The purpose of this document is to update 190-C pump capacity information previous released in HW-52449{sup 1} and HW-58580{sup 2}. Improvements in motor cooling has resulted in raising the previous 3500 HP limit to 3660 HP{sup 3} thus increasing total pumping capacity.

This report supersedes K-1-3425 and LA-2327(Prelim). The first experiment (LAMPRE I) in a program to develop molten plutonium fuels for fast reactors is described and the hazards associated with reactor operation are discussed and evaluated. The reactor desc=iption includes fuel element design, core configuration, sodium coolant system control, safety systems, fuel capsule charger, cover gas system, and shielding. Information of the site comprises population in surrounding areas, meteorological data, geology, and details of the reactor building. The hazmalfunction of the several elements comprising the reactor system. A calculation on the effect of fuel element bowiing appears in an appendix. (auth)

The feasibility of using electroless nickel, a chemical deposit containing about 10 wt.% phosphorous in nickel, as the brazing alloy for assembling tubular stainless steel fuel elements of the type specified in Core I of the N. S. Savannah was investigated. This material was selected primarily because of the ease of braze-metal preplacement by chemical deposition of the alloy on type 304 stainiess steel ferrule spacers, prior to fuelbundle assembly. Brazed joints produced by this method were generally characterized by a relatively ductile solid-solution region at the thinnest portions of the fillet. This ductile zone should minimize the possibility of complete propagation of hairline cracks, which form in the brittle, eutectic regions of fillet. The microstructural appearance of the electroless-nickel joints was not appreciably affected by variations in the brazing temperature from 1750 to 1900 deg F or the brazing time from 15 to 60 min. Several plating solutions were evaluated and all were found to be capable of producing deposits suitable for brazing applications. Corrosion tests conducted in static 525 deg F water indicated that no significant attack of joints brazed with electroless nickel had occurred after 300-hr exposure. A small fuel bundle was successfully assembled by brazing with …

This report presents a reactor design to meet the needs of the basic research program at Argonne National Laboratory. The program requires some irradiations in thermal neutron flux approaching 5 x 10/sup 15/n/ (cm/sup 2/ )(sec) and some beam experiments having over 10/sup 15/ n/(cm/sup 2/) available. Parametric studies indicate that the highest peak thermal flux per unit of reactor power can be obtained in an internal H/sub 2/ reflector having a radius of about 6 cm. Since the design effonts recognized total reactor cost as a mator parameter, an all H/sub 2/O-cooled system was adopted as most expedient. Further, the design and calculations effont was directed toward operating conditions which require little or no research effort. This latter point is a recognition both of the over-all cost plus a desire to meet the needs of research in the shortest possible time. (auth)

Four reinforced-concrete arch structures, with the top of arch crown 4 ft below ground surface, were exposed at high overpressure ranges from Priscilla Burst in order to obtain data on their resistance to blast, radiation, and missile hazards. The four structures received actual air overpressures of 56, 124, and 199 psi and suffered only minor damage, all remaining structurally serviceable. The entranceway used for the structures sealed out the air pressure. It was not designed to attenuate radiation and thus did not provide adequate radiation protection for personnel. There were no missile and apparently no dust hazards in any of the structures. Results of the test indicate that an underground reinforced-concrete arch is an excellent structural shape for resisting the effects of a kiloton-range air burst. (C.H.)

An analytical prediction of the burnout distributions in particular SM-1 fuel elements and absorber sections after 10.5 MWYR of core energy release is given. The distributions are based on the results of a one-shot, non-uniform burnout calculation, and are pre sented for both fuel and boron -10 depletion. Particular emphasis is placed on those elements removed from the SM-1 core in March 1959, since their subsequent burnup analysis by ORNL should provide a valuable check on the analytical models employed. (auth)

Calculations are given for eighteen stainless steel clad helium bonded specimens of UO/sub 2/-ThO/sub 2/ containing normal U to be placed in 6 holes in a holder in a position of the ORR not to exceed a peak unperturbed flux of 4 x 10/ sup 14/ n/ cm/sup 2//sec and irradiated to a peak nvt of 1.96 x 10/sup 21/

