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Quarterly Status Report on LAMPRE Program for Period Ending May 20, 1962
During the report period 110 Mw-hr of operation were completed by LAMPRE I at various power levels, and observations were made of control element performance and reactivity losses. Metallographic examination of a Core I capsule having 120 Mw-hr of irradiation exposure disclosed no major attack. Core II utilizes Ta alloy melts and 13 fuel melts. Starting materials for Core II capsules are tabulated. In research and development a gamma-ray conversion television was used to observe the flow of molten Pu in the PTA experiment. Dilute fuels being studied for use in a high-performance reactor include Pu --Co - -Ce alloys, ing exposure of test containers to U --Pu -Mn fuel at 900 deg C for 2 hrs are tabulated. In fuel reprocessing data on decontamination of Pu metal by electrorefining and by solvent extraction were obtained. Details of maintenance and development of the core test facility are included. (J.R.D.)
Date:
June 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
REPORT ON GLOVE BOXES AND CONTAINMENT ENCLOSURES
Criteria and guide lines are presented for the design, construction, and operation of safe, economical, and efficient glove boxes and associated facilities based upon present conditions and anticipated changes. Comprehensive discussion of glove box materials and components, safety and fire prevention methods, health physics problems, operational considerations, and brief descriptions of AEC installations are included. (39 references) (C.H.)
Date:
June 20, 1962
Creator:
Garden, N.B. ed.
Object Type:
Report
System:
The UNT Digital Library
Improved Sample Bonding and Emission With Tantalum Surface Ionization Filaments
Techniques for conditioning of Ta filaments for improved bonding and emission with a Ta metal powder-Ta/sub 2/O/sub 5/ mixture are described. A porous Ta metal layer is deposited which restricts sample to the filament. Metal- oxide ion emission is enhanced with additional Ta/sub 2/O/sub 5/ to the porous layer. Reduction of fractionation through action of liquid Ta/sub 2/O/sub 5/ is discussed in particular for Sr+ emission. Use of conditioned filaments for rapid U concentration analysis with a single-filament mass spectrometer is emphasized. (auth)
Date:
June 29, 1962
Creator:
Goris, P.
Object Type:
Report
System:
The UNT Digital Library
FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962
The permanent shutdown of the Westinghouse Testing Reactor at the end of the quarter forced a revision in plans for programming the remainder of the irradiation of the two uninstrumented capsules whose testing was in progress. The minimum estimated burnup at this point, based on hot cell data obtained from the pilot capsule, was 13,100 MW-d/ton U. It was decided to continue the testing of only one capsule in another reactor until the original goal of 20,000 MW-d/ton U is reached. The irradiation of the second capsule is to be terminated so that it can serve as a control. Fabrication was initiated on enriched UO/sub 2/ pellets for incorporation in full scale fuel rods to be irradiated in the Vallecitos Boiling Water Reactor. A wet nitrogen pyrohydrolysis step in conjunction with oxidationreduction cycling is being used to attain a satisfactory density exceeding 95% of theoretical at 1150 deg C. Apparatus and procedures being used for measurement of thermal conductivity and thermal expansion of sintered and cast uranium carbide are described. The coefficient of linear thermal expansion for a single specimen of 4.37 wt.% carbon sintered uranium carbide was determined to be 11.8 x 10/sup -6/ mm/mm- deg C, while that …
Date:
June 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Corrosion Tests in Molten Lead-Lead Chloride
Corrosion tests were run on some commercial grade metals, an alloy steel, stainless steels, chromium-- nickel-iron alloys, nickel base alloys, cobalt base alloys, and a chromium-- nickel-- cobalt-- iron ailoy in the system: leadlead chloride-lead chloride vapor at 528 deg C under an argon atmosphere. The following metals and alloys showed a corrosion rate of nine mils per month or less and did not suffer intergranular or other localized attack: tantalum, Incoloy 804, Hastelloy F, Carpenter-20 (Cb), stainless steels 316L, 318 Cb, Haynes Alloy 25, and Haynes Multimet (auth)
Date:
June 1, 1961
Creator:
Stolica, N. D.; Adams, G. S. & Bomar, M. R.
