The following document describes the usage of the LRL 9-ZOOM code, a neutron diffusion theory reactor code for the IBM 709. The code has been modified to solve configuration of a series of stacked cylindrical disks, designating a new geometry case.

None

Report containing an analysis of the Aerial Radiological Monitoring Survey (ARMS) program detailing the feasibility of improvements. Includes suggestions for revisions of the existing ARMS system.

An idealized model is postulated embodying the essential features of industrial caustic-chlorine cells with horizontal flowing-mercury cathodes. This model is examined in detail, and relations expressing the local anode potential, cathode potential, and ohmic potential drop in the electrolyte in terms of local current density and other parameters are established. These relations are combined to give a system of equations relating current density at any location along the cell to applied total potential and to operating conditions in the cell upstream of the point in question. Numerical solutions of these equations for several cases of cell operating conditions are carried out on a digital computing machine. The effects of changes in operating parameters upon average current density, individual electrode potentials, and current distribution are evaluated. (auth)

An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)

The Army Package Power Reactor (APPR-1) control system was analyzed to find its response to a startup accident. The system was found to be adequate, at 0.086%/sec rod withdrawal, for operation of at least twice the rated power. (C.J.G.)

Report documenting the progress of the Army Gas-Cooled Reactor Systems Program to develop a mobile nuclear power plant for military field operation.

Report documenting the progress of the Army Gas-Cooled Reactor Systems Program to develop a mobile nuclear power plant for military field operation.

Several recent modifications of the ART program for the study of the behavior of a nuclear reactor during various thermal transients are described. The program requires a 32,000-word IBM-704 computer with six tape units. The major modifications are provision for a slip flow model and for void reactivity contribution. (auth)

A solid-state bonding technique involving the use of gas pressure at elevated temperatures was utilized for the self-bonding of molybdenum and niobium. Bonding conditions and surface preparation as a function of the integrity of the bond achieved were evaluated for each material. Optimum self-bonding of niobium was achieved by bonding parameters of 2100 to 2300 deg F at 10,000 psi for 3 hr with surfaces which had been prepared by etching in a nitrichydrofluoric acid solution prior to bonding. The process as developed was used to prepare niobium- clad flat-plate- and rod-type fuel elements and flat-plate subassemblies. Niobium tubing was also fabricated by this technique. (Molyb denum self-bonding was most readily achieved by gaspressure bonding at temperatures of 2300 to 2600 deg F at 10,000 psi for periods of 3 hr. With these bonding conditions a number of different surface preparations were satisfactory. Directional ductility of the molybdenum was encountered after bonding and methods to eliminate this were evaluated. Cross rolling with respect to the original rolling direction was shown to improve the ductility of molybdenum-clad specimens. (auth)

N/A

The chemical decontamination of the ANL-2 Highpressure Water Loop at the MTR is described. A mixture of sodium hydroxide and potassium permanganate was first applied to oxidize partially and loosen the magnetite corrosion film on the loop surfaces. Next, the system was treated with a solution of Versenol 120 to remove and suspend the loosened oxide film. The third chemical treatment was a mixture of sulfamic acid, HEDTA, hydrazine, and Rodine 80. After the first two chemical treatments a brown residue remained, which was presumed to be manganese dioxide. The third solution completely removed all traces of residue from the first two treatments. After a delay of one MTR fuel cycle (three weeks) due to mechanical difficulties, the loop was operated for approximately 300 hours at 420 deg F and 380 psig with excess hydrogen to passivate the loop surfaces. Approximately 70% of the corrosion film was removed from the loop surfaces as a result of the chemical decontamination. (auth)

This report for June 1960, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.

