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Rediscovery of the Elements: Soddy and Isotopes

Description: Article describing Frederick Soddy's career and research of isotopes. Tourist information is included for areas significant to Soddy's life and work.
Date: Winter 2010
Creator: Marshall, James L., 1940- & Marshall, Virginia R.
Item Type: Article
Partner: UNT College of Arts and Sciences

Rediscovery of the Elements: Thorium

Description: Article recounting the discovery of the element Thorium in Norway by Hans Morten Thrane Esmark. Maps and tourist information regarding the area are provided.
Date: Winter 2001
Creator: Marshall, James L., 1940- & Marshall, Virginia R.
Item Type: Article
Partner: UNT College of Arts and Sciences

Characteristics of a Mixed Thorium - Uranium Dioxide High-Burnup Fuel

Description: Future nuclear fuel must satisfy three sets of requirements: longer times between refueling; concerns for weapons proliferation; and development of a spent fuel form more suitable for direct geologic disposal. This project has investigated a fuel consisting of mixed thorium and uranium dioxide to satisfy these requirements. Results using the SCALE 4.3 code system indicated that the mixed Th-U fuel could be burned to 72 MWD/kg or 100 MWD/kg using 25% and 35% UO2 respectively. The uranium remained below 20 % total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MWD was a factor of 4.5 less in the Th-U fuel than in the conventional fuel; Pu-239 production per MWD was a factor of 6.5 less; and the plutonium produced was high in Pu-238, leading to a decay heat 5 times greater than that from plutonium derived from conventional fuel and 40 times greater than weapons grade plutonium. High decay heat would require active cooling of any crude weapon, lest the components surrounding the plutonium be melted. Spontaneous neutron production for plutonium from Th-U fuel was 2.3 times greater than that from conventional fuel and 15 times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO2 is the highest oxide of thorium, while UO2 can be oxidized further to U3O8, ThO2- UO2 fuel may be a superior wasteform if the spent fuel is ever to be exposed to oxygenated water. Even if the cost of fabricating the mixed Th-U fuel is $100/kg greater, the cost of the Th-U fuel is 13% to 25% less than that of the fuels using uranium only.
Date: June 1, 1999
Creator: Herring, James Stephen & Mac Donald, Philip Elsworth
Item Type: Article
Partner: UNT Libraries Government Documents Department

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT FOR JANUARY, FEBRUARY, AND MARCH 1957

Description: Fluoride Volatilization Separations Process. Development of a fused fluoride process for dissolution of uranium-- zirconium fuel alloys continued. In corrosion tests to find a suitable container material, Ni was found to be susceptible to a sulfur-type attack. Hastelloy B showed promise, and graphite offers excellent chemical resistance but poor mechanical strength. The dissolution rate of Zr in NaF-- ZrF as affected by impingement of the HF sparge was studied. Production of UF/sub 6/ by fluidized bed fluorination of UF/sub 4/ from ore concentrates was studied. The preparation, melting point, vapor pressure, and vapor density of UF/sub 5/ are given. Preliminary dissolution and recovery runs in semi-works equipment are discussed. Fluidization. Fluidized- bed techniques have been applied to conversion of UO/sub 2/(NO/sub 3/)/sub 2/ to UF/sub 4/ and to calcination of radioactive liquid wastes. Activities of the Green Salt Pilot Plant and shakedown runs of the shielded waste calciner are described. Reactor Chemistry. Studies continued on the kinetics and mechanism of oxidation of U, Th, and Zr. Data are given for oxidation of U in oxygen from 125 to 295 tained C and 20 to 800 mm pressure, and for Zr from 400 to 900 tained C and 200 nan O/sub 2/ pressure. The ratio of capture to fission cross sections for U/sup 233/ and U/sup 238/ in EBR-I have been determined as a function of position. ChemicalMetallurgical Separations Processes. Development of pyrometullurgical processing of spent reactor fuels continued. Work is repcrted on: melt refining and casting of U--Pu; iodine volatility problem; the system U--B-- Ta; the distribution coefficients for Pu between U--Cr and Mg and U and Mg; extraction of Pu from U by liquid Mg; Ce removal by dross refining; adsorption of volatilized metuls on surface active materials; and fractional crystallization of U with Zn. Analytical Research. A study ...
Date: October 31, 1960
Item Type: Report
Partner: UNT Libraries Government Documents Department

