This is the monthly report for the Hanford Laboratories Operation, November 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This is the monthly report for the Hanford Laboratories Operation, July 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities,, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This is the monthly report for the Hanford Laboratories Operation, January 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This is the monthly report for the Hanford Laboratories Operation, February, 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

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 May 1960.

As set forth in the CPD Waste Management Program (HW-63958), new and improved methods of waste processing are planned to be initiated at Hanford. A major part of this program is concerned with the high-level wastes from the Purex Plant, with major objectives being the control of long-term hazards, the insurance of production continuity, and the attainment of positive waste accountability at reasonable over-all costs. Specific plans for the disposal of Purex wastes are based on the use of high temperature calcination for immobilization and volume reduction of the solvent extraction raffinates (1WW). Four major steps are involved, each of which is of basic importance to the program objectives. These are: (1) interim 1WW storage and removal for calcination, (2) calcination, (3) primary containment of the calcined waste, and (4) long-term storage. Immobilization of coating wastes and organic wash wastes is also planned, either by tank solidification methods or possibly by inclusion in the calcination feed stream. This document was dated September 15, 1960.

This is the monthly report for the Hanford Laboratories Operation, October, 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

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This document presents a summary of work and progress at the Hanford Engineer Works for July, 1969. The report is divided into sections by department. A plant wide general summary is included at the beginning of the report, after which the departmental summaries begin. The manufacturing Department reports plant statistics, and summaries for the Metal Preparation, Reactor and separation sections. The Engineering Department`s section summarizes work for the Technical, Design, and Project sections. Costs for the various departments are presented in the Financial Department`s summary. The Medical, Radiological Sciences, Utilities and General Services, Employee and Public Relations, and Community Real Estate and Services Departments have sections presenting their monthly statistics, work, progress, and summaries.

R and D is reported in the following: Reactor and Fuels (PRTR, Pu fabrication pilot plant, KER, NPR, materials); Chemical R and D (Pm recovery, fission products, Purex column, non-production fuels reprocessing, Salt Cycle process); Physics and Instrument R and D (PCTR, NPR, critical experiments, PRTR); and Biology (monitoring, irradiation experiments).

This document details activities of the irradiation processing department during the month of June 1960. 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 Fuels Preparation Department during the month of July 1960. (FI)

Three small samples of zircalloy-2, obtained from a KER tube, were irradiated in the Quickie Facility in F Reactor for one, four, and nine months, respectively. Each sample weighed approximately two grams. One day after discharge from the reactor, the specific activity of each of the samples was essentially the same (about 150 mr/hr/gram at one foot in air). Attenuation measurements taken one day after discharge using lead plates yielded a ten-fold reduction in gamma intensity for each 0.95 inches of lead. This indicates that an enclosure for kilogram samples of zircalloy which have been irradiated for long periods of time at fluxes up to 10{sup 14} neutrons per square centimeter per second should have six inch thick lead walls or the equivalent for gamma shielding. Based on the effective gamma energy and the half-life for the first day, it appears that Zr-89 may be the predominant isotope initially. After a few days, the half-life become long and gamma spectrometer measurements indicate the gamma softens revealing that Zr-95 might become the predominant isotope. However, isotopes were not determined specifically; impurities, other isotopes, or combination of isotopes could yield the same results.

The objective of this production test is to measure deposition on fuel element temperatures during irradiation in high pH water. A Zircaloy-2 jacketed natural uranium solid cylindrical fuel element with thermocouples located at various positions in the element and adjacent coolant channel will be irradiated in the KER Loops to measure the effect of crud film buildup on fuel temperature. Two 1.82 by 1.45 inch enriched 36 inch long tubular elements will be used to provide enough heat to permit operation of the loops at temperature. Duration of the test will depend upon the results obtained; however, the maximum exposure of any of the uranium in the loading will not exceed 1000 MWD/T.

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 March 1960.

This document details activities of the Hanford Laboratories Operation for the month of August 1960.

