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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)

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)

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 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.)

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)

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 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 problem of determining the dynamic properties of the steam void fraction undergoing random variations at a particular location in a boiling channel was studied. Emphasis was placed on a gamma attenuation method and on a method employing sensitive flowmeters at suitable locations of the channel. The dynamic properties of interest were the autocorrelation function and the power density spectrum of the variations. Equations were derived for computing the desired quantities on the basis of gamma records obtained during actual boiling experiments and from runs with empty and full, nonboiling channels. The equations consider the statistical variations of the gamma source and detector. A procedure was outlined for estimating the measurement effort to obtdin a predetermined accuracy. Two models were developed to correlate the variations of flow with fluctuations of the steam void. Both models were based on the mass continuity equation for a two-phase fluid. The simpler model leads to a first- order, linear differential equation with randomly varying coefficients. The desired quantities can be computed from the equation. The second model leads to a linear integral equation of the third kind. This equation relates the power density spectrum of the void fraction with the spectra of the water …

The limitations and operating techniques which were in effect at ANPD for the prevention of criticality accidents are summarized. The standards followed by the atomic industry were followed; however, the safe mass of U/sup 235/ moderated with beryllium oxide and hydrogeneous materials was based upon criticality experiments conducted at ANPD. Although the guide was primarily for the use of the ANPD nuclear safety control organization, it may also be of assistance to designers and operating management in maintaining nuclear safety. (auth)

The need for the first core for EBR-11 resulted in the design and construction of the Initial Fuel Loading Facility for this reactor. The plant was built to provide the required initial loading, to train personnel, and to test prototype equipment for the remote reprocessing of fuel materials in the EBR- II Fuel Cycle Facility. The facilities include: remotely manipulated melting, casting, and pin processing equipment, a degreaser, hoods and their atmospheric control system, a gas-purification system, fuelelement-assembly equipment, mold- preparation and balance room, bonding furnaces, a maintenance shop, and a change area. (auth)

A method is presented for the aerothermodynamic optimization of the net power and/or propulsive thrust per unit reactor free flow area of a nuclear power plant operating on the Brayton cycle. A system so optimized will translate into the minimum size, therefore the minimum weight, nuclear system for any selection of reactor materials, lifetime, and fuel loading. The theory and development of the thermodynamic optimization process, the importance and effect of various parameters, and specific methods to be employed in the optimization of the various forms of the Brayton cycle are discussed. A sample calculation for the case of the ramjet application is included. The results of the application of these techniques to any Brayton cycle system may be used in conjunction with nuclear sizing methods, for beryllia-moderated reactors, to determine the required reactor size as a function of fuel loading and reactivity requirements. (auth)

A semicontinuous process was developed for dissolving the solid residues that are produced in the Ca-reduction process for converting plutonium fluoride to plutonium metal. The slag and crucible residues are charged to a column-type dissolver and are there exposed to a flowing solution of hot nitric acid followed by a nitric acid-aluminum nitrate solution. The resultant solution is suitable as feed to an ion exchange process for recovering Pu. A method was demonstrated for recovering Pu from the dissolver solution by anion exchange in agitated beds of resin. The use of agitated beds of resin eliminates the need for filtering the feed. (auth)

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This report talks about Analytical and preliminary design studies of nuclear rocket propulsion systems. Volume VI. Compilation of hydrogen properties,

Fuel Fabrication and Evaluation. Nearly stoichiometric, low nitrogen- content UC was synthesized in the plutonium glove box atmosphere. PuC and PuC- UC preparation studies were initiated by the oxide-carbon reaction. UC with nickel additions lowered required sintering temperatures, improved density over control straight UC samples, but did not improve densities over those obtained previously with UC. Microprobe analysis of niobium and 21/4 Cr- 1 Mo compatibility samples tested 4,000 hr at 820 deg C, showed no uranium penetration. Irradiation Test. The W1-1 capsule containing two clad UC specimens continued operation to 11,000 MW-d/ton by the end of April. The W1-2 capsule was inserted in WTR, but had to be removed after a short time because of a radioactive gas leak. Plutonium Facilities. The facility at The Carborundum Company started operation with plutonium in March. Modifications and maintenance continue to be required and are being completed with minimum effect on experimental work schedules. The major modifications of the facility at NDA, the once-through nitrogen and helium systems, were completed and initial operation is satisfactory. A performance test was initiated prior to operations with plutonium. Design and construction of equipment for the furnace box are in progress. (auth)

