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Oscillating Vertical Magnetic Dipole Above a Conducting Half-Space

Description: The electromagnetic field produced by a vertical oscillating magnetic dipole above a plane conducting earth is obtained in integral form. An exact solution in closed form is obtained for the case in which the dipole and the point of observation are both located on the surface of the earth. (auth)
Date: April 1, 1961
Creator: Wesley, J. P.
Partner: UNT Libraries Government Documents Department
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Response of Dual-Purpose Reinforced-Concrete Mass Shelter

Description: BS>A reinforced-concrete dual-purpose underground parking garage and personnel sheiter designed for a long-duration incident pressure of 40 psi was tested. The sheiter was exposed to shot Priscilla, an approximately 37-kt 700-ft balloon burst (June 24, 1957), at a ground range of 1600 ft (predicted 35-psi peak incident-pressure level). The recorded peak incident pressure at the shelter was approximately 39 psi. Postshot soil borings were made to obtain undisturbed samples for determining soil characteristics. Preshot and postshot field surveys were made to determine the total lateral and vertical displacement of the structure. The test structure provided adequate protection from the effects of the test device at the test GZ distance. Despite failure of the door sealing gasket, a rise in pressure in the interior did not exceed 1.0 psi. The flat-slab roof and supporting structure were more than adequate to resist the 39psi peak incident test loading. (P.C.H.)
Date: April 1, 1961
Creator: Cohen, E.; Laing, E. & Bottenhofer, A.
Partner: UNT Libraries Government Documents Department
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STEAM-COOLED POWER REACTOR EVALUATION, STEAM-COOLED FAST BREEDER REACTOR

Description: Conceptual design and economic studies of a steamcooled fast breeder reactor that can also be used as a source of power are presented. Two reactor plant sizes were considered: a 300-Mw(e) central power station plant and a 40 Mw(e) plant. It was concluded that attractive economics and good breeding characteristics breeding ratios from 1.27 to 1.42) can be achieved in steam- cooled PuO/sub 2/UO/sub 2/ fueled fast reactors. Low capital costs can be obtained by a compact reactor core and the absence of large heat exchangers and complicated process systems. Reactor design data are discussed. Analysis showed that these reactors can be prevented from going prompt critical, when fully flooded, by incorporating a tolerable amount of high resonance absorption materials such as hafnium or indium. An increase in reactivity on loss of coolant was indicated by preliminary calculations. (M.C.G.)
Date: April 15, 1961
Creator: Sofer, G.; Hankel, R.; Goldstein, L. & Birman, G.
Partner: UNT Libraries Government Documents Department
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Flexible Monte Carlo Programs FMC-N and FMC-G

Description: Flexible Monte Carlo programs FMC-N and FMC-G (GE-ANPD Programs 516 and 515) are digital computer programs which apply Monte Carlo methods to simulate neutron and gamma ray life histories, respectively, in a source-shield configuration. The programs were designed for flexlbility in the geometrical, material, nuclear, and source descriptions of source-shield configurations and variance reduction techniques. The programs were also designed to optimize the use of fast memory and to provide complete freedom in the dimensions of the various input quantities. The programs are coded for an IBM-704 computer with a fast memory capacity of 32,768 storage locations and eight magnetic tape units, and for an IBM7090 computer wlth a fast memory capacity of 32,768 storage locations and ten magnetic tape unlts on two date channels. No magnetic drum storage is necessary for either computer. (auth)
Date: April 28, 1961
Creator: Loechler, J. J. & MacDonald, J. E.
Partner: UNT Libraries Government Documents Department
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FUEL ELEMENT DEVELOPMENT PROGRAM FOR THE PEBBLE BED REACTOR. Final Report

