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Development of processes and equipment for the refabrication of HTGR fuels

Description: Refabrication is in the step in the HTGR thorium fuel cycle that begins with a nitrate solution containing /sup 238/U and culminates in the assembly of this material into fuel elements for use in an HTGR. Refabrication of HTGR fuel is essentially a manufacturing operation and consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and SiC, preparation of fuel rods, and assembly of fuel rods in fuel elements. All the equipment for refabrication of /sup 238/U-containing fuel must be designed for completely remote operation and maintenance in hot cell facilities. This paper describes the status of processes and equipment development for the remote refabrication of HTGR fuels. The feasibility of HTGR refabrication processes has been proven by laboratory development. Engineering-scale development is now being performed on a unit basis on the majority of the major equipment items. Engineering-scale equipment described includes full-scale resin loading equipment, a 5-in.-dia (0.13-m) microsphere coating furnace, a fuel rod forming machine, and a cure-in-place furnace.
Date: June 1, 1976
Creator: Sease, J. D. & Lotts, A. L.
Partner: UNT Libraries Government Documents Department

Drying of uranium-loaded cation exchange resin with microwave heating

Description: The reference fuel kernel for recycle of /sup 233/U to HTGRs (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation exchange resin with uranium and carbonizing it at controlled conditions. The wet, uranium-loaded resin must be dried to a water content of 10 to 16 wt percent prior to carbonization to minimize handling problems. Microwave heating was demonstrated to give controlled and reproducible dried resin in a vessel whose dimensions were safe for nuclear criticality (12.4 cm ID). A standard drying procedure was developed. The duration of microwave heating is controlled either by using an experimentally derived drying factor or by monitoring the amount of water removed from the resin. No significant difficulties were encountered in the operation of the dryer. Heat balance, microwave coupling efficiency, and minimum fluidization velocity were calculated and compared with experimental results or with the literature. Mixing of the resin during drying was required to ensure a uniformly dried product.
Date: December 1, 1976
Creator: Drago, J. P. & Haas, P. A.
Partner: UNT Libraries Government Documents Department

Updated projections of radioactive wastes to be generated by the U. S. nuclear power industry

Description: Eleven types of radioactive wastes to be generated within the fuel cycle operations of the U.S. nuclear power industry are defined, and projections are presented of their annual generation rates, shipping requirements, and accumulated characteristics over the remainder of this century. The power reactor complex is assumed to consist of uranium- and plutonium-fueled LWRs, HTGRs, and LMFBRs, and the installed nuclear electric capacity of the U.S. is taken as 68.1, 252, and 510 GW at the ends of calendar years 1980, 1990, and 2000, respectively. 72 tables.
Date: December 1, 1976
Creator: Kee, C. W.; Croft, A. G. & Blomeke, J. O.
Partner: UNT Libraries Government Documents Department

Application of Hastelloy X in gas-cooled reactor systems

Description: Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given.
Date: October 1, 1976
Creator: Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W. R.; Strizak, J. P.; Lystrup, A. & DiStefano, J. R.
Partner: UNT Libraries Government Documents Department

Assessment of very high-temperature reactors in process applications

Description: An overview is presented of the technical and economic feasibility for the development of a very high-temperature reactor (VHTR) and associated processes. A critical evaluation of VHTR technology for process temperatures of 1400 and 2000/sup 0/F is made. Additionally, an assessment of potential market impact is made to determine the commercial viability of the reactor system. It is concluded that VHTR process heat in the range of 1400 to 1500/sup 0/F is attainable with near-term technology. However, process heat in excess of 1600/sup 0/F would require considerably more materials development. The potential for the VHTR could include a major contribution to synthetic fuel, hydrogen, steel, and fertilizer production and to systems for transport and storage of high-temperature heat. A recommended development program including projected costs is presented.
Date: November 1, 1976
Creator: Spiewak, I.; Jones, J. E. Jr.; Gambill, W. R. & Fox, E. C.
Partner: UNT Libraries Government Documents Department

ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

Description: ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core.
Date: August 10, 1977
Creator: Cleveland, J. C.; Hedrick, R. A.; Ball, S. J. & Delene, J. G.
Partner: UNT Libraries Government Documents Department

Monthly highlights for Office of Nuclear Regulatory Research programs at Oak Ridge National Laboratory

