty quantification has recently been receiving much attention from aerospace engineering community. With ever-increasing requirements for robustness and reliability, it is crucial to quantify multidisciplinary uncertainty in satellite system design which dominates overall design direction and cost. However, coupled multi-disciplines and cross propagation hamper the efficiency and accuracy of high-dimensional uncertainty analysis. In this study, an uncertainty quantification methodology based on active subspaces is established for satellite conceptual design. The active subspace effectively reduces the dimension and measures the contributions of input uncertainties. A comprehensive characterization of associated uncertain factors is made and all subsystem models are built for uncertainty propagation. By integrating a system decoupling strategy, the multidisciplinary uncertainty effect is efficiently represented by a one-dimensional active subspace for each design. The identified active subspace is checked by bootstrap resampling for confidence intervals and verified by Monte Carlo propagation for the accuracy. To show the performance of active subspaces, 18 uncertainty parameters of an Earth observation small satellite are exemplified and then another 5 design uncertainties are incorporated. The uncertainties that contribute the most to satellite mass and total cost are ranked, and the quantification of high-dimensional uncertainty is achieved by a relatively small number of support samples. The methodology with considerably less cost exhibits high accuracy and strong adaptability, which provides a potential template to tackle multidisciplinary uncertainty in practical satellite systems.

Full recycling of transuranic (TRU) isotopes can in theory lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited timeframe, the radiotoxicity of the ‘final’ core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In Part I of this paper, TRU recycle over up to 5 generations of light water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. U-fuelled SFRs operating a break-even fuel cycle are not particularly effective at reducing repository radiotoxicity as the final core load dominates over a very long timeframe. In this paper, the analysis is extended to the thorium (Th) fuel cycle. Closed Th-based fuel cycles are well known to have lower equilibrium radiotoxicity than U-based fuel cycles but the time taken to reach equilibrium is generally very long. Th burner fuel cycles with SFRs are found to result in very similar radiotoxicity to U burner fuel cycles with SFRs for one less generation of reactors, provided that protactinium (Pa) is recycled. Th-fuelled reduced-moderation boiling water reactors (RBWRs) are also considered, but for burner fuel cycles their performance is substantially worse, with the waste taking ∼3–5 times longer to decay to the reference level than for Th-fuelled SFRs with the same number of generations. Th break-even fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years. While this is a very long timeframe, it is roughly half that required for waste from the Th or U burner fuel cycle to decay to the reference level, and less than a tenth that required for the U break-even fuel cycle. The improved performance over burner fuel cycles is due to a more substantial contribution of energy generated by 233U leading to lower radiotoxicity per unit energy generation. To some extent this an argument based on how the radiotoxicity is normalised: operating a break-even fuel cycle rather than phasing out nuclear power using a burner fuel cycle results in higher repository radiotoxicity in absolute terms. The advantage of Th break-even fuel cycles is also contingent on recycling Pa, and reprocessing losses are significant also for a small number of generations due to the need to effectively burn down the TRU. The integrated decay heat over the scenario timeframe is almost twice as high for a break-even Th fuel cycle than a break-even U fuel cycle when using SFRs, as a result of much higher 90Sr production, which subsequently decays into 90Y. The peak decay heat is comparable. As decay heat at vitrification and repository decay heat affect repository sizing, this may weaken the argument for the Th cycle.

© 2015 Elsevier Ltd.All rights reserved. In this paper the possibility of stimulated self-induced electrostatic fields in radioisotope heat sources for power enhancement is discussed. Because electrons have higher mobility than the positively charged fragments from radioactive decay, a build-up of positive charges can be promoted, leading to an internal induced electrostatic field. This, in turn, results in a repulsive force acting on positive charges, endowing them with additional kinetic energy, and heat release after these charged particles are stopped by inelastic collisions with the boundary wall. Utilizing a simplified geometrical model, an analytical expression for the attainable power enhancement is derived. It is shown that the proposed concept could result in power enhancements of 5-10% for beta sources but enhancements are negligible for alpha sources.

