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Appendix to Preliminary Design 30,000 KW Prototype Partially Enriched Uranium, Gas Cooled, Graphite Moderated Nuclear Power Plant (Prototype for an Optimum Plant) for United States Atomic Energy Commission Idaho Operations Office Contract No. AT(10-1)-925

Description: Report containing outlines of operating procedures, costs, construction, safety analysis, studies, and design specifications of a 30,000 kilowatt prototype partially enriched uranium nuclear power plant.
Date: March 1959
Creator: Kaiser Engineers
Partner: UNT Libraries Government Documents Department

Preliminary Design 30,000 KW Prototype Partially Enriched Uranium, Gas Cooled, Graphite Moderated Nuclear Power Plant (Prototype of an Optimum Plant) for United States Atomic Energy Commission, Idaho Operations Office contract no. AT(10-1)-925

Description: Report containing the preliminary design for a nuclear reactor plant and its facilities. Includes a description of systems, safety procedures, costs, and a glossary.
Date: March 1959
Creator: Kaiser Engineers
Partner: UNT Libraries Government Documents Department

Savannah River Project Site Report: 30,000 KW Prototype Partically Enriched Uranium Gas Cooled, Graphite Moderated Nuclear Power Plant for United States Atomic Energy Commission Idaho Operations Office

Description: Report describing a modified prototype of a nuclear reactor that uses partially uranium-enriched fuel and is cooled by helium. The construction site, site safety aspects, and design and construction costs are included.
Date: March 1959
Creator: ACF Industries Incorporated. Nuclear Products - ERCO Division.
Partner: UNT Libraries Government Documents Department

Appendix to Preliminary Design 55,000 KW Prototype Natural Uranium, Gas Cooled, Graphite Moderated Nuclear Power Plant (Prototype for an Optimum Plant) for United States Atomic Energy Commission, Idaho Operations Office

Description: Report containing documentation of projects undertaken by the multiple branches of the Health and Safety Division of the Idaho Operations Office at the Nuclear Reactor Testing Station (NRTS).
Date: March 1958
Creator: U.S. Atomic Energy Commission. Idaho Operations Office.
Partner: UNT Libraries Government Documents Department

Metals technology development plan

Description: This document presents the plan for the metals technology development required to support the design of the MHTGR within the US National Gas-Cooled Reactor Program. Besides descriptions of the required technology development, cost estimates, and schedules, the plan also includes the associated design functions and design requirements.
Date: March 1, 1987
Creator: Betts, W.S.
Partner: UNT Libraries Government Documents Department

ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs

Description: Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.
Date: March 10, 2004
Creator: Murphy, BD
Partner: UNT Libraries Government Documents Department

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

Description: A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.
Date: March 14, 2002
Creator: Grover, S.B. (INEEL) & Metcalfe, M.P. (BNFL, United Kingdom)
Partner: UNT Libraries Government Documents Department

HTGR fuels and core development program. Quarterly progress report for the period ending February 28, 1978. [Graphite and fuel irradiation; fission product release]

Description: The work documented here includes the design, analysis, and testing of the reactor core and its components comprising the fuel elements, hexagonal reflector elements, plenum elements, neutron sources, control rods, and reserve shutdown material. Also included are studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and, where appropriate, the data are presented in tables, graphs, and photographs.
Date: March 1, 1978
Partner: UNT Libraries Government Documents Department

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

Description: Results of work performed from October 1, 1978 through December 31, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program is presented. Objectives 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, and selection of materials for future test facilities and more extensive qualification programs. The activities associated with the characterization of the materials for the screening test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are included. The status of the data management system is presented.
Date: March 9, 1979
Partner: UNT Libraries Government Documents Department

High-Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration Lead Project strategy plan

