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Gas cooled fast reactor fuel cost assessment. Final report, October 1978-September 1979

Description: Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than a few years ago. Fuel assembly fabrication costs represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made. The fabrication costs are calculated based on an equation developed by HEDL.
Date: January 1, 1979
Creator: Thompson, M.L.
Partner: UNT Libraries Government Documents Department

Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

Description: This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.
Date: January 1, 1979
Creator: Thompson, M.L.
Partner: UNT Libraries Government Documents Department

Tests of alternative reductants in the second uranium purification cycle

Description: Miniature mixer-settler tests of the second uranium purification cycle show that plutonium cannot be removed by hydroxylamine-hydrazine (NH/sub 2/OH-N/sub 2/H/sub 4/) because the acidity is too high, or by 2,5-di-t-pentylhydroquinone because HNO/sub 3/ oxidizes the hydroquinone. Plutonium can be removed satisfactorily when U(IV)-hydrazine is used as the reductant.
Date: May 1, 1980
Creator: Thompson, M.C.
Partner: UNT Libraries Government Documents Department

Distribution of selected lanthanides and actinides between 30% TBP in n- paraffin and various metal nitrate solutions

Description: Distributions were measured for nitric acid, americium, curium, cerium, neodymium, samarium, europium, manganese, and mercury between 30 vol % THP in n- paraffin and aqueous solutions containing nitric acid, aluminum nitrate, lithium nitrate, and/or sodium nitrate. Equations for the distributions were derived from the data and used in designing solvent extraction flowsheets for recovery and decortamination of americium and curium from irradiated plutonium- aluminum alloy. Also investigated was the effect of DTPA acid on the distribution of the actinides and lanthanides in the same systems. The actinides (Am and Cm) are more strongly complexed than the light lanthanides (La, Ce, Pr, and Nd) by DTPA acid. By controlling the solution pH in the range of 1 to 3, separation of actinides from lanthanides by factors of 10 to 100 may be obtained by extraction with 30% THP. (auth)
Date: November 1, 1973
Creator: Thompson, M.C.
Partner: UNT Libraries Government Documents Department

Pretreatment/Radionuclide Separations of Cs/Tc from Supernates

Description: Significant improvements have been made in ion exchange and solvent extraction materials and processes available for separation of the radionuclides cesium and technetium from both acid and alkaline waste solutions. New ion exchange materials and solvent extraction reagents are more selective for Cs over sodium and potassium than previous materials. The higher selectivity gives higher Cs capacity and improved separation processes. Technetium removal has been improved by new ion exchange resins, which have either improved capacity or easier elution. Several different crown ethers have been shown to extract pertechnetate ion selectively over other anions. Organic complexants in some waste solutions reduce pertechnetate ion and stabilize the reduced species. Selective oxidation allows conversion to pertechnetate without oxidation of the organic complexants.
Date: September 1, 1998
Creator: Thompson, M.C.
Partner: UNT Libraries Government Documents Department

Demonstration of the UREX Solvent Extraction Process with Dresden Reactor Fuel Solution

Description: A solvent extraction process to recover uranium and technetium from solutions of irradiated commercial reactor fuel while sending the plutonium to waste with the fission products and higher actinides was tested with actual fuel solution. Demonstration of the uranium extraction (UREX) process at baseline conditions showed that the process meets all goals for recovery and decontamination.
Date: November 1, 2002
Creator: Thompson, M.C.
Partner: UNT Libraries Government Documents Department

PRISM (Power Reactor Inherently Safe Module) design concept enhances waste management

Description: PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs.
Date: January 1, 1989
Creator: Thompson, M.L. & Berglund, R.C.
Partner: UNT Libraries Government Documents Department

Equilibrium distribution of uranyl nitrate between nitric acid and 7.5 vol % TBP

Description: The distribution of nitric acid and uranyl nitrate between aqueous solution and 7.5 vol percent tributyl phosphate in normal paraffin diluent was measured at 23, 45, and 60$sup 0$C. The data are consistent with behavior expected from theory and with previously published data. The data have been used to design a solvent extraction flowsheet for recovery of enriched uranium. (auth)
Date: October 1, 1975
Creator: Thompson, M.C.; Murphree, B.E. & Shankle, R.L.
Partner: UNT Libraries Government Documents Department

DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

Description: The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.
Date: August 7, 2009
Creator: Kessinger, G. & Thompson, M.
Partner: UNT Libraries Government Documents Department

