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Gas purification facilities at Purex: Process study

Description: This report provides a summary of the results of a process study, requested by the Atomic Energy Commission an the recovery of krypton and xenon from irradiated uranium at the Hanford Purex Plant. This request was prompted by original Commission forecasts of the expanded requirements for Krypton-85 for commercial phosphorescent signal lights and markers and for xenon isotopes of low neutron cross-section for use in liquid xenon scintillation counters, in connection with D.M.A., government and university-sponsored work. It was requested that both Hanford and Savannah River submit order of magnitude cost estimates for recovery facilities at the respective sites for three separate design cases. The cost information developed, along with market survey information obtained-through the A. D. Little Company and Department of Defense market surveys, would serve as the basis for scheduling of the Hanford and Savannah River participation in the Commission`s overall fission rare gas recovery program.
Date: December 31, 1958
Creator: Michels, L. R. & Gerhart, J. M.
Partner: UNT Libraries Government Documents Department
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PRELIMINARY REPORT ON HEAT GENERATION AND STRESSES IN THE WALL OF A SPHERICAL HRE-4 PRESSURE VESSEL

Description: The heat geueration and thermal stresses in spherical HRE-4 vessels 3 to 4 1/2 ft in diameter with clad and solid stainless steel walls were invegtigated. Parameters included thorium slurry concentrations and moderator material (D/sub 2/ O and H/sub 2/O). The prirnary purpose of this study was to determine the influence of thermal stresses on the selection of the core size for the HRE4 reactor. Curves are presented which facilitate relatively rapid determination of stresses for the range of vessels considered. It is concluded that steady- state thermal stresses in the clad or solid stainless steel vessels considered will not have to be a determining factor in the selection of a core size, provided the power density does not exceed approximately 15 kw/l in the clad vessels and 8 kw/l in the solid stainless steel vessels. (auth)
Date: December 31, 1958
Creator: Cheverton, R.D.
Partner: UNT Libraries Government Documents Department
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Reactor calculations for 105-D

Description: This report provides raw data reactor calculations for the time period of September 12, 1958 through July 21, 1960.
Date: December 31, 1958
Creator: Vaughn, A. D.
Partner: UNT Libraries Government Documents Department
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SDR Project Quarterly Technical Progress Report No. 6 for the Period August 1, 1958 Through October 31, 1958

Description: A family of 200-Mw(e) SDR's was designed which could be operable ln 1965. They give total energy costs in the range 10 to 12 mills/kwh. The most promising of these reactors has the following characteristics: single-region design, moderate steam conditlons, natural uranium fuel, Zircaloy cladding, and stainless steel fuel tubes. A study of the applicability of SDR's to a broad range of power outputs (10 to 500 Mw(e)) was made. Although it appears reasonable to design a natural-uranium SDR with a power output down to 10 Mw(e), present studies indicate significant and wide economic interest at 40 Mw(e) and higher. Difficulties were encountered in interpreting the results of natural uranlum--D/ sub 2/0 lattice experiments in the Process Development Pile, and studies were initiated to determine the sources of difficulty and corrective measures. A code, called PALINDROME, which solves the Boltzmarm transport equation in the P/ sub 3/ approximation, was written. Work during the quarter on the Chugach 10- Mw(e) reactor was mainly concerned with completing the layout design of the more important reactor components and systems. A listing of the current design data is given, and a cross section of the reactor is shown. Two major changes were made during the quarter: (1) the substitution of steel ball-filled organiccooled neutron shield disks for the original concretefilled, water-cooled designs and (2) the reduction in size of the shutdown gas cooling systems. The preliminary safety analysis is briefly outlined. The development of maintenance techniques is discussed. Approximately 280 additional hours of operation of the SDR mockup facility were logged during the quarter. Modifications were made of the barrier test apparatus, and three successful experiments were performed on 6061 Al alloy. A liquid sodium leak detector was constructed, and also a rig for testlng fuel -- coolant tube closures. Sodium --liquid water and …
Date: December 31, 1958
Partner: UNT Libraries Government Documents Department
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STABILITY TESTS ON THE TYPES IN430A AND IN430B AVALANCHE DIODE REGULATORS

