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Fuel Fabrication for Surrogate Sphere-Pac Rodlet

Description: 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 ...
Date: July 19, 2005
Creator: Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

Prototype Tests for the Recovery and Conversion of UF<sub>6</sub>Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

Description: 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.
Date: June 7, 2000
Creator: Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

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

Description: 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 ...
Date: October 1, 2002
Creator: Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

Molten fluoride fuel salt chemistry

Description: 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.
Date: February 1, 1995
Creator: Toth, L.M.; Del Cul, G.D.; Dai, S. & Metcalf, D.H.
Partner: UNT Libraries Government Documents Department

Hydrofluoric Acid Corrosion Testing on Unplated and Electroless Gold-Plated Samples

Description: The Molten Salt Reactor Experiment (MSRE) remediation requires that almost 40 kg of uranium hexafluoride (UF6) be converted to uranium oxide (UO). In the process of this conversion, six moles of hydrofluoric acid (HP) are produced for each mole of UF6 converted.
Date: August 1, 2000
Creator: Osborne, P.E.; Icenhour, A.S. & Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: April 1, 2000
Creator: Del Cul, G.D.; Icenhour, A.S. & Simmons, D.W.
Partner: UNT Libraries Government Documents Department

Evaluation of possible physical-chemical processes that might lead to separations of actinides in ORNL waste tanks

Description: The concern that there might be some physical-chemical process which would lead to a separation of the poisoning actinides ({sup 232}Th, {sup 238}U) from the fissionable ones ({sup 239}Pu, {sup 235}U) in waste storage tanks at Oak Ridge National Laboratory has led to a paper study of potential separations processes involving these elements. At the relatively high pH values (>8), the actinides are normally present as precipitated hydroxides. Mechanisms that might then selectively dissolve and reprecipitate the actinides through thermal processes or additions of reagents were addressed. Although redox reactions, pH changes, and complexation reactions were all considered, only the last type was regarded as having any significant probability. Furthermore, only carbonate accumulation, through continual unmonitored air sparging of the tank contents, could credibly account for gross transport and separation of the actinide components. From the large amount of equilibrium data in the literature, concentration differences in Th, U, and Pu due to carbonate complexation as a function of pH have been presented to demonstrate this phenomenon. While the carbonate effect does represent a potential separations process, control of long-term air sparging and solution pH, accompanied by routine determinations of soluble carbonate concentration, should ensure that this separations process does not occur.
Date: September 1, 1997
Creator: Del Cul, G.D.; Toth, L.M.; Bond, W.D. & Dai, S.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: January 1, 1996
Creator: Williams, D.F.; Del Cul, G.D. & Toth, L.M.
Partner: UNT Libraries Government Documents Department

Immobilization of technetium and nitrate in cement-based materials

Description: The leachabilities of technetium and nitrate wastes immobilized in cement-based grouts have been investigated. Factors found to affect the leachabilities include grout mix ratio, grout fluid density, dry solid blend composition, and waste concentration. 10 refs., 7 figs., 3 tabs.
Date: January 1, 1987
Creator: Tallent, O.K.; McDaniel, E.W.; Del Cul, G.D.; Dodson, K.E. & Trotter, D.R.
Partner: UNT Libraries Government Documents Department

Visible and near-IR spectroscopic studies of UC1{sub 4} in a basic ambient temperature melt: The observation of a possible geometric distorted UC1{sub 6}{sup 2{minus}} species and the evidence for the hydrogen-bond in the melt

Description: Since high temperatures can lead to broadening of absorption spectra, ambient temperature chloride melts were used; the system used was AlCl{sub 3}-1-ethyl-3-methyl-imidazolium chloride (EMIC). The uv-visible spectrum of UCl{sub 4} in basic melt had many peaks with the most intense (triplet) ones around 2000 nm, similar to these at high temperature and indicating the same species. The electronic transition is allowed by a static rather than a vibronic mechanism. The central peak in the UCl{sub 6}{sup 2-} spectrum indicates distortion of geometry from the O{sub h} symmetry by the solvent medium. Very strong hydrogen bonding between UCl{sub 6}{sup 2-} and solvent EMI is suggested.
Date: September 1, 1994
Creator: Dai, S.; Toth, L. M.; Del Cul, G. D. & Metcalf, D. H.
Partner: UNT Libraries Government Documents Department

