Activities in a program to obtain data on thermal-stressfatigue in various stainless steels under reactor conditions are reported. Information concerning procurement of equipment and materials is given along results of initial testing. A specimen of 304 stainless steel was strain-rate tested at 100 to 900 deg C. The averaged result of this test was 0.0005 in./in.-sec. Tensile testing of standard ASTM 1/2-in.-dia stainless steel rods is underway. Data obtained so far in the program are compared with that obtained earlier by other investigators. (J.R.D.)

The SNAP-7C 10-watt Sr/sup 90/ thermoelectric generator, the converter, batteries, and weather station housing that were delivered to Antarctica in December 1961 are described. Thermoelectric analysis, thermal analysis, fuel form and shielding requirements, generator assembly, electrical system, operational tests, and environmental testing are discussed. (M.C.G.)

For the radioisotopes with half-lives over a year, only eight appear to be obtainable in the foreseeable future. The fission products, strontium-90, cesium-137, and promethium-147, exist in wastes from reactor processing, diluted with enormous volumes of other elements and salts. Among those isotopes producible by irradiation of special target materials (cobalt-60, uranium-232, plutonium-238, and curium-244) cobalt-60, though easy to produce, requires a special design for the heat source generator because much of its emitted energy is penetrating gamma radiation. Cobalt-60 appears, therefore, to be rather limited in its prospects for use. Plutonium-238 is favored as a heat source because of its long half-line and no need for special shielding. However, its projected high cost, scarcity, and biological hazard encourages the search for a competitive material. When plutonium assumes a significant place as a recycled fuel in thermal reactors for power production, curium-244 can then become available at costs below that for plutonium-238. Curiunm-244 has five times the specific power of plutonium-238 and appears to be just as easy to handle. Promethium-147, although probably on the ''short end'' of the half-life scale, can be considered for some uses as a substitute for plutonium-238. Although the factors of availability, gamma activity, and …

This report covers the following titles: (1) The Effects of 8-Methyl Lipoic Acid on the Evolution of Oxygen and Reduction of Carbon Dioxide during Photosynthesis; (2) Further {sup 14}C and {sup 15}N Tracer Studies of Amino Acid Synthesis during Photosynthesis by Chlorella Pyrenoidosa; (3) Two-Dimensional High Voltage, Low-Temperature Paper Electrophoresis of {sup 14}C-Labeled Products of Photosynthesis with {sup 14}CO{sub 2}; (4) A Search for Enzymic and Nonenzymic Reactions Between Thiamine Derivatives and Sugar Phosphates; (5) The Cytochrome Content of Purified Spinach Chloroplast Lamellae; (6) The Osmium Tetroxide Fixation of Chloroplast Lamellae; (7) Kinetics of Exoenzymes and Applications to the Determination of the Sequence of Nucleic Acids; (8) Brain Biochemistry and Behavior in Rats; (9) Experiments on Classical Conditioning and Light Habituation in Planarians; (10) Operant Conditioning in Planarians; (11) Manganese Porphyrin Complexes; (12) EPR Studies of Some Complex Organic Solutions; (13) Transient Response of Light-induced Photosynthetic Electron Paramagnetic Resonance Signals: Rhodospirillum rubrum Chromatophores; (14) Studies of the Tautomerism of Amides; (15) Structure and Mechanism of Hydrolysis of the Product of Reaction of PZ05 and Ethyl Ether; (16) A Study of the Irradiation Products of Several Nitrones; (17) Biosynthesis of the Opium Alkaloids; (18) Synthesis of methyl-{beta}-D-thiogalactoside-{sup 35}S; (19) Effect of …

From abstract: The temperature of molten uranium was measured indirectly by an infrared radiation pyrometer. The emissivities of graphite and molten uranium were determined, and a series of equations correlating pyrometer readings with melt temperatures was derived.

From abstract: A high speed motion picture study was conducted of boiling flow in a vertical, rectangular channel at atmospheric pressure. Three different visual flow regimes were defined and described.

This document details activities of the Irradiation Processing Department during the month of August, 1958. 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 report, from the Chemical Processing Department at HAPO for September 1963, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; weapons manufacturing operation; and power and crafts operation.

This document details activities of the N-Reactor Department during the month of September 1963.

This document details the process of casting plutonium for the purpose of reducing the in-process inventory. The plutonium is cast, then rough machined to produce a storable form, then stored for the ten-day waiting period.

Hot die sizing (HDS) is a process being considered at Hanford to replace or supplement the existing AlSi brazing process. Hot die sizing consists of passing a preheated core-component fuel assembly through a cold the to bond the aluminum jacket to the core while passing a die plug through the internal bore to form the internal bond. Fuel end bonding is accomplished in a following step by applying heat and pressure to the sized fuel element. This report summarizes the fabrication of fuel elements for irradiation testing of hot die sized fuel elements as authorized by ``Production Test IP-546-A, Irradiation of Hot Die Size Diffusion Bonded Fuel Elements,`` HW-75465.

The initial fuel landing for the N-Reactor and the use of a specific alternative fuel element are described in a previous report. It is highly probable that it will also be desirable in the future to irradiate still other types of fuel elements in the reactor. If the test is of small scale, fever than five channels, or if these elements vary only slightly from the initial fuel load (or any other fuel type approved for large-scale use at some future date), the irradiation will be accomplished within the normal administrative procedures and controls established for the Production Testing Program. In this context, in-reactor testing of minor changes in the fuel cladding alloy, alternate methods of affixing end closures, different fuel support designs or similar changes are considered to be within the scope of normal productiontesting. It is possible, however, that it will be desirable to irradiate test fuel elements which are significantly different from any element approved for general use and in quantities of greater than five fuel columns at a given time. A more formalized review of such tests including approval by the AEC would appear appropriate in these instances. The purpose of this document is to outline …

This memorandum discusses the problems associated with the production and use of Uranium 233 for an early test shot.

