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WELDING AND BRAZING OF HIGH-TEMPERATURE RADIATORS AND HEAT EXCHANGERS
Procedures were developed for fabricating highperformance radiators and heat exchangers for the Aircraft Nuclear Propulsion (ANP) Program. These components, which contain multitudes of tube-to-tube sheet and tube-to-fin joints, are similar in design to those under consideration for a variety of space vehicle applications. In order to ensure reliability of the tube-to-tube sheet joints, techniques producing welds of extremely high quality were used and back brazing of the welds with a suitable alloy was incorporated. High-temperature brazing was also incorporated to attach high-conductivity fins to Inconel tubes in the radiators. The selection of a suitable brazing alloy for these applications was dependent upon several factors, including corrosion and oxidation resistance, flow point, and mechanical properties. A Ni- Si-B alloy was found to be adequate from all these considerations. Special brazing procedures were developed to obtain satisfactory flowability of this brazing alloy on tube-to-fin joints. The suitability of these fabrication procedures for the very stringent service conditions to which the radiators and heat exchangers were subjected was demonstrated by testing full-size components under operating conditions. (auth)
Date:
February 20, 1962
Creator:
Slaughter, G.M. & Patriarca, P.
Object Type:
Report
System:
The UNT Digital Library
Strontium-90 Fueled Thermoelectric Generator Power Source for Five-Watt U.S. Coast Guard Light Buoy. Final Report
The objectives of the SNAP 7A program were to design, manufacture, test, and deliver a five-watt electric generation system for a U. S. Coast Guard 8 x 26E light buoy. The 10-watt Sr/sup 90/ thermoelectric generator, the d-c-to-d-c converter, batteries and the method of installation in the light buoy are describcd. The SNAP 7A generator was fueled with four capsules containing a total of 40,800 curies of Sr/sup 90/ titanate. After fueling and testing, the SNAP 7A electric generating system was installed in the Coast Guard light buoy at Baltimore, Maryland, on December 15, 1961. Operation of the buoy lamp is continuous. (auth)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 30, 1961
Progress is reported on investigations in support of the Experimental Gas-Cooled Reactor, the Pebble-Bed Reactor Experiment, Advanced reactor design and development, test facilities, components, and materials. Topics covered include EGCR physics, EGCR performance analyses, structural investigations, EGCR component and materials development and testing, EGCR experimental facilities, PBRE physics and design studies, fueled-graphite investigations, clad fuel development, design studies of advanced power plants, experimental investigations of heat transfer and fluid flow, development of equipment anmd test facilities. and fabrication studies. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Examination of Uranium-2 w/o Zirconium Experimental Fuel Slugs Irradiated in EBR-I. Final Report-Program 6.1.11
Six groups of U-2 wt% Zr fuel slugs were irradiated in the first core of the EBR-I to burnups of 0.080 to 0.189 at.% at calculated temperatures of 307 to 353 deg C. Two groups of cast specimens were found to be more dimensionally stable than four groups of wrought slugs. Of the wrought slungs, the as quenched group showed less tendency to grow than the three groups which had some annealing after quenching. Specimens at burnups of about 0.189 at.% and at 383 deg C showed the onset of swelling as indicated by density measurements. The hardnesses of these specimens seemed but little affected by radiation, but there was an indication of softening with increasing irradiation temperature. (auth)
Date:
February 1, 1962
Creator:
Murphy, W. F.; Klank, A. C. & Paine, S. H.
Object Type:
Report
System:
The UNT Digital Library
Startup and Initial Testing of SM-1 Core II With Special Components
The loading operation for SM-1 Core II is described. Results of startup physics measurements (Test A-300 (Series) and fission product iodine monitoring in the primary coolant are given. The SM-1 Core II initial loading progressed satisfactorily, fulfilling the predictions of the zero power experiment performed at the Alco Criticality Facility. The initial cold clean five rod bank position was 6.53 in.; the initial hot, no xenon, five rod bank position was 9.62 in.; the initial hot, equilibrium xenon, five rod bank position was 11.41 in.; and the initial hot, peak xenon, five rod bank position was 12.14 in. Rods A and B were 19.00 in. in all four measurements. Stuck rod measurements indicated that an adequate shutdown margin was available with 20% of the rods fully withdrawn. All rod calibrations indicated a distinct shift and broadening of the peaks when compared with similar Core I calibrations. The temperature coefficient for Core II was 3.5 cents/ deg F at 440 deg F. Equilibrium xenon was worth approximately - 00 while peak xenon was worth - 43, both relative to the hot, no xenon core condition. During the period June 2, 1961 through September 30, 1961, the reactor operated at a total …
Date:
February 28, 1962
Creator:
Moote, F. G. & Schrader, E. W.
