Early results from actual laboratory operation of a more versatile magnet supply.

The purpose of this study was to provide a means for evaluating the various parameters involved in the design and performance of levitation coils.

The chemical potential of the brine from I.I.D. Well No. 1 was evaluated and reported in Rogers Engineering Coo, Inc. March 1962 Flow Test Report. This evaluation considered only the recovery of sodium chloride, as the prime salable product and alternately as the raw material for conversion to salable chlorine and caustic. Both cases showed low net revenue relative to the estimated capital investment required. The subsequent three month well flow test indicates continued production of a consistent type brine can be expected from this well. Based on the flow and composition data established during this test period, chemical production potential has been examined. The separation and sale of high quality potassium chloride appears attractive. Attached are sketches of each of the two production schemes considered with estimates of capital and operating costs and revenue.

From abstract: This report presents information generated during a six month feasibility study of an engine which uses a supercritical working fluid as the secondary portion of nuclear powered electric generating system.

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This report addresses the chemical characterization of selected core samples from the vicinity of Tatum Salt Dome, Lamar County, Mississippi.

From abstract: This report is on the geologic studies in the U12e.07 tunnel for the purpose of determining the structural features, lithographic character, chemical composition, and physical properties of the tuff exposed in the tunnel.

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Estimates are made relating radionuclide movement to ground water velocity as part of the safety program for a proposed experiment to detonate nuclear devices within the Tatum salt dome. The estimates are based on distribution coefficients obtained from laboratory studies. Core samples obtained from hydrologic test well HT-3, Tatum salt dome, Lamar County, Mississippi, were equilibrated with radionuclides in solutions simulating aquifer waters found in the area. The combinations of Cenozoic sand and silty clay, and quality of water of the area were studied and summarized. The distribution coefficients obtained for different radionuclides were tested and indicate retardation factors from 1.3 to 857 for the travel time of these radionuclides when compared to the travel time of water in the aquifer system. Laboratory results indicate that migration of any radioisotope inadvertently introduced to the aquifers in the vicinity of the dome as a result of proposed nuclear test explosions would be extremely slow. Revised estimates of the rate of dissolved radioisotope movement will be made on the basis of further laboratory studies utilizing chromatographic adsorption columns of 0.5 to 4.0 feet in length.

On September 13 and 14, 1962, 8.3 curies of iodine-131 were emitted from the Hanford Redox Plant at a rate of from 0.35 to 0.65 curies per hour for a period of approximately 18 hours. During the emission, the plume trajectories were plotted from meteorological data and samples were collected across the predicted plume trajectories at several altitudes and at distances up to 50 miles from the Plant. Filter, charcoal trap, and air samples were collected during the emission using aircraft; and caustic scrubber, vegetation, and milk samples were collected during and after the emission at selected ground stations in the usual manner. Appropriate background samples were collected before the test began. The data and conclusions are given in this report.

This document details activities of the N-Reactor Department during the month of March 1963.

It is the purpose of this document to provide the scope of piping changes that are needed to supply gravity-fed water to the reactor storage basins on an unattended basis following complete reactor discharge of metal into these basins.

The purpose of this document is to describe an engineering program being conducted by Reactor Design, Facilities Engineering Section to provide a solution for current and anticipated problems of binding and inoperative horizontal control rods (HCR`s) at the ``K`` Reactors. Primary emphasis in the program is On alleviating problems at the K Reactors because of the more advanced stage of graphite distortion and hence control rod problems existing at the K Reactors as compared to the other reactors. It is recognized, however, that problems of similar severity can reasonably be anticipated at the other reactors in the near future. With this in mind, design of remedial measures for K will incorporate consideration of adaptability to the other reactor HCR systems. Also, primary concern of this program is to provide an HCR system which will fulfill functional requirements for safe operation of the reactors at current power levels under conditions of severe HCR channel distortion. Suitability of the system for producing by products such as tritium is considered to be of secondary importance at this time although still a point to be kept in mind in designing the new rod tip and to be studied in detail at a later date.

