Molten Salt Fuel Cycle Requirements for ADTT Applications

The operation of an ADT system with the associated nuclear reactions has a profound effect upon the chemistry of the fuel - especially with regards to container compatibility and the chemical separations that may be required. The container can be protected by maintaining the redox chemistry within a relatively narrow, non-corrosive window. Neutron economy as well as other factors require a sophisticated regime of fission product separations. Neither of these control requirements has been demonstrated on the scale or degree of sophistication necessary to support an ADT device. We review the present situation with respect to fluoride salts, and focus on the critical issues in these areas which must be addressed. One requirement for advancement in this area - a supply of suitable materials - will soon be fulfilled by the remediation of ORNL�s Molten Salt Reactor Experiment, and the removal of a total of 11,000 kg of enriched (Li-7 > 99.9%) coolant, flush, and fuel salts

Molten fluoride fuel salt chemistry

The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior which can be used effectively to reduce the amount of development required for future systems, some significant molten salt chemical questions must still be addressed

Review of ORNL`s MSR technology and status

The current status of molten salt reactor development is discussed with reference to the experience from the Oak Ridge Molten Salt Reactor Experiment. Assessment of the future for this reactor system is reviewed with consideration of both advantages and disadvantages. Application of this concept to ADTT (accelerator driven transmutation technology) needs appears to be feasible by drawing on the MSRE experience. Key chemical considerations remain as: solubility, redox behavior, and chemical activity and their importance to ADTT planning is briefly explained. Priorities in the future development of molten salts for these applications are listed, with the foremost being the acceptance of the 2LiF-BeF{sub 2} solvent system. 8 refs, 2 figs

Some Investigations of the Reaction of Activated Charcoal with Fluorine and Uranium Hexafluoride

The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has been shut down since 1969, when the fuel salt was drained from the core into two Hastelloy N drain tanks at the reactor site. Over time, fluorine (F{sub 2}) and uranium hexafluoride (UF{sub 6}) moved from the salt through the gas piping to a charcoal bed, where they reacted with the activated charcoal. Some of the immediate concerns related to the migration of F{sub 2} and UF{sub 6} to the charcoal bed were the possibility of explosive reactions between the charcoal and F{sub 2}, the existence of conditions that could induce a criticality accident, and the removal and recovery of the fissile uranium from the charcoal. This report addresses the reactions and reactivity of species produced by the reaction of fluorine and activated charcoal and between charcoal and F{sub 2}-UF{sub 6} gas mixtures in order to support remediation of the MSRE auxiliary charcoal bed (ACB) and the recovery of the fissile uranium. The chemical identity, stoichiometry, thermochemistry, and potential for explosive decomposition of the primary reaction product, fluorinated charcoal, was determined

TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the technologies used in processing the fuel. Potential processing methods will be identified both by a review of the literature regarding the processing of similar fuels and by a reliance on the experience and innovation of the authors. The objective is not to generate an exhaustive list of options but rather to identify a number of potentially practical processing options. These options necessarily take into consideration the chemical characteristics of the entire fuel element and its component parts. Once the practical options are identified, a qualitative assessment of the technical merit and maturity, relative costs, and relative quantity of waste generation will be used to rank the various options. Through this form of analysis, a base-case flow sheet will be identified for further study and development. A fallback flow sheet will also be selected to reduce the overall technical risk of the development plan. To support the base-case flow sheet, a technical development plan will be used to identify the key issues for the highest-rated option(s). In this effort the technical uncertainties will be more fully articulated, and research and development activities will be recommended to reduce the technical risks

