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0. 20-m (8-in.) Primary Burner Development Report

Description: High-Temperature Gas-Cooled Reactors (HTGRs) utilize graphite-base fuels. Fluidized-bed burners are being employed successfully in the experimental reprocessing of these fuels. The primary fluidized-bed burner is a unit operation in the reprocessing flowsheet in which the graphite moderator is removed. A detailed description of the development status of the 0.20-m (8-in.) diameter primary fluidized-bed burner as of July 1, 1977 is presented. Experimental work to date performed in 0.10; 0.20; and 0.40-m (4, 8, and 16 in.) diameter primary burners has demonstrated the feasibility of the primary burning process and, at the same time, has defined more clearly the areas in which additional experimental work is required. The design and recent operating history of the 0.20-m-diameter burner are discussed, with emphasis placed upon the evolution of the current design and operating philosophy.
Date: December 1977
Creator: Stula, R. T.; Young, D. T. & Rode, J. S.
Partner: UNT Libraries Government Documents Department

350 MW(t) design fuel cycle selection. Revision 1

Description: This document discusses the results of this evaluation and a recommendation to retain the graded fuel cycle in which one-half of the fuel elements are exchanged at each refueling. This recommendation is based on the better performance of the graded cycle relative to the evaluation criteria of both economics and control margin. A choice to retain the graded cycle and a power density of 5.9 MW/m{sup 3} for the upcoming conceptual design phase was deemed prudent for the following reasons: the graded cycle has significantly better economics, and essentially the same expected availability factor as the batch design, when both are evaluated against the same requirements, including water ingress; and the reduction in maximum fuel pin power peaking in the batch design compared to the graded cycle is only a few percent and gas hot streaks are not improved by changing to a batch cycle. The preliminary 2-D power distribution studies for both designs showed that maximum fuel pin power peaking, particularly near the inner reflector, was high for both designs and nearly the same in magnitude. 10 figs., 9 tabs.
Date: January 1986
Creator: Lane, R. K.; Lefler, W. & Shirley, G.
Partner: UNT Libraries Government Documents Department

350 MW(t) MHTGR preassembly and modularization

Description: The Modular High Temperature Gas Cooled Reactor (MHTGR) provides a safe and economical nuclear power option for the world's electrical generation needs by the turn of the century. The proposed MHTGR plant is composed of four 350 MW(t) prismatic core reactor modules, coupled to a 2(2 {times} 1) turbine generator producing a net plant electrical output of 538 MW(e). Each of the four reactor module is located in a below-ground level concrete silo, and consists of a reactor vessel and a steam generator vessel interconnected by a cross duct vessel. The modules, along with the service buildings, are contained within a Nuclear Island (NI). The turbine generators and power generation facilities are in the non-nuclear Energy Conversion Area (ECA). The MHTGR design reduces cost and improves schedule by maximizing shop fabrication, minimizing field fit up of the Reactor Internals components and modularizing the NI ECA facilities. 3 refs., 6 figs., 2 tabs.
Date: May 1, 1991
Creator: Venkatesh, M.C. (General Atomics, San Diego, CA (USA)); Jones, G. (Gas-Cooled Reactor Associates, San Diego, CA (USA)); Dilling, D.A. (Bechtel International Corp., San Francisco, CA (USA)) & Parker, W.J. (Stone and Webster Engineering Corp., Boston, MA (USA))
Partner: UNT Libraries Government Documents Department

800-MWe HTGR-gas-turbine two-loop electric plant

Description: The design and development of the electric plant for the HTGR-Gas Turbine (GT) present a radical departure in accommodating the electrical requirements of the nuclear heat system and the balance of plant. To address not only these requirements but also codes and standards (for power plants and safety-grade systems), the evolution of the design dictates the application of innovative and unique solutions. The complexity of the GT plant and the criteria for nuclear power stations are substantial restraints and limit the availability of alternatives normally present in the development of a design and configuration. Objective of this report is to acquaint the reader with the electric design and to highlight those aspects of the electric plant which are unique. In addition, some insight will be presented on the impact of the plant configuration evolution on electric systems.
Date: June 1, 1980
Partner: UNT Libraries Government Documents Department

850/sup 0/C VHTR plant technical description

Description: This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 950/sup 0/C (1742/sup 0/F) core outlet temperature.
Date: June 1, 1980
Partner: UNT Libraries Government Documents Department

