An analytical trajectory and aerothermodynamic analysis of a satellite containing a Task 2 thermoelectric generator was completed. A 300-statute mile circular polar orbit was used for this analysis and the launch was assumed to be from Vandenberg Air Force Base. Results of this study show that upon natural decay from a successful mission, the radio-cerium fuel will burn up in space at high altitude, thus only a very minor amount of radio cerium will be released to the stratosphere. A complete analyses of the fate of the radio-cerium fuel following various aborted launching attempts also was carried out. Charts summarizing the various assumed failures and locations of the fuel following failure are shown. A technical discussion of the methods used in performing the analysis is included in the report. (auth)

A study was made of two problems of salt-cavity storage, namely, seepage of wastes out of formations and closure of cavities due to plastic flow of salt. The results indicate that both problems are negligible; bedded salt is more impermeable than dome salt. Kerosene was found to be nonreactive with dome salt whereas brine solutions showed some interaction. It is concluded that storage of radioactive wastes in salt cavities is feasible. (D.L.C.)

This report addresses the control design review of P&W quarterly for July to September 1960.

This report, from the Chemical Processing Department at HAPO for November 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

Experimental data on plutonium yield and U{sup 235} burnout are now available on the E metal portion of three central zone striped E-N columns and one fringe blanket E column. These data and an overall E-N load conversion ratio based on experimental data are now reported.

This document provides the loading and operating conditions for eight 16-inch or five 24-inch elements, either natural NIN-1 or .947% enriched NIE-1, in any of the four KER loops. The loadings are tabulated; operating conditions for either charge in KER 2, 3, 4, or 1 are given in figures.

The projected E-N load at H Reactor will require a complete change in the type of charge placed in most flattened zone process tubes. Objective of this test is to determine the reactivity of the proposed E-N charge relative to the known natural uranium charge reactivity.

Objective of the test was to determine and evaluate the behavior of 8 20-inch fuel elements during irradiation using the hot-headed end closures on the inner tube.

In order to reduce both charge-discharge shutdown time and the number of manhours of radiation exposure, Project CGI-884 is being completed at the B, D, DR, F and R Reactors. This consists essentially of installing a large drain line at the bottom of one rear reactor riser. This drain line passes to a control valve and then to the effluent line beyond the downcomer. This system by-passes the crossover downcomer part of the effluent system and eliminates the need for intermittent rear crossheader valving during reactor charge-discharge procedures. Two aspects of this system have been considered, its basic design requirements, and operating restrictions to ensure adequate process tube cooling. Because of the complexity of the reactor flow system approximate solutions were used to compare different methods or degrees of operation and establish limits. Despite these approximations, there was sufficient difference in the case results to justify the specific conclusions presented in this report. This report should serve the dual purpose of providing design requirements for the crossunder and also providing the technical criteria necessary for the operating standards for the use of this new system.

Irradiation Processing Department opinions regarding HCR sleeves. It is believed that Zr-2 would be a satisfactory sleeve material providing it is limited to maximum service temperatures the order of 400 degrees C. Both Inconel and a 300 series stainless steel are believed possibilities for uncooled sleeve materials, with Inconel our preference. Beryllium, in virtue of its outstanding nuclear properties and probably satisfactory chemical and mechanical properties, warrants consideration on a pilot basis if fabrication of a suitable sleeve form is practicable. Consideration of any sleeve material should include techniques for minimizing the friction between the sleeve and the control rod -- particularly when a fretting -- prone material, such as Zr-2, is involved. A ball, or roller, bearing rod support, with a ceramic bearing contacting the sleeve, might be considered as a device for reducing sleeve wear and improving rod manipulation.

The exposures calculated from these plans are designed to maximize plant return on the basis that the only restraint on metal usage is that imposed by the economics of the plutonium production process and associated uranium cycle. When metal throughput in limited by ex-reactor physical capabilities to a level lower than the unrestrained requirements, exposures higher than those recommended in this document would be indicated. In essence then, these plans represent minimum economical exposures for current production fuel types, on the basis of attempting to maximize plant returns. The adjustments recommended herein, stem solely from revised estimates of the metal performance level indicies (C{sub R} values) of the various metal types.

This monthly report details activities of the Hanford Operations Office during the month of December 1960. (FI)

This document contains information about flooding of the Columbia River. It focuses attention on the following; increased elevation due to rainfall, seepage which destabilized the constructed dike, flooding of cellars, evacuation of people to emergency shelters, tug boat collision damage to power lines, and the washout of the Van Giesen Street Bridge on Yakima River.

