A number of metals, carbides, nitrides, oxides, borides, and uranium gulfides are briefly evaluated concerning their possible applications as advanced nuclear rocket materials. (R.J.S.)

Information of interest to analytical chemists is presented in a report containing both positive and negative results obtained in a total of 58,467 determinations. The data and information are presented in sections concernlng the work of the shift laboratory, special analysis laboratory, spectral analytical group, analytical development group, quality control and standards laboratory, and analytical service for EOCR. Details of methods added to the ICPP analytical manual, and to the ICPP analytical radiochemlcal manual are included. (J.R.D.)

S>A program will be required for controlling the wall temperature of a pot calciner during calcination in order to prevent an excessive temperature rise from internal heat generation while assuring complete calcination of the cake at 1650 deg F. A linear approximation to a pot wall temperature program indicates that this can be accomplished by holding the wall temperature at 1650 deg for 0.8 to 1.0 hrs and then cooling the wall at rates dependent upon the cake internal heat generation rate and average thermal conductivity. A similar study indicates that if it is necessary to reheat the cake for complete calcination, the wall heating time must be limited to prevent excessive temperature rise in the cake. Reheating times are also dependent upon the cake internal heat generation and thermal conductivity. (auth)

Preparation of enriched fuel elements and target (N) elements for a full H-Reactor E-N demonstration loading was authorized in April, 1960. The objective of this loading was to demonstrate the feasibility of operating a Hanford reactor economically as a plutonium-tritium producer to increase over-all conversion ratio and diversify product output. Irradiation of the E-N load began in May, 1961 and was completed in January, 1962.

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Appendices C and D may further be identified as SL1925 and CF-61-12- 20(Rev.), respectively. A comparative report is presented in which the economics and feasibility of plant protection from nuclear attack by plant hardening, remote siting, and utilization of optional fueling concepts for the coal-fired plant are evaluated. (J.R.D.)

A series of computer programs are described which Were successfully applied to the reduction of pulse-height data using digital computers. Calculations carried out by these programs include: determination of pulse height vs energy response of an NaI detector; calculation of coincidence sum spectra; and the analysis of complex pulse-height spectra to obtain energies and relative intensities of individual gamma rays or relative abundances from a mixture of radionuclides. Measurement problems and instrumental requirements are discussed in detail. (auth)

Calculations are presented to show the effect of uraniumbearing scale deposits on the HRE-2 core wall tempera ture, heat flux, and the temperature. of patches proposed for sealing the holes of the HRE-2 core vessel. Flow behavior in the HRE-2 core mockup with simulated patches installed and simulated corrosion pits drilled in the vessel wall was determined. A plausible mechanism was advanced to account for hot spots due to sedimentation of solids on metal surfaces projecting from the core wall. Patch designs with reduced metal projections from the core wall are presented. (auth)

Calcium-magnesium precipitation apparatus was used to reduce the concentrations of these elements in ORNL tap water, used as a substitute for waste water of low level of radioactivity, prior to strontium removal by foam separation. With and without alkali and flocculator chambers and with a stirred sludge of ratio height to diameter equal to 1/1 to ~4/1, use of 5 x 10/sup -3/ M each of NaOH and Na/sub 2/CO/sub 3/ and 2 ppm Fe/sup 3+/ reduced the dissolved Ca + Mg concentrations to 1 to 2 ppm as calcium. Simultaneously, a strontium DF of 20 to 200 was achieved, and, by adding Grundite clay in the proportion ~0.5 1b/ 1000 gal, a cesium DF of 10 to 40 was achieved. (auth)

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Reproducibly high density UC-PuC pellets containing minor amounts of (U,Pu)/sub 2/C/sub 3/ were made with nickel sintering aid. Average densities of 95% and maximum densities of 96.6% of theoretical were obtained. High density, singlephase UC-PuC pellets were made by sintering at high temperatures. Densities of 94% of theoretical appear to be attainable. Improved identification of microstructural constituents was made by means of metallography, x-ray diffraction, and microprobe analysis. Development of specialized chemical analysis techniques continued. The post-irradiation examination of the UC specimens was completed. Mass spectrometer analysis showed that 0.4% of the fission gases was released. Metallographic examination of the UC did not reveal any channges in structure. Metallographic examination of the metal specimen parts that were in contact with UC was completed. No evidence for interaction with niobium and slight evidence for interaction with stainless steel was found. The latter was not deleterious. Perturbation factors were calculated for the UC- PuC experiments. Negotiations continued, in order to select the most suitable test reactor. (auth)