Object Type:
Report
System:
The UNT Digital Library
Chemical Technology Division, Unit Operations Section Monthly Progress Report, January 1962
None
Date:
June 1, 1962
Creator:
Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Object Type:
Report
System:
The UNT Digital Library
Physics Performance of the EBWE in Its Zero to 100 Mw Operation
The distribution of fuel elements in the EBWR core is described. Spike elements with boron steel strips were provided in the core for burnup allowance. A calibration of the control rod system with Zircaloy followers was made. Void coefficient measurements were made for two different H/sub 3/BO/sub 3/ concentrations. Power runs were made, and reactivity loss vs. power characteristic was obtained. Steam remaining in the core, or carryunder,'' was indicated. Some of the spikes were removed for increasing the power to 80 Mw. Calculations were made of the reactivity loss with coolant void and of the average void vs. reactor power. (D. L.C.)
Date:
June 1963
Creator:
Iskenderian, H. P.
Object Type:
Article
System:
The UNT Digital Library
Superconductivity in solid solutions of transition metal carbides. [NbC-TaC]
None
Date:
June 10, 1964
Creator:
Wells, M.
Object Type:
Report
System:
The UNT Digital Library
Primary Shield Optimization Survey for the PL-3 Reactor
A detailed study of four reactor and shield configurations was made. Two basic reactor types, the boiling water and pressurized water reactors were considered. Shield materials of lead-water and iron-water were used with varying thicknesses for determining the optimum shield configuration for the PL-3 reactor. Also presented is a survey of available shielding codes. (auth)
Date:
June 28, 1962
Creator:
Scoles, J. F. & Crouch, A. N.
Object Type:
Report
System:
The UNT Digital Library
Experimental and Analytical Reactivity Studies of Clean Critical Stainless Steel Cores
ABS>The results are presented of critical water height measurements made on close-packed lattices of Spert III, highly enriched, plate-type, stainless- steel-clad fuel elements. Experiments were conducted with cores containing no control rods and with cores containing a single, fully-inserted control rod. The "clean critical" data obtained in these experiments were used to test the validity of various aspects of a four-group, diffusion theory analysis of the full scale Spert III reactor. The results of the analyses of the rod-free and single-rodded critical lattices show that for such stainless steel cores k/sub eff/ can be calculated to within 1% DELTA k and that the Spert III control rod worth is calculable to a few tenths % DELTA k. (auth)
Date:
June 16, 1961
Creator:
Spano, A. H.
Object Type:
Report
System:
The UNT Digital Library
SNAP 2 PRIMARY SYSTEM TEST-OBJECTIVES, SYSTEM DESCRIPTION, AND PROCEDURES
The SNAP-2 Primary System Test loop fabrication was completed with associated flight prototype components including reactor core and boiler mockups for volume and DELTA P simulation, CRU-IIII NaK pump, compact heater, and expansion compensator. A mobile loading system was designed and fabricated with the capability of cleaning the NaK prior to final loop sealing. Loop descriptions, test objectives, and operating procedures are presented. (auth)
Date:
June 12, 1961
Creator:
Kikin, G.M.
Object Type:
Report
System:
The UNT Digital Library
A Small Scale Countercurrent Liquid-Liquid Extractor
Details of design and operation are given for a laboratorysize, 20 - stage, multiple-contact, "unlimited-feed'', countercurrent-flow, liquid-liquid extractor. The apparatus consists mainly of glass parts that are joined by polyethylene tubing and are mounted in a steel cradle that can rotate on its horizontal axis. The reservoirs for feed liquids are an integral part of the assembly, and proper rotation of the assembly causes the flow of liquids to, through and from the extractor. Testing and developing of, and small scale production by, extraction systems are conveniently carried out in this extractor. Stagewise and product samples can be readily obtained for study of the extraction behavior of the components of a liquid-liquid system. (auth)
Date:
June 1, 1961
Creator:
Wilhelm, H. A.
Object Type:
Report
System:
The UNT Digital Library
A PULSED EDDY CURRENT TEST SYSTEM USING REFLECTED FIELDS
An eddy current test system is described in which the test information is detected as a series of fields reflected from the metal surface and interior. Pulsed electromagnetic fields are caused to impinge upon the test specimen. These fields are restricted to a small cross sectional area over a path in space long enough to be useful for test purposes by devices called mask-apenture assemblies. This approach provides a number of advantages over conventional eddy current methods of comparable capabilities, including superior surface resolution, a reduction in circuit complexity, and an improvement in stability and reliability. Various applications and test results are discussed. (auth)
Date:
June 11, 1962
Creator:
Renken, C. J.