In this paper the utility of using fission gas plenums in a lead-bismuth-cooled fast reactor (LBR) is called into question. It is shown that the primary coolant radiological activity levels due to the generation of 210Po from the neutronic activation of bismuth clearly overshadow the total radiological impact from the fission gas products even if only a tiny fraction of available unbound elementary polonium is considered, and then it does not make any difference if fission gases are collected in a gas plenum or if they are vented directly into the coolant. Moreover, the use of fission gas plenums in a LBR is not only futile from a radiological viewpoint but can neutralize the passive safety mechanism of capture and retention of the majority of polonium captured as lead-polonide (PbPo): the outrush of highly pressurized fission gases during a cladding rupture can act as a spray-like mechanism, and then could vaporize and disperse the retained PbPo in the form of a superfine dust or aerosol. The elimination of fission gas plenums will also result in direct mechanical benefits such as a reduction in pumping power requirements, potentially a smaller reactor vessel and related vessel internal components, deep burnups, a reduction in cladding failures, as well as an important reduction in the amount of high-level waste generated. All these benefits will translate into better reactor management and both social and economic benefits.

A new, alternative approach is proposed for natural convection simulation by means of stimulated thermocapillary currents created by virtual walls. In contrast to the well-known effective thermal conductivity model, in the proposed approach it is the mass motion due to the convective currents which is intended to be simulated and the heat flux is a consequence of such flows. As a result, no a priori knowledge of the Nusselt number is needed and thus the approach is more suitable for complex geometries. Utilizing a simplified physical model and the definition of hydraulic diameter, a generalized expression for enclosed geometries is derived which offers thermal engineers a powerful analysis tool that can use virtual walls with an associated fictitious Marangoni stress for pre-screening and estimation of Nusselt numbers.

The use of zirconium hydride (Th–ZrH1.6) blankets in a thorium-fuelled sodium-cooled reactor for void reactivity control with particular reference to UK's plutonium disposition problem is proposed and considered. It is shown that, with the use of such blankets, a mild moderation effect is produced during voiding which compensates for the general hardening of the spectrum, enabling a net negative void coefficient at pin level to be attained without the need to rely on traditional neutron leakage enhancement techniques or neutron poisons, and with negligible impact on transmutation capabilities. One important difference in comparison with the traditional methods is that the void coefficient is obtained at the pin level, eliminating or mitigating substantially the spatial dependencies on the location of the void. Combining the use of such blankets with a suitable n-batch fuelling scheme yields a negative void reactivity coefficient throughout the life of fuel. Additional research and development are required to explore further this concept's potential.

Multidisciplinary Design Optimization, also known as MDO, deals with the optimization of complex but typically single-objective design problems. As such, it would be valuable to combine it with Multi-Objective Optimization (MOO). Within MOO, Weighted Sums (WS) is a standard solution method, but it may not produce a good spacing of solution points along the Pareto Front (PF). Using our differential geometry MDO framework, we present a technique for intelligently spacing out Pareto solutions produced with WS in a multidisciplinary MOO formulation. Following an initial demonstration, we present modifications to our method to handle ill-conditioning better, to produce fewer dominated points in PF refinement, and to use MOO solution methods other than WS.

This paper considers a linear mass spectrometer, dubbed the cambutron, boosted by a self-induced Hall potential which is able to work with higher pressures than a classical calutron, thereby obtaining a higher throughput and allowing isotopic separation inside an ionization chamber where deuterium and protium have been ionized due to the intense neutronic flux coming from a fusion reactor. The use of a cambutron opens up the possibility of producing significant quantities of deuterium and thus, if it is combined with the use of lithium blankets for tritium breeding, for the in-situ production of all the nuclear fuel needed by a fusion reactor.