The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

The assumption of tube replacement losses can have a significant effect in the determination of optimum power levels and goal exposures. The tube replacement loss term in the reactor process optimization model is based on calculating the average projected tube replacement losses that will result from operation at given power and temperature conditions. Tube replacement losses associated with external corrosion, Van Stone flange failure, etc. (i.e., basically non-power level-temperature dependent) are assumed to be miscellaneous losses and are not included in the tube replacement term. Over a period of several years the experienced tube replacement losses (due to internal corrosion) should compare reasonably well with the losses predicted by the optimization model. Planned tube replacement project action which would require, in some cases, premature tube removal is a special situation which requires a modified approach to the prediction of tube replacement losses and to the optimization of reactor power levels prior to project action. A method has been developed for optimizing reactor power levels and goal exposures based on any assumed calendar date for major tube replacement project action. It is the purpose of this document to discuss the general application of this method in the optimization and illustrate the …

The cap-spire pulsing technique of preheating the cap-spire portion of the fuel assembly does significantly improve the brazing of the cap-spire assembly. The air pocket at the spire wafer junction is fully removed. The cap side wafer is essentially 100% wetted with brazing alloy. Destructive tests show that a 4 to 1 improvement in most quality measurements is achieved over present cap preheating techniques, without using additional cleaning step of spire etching. The pulsing is accomplished by a cammed drive system, using a stroke of three-fourths of an inch with a spring return. The system is driven by an electrical gear reduction motor at a rate of 1.4 pulses per second. A preheating cycle of 21 {plus_minus} 2 seconds is used for the current I&E cap designs. The cap-spire assembly does not require any special treatment other than the normal chemical cleaning.

This report details activities of the Fuels Development Operation during the months of July, August, and September 1960.

The program`s objective is to develop the technical foundation for use of projection fuel elements through an accelerated testing program. Two types of projection fuel elements are being considered (1) Self-supported fuel elements for use in ribless aluminum or zirconium process tubes; and (2) Bumper fuel elements for use in ribbed process tubes. A further objective of the program is to determine the most desirable method of fabrication and type of support for projection fuel elements. The program must develop the self-supported concept in time to meet the proposed ribless zirconium process tube installation schedule. Parallel development of the bumper fuel element is an interim program to realize immediate gains in reduced hot spot rupture tendency, during the change over to zirconium tubes. The appended charts indicate the proposed accelerated testing and development schedules being followed. Testing of the self-support concept has indicated that the collapsible supports are adequate to maintain fuel centering, do not introduce additional corrosion problems resulting from the supports, and do in fact greatly reduce the incidence of hot-spot flow patterns. Reductions in rupture potential of at least a factor of 30 have been demonstrated at the 95% confidence level. Bumper fuel testing has indicated that …

Reactor maintenance problems and engineering studies have identified various reactor rear face piping problems at 105-B, D, DR, F, and H. The long-range solutions to these problems are outlined in classified document HW-65269, ``Existing Reactor Rear Face Piping Review.`` These problems, however, must also be considered from the short-range view to minimize their effects on continuity of reactor operation during the interim until the permanent remedy can be accomplished. This is expected to be a minimum period of approximately four years. Flow and bulk outlet temperature up to 90,000 gpm and 95{degrees}C respectively are assumed to prevail in these short-range considerations. The purpose of this report is to review Reactor, Plant Engineering`s efforts on the short-range aspects of this problem.

The demonstrated operating efficiency performance which has resulted from poison spline usage forces an economic decision concerning the self-supported and bumper fuel element programs. As originally conceived the projection fuel elements would preclude the insertion of a spline under the fuel charge; thus it is very important that means be devised either to make poison spline usage compatible with future pile loadings or to demonstrate that some other supplementary control system, which is compatible with future pile loadings, can approximate the effect the splines have an operating efficiency. This report shows the appreciable performance improvement which has been achieved at KE Reactor through the application of the poison spline system.

Three tests have been run in the ETR in high temperature, high pressure, recirculating water. In one test, previously unirradiated fuel elements were used and in the other two the fuel was irradiated to 2400 MWD/T at HAPO prior to insertion in the ETR. Failure was initiated by shearing off a projection on the surface of one rod of a fuel element, thus opening a 25-mil hole through the cladding. The projection was sheared off by a hydraulically operated chisel controlled from outside the reactor. The first test was operated seven hours after the defect was opened with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second test was operated for fourteen hours after the defect was opened with again no failure. The third test was operated for only 33 minutes after the defect cap was sheared off before fission product activity in the loop water caused the test to be terminated.