This document summarizes the amounts of radioactive contamination discharged to ground from separations facilities through December 1962. Detailed data for individual disposal sites are presented on a month-to-month basis for the period of January through December 1962. Previous publications of this series are listed in the bibliography and may be referred to for specific information on measurements and radioactivity totals prior to December 1962. Tables list the major disposal sites in the separation facilities, total volume of waste discharged to each location, and the gross amounts of plutonium and beta particle emitters discharged to ground since startup. This same data is presented on a monthly basis for cribs still in use. Information is presented on the source of the waste stream and the settling facility if used. Isotopic data are included for disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to swamps are also included.

The ``machined parts`` casks are shipped to Hanford by Savannah River for loading with irradiated lithium alloy slugs and returned for processing. The casks have been in service for about 10 years and have shielding material more than what is required by regulations for radiation attenuation. Evaluation shows that rupture of the shell is probable for edge or corner drop, with considerable displacement of shielding. The cask would fail in a 2-hour fire of 2000 F with loss of shielding and probable dispersal of the product. Bolts securing cover to cask are inadequate in event of accident.

Following the startup of 11-6-63 at 105-B, there was an operating period where graphite stringer temperatures were above allowable, and a PCA was authorized for continued operation. It is probable that the combined effects of raising the CO{sub 2} rapidly along with the fast approach to full power level caused the temperature effects. The CO{sub 2} reached 67%. As an interim step, it is suggested that startup CO{sub 2} additions be tailored to level off at 5--10% below the equilibrium level, and that resulting graphite temperatures limit the power level accordingly.

With the use of smooth-bore Zircaloy-2 process tubes and self- supported fuel elements on a large scale at the K Reactors, establishment of the operating characteristics of this fuel element-process tube configuration is imperative. Under authority of Production Test-IP-409-A, two smooth-bore Zircaloy-2 tubes were routinely charged with KVNS fuel elements. Approximately 520 KVNS fuel elements have been irradiated in this facility. In Production Test-IP-572-A, the coolant temperature distribution in the annulus of the KVNS fuel element-process tube system is being evaluated as a function of fuel element support height. This production test supplement will provide data for further defining the operational characteristics of the self-support fuel element concept.

The objective of this test is to produce 50 milligrams of neptunium 239 for a process stream study in the Purex plant. The efficient recovery of neptunium 237 as an alternate product produced at HAPO is presently impaired by the lack of knowledge of the distribution of neptunium 237 in the several Purex process streams. Present methods of following the neptunium 237 through the process streams involves inadequate alpha counting techniques. However, neptunium 239, because of its discrete gamma decay energy, can be readily identified in the process stream. As a result, the neptunium 239 produced by this irradiation will serve as a gamma-emitting tracer for the neptunium 237 in the Purex process stream.

This document defines the objectives, bases, and functional requirements that shall govern the preparation of detail design of the gravity fed water supply to reactor storage basins for all eight reactors. In the event that appears advisable and feasible to discharge all metal from the reactors into the metal storage basins, it would be necessary to add water to the storage basins to prevent overheating of the fuel elements. Without the addition of cool water the storage basin water would soon start to boil and evaporate, eventually exposing the metal to the air. Existing facilities do not permit assurance that sufficient water can be added to the storage basins for the required period of time to protect a complete discharge of fuel elements in the storage basin if the area is left unattended and pumps are shut down. A positive gravity fed system to the metal storage basins from existing supplies of stored water shall be provided by this project. This new gravity fed system, once it is started, shall operate unattended and shall supply adequate water to the storage basins for the required period of time. It shall not be dependent on electric or steam-driven pumps for its continuous …

The growing concern of government agencies for safe shipment of radioactive materials, both in this country and in Europe, has resulted in a number of proposed regulations governing every phase of shipment including the design of the cask or container, There have been requests during the last year for increased shipments of irradiated bismuth from Hanford. This radioactive material has been shipped in the B-cask to the off-site processing plant for several years. These two factors, proposed regulations and increased demand for the product, have created a need for an engineering evaluation of the cask currently being used. While it is evident on casual inspection that the B-cask design would not meet certain requirements of the proposed regulations, it is necessary that a detailed study of the design be made to determine if modifications to the cask are practical for continued use under new regulations and whether additional casks should be of a new design as required by increased shipments of the product. The purpose of this design evaluation study is to: (a) Evaluate the integrity of the B-cask design in accordance with criteria published in document HW-76476, which is based primarily upon proposed regulations of the Atomic Energy Commission …

The discharge of radioactivity from Hanford production reactors to the Columbia River has been studied for a number of years. The primary purposes of these studied have been to learn of the factors influencing production reactor effluent radioactivity concentrations and to develop methods for lowering the discharge concentrations of these radionuclides. In 1962 the Coolant Systems Development Operation (CSDO) began work on this program to determine the in-reactor film compositions and the inventory of radionuclides held in these films with the long-term objective of learning more about the complex in-reactor retention reactions that influence radionuclide effluent concentrations.