Object Type:
Report
System:
The UNT Digital Library
STAINLESS STEEL WASTES. III. LABORATORY STUDIES OF THE RATE OF REMOVAL OF STAINLESS STEEL IONS BY MERCURY CATHODE ELECTROLYSIS
ABS> The removal rates of iron, nickel, and chromium from synthetic stainless steel waste solutions during electrolysis over a mercury cathode were studied. The loading capacity of the mercury for the stainless steel metals was estimated on the basis of laboratory experiments to be about two% by weight. The laboratory data indicated that, at an electrode potential of --1.80 voits vs S.C.E., 85 ampere-hours per liter of waste removed essentially all of the stainless steel ions from a sulfuric acid solution containing 0.13M metal ions at 35 deg C. (auth)
Date:
February 12, 1962
Creator:
Anderson, D. R. & Rhodes, D. W.
Object Type:
Report
System:
The UNT Digital Library
KINETIC EXPERIMENTS ON WATER BOILERS-"A" CORE REPORT-PART II. ANALYSIS OF RESULTS
The status of the analytic portion of the KEWB program at the time of completion of the spherical core experiments is summarized. Three computer programs were developed for use in this analytic effort. The first reassembles and smooths three decades of reactor power data read separately from oscillogram records of reactor excursions. It then computes the logarithmic derivative of the power, energy release, fuel solution temperature, and temperature compensated reactivity. The second program utilizes the space-independent neutron kinetics equations with any number of delayed neutron groups to determine the reactivity in the reactor from the power and its derivative. The third program solves the space-independent kinetics equations for the neutron flux from an input reactivity or initial period. Up to 50 reactivity feedback equations includirg delayed neutrons are provided for in this program. A mathematical model of the reactor investigated extensively was one containing six delayed neutron groups, conventional treatment of temperature reactivity compensation, and void compensation of reactivity induced by radiolytic gas void growth proportional to the product of reactor power and energy release. Partial mathematical solutions to the kinetic equations were derived for reactivity feedback proportional to prompt temperature and void growth according to the product of power …
Date:
February 1, 1962
Creator:
Dunenfeld, M. comp.
Object Type:
Report
System:
The UNT Digital Library
EQUIPOISE-3: A TWO DIMENSIONAL, TWO-GROUP, NEUTRON DIFFUSION CODE FOR THE IBM-7090 COMPUTER
EQUIPOISE-3 is an IBM-7090 FORTRAN programmed code for the solution of two-group, two-dimensional, neutron diffusion equations. A maximum of 2l00 mesh points may be used, and the code will solve problems in either rectangular or cylindrical geometry. Logarithmic derivative boundary conditions are allowed, and removal of neutrons from both groups is permitted. Adjoint fluxes with the associated fluxadjoint flux regional integrals may be calculated automatically if desired. A constant buckling, group-dependent buckling, or region-dependent buckling may be specified for rectangular geometry. This program is intended to fill the need for a rapid two-dimensional calculation suitable for survey calculations. During the iterative part of the computations, all operations are carried out in the core memory. The magnetic tape memory is used only for input, output, and program storage. The running time for a 1000-point problem requiring 100 iterations would be about 3 min.(auth)
Date:
February 21, 1962
Creator:
Fowler, T.B. & Tobias, M.L.