Molten Salt Fuel Cycle Requirements for ADTT Applications

Description: 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.
Date: June 7, 1999
Creator: Del Cul, G.D.; Toth, L.M. & Williams, D.F.
Partner: UNT Libraries Government Documents Department

Review of ORNL`s MSR technology and status

Description: 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.
Date: August 1, 1996
Creator: Toth, L.M.; Gat, U.; Del Cul, G.D.; Dai, S. & Williams, D.F.
Partner: UNT Libraries Government Documents Department

Laboratory tests using chlorine trifluoride in support of deposit removal at MSRE

Description: Experimental trials were conducted to investigate some unresolved issues regarding the use of chlorine trifluoride (ClF{sub 3}) for removal of uranium-bearing deposits in the Molten Salt Reactor Experiment (MSRE) off-gas system. The safety and effectiveness of operation of the fixed-bed trapping system for removal of reactive gases were the primary focus. The chief uncertainty concerns the fate of chlorine in the system and the potential for forming explosive chlorine oxides (primarily chlorine dioxide) in the trapping operation. Tests at the MSRE Reactive Gas Removal System reference conditions and at conditions of low ClF{sub 3} flow showed that only very minor quantities of reactive halogen oxides were produced before column breakthrough. Somewhat larger quantities accompanied breakthrough. A separation test that exposed irradiated MSRE simulant salt to ClF{sub 3} confirmed the expectation that the salt is basically inert for brief exposures to ClF{sub 3} at room temperature.
Date: April 1, 1997
Creator: Williams, D.F.; Rudolph, J.C.; Del Cul, G.D.; Loghry, S.L.; Simmons, D.W. & Toth, L.M.
Partner: UNT Libraries Government Documents Department

Passivation of fluorinated activated charcoal

Description: The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C{sub x}F to carbon and ammonium fluoride, NH{sub 4}F. The charcoal laden with NH{sub 4}F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH{sub 4}F as a mixture of NH{sub 3} and HF, which would primarily recombine as NH{sub 4}F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH{sub 3} concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes ...
Date: October 1, 1997
Creator: Del Cul, G. D.; Trowbridge, L. D.; Simmons, D. W.; Williams, D. F. & Toth, L. M.
Partner: UNT Libraries Government Documents Department

Laboratory tests in support of the MSRE reactive gas removal system

Description: The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has been shut down since December 1969, at which time the molten salt mixture of LiF-BeF{sub 2}-ZrF{sub 4}-{sup 233}UF{sub 4} (64.5-30.3-5.0-0.13 mol%) was transferred to fuel salt drain tanks for storage. In the late 1980s, increased radiation in one of the gas lines from the drain tank was attributed to {sup 233}UF{sub 6}. In 1994 two gas samples were withdraw (from a gas line in the Vent House connecting to the drain tanks) and analyzed. Surprisingly, 350 mm Hg of F{sub 2}, 70 mm Hg of UF{sub 6}, and smaller amounts of other gases were found in both of the samples. To remote this gas from above the drain tanks and all of the associated piping, the reactive gas removal system (RGRS) was designed. This report details the laboratory testing of the RGRS, using natural uranium, prior to its implementation at the MSRE facility. The testing was performed to ensure that the equipment functioned properly and was sufficient to perform the task while minimizing exposure to personnel. In addition, the laboratory work provided the research and development effort necessary to maximize the performance of the system. Throughout this work technicians and staff who were to be involved in RGRS operation at the MSRE site worked directly with the research staff in completing the laboratory testing phase. Consequently, at the end of the laboratory work, the personnel who were to be involved in the actual operations had acquired all of the training and experience necessary to continue with the process of reactive gas removal.
Date: July 1, 1997
Creator: Rudolph, J.C.; Del Cul, G.D.; Caja, J.; Toth, L.M.; Williams, D.F.; Thomas, K.S. et al.
Partner: UNT Libraries Government Documents Department

Removal of uranium and salt from the Molten Salt Reactor Experiment

Description: In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.
Date: June 1, 1998
Creator: Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L. & Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: September 1, 1998
Creator: Del Cul, G.D.; Fiedor, J.N.; Simmons, D.W.; Toth, L.M.; Trowbridge, L.D. & Williams
Partner: UNT Libraries Government Documents Department