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[en] 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 an 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.
[en] This report describes the background and framework for both organizing the discussion and providing information on the potential for nuclear energy R and 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.
[en] The Liquid-Salt-Cooled Very High Temperature Reactor (LS-VHTR), also known as the Advanced High Temperature Reactor (AHTR), is a new, large [>2400 MW(t)], passively safe, high-temperature reactor concept. It uses a graphite-matrix coated-particle fuel and a graphite moderator similar to the fuel used in modular high-temperature gas-cooled reactors, but with a clean liquid-fluoride salt coolant. The neutronics properties of various salt coolant options are considered with respect to their coefficients of reactivity for various reactor configurations. In addition, several variations on the basic graphite block design of the AHTR are considered that would simplify refueling. The results show that the coolant coefficients of reactivity are negative or very small relative to other reactivity feedbacks, such as the fuel Doppler feedback. This allows several salt-coolants, even some with a positive coolant density coefficient, to be considered for use in the AHTR. In addition, parametric studies of assembly-type clustered rod configurations show that there is minimal impact on the reactivity coefficients and multiplication factors with appropriate cluster design choices. (authors)
[en] The Office of Fuel Cycle Technologies (FCT) of the DOE Office of Nuclear Energy is performing an evaluation and screening of potential fuel cycle options to provide information that can support future research and development decisions based on the more promising fuel cycle options.  A comprehensive set of fuel cycle options are put into evaluation groups based on physics and fuel cycle characteristics. Representative options for each group are then evaluated to provide the quantitative information needed to support the valuation of criteria and metrics used for the study. Included in this set of representative options are Molten Salt Reactors (MSRs), the analysis of which requires several capabilities that are not adequately supported by the current version of SCALE or other neutronics depletion software packages (e.g., continuous online feed and removal of materials). A new analysis approach was developed for MSR analysis using SCALE by taking user-specified MSR parameters and performing a series of SCALE/TRITON calculations to determine the resulting equilibrium operating conditions. This paper provides a detailed description of the new analysis approach, including the modeling equations and radiation transport models used. Results for an MSR fuel cycle option of interest are also provided to demonstrate the application to a relevant problem. The current implementation is through a utility code that uses the two-dimensional (2D) TRITON depletion sequence in SCALE 6.1 but could be readily adapted to three-dimensional (3D) TRITON depletion sequences or other versions of SCALE. (authors)
[en] A computational study has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U3O8 mixed with aluminum. An LEU core design has been obtained and requires an increase in 235U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)
[en] The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E/S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E/S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups and evaluating their performance according to a set of 22 standardized metrics. Following the E/S study, additional studies are being conducted to assess transition period from the current US fuel cycle to future fuel cycle options identified by the E/S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle - the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) - to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle. (authors)
[en] The TSUNAMI computational sequences currently in the SCALE 5 code system provide an automated approach to performing sensitivity and uncertainty analysis for eigenvalue responses, using either one-dimensional discrete ordinates or three-dimensional Monte Carlo methods. This capability has recently been expanded to address eigenvalue-difference responses such as reactivity changes. This paper describes the methodology and presents results obtained for an example advanced CANDU reactor design. (authors)
[en] The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from spent nuclear fuel. The proposed fuels include metal and oxide forms mixed actinides (U-Np-Pu-Am-Cm) as well as target concepts with perhaps both Am-Cm. The High Flux Isotope Reactor (HFIR) was built for the purpose of transmuting plutonium to various higher actinides including Am, Cm, and Cf Since a fast-spectrum irradiation facility does not exist in the United States, HFIR can fulfill a first step in the GNEP- mission that being to establish a near-term domestic capability to irradiate materials in a fast neutron spectrum. Modifications to the HFIR central target region to accomplish this goal are described. A second ongoing project for HFIR is to design capsules and installation tools and procedures to irradiate short rods of innovative nuclear fuel types and cladding materials under prototypic light water reactor (LWR) operating conditions at an accelerated rate relative to expected reactor performance. This second proposal would be for a facility representative of thermal reactor conditions rather than the GNEP concept. In order to maintain power densities within the fuel at levels normally seen by LWR reactors, an entirely new experiment and test capsule design will be needed. (authors)
[en] This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)
[en] This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)