PATHFINDER ATOMIC POWER PLANT STEAM SEPARATOR DEVELOPMENT (open access)

PATHFINDER ATOMIC POWER PLANT STEAM SEPARATOR DEVELOPMENT

Development of a steam separator the Pathfinder Reactor is reported. A full-scale separator model was developed through the combination of scale-model testing and the application of principles associated with the existing theory of centrifugal separation. This model was put through full-scale air-water tests which led to modifications and a final design which meets Pathfinder requirements. Design data are included for the reactor and the steam separator. (J.R.D.)
Date: June 15, 1962
Creator: Kutsch, G. C.; Swanson, D. H. & Yant, H. W.
Object Type: Report
System: The UNT Digital Library
A STUDY OF THE VISCOSITY OF THORIA SOLS (open access)

A STUDY OF THE VISCOSITY OF THORIA SOLS

None
Date: June 13, 1962
Creator: Sturch, E.
Object Type: Report
System: The UNT Digital Library
EQUIPOISE 3A (open access)

EQUIPOISE 3A

None
Date: June 1, 1962
Creator: Nestor, C. W. Jr.
Object Type: Report
System: The UNT Digital Library
COMPILER INTO GEORGE ASSEMBLY ROUTINE (open access)

COMPILER INTO GEORGE ASSEMBLY ROUTINE

This program of the GEORGE Assembly Routine (GAR) will accept Fortran- like statements from paper tape and create the GAR language program on tape. This includes the needed calls for common subroutines and the reservations for the named variables and temporaries. The original statements in Fortran are carried along as remarks. The GAR language program may then be processed in the usual way by the GEORGE Assembly Routine, giving machine-language code. The level of sophistication of the source language is roughly equal to that of Fortransit or SALT. (auth)
Date: June 1, 1962
Creator: George, R.
Object Type: Report
System: The UNT Digital Library
Effects of Seismic Vibrations on the Experimental Gas-Cooled Reactor (open access)

Effects of Seismic Vibrations on the Experimental Gas-Cooled Reactor

The effects of seismic vibrations on the dynamic behavior of a composite system were analyzed. The equations of motion were derived and soIved with special emphasis on determining the resulting stresses. The method of analysis thus developed was applied to the composite structure consisting of the core, pressure vessel, and supporting skirt of the Experimental Gas-Cooled Reactor (EGCR). A system with three degrees of freedom was considered in order to determine the effects of an earthquake of the maximum intensity expected in the area surrounding Oak Ridge, Tennessee. The system of equations of motion was solved both numerically and analytically, and the resonant frequencies were determined. The seismic effect was shown to be small when the frequency of the seismic disturbance coincided with a natural frequency of the system. In particular, the shear stresses in the graphite core were shown to be negligible. (auth)
Date: June 22, 1962
Creator: Witt, F.J.; Carver, D.R. & Maxwell, R.L.
Object Type: Report
System: The UNT Digital Library
Periodic Characterization of Radioactive Waste Disposal Effluents. Core I Seed 3. Test Evaluation (open access)

Periodic Characterization of Radioactive Waste Disposal Effluents. Core I Seed 3. Test Evaluation

Data are given on the radioactive nuclides present in waste disposal effluents during Nov. 1961. The concentrations of all activities was well within established limits for discharge to the environment. (C.H.)
Date: June 13, 1962
Creator: unknown
Object Type: Report
System: The UNT Digital Library
Pathfinder Atomic Power Plant Technical Progress Report, January 1962-March 1962 (open access)

Pathfinder Atomic Power Plant Technical Progress Report, January 1962-March 1962

S>Progress is reported on the research and development program being performed in connection with the design of the Pathfinder Atomic Power Plant. The topics covered include: fuel material cladding, bonding, and irradiation testing; heat transfer and fluid flow; fuel element manufacturing research and development; low-enrichment superheater fuel element design; mechanical studies of control rods, guide tubes, and control rod drives; reactor physics; and reactor and system dynamics. (C.H.)
Date: June 30, 1962
Creator: unknown
Object Type: Report
System: The UNT Digital Library
Propane Vibrational Analysis (open access)

Propane Vibrational Analysis

Using the group vibratibn method of McMurry, the normal frequencies and coordinates of propane and three of its symmetrical deuterium substituted compounds were obtained. The force constants used were taken from a variety of previous works on hydrocarbons. The results give reasonable agreements with the experimental frequency and mode assignments of others. (auth)
Date: June 12, 1962
Creator: Marshall, G. D.
Object Type: Report
System: The UNT Digital Library
Quarterly Status Report on LAMPRE Program for Period Ending May 20, 1962 (open access)

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 (open access)

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 (open access)

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 (open access)

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
Chemical Technology Division, Unit Operations Section Monthly Progress Report, January 1962 (open access)

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
Primary Shield Optimization Survey for the PL-3 Reactor (open access)

