An anion exchange process was demonstrated for the recovery of Pu from waste produced in the reduction of Pu salts to the metal. Pu in a highiy salted 6M nitric acid solution, derived from the dissolution of slag and crucible waste, was separated from impurities by absorbing the Pu(IV) nitrate complex on the anion exchange resin and subsequentiy eluting with dilute nitric acid. A flowsheet for plant operation is presented. (auth)

Aqueous slurries of thorium oxide and thorium oxide containing urarium were investigated for their catalytic activity for the reaction of hydrogen and oxygen to form water. Pure thorium oxide. thorium-uranium oxide mixed crystals prepared by calcining coprecipitated oxalates, and thorium oxide with uranium oxide sorbed on the surface were used after calcining at 650, 800, and 1000 deg . The reaction rates were found to be first order with respect to hydrogen pressure and zero order with respect to oxygen pressure in all cases at temperatures from 230 to 300 deg - and total gas pressures from 100 to 2000 psi. For the pure thorium oxide an average activation energy of 41 kcal/mole and an average frequency factor of 4.6 x lO/sup 8/ moles/psi H/sub 2/hr-g of ThO/sub 2/ were found. Addition of uranium lowered both factors, the maximum effect giving a DELTA E/sub a/ of approximately 14 kcal with an A of approximately 10/sup -2/. Actual rates for all catalysts were within one order of magnitude when compared on a unit surface area basis. This compensation effect was explained on the basis of a two-site process, one site being related to the uranium concentration on the catalyst surface and …

The KEWB Facility is described, and an over-all technical evaluation is made of the hazards associated with changing from a spherical to a cylindrical core. The characteristics of the two systems, the operation and emergency procedures, a brief history of the program, a summary of the data obtained, and maximum accident analyses are given. The conclusion is that the core change does not represent an adverse change with respect to associated hazards. (T.R.H.)

Chemical-Metallurgical Processing. A direct-cycle pyrometallurgical fuel-processing plant is being constructed in conjunction with EBR-II. The gamma- irradiation testing of the 175-watt white fluorescent mercury vapor lamp was continued to an integrated exposure of 2 x 10/sup 9/ rad. Irradiation tests of Shell APL grease were completed, and the estimated useful life of this grease in the Air and Argon Cells is 2 and 3 years, respectively. Tests of the three - types of d-c motors used in the operating manipulator of the Argon Cell indicate an expected useful cell operating life of 2 to 3 years. Continued irradiation tests of mineralinsulated cable show no catastrophic breakdown of the electric insulation even after an accumulated gamma dose of 8.4 x 10/sup 9/ rad. The scheme currently under consideration for processing melt-refining residues involves a reduction of skull oxides by a solution of Mg in liquid Cd. Molten salt fluxes had variable effects on the rate of oxide reduction in dilute Mg systems. Work was continued on development of processes for EBR-II blanket materials. A large-scale metal-distillation unit to demonstrate metal distillation at rates up to 100 kg/hour is under construction. Two medium carbon steel thermal-convection loops were built and operated to …

None

The diffusion of Kr/sup 85/ in two UO/sub 2/ powders was studied by performing a series of post-irradiation anneals on the powders. The emanation data were analyzed by considering the effect of sintering as well as the effect of a distribution of particle sizes within the sample. Measurements were made at 900 to 1500 deg C. The time at a temperature was between 8 and 24 hours. The diffusion coefficients for Kr/sup 85/ in the two powders are represented by the equations: D = 2.65 x 10/sup -4/ exp - 65,500/RT for UO/sub 2/ prepared from crushed UO/sub 2/ pellets and, for a chemically prepared UO/sub 2/ powder, D = 4.9 x 10/sup -4/ exp - 73,800/RT. (auth)

A drip casting process initiated to obtain zirconium castings uncontaminated by the melting process and to remove volatile impurities from the zirconium feed rod is described. A feed rod of zirconium is held above the mold, and the bottom of the rod is melted rapidly off into a mold to produce the casting. The melting process is carried out under high vacuum, so that very little atmospheric contamination can result, and some removal of volatile impurities is possible. Since no crucible is used to contain the molten metal, no contamination can result from this source. (auth)

