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Data Relating to Hanford Mined Graphite (2273-D) Samples Annealed at NAA

Description: On 2/8/1950, there was mined from process tube 2273 in D pile at Hanford a quantity of graphite power, which was expected to show the most extensive radiation damage of any graphite available at that time. A series of samples of this powder were annealed in 100 degrees increments from 100 degrees to 2000 degrees C at this labaoratory. There were returned to Hanford and shipped by them to the National Bureau of Standards for total stored energy measurements. The present memorandum is comprised of a description of the annealing procedure used here, curves giving the detailed annealing history of each sample, and various curves derived from data obtained from these samples at Hanford and at the National Bureau of Standards.
Date: March 15, 1951
Creator: Smith, C.A. & Carter, R.L.
Partner: UNT Libraries Government Documents Department

US graphite reactor D&D experience

Description: This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).
Date: February 1, 1997
Creator: Garrett, S.M.K. & Williams, N.C.
Partner: UNT Libraries Government Documents Department

Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

Description: This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept.
Date: October 20, 1999
Creator: Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R. et al.
Partner: UNT Libraries Government Documents Department

Feasibility of Isotopic Measurements: Graphite Isotopic Ratio Method

Description: This report addresses the feasibility of the laboratory measurements of isotopic ratios for selected trace constituents in irradiated nuclear-grade graphite, based on the results of a proof-of-principal experiment completed at Pacific Northwest National Laboratory (PNNL) in 1994. The estimation of graphite fluence through measurement of isotopic ratio changes in the impurity elements in the nuclear-grade graphite is referred to as the Graphite Isotope Ratio Method (GIRM). Combined with reactor core and fuel information, GIRM measurements can be employed to estimate cumulative materials production in graphite moderated reactors. This report documents the laboratory procedures and results from the initial measurements of irradiated graphite samples. The irradiated graphite samples were obtained from the C Reactor (one of several production reactors at Hanford) and from the French G-2 Reactor located at Marcoule. Analysis of the irradiated graphite samples indicated that replicable measurements of isotope ratios could be obtained from the fluence sensitive elements of Ti, Ca, Sr, and Ba. While these impurity elements are present in the nuclear-grade graphite in very low concentrations, measurement precision was typically on the order of a few tenths of a percent to just over 1 percent. Replicability of the measurements was also very good with measured values differing by less than 0.5 percent. The overall results of this initial proof-of-principal experiment are sufficiently encouraging that a demonstration of GIRM on a reactor scale basis is planned for FY-95.
Date: April 30, 2001
Creator: Wood, Thomas W.; Gerlach, David C.; Reid, Bruce D. & Morgan, W. C.
Partner: UNT Libraries Government Documents Department

The Graphite Isotope Ratio Method (GIRM): A Plutonium Production Verification Tool

Description: Over the lifetime of a production reactor, neutrons from the fission process not only convert U-238 into plutonium but also bring about changes in the elements of the reactor's core components. Components such as shielding, pressure vessels, coolant piping, control rods, structural supports, and, in the case of graphite moderated reactors, the solid graphite moderator are all affected. Because a reactor's total plutonium production is directly related to total neutron fluence, and, likewise, changes in the elements and isotopes of a reactor's core components are directly related to fluence; it was argued that measuring these changes could provide an accurate estimate of a reactor's total plutonium production. The U.S. Department of Energy funds a project at Pacific Northwest National Laboratory (PNNL) to develop this concept into a practical plutonium production verification tool for graphite moderated reactors. The following sections describe the GIRM project development process. The purpose of this document is to provide a simple, concise description of the graphite isotope ratio method (GIRM) for use as a verification tool in estimating a graphite-moderated reactor's total plutonium production. The description covers the theory behind the technique and how the method is actually applied.
Date: January 1, 1999
Creator: McNeece, JP; Reid, BD & Wood, TW
Partner: UNT Libraries Government Documents Department

Status of ASME Section III Task Group on Graphite Support Core Structures

Description: This report outlines the roadmap that the ASME Project Team on Graphite Core Supports is pursuing to establish design codes for unirradiated and irradiated graphite core components during its first year of operation. It discusses the deficiencies in the proposed Section III, Division 2, Subsection CE graphite design code and the different approaches the Project Team has taken to address those deficiencies.
Date: August 1, 2005
Creator: Bratton, Robert L. & Burchell, Tim D.
Partner: UNT Libraries Government Documents Department

