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Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors and the Experimental Breeder Reactor-II for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor concepts. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output. It would use tristructural-isotropic (TRISO) particle fuel in compacts within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is an intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include: • core design methodologies, • heat exchanger performance (including passive decay heat removal), • pump performance, • reactivity control, • salt chemistry control to maximize plant life, • salt procurement, handling, maintenance and ultimate disposal, and • tritium management. Non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each ...

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capa...

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride based or chloride based, as either a coolant with a solid fuel (such as fluoride salt–cooled high-te...

Un objectif cle de la feuille de route du DOE (US Department of Energy) dans le domaine de la recherche et du developpement (R&D) pour l'energie nucleaire est le developpement de cycles de combustibles nucleaires durables qui peuvent ameliorer l'utilisation des ressources naturelles et offrir des solutions perennes a la gestion des dechets nucleaires. Une etude recente d'evaluation des differents cycles du combustible a identifie les options offrant les meilleures opportunites pour parvenir a ces ameliorations et identifie les activites de R&D necessaires pour soutenir le developpement de cycles du combustible avances.

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 105 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

This study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR l...

The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

An investigation of the utilization of TRistructural-ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized wa...

A thermal-hydraulics (TH)/neutronics/crud multiphysics coupling framework to simulate the crud deposits’ impact on crud-induced power shift (CIPS) phenomenon is proposed in this paper. The coupling...

A study on the Fukushima Daiichi nuclear power station spent-fuel pool (SFP) at Unit 4 (SFP4) is presented in this paper. We discuss the design characteristics of SFP4 and its decay heat load in detail and provide a model that we developed to estimate the SFP evaporation rate based on the SFP temperature. The SFP level of SFP4 following the March 11, 2011, accident is predicted based on the fundamental conservation laws of mass and energy. Our predicted SFP level and temperatures are in good agreement with measured data and are consistent with Tokyo Electric Power Company evaluation results.

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includesmore » information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.« less

The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major “Challenge Problems” in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

Solutions to the OECD/NEA three-dimensional fuel assembly benchmark problem, C5G7MOX, obtained using the discrete-ordinates neutron transport code TORT are presented. The accuracy and convergence of the effective multiplication factor and maximum pin power are examined while varying the fidelity of the geometric model and the order of the angular quadrature, specifically the Square Legendre–Chebychev quadrature. The calculated value of the effective multiplication factor is shown to converge asymptotically as the order of the quadrature is increased. The behavior of that asymptotic limit of k eff with decreasing computational cell size and increasing staircase resolution of the curved rod-moderator is less converged. Moreover, rapidly increasing computational cost with geometric configuration refinement prevented us from achieving asymptotic convergence with the spatial approximation. The maximum pin power results are generally shown to asymptotically converge to within the statistical error of the reference solution for the problem slices where control rods are not present. For the zones that contain control rods, the solution does not appear to be tightly converged for the meshes and quadrature orders employed.

For many years various countries with access to commercial reprocessing services have been routinely recycling plutonium as UO2/PuO 2 mixed oxide (MOX) fuel in light water reactors (LWRs). This LWR MOX recycle strategy is still widely regarded as an interim step leading to the eventual establishment of sustainable fast reactor fuel cycles. The OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR) has recently completed a review of the technical options for plutonium management in what it refers to as the "medium terra." For the purpose of the review, the WPPR considers the medium term to cover the period from now up to the point at which fast reactor fuel cycles are established on a commercial scale. The review identified a number of different designs of innovative plutonium fuel assemblies intended to be used in current LWR cores, in LWRs with significantly different moderation properties, as well as in high-temperature gas reactors. The full review report describes these various options and highlights their respective advantages and disadvantages. This paper briefly summarizes the main findings of the review.

We present a comprehensive study of the solution error for the C5G7MOXBenchmark problem using the two-dimensional transport code DORT. We set a stringent criterion that a successful solution to the benchmark exercise must satisfy, namely that at least for some of the benchmark quantities convergence to the reference solution with computational model refinement must be demonstrated. In the present exercise this amounts to examining the evolution of the DORT solution, e.g. the multiplication factor, with increasing angular quadrature order, decreasing computational cell size, and tighter representation of the curved rod-moderator interface for all circular rods in the core model. In addition we explored the effect of angular quadrature type, comparing solution accuracy for the fully symmetric and the Square Legendre-Chebychev quadratures establishing superiority of the latter. We establish the high quality of DORT's solution to this benchmark exercise by demonstrating that the multiplication factor asymptotically approaches the reference solution as stated in our self-imposed success criterion. High accuracy of the pin power distribution is also attained, however, convergence to the reference values is not realized. We conjecture that this is due to the lack of error information in the reference values. Our results also illustrate DORT's high accuracy on reasonable meshes and quadrature orders, as well as the sufficiency of a crude, square, geometric approximation of the rod-moderator interface to achieve high accuracy.