Description: The manufacturers' specifications for temperature coefficient of voltage and internaI impedance of the compensated avalanche diode types IN430A and IN430B appear quite promising. These devices couId be used as shunt regulators in high stability power supplies if the noise and drift rate were sufficiently small. One investigator reported the voltage did not drift more than the J-57 engin 0.002 per cent over a 7,000 hour period. Stability tests were performed on two diode samples under resonable laboratory conditions. The measured drift rate did not exceed 0.005 per cent per month, and short term noise was less than the J-57 engin 0.002 per cent. The actual diode drift rate may be even lower than the measured rate. (auth)
Date: December 31, 1958
Creator: Blankenship, J.L.
Partner: UNT Libraries Government Documents Department
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Test of Buried Structural-Plate Pipes Subjected to Blast Loading

Description: Abstract: Two 20-foot-long, 7-foot-diameter, 10-gauge structural-plate pipes having longitudinal joints with 8 bolts per foot were buried and tested in the Smoky event of operation Plumbob at nominal predicted pressure levels of approximately 170 and 195 psi.
Date: December 31, 1958
Creator: Williamson, R. A.
Partner: UNT Libraries Government Documents Department
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United States Atomic Energy Commission Assistance Program to the Eurochemic Company, Mol, Belgium

Description: The United States Atomic Energy Commission program of assistance to the European Company for the Chemical Processing of Irradiated Fuels ("Eurochemic"), Mol, Belgium, is presented. Included are: background, formation, purpose, and structure of the Company; basic design considerations and a brief description of the proposed plant; present status of the and a list of participating organizations and members. (auth)
Date: December 31, 1958
Creator: Shank, E. M.
Partner: UNT Libraries Government Documents Department
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100-N temporary construction line considerations

Description: Present thinking and planning appears to be developing from the following factors as concern the 13.8 KV temporary construction power limit. 1. It is understood that the present intent is to supply 100-N operating requirements from a single stub source in the 230 KV loop. 2. The original thoughts were to obtain construction power over a 13.8 KV line from 151-D substation. 3. Construction load requirements are now less than originally planned since steam has been substituted for electrical drive of primary loop pumps and 5500 hp motor tests are no longer necessary. 4. An extreme emergency backup source for the K plants has always been of concern, although minimized in recent planning. It is desirable to review the temporary construction line requirements from a future operating viewpoint to determine if the line could be useful to the operating plants after completion of construction. It is highly desirable to provide T.C. power source from K plants rather than 151-D and then leave the line and breakers in place for future maintenance assistance and as extreme emergency backup to K plants.
Date: December 30, 1958
Creator: Mollerus, F. J.
Partner: UNT Libraries Government Documents Department
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MATERIAL SPECIFICATIONS FOR APPR-1 CORE II CONTROL ROD FUEL ELEMENTS AND ABSORBER SECTIONS

Description: The control rod fuel elements consist of 16 flat composite fuel plates joined to a pair of side plates to form an integral assembly with a nominal water gap spacing of 0.133 inches between fueL plates. The core section of the fuel plate contains an active fuel section and a flux suppressor section. The fuel section is composed of UO/sub 2/, B/sub 4/C, and type 304LB stainless steel. The suppressor section is composed of Eu/sub 2/O/sub 3/ and elemental stainless steel powder. The absorber section of the assembly contains four absorber plates. The core section of the absorber plate is cermet, composed of Eu/sub 2/O/sub 3/ dispersed in stainless steel. (W.L.H.)
Date: December 30, 1958
Creator: Edgar, E.C. & Robertson, R.D.
Partner: UNT Libraries Government Documents Department
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POWER RESPONSE FOLLOWING REACTIVITY ADDITIONS IN HRT

Description: Calculations have been performed relating the magnitude of an HRT power excursion to an instantaneous addition of reactivity (up to 1% DELTA k/sub e/). Five delayed neutron groups as well as separate heat balances over the core and blanket were accounted for in the calculation. (auth)
Date: December 30, 1958
Creator: Jaye, S. & Lietzke, M.P.
Partner: UNT Libraries Government Documents Department
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A PRELIMINARY INVESTIGATION OF A CHROMATOGRAPHIC COLUMN SEPARATION OF RARE EARTHS USING Di(2-ETHYLHEXYL)PHOSPHORIC ACID

Description: A new chromatographic separation of small amounts of rare earths was devised and tested with carrier-free radiotracers of Nd/sup 147/, Pm/sup 147/, and Eu/sup 155/. The method uses di(2-ethylhexyl)orthophosphoric acid (HDEHP) as a stationary phase supported on a column of aluminum oxide. and elution is with dilute aqueous hydrochloric acid. Column behavior is similar to solvent extraction using HDEHP where separation factors average 2.5 for adjacent rare earths. (auth)
Date: December 30, 1958
Creator: Winchester, J.W.
Partner: UNT Libraries Government Documents Department
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Further Studies with the GCRE Critical-Assembly