EURIPUS-3 calculates the one-dimensional spatial density of neutrons slowing-down past a given energy in an infinite homogeneous medium consisting of hydrogen and one other isotope with arbitrary mass and energydependent differential-elastic and absorption cross sections. DAEDALUS determines the corresponding spatial distribution of angular integrals of an arbitrary function times the vector flux density. Spatial moments of all density functions are furnished directly. Although scattering angles are calculated by Monte Carlo, the spatial distributions and, in DAEDALUS, the energy distribution are obtained partly from an analytic treatment which, besides saving tinne, enables the output to be in the form of actual density functions at specified planes and energies, rather than histograms covering finite intervals. At certain steps in the computation of both the spatial and energy distributions, part of the analytic treatment is replaced by Monte Carlo in order either to maximize efficiency and/ or to avoid round-off error. The neutron source may be monoenergetic with either isotropic or monodirectional angular distributions, or else the source may be that from deuterons bombarding deuterons. The volume displaced by a cylindrical tube from an accelerator to the source can be accounted for in the neutron first flight but not thereafter. (auth)

>An experimental measurement of the effective delayed neutron fraction ( beta -bar) was made for a clean critical assembly by determining the asymptotic period associated with introduction of a known amount of reactivity. The "known amount" of reactivity was obtained by replacing, uniformly throughout the reactor, a small quantity of U/sup 235/ with an alloy of B/sup 10/ and Hf designed to match the absorption properties of U/sup 235/. The replacement was thus equivalent to a uniform reduction in nu , the number of neutrons emitted per fission from the fuel. Such a reduction introduces a reactivity change equal exactly to delta nu / nu /sub 0/. Two analyses of the experiment were made using different high energy cross sections in conjunction with four group, two dimensional diffusion theory. The measured value of beta lay between the results of these computations, the error spread (an average rms error of plus or minus 5.2%) being too great to permit any conclusion regarding the significance of the comparison. (auth)

None

This report on fuel element handling presents in some detail the current status of an engineering study which has been underway for some time, and which is continuing. The study was undertaken to determine if it is feasible, and if it is practicable, to revise the method and equipment used for fuel element handling with existing charging machines.

An improved method for precipitating manganese dioxide was demonstrated that significantly increases the allowable feed rate of the Purex head-end centrifuge. The effects of several process variables are discussed. (auth)

None

Material balance flowsheets are presented for the dissolution of UO/sub 2/, UO/sub 2/-ThO/sub 2/, and U-Mo fuels clad in stainless steel or zirconium by the Sulfex, Darex, and Zinflex process. The mechanics of the three processes are discussed. Basic assumptions upon which the flowsheets are based are contained. (auth)

The advantages and shortcomings of the codes currently in use at Argonne (RE-13 and RE-129) are discussed. A new method of solution, which has increased accuracy, stability for exceptionally large integration intervals, and a procedure for automatically changing the integration interval as the nature of the problem changes, is developed. (auth)

Preliminary report regarding an experimental boiling reactor facility (BORAX V): "The primary objectives of the proposed BORAX V program are to test nuclear superheating concepts, and to advance the art of boiling water reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities" (p. 9). This report discusses the completion of the reactor design and start of construction.

Flowsheets are presented for the dissolution of Borax IV reactor fuel (6.35% UO/sub 2/)--ThO/sub 2/ pellets encased in 1% nickel--aluminum alloy and bonded with lead). In the preferred method the alurninum is dissolved first in boiling 2 M NaOH--1.78 M NaNO/sub 3/, with a urarium loss of approximately 0.07%. The lead and nickel are then dissolved in boiling 1.5 M HNO/sub 3/, with uranium losses of <0.2%. The oxide core is dissolved in two successive digestions with boiling 13 M HNO/sub 3/--0.04 M NaF--0.1 M Al(NO/sub 3/)) to produce a solution 0.6 M in thoriunn and 0.04 M in uranium. Dissolution of only the aluminum in sodium hydroxide or sodium hydroxide--sodium nitrate solution prior to core dissolution is unattractive since the rate of core dissolution is lowered greatly if lead is present and a product solution containing greater than 0.2 M thorium is unattainable owing to the low solubillty of lead nitrate in nitric acid-- thorium nitrate solutions. Simultaneous dissolution of aluminum and lead in mercury-catalyzed nitric acid appears costly since a mercury concentration of at leset 0.5 M is required to ensure an adequate dissolution rate. The solubility of lead nitrate in nitric acid and nitric acid--thorium nitrate solutions …