A large number of irradiation tests have been conducted with UO{sub 2} fuel elements to characterize the behavior of Zircaloy cladding when cooling water is allowed to penetrate the cladding to the fuel core. The test element, is designed so that cooling water will be in contact with the surface of the UO{sub 2} fuel core during irradiation. The irradiation behavior of the UO{sub 2} will be compared to the behavior of similarly irradiated, non-defected UO{sub 2} fuel-elements. Several characteristics are of particular interest: (1) Previous investigations have shown that fission fragments formed in UO{sub 2} fuel cores tend to migrate from their point of formation. The fission fragment distribution in this element will be compared with that observed in other irradiated UO{sub 2} fuel cores. (2) This test will provide a direct measure of fission products released from UO{sub 2} during irradiation. (3) Sintered UO{sub 2} is quite resistant to erosion and is, in general, chemically stable in a water environment. This test will provide data concerning the dissolution rate of UO{sub 2} in water during irradiation. (4) Since the coolant will be in direct contact with the fuel, the average temperature of the fuel will be comparatively low. …

A survey of the water requirements is presented for a hypothetical plant to process all the fuel from a 15,000Mwe nuclear economy. For each processing plant, specific requirements must be based on a detailed water survey which includes water quality, process requirements, and in-plant conservation plans. These considerations are discussed and the quantitative requirements are listed. (J.R.D.)

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On September 19, 1961, the No. 3 horizontal control rod at KW reactor was removed after an apparent failure of the wall separating the boron carbide powder from the coolant water. The instrument indications that were interpreted as a rod failure include: high radiation alarms in the exhaust air, unexplained gain in reactivity, increasing coolant outlet temperature on No. 3 HCR, and high radiation readings of the outlet water from No. 3 HCR. The possible reactor safety aspects of such a failure made it necessary to obtain a thorough examination of the rod and inner coolant tube. A complete borescope examination of the rod and partial visual examination of the inner coolant tube have recently been completed. This document is intended to summarize the inspection results, discuss the safety and costs aspects of a horizontal rod failure, and suggest courses of action for the remaining rods at KE and KW reactors.

An addition is presented for the IBM-7090 program MURGATROYD to produce a rough graph of reactor power versus time. A sample of output is included for the case given as an example. (J.R.D.)

This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company.

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A Monte Carlo IBM-7090 program (OGRE-P1) was written for calculation of the dose rate on one side of a slab owing to an isotropic, cosine, or collimated monoenergetic gamma radiation source on the other side of the slab. A maximum of 50 homogeneous regions are permitted. (auth)

The three alterations discussed in this report have an effect on the parameters used in Total Control Calculations for C Reactor: Conversion to smaller diameter and shorter vertical safety rods in six outer control channels; installation of forty-four overbored tubes with short fuel charges in the flat zone and an associated increase in spike enrichment; and centralization of the enrichment ring into a three-lattice unit width in place of three and one-half lattice units previously in use. New parameters are derived in this document for future total control calculations.

We present the results of measurements of the angular distribution of fission fragments produced by irradiation of Au{sup 197} and Bi{sup 209} with various heavy ions. The projectiles, B{sup 11}, C{sup 12}, N{sup 14}, and O{sup 16}, had energies from a few MeV above the Coulomb barrier to 10.4 MeV per nucleon. The gross features of these results can be explained by use of a model and parameters that have been used by others to account for angular distributions of fission fragments from helium-ion bombardments. In detail, however, these results appear to indicate that the models used to predict the average value of the angular momentum of the compound nucleus give values too low near the Coulomb barrier. Furthermore, at high bombarding energies it is necessary to consider the fact that appreciable direct interaction is taking place.

This report, from the Chemical Processing Department at HAPO, for April 1962 discusses the following: Production operation; Purex and Redox operation; finished products operation; maintenance; financial operations; facilities engineering; research; employee relations; special separation processing; and auxiliaries operation.

During the past four years, IBM data processing computers have been utilized to perform countless studies involving the optimization of the HAPO plutonium production complex (RA.PO and off-site supporting plants). During this time, three process optimization models have been constructed and associated Fortran programs have been written. These three are (1) maximum net return, (2) minimum reactor unit cost, and (3) maximum grams. This document will be concerned only with the first program: net return optimization. In January of 1960, a new program was written for the IBM EDPM 709. This new program incorporated the desirable features of the previous programs and provided additional output data. During the two years since January, 1960, the net return optimization program has undergone many changes. These changes have been caused by improvements in the process technology and a major change in the definition of the Pu-240 specification. It is the purpose of this document to describe the current version of the net return optimization program.

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