Description: >The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found-that molecularly deposited'' ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces and where a larger thickness-todiameter ratio could be used than if the coating were on the surface of the graphite sphere. Fuel elements were irradiated to burnups ranging up to about 6 at.% U/sup 235/. In all specimens containing a uniform dispersion of fuel, the graphite spheres were found to retain their structural properties after irradiation. Data are given on fuel particle coatings of A1/sub 2/O/sub 3/, pyrolytic carbon, and metals: surface coatings of siliconized silicon carbide, pyrolytic carbon, and metal carbides; properties of and the effects of irradiation on graphite spheres; the use of natural graphite in preparing a high-density matrix material; graphite fueling by thorium nitrate infiltration; subsurface metal and metal carbide coatings for graphite; and an in-pile loop program on the behavior of fission products in a recycle helium stream. (auth)
Date: April 30, 1961
Partner: UNT Libraries Government Documents Department
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Nuclear Analysis of Various Spert-III Critical Experiments

Description: Editor please delete 26456.<><DSN>16:026457<ABS>Work done in the P122 reactor control actuator area is summarized. Actuators were required to radially position the absorber blades in the core of the reactor. The P122C1 was a subsonic power plant and temperatures were low enough to permit the use of hydraulics in the actuator area. The program was reoriented and the power plant designated P122C3 which was a supersonic version of the folded flow power plant. The ambient temperature at maximum power was high enough to require pneumatic actuation of the control blades. The program was reoriented after two design iterations of the subsonic power plant. A test model of the actuating equipment and the entire linkage assembly was on hand and completed when the program was cancelled. The linkage was being redesigned for the supersonic application and special bearings were ordered for fabrication into the lower temperature rig. The actual mechanical concepts of the pneumatic actuator were under study when the program was cancelled. (auth)
Date: April 27, 1961
Creator: Paluszkiewicz, S.
Partner: UNT Libraries Government Documents Department
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The H-1 High Temperature Graphite Irradiation Experiment

Description: A high temperature graphite irradiation experinient was performed in the GETR core to determine the effects of differences in manufacturing, formulation, and graphitization temperatures on radiation-induced eontraction. The experiment was performed at temperatures of 800 to 1200 deg C in an intense fast neutron flux. The maximum integrated exposure of the sample positions was 3.2 x 10?sup 21/ nvt, E> 0.18 Mev, corresponding to approximately 24,000 MWD/AT in a conventional graphite-moderated reactor. All the graphites tested, with the exception of the controls, were needle coke filler, coal tar pitch binder graphites varying mn particle size, graphitization temperature, and impregnation. From theoretical and experitnental considerations, the formulations and treatments were expected to result in a relatively stable graphite in the direction transverse to extrusion. For comparison of the experimental results to existing experience, a conventional graphite, CSF, was used at each irradiation position. The results showed that the graphite most stable to contraction was graphaitized at a high temperature(>3100hC) and made from small particle size (all flour) filler. In all cases, the needle coke graphite contracted at a lower rate than the CSF graphite. Differences attributable to the size of extrusion and/or post graphitization cooling rate were discerned readily. Auxil iary to the purposes of the experiment, the apparent thermnal neutron cross section for Co/sup 58/ (plus Co /sup 58m) was determined. Co/sup 58/ and Co/sup 58m/ are the products of the Ni/sup 58/ (n,p) reaction, which is used widely for fast flux monitoring. Both have large thermal neutron capture cross sections which must be accounted for to prevent error in fast neutron dosimetry. In this experiment, a value was determined for the apparent burn-out cross section of 3750 barns. (auth)
Date: April 1, 1961
Creator: Davidson, J.M. & Helm, J.W.
Partner: UNT Libraries Government Documents Department
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Summary of the APDA Fuel Development Programs

Description: Report issued by the APDA over fuel development programs. Designs for fast breeder reactors, safety tests, and methods are presented and discussed. This report includes tables, illustrations, and photographs.
Date: April 1961
Creator: Blessing, W. G.; Busch, J. S.; Duffy, J. G.; Hennig, R. J.; Jens, W. H.; Knight, F. W. et al.
Partner: UNT Libraries Government Documents Department
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Preliminary Studies of Scavenging Systems Related to Radioactive Fallout : Summary Report, May 1, 1960 to March 31, 1961

