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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

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

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

Gas-cooled reactors

Description: Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.
Date: January 1, 1976
Creator: Schulten, R. & Trauger, D. B.
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

Installation of reactor gas refrigeration system -- 105-C, Project C-431

Description: It is recommended that the Design Committee approve the installation of a refrigeration system in the 105-C gas circulation system for removal of moisture from the reactor following a process tube leak as was previously approved by the C-431 Project Committed. Engineering studies show that this refrigeration system is necessary to provide sufficient water removal capacity in order that the water absorbing capacity of the system furnished by the silica gel towers will not be a limit to the rate at which a reactor can be rehabilitated following a serious leak.
Date: March 5, 1953
Creator: Wells, H. T.
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

Technical activities report - July 1952 graphite development - pile graphite

Description: Physical data are presented for transverse CSF samples with capsule exposures of 568, 1049, and 1617 MD/CT. The higher exposures indicate a sharper damage gradient toward the front of the pile. Additional casings of various types of graphite were loaded into test holes during this month. Average values of the thermal conductivity and electrical resistivity for several types of virgin graphites are presented. Data of this nature will be a regular portion of this report henceforth. Process tube channel 2677-H was mined and traversed for bore diameter. Although several of the tube block junctions were obscured, the channel was quite uniform. Examination of all previously mined graphite powder samples for aluminum oxide corrosion product has been completed and the results are reported.
Date: August 11, 1952
Creator: Music, J. F. & Zuhr, H. F.
Partner: UNT Libraries Government Documents Department

CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

Description: In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records ...
Date: August 19, 2005
Creator: Jensen, M. A.
Partner: UNT Libraries Government Documents Department

Studies of alternative nuclear technologies

Description: This report is a summary of tasks performed for the U.S. Arms Control and Disarmament Agency under Contract AC7NC114. The work is directly related to the Agency effort to examine potential alternative fuel cycles that might enhance uranium resource utilization, minimize plutonium production, and reduce the weapons proliferation risk from spent fuel reprocessing or early introduction of fast breeder reactors. Reported herein are summaries of various inter-related task assignments, including: fuel utilization in current light water reactors operating with the uranium fuel cycle; alternate fuel cycles, including the use of denatured fuel in LWRs and of the spectral shift concept for reactivity control; fuel utilization in high temperature graphite moderated reactors using the denatured fuel cycle; fuel utilization in heavy water reactors (CANDU type), including the use of enriched fuel, denatured fuel, and recycle of plutonium and U-233; the tandem fuel cycle (recovery of spent fuel and further irradiation in a CANDU type reactor); issues in the utilization of denatured fuel in LWRs; preliminary conceptual evaluation of a heavy water moderated reactor suitable for use in the United States.
Date: April 1, 1978
Creator: Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W. III & Roach, K.E.
Partner: UNT Libraries Government Documents Department

Gas reactor international cooperative program interim report: United States/Federal Republic of Germany nuclear licensing comparison

Description: In order to compare US and FRG Nuclear Licensing, a summary description of United States Nuclear Licensing is provided as a basis. This is followed by detailed information on the participants in the Nuclear Licensing process in the Federal Republic of Germany (FRG). FRG licensing procedures are described and the rules and regulations imposed are summarized. The status of gas reactor licensing in both the U.S. and the FRG is outlined and overall conclusions are drawn as to the major licensing differences. An appendix describes the most important technical differences between US and FRG criteria.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, January 1, 1979-March 31, 1979

Description: This report presents the results of work performed from January 1, 1979 through March 31, 1979 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. Work covered in this report includes the activities associated with the creep-rupture testing of the test materials for the purpose of verifying the stresses selected for the screening creep test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment.
Date: July 19, 1979
Partner: UNT Libraries Government Documents Department

New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

Description: Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR.
Date: January 1, 1983
Creator: McDonald, C.F. & Sonn, D.L.
Partner: UNT Libraries Government Documents Department