Description: This report follows ciritical-assembly studies on: the effect on reactivity caused by changes in axial reflector materials; the effect on reactivity and the power perturbation caused by fast safety control-blade guides; the effect of changes in fuel-element material composition; the effect of changes in fuel-elements spacing designed to produce uniform radial power-generation rates.
Date: December 29, 1958
Creator: Dingee, David A.; Ballowe, William C.; Egen, R. A.; Jankowski, Francis J. & Chastain, Joel W.
Partner: UNT Libraries Government Documents Department
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Further Studies With the GCRE Critical Assembly

Description: Further engineering and physics data to aid in constructing GCRE-1 were obtained in critical-assembly studies. Four major experiments were performed to investigate the effect on reactivity caused by changes in axial reflector materials, the effect on reactivity and the power perturbation caused by fast safety control-blade guides, the effect of changes in fuel-element material composition, and the effect of changes in fuel-element spacing designed to produce uniform radial power-generation rates. All studies were performed with a 4-in.-thick lead reflector at the core perimeter. Axial-reflector-material studies employed combirations of aluminum and steel reflectors. The reactivity worth of a 2 3/4-in.-thick steel reflector was +0.414% DELTA k/k compared with 0.175% DELTA k/k for a similar aluminum reflector. The perturbation in the flux distribution caused by the safety-blade guides was localized, and affected only the regions immediately adjacent to the guides. The combined reactivity worth of two guides was -0.281% DELTA k/k. Fuel-element material compositions were changed by separate additions of fuel and stainless steel. An increase in uranium loading from an average value of 303 to 404 g per element would provide, based on extrapolations from experimental data, a reactivity of about 4.5% DELTA k/k. An increase in steel from 1708 to 2093 g per element decreased the core reactivity by abeut 1.1% DELTA k/k. A change in fuelelement spacing reduced the ratio of maximum to average power generation from 1.46 to 1.24. (auth)
Date: December 29, 1958
Creator: Dingee, David A.; Ballowe, William C.; Egen, Richard A.; Jankowski, Francis J. & Chastain, Joel W., Jr.
Partner: UNT Libraries Government Documents Department
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Inspection and Maintenance Experience With Hre-2

Description: Experience in replacement of highly radioactive components of Homogeneous Reactor Experiment No. 2 is discussed, with particular emphasis on containment of air-bonne contamination and control of personnel exposure. The design and operation of tools and viewing devices developed to observe the hole in the HRE-2 core tank are described. (auth)
Date: December 26, 1958
Creator: Culver, J. S.; Shepherd, D. M. & Collins, C. W.
Partner: UNT Libraries Government Documents Department
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Measurements of Mechanical Properties of Pure and Uranium-Loaded Graphites at Elevated Temperatures. Final Report for June 27, 1957 Through December 15, 1958

Description: Various mechanical-property measurements on several grades of manufactured (polycrystalline) graphite stock include tensile-strength determinatftons at temperatures from 20 to 2400 deg C (plus a few at 2500 deg C), tensile stress-strain and creep tests in the range 2000 to 2400 deg C, torsional strength and stress-strain measurements from 20 to 2600 deg C, torsional creep and stress-relaxation tests in the range from 2000 to 2500 deg C, dynamic Young's- modulus determinations from 20 to 2450 deg C, and evaluation of average coefficients of linear thermal expansion for the ranges 20 to 1050 deg C and 1300 to 2400 deg C. The materials studied include a Grade H4LM) molded in 36-in.-diam pieces, both as the standard stock with a nominal density of 1.74 g/cc and as stock re-impregnated to a nominal density of 1.80 to 1.85 g/cc, and four grades of graphite manufactured by the Los Alamos Scientific Laboratory containing normal uranium (added as UO/sub 2/ during the manufacture) in varying concentrations. The tensile-strength and creep data obtained are compared with the results of Los Alamos measurements upon similar materials. (auth)
Date: December 23, 1958
Creator: Green, L., Jr.; Stehsel, M.L. & Waller, C.E.
Partner: UNT Libraries Government Documents Department
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MILLIMICROSECOND DISCRIMINATOR

Description: A discriminator circuit for use with millimicrosecond counting equipment is described. The main characteristics of this unit are its good response to pulses as short as 3 millimicroscconds and the fast recovery time which is less than 0.15 microsec. (auth)
Date: December 23, 1958
Creator: Swift, D.F. & Perez-Mondez, V.
Partner: UNT Libraries Government Documents Department
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Bremsstrahlung Theories Since 1945 : [Bibliography]