Description: Technical highlights and cost/budget data are presented for the following programs: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, and design criteria for piping and nozzles. (DG)
Date: March 1, 1976
Creator: Fee, G. G. (comp.)
Partner: UNT Libraries Government Documents Department

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1976

Description: Technical highlights are presented for the following activities: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, Zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and dose conversion factors for inhalation of radionuclides.
Date: October 1, 1976
Creator: Fee, G. G. (comp.)
Partner: UNT Libraries Government Documents Department

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

Description: Brief highlights are presented for the following activities: heavy section steel technology program, fission product ..beta.. and ..gamma.. energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, and design criteria for piping and nozzles.
Date: February 1, 1976
Creator: Fee, G. G. (comp.)
Partner: UNT Libraries Government Documents Department

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

Description: Brief highlights are presented for the following programs: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis, design criteria for piping and nozzles, and dose conversion factors for inhalation of radionuclides.
Date: August 1, 1976
Creator: Fee, G. G. (comp.)
Partner: UNT Libraries Government Documents Department

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1977

Description: Technical highlights are presented for the following safety-related studies: heavy section steel technology, fission product beta and gamma energy release, fission product release from LWR fuel, fission product transport tests, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and noise diagnostics for safety assessment.
Date: September 13, 1977
Creator: Fee, G. G. (comp.)
Partner: UNT Libraries Government Documents Department

HTCAP: a FORTRAN IV program for calculating coated-particle operating temperatures in HFIR target irradiation experiments. [HTGR]

Description: A description is presented of HTCAP, a computer code that calculates in-reactor operating temperatures of loose coated ThO/sub 2/ particles in the HFIR target series of irradiation tests. Three computational models are employed to determine the following: (1) fission heat generation rates, (2) capsule heat transfer analysis, and (3) maximum particle surface temperature within the design of an HT capsule. Maximum particle operating temperatures are calculated at daily intervals during each irradiation cycle. The application of HTCAP to sleeve CP-62 of HT-15 is discussed, and the results are compared with those obtained in an earlier thermal analysis on the same capsule. Agreement is generally within +-5 percent, while decreasing the computational time by more than an order of magnitude. A complete FORTRAN listing and summary of required input data are presented in appendices. Included is a listing of the input data and a tabular output from the thermal analysis of sleeve CP-62 of HT-15.
Date: May 1, 1976
Creator: Kania, M. J.
Partner: UNT Libraries Government Documents Department

In-line monitoring of effluents from HTGR fuel particle preparation processes using a time-of-flight mass spectrometer

Description: The carbonization, conversion, and coating processes in the manufacture of HTGR fuel particles have been studied with the use of a time-of-flight mass spectrometer. Non-condensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes which have been monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO/sub 2/ to UC/sub 2/, buffer and low temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures.
Date: August 1, 1976
Creator: Lee, D. A.; Costanzo, D. A.; Stinton, D. P.; Carpenter, J. A.; Rainey, W. T. Jr.; Canada, D. C. et al.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

Description: Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.
Date: July 1, 1977
Creator: Cole, T. E.; Sanders, J. P. & Kasten, P. R.
Partner: UNT Libraries Government Documents Department

HTGR fuel and fuel cycle technology

Description: The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.
Date: August 1, 1976
Creator: Lotts, A. L. & Coobs, J. H.
Partner: UNT Libraries Government Documents Department

HTGR steam generator modeling

Description: Research activities at The University of Tennessee on gas cooled reactor dynamics are described. The main activity is on steam generator modeling, using approaches ranging from a relatively simple linear representation to a detailed nonlinear representation. Model comparisons will involve simulations of the Fort St. Vrain reactor steam generator, with emphasis on the evaluation of accuracy vs computation costs. A smaller effort is also in progress for modeling the reactor core, the main turbine and the blower turbines. Preparations are described for using test data from Fort St. Vrain for validating the dynamic models and identifying important design parameters in the plant.
Date: July 1, 1976
Creator: Kerlin, T. W.
Partner: UNT Libraries Government Documents Department

Computer model for the KALC process studies in the ORGDP Off-Gas Decontamination Pilot Plant. [Krypton adsorption in liquid carbon dioxide]