Description: Abstract: "Experimental investigation of aerosol particle capture by evaporating and condensing water drops has shown that capture is a function of the rate of water drop growth and aerosol particle diameter. Capture was found to be proportional to the rate of water vapor condensation and inversely proportional to aerosol particle diameter. The influence of water vapor gradient and particle size on aerosol particle capture during evaporation is insignificant. The experimental results are explained on the basis of particle penetration through the boundary layer of a water drop. An analysis of previous research on radioactivity of dry particulate matter in an urban atmosphere is included."
Date: April 28, 1961
Creator: Rosinski, John
Partner: UNT Libraries Government Documents Department
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Improved Zirconium Alloys Quarterly Report: January - March 1961

Description: The following report is one of a series of quarterly reports following the progress and development of improved zirconium alloys for service in superheated water and steam. This report covers the period between January 1 to March 31, 1961.
Date: April 11, 1961
Creator: Weinstein, Daniel & Holtz, F. C.
Partner: UNT Libraries Government Documents Department
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Response of dual-purpose reinforced-concrete mass shelter. Project 30. 2 of Operation Plumbbob

Description: A reinforced-concrete dual-purpose underground parking garage and personnel shelter designed for a long-duration incident pressure of 40 psi was tested. The shelter was exposed to shot Priscilla, an approx. 37-kt 700-ft balloon burst (June 24, 1957), at a ground range of 1600 ft (predicted 35-psi peak incident-pressure level). The recorded peak incident pressure at the shelter was approximately 39 psi. Postshot soil borings were made to obtain undisturbed samples for determining soil characteristics. Preshot and postshot field surveys were made to determine the total lateral and vertical displacement of the structure. The test structure provided adequate protection from the effects of the test device at the test GZ distance. Despite failure of the door sealing gasket, a rise in pressure in the interior did not exceed 1.0 psi. The flat-slab roof and supporting structure were more than adequate to resist the 39-psi peak incident test loading.
Date: April 1, 1961
Creator: Cohen, E.; Laing, E. & Bottenhofer, A.
Partner: UNT Libraries Government Documents Department
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Irradiation Processing Department monthly report, March 1961

Description: This document details activities of the irradiation processing department during the month of March, 1961. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; Financial Operation; and NPR project.
Date: April 14, 1961
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department Monthly Report: March 1961

Description: This report for March 1961, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.
Date: April 21, 1961
Creator: Hanford Atomic Products Operation. Chemical Processing Department.
Partner: UNT Libraries Government Documents Department
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Production test IP-310-A-FP, determination of the dimensional stability of uranium fuel cores classified by the fuel core tester (UT-2)

Description: Since it is now possible to obtain a heat-treated U core that is randomly oriented and has a finer average grain size, it is necessary to irradiate measured, transformed fuel cores over the full range of grain sizes, in order to compare relative dimensional stabilities. An improved ultrasonic tester, Fuel Core Tester-UT-2, is used to test all fuel cores.
Date: April 27, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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40-tube overbore facility location, C Reactor

Description: Possible locations of the projected 40-tube overbore facility at the C Reactor are discussed from the standpoint of obtaining conversion ratio data applicable to a full-reactor overbore program.
Date: April 19, 1961
Creator: Nilson, R.: Nechodom, W. S.
Partner: UNT Libraries Government Documents Department
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Loading and operating conditions for PT-IP-401-A and PT-IP-363-A in KER-1

Description: IP-401-A authorized the irradiation of 18-inch UO{sub 2} elements and IP-363-A authorized 18-inch KSE-3 elements. This document provides specific loading and operating conditions for a charge of one UO{sub 2} element and two KSE-3 elements in KER-1.
Date: April 5, 1961
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Process Water Flow Tests, C Reactor

