The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)

Basic properties of several types and grades of commercial UO/sub 2/ have been determined. Compacting characteristics of these powders were evaluated with the objective of obtaining a minimum cold-pressed density of 70% of theoretical prior to gas-pressure bonding. Fused and special dense grades of UO/ sub 2/ powders were compacted to a density of 85% of theoretical by use of a 50- tsi compacting pressure. Cold-pressed compacts were simultaneously clad and densified to a maximum he UO/sub 2/ powders that were capable of being pressed to the hightest coldpressed density exhibited the least amount or densification during gus-pressure bonding. A small stainless essure bonded at 2100 F for 3 hr at 10,000 psi. Examination of this assembly indicates that it is feasible to prepare fuel ep operation by use of the gas-pressure-bonding process. (auth)

Bench-scale studies were made on use of di(2ethylhexyl)-phosphoric acid in an organic diluent (Dapex process) for solvent extraction recovery of vanadium from acid leach liquors. Vanadium may be stripped from the solvent by either acidic or alkaline reagents, the former having been studied in considerably greater detail. A process for single-cycle recovery and separation of uranium and vanadium from sulfate leach liquors was shown to be attractive both from the standpoint of operation and chemical costs. Process schemes for recovery of vanadium from uranium-barren liquors are also described. On the basis of the encouraging laboratory results, pilot scale tests for specific applications are recommended. (auth)

Uranium and plutonium were recovered by liquid-liquid extraction from simulated sulfuric acid stainless steel decladding solution with several extractants. Consecutive extraction of U(IV) and Pu(III) or (IV) by 0.1 to 0 3 M primary amine in hydrocarbon-- alcohol diluent appeared promising, and chemical flowsheets were demonstrated in laboratoryscale continuous countercurrent extraction. Extraction of U(VI) with a dialkylphosphoric acid appeared promising when plutonium recovery is not needed. Recovery is also chemically feasible by extraction of U(VI) and Pu(IV) with an N-benzyl secondary alkyl amine or a trialkylphosphine oxide. The amine extracts are stripped with nitric acid, giving a sulfate-nitrate product solution. The organophosphorus extractants permit elimination of the sulfate but require sodium carbonate for stripping. (auth)

Eurochemic reports were reviewed by BNL personnel concerning aqueous reprocessing, and several comments are offered for process improvements. Also, pertinent BNL documents related to waste and head end processes and corrosion are listed. (J.R.D.)

None

Some calculations have been made to estimate the characteristics of a neutron-burst type fast reactor similar to Godiva but made up of relatively small component parts--the so-called "layered assembly." One spherical and three cylindrical assemblies have been considered. Critical masses, assuming 5% voids, range from 58 to 65 kg of 93.4% enriched U/sup 235/. For a reactivity addition of 0.33 dollars above prompt criticals bursts between 2 x 10/sup 17/ and 6.7 x 10/ sup 17/ fissions were computed with accompanying temperature rises varying from 514 to 1600 deg C. The burst width at half-maximum was about 12 microseconds. To obtain an idea of the possibilities of stress reduction which might be achieved by layerings an assembly made of small rings was considered. While the critical masses obtained here are believed to be fairly accurates the predictions concerning mechanical energy generated, total fissions, and burst width may be subject to sizeable error due to the many simplitications required to allow hand computations. Neverthelesss considerable improvement in safety and burst-size is indicated by the use of a "layered assembly" instead of an assembly composed of relatively thick parts. (auth)

The Army Gas Cooled Reactor System Program includes water moderated heterogeneous reactor (Gas Cooled Reactor Experiment I), a graphite moderated homogeneous reactor (Gas Cooled Reactor Experiment II), a mobile gas cooled reactor (ML-1), and the co ordination of thc Gas Turbine Test Facility. The progress of each project, the associated tests and data evaluation, the applicabie design criteria, and the fabrication of reactor components are briefly summarized. (For preceding period see IDO-28538.) (W.D.M.)

Diffusion coefficients for hydrogen in beta zirconium were determined from permeation rates in the range 650 to 850 deg C. Both the steady-state method, which is dependent upon the hydrogen concentration, and the time-lag method, which is independent of hydrogen concentration, were employed to obtain diffusion data. Zirconium disks, 0.03 to 0.1 cm thick and varying in hydrogen concentration from 9 to 33 at.%, were used to measure permeation rates. The diffusion coefficients determined by the steady-state and time-lag methods on samples of differing thickness were in agreement. It was concluded that the permeation process was diffusion controlled. The diffusion coefficients were found to be independent of concentration and can be expressed by D = 6.14 x 10/ sup 4/ exp (--45,900/RT). (auth)

Annual report of the Liquid Metal Fuel Reactor Experiment describing progress during calendar year 1958 and as well as an evaluation of progress and plans for future work.

Abstract: "The cross section for the production of 2.61-Mev gamma rays from the Pb-208(n, n')Pb-208* reaction has been measured for five different angles from 50 to 130 degrees. The angular distribution appears to be isotropic within the precision of the experiments, and leads to an integrated cross section of 25.5 +/- 4.6 millibarns."

The following document describes the usage and purpose of the neutron diffusion theory reactor program 9-Zoom, a memory-contained program that takes advantage of 709 features such as, for example, preferential order of multiply by zero, and for small problems approaches input-output limitations with excellent convergence properties.