4 -Metallurgical Processing. A direct-cycle fuel reprocessing plant is being designed for pyrometaliurgical processing of discharged power reactor fuel elements. Irradiation tests on cerium-bearing glass samples for shielding windows revealed that the optimum cerium content is less than the nominal amount originally specified. The light output of gammairadiated mercury vapor lamps was determined to be about 55% of original after an exposure of 1.1 x 10/sup 9/ rads. Analyses of the composition of 30 ten-kilogram ingots of natural U-5% fissium alloy prepared in the melt-refining furnace indicate that not all of the added Zr and Mo went into solution. Irradiation tests have shown that natural rubber formulated with an antioxidant (Antiox 4010) is satisfactory for cable insulation at radiation levels to 2 x 10/sup 8/ rads. Four 2-kilogram scale runs were made to study the meltrefining characteristics of high Pu (20%) -- U --flssium alloys. A further investigation was made of the possible Zr contamination of molten U and its Ce alloy resulting from prolonged holding at 1400 deg C in stabilized ZrO/sub 2/ crucibles. Experiments at 1700 deg C showed considerable evolution of CO as a result of reaction of ZrO/sub 2/ with graphite. Reduction of oxide coatings on …

Pu production from separation plants during May was 116% commitments. UO{sub 3} production and shipments met schedules. Button output and shape production was 97 and 121% of schedule/forecast. Recuplex (product recovery) operated at record rates. Processing at Purex was carried out with the HS column bypassed. Palm processing resulted in excellent product quality but with low yield. A sample of fission products was prepared for Curtiss-Wright. Piping modifications were made to the Purex Pu ion exchange units. One Redox feed batch was prepared with dichromate oxidation; the U and Pu streams increased (Ru) as anticipated. Containers and casks were designed for fission product recovery. Design of installation for subassembly of Pit 65 weapon components was begun.

A gamma scintillation spectrometer was used to measure diffusivity of uranyl nitrate in water during preliminary capillary experiments. During Fluorox run FBR-22, 90.4% of the theoretical amount of UF/sub 6/ formed was collected in cold traps and chemical traps. Toroid tests of flame calcined mixed Th-U oxide showed low corrosion rates, small changes in particle size and a low solubilization of uranium, while denitration of uranyl nitrate in a fluidized bed resulted in particle growth with uniform layers of uranium oxide. A half-time of 30 min for uranium anion exchange was measured in differential bed studies of uranium sorption on Dowex 21K. The Darex Reference flowsheet operation resulted in chloride removal to less than 50 ppm in solvent extraction feed from APPR head- end treatment. Unirradiated prototype Consolidated Edison pins were dejacketed with 6 M H/sub 2/SO/sub 4/ with uranium losses to the dejacketing solution of approximately 0.2%. An optimum procedure was developed for clarifying large batches of solvent extraction feed by sand bed filtration. Sheared sections of stainless steel clad UO/sub 2/ were completely leached in onehalf the time required for equal lengths of stainless tubes containing uncrushed pellets. Abrasive disc wheel to metal removal ratios were measured at …

In a preliminary experiment, the integral diffusivity of 1 M FeCl/sub 3/ solution varied uniformly with the fraction difiused: from 0.3 x 10/sup -6/ cm/ sup 2//sec for 10 per cent diffused to 0.8 x 10/sup -6/ for 30 per cent diffused. A short Fluorox run made with crude UF/sub 4/ in the 4-in. fluidized bed showed that UF/sub 6/ could be produced from the impure feed. Denitration of Th(NO/sub 3/)/sub 4/ solutions on fluidized or mechanically agitated beds of ThO/sub 2/ and gave fine ThO/sub 2/ particles for all conditions tested. The rate of sorption of uranium into 40 micron Dowex 21K resin particles from a 0.0042 M uranyl sulfate solution was studied by measuring the uranium loading on individual beads as a function of time. Chloride concentrations of 28 to 51 ppm were produced in the solvent extraction feeds during five feed adjustment runs made with the Darex Reference flowsheet. Decladding of SS-clad /sub 4/ gave essentially identical results as batch decladding. When a Mark I prototype assembly was sheared into 0.75-in. lengths with a"plane of contact" blade in the 126-ton Manco shear, 22.2 g of metal fines (304L stainless steel) 1680 microns or less in size was …

