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.)

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.)

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.)

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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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)

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)

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)

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)

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 …

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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.)

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 …

The influence of temperature on the critical buckling or bare graphite assemblies with various carbon-to-uranium235 molar ratios has been measured. A range from l185: 1 to 2l,690: 1 was covered, for 45 to 1205'F. Preliminary results indicate that the fractional rate of change of critical buckling with core temperature varies monotonically with C/U2as ratio by a factor of five over the factor-of-eighteen range in gross C/U2as ratio. This quantity appears to approach asymptotically a value near 2%/l00"F at very high C/U2ss ratios. (auth)

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 …

A force multiplier was designed. This piece of equipment was made to increase the gripping force presently available in the Model 8 master slave. The force multiplier described incorporates a clamp which can be quickly attached to and detached from the master slave hand. (auth)

The need for a source of pulses randomly distributed in time according to the Poisson distribution is discussed. A theoretical analysis showed that these pulses may be generated by employing an electrical white noise'' source. Calculations indicated the feasibility of a generator capable of producing Poisson pulse at a maximum average rate of one megacycle. The results of laboratory experiments support the calculations and demonstrate the nadequacy of using periodic pulses to determine instrument response to random pulses. A design of a nuclear signal simulator incorporating the Poisson pulse generator is suggested. (auth)

The pressure drop of a new fuel element design concept of spacing rods by means of helical wire springs was investigated experimentally and analytically. Extensive single- and two-phase pressure drop data at 1,000 psia were obtained for one flow geometry and helical spring spacer. Test conditions ranged from 0.7 to 1.2 x 10/sup 6/ lb/hr ft/sup 2/ in mass velocity and from 0 to 15% in quality. The effect of the specific spring which was tested was to increase the over-all pressure drop by 70%. A general analytical model was developed to predict the pressure drop of an element with helical spring spacers when the pressure drop without springs is known. The accuracy of the model, compared to the experimental data, was better than plus or minus 22%. The analytical model allows determination and evaluation of an optimum helical spring spacer design, so that pressure drop will not be a serious disadvantage. (auth)

The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.