1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

Description: A conceptual design and cost study is presented for intermediate size high temperature gas-cooled reactor (HTGR) for industrial energy applications performed by United Engineers and Constructors Inc., (UE and C) and The General Atomic Company (GAC). The study is part of a program at ORNL and has the objective to provide support in the evaluation of the technical and economic feasibility of a single unit 1170 MW/sub t/ HTGR steam cycle cogeneration plant (referred to as the Steamer plant) for the production of industrial process energy. Inherent in the achievement of this objective, it was essential to perform a number of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.
Date: August 1, 1980
Partner: UNT Libraries Government Documents Department

1170-MW(t) HTGR-PS/C plant application-study report: alumina-plant application

Description: This report considers the HTGR-PS/C application to producing alumina from bauxite. For the size alumina plant considered, the 1170-MW(t) HTGR-PS/C supplies 100% of the process steam and electrical power requirements and produces surplus electrical power and/or process steam, which can be used for other process users or electrical power production. Presently, the bauxite ore is reduced to alumina in plants geographically separated from the electrolysis plant. The electrolysis plants are located near economical electric power sources. However, with the integration of an 1170-MW(t) HTGR-PS/C unit in a commercial alumina plant, the excess electric power available (approx. 233 MW(e)) could be used for alumina electrolysis.
Date: May 1, 1981
Creator: Rao, R.; McMain, A.T. Jr. & Stanley, J.D.
Partner: UNT Libraries Government Documents Department

1170-MW(t) HTGR-PS/C plant application study report: Geismar, Louisiana refinery/chemical complex 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 an industrial complex at Geismar, Louisiana. This study compares the HTGR with coal and oil as process plant fuels. This study uses a previous broad energy alternative study by the Stone and Webster Corporation on refinery and chemical plant needs in the Gulf States Utilities service area. The HTGR-PS/C was developed by General Atomic (GA) specifically for industries which require both steam and electric energy. The GA 1170-MW(t) HTGR-PC/C design is particularly well suited to industrial applications and is expected to have excellent cost benefits over other energy sources.
Date: May 1, 1981
Creator: McMain, Jr., A. T. & Stanley, J. D.
Partner: UNT Libraries Government Documents Department

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

Description: The US has large shale oil energy resources, and many companies have undertaken considerable effort to develop economical means to extract this oil within environmental constraints. The recoverable shale oil reserves in the US amount to 160 x 10/sup 9/ m/sup 3/ (1000 x 10/sup 9/ bbl) and are second in quantity only to coal. This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to a shale oil recovery process. Since the highest potential shale oil reserves lie in th Piceance Basin of Western Colorado, the study centers on exploiting shale oil in this region.
Date: May 1, 1981
Creator: Rao, R. & McMain, A.T. Jr.
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

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

3000 MW(t) HTGR - gas turbine non-intercoolel. Technical evaluation report

Description: This report summarizes all the technical work performed on the 3000-MW(t) 3-loop High-Temperature Gas-Cooled Reactor Gas Turbine design as of June 1979. Although the plant configuration has changed to a 2000-MW(t) 2-loop plant, most of the technical assessments described in this report are still applicable to the 2000-MW(t) plant. The report covers the criteria under which the plant was designed, the technical feasibility problems associated with the plant and their potential solutions, and other potential applications and improvements which could make the gas turbine concept more attractive economically.
Date: December 1, 1979
Partner: UNT Libraries Government Documents Department

Accelerator breeders: will they replace liquid metal fast breeders

Description: Investigation of accelerator breeders at Brookhaven National Laboratory indicate that the AB-LWR fuel cycle is economically competitive with the LMFBR fuel cycle. The same can be said about the accelerator breeder-High Temperature Gas Reactor symbiosis. This system appears to be very competitive with the added real advantage of superior safety and proliferation resistance. This discussion would be incomplete if the real competitor to accelerator breeding was not mentioned, namely Fusion Hybrid Breeding (FHB). Fusion Hybrid Breeding is a nearer option than pure fusion, as the breakeven Q value requirements are much more modest. Fusion Hybrid Breeding, if successful and practical, has the potential for highly efficient fissile fuel breeding, leading to cheaper fuel. The system, however, has yet to be demonstrated scientifically and to be shown commercially feasible. This is in contrast with the AB system which is an extension of proven, state-of-the-art technology with implementation possible within twenty years. 25 references, 4 figures, 5 tables.
Date: June 1, 1983
Creator: Grand, P.; Powell, J.R.; Steinberg, M. & Takahashi, H.
Partner: UNT Libraries Government Documents Department

