1,475 Matching Results

Search Results

Advanced search parameters have been applied.

Uranium Oxide Aerosol Transport in Porous Graphite

Description: The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.
Date: January 23, 2012
Creator: Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A. et al.
Partner: UNT Libraries Government Documents Department

The H-3 Irradiation Experiment: Irradiation of Experimental Gas Cooled Reactor Graphite, Number 2

Description: Report documenting "the long-term irradiation stability of the graphite used as the moderator in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge" (p. 1) by irradiating capsules at the General Electric Test Reactor. This includes the design and construction of the experiment, experiment procedures, and results of irradiation. Appendices begin on page 83.
Date: September 1964
Creator: Helm, J. W.
Partner: UNT Libraries Government Documents Department

Statistical Comparison of the Baseline Mechanical Properties of NBG-18 and PCEA Graphite

Description: High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR), a graphite-moderated, helium-cooled design that is capable of producing process heat for power generation and for industrial process that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is endeavoring to minimize the conservative estimates of as-manufactured mechanical and physical properties by providing comprehensive data that captures the level of variation in measured values. In addition to providing a comprehensive comparison between these values in different nuclear grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons and variations between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between the two grades of graphite that were initially favored in the two main VHTR designs. NBG-18, a medium-grain pitch coke graphite from SGL formed via vibration molding, was the favored structural material in the pebble-bed configuration, while PCEA, a smaller grain, petroleum coke, extruded graphite from GrafTech was favored for the prismatic configuration. An analysis of the comparison between these two grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in variability in properties within each of the grades that will ultimately provide the basis for the prediction of in-service performance. The comparative performance of the different types of nuclear grade graphites will continue to evolve as thousands more specimens are fully characterized from the numerous grades of graphite being evaluated.
Date: August 1, 2013
Creator: Carroll, Mark C. & Rohrbaugh, David T.
Partner: UNT Libraries Government Documents Department

AGC-1 Irradiation Experiment Test Plan

Description: The Advanced Graphite Capsule (AGC) irradiation test program supports the acquisition of irradiated graphite performance data to assist in the selection of the technology to be used for the VHTR. Six irradiations are planned to investigate compressive creep in graphite subjected to a neutron field and obtain irradiated mechanical properties of vibrationally molded, extruded, and iso-molded graphites for comparison. The experiments will be conducted at three temperatures: 600, 900, and 1200°C. At each temperature, two different capsules will be irradiated to different fluence levels, the first from 0.5 to 4 dpa and the second from 4 to 7 dpa. AGC-1 is the first of the six capsules designed for ATR and will focus on the prismatic fluence range.
Date: May 1, 2006
Creator: Bratton, R. L.
Partner: UNT Libraries Government Documents Department


Description: Twenty-eight of the thirty-five papers presented at the conference are included with discussions. Abstracts of the remaining seven papers are given. The seven sessions were devoted to: radiation effects on graphite, nuclear properties of graphitc, graphite lattice reactivities, chemistry of graphite, chemical reactions between liquid Na and Zr, slug canning fer the ORNL Graphite Reactor, and critical assemblies. Separate abstracts have been prepared for each of the twentyeight papers. (T.R.H.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department

Testing of Small Graphite Samples for Nuclear Qualification

Description: Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.
Date: November 1, 2010
Creator: Chapman, Julie
Partner: UNT Libraries Government Documents Department

Effect of Massive Neutron Exposure on the Distortion of Reactor Graphite

Description: Distortion of reactor-grade graphites was studied at varying neutron exposures ranging up to 14 x 10/sup 21/ neutrons per cm/sup 2/ (nvt)/sup */ at temperatures of irradiation ranging from 425 to 800 deg C. This exposure level corresponds to approximately 100,000 megawatt days per adjacent ton of fuel (Mwd/ At) in a graphite-moderated reactor. A conventionalcoke graphite, CSF, and two needle-coke graphites, NC-7 and NC-8, were studied. At all temperatures of irradiation the contraction rate of the samples cut parallel to the extrusion axis increased with increasing neutron exposure. For parallel samples the needle- coke graphites and the CSF graphite contracted approximately the same amount. In the transverse direction the rate of cortraction at the higher irradiation temperntures appeared to be decreasing. Volume contractions derived from the linear contractions are discussed. (auth)
Date: May 28, 1963
Creator: Helm, J. W. & Davidson, J. M.
Partner: UNT Libraries Government Documents Department

Status of Graphite Oxidation Work

Description: Data were developed to compare the extent of structural damage associated with high temperature exposure to an air leak. Two materials, NBG-18 graphite and unpurified PCEA graphite have been tested as of this report. The scope was limited to isothermal oxidation at a single temperature, 750°C. Ambient post-oxidation compression strength testing was performed for three levels of burn off (1%, 5%, and 10% mass loss) for two leak scenarios: 100% air and 10% air in helium. Temperature, gas flow, and dynamic mass loss oxidation conditions were monitored and recorded for each sample. The oxidation period was controlled with flow of inert gas during the thermal ramp and upon cool down with a constant 10 liter per minute flow maintained throughout furnace operation. Compressive strengths of parallel un-oxidized samples were tested to assess the relative mass loss effects. In addition to baseline samples matching the un-oxidized dimensions of the oxidized samples, two sets of mechanically reduced samples were prepared. One set was trimmed to achieve the desired mass loss by removing an effectively uniform depth from the geometric surface of the sample. The other set was cored to produce a full penetration axial hole down the center of each sample.
Date: May 1, 2010
Creator: Smith, Rebecca
Partner: UNT Libraries Government Documents Department