In the reaction of silicon halides with graphite to form silicon carbide, thermodynamic conditions were determined for the formation of SiC, free of elemental silicon. The use of these conditions was designed to limit the rate of coating formation by the rate of diffusion of carbon through the coating, and render the operation independent of the vapor-flow factors which normally limit the uniformity of vapor-deposited coatings. Although a wide range of pressure- temperaturecomposition conditions was explored, it was not possible to duplicate the success previously obtained in applying the method to NbC, TaC, TiC, and ZrC coatings. Fundamental differences in the characteristics of the carbides which may account for the difference in behavior are the lack of a range of homogeneity in beta SiC crystal structure, and the fact that SiC undergoes a modification from the cubic beta to a hexagonal form at l900 to 2000 deg C.There remains the prospect of forming a uniform SiC ''sponge'' by the present process which can be subsequently impregnated with metallic silicon to form an oxidation-resistant cpating. (auth) l6200 Preliminary results were obtained on the value that commercially pure Pu (95% Pu/sub 235/ and 5% Pu/subp 240/) has when used as nuclear fuel. …

Many of the features of plutonium facilities have been covered, and a number of them have been omitted. A great variety of safety equipment is available, together with trained personnel to operate it. All of these devices, however, do not assure a contamination-free operation. Basically, the careful design of enclosures, experimental equipment, and procedures when handled by trained personnel represents the only approach to the problem of plutonium handling. (auth)

Studies are under ray in Chemical Research and Development to-provide the technical ``know-how`` necessary to calcine Purex 1WW and safely store the solid product. The solid product can be stored as the particulate solid from the calciner or, after further treatment, as a solid matrix or melt. In this regard, Chemical Research is studying a PO{sub 4}BO{sub 2} melt. This document presents expected fission product heat generation rates in the solid product, shows the effect of various variables on heat transfer in the stored solid, and emphasizes certain items for further study. Also a vault cooling method is suggested for further study.

Several experiments have been performed at ORNL with light water solutions of uranyl nitrate (highly enriched in either U/sup 233/ or U/sup 235/) in an essentially bare sphere 27 inches in diameter. Results are presented of several calculations with elementary bare reactor theory and a discussion of the observed discrepancies between the calculated and experimental results. If the observed critical concentration is used in the calculations, the calculated effective multiplication constant is less than unity; thus a higher critical concentration would be predicted than is actually observed. ( auth)

Batch extraction scouting tests were performed to establish comparisons of distribution coefficients for uranium, thorium, fission product, and/or plutonium in systems involving several classes of organic phosphorus compounds (diluted in Amsco 125-82 and/or xylene) and aqueous nitrate or nitric acid solutions. Results have substantiated previous conclusions which suggested (1) that the branched secondary alkylphosphates and alkylphenylphosphonates would generally afford uranium separation factors (from thorium and fission products) superior to those obtainable by tributylphosphate (TBP, a normal alkylphosphate); and (2) that the phenylphosphonates would afford reasonably higher extractability of uranium. Preliminary data from irradiation tests with di-sec-butyl phenylphosphonate also support a previous conclusion (3) that the phenyl group affords greater radiation stability of the organo-phosphorus compounds. Since the compound di-sec-butyl phenylphosphonate (DSBPP) effectively combines the above advantages (1), (2), and (3), it has received especial attention as a potential practical competitor for TBP as a recovery process extractant. Results of preliminary counter-current extraction tests