The Fission of Thorium with Alpha Particles

Description: The fission distribution of fission of thorium with alpha particle of average energy 37.5 Mev has been measured by the chemical method. The distribution found shows that the characteristic dip in the fission yield mass spectrum has been raised to within a factor of two of the peaks compared to a factor of 600 in slow neutron fission of U{sup 235}. The raise in the deip has caused a corresponding lowering in fission yield of these elements at the peaks. The cross section for fission of thorium with 37.5 Mev alphas was found to be about 0.6 barn, and the threshold for fission was found to be 23 to 24 Mev.
Date: April 15, 1948
Creator: Newton, Amos S.
Item Type: Article
Partner: UNT Libraries Government Documents Department

The Fission of thorium with Alpha Particles

Description: Soon after the discovery of fission, Meitner, Bretscher and Cook found differences in the decay of various chemical fractions separated from uranium irradiated with slow neutrons and thorium irradiated with fast neutrons respectively and suggested that a difference existed in the distribution of fission products in the two cases. In 1940, Turner suggested that the distribution in various modes of fission should be investigated. The fact that elements such as tin, cadmium, palladium, and silver were found in fast neutron and deuteron fission of uranium and thorium before they were found in slow neutron fission of uranium suggested that the middle region of the distribution was raised as the energy of the incident particle was increased. Since the compound nucleus formed in the fission of thorium with alpha particles is U{sup 236}, the same compound nucleus formed in the fission of U{sup 235} with neutrons, it is of interest to study the fission of thorium with alphas and compare the resulting distribution of fission products with that found with uranium with slow and thorium with fast neutrons. Any difference between the various results where the same compound nucleus is formed must be due to differences in energy content and possible differences in distribution of the nucleons in the compound nucleus at the time of fission.
Date: October 15, 1948
Creator: Newton, Amos S.
Item Type: Report
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING DECEMBER, 1960

Description: Research areas include: reactor materials and components; alloy fuels; fission-gas release from refractory fuels; fuel-element development; gas-pressure bonding of ceramic, cermet, and dispersion fuel elements; development of uranium carbide; physical research; radioisotope and radiation applications; void- distribution and heattransfer studies; development of uranium mononitride; materials development and evaluation; coated-particle fuel materials; problems associated with recovery of spent fuel elements; pebble-bed reactor materials; development of fabrication processes for cold bonding of Zircaloy-2 to type 410 stninless steel; development and evaluation of fuel elements for MGCR; development studies for SM-2; gas-cooled reactor program; corrosion of thorium and uranium under storage conditions; and gas-pressure bonding of berylliumclad fuel elements. (For preceding period see BMI-1480.) (B.O.G.)
Date: January 1, 1961
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Item Type: Report
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF THORIUM-URANIUM-BASE FUEL ALLOYS

Description: Thorium-uranium alloys were studied with the aim of developing alloys with improved irradiation behavior by control of microstructure. The effect of thorium purity, melting technique, hot and cold working, and heat treatment on microstructure was investigated. The most signifi- . cant microstructural differences occurred as a result of casting technique, The arc-melted alloys exhibited the most nearly ideal structure, that of a homogeneous dispersion of small-diameter uranium particles in a thorium matrix, In addition, the rate of work hardening, recrystallization behavior, density, and hot hardness of thoriumuranium alloys were determined. As uranium content increases, the rate of work hardening increases, The recrystallization temperature of thorium was found to increase by over 100 deg C when uranium is present. Molybdenum, niobium, zirconium, and zirconium in conjunction with niobium were added to thorium- uranium with the aim of increasing irradiation resistance by stabilizing the gamma-uranium phase and/or improving the hightemperature strength of the alloy. It was found that small additions of molybdenum or niobium were effective in stabliizing the gamma-uranium phase, while zirconium was an effective hardener at temperatures up to 600 deg C, Zirconium additions to thorium-uranium alloys were effective in improving the 300 deg C water corrosion resistance of thorium by a factor of two. (auth)
Date: March 18, 1960
Creator: Farkas, M.S.; Bauer, A.A. & Dickerson, R.F.
Item Type: Report
Partner: UNT Libraries Government Documents Department