Description: The strategy for developing the HTGR system and introducing it into the energy marketplace is based on using the most developed technology path to establish a HTGR-Steam Cycle/Cogeneration (SC/C) Lead Project. Given the status of the HTGR-SC/C technology, a Lead Plant could be completed and operational by the mid 1990s. While there is remaining design and technology development that must be accomplished to fulfill technical and licensing requirements for a Lead Project commitment, the major barriers to the realization a HTGR-SC/C Lead Project are institutional in nature, e.g. Project organization and management, vendor/supplier development, cost/risk sharing between the public and private sector, and Project financing. These problems are further exacerbated by the overall pervading issues of economic and regulatory instability that presently confront the utility and nuclear industries. This document addresses the major institutional issues associated with the HTGR-SC/C Lead Project and provides a starting point for discussions between prospective Lead Project participants toward the realization of such a Project.
Date: March 1, 1982
Partner: UNT Libraries Government Documents Department

Deposition of silicon carbide coatings on particles in a fluidized bed using silane and tetramethylsilane: a feasibility study

Description: Deposition of silicon carbide (SiC) coatings from compounds such as methyltrichlorosilane (CH/sub 3/SiCl/sub 3/) results in corrosive chlorine-containing byproducts. The feasibility of depositing coatings from SiH/sub 4/ and Si(CH/sub 3/)/sub 4/, which contain no chlorine, was briefly studied. Coatings of SiC were deposited from Si(CH/sub 3/)/sub 4/ over the temperature range 1025 to 1525/sup 0/C in beds of particles fluidized with either H/sub 2/ or Ar. The densest coatings were deposited at 1025/sup 0/C, but none approached the theoretical density of SiC. A SiC coating was also deposited at 800/sup 0/C from a mixture of SiH/sub 4/, C/sub 2/H/sub 4/, and H/sub 2/.
Date: March 1, 1978
Creator: Federer, J.I.
Partner: UNT Libraries Government Documents Department

Experiments with a lime slurry in a stirred tank for the fixation of carbon-14-contaminated CO/sub 2/ from simulated HTGR fuel reprocessing off-gas

Description: The fixation of CO/sub 2/ with a lime slurry in a stirred tank reactor appears to be feasible. The rate of reaction is fast, and virtually complete removal of CO/sub 2/ can be attained. At a gas residence time of <1 min, the decontamination factor (DF) is >100 in a single stage reactor for CO/sub 2/ concentrations ranging from 5 to 100%. It has been determined that two-stage contacting sequences which result in a cumulative DF > 10/sup 4/ are feasible. The reaction rate is constant up to 90% utilization of the lime and then rapidly decreases, as does the pH for the remainder of the reaction. The reaction appears to be liquid-phase-controlled, and the overall gas-side mass transfer coefficient (K/sub G/..cap alpha..) increases with impeller speed and gas flow rate, ranging from 0.4 x 10/sup -6/ to 6 x 10/sup -6/ g-moles of CO/sub 2/ per (cm/sup 3/-sec-atm). The reaction rate data are also correlated by a model of mass transfer accompanied by a fast pseudo first-order chemical reaction from which good agreement of calculated and predicted interfacial area is obtained. It was noted that temperature (21 to 46/sup 0/C) and lime concentration (0.5 to 1.5 M) had very little effect on mass transfer rate and DF. The settling rate of the CaCO/sub 3/ product increased significantly with impeller speed and temperature and decreased with gas flow rate. Scale-up calculations indicate that reasonably sized equipment would provide adequate removal of CO/sub 2/ for a full-scale reprocessing plant.
Date: March 1, 1978
Creator: Holladay, D. W.
Partner: UNT Libraries Government Documents Department