Object Type:
Report
System:
The UNT Digital Library
THE DISTORTED-WAVE THEORY OF DIRECT NUCLEAR REACTIONS. I. "ZERO-RANGE" FORMALISM WITHOUT SPIN-ORBIT COUPLING, AND THE CODE SALLY
The distorted-wave theory of direct nuclear reactions is presented in a unified manner, in which the effects of assuming various reaction mechanisms and nuclear models appear only in certain radial form factors. The zero-range approximation is used, and spin-orbit coupling is neglected in the distorted waves. Formulas are given for transition amplitudes, cross sections, and polarizations. A description is given of the IBM-704 computer code SALLY that is based on these formulas. (auth)
Date:
February 1, 1962
Creator:
Bassel, R.H.; Drisko, R.M. & Satchler, G.R.
Object Type:
Report
System:
The UNT Digital Library
HYDROLOGIC AND GEOLOGIC STUDIES FOR PROJECT GNOME. Preliminary Report
Geologic lnformation required to define the pre- and post-shot physical and chemical characteristics of the sait and other rocks affected by the exploslon was gathered. Information on the pre- and post-shot hydrologic condltlons at the site and the surrounding area was also obtalned. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
A Theoretical Study of Destructive Nuclear Bursts in Fast Power Reactors
The calculation of destructive nuclear bursts in fast reactors by an improved Bethe-Tait method, which, for purposes of calculation, neglects propagation of the pressure wave is described. TMen exact numerical calculations for hydrodynamic and neutronic conditions during the power burst are performed in order to assess the importance of this neglect. (auth)
Date:
February 1, 1962
Creator:
Jankus, V. Z.
Object Type:
Report
System:
The UNT Digital Library
NIGHTMARE-AN IBM 7090 CODE FOR THE CALCULATION OF GAMMA HEATING IN CYLINDRICAL GEOMETRY
The NIGHTMARE program calculates the gamma-dose rate at any point in or near a reactor by means of the NDA buildup-factor method. The source distributton is obtained from a two-group, two-dimensional diffusion code (EQUIPOISE-2). Seven gamma-energy groups are used. As many as nine different concentric cylindrical regions may be considered, but only radial variation of attenuation and buildup properties is permdtted. Two schemes of estimating buildup factors through a succession of materials are built into the code, and results may be obtained by using either or both methods. An IBM 7090 computer wtth seven tape untts is required. Computations are performed at the approximate rate of l0,000/n mesh points per minute, where n is the number of axial, radial, and angular increments. The program is run under the control of the IBM 7090 MONITOR system. (auth)
Date:
February 26, 1962
Creator:
Tobias, M.L.; Vondy, D.R. & Lietzke, M.P.
Object Type:
Report
System:
The UNT Digital Library
Studies of Nuclear Resonant Absorption of Gamma Rays. Quarterly Report No. 4 Covering Period June 1, 1961 to August 31, 1961
The effect of polarizing mngnetic field intensity on the nuclear resonant absorption was studied by varying the field strength at a 1-mc Co/sup 57/ source from 0 to 1000 gauss while keeping the absorber between the poles of a magnet having a fixed field of 800 gauss. The rates of resonance absorption change with field intensity were greatest in the region of 300 to 1000 gauss, and the% nuclear resonant absorption for 1000-gauss fields was 8.5 and 26% for perpendicular and parallel fields, respectively, as compared with 15% for no fields. Other absorption measurements for Co/sup 57/ sources are also reported. Calculations on the use of nuclear resonant absorption to measure gravitational fields and altitudes were made which indicates that this application is not promising. (D.L.C.)
Date:
February 23, 1962
Creator:
Ezop, J. J.
Object Type:
Report
System:
The UNT Digital Library
SALT PHASE CHLORINATION OF REACTOR FUELS. III. CATALYZED DISSOLUTION OF URANIUM DIOXIDE IN LEAD CHLORIDE-CHLORIDE SYSTEMS
The rapid dissolution of uranium dioxide is described, wherein copper is added to molten lead chloride at 550 deg C. and chlorine is passed through the melt. The integral dissolution of zirconium-clad uranium dioxide fuels is also described. The dissolution rate of uranium dioxide is directly proportional to the concentration of cuprous chloride if an excess of chlorine is used; the value for the rate constant is approximately 100 mg (UO/sub 2/) cm/sup -2/ min/sup -1/ (CuCl molality). The uranium dioxide is converted to water soluble uranyl chloride. The dissolution rate can be controlled by three factors: the copper concentration, the flow rate of chlorine, and the surface area of uranium dioxide. Scoping work indicates that iron chloride may be as effective catalytically as copper chloride for uranium dioxide dissolution in the lead chloride -chlorine system at 550 deg C; thallium chloride is not as effective. (auth)
Date:
February 1, 1962
Creator:
Vander Wall, E. M.; Bauer, D. L. & Hahn, H. T.