Ridge functions have recently emerged as a powerful set of ideas for subspace-based dimension reduction. In this paper we begin by drawing parallels between ridge subspaces, sufficient dimension reduction and active subspaces, contrasting between techniques rooted in statistical regression and those rooted in approximation theory. This sets the stage for our new algorithm that approximates what we call a Gaussian ridge function---the posterior mean of a Gaussian process on a dimension-reducing subspace---suitable for both regression and approximation problems. To compute this subspace we develop an iterative algorithm that alternates between optimizing over the Stiefel manifold to compute the subspace and optimizing the hyperparameters of the Gaussian process. We demonstrate the utility of the algorithm on two analytical functions, where we obtain near exact ridge recovery, and a turbomachinery case study, where we compare the efficacy of our approach with three well-known sufficient dimension reduction methods: SIR, SAVE, and CR. The comparisons motivate the use of the posterior variance as a heuristic for identifying the suitability of a dimension-reducing subspace., The first author's work was supported by a Rolls-Royce fellowship. The second author's work was supported by Magdalene College, Cambridge

To date the design of structures using topology optimization methods has mainly focused on single-objective problems. Since real-world design problems typically involve several different objectives, most of which counteract each other, it is desirable to present the designer with a set of Pareto optimal solutions that capture the trade-off between these objectives, known as a smart Pareto set. Thus far only the weighted sums and global criterion methods have been incorporated into topology optimization problems. Such methods are unable to produce evenly distributed smart Pareto sets. However, recently the smart normal constraint method has been shown to be capable of directly generating smart Pareto sets. Therefore, in the present work, an updated smart Normal Constraint Method is combined with a Bi-directional Evolutionary Structural Optimization (SNC-BESO) algorithm to produce smart Pareto sets for multiobjective topology optimization problems. Two examples are presented, showing that the Pareto solutions found by the SNC-BESO method make up a smart Pareto set. The first example, taken from the literature, shows the benefits of the SNC-BESO method. The second example is an industrial design problem for a micro fluidic mixer. Thus, the problem is multi-physics as well as multiobjective, highlighting the applicability of such methods to real-world problems. The results indicate that the method is capable of producing smart Pareto sets to industrial problems in an effective and efficient manner.

In recent years, a number of authors have studied entropy generation in Wells turbines. This is potentially a very interesting topic, as it can provide important insights into the irreversibilities of the system, as well as a methodology for identifying, and possibly minimizing, the main sources of loss. Unfortunately, the approach used in these studies contains some crude simplifications that lead to a severe underestimation of entropy generation and, more importantly, to misleading conclusions. This paper contains a re-examination of the mechanisms for entropy generation in fluid flow, with a particular emphasis on RANS equations. An appropriate methodology for estimating entropy generation in isolated airfoils and Wells turbines is presented. Results are verified for different flow conditions, and a comparison with theoretical values is presented., Comment: submitted to Meccanica

This paper revisits an old problem of validating computational fluid dynamics simulations with experiments in turbomachinery. The case considered here is NASA rotor 37. Prior computational fluid dynamics studies of this blade have been unable to predict a total pressure deficit at the hub as observed in the experiments. A possible explanation for this discrepancy is a small hub leakage flow emanating fore of the leading edge, between the forward stationary center body and the rotating disk. In this work, a large-scale high-fidelity uncertainty quantification study is carried out to investigate whether this indeed was the case. Computations are carried out on a 4.5-million-cell rotor 37 mesh with a small cavity fore of the leading edge. This cavity has an inlet with three boundary conditions: all assumed to be uncertain. A nonintrusive, orthogonal polynomial-based technique using sparse grids is used to propagate these three uncertainties, namely, leakage mass flow, leakage whirl velocity, and radial flow ...

Recently, the study of micro fluidic devices has gained much interest in various fields from biology to engineering. In the constant development cycle, the need to optimise the topology of the interior of these devices, where there are two or more optimality criteria, is always present. In this work, twin physical situations, whereby optimal fluid mixing in the form of vorticity maximisation is accompanied by the requirement that the casing in which the mixing takes place has the best structural performance in terms of the greatest specific stiffness, are considered. In the steady state of mixing this also means that the stresses in the casing are as uniform as possible, thus giving a desired operating life with minimum weight. The ultimate aim of this research is to couple two key disciplines, fluids and structures, into a topology optimisation framework, which shows fast convergence for multidisciplinary optimisation problems. This is achieved by developing a bi-directional evolutionary structural optimisation algorithm that is directly coupled to the Lattice Boltzmann method, used for simulating the flow in the micro fluidic device, for the objectives of minimum compliance and maximum vorticity. The needs for the exploration of larger design spaces and to produce innovative designs make meta-heuristic algorithms, such as genetic algorithms, particle swarms and Tabu Searches, less efficient for this task. The multidisciplinary topology optimisation framework presented in this article is shown to increase the stiffness of the structure from the datum case and produce physically acceptable designs. Furthermore, the topology optimisation method outperforms a Tabu Search algorithm in designing the baffle to maximise the mixing of the two fluids.