Description: The power calculator at C Reactor and the process pump flowmeters at 190-C use orifice plates as primary sensing elements. These orifices were designed for considerably less flow and at the present higher flows, the reactor riser pressure is appreciably reduced as the result of the inefficiency of the orifices. Only a portion of the differential pressure across an orifice is recovered. The amount is dependent upon the ratio of the orifice diameter to pipe diameter. The permanent loss across these two orifices is 16 psi. The permanent pressure loss resulting from properly sized venturi elements would be 1.6 psi, for a resultant decrease in system pumping resistance of 14 to 15 psi. This would permit higher process water flow and reactor power levels.
Date: April 11, 1961
Creator: Hamilton, W. D.
Partner: UNT Libraries Government Documents Department
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In-tank solidification of intermediate-activity wastes

Description: Solidification of intermediate-activity wastes is a major goal of the CPD Waste Management Program. Plans are to reduce the wastes, by evaporation, to salt cakes in existing tanks, thereby insuring safe, long-term storage of contained fission products regardless.of tank integrity. Initiation of these plans at an early date is necessary to offset the expected increase in tank failures and to provide space for future wastes. Major decisions of the program relate to selection of the evaporative method to be employed. The requirements of in-tank solidification were therefore reviewed to determine if the choice of evaporative systems can be made at this time. The relative potential of Bentube evaporation and submerged combustion for meeting these requirements were analyzed on the basis of available information, including actual performance of the Bentube facility at the Savannah River Plant (SRP).
Date: April 3, 1961
Creator: Campbell, B. F.
Partner: UNT Libraries Government Documents Department
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Processing data for NIE and NIN KER Loop charges (PT 377)

Description: The following data represents processing conditions used in fabricating prototypic (except for supports) natural and enriched NPR inner fuel elements for Production Test 377. The prototypic NPR inner fuel elements were inserted into zircaloy-sleeves prior to charging. In this way it was possible to simulate, in the KER loops, the conditions under which NPR fuel elements would be subjected under irradiation. The purpose in documenting the data is to provide a permanent record of processing conditions and dimensions which may be referred to for post irradiation analysis and possible future process development work. Post irradiation results will be issued by the Fuels Development Operation, Hanford Laboratories Operation, and the test loop operating conditions will be issued by Process and Reactor Development Operation, Irradiation Processing Department, as outlined in the Production Test Procedure.
Date: April 19, 1961
Creator: Kusler, L. E. & Hays, D. D.
Partner: UNT Libraries Government Documents Department
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Proposal for charging GEH-10-43 NINI interim measurement test in the ETR

Description: The objective of this test is to irradiate an NPR fuel element inner tube in the ETR 3 {times} 3 Loop. the NPR fuel element is a tube in tube type. Both tubes are Zircaloy-2 clad uranium metal. The inner tube is identified by the Hanford designation NINI. In GEH-10-43 we plan to measure an NINI tube after each cycle for four cycles. An NPR inner tube with slightly larger diameter has been run in the ETR Hanford 3 {times} 3 Loop. Two NINI elements are now running in the 3 {times} 3 Loop.
Date: April 7, 1961
Creator: Geering, G. T. & Heck, E. N.
Partner: UNT Libraries Government Documents Department
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Supplement G, production test IP-314-A, measurement of fuel element temperature changes as the result of film deposition

Description: Irradiation of a third thermocouple train with a thermocouple element of the same design as used on the previous two trains, but with different train design and heater elements, is authorized to an exposure no greater than 1000 MWD/T. A third decontamination of KER-1 is also authorized.
Date: April 25, 1961
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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C-Reactor graphite burnout interim report, 1P-25A(PT-105-532-E)

Description: The oxidation of graphite in the Hanford reactors is of consequence since graphite burnout affects the strength of the moderator. As a means for indication of any highly oxidizing condition within the stack, containers or boats of small weighed samples of reactor-grade graphite are positioned along the length of an empty process channel in each reactor. The rate of oxidation of the monitoring samples, referred to as the burnout rate, is reported as percent weight lose per 1000 operating days (%/KOD). Currently the burnout rate limit is 2%/KOD. This document presents recent burnout data at the C-reactor.
Date: April 5, 1961
Creator: Ryan, B. A.
Partner: UNT Libraries Government Documents Department
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