Object Type:
Article
System:
The UNT Digital Library
Dynamic Simulation of Multi-Pass Pressurized Water Nuclear Power Plants by Analog Computer Techniques
A kinetic model of the primary loop of a multi-pass pressurized water reactor power plant is developed to evaluate, by analog computer techniques, the transient response characteristics under conditions of steam generator load and reactor control rod perturbations. Using the 2-pass 28 Mw(t) SM-2 reactor as a typical plant, transient behavior patterns are illustrated and examined for a variety of load inputs, variations in plant constants, and analog model simplifications. (auth)
Date:
June 1, 1961
Creator:
Brondel, J. O.
Object Type:
Thesis or Dissertation
System:
The UNT Digital Library
FEASIBILITY OF Pu$sup 239$-U$sup 235$-FUELED CORES TO PREDICT Pu$sup 239$- FUELED CORE DIMENSIONS
Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region and U/sup 235/ in the outer core region. Core diluents, simulating fertile, structural, and coolant materials, are assumed identical in all cases. it is estimated that construction of the reflected two- region core with ratio of central core region volume to total core volume of 0.1 will theoretically decrease the calculated error in prediction of the critical size of a corresponding solely Pu/sup 239/-fueled assembly by a factor of about 10 to 20. (auth)
Date:
June 1, 1962
Creator:
Meneghetti, D. & Ishikawa, H.
Object Type:
Report
System:
The UNT Digital Library
PROGRAM ODD--A ONE-DIMENSIONAL MULTIGROUP CODE FOR THE IBM-7090 (ANP PROGRAM NO. 657)
The physical and mathematical reactor models which are used in Program ODD are discussed. In addition, the FORTRAN II source program listings, decimal data input sheets, and input and output for a sample case are given. Program ODD was designed to raake use of the Revised Nuclear Data System at ANPD which consists of twenty-five energy group cross-section data including high energy inelastic scattering matrices, resonance parameters for the resolved resonances, and thermalization scattering matrices for the near thermal energy region. The most unique aspect of the program is the mathematical technique employed for eliminating inner iterations and slow convergenc rates occasioned by the up- scattering'' in the thermalization region of the energy lattice. Direct inversion of the energy matrix coupling the thermal and last four epitherma groups provides simultaneous consistent solutions for thes groups within each power iteration. (auth)
Date:
June 30, 1961
Creator:
Fischer, P.G.; Wenstrup, F.D. & Hoffman, T.A.
Object Type:
Report
System:
The UNT Digital Library
ELECTROLYTIC DISSOLUTION OF NUCLEAR FUELS. PART III. STAINLESS STEEL (304) IN NITRATE SOLUTIONS
The potential-current density relationships for 304 stainless steel dissolution in a nitrate system were studied as a function of solution composition and temperature in order to optimize the conditions for electrolytic dissolution of ihis material. In the nitrate system, the anodic dissolution of steel takes place in the transpassive region. Under some conditions, deviations from Tafel behavior are observed which depend greatly on the nitrate and hydrogen ion concentration, and on temperature. A discussion of passivity, transpassivity, secondary passivation, the limiting current density, and the effect of alloy composition on the dissolution behavior is given. It was found that at temperatures above 60 deg C efficient dissolver operation should be possible over a wide range of solution compositions and at current densities up to 2 amp/cm/sup 2/. (auth)
Date:
June 1, 1962
Creator:
Aylward, J. R. & Whitener, E. M.
Object Type:
Report
System:
The UNT Digital Library
Thermal Stress Testing of Type 1 Fuel Plates
Thermal stress tests on Type 1, SM-1A Core II fuel ele-ment sections were performed to study plate distortion and determine its dependency on temperature distribution, temperature differential, initial flatness, and ripple length. Test results will be correlated with the analytical model and used to predict ripple growth in other plate-type fuel elements. The tests showed that ripple growth is dependent on initial flatness of the plate and that the characteristic shape of ripples is maintained at all temperature differentials: The tests also showed that the ripple growth rate for a ripple of 5 mil initial magnitude is approximately 0.12 mils/ deg F for a peak temperature differential of 103 deg F and that the apparent relationship between ripple net growth and length is 1.3 mil/in. of ripple for a peak temperature differential of 103 deg F. A permanent distortion of 2 mils for a complete temperature cycle from 0 to 103 to O deg F differential was found. The temperature profile across the plate width was found to affect the magnitude of ripple growth. (auth)
Date:
June 27, 1962
Creator:
Gebhardt, F. G.