Copyright © 2017 ASME. Two new multi-objective differential evolution (DE) algorithms are used to optimize heterogeneous low-enriched uranium + mixed oxide fuel assemblies for use in a pressurized water reactor. The objectives were to maximize plutonium content and minimize the power peaking factor. A performance comparison to a genetic algorithm is used to evaluate the applicability of DE algorithms to nuclear fuel assembly design optimization problems. Results show that DE performs highly competitively against a more established algorithm and can arguably better represent the trade-off between both objectivesthrough greater variety in the number of different pin arrangements explored and a higher reliability in finding thetrue' Pareto-front.

Turbomachinery active subspace performance maps are 2D contour plots that illustrate the variation of key flow performance metrics with different blade designs. While such maps are easy to construct for design parameterizations with two variables, in this paper maps will be generated for a fan blade with twenty-five design variables. Turbomachinery active subspace performance maps combine active subspaces — a new set of ideas for dimension reduction — with fundamental turbomachinery aerodynamics and design spaces. In this paper, contours of (i) cruise efficiency, (ii) cruise pressure ratio, (iii) maximum climb flow capacity and (iv) sensitivity to manufacturing variations, are plotted as objectives for the fan. These maps are then used to infer pedigree design rules: how best to increase fan efficiency; how best to desensitize blade aerodynamics to the impact of manufacturing variations? In the present study, the former required both a reduction in pressure ratio and flow capacity — leading to a reduction of the strength of the leading edge bow wave — while the latter required strictly a reduction in flow capacity. While such pedigree rules can be obtained from first principles, in this paper these rules are derived from the active subspaces. This facilitates a more detailed quantification of the aerodynamic trade-offs. Thus, instead of simply stating that a particular design is more sensitive to manufacturing variations; or that it lies on a hypothetical ‘efficiency cliff’, this paper seeks to visualize, quantify and make precise such notions of turbomachinery design.Copyright © 2017 by Rolls-Royce plc

Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in reactor physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15% 235U enriched homogeneously mixed all-UO2 fuel and 18% 235U enriched micro–heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13 × 13 arrangement. The eigenvalue/reactivity ( k ∞ ) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k ∞ values was observed among the codes for both the candidate fuels. The differences in k ∞ between the codes are ~ 200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ~ 450 pcm) in the k ∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes.

Analysis within the field of Multidisciplinary Design Optimization (MDO) generally falls under the headings of architecture proofs and sensitivity information manipulation. We propose a differential geometry (DG) framework for further analyzing MDO systems, and here, we outline the theory undergirding that framework: general DG, Riemannian geometry for use in MDO, and the translation of MDO into the language of DG. Calculating the necessary quantities requires only basic sensitivity information (typically from the state equations) and the use of the implicit function theorem. The presence of extra or non-differentiable constraints may limit the use of the framework, however. Ultimately, the language and concepts of DG give us new tools for understanding, evaluating, and developing MDO methods; in Part I, we discuss the use of these tools and in Part II, we provide a specific application.