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
A PULSED EDDY CURRENT TEST SYSTEM USING REFLECTED FIELDS (open access)

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
FEASIBILITY OF Pu$sup 239$-U$sup 235$-FUELED CORES TO PREDICT Pu$sup 239$- FUELED CORE DIMENSIONS (open access)

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
ELECTROLYTIC DISSOLUTION OF NUCLEAR FUELS. PART III. STAINLESS STEEL (304) IN NITRATE SOLUTIONS (open access)

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 (open access)

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
Spert I Destructive Test Program Safety Analysis Report (open access)

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 (open access)

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 (open access)

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
CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, JANUARY-MARCH 1962 (open access)

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, JANUARY-MARCH 1962

Additional experiments conducted on nitridation of irradiated U-fissium fuel pins revealed that irradiation does not greatly affect the nitridation rate at 300 deg C. In skullreclamation development, a phase separation common to both the blanket- and skull-processes was investigated in which a 50% Mg-- Zn supernatant solution was removed from precipitated U metal. In most runs the supernatant phase was removed with negligible U entrainment. The reduction rate of ZrO by Zn--Mg solution under skull-recovery process conditions was found to be lower than that of U oxides. It may be possible in this process to effect some Zr separation by limiting the reduction time to that necessary for U. Methods of procesging EBR-II fuels are being investigated to establish methods of separating rare earths from Pu. Development work on preparation of UC by addition of C to U dissolved in liquid metal media showed that the limited addition of alkali metals improves C wetting and the tests showod high C-to-U ratios and high O/sub 2/ contamination; procedure and equipment improvements are being made. Studies are in progress to evaluate the compatibility of various materials with the liquid metal-salt systems contemplated for reactor fuel reprocessing. Tungsten appears to have high corrosion …
Date: June 1, 1962
Creator: unknown
Object Type: Report
System: The UNT Digital Library
Irradiation Behavior of Restrained and Vented Uranium-2 w/o Zirconium Alloy. Final Report-Programs 6.1.22 and 6.1.27 (open access)

Irradiation Behavior of Restrained and Vented Uranium-2 w/o Zirconium Alloy. Final Report-Programs 6.1.22 and 6.1.27

Twelve 0.22-in.-diameter fuel specimens containing a longitudinal central vent and clad with 0.010 in. of Type 304 stainless steel were irradiated to evaluate the effect of restraint and a central vent on fuel element stability. The cladding of 10 of the specimens contained porous end plugs to vent any released fission gas and thus to minimize the buildup of gas pressure within the stainless steel cladding. The specimens consisted of a 20% enriched uranium--2 wt% zirconium alloy core surrounded by a natural uranium--2 wt% zirconium alloy sleeve. Eight of the specimens were irradiated to burnups of the enriched core of 6.9 to 12.8% of all atoms (1.2 to 2.2 at.% of the duplex assembly) at maximum fuel temperatures ranging from 280 to 760 deg C. Most of the clad specimens exhibited negligible volume increases as a result of irradiation. Two specimens containing central vents but unclad were irradiated together with the clad specimens in an attempt to differentiate between the effects due to a central vent and the effects due to cladding. The central vent in itself did not appear to reduce the swelling characteristics of the alloy. Mechanical restraint appeared to have extended the useful operating temperatures of the …
Date: June 1, 1962
Creator: Horak, J. A.; Kittel, J. H. & Yaggee, F. L.
Object Type: Report
System: The UNT Digital Library
Chemical Processing Technology Quarterly Progress Report, January-March 1962 (open access)

Chemical Processing Technology Quarterly Progress Report, January-March 1962

The processing of Al fuel, principally of the MTR-ETR type, is reported. Processing rate averaged 90% of flow sheet values for the entire operating period, and a U recovery of 99.85% was achieved. Aqueous Zr fuel processing studles continued with the objective of adapting the HF process to continuous dissolution-complexing in order to increase the capacity of the ICPP process while using as much existing equipment as possible to minimize costs. Good results were indicated in a 190-hr run dissolving 2% U-Zr fuel in a Monel dissolver using 4.8M HF-0.03M HNO/sub 3/ dissolvent at 200 deg F; insoluble material did not accumulate in the dissolver, although a U-containing film was formed, apparently in small, equilibrium quantity. Shorter term continuous laboratory dissolutions indicated that 4.8M acid was preferable to 10M acid for the acid feed rate/fuel surface ratios proposed, resulting in dissolver products of greater stability and higher uranium content. Additional laboratory data are presented on UF/sub 4/ hydrate form and solubility, together with maximum dissolvable U compositions with Zircaloy under various flowsheet condltions. Processing of Al alloys containing high Si was found to present no unusual problems in laboratory studies. Siliceous residues resulting from dissolution of Al-U alloys containing 2% …
Date: June 29, 1962
Creator: Bower, J. R.
Object Type: Report
System: The UNT Digital Library