Accident simulation and consequence analysis in support of MHTGR safety evaluations

Description: This paper summarizes research performed at Oak Ridge National Laboratory (ORNL) to assist the Nuclear Regulatory Commission (NRC) in preliminary determinations of licensability of the US Department of Energy (DOE) reference design of a standard modular high-temperature gas-cooled reactor (MHTGR). The work described includes independent analyses of core heatup and steam ingress accidents, and the reviews and analyses of fuel performance and fission product transport technology.
Date: January 1, 1991
Creator: Ball, S.J.; Wichner, R.P.; Smith, O.L.; Conklin, J.C. (Oak Ridge National Lab., TN (United States)) & Barthold, W.P. (Barthold Associates, Inc., Knoxville, TN (United States))
Partner: UNT Libraries Government Documents Department

Accuracy of finite-element models for the stress analysis of multiple-holed moderator blocks

Description: Two steps have been taken to quantify and improve the accuracy in the analysis. First, the limitations of various approximation techniques have been studied with the aid of smaller benchmark problems containing fewer holes. Second, a new family of computer programs has been developed for handling such large problems. This paper describes the accuracy studies and the benchmark problems. A review is given of some proposed modeling techniques including local mesh refinement, homogenization, a special-purpose finite element, and substructuring. Some limitations of these approaches are discussed. The new finite element programs and the features that contribute to their efficiency are discussed. These include a standard architecture for out-of-core data processing and an equation solver that operates on a peripheral array processor. The central conclusions of the paper are: (1) modeling approximation methods such as local mesh refinement and homogenization tend to be unreliable, and they should be justified by a fine mesh benchmark analysis; and (2) finite element codes are now available that can achieve accurate solutions at a reasonable cost, and there is no longer a need to employ modeling approximations in the two-dimensional analysis of HTGR fuel elements. 10 figures.
Date: January 1, 1981
Creator: Smith, P.D.; Sullivan, R.M.; Lewis, A.C. & Yu, H.J.
Partner: UNT Libraries Government Documents Department

Advanced converter reactors

Description: Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.
Date: January 1, 1979
Creator: Kasten, P.R.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program

Description: Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.
Date: January 1, 1977
Partner: UNT Libraries Government Documents Department

Advanced gas-cooled nuclear reactor materials evaluation and development program: corrosion behavior of experimental alloys in controlled-purity helium at temperatures in the 750 to 1050/sup 0/C range. [Ni-20Cr + Al, Ti, Si, Nb and/or Y]

Description: A series of 10 experimental alloys (basically Ni-20Cr with addition of one or more of the elements Al, Ti, Si, Nb and Y) has been examined after exposure to controlled purity helium for periods of 1000 to 6000 hours at temperatures of 750 to 1050/sup 0/C. Alloys containing aluminum were particularly susceptible to internal oxidation at the lower temperatures, but at 950/sup 0/C and above carburization became the dominant corrosive mechanism. The most corrosive resistant alloys were Ni-Cr20 and Ni-Cr containing Si, Ti and Nb. The presence of small amounts of yttrium dramatically promoted the occurrence of carburization, even at temperatures as low as 850/sup 0/C. 69 figures, 7 tables.
Date: May 1, 1982
Creator: McKee, D.W.
Partner: UNT Libraries Government Documents Department

Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

Description: The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.
Date: January 25, 1980
Partner: UNT Libraries Government Documents Department

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

Description: Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.
Date: November 14, 1980
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

Description: The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.
Date: August 31, 1978
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report for period, 1 October 1977--31 December 1977

Description: The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered includes the activities associated with the procurement of the materials for the screening test program and information from vendor certification for the materials received for the nuclear process heat candidate alloys. The design modifications to the helium purification system and the construction status of the simulated reactor helium supply system, testing equipment, and analysis instrumentation and equipment are discussed. Finally, the status and details of the data management are presented.
Date: March 20, 1978
Partner: UNT Libraries Government Documents Department

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1978--March 31, 1978

Description: The activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials received, and preliminary information from the materials characterization tests performed by GE are reported. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The final recommended impurity levels for the screening phase helium are presented and the rational behind this gas chemistry is discussed. The status of the data management system is presented.
Date: June 26, 1978
Partner: UNT Libraries Government Documents Department

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

Description: Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.
Date: June 25, 1980
Partner: UNT Libraries Government Documents Department