Tenth ORNL Personnel Dosimetry Intercomparison Study

Description: The Tenth Personnel Dosimetry Intercomparison Study was conducted at the Oak Ridge National Laboratory during April 9-11, 1984. Dosemeter badges from 31 participating organizations were mounted on 40cm Lucite phantoms and exposed to a range of dose equivalents which could be encountered during routine personnel monitoring in mixed radiation fields. The Health Physics Research Reactor served as the only source of radiation for eight of the ten irradiations which included a low (approx. 0.50 mSv) and high (approx. 10.00 mSv) neutron dose equivalent run for each of four shield conditions. Two irradiations were also conducted for which concrete- and Lucite-shield reactor irradiations were gamma-enhanced using a /sup 137/Cs source. Results indicated that some participants had difficulty obtaining measurable indication of neutron and gamma exposures at dose equivalents less than about 0.50 mSv and 0.20 mSv, respectively. Albedo dosemeters provided the best overall accuracy and precision for the neutron measurements. Direct interaction TLD systems showed significant variation in accuracy with incident spectrum, and threshold neutron dosemeters (film and recoil track) underestimated reference values by more than 50%. Gamma dose equivalents estimated in the mixed fields were higher than reference values with TL gamma dosemeters generally yielding more accurate results than film. Under the conditions of this study in which participants had information concerning exposure conditions and radiation field characteristics prior to dosemeter evaluation, only slightly more than half of all reported results met regulatory standards for neutron and gamma accuracy. 19 refs., 2 figs., 29 tabs.
Date: March 1, 1985
Creator: Swaja, R.E.; Chou, T.L.; Sims, C.S. & Greene, R.T.
Partner: UNT Libraries Government Documents Department

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

Description: The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.
Date: March 1, 1983
Creator: McCulloch, R.W. & MacPherson, R.E.
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

Summary of ORNL work on NRC-sponsored HTGR safety research, July 1974-September 1980

Description: A summary is presented of the major accomplishments of the Oak Ridge National Laboratory (ORNL) research program on High-Temperature Gas-Cooled Reactor (HTGR) safety. This report is intended to help the nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR. The ORNL program includes extensive work on dynamics-related safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification. Cooperative efforts with other programs are also described. Suggestions for near-term and long-term research are presented.
Date: March 1, 1982
Creator: Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Delene, J.G.; Harrington, R.M.; Hatta, M. et al.
Partner: UNT Libraries Government Documents Department

Reactor physics studies in the GCFR Phase III critical assembly

Description: The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO/sub 2/-UO/sub 2/ with radial and axial blankets of UO/sub 2/. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction rate distributions, absorption-to-fission ratios and the central point conversion ratio and the worth of steam entry into a small central zone. The reactivity change associated with the construction of a central pin zone in the core and axial blanket was measured. Reaction rate and steam entry measurements were repeated in the pin environment. Standard analysis methods using ENDF/B-IV data are described and the results are compared to measurements performed during the program.
Date: March 1, 1980
Creator: Morman, J A
Partner: UNT Libraries Government Documents Department

Reactor-specific spent fuel discharge projections, 1987-2020

Description: The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.
Date: March 1, 1988
Creator: Walling, R.C.; Heeb, C.M. & Purcell, W.L.
Partner: UNT Libraries Government Documents Department

RCRA Facilities Assessment (RFA)---Oak Ridge National Laboratory

Description: US Department of Energy (DOE) facilities are required to be in full compliance with all federal and state regulations. In response to this requirement, the Oak Ridge National Laboratory (ORNL) has established a Remedial Action Program (RAP) to provide comprehensive management of areas where past and current research, development, and waste management activities have resulted in residual contamination of facilities or the environment. This report presents the RCRA Facility Assessment (RFA) required to meet the requirements of RCRA Section 3004(u). Included in the RFA are (1) a listing of all sites identified at ORNL that could be considered sources of releases or potential releases; (2) background information on each of these sites, including location, type, size, period of operation, current operational status, and information on observed or potential releases (as required in Section II.A.1 of the RCRA permit); (3) analytical results obtained from preliminary surveys conducted to verify the presence or absence of releases from some of the sites; and (4) ORNL's assessment of the need for further remedial attention.
Date: March 1, 1987
Partner: UNT Libraries Government Documents Department