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1170-MW(t) HTGR-PS/C plant application study report: tar sands oil recovery application

Description: This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to tar sands oil recovery and upgrading. The raw product recovered from the sands is a heavy, sour bitumen; upgrading, which involves coking and hydrodesulfurization, produces a synthetic crude (refinable by current technology) and petroleum coke. Steam and electric power are required for the recovery and upgrading process. Proposed and commercial plants would purchase electric power from local utilities and obtain from boilers fired with coal and with by-product fuels produced by the upgrading. This study shows that an HTGR-PS/C represents a more economical source of steam and electric power.
Date: May 1, 1981
Creator: Rao, R. & McMain, Jr., A. T.
Partner: UNT Libraries Government Documents Department

1170-MW(t) HTGR-PS/C plant application study report: SRC-II process application

Description: The solvent refined coal (SRC-II) process is an advanced process being developed by Gulf Mineral Resources Ltd. (a Gulf Oil Corporation subsidiary) to produce a clean, non-polluting liquid fuel from high-sulfur bituminous coals. The SRC-II commercial plant will process about 24,300 tonnes (26,800 tons) of feed coal per stream day, producing primarily fuel oil plus secondary fuel gases. This summary report describes the integration of a high-temperature gas-cooled reactor operating in a process steam/cogeneration mode (HTGR-PS/C) to provide the energy requirements for the SRC-II process. The HTGR-PS/C plant was developed by General Atomic Company (GA) specifically for industries which require energy in the form of both steam and electricity. General Atomic has developed an 1170-MW(t) HTGR-PS/C design which is particularly well suited to industrial applications and is expected to have excellent cost benefits over other sources of energy.
Date: May 1, 1981
Creator: Rao, R. & McMain, A. T., Jr.
Partner: UNT Libraries Government Documents Department

Krypton absorption in liquid CO/sub 2/ (KALC): effects of the minor components N/sub 2/, CO, and Xe

Description: Results are presented for the fourth major campaign for quantifying krypton removal from simulated High-Temperature Gas-Cooled Reactor reprocessing off-gas by the Krypton Absorption in Liquid CO/sub 2/ (KALC) process. This process utilizes the high solubility of krypton in liquid CO/sub 2/. Mass transfer experiments for the absorption, fractionation, and stripping operations of the KALC process indicate that the addition of N/sub 2/ and CO do not alter the mass transfer characteristics exhibited by O/sub 2/ and krypton in the basic CO/sub 2/--O/sub 2/--Kr system. Decontamination factors for xenon in the absorber and stripper were several orders of magnitude less than those for krypton under similar conditions. Indications are that the fate of xenon is controlled by the heat input to the stripper reboiler. Experiments on the solubility of O/sub 2/ and CO indicate that CO is more soluble than O/sub 2/ at temperatures below -21/sup 0/C.
Date: February 1, 1979
Creator: Gilliam, T. M.; Fowler, V. L. & Inman, D. J.
Partner: UNT Libraries Government Documents Department

MEU/Th fuel cycle optimization for the Lead Plant

Description: The reference equilibrium cycle fuel composition for the Lead Plant was specified previously by a C/Th ratio of 850 and a fuel rod diameter of 1.17 cm, which is optimal for non-recycle operation and close to optimal for recycle of bred U-233. Subsequent work has emphasized the importance of full recycle of all discharged uranium to maintain the competitive advantage of the MEU/Th cycle. Cycles with full recycle optimize at higher thorium loadings and larger rod diameters. This is an additional benefit for core design and reduces fabrication problems. New optimization studies based on full recycle lead to an equilibrium cycle composition characterized by a C/Th ratio of 600 and a rod diameter of 1.35 cm. The average packing fraction of fuel particles in the rod is 0.43. The C/Th ratio for the initial core is 350, which can also be accommodated with the 1.35 cm rod diameter. Mass flow data for 30 year operation and fuel cycle cost data have been obtained for this cycle and for several other thorium loadings.
Date: December 1, 1978
Creator: Merrill, M.H. & Lane, R.K.
Partner: UNT Libraries Government Documents Department