Description: Report detailing various Bremsstrahlung theories since 1945. Citations for different theories are included.
Date: December 22, 1958
Creator: Maynard, Glenn R. & Lane, Zanier
Partner: UNT Libraries Government Documents Department
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Emergency cooling and air filtration systems for HAPO reactors

Description: This report represents a critical review of existing reactor cooling systems and discusses suggested supplementary-cooling system which might be employed in emergencies resulting from such natural hazards as earthquake, equipment failure, or personnel error. In addition the subject of building filtration is discussed. Maintenance of an uninterrupted flow of cooling water is of major concern to the safety of any HAPO reactor. For some time supplementary cooling systems which would be capable of removing heat output in the event of failure in the existing emergency backup systems have been under scrutiny. Loss of coolant may cause damaging power excursion (should this occur during operation) or will inevitably result in fuel melting and a subsequent release of fission products to the atmosphere, even if the reactor is shut down prior to the loss of coolant.
Date: December 22, 1958
Creator: Adams, O. E. Jr.
Partner: UNT Libraries Government Documents Department
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Full reactor-E-N-load-review of the hazards from a chemical viewpoint

Description: chemical Research and Development Operation was contacted by R. Nilson of Irradiation Processing Department in order to obtain an analysis from a chemical viewpoint of the worst credible accident to a full E-N reactor loading. This analysis would be used to present to the ACRS directly or would be abstracted to prepare a supplementary report. The primary purpose of this report is to present an analysis of the chemical behavior of an E-N loading in the event of the sudden loss of coolant to the entire reactor, with, however, normal operation of the control devices such that the reactor is initially driven sub-critical. We are here concerned wit the effect of a rising temperature upon the E-N load, the rise in temperature being primarily caused by the release of fission product energy.
Date: December 22, 1958
Creator: DeHollander, W. R.
Partner: UNT Libraries Government Documents Department
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K-Reactor water plant analysis

Description: The reliability of the K-Reactor water plants has been reviewed by Research and Engineering personnel; this study augments a brief analysis of the KW generator failures incident given in my letter of October 6, 1958. Our attention has been directed largely to ascertaining the requirements of the system, the consequences of possible component failures, giving general assessment of the primary features upon which reliability depends and identifying potential improvements which would increase reliability or moderate the consequences of a major failure. Our attention is restricted to failures which might occur as a result of equipment failure, operator error or certain natural causes such as earthquakes; we have not weighed special considerations brought on by enemy action such as sabotage, for example, in this evaluation. R. S. Bell is providing an analysis of the operating and maintenance aspects of K water plant reliability to complement this engineering study. It is concluded that the basic elements of the system lend themselves to reliable operation. The lines of defense appear adequate in depth, that is, a primary system operated with BPA power, an emergency ``backup`` system consisting of two essentially independent steam driven generators driving selected pumps, and a ``last ditch`` system providing coolant from an independent area through hydraulic cross-tie lines. It is further concluded that the reliability of the system currently is good and is very intimately associated with level of the maintenance and testing programs and competence with which they are carried out.
Date: December 22, 1958
Creator: Dickeman, R. L.
Partner: UNT Libraries Government Documents Department
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Model 1C shield design

Description: No Description Available.
Date: December 22, 1958
Creator: Henry, A. H.
Partner: UNT Libraries Government Documents Department
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DESIGN AND PERFORMANCE OF INDUCTION PUMP FOR SRE MODERATOR SYSTEM

Description: A three-phase linear induction pump was designed, constructed, and installed in the Sodium Reactor Experiment to control moderator temperature. A maximum flow rate of 91 gpm was obtained at 760 deg F, at 12.4 psi, and an efficiency of 1.7%. (C.J.G.)
Date: December 20, 1958
Creator: Baker, R.S.
Partner: UNT Libraries Government Documents Department
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Monthly Progress Report for the Period November 1 to 30, 1958

Description: Work directed toward developing a stainless steel clad UO/sub 2/ fuel element is reported. Methods are being developed for utilizing chemical poisoning for reactor in the reference environment. The design and development of mechanical features of fuel assemblies, control rods, baffles, the support structure, the reactor vessel closure, and fuel handling tools are presented. (For preceding period see YAEC-101.) (W.L.H.)
Date: December 20, 1958
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department
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