Experiments showed that 30% tributyl phosphate will not extract acid- deficient species of uranyl nitrate. Flooding throughputs for the Mark I stacked- clone contactor ranged from 600 cc/min organic at zero aqueous to 60 cc/min organic at 1950 cc aqueous. A large electronic vibrator of 5000-lb thrust was found somewhat inferior to pneumatic vibration for compacting oxide fuels into stainless-steel tubes. Tests were started on the use of fixed-bed CuO oxidizers for removing hydrogen contamination from helium gas streams. None of the variables studied within this period effected an increase in particle size in the denitration of thorium nitrate to produce ThO/sub 2/. The rate of uranyl sulfate loading on nitrate equilibrated Dowex 21K appears to be essentially independent of the loading solution sulfate concentration. The nitric acid concentrations corresponding to maximum UO/sub 2/-ThO/sub 2/ pellet dissolution rate were 15.5 M for Thorex solution and 13 M for the adjusted Darex solution. Two additional Semicontinuous Sulfex declad and Thorex core dissolutions of prototype Consolidated Edison fuel assemblies were made to complete the series of runs. The effective area of cylindrical UO/sub 2/ pellets dissolving in nitric acid was estimated from experimental rate measurements as a function of the fraction dissolved. …

An economical process was successfully demonstrated in bench-scale continuous equipment for stripping U from amines with ammonium carbonate solution. A continuous countercurrent mixer-settler extraction system was set up for further testing of the process for recovery of Te, Np, and U by tertiary amine extraction from UF/sub 6/ transfer cylinder was solutions. The effect of Purex aqueous feed adjustment procedures on Pu extraction by 1 M di-secbutyl phenylphosphonate (DSBPP) was studied. Work was continued on plutonium(IV) nitrate extraction with TBP and phenylphosphonate esters. The response of Ru/sup 106/ extraction to variations in the treatment of TBP-Amsco 125-82 solvent was tested. Two solvents have shown ability to extract cesium. (For preceding period see CF-80-3-136.) (W.L.H.)

Abstract: "Methods developed by [the Systems Evaluation Department] to analyze circular targets are extended to handle a particular family of casualty curves."

The Hanford water treatment plants were operated for a number of years with alum as a coagulant and activated silica as a coagulant and a polyacrylamide as a filter conditioning agent. (other polyelectrolytes may be used.) Sufficient time has elapsed to make a comparison of the two methods of operation valid and useful. Such a comparison is the purpose of this paper.

A calculation is made of the interaction of a beam of particles in an accelerator with the radio-frequency cavity that provides the accelerating mechanism of the machine. A Hamiltonian for synchrotron motion is employed that makes possible the simultaneous solution of Maxwell's equations and the Vlasov equation, so that a self-consistent distribution of particles in synchrotron phase space is determined. The effective voltage on the cavity due to the beam of charged particles is of the order of magnitude of the product of the total circulating current in the accelerator and the shunt impedance of the rf cavity. It has the net effect of producing a total voltage on the cavity which is both less than the applied voltage, and shifted in phase with respect to it. The increase in the stable phase angle required so the particles will remain in phase with the accelerating radio frequency is calculated. The decrease in total voltage and increase in stable phase angle result in a decrease in stable phase space available for acceleration, and convenient expressions are given for these quantities in terms of parameters of the accelerator. It is shown that the consequences of the induced voltage may be alleviated by …

The nuclear characteristics of the CETR are described. Core operating lifetime, control-rod worth, and power density distribution are discussed in relation to maximizing the core operating life. Other objectives of nuclear design are to minimize the power-density variation and to assure control of the reactor.

Critical experiments with pin-type oxide fuel elements for the CETR are described. These experiments were conducted to provide data on the infinite medium properties of the lattice inside the canned elements in the various CETR loading zones and to obtain specific information on the fullsize CETR core. Measurements are described and results are interpreted. (J.R.D.)

This report describes the critical experiments with pin-type oxide fuel elements for the Consolidated Edison Thorium Reactor (CETR). This report also describes the measurements and gives a brief interpretation of the results in some cases

The design functions and fabrication details for internal components of the CETR are presented and pertinent analytical stress studies are summarized. Functions of the internal components include proper orientation and support for the fuel elements, proper distribution of primary coolant within the reactor vessel, and the establishment of guide channels for the control rods. (J.R.D.)

The design functions and fabrication details for internal components of the CETR are presented and pertinent analytical stress studies are summarized. Functions of the internal components include proper orientation and support for the fuel elements, proper distribution of primary coolant within the reactor vessel, and the establishment of guide channels for the control rods.

The series of experiments described in this paper was performed to determine reactivity values of various patterns of fully inserted cruciform control blades. The experiments yielded the worth of fully inserted blade patterns in terms of their holddown characteristics.