Fuel Fabrication for Surrogate Sphere-Pac Rodlet

Sphere-pac fuel consists of a blend of spheres of two or three different size fractions contained in a fuel rod. The smear density of the sphere-pac fuel column can be adjusted to the values obtained for light-water reactor (LWR) pellets (91-95%) by using three size fractions, and to values typical of the fast-reactor oxide fuel column ({approx}85%) by using two size fractions. For optimum binary packing, the diameters of the two sphere fractions must differ by at least a factor of 7 (ref. 3). Blending of spheres with smaller-diameter ratios results in difficult blending, nonuniform loading, and lower packing fractions. A mixture of about 70 vol% coarse spheres and 30 vol% fine spheres is needed to obtain high packing fractions. The limiting smear density for binary packing is 86%, with about 82% achieved in practice. Ternary packing provides greater smear densities, with theoretical values ranging from 93 to 95%. Sphere-pac technology was developed in the 1960-1990 period for thermal and fast spectrum reactors of nearly all types (U-Th and U-Pu fuel cycles, oxide and carbide fuels), but development of this technology was most strongly motivated by the need for remote fabrication in the thorium fuel cycle. The application to LWR fuels as part of the DOE Fuel Performance Improvement Program did not result in commercial deployment for a number of reasons, but the relatively low production cost of existing UO{sub 2} pellet fuel is probably the most important factor. In the case of transmutation fuels, however, sphere-pac technology has the potential to be a lower-cost alternative while also offering great flexibility in tailoring the fuel elements to match the exact requirements of any particular reactor core at any given time in the cycle. In fact, the blend of spheres can be adjusted to offer a different composition for each fuel pin or group of pins in a given fuel element. Moreover, it can even provide a vertical gradient of composition in a single fuel pin. For minor-actinide-bearing fuels, the sphere-pac form is likely to accept the large helium release from {sup 241}Am transmutation with less difficulty than pellet forms and is especially well suited to remote fabrication as a dustless fuel form that requires a minimum number of mechanical operations. The sphere-pac (and vi-pac) fuel forms are being explored for use as a plutonium-burning fuel by the European Community, the Russian Federation, and Japan. Sphere-pac technology supports flexibility in the design and fabrication of fuels. For example, the blend composition can be any combination of fissile, fertile, transmutation, and inert components. Since the blend of spheres can be used to fill any geometric form, nonconventional fuel geometries (e.g., annular fuels rods, or annular pellets with the central region filled with spheres) are readily fabricated using sphere-pac loading methods. A project, sponsored by the U.S. Department of Energy Advanced Fuel Cycle Initiative (AFCI), has been initiated at Oak Ridge National Laboratory (ORNL) with the objective of conducting the research and development necessary to evaluate sphere-pac fuel for transmutation in thermal and fast-spectrum reactors. This AFCI work is unique in that it targets minor actinide transmutation and explores the use of a resin-loading technology for the fabrication of the remote-handled minor actinide fraction. While there are extensive data on sphere-pac fuel performance for both thermal-spectrum and fast-spectrum reactors, there are few data with respect to their use as a transmutation fuel. The sphere-pac fuels developed will be tested as part of the AFCI LWR-2 irradiations. This report provides a review of development efforts related to the fabrication of a sphere-pac rodlet containing surrogate fuel materials. The eventual goal of this activity is to develop a robust process that can be used to fabricate fuels or targets containing americium. The report also provides a review of the materials, methods, and techniques to be used in the fabrication of the surrogate fuel rodlet that will also be used in the actual LWR-2 irradiation specimen

Some Investigations of the Reaction of Activated Charcoal with Fluorine and Uranium Hexafluoride

The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has been shut down since 1969, when the fuel salt was drained from the core into two Hastelloy N drain tanks at the reactor site. Over time, fluorine (F{sub 2}) and uranium hexafluoride (UF{sub 6}) moved from the salt through the gas piping to a charcoal bed, where they reacted with the activated charcoal. Some of the immediate concerns related to the migration of F{sub 2} and UF{sub 6} to the charcoal bed were the possibility of explosive reactions between the charcoal and F{sub 2}, the existence of conditions that could induce a criticality accident, and the removal and recovery of the fissile uranium from the charcoal. This report addresses the reactions and reactivity of species produced by the reaction of fluorine and activated charcoal and between charcoal and F{sub 2}-UF{sub 6} gas mixtures in order to support remediation of the MSRE auxiliary charcoal bed (ACB) and the recovery of the fissile uranium. The chemical identity, stoichiometry, thermochemistry, and potential for explosive decomposition of the primary reaction product, fluorinated charcoal, was determined

Prototype Tests for the Recovery and Conversion of UF₆Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design

A descriptive model of the molten salt reactor experiment after shutdown: Review of FY 1995 progress

During FY 1995 considerable progress was made toward gaining a better understanding of the chemistry and transport processes that continue to govern the behavior of the Molten Salt Reactor Experiment (MSRE). As measurements in the MSRE proceed, laboratory studies continue, and better analyses are available, our understanding of the state of the MSRE and the best path toward remediation improves. Because of the immediate concern about the deposit in the auxiliary charcoal bed (ACB), laboratory studies in the past year focused on carbon-fluorine chemistry. Secondary efforts were directed toward investigation of gas generation from MSRE salts by both radiolytic and nonradiolytic pathways. In addition to the laboratory studies, field measurements at the MSRE provided the basis for estimating the inventory of uranium and fluorine in the ACB. Analysis of both temperature and radiation measurements provided independent and consistent estimates of about 2.6 kg of uranium deposited in the top of the ACB. Further analysis efforts included a refinement in the estimates of the fuel- salt source term, the deposited decay energy, and the projected rate of radiolytic gas generation. This report also provides the background material necessary to explain new developments and to review areas of particular interest. The detailed history of the MSRE is extensively documented and is cited where appropriate. This work is also intended to update and complement the more recent MSRE assessment reports

Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design