The physics design of accelerator-driven transmutation systems

Description: Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safer, less expensive and more environmentally sound approach to nuclear power.
Date: February 1, 1995
Creator: Venneri, F.
Partner: UNT Libraries Government Documents Department

Metal burning in graphite-moderated reactors

Description: Pinto beans, sweet corn, and zucchini squash (Cucurbita pepo var. black beauty) were grown in a randomized complete-block field/pot experiment at a site that contained the highest observed levels of surface gross gamma radioactivity within Los Alamos Canyon (LAC) at Los Alamos National Laboratory. Soils as well as washed edible and nonedible crop tissues were analyzed for various radionuclides and heavy metals. Most radionuclides, with the exception of {sup 3}H and {sup tot}U, in soil from LAC were detected in significantly higher concentrations (p <0.01) than in soil collected from regional background (RBG) locations. Similarly, most radionuclides in edible crop portions of beans, squash, and corn were detected in significantly higher (p <0.01 and 0.05) concentrations than RBG. Most soil-to-plant concentration ratios for radionuclides in edible and nonedible crop tissues from LAC were within the default values given by the Nuclear Regulatory Commission and Environmental Protection Agency. All heavy metals in soils, as well as edible and nonedible crop tissues grown in soils from LAC, were within RBG concentrations. Overall, the total maximum net positive committed effective dose equivalent (CEDE)--the CEDE plus two sigma for each radioisotope minus background and then all positive doses summed--to a hypothetical 50-year resident that ingested 160 kg of beans, corn, and squash in equal proportions, was 74 mrem y{sup -1}. This dose was below the International Commission on Radiological Protection permissible dose limit (PDL) of 100 mrem y{sup -1} from all pathways; however, the addition of other internal and external exposure route factors may increase the overall dose over the PDL. Also, the risk of an excess cancer fatality, based on 74 mrem y{sup -1}, was 3.7 x 10{sup -5} (37 in a million), which is above the Environmental Protection Agency`s (acceptable) guideline of one in a million. 25 refs.
Date: May 1, 1997
Creator: Wichner, R.P.; Ball, S.J.; Daw, C.S. & Thomas, J.F.
Partner: UNT Libraries Government Documents Department

The first reactor [40th anniversary commemorative edition]

Description: This updated and revised story of the first reactor, or 'pile,' commemorates the 40th anniversary of the first controlled, self-sustaining nuclear chain reaction created by mankind. Enrico Fermi and his team of scientists initiated the reaction on December 2, 1941, underneath the West Stands of Stagg Field at the University of Chicago. Firsthand accounts of the participants as well as postwar recollections by Enrico and Laura Fermi are included.
Date: December 1, 1982
Partner: UNT Libraries Government Documents Department

RBMK thermohydraulic safety assessments using RELAP5/MOD3 codes

Description: The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure tube ruptures. These assessments show that the RELAP5/MOD3 code can predict major phenomena during postulated accidents in the RBMK reactors.
Date: June 1, 1995
Creator: Tsiklauri, G.V. & Schmitt, B.E.
Partner: UNT Libraries Government Documents Department

Report of the special study group

Description: The special study group was activated by a charter letter from Sub-Section Managers of Pile Technology on June 20, 1956. The principal objectives were: to collect the information which is presently available for new reactor design and to determine what information should be developed; to make a guess at pile variables; and to point out development programs which must be pursued to achieve a detailed design start in two years. The study was restricted to graphite moderated reactors with H{sub 2}0, D{sub 2}0, and organic coolants. The program was to determine technical feasibility only and detailed economic considerations were not to be included. This report presents the conclusions of the group and some of the reasoning behind these conclusions.
Date: July 18, 1956
Creator: Brown, J.H.
Partner: UNT Libraries Government Documents Department

Medium-size high-temperature gas-cooled reactor

Description: This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).
Date: August 1, 1980
Creator: Peinado, C.O. & Koutz, S.L.
Partner: UNT Libraries Government Documents Department

Evaluation of EPRI nuclear power division research topics supportive of HTGR technology

Description: For HTGR commercialization studies, an LWR/HTGR Technology Transfer program was devised. Candidate programs were identified out of a total of 208 EPRI NPD (Nuclear Power Division) projects. Of these, 26 project areas presented the highest probability for technology transfer. (DLC)
Date: October 6, 1978
Partner: UNT Libraries Government Documents Department

Critical experiments in support of the CNPS (Compact Nuclear Power Source) program