Description: A computer model of the KALC process is presented for the equipment configuration in use during HTGR off-gas studies at the ORGDP Off-Gas Decontamination Pilot Plant. The model is tailored to require input routinely available during such experimental studies. A program is included to provide McCabe-Thiele plots as an additional convenience.
Date: September 1, 1976
Creator: Glass, R. W. & Barker, R. E.
Partner: UNT Libraries Government Documents Department

HEATUP: a computer program for the thermal anaysis of a LOFC accident in an HTGR

Description: The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short review of preliminary results related to the choice of thermal properties, boundary conditions and initial conditions; a full description of a typical three-dimensional R-Z model and a limited one of a two-dimensional RZ model. HEATUP's additional computations are presented together with the method of input preparation. The three-dimensional model of the Fulton Generating Station Loss of Forced Circulation accident is used as a sample problem. A complete presentation of the input data is made. Also, the computer printout of the sample problem input data and results are given.
Date: November 1, 1976
Creator: Siman-Tov, I. I. & Turner, W. D.
Partner: UNT Libraries Government Documents Department

FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

Description: FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm.
Date: June 1, 1976
Creator: Paul, D. D.
Partner: UNT Libraries Government Documents Department

Coal technology program quarterly progress report for the period ending December 31, 1975

Description: Installation of the bench-scale hydrocarbonization research system was completed and initial pressurization was followed by heat up, leak checking, and testing of instrumentation. Shakedown runs were completed using 5 and 20 lb of pulverized coal as feed and pyrolyzing the coal at approx. 1050/sup 0/F under 5 atm of nitrogen pressure. Some excellent results have been obtained in the additive agglomeration studies using liquid additives designed for the petroleum industry. Characterization studies of SRC oils continue using a number of methods. Studies have continued on several aspects of the physical structure of coal, the sorption and desorption of water vapor on a cobalt--molybdenum catalyst, and the catalytic chemistry of the same catalyst. Synthoil and COED process oil samples have been chemically fractionated for biological and chemical characterization. Development continues on achieving cleaner fractionation of chemical classes than has previously been afforded by more conventional schemes. Studies have begun to characterize aqueous effluents from the synthane process and in-situ shale oil retorting. Engineering evaluations of several processes are reported. Investigation into the fate and availability of a model PAH compound in water has begun. Acquisition of literature on the bioaccumulation and toxicity of trace elements released in aqueous effluents from coal conversion processes has been completed. In the engineering evaluations of nuclear process heat for coal conversion, preliminary design of a plant for steam gasifying coal with VHTR heat has been completed. (auth)
Date: March 1, 1976
Partner: UNT Libraries Government Documents Department

CACHE: an extended BASIC program which computes the performance of shell and tube heat exchangers. [HTGR]

Description: An extended BASIC program, CACHE, has been written to calculate steady state heat exchange rates in the core auxiliary heat exchangers, (CAHE), designed to remove afterheat from High-Temperature Gas-Cooled Reactors (HTGR). Computationally, these are unbaffled counterflow shell and tube heat exchangers. The computational method is straightforward. The exchanger is subdivided into a user-selected number of lengthwise segments; heat exchange in each segment is calculated in sequence and summed. The program takes the temperature dependencies of all thermal conductivities, viscosities and heat capacities into account providing these are expressed algebraically. CACHE is easily adapted to compute steady state heat exchange rates in any unbaffled counterflow exchanger. As now used, CACHE calculates heat removal by liquid weight from high-temperature helium and helium mixed with nitrogen, oxygen and carbon monoxide. A second program, FULTN, is described. FULTN computes the geometrical parameters required as input to CACHE. As reported herein, FULTN computes the internal dimensions of the Fulton Station CAHE. The two programs are chained to operate as one. Complete user information is supplied. The basic equations, variable lists, annotated program lists, and sample outputs with explanatory notes are included.
Date: March 1, 1976
Creator: Tallackson, J. R.
Partner: UNT Libraries Government Documents Department

Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

Description: In April 1974, the U.S. Atomic Energy Commission (now the Energy Research and Development Administration (ERDA)) authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093/sup 0/C (1200, 1400, 1600, 1800, and 2000/sup 0/F). There are a number of large industrial process heat applications that could utilize the VHTR.
Date: December 1, 1976
Creator: Jones, J. E. Jr. & Spiewak, I.
Partner: UNT Libraries Government Documents Department