Present thinking and planning appears to be developing from the following factors as concern the 13.8 KV temporary construction power limit. 1. It is understood that the present intent is to supply 100-N operating requirements from a single stub source in the 230 KV loop. 2. The original thoughts were to obtain construction power over a 13.8 KV line from 151-D substation. 3. Construction load requirements are now less than originally planned since steam has been substituted for electrical drive of primary loop pumps and 5500 hp motor tests are no longer necessary. 4. An extreme emergency backup source for the K plants has always been of concern, although minimized in recent planning. It is desirable to review the temporary construction line requirements from a future operating viewpoint to determine if the line could be useful to the operating plants after completion of construction. It is highly desirable to provide T.C. power source from K plants rather than 151-D and then leave the line and breakers in place for future maintenance assistance and as extreme emergency backup to K plants.

The status of various projects undertaken to assist other scientists at ANL is reviewed. (T.R.H.)

In order to investigate the rupture stability of uranium elements irradiated at power generations per unit length larger than those encountered in natural uranium fuel elements, three partial columns of 1.44% U{sup 235} enriched, internally-externally cooled, uranium elements were irradiated in PT-IP-1-A. After discharge and examination of the fuel elements 25 pieces (182.5 pounds) of this metal were available for special analysis of U{sup 235} burnout and plutonium content. The average exposure of these pieces was 2,187 {+-} 6% MWD/Ton. The purpose of this document is to summarize some calculations of buildup of plutonium isotopes and burnout of U{sup 235} in an attempt to correlate calculations with the results from experimental analysis.

8 2 5 6 7 6 1 1 B 3 7 3 7 reprocessing scheme which may be applicable to most power reactor fuels is described. The irradiated fuel is dissolved in an appropriate aqueous solution which is then calcined to a mixture of dry oxide. Fluorination of the mixture ultimately produces the hexafiuorides of U and Pu, which are volatilized. This process will be called the ADF (aqueous dissolution and fluorination) process. The fluorination rates of UF/sub 4/ and PuF/sub 4/ were investigated. Since oxides, as well as fluorides, are involved in this scheme, fluorination rate studies were conducted on UO/sub 2/. The thermal decomposition of PuF/sub 6/ was investigated at 300 deg C by several methods that showed the reaction products to be fluorine and PuF/ sub 4/. A brief study was made on the rate of fluorine consumption by Ni vessels at temperatures from 300 to 600 deg C. Development work has continued on thc fused flouride process for the recovery of enriched U from Zr-matrix fuel alloys. Studies of alternate methods of contacting in the fused salt fluorination step are under way. In one scheme, dropwise fluorination is carried out by spraying molten salt into a …

This report presents daily production data for the redox, purex and uranium trioxide operations at the Hanford Engineer Works during the period of January 1st, 1958 through March 31st, 1958. (JL)

Tungsten and graphite are unsuitable materials of construction for a UF/ sub 6/ inlet nozzle in a continuous DRUHM reactor. Preparation of feed was completed for an extended Fluorox test. Difficulties were experienced in the operation of a fluidized bed TbNT denitrator. Flame denitration of UNH and TbNO produced mixed oxides of 1 to 14 micron mean particle size. The chloride capacity of Dowex 21K was measured, and equilibria measurements of uranium sorption from sulfate solutions were continued. Siliceous deposits in the stripping column caused the termination of a Darex run with a prototype APPR fuel element. Hot runs were begun on the chemical dejacketing of irradiated PWR blanket pins. The addition of formaldehyde to neutralize "25" waste reduced the recovery of nitrate from the calciner off-gas from 76% to 23%. (For preceding report see CF-58-9-62.) (auth)

Side rupture data experienced to date for I and E fuel elements give rupture rate curves which do not differ significantly from those predicted on the basis of solid fuel elements experience.