In this paper, a reduced-moderation LWR (RMLWR) system is compared to a sodium-cooled fast reactor (SFR) as a reference fast system for sustained transuranic burning and self-sufficient break-even cycles using thorium (Th) as the fertile material. SFRs can achieve a higher discharge burn-up (and incineration rate – in the case of burners) than RMLWRs, which results in a highly significant reduction of ∼40% and ∼60% in fuel reprocessing and fabrication requirements for breeder and burner reactors respectively. With sufficiently reduced moderation, the repository radiotoxicity and decay heat of actinide process waste from RMLWRs is very similar to that from SFRs, assuming same fraction of process losses per reprocessed fuel inventory, for the breeder and burner cases. In all cases, fuel fabrication is challenging due to high energy gamma and spontaneous neutron sources. The gamma source is comparable in the burner cores but the spontaneous neutron source in the burner options is approximately an order of magnitude lower for the SFR after a large number of recycles. Th fuel is highly preferable to uranium fuel in RMLWRs due to a large improvement in the void coefficient, which must be kept negative. This constraint makes it difficult, if not impossible, to fuel an RMLWR burner or breeder with uranium. Similarly, the void coefficient in SFRs is significantly better with Th fuel, although it is probably not essential for this to be negative. Although the neutron spectrum and core geometry are substantially different, RMLWRs are a variant of conventional LWR technology. RMLWRs have never been built, whereas some SFRs have been built and operated. Despite this, SFRs are anticipated to have a higher capital cost, which appears to outweigh the lower reprocessing and fuel fabrication requirements, although this conclusion is sensitive to the reactor size, cost and availability factor assumptions.

It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favourable impact on the MTC fostered by use of thorium (Th), the possibility of performing Th–TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. Heterogeneous fuel design with spatial separation of Th–U and Th–TRU is necessary to improve neutronic performance. This can take the form of a heterogeneous fuel assembly (TPUC), or whole assembly heterogeneity (WATU). Satisfactory discharge burn-up can be maintained while ensuring negative MTC, with the pin diameter of a standard PWR increased from 9.5 to 11 mm. However, the reactivity becomes positive when the coolant density in the core becomes extremely low. This could lead to positive reactivity in some loss of coolant accident (LOCA) scenarios, for example a surge line break, if the reactor does not trip. To protect against this beyond design basis accident, a second redundant set of shutdown rods is added to the reactor, so that either the usual or secondary rods can trip the reactor when there is zero coolant in the core. Even so, this condition is likely to be concerning from a regulatory standpoint. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron. Mechanical shim is therefore employed. As a result, it is difficult to maintain adequate core power peaking over the cycle, and careful lattice and core design is required, particularly for the WATU design. An adequate shutdown margin (SDM) can be maintained in both cases, but this requires use of B 4 C rods with enriched Boron-10 in B. A higher Boron-10 enrichment is required for the TPUC fuel design. The response to a rod ejection accident (REA) is significantly worse than in conventional PWRs due to a lower delayed neutron fraction, but fuel enthalpy deposition limits can be satisfied.

In this study, we have sought to determine the advantages, disadvantages, and viability of open cycle thorium–uranium-fuelled (Th–U-fuelled) nuclear energy systems. This has been done by assessing three such systems, each of which requires uranium enriched to ∼20% 235U, in comparison to a reference uranium-fuelled (U-fuelled) system over various performance indicators, spanning material flows, waste composition, economics, and proliferation resistance. The values of these indicators were determined using the UK National Nuclear Laboratory’s fuel cycle modelling code ORION. This code required the results of lattice-physics calculations to model the neutronics of each nuclear energy system, and these were obtained using various nuclear reactor physics codes and burn-up routines. In summary, all three Th–U-fuelled nuclear energy systems required more separative work capacity than the equivalent benchmark U-fuelled system, with larger levelised fuel cycle costs and larger levelised cost of electricity. Although a reduction of ∼6% in the required uranium ore per kWh was seen for one of the Th–U-fuelled systems compared to the reference U-fuelled system, the other two Th–U-fuelled systems required more uranium ore per kWh than the reference. Negligible advantages and disadvantages were observed for the amount and the properties of the spent nuclear fuel (SNF) generated by the systems considered. Two of the Th–U-fuelled systems showed some benefit in terms of proliferation resistance of the SNF generated. Overall, it appears that there is little merit in incorporating thorium into nuclear energy systems operating with open nuclear fuel cycles.