Object Type:
Report
System:
The UNT Digital Library
ANALYSES OF THE RADIAL SEPARATOR
A mathematical analysis is presented concerning the performance of the radial or vane type steam separator. The analyses are concerned with the spread of flow on the vanes because of centrifugal force (which requires that the vane height be greater than the nozzle), and with separation of the vapor from the liquid. (J.R.D.)
Date:
February 1, 1962
Creator:
Robbins, C.H.
Object Type:
Report
System:
The UNT Digital Library
INSTRUMENTATION AND CONTROLS DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING JULY 1, 1961
The report comprises seven sections. A separate abstract was prepared for each section. (J.R.D.)
Date:
February 19, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Hallam Nuclear Power Facility Preoperational Test Completion Report Dry Criticality
A dry criticality test was carried out to determine the minimum critical mass of the HNPF Core without sodium. A subcritical calibration of the central control rod was performed and the relative reactivity worths of the inner ring of six control rods were determined. The extrapolated critical loading for the various plots after each incremental fuel loading with all rods out is shown. A tabulation is presented of multiplication data taken throughout the dry critical test. In order to find the relative integral reactivity worth of the central control rod in the dry critical loading, subcrltical multiplication data were obtained. The test completion criteria as stated in the test procedure were met. (M.C.G.)
Date:
February 24, 1962
Creator:
Kempt, H. C. & Corcoran, W. P.
Object Type:
Report
System:
The UNT Digital Library
Comparison of Gaseous Waste Handling Systems for PL-3
Gaseous waste handling systems are compared for direct and indirect boiling water and pressurized water reactors for PL-3 application. Areas that are common to the various concepts are not discussed since they do not enter into a comparison study. The major differences present are in the handling of active gases released to or held in the primary system coolant. These gases which could be present, their possibIe release from the system, and the necessary processing requirements are discussed in detail. Sufficient information is presented to enable one to analyze the methcds required to remove gaseous activity present in other size and type pressurized and boiIing water reactors. (auth)
Date:
February 28, 1962
Creator:
Noble, J. H. & Duke, E. E.
Object Type:
Report
System:
The UNT Digital Library
HRT-CHEMICAL PLANT RUN 21 SUMMARY
The multiple hydroclone system removed l83 grams of corrosion product solids in l814 hours of operation during reactor run 2l. The low removal rate was attributed to plugging of multiclone feed ports that presumably occurred during the latter part of run 20. After modifications to the reactor core and removal of the multiclone unit at the end of run 21, the reactor core was backflushed with the flow direction in the core loop reversed. During this period, the single hydroclone removed 205 grams of solids in 10.5 hours of operation. (auth)
Date:
February 12, 1962
Creator:
Yarbro, O.O.