The identification of phases and interpretation of phase transformation through microstructural analysis are still difficult even with metallographic standards as a guide. Macrostructures provide a better guide to determination of the origin of the quenched phase and the degree of quenching. Grain sizes are of some assistance in determining the cooling conditions employed. Twinning concentration and grain boundry shape are inconclusive characteristrcs to use in determining the origin of the quenched phase and the degree of quenching. Subgrains are more prevalent in beta-treated uranium. Hardness profiles provide highly irregular data and are of little assistance in interpretation of heat- treating conditions. Single and multiple gamma-beta-alpha cycles appear severe enough to cause the formation of small voids in thc air-cooled zones of end quench samples. (auth)

For the centrifugal pump under investigation, the static head at pump suction, in feet absolute, at cavitation inception was correlated for water and for 1500 F NaK on the basis of the differences of the vapor pressures of the two liquids. The difference between the vapor pressure of water and NaK, for the same conditions of pump speed and liquid flow, was added to the water-test cavitation inception value, and this estimate proved to be a good approximation to the experimental value found for cavitation inception with NaK. (auth)

BS>The corrosion of type 304 stainless steel in water containing boric acid at 600 deg F and 2000 psig and at a flowvelocity of seven feet per second was investigated as a function of boron concentration and of alkaliknity. At a boron concentration of 39 ppm and at a pH value of 10, controlled with lithium hydroxide, no trend of corrosive attack with time was found after the first tw0 days. At a concentration of 41 ppm boron wlth no pH control (pH = 6.9), a regression tyoe equation of the form Y= A + B log (t), was found to express the relationship between corrosive attack and exposure teme. A similiar relationship was found at 1600 ppm boron withthe pH value maintained at 10 using lithium hydroxide. No evidence of raicrostructural attack was observed on coupon specimens under any of the conditions tested. The effects of surface finish and of position were studied. The effect of orientation of the rolling directon of coupon specimens with respect to the direction of fluid flow was also investigsted. (auth)

A study was made of corrosion product transport and deposition on Zircaloy-2 and AISI 304 stainless steel in the presence and absence of ionizing radiation. Three 100-hour irradiation tests, using 2-Mev electrons from a Van de Graaff accelerator, and four 100-hour nonradiation tests were performed in 6O0 deg F pressurized water. Data from the seven runs and an additional exploratory run are presented. In addition, complete experimental procedures and descriptions of the apparatus are included. Control of pH was obtained by using H and OH form ion exchange resias for pH 7 and Li and OH form ion exchange resins for pH 10. The major conclusion to be drawn from the present work is that the deposition on Zircaloy-2 at pH 10 and 600 deg F is higher than on AISI 304 stainless steel at the same conditions. (auth)

None

A three-phase linear induction pump was designed, constructed, and installed in the Sodium Reactor Experiment to control moderator temperature. A maximum flow rate of 91 gpm was obtained at 760 deg F, at 12.4 psi, and an efficiency of 1.7%. (C.J.G.)

Russian developments in ultracenttifuges for U isotope separation are described. Progress on development of the short-bowl certrifuge is reported (For preceding period see ORO-202.) (T.R.R.)

A description is given of the equipment and method of operation of a distillation system to remove light water from heavy water moderator. The basic principles of the theory of distillation are reviewed. ( auth)

A three-dimensional few-group depletion code prograrnmed for the IBM-704 is presented, The code, called DRACO, is used in studying the neutron flux, the power level, and the related buildup and depletion of materials at different stages in a reactor lifetime. The three sections of the code are described, and the background documents are referenced. (auth)

Beginning in February, 1958, a sufficient quantity of seven-inch dip canned I & E depleted uranium fuel elements was prepared for irradiation to produce eleven kilograms of plutonium, containing at least twenty per cent of the Pu-240 isotope, as authorized by the Atomic Energy Commission. Subsequently, eighty-four columns in C reactor were partially charged with the finished depleted fuel under PT-IP-132-AC. To date, ten depleted ruptures have been sustained, after being irradiated six to eight months toward a planned accoumulated goal exposure of 210 MWD per column or a total irradiation time approximating 12--14 months. The mechanism and cause of these failures is being thoroughly investigated. This document summarizes the fabrication history, irradiation experience to date, rupture examinations, and an investigation of process conditions which may have contributed to the high incidence of ruptures.

This report represents a critical review of existing reactor cooling systems and discusses suggested supplementary-cooling system which might be employed in emergencies resulting from such natural hazards as earthquake, equipment failure, or personnel error. In addition the subject of building filtration is discussed. Maintenance of an uninterrupted flow of cooling water is of major concern to the safety of any HAPO reactor. For some time supplementary cooling systems which would be capable of removing heat output in the event of failure in the existing emergency backup systems have been under scrutiny. Loss of coolant may cause damaging power excursion (should this occur during operation) or will inevitably result in fuel melting and a subsequent release of fission products to the atmosphere, even if the reactor is shut down prior to the loss of coolant.