DISSOLUTION OF IRRADIATED CONSOLIDATED EDISON POWER-REACTOR FUEL BY THE SULFEX AND DAREX PROCESSES

Description: Losses of fertile and fissile materials during chemical decladding of irradiated prototype Consolidated Edison power-reactor fuel pins by the Sulfex and Darex processes were determined, on a laboratory scale, in all-glass apparatus. For air-fired low-density (-85 per cent of theoretical) fuel cores, minimum losses of uranium, thorium, and plutonium were in the 0.1 to 0.2 per cent range, by either process. These losses increased if the dejacketed cores were allowed to remain in contact with the cladding solution. No selectivity of dissolution of core components was apparent. Comparable losses were obtained with similar unirradiated fuel pins, irradiated core pellets showed a tendency to shatter. When shattered core pellets were present, losses to the cladding solution were excessive. Losses of from 0.5 to 4.5 per cent were observed, depending on the extent of core fragmentation and the time of contact with the cladding solution. No correlation between burnup and extent of shattering was discernible. Core dissolution times were not lengthened by irradiation to the 175 to 300-Mwd/t core level. (auth)
Date: March 10, 1960
Creator: Ewing, R.A.; Brugger, H.B. & Sunderman, D.N.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Studies with Colloids Containing Radioisotopes of Yttrium, Zirconium, Columbium and Lanthaum: 2. The Controlled Selective Localization of Radioisotopes of Yttrium, Zirconium, Columbium in the Bone Marrow, Liver and Spleen

Description: Several workers have shown that certain colloidally dispered materials are removed from the blood stream by the liver and spleen. Jones, Wrobel, and Lyons have utilized suspensions of anhydrous chromic phosphate for the selective irradiation of the liver and spleen with p{sup 32} beta particles. Gersh demonstrated that colloidal calcium phosphate is taken up by the liver and spleen. He stressed the failure of bone marrow phagocytes to take up this colloid in rats and dogs (though he referred to possible uptake in the marrow of rabbits under special conditions), and commented on the relative 'refractoriness' in general of the bono marrow as compared with liver and spleen with respect to the uptake of colloidal dyes from the blood stream. Some histological data indicate that 'Thorotrast' (a colloidal thorium dioxide preparation) is deposited in the bone marrow as well as in the liver and spleen, but no quantitative data as to the relative distribution are available. In the preceding communication the methods for the preparation of colloids incorporating radioisotopes of yttrium, columbium, and zirconium were given. The present studies are concerned with the localization of such colloids primarily in the bone marrow or primarily in the spleen and liver, with an analysis of some of the factors which may be responsible for differences in localization.
Date: April 21, 1948
Creator: Dobson, E.L.; Gofman, J.W.; Jones, H.B.; Kelly, Lola S. & Walker, L.
Item Type: Report
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MARCH 1960

Description: Progress is reported for investigations of: reactor materials and components; development of fuel elements; fissiongas release from refractory fuels; gas-pressure bonding in fuel elements; development of UC; radioisotope and radiation applications; heat transfer and void distributions; materials development and evaluation; reflective insulations; recovery of spent fuel elements; variable-moderator reactor critical assemblies; Pebble-Bed Reactor materials; tantalum and tantalum alloys for LAMPRE applications; materials development for HTGR and MGCR; development of SM-2; gas-cooled reactor program; and corrosion of thorium and uranium under storage conditions. (B.O.G.)
Date: April 1, 1960
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Item Type: Report
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING SEPTEMBER, 1960

