This document considers the costs of rhenium formation as produced by irradiating tungsten. Two isotopic compositions of tungsten are considered for the study. The cost for reactor-formed rhenium appears to be prohibitively high -- over $20 per gram. This cost would exist for tungsten containing 90% 186, 9, 184, and 1% tungsten 182 and 183. The cost of alloy made fro natural isotopic compositions of tungsten would be higher by a factor of 3, and would take prohibitively long to produce significant quantities of rhenium. Thus, detailed numbers are not shown or considered for the natural isotopic composition of tungsten.

None

This report contains the feasibility and cost data for the production of tungsten-rhenium alloys from tungsten targets in the N-Reactor. The two types of target elements assumed were: (a) tungsten containing 90 a/o tungsten-186, 9 a/o tungsten-184 and 1 a/o tungsten-183 and 182, and (b) tungsten of natural isotopic composition.

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.

This report summarizes and highlights the more importan milestones, development programs, performance and characteristics, and properties of dingot uranium and its use in the Hanford reactors from 1955 to the present. For the benefit of those unfamiliar with the terms ingot and dingot uranium as used in this report, ingot uranium refers to the metal made at the Fernald Plant of the National Lead Company by remelting under vacuum a mixed charge of solid scrap, briquetted scrap, and derby metal (the product of the UF{sub 4} bomb reduction), while dingot metal refers to the metal made by Mallinckrodt Chemical Works by reduction of UF{sub 4} to metal which is used directly (after suitable forming into convenient size for rolling), without the intermediate vacuum remelting step.

A half-plant low alum treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by results obtained from a statistical analysis of fuel element ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge and groove corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in process water treated with two different alum feed rates. This report presents the results from 20 fuel charges, ten from each side of F Reactor, which were discharged such that the near side fuel pieces had been exposed for 75 days to water treated with 18 ppm alum and the far side pieces had been exposed 75 days to water treated with 8 ppm alum.

A half-plant low alum water treatment test began at F Reactor on January 16, 1963 at startup from the scheduled January 3 tube replacement outage. The test, which was prompted by results obtained from a statistical analysis of fuel ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in two different alum treated process waters. This report discusses the results obtained from twenty fuel charges, ten from each side of F Reactor, which were discharged prior to the reduction in alum feed to establish the pre-test corrosion environment.

A half-plant high flocculation pH test began at B Reactor on September 24, 1963. The purpose of the test was to determine whether operation of the water plant flocculation basins and filters at a pH higher than 7.0 with subsequent coolant pH adjustment to 6.6 is beneficial as far as radioactivity in the effluent is concerned. It is planned to increase the flocculation basin pH in steps to a maximum of 7.6. Although primarily an effluent activity test, fuel discharged after a metal cycle at each water plant condition will be monitored by visual examination. Thus, it will be possible to determine whether the incidence of localized corrosion is influenced by the various modes of operation. One 20 column fuel discharge, ten columns from each side of the reactor, was obtained prior to the start of the test. This report discusses the results obtained from the first discharge of fuel under test conditions.

A computer program was written to calculate the initial ratio of isotopes which should be used in an in-core neutron flux detector to obtain the maximum use lifetime where the useful lifetime is defined as that period of time during which the sensitivity of the detector remains within specified limits. Input data for the program include cross sections of the isotopes to be used. In order that various reactor environments might be considered, Westcott cross sections have been employed. Westcott cross sections assume that the neutron spectrum in a thermal reactor can be characterized by the temperature, T, and by a spectral parameter, r. An experimental technique was developed for measuring these parameters. The evaluation of the parameter, r, is based on a standard technique comparing the relative radioactivity induced in bare and cadmium-covered cobalt samples. The determination of the neutron temperature, T, was made using a mass spectrometer to determine the isotopic changes in uranium and plutonium samples as a function of.exposure in the neutron flux. This is a new technique which appears to be both simple and accurate. The irradiations were made in a water-cooled irradiation facility located in the KE Hanford production reactor. Based on the results …

We have estimated the effects on production efficiency of raising the plutonium product specification. The various cases have been considered using a pH of 6.6 at all reactors. In addition to this, the following estimates and assumptions have been used: (1) failure rates will vary with the fifth power of exposure and directly with fuel throughput, (2) charge-discharge losses will be proportional to throughput, (3) throughput will vary inversely with exposure, (4) the failure rates of the K5 self-support fuel are unknown, but are assumed to equal present K4 rate.

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 document details activities of the Reactor Branch during the months of January through December 1963. (FI)

This report discusses nuclear processes involved in heavy isotope production in the N-Reactor. Production methods, production rates, and the nuclear advantages of the N-Reactor are presented.

None

None

None

This report is an explanation of the response characteristics, some typical curves of displacements, velocities, etc., and the computer circuit diagram.

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodying hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)

A simple method is developed for calculating or predicting temperature distributions in spent reactor fuels in shipping casks. The method accounts for radiant heat transfer between all the individual pins in a square array. With the dimensions of the fuel bundle, the configuration factors for radiation between various tubes in the bundle can be obtained from the tabulated numerical calculations presented. The configuration factors, along with the heat generation rates, surface emissivity, and the temperature of the wall of the cask can be used to estimate the temperature distribution automatically with the computer code presented or possibly by hand calculations by the method outlined. Experimental measurements of temperature distribution in electrically heated tube arrays in steel shells that simulated shipping casks were made to test the proposed calculational procedure. Several heat generation rates and bundles containing up to 64 tubes were tested in 12-in.- and 6-in,-inner diameter shells. Tests were made with the casks in horizontal and vertical positions. The predicted temperatures were very near those observed experimentally under the conditions in which heat transfer is likely to be a problem in fuel shipment, that is, when the temperatures are near or above 200 c- C and the casks do …

None

The neutron tissue dose at large distances from a fission source was studied by using a water-filled phantom and four different detectors: a BF/sub 3/ counter, a polyethylenelined ethylene-filled proportional counter, indium foils, and nuclear emulsions. The source of fission neutrons was the ORNL Health Physics Research Reactor which was attached to a hoist which was in turn installed on a 1530-foot tower. The reactor could be operated at any elevation from 27 to 1500 ft. The phantom studies were made at horizontal distances from 250 to 1500 yards from the tower. Dose contributions from recoil protons H/sup 1/ (n, gamma )D/sup 2/ and N/sup 14/(n,p)C/sup 1 reactions are considered. (auth)

The listing of a FORTRAN program for calculating the scattering of nucleons from a nonlocal optical potential is given. The mathematical formulation of the problem is presented, together with the numerical methods used in the code. The input to the program is explained, and a brief functional description of each subroutine of the code is included. (auth)

A total of 74 subsections are included in the report. The information in 4 subsections was previously abstracted in NSA. Separate abstracts were prepared for 38 of the subsections. Those sections for which no abstracts were prepared contain information on prompt neutron lifetime, Rover critical experiments, Pu/sup 239/ fission, neutron decay, the O5R code, alpha scattering, 8 and P wavelengths, proton scattering, deuteron scattering, local optical potentials, N. S. Savamah radiation leakage, reactor shielding, cross section data analysis, gamma transport, gamma energy deposition, gaussian integration, data interpolation, neutron scattering, neutron energy deposition, space vehicles, computer analyses, shielding, positron sources, and secondary particles. (J.R.D.)

None