Due to sluggish transformation rates of metastable phases, unalloyed Pu continues to shrink after casting. Therefore, a pre-machining holding period is required to promote finished part dimensional stability. This test will determine the holding period commensurate with both this requirement for stability and an optimum material inventory level. Four storage periods will be evaluated (9, 8, 7, 6 days) for dimensional changes and related to parts stored for the standard 10-day period.

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 ad 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 frequency of occurrence of localized corrosion is influenced by the various modes of operation. This report discusses the results obtained from a 20 column fuel discharge, obtained prior to the start of the test.

Three overbore elements, 75Zl2, 75Zl4, and 75Zl5 were discharged from 3062C on April 16, 1962 with longitudinal splits after an exposure of 676. The three elements, two of which had completely split apart, were received on July 5, 1963 to determine the cause of failure. All three elements had failed in the same manner with several full length longitudinal cracks in the uranium extending radially from the canwall to about midway in the core. The failure mechanism was found to be the same as that observed during the examination of worm-tracked overbore elements under RM 456 in which incipient failures had occurred from excessive core temperatures. In two out of the three worm-tracked elements the interior core temperatures had exceeded the beta transformation point as a result of severe cycling conditions. In these three ruptures from 3062C the maximum core temperature had not reached the beta phase; however, the failures were undoubtedly caused by excessive hoop stresses resulting from high internal core temperatures.

A study was made to determine operating temperatures of old pile HCR sheaths. The calculations were performed with the aid of a FORTRAN coded program for the IBM 7090. The difficulty of determining the correct value for the contact coefficient between the HCR and the graphite channel at any point in the channel length resulted in performing the calculations in a parametric style. The independent system parameters were varied during the calculations. These parameters are diametral spacing between the poison rings and the outer aluminum sheath, heat generation rate or reactor power level, the heat transfer contact coefficient between the rod and the graphite channel, and the cooling water temperature or cooling flow rate. For contact coefficients between 200 and 700 Btu/hr -- ft{sup 2} -- {degree}F and for diametral clearances around the poison of 5 mils to 20 mils, the maximum sheath temperature was calculated to vary between 350{degree}F and 650{degree}F. The dependence of sheath temperature upon the rod coolant temperature or coolant flow rate was found to be small enough to be neglected in the normal flow range of 12 gpm to 15 gpm. For a given increase of the average coolant temperature, the maximum sheath temperature increased …

The purpose of this study was to determine the maximum forces that can be imposed by the graphite moderator on prospective VSR channel sleeves. In order to do this, both the origins and modes of transmission of the forces were determined. Forces in the moderator stack that are capable of acting on a block or group of blocks may originate from any of the following primary effects: Contraction of graphite due to irradiation; thermal expansion of graphite; frictional resistance to motion; resistance from keys; gravity; and other.

The purpose of this letter is to outline the immediate and long-range plans for increasing the H Reactor coolant flow. The flow can be increased, immediately from the present base of about 84,000 gpm to about 88,000 gpm via the following changes: routine purges; use of non-bumpered fuel; increasing the flow through Fringe enrichment and Fringe Poison tubes by changing orifices; and reaming damaged rear fittings to their original size. The flow can be increased to about 94,000 gpm via the following changes: installing larger Venturis throughout the reactor; reaming rear fittings to a larger size; installing rear nozzles that permit higher tube flows; charging I&E Poison elements in Fringe tubes; and some changes in the water pumping facilities may be necessary.

None

The always critically-safe ratio of plutonium to enriched uranium was exceeded in the F-2 and V-1 vessels on June 6, 1963. The allowable always-safe ratio at the time of the incident vas 627 grams per ton. The actual ratio achieved vas 673 grams per ton. This report details the investigation into events leading to the over-batch.

A half-plant low alum water treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by the analysis of ledge corrosion attack on fuel elements, 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 the normal production fuel irradiated in process water treated with two different alum feed rates. Four, 20 column fuel discharges, ten columns from each side of the reactor, have been taken during the test as follows: (1) One discharge prior to the start of the test. (2) One discharge such that the test side was exposed to coolant treated with both high and low alum feed. (3) Two discharges under test conditions. This report discusses the results obtained from the third discharge under test conditions.

In the summer of 1962, a brief scoping study of the N-Reactor`s capability as a producer of tritium concurrently with plutonium and steam was completed and documented. This study indicated that there might be considerable incentive in pursuing the development of a multi-product mode of operation of the N-Reactor. Early in 1963 reports presented the preliminary production and cost data, and brief discussions of the potential technical problems which mav arise in a multi-product mode of operation. On September 9, 1963, a report presented a program for the conversion of the N-Reactor to the concurrent production of weapon grade (6%) plutonium and substantial quantities of tritium. The production and cost data showed that the amount of tritium can be varied over a wide range and that weapon grade plutonium can be economically generated. The purpose of this report is to evaluate the feasibility and incentive for operation of the N-Reactor with fully enriched fuel producing only two major products tritium and steam. Neptumium 237 is an important by-product.

The activities included the single tube fuel element study, production losses encountered upon simultaneous irradiation of Np{sup 237}target material, startup measurements of lattice parameters and spectral indices, neutron lifetime in N-Reactor, production estimates for tritium only, FLEX code, etc.