There is a growing sociopolitical desire to develop cleaner energy sources in the United States and maintain energy security. Regardless of politics, many coal-fired electric plants have already been shut down and many utilities are vowing to retire their current coal-fired assets within the next two decades. Replacement power assets require consideration of appropriate siting. A geographic information system (GIS)-based multicriteria decision analysis approach is useful to assist utility and energy companies, as well as policymakers, to evaluate potential areas for siting new plants in the contiguous United States. A GIS-based framework is simply a database of location information that allows for mapping, querying, modeling, and analyzing data based on location. The spatial output can be structured to be visual, allowing for easier analysis of location data. The need to site additional power assets, including renewable resources and clean power sources, such as nuclear, led to the development of the Oak Ridge Siting Analysis for power Generation Expansion (OR-SAGE) tool discussed in this paper. The tool takes inputs such as population growth, water availability, environmental indicators, and tectonic and geological hazards to provide an in-depth visual analysis for siting options. Energy companies and other stakeholders can use OR-SAGE to procure feedback quickly and effectively on land suitability based on technology specific inputs. Policymakers can use OR-SAGE to analyze the impacts of future energy technology decisions, while balancing competing resource use. This paper discusses the recent use of OR-SAGE for these purposes and plans for future development.

Nuclear design of a reactor system involves the successful demonstration and optimization of safety, economics, reliability, and operations. In order to achieve these objectives, iPWRs share many of the same nuclear design principles as any other reactor types, in particular those of the large PWRs. This chapter describes (i) the important safety design criteria and principles in nuclear design of reactors, and in particular, iPWRs, (ii) what design features are used to achieve a viable and economic nuclear design, and (iii) how the iPWR designers and vendors have addressed the design principles and features in their respective reactors.

assessment of the detection research and development (R&D;) necessary to transition the current safeguards technology toolkit to meet the verification needs of thorium fuel cycles, and to formulate the scientific basis for building new instrumentation to fill any potential capability gaps. The purpose of this technology road map is to define and inform on the safeguards technology needs for thorium fuel cycles, reflecting only the leading candidate thorium fuel cycles prioritized based on implementation timescales and the current direction of international programs. This work provides a guide to the priorities for future directed R&D; needed to bring the technology readiness levels (TRLs) of safeguards detection solutions in line with the higher TRLs of the most promising thorium fuel cycles. Herein, a summary is presented of thorium fuel cycle options and activities currently under way worldwide, which provides the technical basis for the needs evaluation. The key synergies and differences between these fuel cycles and current conventional uranium- and plutonium-based fuel cycles will be discussed from a nuclear material accountancy and detection standpoint. Reactor inventory calculations, fuel cycle simulations, and conclusions from the safeguards technology needs assessment will be presented, together with plans for experimental validation.

This report summarizes the application of the OR-SAGE tool to the potential backfit of a small modular reactor (SMR) or an advanced non–light-water reactor (non-LWR) at a current or former TVA coal plant site to serve as an example in support of the DOE-NE SA&I Campaign. The OR-SAGE development funded by this campaign has led to additional development and application of the tool.

This paper presents the major takeaways from studies conducted over several years that were focused on transitioning the U.S. nuclear infrastructure from the current once-through fuel cycle to one in which fuel is continuously recycled in fast reactors. These studies involved simulating and analyzing numerous example scenarios of fuel cycle transition with various assumptions on technology, policy, and material utilization strategies. Among the many findings, perhaps the most important is that under certain conditions, the use of high-assay low-enriched uranium to start up a fleet of fast reactors may be more favorable compared to using recycled Pu from thermal reactors since it is less constrained by other technologies and may even be more economical. (C) 2020 Elsevier Ltd. All rights reserved.

Oak Ridge National Laboratory (ORNL) has identified a number of previously reported1,2 technical factors impacting the implementation of safeguards for Molten Salt Reactors (MSRs), which include: (1) the homogeneous mixture of fuel, coolant, fission products (FPs), and actinides; (2) continuous variation of isotopic concentrations in the fuel salt, including removal (passive or active) of FPs, rare earth elements, and noble metals; (3) the potential for online reprocessing whereby some fraction of the inventory can be removed while the reactor is operational; (4) unique refueling schemes, including the ability to continuously feed the core with fresh fissile or fertile material; and (5) the need for measurements to occur in high-radiation, high-dose, high-temperature environments. These factors necessitate the use of advanced modeling and simulation for tracking the isotopic masses and signatures (i.e., chemical, elemental, isotopic, and radiation) throughout the reactor and associated auxiliary processing, and importantly, tracking must be accomplished as a function of time as the fuel salt evolves during reactor and fuel cycle operations. Determining what needs to be measured, what can be observed, and where to make the measurement(s) are the first steps towards the development of the safeguards technology for the MSR family of reactors. This paper presents insight into the necessity of advanced modeling and simulation methods and tools, highlighting the tight coupling between the reactor and fuel cycle operations and the resulting fuel inventory and associated signatures. This paper also demonstrates the importance of a comprehensive understanding of the MSR and fuel cycle technologies, as well as the safeguards approaches and technologies that need to be applied. Using ORNLdeveloped tools designed to model dynamic, complex systems such as salt-fueled MSRs (e.g., source term accountancy), several scenarios are presented that demonstrate the data and modeling fidelity required and highlights the physical behavior of the critical factors identified above, illustrating the need to capture the tight coupling between MSR behavior and safeguards assessments. A preliminary evaluation of the implications for safeguards technology development is also presented.