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Fuel elements clad with Zr or containing Zr as a diluent can be recovered by a fluoride-volatility process. The first step consists of hydrofluorination of the elements in a bath of molten fluoride salts using an HF sparge. In this case the two salt systems considered were NaF-ZrF/sub 4/ and NaF- LiF. Materials evaluated at Battelle for possible use in the construction of this hydrofluorinator include Inconel, A'' Nickel, copper, silver, Monel, Hastelloy B, Hastelloy W, INOR-1, and INOR-8. The metals were exposed to molten fluoride salts through which HF was bubbled continuously. The data indicate that the NaF-LiF systems are much more corrosive than the NaF-ZrF/sub 4/ system. The systems are most corrosive when the alkali fluoride component is high. An elevation in temperature increases the corrosion significantly as does an increase in the HF flow rate. Hydrogen in the HF flow stream retards the corrosion of the sodiumzirconium salts significantly, but appears to have less effect on the sodium -lithium systems. The areas at the interface of the liquid and vapor phases were most seriously damaged under the exposure conditions usually used. However, appreciable reduction in attack was experienced when zirconium was actually hydrofluorinated. INOR-8 was the most …

The ''absolute'' containment of ions in the DCX magnetic mirror field resulting from the cylindrical symmetry of the field is discussed. The regions of confine;, ment in space and momentum are plotted for 300-kev deuterons. (auth)

>Tests on a coupled transmission line resonator show that a fairly accurate value of the coupling coefficient can be predicted by considering line geometry. The measured values of V/sub 02/ and I/sub 02/ appear to agree with predicted values. Development of a resonator for ORIC with two coupled transmission lines is under way. (W.L.H.)

The Argonne Fast Source Reactor is designed to operate at low power (nominally 1000 watts) to supply neutron fluxes, both fast and thermal, for laboratory experiments. It is built around a cylindrical core (with vertical axis) of solid, highly enriched uranium approximately 4 1/2 in. in diam. by 4/4 in. high. The blanket is of solid depleted uranium with a minimum thickness of eight in.; its outer form is cylindrical, 20 5/8 in. in diam. by 20 5/8 in. high. The reactor, contained in a shield of high density concrete of minimum thickness 4 1/2 ft, is freestanding on the floor of the reactor building. A graphite thermal column 4 x 4 x 6 ft is provided. All control and safety mechanisms are located in a pit beneath the reactor. The hazards associated with operation of the reactor have been analyzed. A number of potentially dangerous circumstances were studied to determine the probable severity of the resultant excursions. As an upper limit, a detailed study was made of the extreme case in which the reactor, overloaded by five kilograms of U/sup 235/, goes critical at the air cylinder speed of 18 in./min. It is estimated that the excursion would amount …

For several years, a major category of fuel element failures has been the side corrosion type, characterized by localized accelerated fuel element jacket corrosion. Since it has been demonstrated {sup 1} that misalignment of fuel elements in a process tube will produce flow patterns and accelerated corrosion, termed ``hot spots``, failure to align the fuel elements in process tubes is considered a contributing factor in the production of side corrosion failures. Preliminary testing of both self-supporting and ``bumper`` fuel elements is underway. Data on the self-supporting fuel elements have demonstrated that the bridge-rail projections have sufficient support strength, do not of themselves create a corrosion problem and in actuality probably eliminate any hot-spot areas. Although one tube of bumper fuel elements in KW Reactor {sup 3} has been discharged, data are not as yet available. Potentially, the most sever corrosion conditions exist during the summer months when reactor inlet temperatures are high. It is desirable then, provided bumper fuel elements limit hot- spot corrosion, to evaluate the bumper concept for large scale use possibly by the summer of 1960. To accomplish this, a demonstration loading of the bumper type fuel elements must be underway by about July, 1959. The purpose …

The temperature coefficient of reactivity at the plant operating temperature was --1.92 DELTA K/ DELTA T x 10/sup -//sup 4/, as obtained from the curve of the temperature coefficient plotted as a function of the temperature. (B.O.G.)

The angular relationships of warp to rib marks and warp to hot spots were measured for a large number of fuel elements. The data was analyzed and found to provide information regarding the formation of warp and hot spots.