A description of the new Weldon Spring Feed Preparation and Sampling Plant for uranium concentrates is given. Prior to the startup of this plant the auger to be later installed was used in an evaluation program to test reliability for representative sampling and uniformity both within drums and between drums of various concentrates. Results of this program were used as a reference for the sampling plant evaluation which involved successive auger and mechanical sampling of a series of lots of several concentrates, followed by moisture determinations, uranium assays, and statistical analyses of the data. From the final results conclusions are drawn concerning the suitability of the mechanical sampling system for the concentrates examined.

Cold-worked specimens of iodide zirconium, Zircaloy-2, iodide titanium, and A-55 titanium were tested in a high-pressure recombiner loop in an attempt to duplicate anomalous results obtained in a prior recombiner loop. Hydrogen analyses and metallographic examinations were made on all specimens. The titanium materials and Zircaloy-2 picked up major amounts of hydrogen in the cell section. None of the materials tested showed appreciable hydrogen absorption in the recombiner section. Complete recrystallization occurred in all cell specimens while only Zircaloy-2, of the recombiner specimens, showed any degree of recrystallization. No explanation for this behavior can be given. A survnnary of the data obtained in previous recombiner loops is compared with the results of this loop. Conclusions were based on the results of three recombiner loops. Primarlly because of the hydrogen absorption data obtained in all three recombiner loops it is recommended that the zirconium and titunium materials tested not be used in environments similar to those encountered in high pressure recombiner loops. (auth)

chemical Research and Development Operation was contacted by R. Nilson of Irradiation Processing Department in order to obtain an analysis from a chemical viewpoint of the worst credible accident to a full E-N reactor loading. This analysis would be used to present to the ACRS directly or would be abstracted to prepare a supplementary report. The primary purpose of this report is to present an analysis of the chemical behavior of an E-N loading in the event of the sudden loss of coolant to the entire reactor, with, however, normal operation of the control devices such that the reactor is initially driven sub-critical. We are here concerned wit the effect of a rising temperature upon the E-N load, the rise in temperature being primarily caused by the release of fission product energy.

Further engineering and physics data to aid in constructing GCRE-1 were obtained in critical-assembly studies. Four major experiments were performed to investigate the effect on reactivity caused by changes in axial reflector materials, the effect on reactivity and the power perturbation caused by fast safety control-blade guides, the effect of changes in fuel-element material composition, and the effect of changes in fuel-element spacing designed to produce uniform radial power-generation rates. All studies were performed with a 4-in.-thick lead reflector at the core perimeter. Axial-reflector-material studies employed combirations of aluminum and steel reflectors. The reactivity worth of a 2 3/4-in.-thick steel reflector was +0.414% DELTA k/k compared with 0.175% DELTA k/k for a similar aluminum reflector. The perturbation in the flux distribution caused by the safety-blade guides was localized, and affected only the regions immediately adjacent to the guides. The combined reactivity worth of two guides was -0.281% DELTA k/k. Fuel-element material compositions were changed by separate additions of fuel and stainless steel. An increase in uranium loading from an average value of 303 to 404 g per element would provide, based on extrapolations from experimental data, a reactivity of about 4.5% DELTA k/k. An increase in steel from 1708 to …

This report provides a summary of the results of a process study, requested by the Atomic Energy Commission an the recovery of krypton and xenon from irradiated uranium at the Hanford Purex Plant. This request was prompted by original Commission forecasts of the expanded requirements for Krypton-85 for commercial phosphorescent signal lights and markers and for xenon isotopes of low neutron cross-section for use in liquid xenon scintillation counters, in connection with D.M.A., government and university-sponsored work. It was requested that both Hanford and Savannah River submit order of magnitude cost estimates for recovery facilities at the respective sites for three separate design cases. The cost information developed, along with market survey information obtained-through the A. D. Little Company and Department of Defense market surveys, would serve as the basis for scheduling of the Hanford and Savannah River participation in the Commission`s overall fission rare gas recovery program.

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operation research, inventions, visits, and personnel status are discussed. This report is for November 1958.