The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 x 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% U-235 throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel center-line temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory. Published by Elsevier Ltd.

Nuclear fuel cycle simulators (NFCSs) are fundamental in guiding policy and economic decisions regarding nuclear fuel cycle (NFC) options. This paper introduces a new method to predict the mixed oxide fuel (MOX) criticality value using an artificial neural network (ANN) model, while most current NFCSs use simple MOX fabrication estimations that do not account for burnup effects. The authors generated over one million depletion simulation results of MOX fuel with varying plutonium vectors and plutonium content to train an ANN network to predict the fuel’s Beginning of Cycle (BOC) and End of Cycle (EOC) criticality. Results show that the trained ANN can predict criticality of MOX fuel within 1% error compared with the test data. The trained ANN is implemented into Cyclus, an agent-based NFC, to demonstrate capabilities to model dynamic reactor behavior such as reactor power and incoming plutonium vector variation. This paper concludes with the discussion of the shortcomings of the ANN approach and potential ways to mitigate them.

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. A solid-fueled reactor only discharges a fraction of a core every year. (C) 2019 Elsevier Ltd. All rights reserved.

The fuel cycle performance and core design of the Transatomic Power liquid -fueled molten salt reactor concept is analyzed. This advanced reactor concept uses configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from intermediate at beginning of life to thermal at end of life. With a harder spectrum during the early years of reactor operation, this spectral shift design drives captures in fertile U-238. The converted fissile plutonium makes up over 50% of the fissile material in the fuel salt over the last half (similar to 15 years) of reactor operation. A softer spectrum late in reactor life helps drive the fuel to a burnup of 90 GWd/MTU. Continuously changing physics necessitates time-dependent analyses resolved over long timescales (i.e., months to years), as this concept does not meet an equilibrium condition. The spectral shift and molten salt reactor material feeds and removals enable this concept to perform better in fuel cycle metrics, increasing resource utilization by more than 50% compared with a typical light water reactor (i.e., from 0.6% to 1%). These metrics are compared to similar fuel cycles using alternate technologies. Additional core design and analysis challenges associated with the spectral shift and use of molten salt reactor technology are identified and discussed. (C) 2018 Elsevier Ltd. All rights reserved.

Certain characteristics of heavy water reactors (HWRs), such as a more flexible neutron economy compared to light water (due to reduced absorptions in hydrogen), online refueling capability, and having a thermal neutron spectrum, make them potentially attractive for use with a thorium fuel cycle. Three options that combine HWRs with thorium-based fuels are considered in this paper: a Near-Term option with minimal advanced technology requirements, an Actinide Management option that incorporates the recycle of minor actinides (MAs), and a Thorium-Only option that uses two reactor stages to breed and consume 233U, respectively. Simplified, steady-state simulations and corresponding material flow analyses are used to elucidate the properties of these fuel cycle options. The Near-Term option begins with a low-enriched uranium oxide pressurized water reactor (PWR) that discharges spent nuclear fuel, from which uranium and plutonium are recovered to fabricate the driver fuel for an HWR that uses thorium oxide as a blanket fuel. This option uses 28% less natural uranium (NU) and sends 33% less plutonium to disposal than the conventional once-through uranium fuel cycle on an energy-normalized basis. The ActinideManagement option also uses spent nuclear fuel from a PWR using enriched uranium oxide fuel (both a low-and high-enrichment variant are considered), but the uranium is recycled for reuse in the PWR while the plutonium andMAs are recycled and used in conjunction with thorium in an HWR with full recycle. Both enrichment variants of this option achieve a more than 95% reduction in transuranic actinide disposal rates compared to the once-through option and a more than 60% reduction compared to closed transuranic recycle in a uranium-plutonium-fueled sodium fast reactor. The Thorium-Only option breeds a surplus of 233U in a thorium-based HWR to supply fissile material to a high-temperature gas-cooled reactor, both of which recycle uranium and thorium. This option requires no NU and produces few transuranic actinides at steady state, although it would require a greater technology maturation effort than the other options studied. Collectively, the options considered in this study are intended to illustrate the range of operational missions that could be supported by fleets that integrate thorium and HWRs.

The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify results from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.

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 (LiF)-Li-7-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, andtritium 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. (C) 2016 Elsevier Ltd. All rights reserved.

Small modular reactors (SMRs) may offer potential benefits relative to large light water reactors, such as enhanced flexibility in deployment and operation. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary neutronics and fuel cycle analyses conducted in support of the US Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a "cartridge type" one-batch core with slightly less than 5.0% enrichment.The high-level issues identified and preliminary scoping calculations in this paper are intended to inform decision makers regarding potential fuel cycle impacts of one-batch thermal-spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and a reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for the simplified example SMR analyzed herein are compared with those for a conventional three-batch light water reactor (LWR) in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of per energy generated, mass of depleted uranium disposed of per energy generated, land use per energy generated, and carbon emissions per energy generated.Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores, which mean high leakage (there is a radial and an axial component); (2) a heterogeneous core and extensive use of control rods and burnable poisons; and (3) single-batch cores. But not all SMR designs have all of these traits. The approach used in this study is an example bounding case, and not all SMRs may be impacted to the same extent. (C) 2016 Elsevier Ltd. All rights reserved.