Methods for very high temperature design

Description: Design rules and procedures for high-temperature, gas-cooled reactor components are being formulated as an ASME Boiler and Pressure Vessel Code Case. A draft of the Case, patterned after Code Case N-47, and limited to Inconel 617 and temperatures of 982/degree/C (1800/degree/F) or less, will be completed in 1989 for consideration by relevant Code committees. The purpose of this paper is to provide a synopsis of the significant differences between the draft Case and N-47, and to provide more complete accounts of the development of allowable stress and stress rupture values and the development of isochronous stress vs strain curves, in both of which Oak Ridge National Laboratory (ORNL) played a principal role. The isochronous curves, which represent average behavior for many heats of Inconel 617, were based in part on a unified constitutive model developed at ORNL. Details are also provided of this model of inelastic deformation behavior, which does not distinguish between rate-dependent plasticity and time-dependent creep, along with comparisons between calculated and observed results of tests conducted on a typical heat of Inconel 617 by the General Electric Company for the Department of Energy. 4 refs., 15 figs., 1 tab.
Date: January 1, 1989
Creator: Blass, J.J.; Corum, J.M. & Chang, S.J.
Partner: UNT Libraries Government Documents Department

Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

Description: Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the CB atmosphere temperature and pressure, as overpressurization is the most likely failure mode.
Date: January 1, 1985
Creator: Kroeger, P.G. & Chan, B.C.
Partner: UNT Libraries Government Documents Department

Applications of high-strength concrete to the development of the prestressed concrete reactor vessel (PCRV) design for an HTGR-SC/C plant

Description: The PCRV research and development program at ORNL consists of generic studies to provide technical support for ongoing PCRV-related studies, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities under this program have concentrated on the development of high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C plant, and the testing of models to both evaluate the behavior of high-strength concretes (plain and fibrous) and to develop model testing techniques. A test program to develop and evaluate high-strength (greater than or equal to 63.4 MPa) concretes utilizing materials from four sources which are in close proximity to potential sites for an HTGR plant is currently under way. The program consists of three phases. Phase I involves an evaluation of the cement, fly ash, admixtures and aggregate materials relative to their capability to produce concretes having the desired strength properties. Phase II is concerned with the evaluation of the effects of elevated temperatures (less than or equal to 316/sup 0/C) on the strength properties of mixes selected for detailed evaluation. Phase III involves a determination of the creep characteristics and thermal properties of the selected mixes. An overview of each of these phases is presented as well as results obtained to date under Phase I which is approximately 75% completed.
Date: January 1, 1984
Creator: Naus, D.J.
Partner: UNT Libraries Government Documents Department

Technical Division quarterly progress report, April 1--June 30, 1978

Description: Fuel cycle research and development: results are presented on fluidized-bed calcination and on post-treatment of commercial wastes; study was done on the use of microwave energy in processing wastes and on the use of bidentate compounds for separation of actinides from commercial power reactor reprocessing waste. Work on the krypton-85 storage development program, including the results of rubidium corrosion tests, is reported. In the HTGR fuel reprocessing section, the results of x-ray and Auger spectroscopy analysis of CO oxidation catalyst are reported. Special materials production: the long-term management of high-level ICPP wastes is reported: development of a calcine pelletizing pilot plant, actinide removal, actinide extraction by DHDECMP, and calcined solids retrieval and handling. Design work was completed for the fluorinel pilot-plant upgrade. Other development results reported are on the progress of the Rover plant, and on flowsheet development for electrolytic and second-cycle waste, for Fluorinel waste, and for Tank WM-183. Other results reported include: assistance to the Waste Calcining Facility and to the New Waste Calcining Facility, methods for the monitoring of gaseous effluents, and a mathematical model to describe chloride buildup in the waste calcining scrubbing solution. Other projects supporting energy developments: results are reported on nuclear materials safety, the installation and operation of a geothermal fluidized-bed dryer, the in-plant source-term measurement at the Turkey Point station, burnup methods for fast breeder reactor fuels, absolute thermal fission yields, analytical support to light-water breeder reactor developments, cerium analysis of actinide removal project solutions, a spark source mass spectrometric computer program, and on environmental iodine species behavior.
Date: December 1, 1978
Creator: Plung, D.L. (ed.)
Partner: UNT Libraries Government Documents Department

Nondestructive assay of green HTGR fuel rods

Description: This report describes the nondestructive (NDA) work done at Los Alamos during 1979 and 1980 as part of the New Brunswick Laboratory-sponsored evaluation of NDA of the uranium content of fabricated fuel rods for high-temperature gas-cooled reactors (HTGR). The methods used (delayed neutron and passive gamma ray) are concisely described, and the results are summarized and compared in graphical and tabular form. The results indicate that, with the use of proper physical standards, accuracies within about 1 percent should be achievable by NDA procedures.
Date: May 1, 1981
Creator: Barschall, H.H.; Meier, M.M. & Parker, J.L.
Partner: UNT Libraries Government Documents Department