The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (

During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

Fully ceramic microencapsulated (FCM) fuels are identified as suitable fuel forms for development and deployment of SMR technologies. High thermal conductivity of the composite fuel coupled with its exceptional radiation and environmental stability enable monolithic compact cores for SMR applications. Meanwhile the built-in redundancy of multiple layers for fission-product retention alleviates the need for structurally demanding pressure vessels and containment buildings. All of the above will enable the reactor designer to achieve a compact and passively safe SMR through utilization of a flexible fuel technology that is currently under active development.Copyright © 2011 by ASME

During FY 2010, five studies were conducted of technology families’ applicability to various fuel cycle strategies to fill in knowledge gaps in option space and to better understand trends and patterns. Here, a “technology family” is considered to be defined by a type of reactor and by selection of which actinides provide fuel. This report summarizes the higher-level findings; the detailed analyses and results are documented in five individual reports, as follows: • Advanced once through with uranium fuel in fast reactors (SFR), • Advanced once through (uranium fuel) or single recycle (TRU fuel) in high temperature gas cooled reactors (HTGR), • Sustained recycle with Th/U-233 in light water reactors (LWRs), • Sustained recycle with Th/U-233 in molten salt reactors (MSR), and • Several fuel cycle missions with Fusion-Fission Hybrid (FFH). Each study examined how the designated technology family could serve one or more designated fuel cycle missions, filling in gaps in overall option space. Each study contains one or more illustrative cases that show how the technology family could be used to meet a fuel cycle mission, as well as broader information on the technology family such as other potential fuel cycle missions for which insufficient information was availablemore » to include with an illustrative case. None of the illustrative cases can be considered as a reference, baseline, or nominal set of parameters for judging performance; the assessments were designed to assess areas of option space and were not meant to be optimized. There is no implication that any of the cases or technology families are necessarily the best way to meet a given fuel cycle mission. The studies provide five examples of 1-year fuel cycle assessments of technology families. There is reasonable coverage in the five studies of the performance areas of waste management and uranium utilization. The coverage of economics, safety, and proliferation resistance and physical protection in the five studies was spotty. Some studies did not have existing or past work to draw on in one or more of these areas. Resource constraints limited the amount of new analyses that could be performed. Little or no assessment was done of how soon any of the technologies could be deployed and therefore how quickly they could impact domestic or international fuel cycle performance. There were six common R&D needs, such as the value of advanced fuels, cladding, coating, and structure that would survive high neutron fluence. When a technology family is considered for use in a new fuel cycle mission, fuel cycle performance characteristics are dependent on both the design choices and the fuel cycle approach. For example, the use of the sodium-cooled fast reactor to provide recycle in either breeder or burner mode has been studied for decades, but the SFR could be considered for once-through fuel cycle with the physical reactor design and fuel management parameters changed. In addition, the sustained recycle with Th/U-233 in LWR could be achieved with a heterogeneous assembly and derated power density. Therefore, it may or may not be adjustable for other fuel cycle missions although a reactor intended for one fuel cycle mission is built. Simple parameter adjustment in applying a technology family to a new fuel cycle mission should be avoided and, if observed, the results viewed with caution.« less

The Options Study has been conducted for the purpose of evaluating the potential of alternative integrated nuclear fuel cycle options to favorably address the issues associated with a continuing or expanding use of nuclear power in the United States. The study produced information that can be used to inform decisions identifying potential directions for research and development on such fuel cycle options. An integrated nuclear fuel cycle option is defined in this study as including all aspects of the entire nuclear fuel cycle, from obtaining natural resources for fuel to the ultimate disposal of used nuclear fuel (UNF) or radioactive wastes. Issues such as nuclear waste management, especially the increasing inventory of used nuclear fuel, the current uncertainty about used fuel disposal, and the risk of nuclear weapons proliferation have contributed to the reluctance to expand the use of nuclear power, even though it is recognized that nuclear power is a safe and reliable method of producing electricity. In this Options Study, current, evolutionary, and revolutionary nuclear energy options were all considered, including the use of uranium and thorium, and both once-through and recycle approaches. Available information has been collected and reviewed in order to evaluate the ability of anmore » option to clearly address the challenges associated with the current implementation and potential expansion of commercial nuclear power in the United States. This Options Study is a comprehensive consideration and review of fuel cycle and technology options, including those for disposal, and is not constrained by any limitations that may be imposed by economics, technical maturity, past policy, or speculated future conditions. This Phase II report is intended to be used in conjunction with the Phase I report, and much information in that report is not repeated here, although some information has been updated to reflect recent developments. The focus in this Options Study was to identify any nuclear fuel cycle technology or option that may result in a significant beneficial impact to the issues as compared to the current U.S. approach of once-through use of nuclear fuel in LWRs or similar reactors followed by direct disposal of UNF. This approach was taken because incremental differences may be difficult to clearly identify and justify due to the large uncertainties that can be associated with the specific causes of the issues. Phase II of this Options Study continued the review of nuclear fuel cycle options that was initiated and documented during Phase I, concentrating on reviewing and summarizing the potential of integrated nuclear fuel cycles. However, based on the reviews of previous studies and available data, it was not always possible to clearly determine sufficiently large differences between the various fuel cycle and technology options for some of the issues or evaluation measures, for example, in cases where only incremental differences with respect to the issues might be achieved regardless of the fuel cycle option or technologies being considered, or where differences were insufficient to clearly rise above the uncertainties.« less

This report describes the background and framework for both organizing the discussion and providing information on the potential for nuclear energy R&D to develop alternative nuclear fuel cycles that would address the issues with the current implementations of nuclear power, including nuclear waste disposal, proliferation risk, safety, security, economics, and sustainability. The disposition of used fuel is the cause of many of the concerns, and the possible approaches to used fuel management identify a number of basic technology areas that need to be considered. The basic science in each of the technology areas is discussed, emphasizing what science is currently available, where scientific knowledge may be insufficient, and especially to identify specific areas where transformational discoveries may allow achievement of performance goals not currently attainable. These discussions lead to the wide range of technical options that have been the basis for past and current research and development on advanced nuclear fuel cycles in the United States. The results of this work are then briefly reviewed to show the extent to which such approaches are capable of addressing the issues with nuclear power, the potential for moving further, and the inherent limitations.

A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation andmore » accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).« less

This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with amore » modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.« less

As an alternative to uranium as a fuel for reactors, thorium has been debated and evaluated since the dawn of the nuclear era.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1992.

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.