In the original publication of the article, Eqs. (26)–(28) are incorrect due to a missing ρ symbol, inadvertently omitted when reporting the equations in the article.

The Rayleigh density limit and its significance with regard to coalescence and growth of radiative bubbles is discussed. It is shown that taking into account the kind of radioactive decay, i.e. alpha or beta decay, the Rayleigh density limit results in a critical radius limit for the coalescence and growth of radiative bubbles. Because the size of bubbles is one of the most important parameters to be considered in transport and extraction technologies, the present study has important application for the computational modelling of tritium extraction from liquid Pb-16Li blankets for fusion reactor technologies.

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in a degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which ...

Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios, which is desirable to maximize the TRU burning rate. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage TRU burning cycle, where the first stage is Th–Pu MOX in a conventional PWR feeding a second stage continuous burn in RMPWR or RBWR, is technically reasonable, although it is more suitable for the RBWR implementation. In this case, the fuel cycle performance is relatively insensitive to the discharge burn-up of the first stage.

The optimization of a near-circular low-Earth-orbit multispacecraft refueling problem is studied. The refueling sequence, service time, and orbital transfer time are used as design variables, whereas the mean mission completion time and mean propellant consumed by orbital maneuvers are used as design objectives. The J2 term of the Earth’s nonspherical gravity perturbation and the constraints of rendezvous time windows are taken into account. A hybrid-encoding genetic algorithm, which uses normal fitness assignment to find the minimum mean propellant-cost solution and fitness assignment based on the concept of Pareto-optimality to find multi-objective optimal solutions, is presented. The proposed approach is demonstrated for a typical multispacecraft refueling problem. The results show that the proposed approach is effective, and that the J2 perturbation and the time-window constraints have considerable influences on the optimization results. For the problems in which the J2 perturbation is not accounted f...

This paper investigates the effects of design parameters, such as cladding and coolant material choices, and operational phenomena, such as creep and fission product decay heat, on the tolerance of Accelerator Driven Subcritical Reactor (ADSR) fuel pin cladding to beam interruptions. This work aims to provide a greater understanding of the integration between accelerator and nuclear reactor technologies in ADSRs. The results show that an upper limit on cladding operating temperature of 550 °C is appropriate, as higher values of temperature tend to accelerate creep, leading to cladding failure much sooner than anticipated. The effect of fission product decay heat is to reduce significantly the maximum stress developed in the cladding during a beam-trip-induced transient. The potential impact of irradiation damage and the effects of the liquid metal coolant environment on the cladding are discussed.

The Accelerator Driven Subcritical Reactor (ADSR) concept is based on the coupling of a particle accelerator to a subcritical reactor core by means of a neutron spallation target interface. This paper investigates the benefits of multiple spallation targets in ADSRs. The motivation behind this is, firstly, to improve the overall reliability of the accelerator–reactor system, and, secondly, to evaluate other potential advantages such as lower beam power requirements. The results show that a system containing two or three spallation targets, coupled to independent accelerators, offers better neutronic performance. This is demonstrated through the increased effective multiplication factor (keff) in the two- and three-target configurations and a more uniform neutron flux distribution. A multiple-target ADSR also proves effective in mitigating the impact of frequent beam interruptions, a pressing issue that needs to be addressed for the ADSR concept to advance. Assuming no simultaneous beam shutdowns, the two- and three-target configurations reduce the risk of fuel cladding failure due to thermal cyclic fatigue.

The potential for countries that currently have a nominal nuclear energy infrastructure to adopt thorium–uranium-fuelled nuclear energy systems, using a once-through ‘open’ nuclear fuel cycle, has been suggested by the International Atomic Energy Agency. This review paper highlights generation II, III and III+ nuclear energy technologies that could potentially adopt an open thorium–uranium fuel cycle and qualitatively highlights the main differences between the open thorium–uranium and open uranium fuel cycles.

