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.)

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Critical experiments were performed with two assemblies simulating a cold clean, and an end-of-cycle,- Argonne High Flux Reactor, core. Data were obtained for flux distributions; cadmium ratios; temperature and void coefficients; and control rod, beam hole, and reflector worths. The data obtained furnished confirmation of theoretical predictions. The peak 2200-m/sec flux per unit power was measured as 3 x 10/sup 7/ n/(cm/sup 2/)(sec)(watt) for both cores. The two cores had internal H/sub 2/O thermal columns, 12.7 cm x 12.7 cm x 50.8 cm. These were enclosed by 100-liter fuel zones. The radial reflector was 90% beryllium containing 10% H/sub 2/0 plus Plexiglas by volume. The top and bottom reflectors were H/sub 2/O. The critical mass was 3.58 kg U/sup 235/ with a 1.16 metal-towater ratio in the fuel zone. The critical mass with a 1.60 metal- to-water ratio, taking into account 34.3 kg Type 304 stainless steel, was 7.15 kg U/sup 235/. (auth)

ent temperature variation in a thermally orthotropic plate which is subjected to an arbitrary heating rate distribution along one face with all other surfaces being insulated. Dimensionless temperature histories and distributions determined from this solution are presented for the special, but representative, case of a linearly varying heating rate distribution on plates with varying degrees of thermal orthotropy. These results establish quantitatively the value of a material with high planar and low normal thermal conductivities for applications where it is desired to maintain minimum temperatures on the rear or unheated surface of a heat shield when the heated surface is subjected to a very non-uniform heating rate distribution. The applicability of simplifying assumptions in analyzing such a system is discussed. Experimental temperature measurements in a pyrolytic graphite plate heated by an oxyacetylene flame were made to verify the analytical results. Achievement of satisfactory agreement wss found to be dependent upon use of thermal property values differing from those presently available for this material. This is not unusual in that differences in production methods are known to introduce substantial property variations in anisotropic materials such as pyrolytic graphite. (auth)

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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)

Drag coefficients were determined for various irregular objects such as glass fragments, stones, steel fragments, and spheres by means of drop tests for use in a mathematical model to correlate nuclear explosion blast experiments. Drop tests were also made on small laboratory animals and extrapolated to estimate the drag properties of man. A method was developed to estimate the average drag properties of man from his total surface area. (D.L.C.)

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)

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.)

Mathematical models of the time-dependent behavior of two-phase natural- circulation loops were used to predict the operation and to explain the unusual instability sometimes observed. The initial results obtained for a loop similar to the Univ. of Minnesota loop were used to formulate a more complex and accurate model, and the predicted transient behavior was in close agreement with the experimental results from the Minnesota loop. For a 300psia, high-pressure loop, unstable oscillatory behavior was predicted under certain conditions and stable behavior under others. Closed unstable regions rather than limits were predicted, and the specifications of stability in terms of a single parameter were found to be impossible. The great difference in oscillatory frequencies observed at low and high pressures was found to be due largely to the system geometry. The criterion for the absence of oscillations was found to be similar to one of the criteria for stability of chemical reaction systems. (D.L.C.)

A test was carried out to determine the adequacy of storage capacity and operating procedures of the radioactive waste disposal system during a normal reactor plant warmup. The capacity and operating procedures were found to be adequate. It was impossible to perform a complete material balance based on existing level instrumentation and using the data required by the test procedure. Approximately 21,290 gal. of waste were received in the system and 13,210 gal. were discharged to the river with a total activity of 1200 mu c. A quantity of 6,670 gal. of reactor coolant effluent was processed. Approximately 634 lb of combustible waste were incinerated. (M.C.G.)

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.)

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)

Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)

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.)

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.)

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)

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 …

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)

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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.)

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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)