SHARING AND DEPLOYING INNOVATIVE INFORMATION TECHNOLOGY SOLUTIONS TO MANAGE WASTE ACROSS THE DOE COMPLEX

Description: There has been a need for a faster and cheaper deployment model for information technology (IT) solutions to address waste management needs at US Department of Energy (DOE) complex sites for years. Budget constraints, challenges in deploying new technologies, frequent travel, and increased job demands for existing employees have prevented IT organizations from staying abreast of new technologies or deploying them quickly. Despite such challenges, IT organizations have added significant value to waste management handling through better worker safety, tracking, characterization, and disposition at DOE complex sites. Systems developed for site-specific missions have broad applicability to waste management challenges and in many cases have been expanded to meet other waste missions. Radio frequency identification (RFID) and global positioning satellite (GPS)-enabled solutions have reduced the risk of radiation exposure and safety risks. New web-based and mobile applications have enabled precision characterization and control of nuclear materials. These solutions have also improved operational efficiencies and shortened schedules, reduced cost, and improved regulatory compliance. Collaboration between US Department of Energy (DOE) complex sites is improving time to delivery and cost efficiencies for waste management missions with new information technologies (IT) such as wireless computing, global positioning satellite (GPS), and radio frequency identification (RFID). Integrated solutions developed at separate DOE complex sites by new technology Centers of Excellence (CoE) have increased material control and accountability, worker safety, and environmental sustainability. CoEs offer other DOE sister sites significant cost and time savings by leveraging their technology expertise in project scoping, implementation, and ongoing operations.
Date: January 31, 2011
Creator: Crolley, R. & Thompson, M.
Partner: UNT Libraries Government Documents Department

ENHANCED CHEMICAL CLEANING OF SRS WASTE TANKS TO IMPROVE ACTINIDE SOLUBILITY

Description: Processes for the removal of residual sludge from SRS waste tanks have historically used solutions containing up to 0.9 M oxalic acid to dissolve the remaining material following sludge removal. The selection of this process was based on a comparison of a number of studies performed to evaluate the dissolution of residual sludge. In contrast, the dissolution of the actinide mass, which represents a very small fraction of the waste, has not been extensively studied. The Pu, Np, and Am in the sludge is reported to be present as hydrated and crystalline oxides. To identify aqueous solutions which have the potential to increase the solubility of the actinides, the alkaline and mildly acidic test solutions shown below were selected as candidates for use in a series of solubility experiments. The efficiency of the solutions in solubilizing the actinides was evaluated using a simulated sludge prepared by neutralizing a HNO{sub 3} solution containing Pu, Np, and Am. The hydroxide concentration was adjusted to a 1.2 M excess and the solids were allowed to age for several weeks prior to starting the experiments. The sludge was washed with 0.01 M NaOH to prepare the solids for use. Following the addition of an equal portion of the solids to each test solution, the concentrations of Pu, Np, and Am were measured as a function of time over a 792 h (33 day) period to provide a direct comparison of the efficiency of each solution in solubilizing the actinide elements. Although the composition of the sludge was limited to the hydrated actinide oxides (and did not contain other components of demonstrated importance), the results of the study provides guidance for the selection of solutions which should be evaluated in subsequent tests with a more realistic surrogate sludge and actual tank waste.
Date: September 20, 2011
Creator: Rudisill, T. & Thompson, M.
Partner: UNT Libraries Government Documents Department

Flowsheet for coprocessing uranium and plutonium

Description: A coprocessing solvent extraction flowsheet for recovering and purifying LWR fuels has been altered to achieve partial partioning of uranium and plutonium to eliminate streams with pure plutonium. Partial partitioning has been demonstrated in the laboratory with simulated feeds and irradiated LWR fuel solutions. Hydroxylamine nitrate was the reductant in these tests. Plutonium was concentrated by factors of 6 to 27.4. Tests have shown that 1 to 2 plutonium atoms are reduced for each hydroxylamine molecule consumed. Nitrite interferes with the reduction of plutonium, unless the hydroxylamine concentration is increased. 12 tables, 11 figures.
Date: January 1, 1979
Creator: Statton, M. A. & Thompson, M. C.
Partner: UNT Libraries Government Documents Department

Purification Testing for HEU Blend Program

Description: The Savannah River Site (SRS) is working to dispose of the inventory of enriched uranium (EU) formerly used to make fuel for production reactors. The Tennessee Valley Authority (TVA) has agreed to take the material after blending the EU with either natural or depleted uranium to give a {sup 235}U concentration of 4.8 percent low-enriched uranium will be fabricated by a vendor into reactor fuel for use in TVA reactors. SRS prefers to blend the EU with existing depleted uranium (DU) solutions, however, the impurity concentrations in the DU and EU are so high that the blended material may not meet specifications agreed to with TVA. The principal non-radioactive impurities of concern are carbon, iron, phosphorus and sulfur. Neptunium and plutonium contamination levels are about 40 times greater than the desired specification. Tests of solvent extraction and fuel preparation with solutions of SRS uranium demonstrate that the UO{sub 2} prepared from these solutions will meet specifications for Fe, P and S, but may not meet the specifications for carbon. The reasons for carbon remaining in the oxide at such high levels is not fully understood, but may be overcome either by treatment of the solutions with activated carbon or heating the UO{sub 3} in air for a longer time during the calcination step of fuel preparation.Calculations of the expected removal of Np and Pu from the solutions show that the specification cannot be met with a single cycle of solvent extraction. The only way to ensure meeting the specification is dilution with natural U which contains no Np or Pu. Estimations of the decontamination from fission products and daughter products in the decay chains for the U isotopes show that the specification of 110 MEV Bq/g U can be met as long as the activities of the daughters of U- ...
Date: June 1, 1998
Creator: Thompson, M.C. & Pierce, R.A.
Partner: UNT Libraries Government Documents Department