Conceptual Design of an Advanced Engineering Test Reactor

Description: From abstract: This report describes a conceptual design for an Advanced Engineering Test Reactor. The reactor is a large graphite assembly penetrated by parallel Zircaloy tubes through which flow as heavy water solution of uranyl sulfate. Reactor power is 220 megawatts.
Date: March 1, 1957
Creator: Mallon, R. G.; Saldick, J. & Gibbons, R. E.
Partner: UNT Libraries Government Documents Department


Description: Graphite-moderated, graphite reflected critical assemblies have been set up in the LASL Honeycomb remotely controlled machine. lnformation has been obtained on the critical masses of systems having C/Oy ratios of 6650 and 4093. A third system at a smaller ratio is planned. The reactivity contribution of channels through the core and reflector was determined. (auth)
Date: January 1, 1957
Creator: Byers, C.C.
Partner: UNT Libraries Government Documents Department


Description: The effect of irradiation on the thermal conductivity and electrical resistivity of U and U0/sub 2/ is being investigated. The creep properties of 15% cold-worked Zircaloy-2 are being investigated in the 290 to 400 deg C temperature range for times exceeding 10,000 hr. The density distribution of crushed graphite is being investigated by the sink-float method. Centrifugal- casting techniques for the production of Al-35 wt.% U casting in the form of hollow cylinders are being investigated. A study of the processes involved in the solidification of U castings in graphite molds is being made. Work continued on electrolytic oxide and electroless oxide coatings on Croloy-2 1/4. Experimental work was continued to determine the effect of additive oxides on the oxidation characteristics and phase stability of U0/sub 2/. The fueled-moderator study has continued with the determination of additional hydrogen-absorption isotherms for the Zr-25 wt.% alloy and high-temperature x-ray diffraction patterns of hydrides of the 1 and 50 wt.% alloys. The irradiation of type 347 stainless steel at ETR process-water temperature. about 120 deg F and at 600 deg F, and subsequent determination of irradiation damage are being done in support of the KAPL-33 loop to be installed at the ETR. AIloys of U and Nb are being considered as possible high-temperature reactor fuels. Thorium-uranium base alloys are the subject of an investigation aimed at improving irradiation stability and corrosion resistance by ternary alloying and control of processing techniques Cermet fuel materials consisting of from 60 to 90 vol.% U0/sub 2/, UN, or UC dispersed in stainless steel are being investigated. Several types of 1 1/ 2-inch-diameter fueled graphite spheres containing 10 wt.% of fully enriched U0/ sub 2/ are being evaluated before and after irradiation in the BRR. Some localized attack has been observed after prolonged exposure of Ti steam tubes ...
Date: December 1, 1958
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

Graphitized needle cokes and natural graphites for lithium intercalation

Description: This paper examined effects of heat treatment and milling (before or after heat treatment) on the (electrochemical) intercalating ability of needle petroleum coke; natural graphite particles are included for comparison. 1 tab, 4 figs, 7 refs.
Date: May 10, 1996
Creator: Tran, T.D.; Spellman, L.M.; Pekala, R.W.; Goldberger, W.M. & Kinoshita, K.
Partner: UNT Libraries Government Documents Department

Support Facility for a Graphite Target Neutrino Factory

Description: The Target Support Facility for a Neutrino Producing Research Facility extends from the pretarget, primary beam focusing region to the end of the decay channel. Technical components include the target, beam absorber, and solenoid magnetic-field focusing system. While the ultimate goal is to target about 4 MW of proton beam in the target area., smaller values and different target materials (e.g., low Z) are considered to facilitate the first step. As detailed in this report, a carbon target was chosen with an incident primary beam power of 1.5 MW, The target is embedded in a high-field solenoid magnet of 20 T, followed by a transition section channel, where the field tapers down to 1.25 T. An iterative design process has been carried out which optimizes Monte Carlo code flux projections with realistic magnetic-field parameters. The severe radiation environment and component shielding requirements strongly influence design choices. The overall system design includes the capture and decay channel solenoids, the design parameters of which were provided by the National High Magnetic Field Laboratory. This design balances resistive and superconducting magnet contributions. Facility requirements, including shielding, remote handling, radioactive water system, etc. are based on the final design goal of 4 MW. The extent of the Target Support Facility and radiation-handling equipment includes the 50-m decay channel, where remote-handling operations are also required.
Date: August 1, 2000
Creator: Spampinato, P.T.; Chesser, J.B.; Gabriel, T.A.; Gallmeier, F.X.; Haines, J.R.; Lillie, R.A. et al.
Partner: UNT Libraries Government Documents Department