Description: The study of possible reactions induced by the application of high pressures at high temperatures was concentrated on U/sub 3/O/sub 8/-Al/sub 2/O/ sub 3/. Data are reported on the effects of temperature on the tensile properties of fully recryatallized Cr-- Hb, Nb-- V, Nb-- Zr, and Cr-- Nb-- Zr alloys. The effect of radiation on the creep properties of Zircaloy-2 at elevated temperatures is being studied. Magnesium oxides are being evaluated as a fuel matrix material for uranium dioxide. Specimens of niobium-base birary alloys containing 10, 20, and 30 wt.% uranium and ternary alloys containing 20 wt.% uranium with 10 or 20 wt.% Zr were prepared with enriched uranium for irradiation tests. Pu--Nb, Pu-- Th, Mo-- Pu-- U, and Nb-- Pu--U alloys were selected as candidates in the search for improved irradiation resistance. Specimens consisting of uranium nitride dispersed in C, Cr, Fe, Mo, Nb, Ta, Ti, W, and Zr were fabricated and heat-treated to evaluate compatibility. A summary of results is presented from 3 hr, 1430 deg C, 5 tsi hydrostatic pressings of various types of uranium carbide and U--C -Nb systems. The effects of heat- treatment on some properties of U-C syatems are presented. Data are presented on the corrosion resistance of UC -Mo/sub 2/C, UC --NbC, UC -TiC, UC --VC, Uc -- ZrC --Mo/sub 2/C, and UC -- ZrC --VC to Santowax R at 350 C for7 days. The development of powder metallurgy and melting and casting techniques for uranium nitrides is reported. Creep and atress-rupture tests are reported on Zircaloy-2 sheet specimens. Fission gas release experiments are reported on fueled-graphite spheres in suppent of the Pebble-bed Reactor. In the development of container materials for LAMPRE applications, the fabrication behavior of 13 birary tantalum- base alloys was investigated. Radiation-effects studies of fuel specimens of UO/ sub 2/ ...
Date: October 1, 1960
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Item Type: Report
Partner: UNT Libraries Government Documents Department

A DISCUSSION OF THE METALLURGICAL ASPECTS OF THORIUM AS A FUEL MATERIAL

Description: Data on thorium are compiled and the use of thorium as a fuel material is discussed. It was assumed that the general reactor configuration will follow the conventional SGR-SRE reactor design of a graphite-moderated sodium-cooled metallic fuel system. Factors outlined which affect the use of thorium as a fuel material are availability of ore, capacity of processing plants, purity specifications, design and use of materials in fuel elements which may affect chemical separation economics, and technical feasibility. General metallurgical properties of thorium are described including melting point, density, chemical purity, thermal conductivity, thermal expansion, phase transformations, corrosion in liquid metals, tensile and creep strength, dimensional stability, and alloying properties. Methods of chemical processing, fabrication, and cladding are outlined. The advantages and disadvantages in the use of thorium as a fuel are summarized. (M.C.G.)
Date: November 1, 1953
Creator: Hayward, B.R.
Item Type: Report
Partner: UNT Libraries Government Documents Department

On the Coordination of Actinides and Fission Products in Silicate Glasses

Description: The local structure around Th, U, Ce and Nd in leached silicate glasses was examined using XAFS spectroscopy at their L3 edges and also at the K edge of Fe, Co, Ni, Zr and Mo. Pellets of inactive borosilicate glasses with a simplified or a complex composition were leached statically at 90 C, at pH buffered to 0 or 6 for 28 days (surface/volume, S/V, ratios of 0.1 cm{sup -1}). These glasses are compared to another SON68 sample (denoted ''SP1'' in this paper) that was statically leached for 12 years under similar conditions, except for a higher S/V of 12 cm{sup -1} and a higher unconstrained pH of 9.6. The speciation of Fe, Co, Ni, Zr and Mo in the simple and the complex unleached are similar. In the statically leached glasses, the speciation of these transition metals is mostly identical to in the unleached glasses, except in the gels formed at the surface of the glasses leached at low pH, where large speciation differences are observed. Surface precipitates, especially for Fe (as ferrihydrite), Mo (possibly sidwillite) and Th (as ThO{sub 2}) were detected. Finally, the drying of the gels considerably affects the metal speciation by enhancing metal polymerization.
Date: December 13, 2006
Creator: Haddi, Anne; U., /Marne la Vallee; Farges, Francois; /Marne la Vallee U. /Museum Nat. Hist., Paris /Stanford U., Geo. Environ. Sci.; Trocellier, Patrick; /Saclay et al.
Item Type: Article
Partner: UNT Libraries Government Documents Department