To determine the effects of gamma radiation upon transducers, switches, and motors, two test assemblies were irradiated in the ANL Gamma Irradiation Facility at levels of 1.50 x 10/sup 5/to 2.10 x 10/sup 6/ rad/hr to a total dose of 1.0 x 10/sup 9/ rad. The components tested were a motor, a straingage pressure pickup, a quartz pressure pickup, and a high-temperature switch. The performances of the straingage pickup and snap switch were highly satisfactory over the test period. The quartz pickup, however, exhibited a sensitivity to radiation which rendered it ineffective for use in such an environment. Radiation damage to the motor was manifest as degradation of insulation, and evaporation and condensation of some of the products of degradation upon the walls of the test containment shell. As a result, the motor lost power, indicating a need for insulation more resistant to gamma radiation. Radistion damage to paints, varnish, grease, and conventional insu lations was noted, but the extent of this effect was not considered detrimental to the test. (auth)

LAMPRE-L The reactor was started up and held st 160 deg C and at hot critical for various measurements. Some of the 160 deg C measurements included rod worth, temperature coefficient, and shim-down extrapolated mass. In the approach to hot critical, multiplication was measured during the temperature increase from 160 to 500 deg C and found to increase markedly when the core meltsd. Reactivity changes due to jarring of the fuel capsules to remove bubbles and voids in the melted fuel were observed. Zero measurements made at 48O deg C are reported. Power coefficient measurements at 0 to 46 kw and power demand experiments were made, and the 50-kw transfer function is plotted. Shielding inadequacies are discussed briefly. The results of a fuel pin scan are given. Observations on the performance of the sodium system are discuaaed. Type 17-4 PH stainless steel was found to undergo additional grain boundary precipitution during long-term exposure to temperatures below the hardening temperature, and measurements of its properties are reported. fuel and Container Development. Impurities in the LAaMPRE-1 fuel which has a corrosive effect on the fuel container are discussed. Observations in the Pu -Ce-Co phase diagram are reported. Different types of LAMPRE-1 capsule …

A detailed description of the site, facilities, and characteristics of the UTR Test Reactor and operations to be performed are presented. Also included is an evaluation of safety considerations and an analysis of potential hazards which might result from installation or operation of this reactor. Based on satisfactory operation of similar prototype reactors in the past and on a detailed hazards analysis, it was concluded that the operation of the reactor at this site wiil not constitute a potential hazard to property or persons in the surrounding area. The application for construction permit and facility license is included. (M.C.G.)

A preliminary chemical flowsheet was developed on a laboratory scale for the preparation of feed and two solvent extraction cycles for the processing of Si-containing Al--U alloy, 20% enriched, fuel elements. Major process steps include dissolution of the fuel assembly in Hg-catalyzed ritric acid, removal of silica by coagulation during feed clarification, and recovery of U and pu by extraction with a tributyl phosphate solvent and selective stripping. Criticality control in existing process equipment is maintained by the use of internal neutron poisons and concentration control throughout the process. (auth)

Methods and techniques for measuring the conductances of reactor solutions at elevated temperatures and to interpret the results with electrolytic solution theories are discussed. Criteria applied to room-temperature conductances for determining complete ionization of 1-1 electrolytes were applied to the Noyes high-temperature conductance data obtdined on NaCl, HCl, KCl, NaOH, AgNO3, and H/sub 3/PO/sub 4/ solutions. lt was found that the first four electrolytes behave as strong electrolytes up through temperatures of 218, 218, 281, and 156 deg C, respectively. Application of the Bjerrum theory showed that, somewhat above each of these temperatures, these electrolytes begin to associate by ion-pair formation. H/sub 3/PO/sub 4/ solutions associated at room temperature and above. AgNO/sub 3/ solutions appeared to asscciate at all temperatures. The equivalent conductances and infinite dilution and the dissociation constants for each of the electrolytes were calculatsd. An apparatus was constructed and measurements were made on the conductances of LiOH and NH/sub 4/OH solutions over the temperature range of 100 to 520 deg F. Equivalent conductances at infinite dilution were calculated. The pH of reactor coolants using LiOH or NH/sub 4/OH was calculated from the disscciation constants for water, LiOH, and NH/sub 4/OH at elevated temperatures. (auth)

"Ultrasene" typically contained 57 wt % branched paraffins and cycloparaffins, 40 wt% n-paraffins, 3 wt% aromatics, and less than 0.1 wt% olefins. The n-paraffinic fraction from "Ultrasene" was stable to chemical and radiolytic degradation, whereas the combined branched paraffinic and cycloparaffinic fraction, and the aromatic fraction, were significantly less stable. (auth)