None

In connection with design of rear face fittings for the plant-expansion study currently being conducted we have been asked to determine if rear face pressurization is required for safety reasons and if so, how much. Pressurization of the rear face piping would be used to provide sufficient reverse flow to prevent process tube burnout in the event of complete loss of coolant supply to a single tube by virtus of a front connector failure. Consideration of the effectiveness of rear face pressurization, however, requires a more general look at the problem of single tube meltdown than that provided by considering front fitting failure alone.

Tests were performed on aluminum alloys to evaluate their behavior in high-temperature, high-pressure. watercooled and -moderated nuclear reactor enviromnents. Test equipment, sample preparation. and test procedures are discussed. Aluminum nickel alloys were found resistant to disintegration for periods up to 60 days in dynamic water at 600 ction prod- F. The corrosion rates of the aluminum alloys M-388 and X-2219 at 600 ction prod- F were found to be too high to merit consideration for cladding materials. The influence of pH. gas content, and velocity of the water on the corrosion of the above alloys was evaluated. Hydrogen addition at startup appeared to increase the degree of corrosion attack on the M-388 alloy. Irradiation tests on aluminum-nickel alloys revealed that the corrosion rate increased with distance from core. In-reactor samples of M-388 exhibited less corrosion attack than out-of-reactor samples. Boiling water corrosion tests were performed on M-388 for 1612 hr at 422 ction prod- F with an average heat flux of 25,000 Btu per hrft/sup 2/. The over- all corrosion rate was 2.9 mil per yr. It is concluded that the corrosion rate of M-388 is acceptable for the specified test conditions: (1) absence of radiation: (2) demineralized water at …

The equipment described is an a-c operated portable scintillation spectrometer consisting of a preamplifier, a linear pulse amplifier, a single- channel pulse-height analyzer, a linear count-rate meter, a scaler, and a highvoltage power supply. The operation and performance of the circuits are discussed. The instrument is accurate and reliable, light in weight, and consumes low power. (auth)

ABS>A recording potentiometer was modified to provide a versatile instrument that can be applied to a variety of problems without time-consuming changes. Ranges may be selected in six spans, from 0.5 to 100 mv. No adjustments of amplifier gain are required when switching from one range to another. Zero suppression is continuously variable over a plus or minus 100 mv range by means of coarse and vernier controls. Cold junction compensation is provided for four standard thermocouples, and chart speeds from 1/2 to 16 im- ./ hr may be selected at will. (auth)

Canning bath Al-Si, supplied from offsite vendors and reclaimed lathe turnings in the 313 building, is used in the production of I & E fuel elements. A study was made of the effect of increasing the Ni content to over 0.5% in the canning baths, in order that all of the X-8001 scrap could be reclaimed. Effect on bond quality, weld integrity, and canning bath operation was studied. Based on adverse weld quality, slight loss in reactivity, and potential for furnace channel plugging, it is recommended that the present Ni specification of 0.5% maximum remain unchanged.

Experiments have been conducted in the Hydraulics Laboratory, at the request of IPD`s Mechanical Development-A Operation, to determine the energy losses of various enlarged outlet fitting combinations. These experiments were conducted an steady state runs and allow the determination of the normal operating point (flow rate) of a reactor process channel under selected conditions of front header pressure and fuel charge. No attempt is made to make a mechanical or economic evaluation of the particular fitting combinations, although observations were noted which might bear on this evaluation. It is very important for the reader to bear in mind that changing outlet fittings will definitely affect the reactor tube power limits and outlet vater temperature limits. The size of the outlet fittings largely determines the present outlet temperature limits of the old reactors. The flow characteristics of these present fittings cause some degree of pressurization to suppress boiling on the fuel charge and also cause dual Panellit trip protection for certain flow changes and for power surges. Enlargement of the outlet fittings may actually reduce the allowable outlet coolant temperature limits. Since these effects cannot be determined on the apparatus used in these experiments, a complete discussion of this point is …