The in-line monitoring instruments installed in the Redox plant in the late 1950s consisted of six gamma monitors, a single channel gamma spectrometer, an alpha monitor, and a neutron counter. A gamma monitor on the steam condensate outlet line and a uranium monitor on the 2DFS stream were to be added at a later date. The first section of this information manual describes the in-line gamma monitors and gives operating instructions for them. The second section covers the alpha monitor, and the third section the neutron counter. Sections on the uranium monitor and the steam header gamma monitor were to be added at a later date.

A major problem in the development of a pressurized water reactor coolant system for the NPS is the rupture performance of the fuel elements. As water temperatures are increased to 300 C, uranium corrosion rates increase rapidly. Swelling of the uranium fuel by corrosion could cause the process tube to burst or reduce the tube cooling water flow below acceptable limits. The desirability of slow cooling of the water to avoid thermal shocks to the reactor piping after a rupture is detected further complicates discharge and decontamination problems as fuel will continue to corrode with attendant fuel element damage during the cooling period. Coextruded uranium-zircaloy-2 clad fuel elements are scheduled for use in the NPR. The rupture behavior of this type fuel material after heat treatment was studied in ELMO-4, an out-of-reactor recirculating water loop. Several types of initial defects were studied. Fuel materials with five different heat treatment histories and with different types of defects were tested to determine their rupture behaviors. The five conditions were (1) as-extruded material as received from Nuclear Metals, Inc., (2) beta heat-treated and water quenched, (3) beta heat-treated and air-cooled, (4) beta heat-treated, isothermally treated at 600 C and air-cooled, and (5) beta …

It is the objective of this study to investigate preliminary expansion program requirements for process water, as supplied by 190 Building equipment; from the point of view of practical pumping, flywheel and pump suction head requirements. These requirements are to be determined at this time in such a form and accuracy as to be useful in refined estimating for budget study purposes. In order to obtain the objectives of this study at this time it has been decided to consider five different conditions of process water flow to a reactor. These conditions are flow to the reactor under summer conditions of operation in gallons per minute with a corresponding top of riser pressure in pounds per square inch: 85,000 gal/min, 580 psi; 100,000 gal/min, 480 psi; 130,000 gal/min, 280 psi; 150,000 gal/min, 280 psi; and 150,000 gal/min, 150 psi.

This document details activities of the Fuels Preparation Department during the month of April 1959. (FI)

Hot press canning of internally and externally cooled fuel elements has been developed to a point where the process is feasible. Complete specifications have been written for the process covering component, dies and punches, furnace construction, nickel plating, component cleaning, component assembly, sizing, hot pressing and inspection. Drawings covering each major item are included.

In early July 1958, it appeared that Columbia River temperatures at HAPO would be near 24--50{degree}C by the end of August. River temperatures were averaging 40 to 50{degree}C above 1957 figures and were 3{degree} to 4{degree} above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent. The large storage of cold water behind Grand Coulee Dam, normally untapped, was a source of possible relief. A plan for use was proposed for the peak high temperature period and agreed to by the Bureau of Reclamation.

The temperatures of the Columbia River was reduced 1 to 30 Centigrade with beneficial effects at HAPO. It is reasonable to expect that future benefits may be possible. It is desirable that the temperature of the river be controlled each year to the maximum extent possible. Instrumentation improvements requested to effect optimum savings. Records of river temperatures and flows should continue to be maintained by IPD as a necessary part of temperature optimization. Where possible, the coincident use of the river cooling technique should be made for the benefit of anadromous fish.

Introduction: Seldom does the actual arrangement of control elements in a nuclear reactor confers to the ideal and convenient mathematical array. In order to achieve shim control. safety and regulation, it is desirable to design with rods of different sizes and materials. With given fuel element arrangement, typically in square or hexagonal lattice spacing, there will be rods located at different distances form the center of the core and from each other. As the reactor operates, absorbers will be withdrawn, leaving further asymmetries in the location of those remaining. It is the purpose of this report to develop in detail a two-group diffusion theory with as complete generality as possible. The method is as yet restricted to the unreflected core, or to the reflected core by use of reflector savings and bare equivalent geometries.

This report describes the design and testing of a 1.17-Bev/c separated K{sup -} beam designed in the fall of 1958 in connection with a 15-in. hydrogen bubble chamber experiment.

The spatial distribution of thermal neutrons in a reactor lattices cell is of fundamental importance for many reasons. First of all, this information allows the determination of the relative absorption rates in each component of the cell which in turn enables a determination of the thermal utilization, and the diffusion length of the lattice. In addition, the observed spatial distributions of thermal neutrons in the lattice cell is of major interest in testing various approximations to the solution of the transport equation such as the P1 and P3 solutions.

Nuclear safety in the dissolution of irradiated 0.95 U235 enriched fuel has been investigated. In particular, critical conditions of fuel of this enrichment in a 52-inch diameter dissolver crib were studied. Since a crib this size is not safe by geometry, dissolution procedures as well as maximum safe batch sizes were analyzed. Uranium-water lattices are normally studied in systems in which rods are uniformly dispersed in the moderator. The results of such a study for 1.34-inch diameter solid rods as well as I. and E. fuel having a 1.37-inch O.D. by a 0.48-inch I.D. have already reported.

The purpose of this procedure is to present a detailed, chronological presentation of the preliminary decontamination and post decontamination steps necessary to fulfill the requirements of the Production Test Authorization IP-239-N. The procedure attempts to present the required operation in sufficient detail to successfully accomplish the intent of the test. Certain procedures involve operations of a standard nature and have not been elaborated upon to any great extent, as it is expected that the reactor operations and radiation monitoring personnel will implement these instructions according to standard operating procedures.