The loss associated with low goaling rupture-prone material is quite small compared to the benefits. There is an optimum goal exposure for a given metal quality. At this optimum exposure the loss associated with irradiating rupture-prone material is minimized. Reasonable decision rules are as follows: Discharge, at the first opportunity, tubes containing rupture-prone material whose exposures are equal to or greater than 50 percent of normal goal. Discharge tubes under 50 percent of normal goal which would be expected to reach 70 percent of normal goal before the next scheduled outage. Irradiate all tubes containing rupture-prone material whose exposures will not be expected to exceed 70 percent of normal goal before the next scheduled outage. Discharge all tubes containing rupture-prone material, regardless of exposure, if an additional rupture is incurred in the lot. Return unirradiated rupture-prone material to Production Fuels Section as reject material.

With the use of smooth-bore Zircaloy-2 process tubes and self- supported fuel elements on a large scale at the K Reactors, establishment of the operating characteristics of this fuel element-process tube configuration is imperative. Under authority of Production Test-IP-409-A, two smooth-bore Zircaloy-2 tubes were routinely charged with KVNS fuel elements. Approximately 520 KVNS fuel elements have been irradiated in this facility. In Production Test-IP-572-A, the coolant temperature distribution in the annulus of the KVNS fuel element-process tube system is being evaluated as a function of fuel element support height. This production test supplement will provide data for further defining the operational characteristics of the self-support fuel element concept.

A testing program designed to provide the necessary data for evaluation of the emergency flow backup adequacy was conducted at the K reactors on February 16, 1963. Three transient or trip-out tests and 13 steady-state tests were performed under authorization of PT IP-573-I. This report contains the numerical results of the pressure and flow data that were obtained during these tests.

The purpose of this report is to permit the charging of the fourth E-N demonstration loading at H Reactor; modifications in the charge makeup and the loading pattern, designed to improve the conversion ratio and to advantageously re-shape the longitudinal flux pattern, are recommended. These gains are available through utilization of the small surplus of excess reactivity, over operating requirements, in the present (third) core.

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rod as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.

This is the monthly report for the Hanford Laboratories Operation March 1963. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

The objective of this test is to produce 50 milligrams of neptunium 239 for a process stream study in the Purex plant. The efficient recovery of neptunium 237 as an alternate product produced at HAPO is presently impaired by the lack of knowledge of the distribution of neptunium 237 in the several Purex process streams. Present methods of following the neptunium 237 through the process streams involves inadequate alpha counting techniques. However, neptunium 239, because of its discrete gamma decay energy, can be readily identified in the process stream. As a result, the neptunium 239 produced by this irradiation will serve as a gamma-emitting tracer for the neptunium 237 in the Purex process stream.

The KER loops are four pressurized water-cooled in-reactor testing loops in which fuel, coolant and corrosion tests are being conducted for the N Reactor program. Two of the loops, KER-3 and KER-4, are equipped with N-size Zr-2 process tubes and have sufficient pumping and heat dissipation capacity to permit testing of 14-element charges of N fuel. The other two loops, HER-1 and KER-2, are fitted with Zr-2 process tubes of 2.1 in. inner diameter (vs 2.7 in. for N). The small loops can be used to test charges of N inner tubes or other elements sized at about 1.8 in. outer diameter or less. An inert sleeve must be used with N inner tubes to assure adequate coolant flow over both surfaces of the elements. The KE reactor active zone is 28 in. long com pared to 35 in. for N so the length of test charges is somewhat shorter than actual N charges will be. The purpose of this report is to review the testing environment, to discuss the testing program and to present the schedule as presently seen.

This report discusses Hanford`s contribution to river water temperatures and describes the Pasco and Richland Water Treatment Plants.

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Waggoner Carr, regarding a legal question submitted for clarification: Constitutionality of Senate Bill No. 212

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Waggoner Carr, regarding a legal question submitted for clarification: Under the submitted facts, whether the Gross Receipts Tax provided for by Article 11.03, Taxation-General, is due by the stated corporations.