Measurement and modelling of postirradiation fission product release from HTGR fuel particles under accident conditions

Description: A study was performed to provide a description of the release of fission products from failed fuel particles during a core heatup event in an HTGR. The need for this study was established in the Accident Initiation and Progression Analysis program. The release of fission products was measured from laser-failed BISO ThO/sub 2/, TRISO UC/sub 2/, and weak acid resin (WAR) particles over a range of burnups. The burnups were 0.25, 1.4 and 15.7% FIMA for ThO/sub 2/ particles, 23.5 and 74% FIMA for UC/sub 2/ particles, and 60% FIMA for WAR particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium. Two types of experiments were performed: isothermal and temperature rise experiments. The range of the temperatures was from 1200/sup 0/ to 2300/sup 0/C. In the temperature rise experiments, the heating rates were between 50/sup 0/ and 450/sup 0/C/h.
Date: December 1, 1978
Creator: Myers, B.F. & Morrissey, R.E.
Partner: UNT Libraries Government Documents Department

Strategy for the practical utilization of thorium fuel cycles

Description: There has been increasing interest in the utilization of thorium fuel cycles in nuclear power reactors for the past few years. This is due to a number of factors, the chief being the recent emphasis given to increasing the proliferation resistance of reactor fuel cycles and the thorium cycle characteristic that bred /sup 233/U can be denatured with /sup 238/U (further, a high radioactivity is associated with recycle /sup 233/U, which increases fuel diversion resistance). Another important factor influencing interest in thorium fuel cycles is the increasing cost of U/sub 3/O/sub 8/ ores leading to more emphasis being placed on obtaining higher fuel conversion ratios in thermal reactor systems, and the fact that thorium fuel cycles have higher fuel conversion ratios in thermal reactors than do uranium fuel cycles. Finally, there is increasing information which indicates that fast breeder reactors have significantly higher capital costs than do thermal reactors, such that there is an economic advantage in the long term to have combinations of fast breeder reactors and high-conversion thermal reactors operating together. Overall, it appears that the practical, early utilization of thorium fuel cycles in power reactors requires commercialization of HTGRs operating first on stowaway fuel cycles, followed by thorium fuel recycle. In the longer term, thorium utilization involves use of thorium blankets in fast breeder reactors, in combination with recycling the bred /sup 233/U to HTGRs (preferably), or to other thermal reactors.
Date: January 1, 1978
Creator: Kasten, P.R.
Partner: UNT Libraries Government Documents Department

Structure interaction due to thermal bowing of shrouds in steam generator of gas-cooled reactor

Description: The design of the gas-cooled reactor steam generators includes a tube bundle support plate system which restrains and supports the helical tubes in the steam generator. The support system consists of an array of radially oriented, perforated plates through which the helical tube coils are wound. These support plates have tabs on their edges which fit into vertical slots in the inner and outer shrouds. When the helical tube bundle and support plates are installed in the steam generator, they most likely cannot fit evenly between the inner and outer shrouds. This imperfection leads to different gaps between two extreme sides of the tube bundle and the shrouds. With different gaps through the tube bundle height, the helium flow experiences different cooling effects from the tube bundle. Hence, the temperature distribution in the shrouds will be non-uniform circumferentially since their surrounding helium flow temperatures are varied. These non-uniform temperatures in the shrouds result in the phenomenon of thermal bowing of shrouds.
Date: January 1, 1981
Creator: Woo, H.H.
Partner: UNT Libraries Government Documents Department

Nondestructive examination of 51 fuel and reflector elements from Fort St. Vrain Core Segment 1

Description: Fifty-one fuel and reflector elements irradiated in core segment 1 of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR) were inspected dimensionally and visually in the Hot Service Facility at Fort St. Vrain in July 1979. Time- and volume-averaged graphite temperatures for the examined fuel elements ranged from approx. 400/sup 0/ to 750/sup 0/C. Fast neutron fluences varied from approx. 0.3 x 10/sup 25/ n/m/sup 2/ to 1.0 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Nearly all of the examined elements shrank in both axial and radial dimensions. The measured data were compared with strain and bow predictions obtained from SURVEY/STRESS, a computer code that employs viscoelastic beam theory to calculate stresses and deformations in HTGR fuel elements.
Date: December 1, 1980
Creator: Miller, C.M. & Saurwein, J.J.
Partner: UNT Libraries Government Documents Department

