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40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Summary Report for the Period January 1, 1959-December 31, 1959 and Quarterly Progress Report for the Period October 1, 1959-December 31, 1959

Description: The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of both fuel compacts and graphite sleeves. The hot-pressing process for making fuel compacts was used successfully to make full-size compacts with a uniform distribution of ThC/sub 2/- UC/sub 2/ particles. Three irradiation capsules were fabricated and inserted in a test reactor to determine fuel compact and sleeve performance under HTGR conditions of irradiation and temperature. Two of these ran satisfactorily for the scheduled time of operation. A scope design study of the in-pile loop that will be used to evaluate the full-diameter graphite-clad element was completed. Experiments to determine the extent of fuel migration within the element were undertaken. Preliminary results indicated that the central fuel-element temperatures must not exceed 2300-C for routine operation. An important start was made in developing an understanding of how to treat the neutron thermalization process in high-temperature graphite reactors. Analytical techniques for calculating the thermal neutron spectra in poisoned graphite media were developed and programmed for the IBM 704 computer. The experimental technique of measuring neutron spectra by using a pulsed linear electron accelerator was demonstrated by measurements made with boron-loaded graphite. A mockup of a small portion of the reactor core was constructed and operated to determine the local heat-transfer coefficients and pressure drop ...
Date: September 1, 1960
Partner: UNT Libraries Government Documents Department

THE ANALYSIS OF REFRACTORY BORIDES, CARBIDES, NITRIDES, AND SILICIDES

Description: Methods are presented for the analysis of 41 refractory materials. An evaluation of the accuracy and the precision of these techniques are also given The materials studied are the borides of hafnium, molybdenum, niobiumL rhenium, tantalum, thorium, titanium, tungsten, uranium vanadium, and zirconium; the carbides of hafnium molybdenum, miobium, silicon, tantalum, thorium, titanium, tungsten, uranium, vanadium, and zircomium; the nitrides of boron, hafnium, niobium, silicon, tantalum, titanium, uranium, and zirconium; the silicides of molybdenum, rhenium, tantalum, titanium, tungsten, vanadium, and zirconium; and mixed carbides of uranium with hafnium, niobium, tantalum, or zirconium. (auth)
Date: March 1, 1959
Creator: Kriege, O.H.
Partner: UNT Libraries Government Documents Department

Aqueous Corrosion Rates for Waste Package Materials

Description: The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.
Date: October 8, 2004
Creator: Arthur, S.
Partner: UNT Libraries Government Documents Department

Determination of thorium in plutonium-thorium oxides and carbides

Description: Thorium is determined in (PuTh)C and (PuTh)O/sub 2/ by complexometric titration with ethylenediaminetetraacetic acid (EDTA) following separation on anion-exchange resin. Carbides are first oxidized by ignition in air at about 800/sup 0/C. Oxide or oxidized carbide samples are dissolved in acids by the sealed-reflux technique or by heating in beakers. The plutonium is selectively sorbed from the 12M hydrochloric acid solution of the fuel on a Bio-Rad AG1-X2 anion-exchange resin column, and the eluted thorium is titrated with EDTA using xylenol orange as the indicator. The average recovery of thorium in 20 samples is 99.98% with a relative standard deviation of 0.07%.
Date: October 1, 1979
Creator: Walker, L.F. & Temer, D.J.
Partner: UNT Libraries Government Documents Department

THE DEVELOPMENT AND EVALUATION OF GRAPHITE-MATRIX FUEL COMPACTS FOR THE HTGR

Description: >A summary of development and evaluation of graphitematrix fuel compacts for the high-temperature gas-cooled reactor is presented. The fuel compacts consist of Th-- U dicarbide particles dispersed in a high-density graphite matrix and are produced by a hot-pressing process. Compacts prepared from Th--U oxide, Th-- U silicide, and Th--U dicarbide coated with pyrolytic C were subjected to radiation and fission-product-retention testing. Results of these tests are included and are discussed in detail. (J.R.D.)
Date: August 1, 1961
Creator: Goeddel, W.V.
Partner: UNT Libraries Government Documents Department

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

Description: There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.
Date: September 28, 2001
Creator: Radulescu, H.
Partner: UNT Libraries Government Documents Department

AN EVALUATION OF DATA ON NUCLEAR CARBIDES

Description: Data on the properties, constitution, compatibility, radiation behavior, fabrication, preparation, storage, and handling of uranium, thorium, and plutonium carbides are reviewed. 187 references. (C.J.G.)
Date: May 31, 1960
Creator: Rough, F.A. & Chubb, W. eds.
Partner: UNT Libraries Government Documents Department

