Nuclear Thermal Propulsion (NTP) systems are actively being developed for future crewed missions to Mars. NTP systems excel in missions where both high thrust and high specific impulse are required, but modern NTP systems currently do not have a Technology Readiness Level (TRL) high enough for use in crewed space exploration. TRLs are used to demonstrate the level of rigor with which a component/system has been tested/demonstrated for its intended use. While space systems technology in general must be qualified as a unit, nuclear technology must be first demonstrated to meet qualification level requirements both at the fuel (component) level and the reactor (subsystem) level. In this paper, historic NTP development programs are surveyed to identify a testing and development strategy that can be effectively implemented to allow for NTP reactor development. Based on this strategy, required facilities to enable such activities are identified. Current domestic experimental capabilities to support NTP qualification are limited to separate effects testing of individual components. Separate effects testing is found extensively in historic NTP development efforts but is not sufficient for full fuel and reactor qualification. Combined effects testing allows for an accurate assessment of fuel performance but is not achievable for NTP conditions in existing facilities. Assessment of historic development programs suggests that an intermediate, subscale test facility is necessary to advance NTP TRLs. A solution to meet this need is proposed, namely the Subscale Maturation of Advanced Reactor Technologies (SMART) facility. SMART will mitigate risk to NTP development by enabling performance and reactor physics demonstrations of NTP subsystems. A SMART facility could be built by modifying existing nuclear test facilities, which may potentially enable schedule and cost savings. To pursue reactor qualification beyond the subscale, a new ground test facility will be necessary. This ground test facility should be developed concurrently with SMART to allow for the facility to be operational in time for expedited NTP engine demonstration.

Various nuclear fuel cycle options have been proposed by developers of advanced nuclear energy systems. Motivations for these options, compared to the current light-water-cooled reactors used in the United States, have included more efficient use of resources, waste reduction, greater safety and security, and lower costs. To provide indications on which of the fuel cycle options are promising for future use in order to yield significant benefits in performance and to identify potential challenges, an assessment of options has been conducted using a consistent set of criteria and associated metrics. Findings of the comprehensive assessment of nuclear fuel cycle options are summarized in this chapter. R&D requirements for implementing the most promising options are identified.

The demand for more practical and innovative nuclear reactor designs capable of providing clean energy to meet new global demands is revolutionizing the nuclear industry. In this paper we present the fuel cycle and neutronics analysis of a design expected to be similar to the Westinghouse eVincim (TM) heat pipe (HP) reactor. The concept considered here uses low enriched urania rods and potassium liquid metal cooled HPs to remove heat from the core. The fuel and HPs are contained in steel monolith to form the core which is surrounded by an alumina reflector. There is no need for forced circulation to remove heat from the active core region, thus eliminating the use of pumps, valves and tube piping. The concept is designed to operate safely and provide a reliable autonomous power supply in support of off-grid missions. This study shows that the HP reactor design can operate for more than 10 years without refueling. Further, we show that the current design of the HP reactor concept is best suited to serve as a nuclear battery rather than a centralized power source. We also studied the nuclear fuel cycle performance of the HP reactor concept in a once-through fuel cycle. The natural resource utilization, waste output, and environmental impact, as defined in a reference study from the literature, did not perform as well on a GWe per year energy basis compared to light water reactors with less than 5% enriched uranium in a once-through fuel cycle. Parasitic absorption in the steel monolith structure and neutron leakage lower the reactivity of the core, which decreases the discharge burnup of the fuel. Analysis of modifications of the monolith material and size of the reference core configuration showed that the neutron leakage impact on the small sized HP core is the most limiting factor on the fuel cycle performance. (C) 2018 Elsevier Ltd. All rights reserved.

This paper discusses modification of the U.S. Department of Energyu0027s (DOE) Technology Readiness Assessment Guide (DOE G 413.3-4A) for use in the Office of Nuclear Energyu0027s Nuclear Technology Research and Development (NTRD) program. DOE 413.3-4A describes an approach for identifying and managing technology risks in large construction projects to ensure technology development is part of the project baseline and technology maturity is sufficiently advanced before including a technology in the facility design. The process includes the identification of Critical Technology Elements (CTEs), followed by in-depth evaluation of Technology Readiness Levels (TRLs) and development of Technology Maturation Plans (TMPs) for those CTEs. This approach was augmented for the NTRD program to include System Readiness Levels (SRLs) as a complement to TRLs, and developing perspective tracks to better organize the TRL/SRL evaluations that determine the level of knowledge/maturation of the CTEs (e.g. Experiments u0026 Modeling, Safety u0026 Hazards, Manufacturing, etc.). The augmented approach was tested on two areas of research in 2017 (metallic nuclear fuels and fuel dissolution for recycling) and is now being applied to a complete advanced nuclear fuel cycle (from uranium mining to waste disposal).

The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects.Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC). (C) 2016 Elsevier B.V. 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.

Dynamic fuel cycle simulation tools are intended to model holistic transient nuclear fuel cycle scenarios. As with all simulation tools, fuel cycle simulators require verification through unit tests, benchmark cases, and integral tests. Model validation is a vital aspect, as well. Although comparative studies have been performed, there is no comprehensive unit test and benchmark library for fuel cycle simulator tools. The objective of this paper is to identify some of the "must test" functionalities of a fuel cycle simulator tool within the context of specific problems of interest to the Fuel Cycle Options Campaign within the U.S. Department of Energy's Office of Nuclear Energy (DOE-NE). This paper identifies the features needed to cover the range of promising fuel cycle options identified in the DOE-NE Fuel Cycle Evaluation and Screening and categorizes these features to facilitate prioritization. Features are categorized as essential functions, integrating features, and exemplary capabilities. A library of unit tests applicable to each of the essential functions should be developed as future work. An international dialog on the functionalities and standard test methods for fuel cycle simulator tools is encouraged. (C) 2016 Elsevier Ltd. All rights reserved.

Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly U-233) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material.These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission hybrid systems per form better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

The safe, reliable, and economic operation of the nation's nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry's success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATE in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop All, concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.

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 10(5) 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. (C) 2015 Elsevier B.V. All rights reserved.

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to N-14 content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation. (C) 2015 Elsevier B.V. All rights reserved.