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Energy Economic Data Base (EEDB) Program. Phase III. Final report and third update

Description: The objective of the USDOE EEDB Program is to provide periodic updates of technical and cost (capital, fuel and operating and maintenance) information of significance to the US Department of Energy. This information is intended to be used by USDOE in evauating and monitoring US civilian nuclear power programs, and to provide them with a consistent means of evaluating the nuclear option and proposed alternatives. The data tables, which make up the bulk of the report, are updated to January 1, 1980. The data in these tables and in the backup data file supercede the information presented in the Second Update (1979). Where required, new descriptive information is added in the text to supplement the data tables.
Date: July 1, 1981
Partner: UNT Libraries Government Documents Department

Review of draft HTGR-SC/C design and technology development plans

Description: GCRA has approached the initial review of the subject plans from the vantage of Utility/User interests in an HTGR-SC/C Lead Project as characterized in the HTGR-SC/C Lead Project Plan. At their current stage of development, the plans are considered to represent the early views of General Atomic Company, acting as a potential NSSS vendor, on the effort necessary to design and license an HTGR-SC/C plant. As such, these plans embody GA's perception of the level of technical development and demonstration needed to support the cost and risk sharing assumptions in the Lead Project Plan. These plans and the plant design which they address will be subjected to a more formal review by other vendor and Utility interests in the course of establishing the Project Decision Package scheduled for June 1982. The culmination of these review activities will be the adoption of this information by all participants as the HTGR-SC/C Program Baseline.
Date: September 22, 1981
Partner: UNT Libraries Government Documents Department

High-temperature process heat. Interim design and cost status report, FY 1981

Description: Studies conductd on HTGR systems in FY 1980 were concluded in Application Study Reports to describe the preconceptual system designs to that point and discuss possible applications for three variations of the systems; the steam cycle/cogeneration plant, the higher temperature reformer plant, and the gas turbine concept. The HTGR-Reformer Application Study was conceived and directed to evaluate the HTGR-R with a core outlet temperature of 850/sup 0/C as a near-term Lead Project and as a vehicle to long-term HTGR Program Objectives. The scope of this effort included evaluations of the HTGR-R technology, evaluation of potential HTGR-R markets, assesment of the economics of commercial HTGR-R plants, and the evaluation of the program scope and expenditures necessary to establish HTGR-R technology through the completion of the Lead Project.
Date: October 1, 1981
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety-reliability program plan

Description: The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.
Date: March 1, 1981
Partner: UNT Libraries Government Documents Department

Outline and schedule for the HTGR-SC/C licensing plan

Description: The Licensing Plan is based on licensing a HTGR-SC/C lead plant in the near term. The plan also provides reference safety material and a basis (requirements, criteria, etc.) for licensing commercial follow-on plants. The plan is structured in two parts: program management and project management, and covers three sequences of licensing activities: pre-application, construction permit application, and operating licensing application. Major activities and a schedule of events are outlined in these three phases indicating the approach, the objective and the documentation involved. The Licensing Plan will be further developed in detail in FY 1982 as part of a Project Decision Package.
Date: September 1, 1981
Partner: UNT Libraries Government Documents Department

Overview of the siting characteristic of the HTGR steam cycle/cogeneration

Description: This report presents an overview of the siting characteristics of the HTGR-SC/C design and compares them to current NRC limiting criteria for radiological impact considerations. A systems advantage comparison is also made to an equivalent size PWR. Data is presented in summary tabular format that is derived from existing plant designs and published safety and environmental analysis for both the HTGR and the PWR.
Date: November 1, 1981
Partner: UNT Libraries Government Documents Department

Status of the United States National HTGR program

Description: The HTGR continues to appear as an increasingly attractive option for application to US energy markets. To examine that potential, a program is being pursued to examine the various HTGR applications and to provide information to decision-makers in both the public and private sectors. To date, this effort has identified a substantial technical and economic potential for Steam Cycle/Cogeneration applications. Advanced HTGR systems are currently being evaluated to determine their appropriate role and timing. The encouraging results which have been obtained lead to heightened anticipation that a role for the HTGR will be found in the US energy market and that an initiative culminating in a lead project will be evolved in the forseeable future. The US Program can continue to benefit from international cooperative activities to develop the needed technologies. Expansion of these cooperative activities will be actively pursued.
Date: January 1, 1981
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

Transient stress wave propagation in HTGR fuel element impacts

Description: A study of transient stress wave propagation in graphite HTGR fuel elements has been undertaken as a step toward developing techniques for the evaluation of seismic impact loads. The objectives of the study were to identify appropriate numerical methods, to understand the influence of the geometry and the multiple holes on the response, and to determine the relative importance of high frequency response and lower mode vibrations. A general review is made of the dynamic contact problem, and the methods available to model impact phenomena and stress wave propagation are evaluated.
Date: February 1, 1981
Creator: Almajan, I.T. & Smith, P.D.
Partner: UNT Libraries Government Documents Department

Fission-product SiC reaction in HTGR fuel

Description: The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels.
Date: July 13, 1981
Creator: Montgomery, F.
Partner: UNT Libraries Government Documents Department

US/FRG umbrella agreement for cooperation in GCR development. Fuel, fission products, and graphite subprogram. Quarterly status report, July 1, 1981-September 30, 1981