Size effect on the irradiation performance of coated fuel particles

Description: Outer coatings that were as near alike as possible were applied to two different sizes of inert TRISO particles that were larger than those commonly used to fuel HTGR reactors, and these particles were then irradiated in a test reactor to observe the influence of particle size on outer coating failures that resulted from irradiation-induced shrinkage of coatings onto the more stable SiC substrates over which they were applied. Outer coatings of plain pyrocarbon and of Si-alloyed pyrocarbon were used to make up two test pairs of particles with diameters of about 1050 ..mu..m and 1300 ..mu..m. For a fast-neutron fluence of 5.5 x 10/sup 25/ n/m/sup 2/ (E > 29fJ) at an irradiation temperature of 1125 K, failure was about twice as high in the larger 1300 ..mu..m particle of each test pair as in the smaller 1050 ..mu..m particle (16% versus 8%), with each of the coating types having roughly the same behavior.
Date: September 1, 1980
Creator: Bullock, R.E.
Partner: UNT Libraries Government Documents Department

Irradiation performance of HTGR fuel in HFIR capsule HT-31

Description: The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600/sup 0/C, fast neutron fluences (0.18 MeV) up to 9 x 10/sup 25/ n/m/sup 2/, and burnups up to 8.9% FIMA for ThO/sub 2/ particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO/sub 2/. Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO/sub 2/ particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200/sup 0/C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles.
Date: May 1, 1979
Creator: Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B. et al.
Partner: UNT Libraries Government Documents Department

Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4

Description: Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized-water-reactor steam-generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models being developed to provide better digital codes to compute the bahavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.
Date: April 1, 1983
Creator: Edler, S.K. (ed.)
Partner: UNT Libraries Government Documents Department

Applications of power beaming from space-based nuclear power stations. [Laser beaming to airplanes; microwave beaming to ground]

Description: Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000/sup 0/K and a liquid drop radiator to reject heat at temperatures of approx. 500/sup 0/K. Higher RBR coolant temperatures (up to approx. 3000/sup 0/K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beaming to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel.
Date: January 1, 1981
Creator: Powell, J.R.; Botts, T.E. & Hertzberg, A.
Partner: UNT Libraries Government Documents Department

Postirradiation examination and evaluation of Peach Bottom fuel test elements FTE-14 and FTE-15

Description: Peach Bottom fuel test elements FTE-14 and FTE-15 were companion nonaccelerated tests of fuel rods and fuel particles representative of the Large High-Temperature Gas-Cooled Reactor (LHTGR). The purpose of the tests was to broaden the data base of H-327 graphite and various fuel types; specifically, UO/sub 2/, UC/sub 2/, weak acid resin UC/sub x//O/sub y/, and several fertile fuel types were tested. The irradiation reached peak fuel temperatures of 1600/sup 0/C volume- and time-averaged temperatures of 1300/sup 0/C, and fast fluence exposures up to 2 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Experimental results were compared with predictions based on accelerated irradiation tests, postirradiation heating, and other Peach Bottom test elements to validate HTGR design codes. The nuclear design predictions were modified by measurements which allowed the verification of thermal design calculations and thermocouple readings.
Date: February 1, 1979
Creator: Holzgraf, J.F.; McCord, F.; Miller, C.M.; Norman, B.L.; Saurwein, J.J. & Wallroth, C.F.
Partner: UNT Libraries Government Documents Department

Postirradiation examination and evaluation of Fort St. Vrain fuel element 1-0743

Description: Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680/sup 0/C, fissile and fertile particle burnups of approximately 6.2% and 0.3%, respectively, and a total burnup of 12,210 MWd/tonne. The postirradiation examination revealed that the element was in excellent condition. No cracks were observed on any of the element surfaces. The structural integrity of the fuel rods was good. No evidence of mechanical interaction between the fuel rods and fuel body was observed. Calculated irradiation parameters obtained with HTGR design codes were compared with measured data. Radial and axial power distributions, irradiation temperatures, neutron fluences, and fuel burnups were in good agreement with measurements. Calculated fuel rod strains were about a factor of three greater than were observed.
Date: May 1, 1981
Creator: Saurwein, J.J.; Miller, C.M. & Young, C.A.
Partner: UNT Libraries Government Documents Department

