This report discusses the conference on direct energy conversion.

Results of experimental investigation for improving intensification and to reduce stress concentrations for anvil redesign.

Description of a data collection and reporting system which was devised and installed in the Union Carbide Nuclear Company's Y-12 Plant Maintenance Division.

This is a map of the geology of Jackass Flats quadrangle.

The objective of this test is to demonstrate and evaluate the design of the proposed N-Reactor fuel surface crud monitor to be used in the N Reactor.

During fuel element assembly, a dip-brazing process., the aluminum can which is ultimately a jacket for a uranium core is protected from the molten AlSi solution by a steel sleeve. Once the can is inserted into the sleeve and thermal expansion is complete, upset metal rings prevent molten AlSi from flowing between the can and sleeve, a situation that warrants rejection. Passage of molten solution beyond upset metal to a point on the can wall not removed during the subsequent facing operation is called an ``AlSi Slopover`` or ``Slop.`` With discovery of a solution which would mask 8001-F aluminum alloy from molten AlSi, efforts were directed towards using this solution to eliminate AlSi stopovers. With proper application and control, PWR 125 proved that cans coated with PX1288 did not alloy AlSi in the molten state to pass beyond the upset metal region during canning. Since this material has been used on total production, the CPR rate has been greatly reduced. Percentage of rejects attributable to the CPR category is 0.5%.

A concentrated study of the present manufacturing process was initiated in the Process Engineering Unit, the purpose of the study being to optimize the present operations, procedures, and methods of the AlSi canning process. It is the purpose of this document to report the findings and accomplishments for that portion of the study concerned with the optimization of component preparation and fabrication within the canning furnace. To date, this study has consisted mainly of heating rate studies of the various components used in the fuel element fabrication with some work being done on the optimization of the heating cycle for these components. This work necessarily serves as the basis for further optimization study.

Prior to the use of venturis on the front of the B, D, DR, P, and H Reactors, much of the pressure and velocity heads developed by the 190 process water pumps were used in overcoming built-in frictional losses in the system external to the process tubes. As Reactor Technology was gained, changes were continually taking place. Venturis replaced orifices installed during original construction. Hydraulically efficient front nozzles, front pigtails, front adaptors, and a second generation of front venturis were designed and installed as a part of Project CG-558. Process changes, such as eliminating upstream dummy charges, were also made to reduce the energy dissipation due to frictional losses. The pressure drop across the active section of a fuel charge was reduced by the internal flow passage in the I&E elements. Little has been done to date, however, to reduce the pressure drops in the rear face fittings. This report is intended to review some of these still existent pressure restrictions and to discuss possible future courses of action.

The current-revision of the Process Optimization Program contains two empirical plutonium-exposure expressions. One relates total Pu conversion to exposure and the other, Pu-240 buildup to exposure. The former was incorporated to save computer time and the latter to provide product quality data. Both of these relationships, in addition to a third which relates Pu-241 buildup to exposure, are used in PULP-3, the data generator for LP-90. EMCON-3, a 7090 computer program which determines arbitrary constants for all three empirical expressions, has been devised. It uses, as input, official conversion data and will facilitate rapid evaluation of the effect, of periodic revisions in official conversion data, changes in reactor loadings, and/or changes in product specification. This document contains a discussion of the empirical expressions, the program mechanics, and program input/output. A Fortran listing of the program logic a deck assembly diagram, and sample program output are included in the appendix to aid the reader in using and/or revising the program.

Certain advantages, such as higher strength at elevated temperature, have led to the use of alloys containing large percentages of nickel as a replacement for aluminum components. One example is the Inconel sheathing of control rods for the Hanford reactors. Since some of these components may remain in a reactor several years prior to removal, the neutron-activation products resulting from multiple chain reactions may become important in determining the radiation exposure levels at discharge. This paper outlines one of the activation chains which may contribute significantly to the {gamma}-activity. Equations are developed from which the concentrations of the product isotopes emanating from nickel-58 can be calculated. ``Best values`` of the parameters used in these calculations are tabulated, and the conversion from isotope concentration to {gamma}-dose rate is outlined.

Introduction: This report covers a year's activity in the continuation of a program designed to explore the missile hazard resulting from a reactor excursion.