Because of difficulties in charging eight-inch fuel elements in bowed process tubes in the upper parts of the old reactors, IPD has requested that four-inch I & E fuel elements be provided for these tubes early in CY1961. Approximately 30,000 fuel elements of this design, Model OVN, will be required per year. Since most of these fuel elements will be charged in the upper fringe zones, they will be in-reactor for an extended period of time. For this reason, the quality of this material should be as high as possible. This report contains a summary of tests made to determine optimum canning conditions and to establish process specifications for canning four-inch I & E fuel elements.

Three E-N loaded tubes were dissolved, sampled and analyzed, starting November 9, 1960. The results of these analyses and an explanation of the methods used are the subject of this report. Each tube loading received an identification code in each facility in which it was processed. All of these codes are listed for future reference. Each batch of slugs was dissolved in a preflushed dissolver. When complete solution was indicated by a leveling off of the specific gravity, two 1 ml samples were taken and analyzed for uranium, specific gravity, and excess nitric acid. The dissolver charge was digested an additional four hours. At the end of the digestion period, two 1 ml samples and one 20 ml pig sample were taken. The three samples were analyzed for U, SpG, and HNO{sub 3}. Agreement between these samples and the previous samples was taken as confirmation of complete dissolution and representative sampling. If agreement was not obtained, sampling was continued. After verification of the pig sample, six 1 ml aliquots were taken for analysis by the Analytical Control group. The remainder was aliquoted to provide material for mass analysis and for analysis by the Process Chemistry group.

The release of radio-iodine from the irradiated uranium during dissolution, and the emission of a portion of this radio-iodine to the environs via the separations plant stacks continues to be a problem in spite of the expenditure of several million dollars for equipment designed to remove the radio-iodine from the gaseous wastes. This report discusses the control of the radio-iodine emissions from the hanford separation processes.(JL)

Although a fuel reactor job procedure has not yet been prepared, techniques and equipment used in a current production test are adequate with minor modifications for full reactor Parker fitting reaming. With existing reactor hardware, available flow increases at B, D, and F Reactors would be limited to about 2,500 gpm. The extent of additional rear face piping problems which may be encountered as a result of reaming cannot in several instances be predicted with confidence at this time. It is expected that preliminary answers to currently foreseen problems will be available by April 1, 1961. To some extent, however, this is dependent on production tests under way or proposed, and will require coordination with various Manufacturing Operations. For vibration and flow problems in the downcomer approach piping, conclusive answers may not be available before July, 196l, although there is no positive evidence that the flow increases cannot be absorbed. Increased pressurization of effluent line junction boxes at 105-D has been solved by improved box venting. Increased flows resulting from reaming may require still greater venting capacity, which would be available for capital expenditures on the order of $3,000--$5,000 per area.

This document consists of length and annealing data sheets.

Recent studies have confirmed that large incentives exist for overboring the reactor process channels approximately 500 mils in the C and five old reactors under the proposed Plant Improvement Program. Conservative estimates of the incentives for overboring indicate a payout period of about two years for the proposed work, an increase in plutonium production of 15--18%, derived from increased conversion ratio and a reduction in plant unit cost. Since sufficient progress has been made in the areas of reactor development to warrant the necessary planning and budgetary action to accomplish the proposed work, a preliminary study has been conducted to determine and assess the technical problems associated with the processing and performance of oversize fuels, to define those areas requiring further development, and to provide cost estimates for the conversion and operation of both offsite uranium fabrication facilities and HAPO fuel manufacturing facilities as required to produce oversize fuel elements. Details of this study are contained in this report.

The objective of the production test detailed in this report is to evaluate the behavior of N-Reactor inner fuel tubes during irradiation in the KER Loops at conditions equivalent to those expected in the N-Reactor. Natural and .947% enriched Zircaloy-2 jacketed N-Reactor inner fuel tubes (NIN-1) and (NIE-1) will be irradiated in the KER Loops to exposures no greater than 3500 MWD/T. The elements have brazed end closures and Zircaloy-2 supports and are contained in Zr-2 flow distributing sleeves.

A nomogram is attached as Figure 1 which may be used to determine TAI limits. The nomogram was developed to meet the need for a rapid and accurate method for calculating these limits. The technical basis for determining individual tube outlet temperature limits is given in HW-65729.

A study was made to establish the validity of using homogeneous-lattice machine program calculations of the relative conversion ratio of the overbored lattice as compared to the present case. It is possible to draw both qualitative and quantitative conclusions from hand calculations utilizing simplified geometries; it is the purpose of this document to summarize such a study. Numbers included herein are not intended to represent absolute quantities, since the required calculations lie outside the scope of this study. Nevertheless, it is felt that a comparison of the overbored to the existing lattice based on the values presented here is a good ``first order`` approximation.

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