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

Description: High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards ...
Date: October 1, 2009
Creator: Durst, Phillip Casey; Beddingfield, David; Boyer, Brian; Bean, Robert; Collins, Michael; Ehinger, Michael et al.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Natural Tracers and Multi-Scale Assessment of Caprock Sealing Behavior: A Case Study of the Kirtland Formation, San Juan Basin

Description: The assessment of caprocks for geologic CO{sub 2} storage is a multi-scale endeavor. Investigation of a regional caprock - the Kirtland Formation, San Juan Basin, USA - at the pore-network scale indicates high capillary sealing capacity and low permeabilities. Core and wellscale data, however, indicate a potential seal bypass system as evidenced by multiple mineralized fractures and methane gas saturations within the caprock. Our interpretation of {sup 4}He concentrations, measured at the top and bottom of the caprock, suggests low fluid fluxes through the caprock: (1) Of the total {sup 4}He produced in situ (i.e., at the locations of sampling) by uranium and thorium decay since deposition of the Kirtland Formation, a large portion still resides in the pore fluids. (2) Simple advection-only and advection-diffusion models, using the measured {sup 4}He concentrations, indicate low permeability ({approx}10-20 m{sup 2} or lower) for the thickness of the Kirtland Formation. These findings, however, do not guarantee the lack of a large-scale bypass system. The measured data, located near the boundary conditions of the models (i.e., the overlying and underlying aquifers), limit our testing of conceptual models and the sensitivity of model parameterization. Thus, we suggest approaches for future studies to better assess the presence or lack of a seal bypass system at this particular site and for other sites in general.
Date: March 15, 2011
Creator: Heath, Jason; McPherson, Brian & Dewers, Thomas
Item Type: Report
Partner: UNT Libraries Government Documents Department

SOLUBILITIES OF SELECTED METALS IN MERCURY: HERMEX PROCESS

Description: The solubilities of uraninm, thorium, gadelinium, sama rinm, and neodymium in mereury were determined from room temperatare to 356 deg C. Equations of the form log of solubility (wt.%) = a + b/T were developed for these metals. Integral heats of solution were calculated for each. The solubilities of ruthenium, palladium, zirconium, and molybdenum in mercury in the presence of excess uranium were determined; the low solubility of zirconium and molybdenum gave solutions with a concentration below the limit of detection in the analytical method used. Their values are reported as an upper solubility limit. Uranium solubility in a 0.1 wt.% magnesium amAlgam was approximately 1.2 to 1.5 times greater than in mercury alone. When uranium and thorium were present in the same mercury solution, their solubilities were mutually depressed. (auth)
Date: June 29, 1960
Creator: Messing, A. F. & Dean, O. C.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Rapid Column Extraction Method for Actinides and Sr-89/90 in Water Samples

Description: The SRS Environmental Laboratory analyzes water samples for environmental monitoring, including river water and ground water samples. A new, faster actinide and strontium 89/90 separation method has been developed and implemented to improve productivity, reduce labor costs and add capacity to this laboratory. This method uses stacked TEVA Resin{reg_sign}, TRU Resin{reg_sign} and Sr-Resin{reg_sign} cartridges from Eichrom Technologies (Darien, IL, USA) that allows the rapid separation of plutonium (Pu), neptunium (Np), uranium (U), americium (Am), curium (Cm) and thorium (Th) using a single multi-stage column combined with alpha spectrometry. By using vacuum box cartridge technology with rapid flow rates, sample preparation time is minimized. The method can be used for routine analysis or as a rapid method for emergency preparedness. Thorium and curium are often analyzed separately due to the interference of the daughter of Th-229 tracer, actinium (Ac)-225, on curium isotopes when measured by alpha spectrometry. This new method also adds a separation step using DGA Resin{reg_sign}, (Diglycolamide Resin, Eichrom Technologies) to remove Ac-225 and allow the separation and analysis of thorium isotopes and curium isotopes at the same time.
Date: June 15, 2005
Creator: MAXWELL III, SHERROD L.
Item Type: Article
Partner: UNT Libraries Government Documents Department