The Purex Plant has been requested to process Z-Plant MR and EPT solutions containing about 200 kilograms of plutonium in order to alleviate processing problem at Z-Plant and to reduce the inventory of MR and EPT rework. Campaign recycle of uranium through the main plant is recommended as an economical and effective means of reworking large quantities of Z-Plant plutonium solutions. This memorandum outlines process alternates and costs for reworking the solutions and itemizes specific recommendations for process and critical mass control.

The purpose of this supplement is to authorize charging the complete central zone core of H Reactor with the modified E-N charge makeup, installed as a 102-tube ``test core`` in the fifth loading. The potential economic and efficiency gains of a simplified charge pattern (a smaller average number of N pieces per tube and a minimum number of different charge types) were discussed in Supplement II to the parent PITA. Accordingly, a test block, consisting of 102 process channels in the center of the reactor, was included in the charging of the present (fifth) E-N core loading for the purpose of evaluation. Based on satisfactory preliminary confirmations of the calculated reactivity, front-to-rear flux distribution and other operating characteristics, extension to a full-scale core loading appears feasible. All resulting changes in pile reactivity and other operating variables will be minor and easily adjustable without significantly affecting operating efficiencies.

Abstract: This report presents a simple, quantitative analytical method for determining the percent of a particular group or structure within a urethane compound by means of the infrared spectrophotometer.

From introduction: The pre-shot and postshot mine surveys indicated that no damage attributable to the Shoal event occurred to any of the sixteen private mining properties.

From introduction: This report documents the preshot and postshot condition of all structures within 10 miles of Ground Zero, plus structures at Salt Wells and surface structures at mines that were included in the mine survey.

The PITA`s (process improvement transition authorizations) present the specific requirements necessary to accomplish the outage work and to control the use of this increased flow capability. The procedures and methods to accomplish the programmed modifications are contained in the approve design changes or in supplemental procedures.

A program for increasing the available coolant flow at the H Reactor has been proposed and supported by documentation. This Process Improvement Transition Authorization (being written in two parts) authorizes the initiation of an immediate phase of the program designed to increase total coolant flow by about 6,000 to 10,000 gallons per minute. Part I of this PITA describes in detail the limitations and requirements peculiar to Process Engineering aspects of the proposed modifications and is presented as a separate document. This portion of the PITA, Part II, is written to establish and define conditions and restrictions in the area of Operational Physics, including loading procedures, instrumentation requirements startup checks and procedures related to reactivity effects, total control and speed of control considerations, and other physics functions. The governing criterion shall be that nuclear safety is not compromised during the proposed transition, which includes a complete shutoff of reactor flow for a limited period of time under closely monitored conditions.

If a metal shipment from the reactors to a chemical processing plant contained irradiated Li-Al pieces, there would be a release of tritium during processing. Two situations were analyzed -- one where a single piece is inadvertently mixed in a bucket of irradiated uranium metal pieces, and the extreme case where three buckets of Li-Al instead of uranium are sent in a regular well car shipment. One piece contains about 600 curies of H{sup 3} and three buckets of 585, pieces each would contain a total of about 10{sup 6} curies. Release of tritium from Li-Al pieces in the CPD irradiated uranium metal dissolvers would occur during the caustic coating removal step. About 100 curies are available in three buckets of metal, which could potentially cause (1) exposure of plant personnel up to 1 rem during sample handling or from inhalation of stack gases, (2) exposure of off-site personnel up to 50 mrem, (3) a five per cent increase in ground water tritium, and (4) some restriction on possible future handling of the contents of an underground coating waste storage tank. It is not believed any occupational or general public exposure limits would be exceeded. It would probably exceed an …

This report for November 1963, 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; employee relations; weapons manufacturing; and power and crafts operation.

It is the purpose of this document to provide, in one well-referenced manual, a measure of the degree of reliability that may be placed on routine nuclear material transfer measurements within the Chemical Processing Department, and the relative effects of each measurement uncertainty on the various nuclear material balances. Although statistical techniques were used as a tool in analyzing the data, the results have been presented in a form that should require a bare minimum of statistical training to comprehend, Revisions and corrections will be provided as needed.

This report details activities of the N-Reactor Reactor Physics Department during the month of November 1963.

The objective of this production tests is to irradiate N Reactor prototype and candidate fuel components and alternate products at simulated N-Reactor pressure and temperature conditions in KER Loops 1, 2, 3, and 4.

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