Description: This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 3, April 1981. The work described was performed during the period July 1, 1981 through September 30, 1981 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at General Atomic Company, and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Julich, HRB Mannheim, HOBEG Hanau and SIGRI Meitingen.
Date: October 1, 1981
Creator: Turner, R.F.
Partner: UNT Libraries Government Documents Department

HTGR high temperature process heat design and cost status report. Volume II. Appendices

Description: Information is presented concerning the 850/sup 0/C IDC reactor vessel; primary cooling system; secondary helium system; steam generator; heat cycle evaluations for the 850/sup 0/C IDC plant; 950/sup 0/C DC reactor vessel; 950/sup 0/C DC steam generator; direct and indirect cycle reformers; methanation plant; thermochemical pipeline; methodology for screening candidate synfuel processes; ECCG process; project technical requirements; process gas explosion assessment; HTGR program economic guidelines; and vendor respones.
Date: December 1, 1981
Partner: UNT Libraries Government Documents Department

Scale-model study of the seismic response of a nuclear reactor core

Description: The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selection, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of (1) Coulomb friction between the blocks, and (2) the clearance gaps between the blocks on the system response to seismic excitation are emphasized. 6 refs., 10 figs., 6 tabs.
Date: May 1, 1981
Creator: Dove, R.C.; Dunwoody, W.E. & Rhorer, R.L.
Partner: UNT Libraries Government Documents Department

Effect of prestress and stress on the strength and oxidation rate of nuclear graphite

Description: Effects of prestress or stress during oxidation on the reactivities and strengths of PGX and H451 graphites were examined at oxidation temperatures of 450 to 600/sup 0/C in air and 550 to 750/sup 0/C in 2% H/sub 2/O/He. Little or no effect of compressive prestress at stress levels of up to 75% of the mean fracture strengths, was found. A small effect on the reactivity of H451 graphite was observed in the case of 2% H/sub 2/O/He, i.e., the rate was enhanced by no more than 30% for compressive prestress levels of up to 90% of the fracture strength. Tensile and compressive stresses during oxidation did not affect the reactivities of these graphites at stress levels of up to 28% of the mean compressive strength and 35% of the mean tensile strength for H451 graphite, and 44% and 56% for PGX graphite, respectively. 13 refs., 18 figs.
Date: September 1, 1981
Creator: Eto, M. & Growcock, F.B.
Partner: UNT Libraries Government Documents Department

Three-dimensional thermoelastic analysis of a Fort St. Vrain core support block

Description: A thermoelastic stress analysis of a graphite core support block in the Fort St. Vrain High-Temperature Gas-Cooled Reactor is described. The support block is subjected to thermal stresses caused by a loss of forced circulation accident of the reactor system. Two- and three-dimensional finite element models of the core support block are analyzed using the ADINAT and ADINA codes, and results are given that verify the integrity of this structural component under the given accident condition. 10 refs., 39 figs.
Date: September 1, 1981
Creator: Butler, T.A. & Anderson, C.A.
Partner: UNT Libraries Government Documents Department

Interpretation of bend strength increase of graphite by the couple-stress theory. [HTGR]

Description: This paper presents a continued evaluation of the applicability of the couple-stress constitutive theory to graphite. The evaluation is performed by examining four-point bend and uniaxial tensile data of various sized cylindrical and square specimens for three grades of graphites. These data are superficially inconsistent and, usually, at variance with the predictions of classical theories. Nevertheless, this evaluation finds that they can be consistently interpreted by the couple-stress theory. This is compatible with results of an initial evaluation that considered one size of cylindrical specimen for H-451 graphite.
Date: May 1, 1981
Creator: Tang, P.Y.
Partner: UNT Libraries Government Documents Department

Relationship between carburization and zero-applied-stress creep dilation in Alloy 800H and Hastelloy X. [HTGR]

Description: Typical HTGR candidate alloys can carburize when exposed to simulated service environments. The carbon concentration gradients so formed give rise to internal stresses which could cause dilation. Studies performed with Hastelloy X and Alloy 800H showed that dilations of up to almost 1% can occur at 1000/sup 0/C when carbon pickup is high. Dilation was normally observed only when the carbon increase was >1000 ..mu..g/cm/sup 2/ and ceased when the diffusing carbon reached the center of the specimen.
Date: January 1, 1981
Creator: Inouye, H. & Rittenhouse, P.L.
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

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

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

Sorption of iodine on low-chromium-alloy steel. [HTGR]

Description: The sorption behavior of iodine on the surfaces of 2 1/4% Cr-1% Mo steel was investigated as a part of the High Tmeperature Gas-Cooled Reactor (HTGR) Chemistry Program at Oak Ridge National Laboratory (ORNL). The primary objective of these tests was to determine the equilibrium sorptive capacity of this alloy, which comprises most of the cooler regions of HTGR coolant circuit, under representative conditions. The data will be used to improve the capability for predicting, with computer programs, iodine deposition as functions of temperature and location in the primary circuit.
Date: January 1, 1981
Creator: Osborne, M.F.; Briggs, R.B. & Wichner, R.P.
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

Correction for gamma-ray self-attenuation in regular heterogeneous materials

Description: A procedure for determining the total correction factor for gamma-ray self-attenuation in regular heterogeneous materials is derived and discussed. The result of a practical application of the procedure to the passive gamma-ray assay of the /sup 235/U content of high-temperature gas reactor fuel is presented.
Date: September 1, 1981
Creator: Parker, J.L.
Partner: UNT Libraries Government Documents Department