A method for VVER-1000 fuel rearrangement optimization that takes into account both cladding durability and fuel burnup and which is suitable for any regime of normal reactor operation has been established. The main stages involved in solving the problem of fuel rearrangement optimization are discussed in detail. Using the proposed fuel rearrangement efficiency criterion, a simple example VVER-1000 fuel rearrangement optimization problem is solved under deterministic and uncertain conditions. It is shown that the deterministic and robust (in the face of uncertainty) solutions of the rearrangement optimization problem are similar in principle, but the robust solution is, as might be anticipated, more conservative.

© 2016 by the American Institute of Aeronautics and Astronautics, Inc. It is well-known that a nuclear thermal rocket (NTR) cannot be abruptly shut down. After a power manoeuver, the reactor has contaminated itself with fission products and the decay heat released must be removed by maintaining an adequate flow of hydrogen through its passages. The objective of this work was to derive a first estimate of how much hydrogen will be needed to prevent the core from overheating after shutdown, and, from this, be able to assess the advantages of using a dedicated decay heat removal system to reduce or eliminate the amount of hydrogen needed to prevent the core from overheating after shutdown. Furthermore, the use of such a heat removal system could be needed by certain special nuclear thermal propulsion concepts, such as the fission fragment rocket or the more recently proposed pulsed nuclear thermal rocket, where significant amplification of specific impulse, Isp, as well as thrust can be obtained by the direct use of fission fragments or by pulsing the nuclear core, respectively.

Nuclear data consists of measured or evaluated probabilities of various fundamental physical interactions involving the nuclei of atoms and their properties. Most fluoride salt-cooled high-temperature reactor (FHR) studies that were reviewed do not give detailed information on the data libraries used in their assessments. Therefore, the main objective of this data libraries comparison study is to investigate whether there are any significant discrepancies between main data libraries, namely ENDF/B-VII, JEFF-3.1 and JEF-2.2. Knowing the discrepancies, especially its magnitude, is important and relevant for readers as to whether further cautions are necessary for any future verification or validation processes when modelling an FHR. The study is performed using AMEC’s reactor physics software tool, WIMS. The WIMS calculation is simply a 2-D infinite lattice of fuel assembly calculation. The comparison between the data libraries in terms of infinite multiplication factor, kinf and pin power map are presented...

A multi-objective optimization approach was proposed for multiphase orbital rendezvous missions and validated by application to a representative numerical problem. By comparing the Pareto fronts obtained using the proposed method, the relationships between the three objectives considered are revealed, and the influences of other mission parameters, such as the sensors' field of view, can also be analyzed effectively. For multiphase orbital rendezvous missions, the tradeoff relationships between the total velocity increment and the trajectory robustness index as well as between the total velocity increment and the time of flight are obvious and clear. However, the tradeoff relationship between the time of flight and the trajectory robustness index is weak, especially for the four- and five-phase missions examined. The proposed approach could be used to reorganize a stable rendezvous profile for an engineering rendezvous mission, when there is a failure that prevents the completion of the nominal mission.

Plutonium loading in a plutonium–thorium (Pu–Th) mixed oxide (MOX) fuelled pressurized water reactor (PWR) core is typically constrained by large maximum radial form factors (RFF) and positive moderator temperature coefficient (MTC). The large form factors in higher Pu content fuels stem from the large differences in burnup, and thus reactivity, between fresh and burnt fuel, while positive MTC can potentially be the result of the high soluble boron concentrations needed to maintain criticality for such reactive fuel. The conventional solution to these problems is the use of burnable poisons (BPs). While BPs are able to reduce RFF, the positive MTC is not entirely due to a large critical boron concentration (CBC) requirement. In fact, analysis shows a positive MTC in Th–Pu fuel is mainly caused by fissioning in the epithermal–fast energy range. A reduction in epithermal–fast fissioning through the use of certain BPs and the strategic employment of loading patterns that encourage leakage are more effective in attaining negative MTC, as a reduction in CBC has a negligible effect on MTC. This paper examines the contributions to positive MTC by isotope and energy and identifies characteristics of BPs that are able to mitigate positive MTC in a Pu–Th MOX PWR core.

This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of 233 U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, 233 U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, 233 U build-up is slower in the homogeneous case and therefore Pu burning is greater. This 233 U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to burn.