Object Type:
Report
System:
The UNT Digital Library
DISSOLUTION OF BeO-AND Al$sub 2$O$sub 3$-BASE REACTOR FUEL ELEMENTS. PART I
Aqueous methods for recovering uranium from BeO- and Al/sub 2/O/sub 3/- base gas-cooled-reactor fuel elements are being evaluated. Two methods for processing Hastelloy-X--clad pelletized BeO-base fuels containing 60 to 70% UO/ sub 2/, such as the GCRE and MGCR, seem feasible. One method involves mechanical stripping or chopping of the cladding followed by leaching of the uranium from the fuel pellets with boiling 6-l3M HNO/sub 3/. In the other method the cladding and UO/sub 2/ are dissolved in boiling 2M HNO/sub 3/-4M HCl. In either case, most of the BeO matrix remains as an undissolved residue. Pellets containing 70% UO/sub 3/ dissolved completely in less than 20 hr in boiling 8M HNO/sub 3/ containing either 2M H/sub 2/SO/sub 4/ or 0.5M HF, producing solutions containing 4 g of uranium per liter. Fuels of high BeO content, e.g. BeO--5% UO/sub 2/, dissolved only slowly in boiling aqueous reagents. Highest initial rates were in sulfuric acid solutions, log (Rate, mg min/sup -1/cm/sup -2/) = 0.223 (H/sub 2/SO/ sub 4, M) - 2.8l and in HF--NH/sub 4/F solutions. ln boiling 5-8M NH/sub 4/F the initial dissolution rate increased from 0.07 to 3.5 mg min/sup -1/cm/sup -2/ as the HF concentration increased from 0 …
Date:
February 14, 1962
Creator:
Warren, K S; Ferris, L M & Kibbey, A H
Object Type:
Report
System:
The UNT Digital Library
SOME CALCULATIONS PERTAINING TO FAST REACTOR SAFETY
A program on fast reactor safety is described in which the theoretical aspects, techniques and accuracy of calculation, and associated safety predictions are considered. One topic deals with the success obtained in analyzing complicated meltdown configurations using two-dimensional, twogroup diffusion theory. Another topic deals with the dependence of energy yield in a fast reactor explosion on the shape of the power distribution in the core. The results of a brief survey are included dealing with the change of dynamic reactor behavior during various startup accidents as a consequence of a reduction in the delayed-neutron fraction. (J.R.D.)
Date:
February 1, 1962
Creator:
O'Shea, D.M.; Okrent, D. & Chaumont, J.M.
Object Type:
Report
System:
The UNT Digital Library
Gcr-Orr Loop No. 2 Filter Tests. Part Ii
Tests of Cambridge absolute filters, Model Sl-071, specified for use in the GCR-ORR Loop No. 2 as full-flow, primary coolant fiiters were completed. kD.O.P/ (dioctylphthalate) efficiency tests were performed on three filters in the as-received condition, on two filters following canning and thermal cycling, and on one of the canned fiIters following bsking out. None of the three units met the design criteria of 99.97% efficiency for removal of 0.3 micron particles in the as-received condition. The postthermal cycle efficiencies of the canned fiIters were slightly higher than their respective as-received efficiencies. At the corapletion of testing, the two fiiters canned for installation in the reactor facility had measured efficiencies of 99.855% and 99.93%. These values were judged acceptable for the intended application/su The thermal cycling of the two canned filters and the subsequent baking out of one of these units demonstrated that a limited amount of off-gas products would be given off/su Pressure drop tests were performed on the canned fiiters with instrument air (ambient temperature, atmospheric pressure) over a flow rate range of 150 to 530 lb/hr. Curves of pressure drop across each fiIter versus Reynolds number were plotted for air and He. (auth)
Date:
February 19, 1962
Creator:
Flint, F. A. & Smith, A. M.
Object Type:
Report
System:
The UNT Digital Library
DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Monthly Progress Letter for Month of January 1962
Development work leading to comparison of co-precipitated and mechanically blended UO/sub 2/-- PuO/sub 2? powders is reported in which special emphasis was placed on blending trials, pellet sintering studies, and subsequent evaluation of pellets made with blended material. Homogeneity studies indicate that currently used procedures are unsatisfactory because particle buildup occurs during blending. Powder preparation via the oxalate route was continued along with PuO/sub 2/ moisture pickup studies. Homogeneous precipitation studies on UO/ sub 2/ were continued to determine feasibility of direct preparation of dense UO/ sub 2/-- PuO/sub 2/ feed materials. Plasma-torch-produced PuO/sub 2/ spheres are being evaluated. (J.R.D.)
Date:
February 1, 1962
Creator:
Puechl, K.H.
Object Type:
Report
System:
The UNT Digital Library
THE CHEMISTRY OF NIOBIUM IN PROCESSING OF NUCLEAR FUELS
The chemical and industrial literature and laboratory work concerned with processing of niobium-containing nuclear fuels are reviewed. 57 references. (auth)
Date:
February 1, 1962
Creator:
Gens, T.A.
Object Type:
Report
System:
The UNT Digital Library