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Report discussing the mapping of the Monument Valley area in Arizona in order to locate uranium-vanadium deposits, which were localized in strata of the Shinarump member of the Chinle formation of Triassic age that fill ancient scour channels.

A study to define the effectiveness limits of sulfamic acid and to discover other better stabilizers for Pu(III) is described. Ascorbic and isoascorbic acids, used in conjunction with sulfamic acid reduced Pu(IV) to stable Pu(III) in nitric acid. Aminoguanidine sulfate also retarded the oxidation of Pu(III) but did not reduce Pu(IV). (J.R.D.)

Report discussing the progress made on a study of electric propulsion for space travel during the period from July 1, 1958 to December 31, 1958 at the Lawrence Radiation Laboratory in Livermore, California.

The following report discusses technical problems with ion cyclotron waves, and to suggest a possible application to ion magnetron work.

From introduction: "This report deals with a part of the research and development studies which preceded the manufacture of fuel elements for the Gas Cooled Reactor Experiment (GCRE)." The studies evaluate the effects of varying the type and size of UO2 particles, stainless steel matrix powders, blending procedures, compacting pressures, sintering times, temperatures and atmospheres, roll-cladding temperatures and reduction rates, total cold reduction, and heat-treating times and temperatures has been made for UO2 stainless steel dispersion fuel elements."

Dogs, pigs, rabbits, guinea pigs, and mice were exposed to nuclear detonations in two open underground partitioned shelters. The shelters were of similar construction, and each was exposed to separate detonations. Each inner chamber filled through its own orifice; thus four separate pressure environments were obtained. An aerodynamic mound was placed over the escape hatch of each structure to determine its effect on the pressure-curve shape inside the chamber. In one test a sieve plate bolted across the top of the mound was evaluated. Wind protective baffles of solid plate and of heavy wire screen were installed in the shelters to compare primary and tertiary blast effects on dogs. The shelters also contained static and dynamic pressure gages, radiation detectors, telemetering devices, and, in one test, air-temperature measuring instruments, dust-collecting trays, and eight pigs for the biological assessment of thermal effects. One dog was severely injured from tertiary blast effects associated with a maximal dynamic pressure (Q) of 10.5 psi, and one was undamaged with a maximal Q of 2 psi. Primary blast effects resulting from peak overpressures of 30.3, 25.5, 9.5, and 4.1 psi were minimal. The mortality was 19% of the mice exposed to a peak pressure of …

The reaction between formaldehyde and nitric acid, in which the acid is destroyed with the production of predominantly gaseous products, has been recognized as of great potential value in the processing of radioactive fuels, particularly during waste treatment. Laboratory studies of the reaction at Harwell and at Hanford have shown that a major fraction of the nitric acid can be readily removed from an acidic solution containing nitrates by the addition of formaldehyde. The process possesses the advantages of low chemical cost; recoverability of nitric acid; and, in the case of waste treatment, the production of a solution relatively low in inert salt concentration suitable for fission product recovery or ultimate disposal. The primary purpose of the study was to confirm and extend existing information on the formaldehyde reaction to the destruction of nitric acid in Purex type waste (1WW) through operation of pilot plant scale apparatus. Operational behavior, formaldehyde utilization efficiency, and safety considerations were particular subjects of study. In addition, destruction of nitric acid in a Darex-type dissolver solution was investigated.

This report for January 1959, 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.

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Changes in pile operating parameters, such as fuel exposure, paver, and reactivity gain upon water loss affect the potential reactivity of the operating, pile and, therefore, the amount of enrichment required to operate under the desired conditions. Similarly a change in ECT (through enrichment location) changes the amount of enrichment.needed to maintain a constant reactivity. Many of the factors which influence the operating reactivity have quite different effects upon the reactivity of the dry pile with the ball 3X system inserted; therefore, changes in operating parameters may cause sizeable changes in requirements for supplementary control to maintain sub-criticality in the dry pile. It is the purpose of this document to summarize these trends in disaster control requirements for the K piles. The trends are presented as graphs of supplementary control requirements against a single varying parameter. All other parameters are held fixed as specified on each graph. The fuel element (central orifice zone) for all examples is the K-2 slug. In general, the slopes of these curves are only slightly dependent upon the specific values assigned the fixed parameters, although the magnitude of the control requirements may be quite dependent on these values. Unless otherwise noted any dependence of the …

Following-the K-pile start-ups in early 1955, a program of planned power raises was begun. The operating level had reached 1700--2000 MW by late 1955, and a severe operational control problem became apparent; the power distribution in the reactor was difficult to control and appeared inherently unstable. A study of available data led to the initiation of a production test so that a more detailed study of the phenomena could be made. This report describes the measures taken which led to an improvement in the operating characteristics of the K-piles; the current status and future outlook are also discussed in a general way.