Object Type:
Report
System:
The UNT Digital Library
Evaluation of macroreticular anion exchange resin - RTA-893-R
The macroreticular anion exchange resin, Amberlite IRA-900-OH, and an experimental resin from Ionac Chemical Company were irradiated by a /sup 60/Co source to doses of 5 x 10/sup 7/ rad and 10/sup 8/ rad. These doses approximate or exceed the dose encountered by the deionizer resins in 100-Area service. The loss in exchange capacity and the volume changes of Amberlite IRA-900 on irradiation were similar to those found previously for Amberlite IRA-400. The Ionac experimental resin, which had a considerably lower initial exchange capacity, likewise was not stable to radiation. It is concluded that Amberlite IRA-900 offers no advantage over Amberlite IRA-400 for 100-Area purification service. Since there is little justification for further evaluation of macroreticular resin for 100-Area use, the present work completes RTA-893-R.
Date:
June 7, 1965
Creator:
Baumann, E. W.
Object Type:
Report
System:
The UNT Digital Library
Spert I Destructive Test Program Safety Analysis Report
The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor period range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation …
Date:
June 15, 1962
Creator:
Spano, A. H. & Miller, R. W.
Object Type:
Report
System:
The UNT Digital Library
ECONOMIC FACTORS OF MFP THERMOELECTRIC GENERATORS. Interim Report
Mixed Fission Products (MFP) for use as a heat source for thermoelectric generators will become increasingly available in the coming years. The Atomic Energy Conamission sponsored program on solidification of nuclear wastes is now entering the hot-bench scale test phase. During this phase approximately 5000 thermal watts of two year old MFP could be produced monthly. Two different types of hot calcination pilot plants are planned for installation at the Hanford National Laboratories in the 1964 to 1966 time period. Each of these plants should be able to produce 160,000 thermal watts of two year MFP and 16,000 thermal watts of ten year MFP on a monthly basis. During this phase, MFP costs should be less than 15 per ihermal watt for two year MFP and 50 for ten year MFP. This cost includes operation of the plant solely to obtain heat sources and sealing the MFP into fuel containers. A full scale plant for a 15,000 Mw(e) nuclear economy is estimated to produce four to five times as much MFP as either of the pilot plants. Costs will be dependent upon AEC policy in effect at the time the plant is operating. lf the policy indicates that the full …
Date:
June 1, 1962
Creator:
Barmat, M.
Object Type:
Report
System:
The UNT Digital Library
Secondary Isotope Effects in Molecular Structure
A study was made to determine whether secondary iso tope effects also occur in molecular structure. Electron diffraction studies were carried out on ethane and deuteroethane. In C/sub 2/H/sub 6/ the mean C-C and C-H bond lengths found agreed very closely with values determined for other paraffin hydrocarbons, and the C--H bond showed a normal primary isotope effect (~ 0.005 A) similar to that found in methane when H is replaced by O. The output of the leastsquares analysis suggested that the mean C-- C bond length in C/sub 2/D/sub 6/ is shorter than in C/sub 2/H/sub 6/ and by about 0.004 A. Th e decrease seemed to be real for the apparent uncertainty was not much greater than 0.001 A. (M.C.G.)
Date:
June 15, 1962
Creator:
Bartell, L. S.
Object Type:
Article
System:
The UNT Digital Library
LA CROSSE BOILING WATER REACTOR. Monthly Operational Report No. 22, April 1969
None
Date:
June 1, 1969
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
A Neutron Diffraction Study of Krypton in the Liquid State
A study was made of the neutron diffraction patterns obtained from Kr liquid under seventeen conditions of temperature and pressure at 117 to 210 deg K. The low temperatures were used because the diffraction patterns and the corresponding radial distribution functions are more sharply defined near the liqdid triple point. (J.R.D.)
Date:
June 1, 1961
Creator:
Clayton, G.T. & Heaton, L.
Object Type:
Report
System:
The UNT Digital Library