EVALUATION OF (Th,U)C$sub 2$, CARBON-COATED (Th,U)C$sub 2$ PARTICLES, AND CARBON COATINGS

Description: Thorium and uranium carbide nuclear fuel particles were evaluated by metallographic and x-ray diffraction techniques. Techniques were developed to etch the polished surface of Th--U carbide to reveal the grain structure. In addition, techniques to determine particle density and coating thickness were developed. Comparison of the data indicates that the use of spherical particles allowed for more precise determination of the coating thickness, density, and strength of coatings. Strength of individual particles was about 700 to 1300 g per particle. A large scatter in crushing-strength values was observed when individual particles were crushed. Consequently, a relative crushingstrength test was developed for comparing coating strength: loads were applied to a column of particles in stepped increments, and broken coatings were detected by observing a weight gain in moist air from hydrolysis of the Th--U carbides. A correlation of relative crushing strength and coating thickness was obtained. The crystal structure of the C coatings was found to depend on the temperature of deposition in the range from 1400 to 2400 deg C; the twodimensional structure became more defined with increasing deposition temperatures. Subsequent annealing at 2400 deg C of the coatings deposited at 1400 deg C and at 2200 to 2400 deg C produced graphitization, whereas those deposited at 1800 to 2000 deg C retained their two-dimensional character. Macrostructure of the coatings as revealed by metallographic techniques showed a concentric shell-like structure in the 1400 deg C deposits and mottled or conical structure in the 1800 deg C deposits. Annealing to higher temperatures did not alter the general appearance of either macrostructure. (auth)
Date: April 26, 1962
Creator: Engle, G. B.; Luby, C. S. & Bokros, J. C.
Partner: UNT Libraries Government Documents Department

Fuel cycle cost studies: fabrication, reprocessing, and refabrication of LWR, SSCR, HWR, LMFBR, and HTGR fuels

Description: The comparative analysis of power generation costs for the various reactor cycles that is being performed in the Nonproliferation Alternative Systems Assessment Program (NASAP) and the International Nuclear Fuel Cycle Evaluation (INFCE) requires that the costs associated with processing of fuel materials for use in these cycles be estimated. The study described here provided unit cost estimates for the fabrication, reprocessing, and refabrication of a variety of fuels for several reactor systems. We examined in detail the facility requirements and operations to estimate capital and operating costs. Unit processing cost determinations were based on a cash flow analysis technique in which income from sales over the life of each facility was equated to the total capital and operating expenses of that facility plus a specified return on equity investment. The effects of plant capacities were determined by application of scaling factors to individual components of the reference plant costs. Capital and operating costs were estimated for 21 reactor and fuel cycle combinations. Based on these estimates, unit costs were determined for fabrication, reprocessing, and refabrication of the fuels. In each instance, the effect of plant capacities on unit costs associated with the processing of fuels was determined. All costs were based on mature industries, and first-of-a-kind costs were not included. Unit cost determinations were based on three financing techniques, which included government financing, typical industrial financing, and high-risk industrial financing. The unit costs recommended for the comparative analysis of power generation costs are those associaated with the economic assumptions of a typical industry.
Date: March 1, 1979
Creator: Olsen, A.R.; Judkins, R.R.; Carter, W.L. & Delene, J.G.
Partner: UNT Libraries Government Documents Department

Head-end reprocessing studies with irradiated high temperature gas-cooled reactor (HTGR) fuels

Description: Fifty (U-2.75 Th)C/sub 2/ and ThC/sub 2/ coated-particle fuel rods irradated in Peach Bottom were crushed and burned. The fertile and fissile fractions were separated using Thorex reagent and chemical analyses conducted for carbon, heavy metals, and fission products. Results were generally consistent with predictions, indicating that the reprocessing of TRISO-BISO fuel can be accomplished by the proposed flowsheet steps of crushing, fluidized-bed burning, coated particle separation and crushing, secondary burning, dissolution, clarification, and solvent extraction. (DLC)
Date: January 1, 1980
Creator: Fitzgerald, C.L. & Vaughen, V.C.A.
Partner: UNT Libraries Government Documents Department

Head-end reprocessing studies with irradiated HTGR-type fuels. III. Studies with RTE-7: TRISO UC$sub 2$-TRISO ThC$sub 2$

Description: Head-end reprocessing studies on irradiated Fort St. Vrain type HTGR fuel (TRISO coated UC$sub 2$-TRISO coated ThC$sub 2$) were performed. Fuel sticks were burned quiescently in oxygen at 750$sup 0$C, the ash was screened to separate the fissile SiC-coated particles from the fertile SiC-coated particles, each particle fraction was crushed and burned separately in oxygen, and the secondary burner ash products were leached. These studies concentrated on the release of radionuclides to the burner off-gases after passing through a 20-$mu$ m sintered metal frit held at 500$sup 0$C. Fission product and heavy metal inventories were made on the process gaseous, liquid, and solid streams. Experimental results are compared with pre-irradiation average fuel stick compositions and with calculated fission product inventories. (auth)
Date: November 1, 1975
Creator: Fitzgerald, C.L.; Vaughen, V.C.A.; Notz, K.J. & Lowrie, R.S.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor fuel recycle development. Annual progress report for period ending September 30, 1977

Description: The status of the following tasks is reported: program management, studies and analysis, fuel processing, refabrication development, in-plant waste treatment, research general support, and major facilities including HTGR recycle reference facility, hot engineering test facility and cold prototype test facility-refabrication. (JRD)
Date: September 1, 1978
Creator: Lotts, A.L. & Kasten, P.R.
Partner: UNT Libraries Government Documents Department