Description: Zero-power static and kinetic measurements have been made on a mock-up of the Compact Nuclear Power Source (CNPS), a graphite moderated, graphite reflected, U(19.9% /sup 235/U) fueled reactor design. Critical configurations were tracked from a first clean configuration (184 most central fuel channels filled and all control rod and heat pipe channels empty) to a fully loaded configuration (all 492 fuel channels filled, core-length stainless steel pipe in the twelve heat-pipe channels, and approximately half-core-length boron carbide in the outer 4 control rod channels. Reactor physics data such as material worths and neutron lifetime are presented only for the clean and fully loaded configurations.
Date: January 1, 1988
Creator: Hansen, G.E.; Audas, J.H.; Martin, E.R.; Pederson, R.A.; Spriggs, G.D. & White, R.H.
Partner: UNT Libraries Government Documents Department

Extensions to SCDAP/RELAP5-3D for Analysis of Advanced Reactors

Description: The SCDAP/RELAP5-3D code was extended to enable the code to perform transient analyses of advanced LWRs (Light Water Reactors) and HTGRs (High Temperature Gas Reactors). The extensions for LWRs included: (1) representation of micro-heterogeneous fuel varying in composition in the radial and axial directions, (2) modeling of two-dimensional radial/axial heat conduction for more accurate calculation of fuel and cladding temperatures during the reflood period of a large break loss-of-coolant accident (LOCA), (3) modeling of fuel-cladding interface pressure and fuel-cladding gap conductance, (4) representation of radial power profiles varying in a discontinuous manner in the axial direction, and (5) addition of material properties for fuel composed of mixtures of ThO2-UO2 and ThO2-PuO2. The extensions for HTGR analyses included: (1) modeling of the transient two-dimensional temperature behavior of graphite moderated reactor cores (pebble bed and block-type), reactor vessel, and reactor containment, (2) modeling of flow losses and convective heat transfer in pebble bed reactor cores, (3) modeling of oxidation of graphite components in reactor cores due to the ingress of air and/or water, and (4) modeling of the affect of oxidation on the composition of gases in the reactor system. The applications of the extended code to LWR analyses showed that advanced fuels intended for proliferation resistance and waste reduction could also be designed to produce calculated peak cladding temperatures during a large break LOCA less than the 1477 K acceptance criterion in 10 CFR 50.46. Fuels composed of ThO2-UO2 and ThO2-PuO2 are examples of such fuels. The applications of the extended code to HTGR analyses showed that: (1) HTGRs can be designed for passive removal of all decay heat, and (2)
Date: April 1, 2003
Creator: Harvego, Edwin Allan & Siefken, Larry James
Partner: UNT Libraries Government Documents Department

Graphite oxidation modeling for application in MELCOR.

Description: The Arrhenius parameters for graphite oxidation in air are reviewed and compared. One-dimensional models of graphite oxidation coupled with mass transfer of oxidant are presented in dimensionless form for rectangular and spherical geometries. A single dimensionless group is shown to encapsulate the coupled phenomena, and is used to determine the effective reaction rate when mass transfer can impede the oxidation process. For integer reaction order kinetics, analytical expressions are presented for the effective reaction rate. For noninteger reaction orders, a numerical solution is developed and compared to data for oxidation of a graphite sphere in air. Very good agreement is obtained with the data without any adjustable parameters. An analytical model for surface burn-off is also presented, and results from the model are within an order of magnitude of the measurements of burn-off in air and in steam.
Date: January 1, 2009
Creator: Gelbard, Fred
Partner: UNT Libraries Government Documents Department

EVALUATION OF THE INITIAL CRITICAL CONFIGURATION OF THE HTR-10 PEBBLE-BED REACTOR

Description: This report describes the evaluation of data from the initial criticality measurement of the HTR-10 pebble-bed reactor at the Institute of Nuclear Energy Technology in China to determine whether the data are of sufficient quality to use as benchmarks for reactor physics computer codes intended for pebble-bed reactor analysis. The evaluation applied the INL pebble-bed reactor physics code PEBBED to perform an uncertainty analysis on the core critical height. The overall uncertainty in k-effective was slightly over 0.5%, which is considered adequate for an experimental benchmark.
Date: November 1, 2005
Creator: Terry, William K.
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

HTGR fuel recycle program. Quarterly progress report for the period ending August 31, 1977

Description: The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.
Date: September 1, 1977
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

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