It was found from graphite irradiations that the changes can be described by a growth component and a contraction component, and that the contraction is a linear function of exposure, while the growth is a function of both exposure and temperature which saturated at high exposures, say 3000 MWD/a ton. A graph is included. Contraction rate is determined for KW, KE, and H piles.

The effects of ``water mixing`` pieces on the axial flux distributions and reactivities of the K Reactors are discussed in this report. Two mixing piece configurations in the K piles are compared to provide a basis for determining the optimum configuration with respect to rupture control, reactivity cost and temperature cycling effects.

This document details activities of the irradiation processing department during the month of January 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.

This document details activities of the Hanford Operations Office during the month of February 1959. (FI)

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

Two Zircaloy-2 jacketed seven-rod cluster elements will be irradiated in the 3674 KE front-to-rear test hole to an exposure of 1000 MWD/T and two elements will be irradiated in the 3674 KW front-to-rear test hole to an exposure of 2000 MWD/T. After irradiation, the elements will be sent to the ETR where they will be ruptured during reactor operation to determine the failure characteristics of co-extruded Zircaloy-2 jacketed cluster elements.

This report discusses the events from a chemical standpoint following a total loss of coolant disaster will not be altered in the melting reactor by the introduction of N metal. The interdiffusion of uranium and aluminum will be the dominating reaction, causing the blockage and tying up of the lithium in UAl{sub 3} which does not melt until after the uranium does. Pressure from the swelling UAl{sub 3} will extrude uranium-aluminum and lithium into graphite weep holes and block interfaces. The migration of lithium by vaporization will not became appreciable until well over 2000{degrees}C, well beyond the time when uranium and UAl{sub 3} have melted. The eventual result will be a diffuse distribution of uranium, lithium, and aluminium in the lattice. The E-N pile has a larger excess over required control capacity than the uranium provided the large reactivity poison tied up in the lithium is not lost. Compared to the natural uranium pile, the gain of reactivity on loss of coolant is less and the net temperature coefficient in the dry pile remains negative to higher exposures. Furthermore, permanent pile poisoning during meltdown is accomplished via two mechanisms both lithium and uranium redistribution in the lattice produce large negative …

The objective of this supplement is to authorize a change in the panellit trip range from 25--75 psi to 5--95 psi. The test hole facility consists of two concentric aluminum tubes which extend from the front face to the rear face of the reactor. The ID of the inner tube is 2--7/8 inch. Water from one crossheader supplies the annulus, water from another crossheader supplies the inner tube. The three-foot-long, .570 inch OD fuel element is centered in a 40-inch long aluminum holder which has an ID of 1.380 inch and an OD of 2.800 inch. The panellit gage which monitors the flow to the inner tube fluctuates to such an extent during start-up that on two occasions the reactor was scrammed. During equilibrium operation the panellit gage reading remains stable. A possible explanation of this behavior is that during start-up aluminum spacers which are in the inner tube as part of the test charge chatter and cause variations in the water path through the tube. It is further surmised that at equilibrium operation the pressure drop across the column in the tube is sufficient to suppress the chattering. It is concluded that extending the trip range to 5--95 psi …

Technical report issued by E. I. du Pont de Nemours & Co. From Abstract: "Information obtained from investigations pertinent to tritium monitoring and protective measures at the Savannah River Plant are given. These findings were used to establish realistic protective techniques and consequently, to insure the safety of workers exposed to this hazard. Topics included are: contamination, permeation of plastics and rubber, instrumentation, and biological and physical aspects."

Technical report. From Abstract: "Precipitation of manganese dioxide in solutions of uranium-aluminum alloy effected maximum removal of fission products when the alloy solutions were acid deficient. Minimum adsorption of uranium and minimum volume of centrifuged MnO2 were obtained by precipitation in acidic solutions with vigorous agitation."

Technical report. From Abstract : "Ascorbic and isoascorbic acids, used in conjunction with sulfamic acid, reduced Pu(IV) rapidly and completely to Pu(III) in solutions of nitric acid. The solutions of Pu(III) were stable. Aminoguanidine sulfate also retarded the oxidation of Pu(III) but did not reduce Pu(IV)."

Preface: The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming from the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to a steam generator. Because of this activity the until will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shut down it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shut down, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The general arrangement of the heat exchanger described in this report presents a conventional design utilizing known materials and existing methods of fabrication. In further consideration of all concepts, designs and analyses developed during this period of the program, it is felt that this preliminary design will provide an intermediate sodium heat exchanger of lower cost …