Three dimensional depletion analysis for the ''as built'' FSV initial core

Description: During fabrication of the initial FSV core fuel elements, some difficulties with assayed quantities of uranium and thorium were encountered. This problem became apparent when about 40 percent of the total fuel rod requirement was in some state of production and about 10 percent of the fuel elements had been assembled. As a result, some of the specified fuel blends were nearing completion and the uranium and/or thorium was 3 to 4 percent too high or low, and some of the already assembled fuel elements were outside the loading tolerances given in the fuel specification. A description is given of the method of resolution of this problem with the minimum perturbation to the fuel production and no loss of integrity on the fuel performance. (auth)
Date: November 15, 1975
Creator: Nirschl, R.J.
Item Type: Report
Partner: UNT Libraries Government Documents Department

A Thermodynamic Model for Acetate, Lactate, and Oxalate Complexation with Am(III), Th(IV), Np(V), and U(VI) Valid to High Ionic Strength

Description: The organic ligands acetate, lactate, oxalate and EDTA have been identified as components of wastes targeted for disposal in the Waste Isolation Pilot Plant (WIPP) located in Southeastern New Mexico. The presence of these ligands is of concern because complexation of the actinides with the ligands may increase dissolved actinide concentrations and impact chemical retardation during transport. The current work considers the complexation of Am(III), Th (IV), Np(V), and U(W) with two of the organic ligands, acetate and lactate, in NaCl media from dilute through high concentration. A thermodynamic model for actinide complexation with the organic ligands has been developed based on the Pitzer activity coefficient formalism and the Harvie-Moller-Weare, Felmy-Weare database for describing brine evaporite systems. The model was parameterized using first apparent stability constant data from the literature. Because of complexation of other metal ions (Fe, Mg, Ni, Pb, etc.) present in the WIPP disposal room with the organic ligands, preliminary results from model calculations indicate the organic ligands do not significantly increase dissolved actinide concentrations.
Date: January 15, 1999
Creator: Bynaum, R.V.; Free, S.J. & Moore, R.C.
Item Type: Article
Partner: UNT Libraries Government Documents Department

ALPHA SPECTROMETRIC EVALUATION OF SRM-995 AS A POTENTIAL URANIUM/THORIUM DOUBLE TRACER SYSTEM FOR AGE-DATING URANIUM MATERIALS

Description: Uranium-233 (t{sub 1/2} {approx} 1.59E5 years) is an artificial, fissile isotope of uranium that has significant importance in nuclear forensics. The isotope provides a unique signature in determining the origin and provenance of uranium-bearing materials and is valuable as a mass spectrometric tracer. Alpha spectrometry was employed in the critical evaluation of a {sup 233}U standard reference material (SRM-995) as a dual tracer system based on the in-growth of {sup 229}Th (t{sub 1/2} {approx} 7.34E3 years) for {approx}35 years following radiochemical purification. Preliminary investigations focused on the isotopic analysis of standards and unmodified fractions of SRM-995; all samples were separated and purified using a multi-column anion-exchange scheme. The {sup 229}Th/{sup 233}U atom ratio for SRM-995 was found to be 1.598E-4 ({+-} 4.50%) using recovery-corrected radiochemical methods. Using the Bateman equations and relevant half-lives, this ratio reflects a material that was purified {approx} 36.8 years prior to this analysis. The calculated age is discussed in contrast with both the date of certification and the recorded date of last purification.
Date: December 6, 2011
Creator: Beals, D.
Item Type: Article
Partner: UNT Libraries Government Documents Department

IMPACT OF URANIUM AND THORIUM ON HIGH TIO2 CONCENTRATION NUCLEAR WASTE GLASSES

Description: This study focused on the potential impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. MST from the Salt Waste Processing Facility (SWPF) is also considered in the study. The KT08-series of glasses was designed to evaluate any impacts of the inclusion of uranium and thorium in glasses containing the SCIX components. All but one of the study glasses were found to be amorphous by X-ray diffraction (XRD). One of the slowly cooled glasses contained a small amount of trevorite, which is typically found in DWPF-type glasses and had no practical impact on the durability of the glass. The measured Product Consistency Test (PCT) responses for the study glasses and the viscosities of the glasses were well predicted by the current DWPF models. No unexpected issues were encountered when uranium and thorium were added to the glasses with SCIX components.
Date: January 11, 2012
Creator: Fox, K. & Edwards, T.
Item Type: Article
Partner: UNT Libraries Government Documents Department