Postirradiation examination of recycle test elements from the Peach Bottom Reactor

Description: The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.
Date: December 1, 1978
Creator: Tiegs, T.N. & Long, E.L. Jr.
Partner: UNT Libraries Government Documents Department

Postirradiation thermal analysis of capsule P13T. [HTGR]

Description: In determining fuel rod temperature histories for the P13T capsule, a technique which combined measured temperature and dimensional data, TAC-2D computer modeling, and a calculational procedure was employed. TAC-2D models were constructed for each of the capsule's four fuel bodies and temperature matching runs were made at five time points of the irradiation history. The agreement between TAC-calculated and measured temperatures was good; at all times the TAC-calculated temperatures were within 20/sup 0/C of the Chromel-Alumel (C/A) measurements and 40/sup 0/C of the corrected tungsten-rhenium (W/Re) temperatures. Thermocouple decalibration was treated in detail and corrected temperatures for all W/Re thermocouples were calculated over the irradiation period.
Date: December 1, 1978
Creator: Ketterer, J.W.
Partner: UNT Libraries Government Documents Department

Chemical reactions in the helium impurities loop. [HTGR]

Description: The Helium Impurities Loop (HIL) at Brookhaven National Laboratory has been run to study reactions between the three metals and the four major helium oxidation and reduction and on carburization of the metals. Preliminary work on hydrogen diffusion through the loop walls and on hydrogen retention in the walls is also presented. During the past year much experience has been accumulated on loop operation and instrument reliability. A small ''bench test'' loop has been built and operated to study reactions under more controlled conditions than possible in the HIL. These efforts are in preparation for more quantitative experiments that will be performed during the coming year.
Date: January 1, 1978
Creator: Epel, L.G. & Schweitzer, D.G.
Partner: UNT Libraries Government Documents Department

Radionuclide distributions and sorption behavior in the Susquehanna--Chesapeake Bay System

Description: Radionuclides released into the Susquehanna--Chesapeake System from the Three Mile Island, Peach Bottom, and Calvert Cliffs nuclear power plants are partitioned among dissolved, particulate, and biological phases and may thus exist in a number of physical and chemical forms. In this project, we have measured the dissolved and particulate distributions of fallout /sup 137/Cs; reactor-released /sup 137/Cs, /sup 134/Cs, /sup 65/Zn, /sup 60/Co, and /sup 58/Co; and naturally occurring /sup 7/Be and /sup 210/Pb in the lower Susquehanna River and Upper Chesapeake Bay. In addition, we chemically leached suspended particles and bottom sediments in the laboratory to determine radionuclide partitioning among different particulate-sorbing phases to complement the site-specific field data. This information has been used to document the important geochemical processes that affect the transport, sorption, distribution, and fate of reactor-released radionuclides (and by analogy, other trace contaminants) in this river-estuarine system. Knowledge of the mechanisms, kinetic factors, and processes that affect radionuclide distributions is crucial for predicting their biological availability, toxicity, chemical behavior, physical transport, and accumulation in aquatic systems. The results from this project provide the information necessary for developing accurate radionuclide-transport and biological-uptake models. 76 refs., 12 figs.
Date: January 1, 1989
Creator: Olsen, C.R.; Larsen, I.L.; Lowry, P.D.; McLean, R.I. & Domotor, S.L.
Partner: UNT Libraries Government Documents Department

Reaction of uranyl nitrate solutions with carboxylic acid cation exchange resins at 30, 40, and 50/sup 0/C. [Amberlite IRC-72; Duolite C-464]

Description: An investigation of the reaction of uranyl nitrate solution with Amberlite IRC-72 and with Duolite C-464 resin was conducted under equilibrium conditions for application to a high-temperature gas-cooled reactor fuel refabrication facility. Test series were conducted at constant nitrate ion concentrations in which the uranyl ion distribution between the phases was varied from values corresponding to approximately 10% resin loading to those for essentially complete resin loading. Equilibrium quotients, calculated as the metathetical exchange of uranyl ion for replaceable hydrogen in the resin phase, were calculated for each test series over the nitrate ion concentration range of 0.2 to 2.0 N and at temperatures of 30, 40, and 50/sup 0/C. These values varied with nitrate ion concentration but were relatively insensitive to changes in temperature.
Date: December 1, 1978
Creator: Shaffer, J. H. & Greene, C. W.
Partner: UNT Libraries Government Documents Department