Both the inner and outer tubes of the N-Reactor fuel elements will have self supports spot welded to the lateral heat-transfer surface of the element. A crevice a few mils thick will exist around the weld between the support tab and the cladding. Because of the heat flux through the cladding at this point and the insulating effect of the support tab, the temperature in this crevice will be higher than that on the free surface away from the support. This can result in boiling in the crevice leading to concentration of LiOH (or impurities in the water) to a level where it can cause severe corrosion of the Zircaloy-2 cladding. The tests described in this report were conducted to determine whether such attack might be encountered in N-Reactor.

Equipment Laboratory and Testing Operation was requested to measure the pressure gradient developed across a nozzle and crossheader assembly used on the rear face during reactor operation. A nozzle and pigtail assembly with pressure taps was installed on process tube 2058, and pressure taps were located at the center, at tube 2058, and at the end of crossheader No. 19-1/2. copper sensing lines were installed between the pressure tape and a tool storage area on zero level, near side of the reactor where our gage could be safely located.

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Self-limiting power excursion tests were made to investigate the dynamic response of the Spert II BD-22/24 expanded D/sub 2/O core. This D/sub 2/O- moderated core is comprised of 3-in.-square x 24-in.-long, highly enriched, uranium-aluminum, plate-type fuel assemblies spaced on 6-in. centers. The power excursions were initiated by stepwise reactivity insertions with the system at ambient temperature and atmospheric pressure, with no forced coolant flow, and with approximately a 50-in. D/sub 2/O head over the core. The reactivity insertions ranged from about 40 cents to 3 dollars and resulted in initial asymptotic reactor periods of from about 10 sec to 50 msec. The basic self- shutdown mechanisms are apparently thermal in nature. For initial periods longer than 600 msec, the self-shutdown is due to nonboiling processes. For initial periods less than 600 msec, steam formation contributes to the self-shutdown and is apparently the predominant shutdown mechanism for power excursions with periods less than 100 msec. Limited mechanical damage to fuel plates is not obtained in tests with periods greater than 70 msec. Extrapolation of the fuel- plate surface-temperature data indicates that partial melting of the fuel plates might occur in tests with periods of 30 msec or less. The high fuel …

Experimental work consisted of one new run started and completed this quarter, and the completion of one of the three continuing runs. Of the scheduled operating time, the beam was on for 69.4% of the time, 2.3% of the time was used for experimental setup, and equipment outage took 29.3% of the time. There were two scheduled and two impromptu shutdowns. During one of the scheduled shutdowns the external-beam extraction magnets were installed in the east and south tangent tanks. The other scheduled shutdown was to readjust the Bevatron magnet elevation to correct for foundation subsidence. Internal magnets were also installed. In the new linac development program the ion source was run at 480 kv with a beam current of 100 ma. The linac tank was partially deplated to provide a clean copper surface, and welds and holes were plated with copper. The r-f losses were thereby reduced 20%. (auth)

The design of the High-Flux Isotope Reactor (HFIR) was supported by a series of preliminary experiments performed at the Oak Ridge Critical Experiments Facility in 1960. The experiments yielded results describing directly some of the expected performance characteristics of the reactor and strengthened the calculational methods used in its design. The critical assembly, like the reactor, was of a flux-trap type in which a central 6-in.-dia column of H/sub 2/O was surrounded by an annulus of fissile material and, in turn, by an annular neutron reflector. The fuel region contained a solution of enriched uranyl nitrate in a mixture of H/sub 2/O and D/sub 2/O and the reflector was a composite of two annuli, the inner one of D/sub 2/O surrounded by one of H/sub 2/O. In most experiments the ends of the assembly were reflected by H/sub 2/O. Important results evaluate the absolute thermal-neutron flux to be expected in the design reactor and describe the flux distributions within this type of assembly. It was also observed that the cadmium ratio along the axis of the assembly was about 100, showing that a highly thermal-neutron flux was truly developed in the trap. It was shown that reduction of the hydrogen …

An investigation was conducted to determine the cause of cracking, during tension, on the surface of tensile specimens of U--Mo alloys. The cracking was observed in alloys that were water quenched from a temperature range in which the body-centered cubic gamma phase was stable and, subsequently, stressed in tension. The investigation was carried out on alloys of U--Mo containing 5 and 4 wt% Mo. It was discovered early in the study that the cracking could be eliminated by testing tensile specimens in a He atmosphere. Because of this result, tensile tests were performed in vacuum O/sub 2/2, impure N/sub 2/, and purified N/sub 2/. Additional experiments were done to establish the relation of stress state, temperature, strain rate, O/sub 2/ pressure, stress magnitude, and grain size to the cracking phenomenon. U--4 wt% Mo and U--5 wt% Mo were compared with respect to microstructure and crystal structure, since both alloys cracked during the tensile test. The origin and propagation of cracks in the microstructure were studied, using motion pictures and metallography. The results of the various experiments showed that the cracking occurred only in tension and was caused by stress-corrosion cracking due to the presence of O/sub 2/ in the atmosphere. …

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Experiments to obtain detailed thermal flux distribution measurements within a unit fuel cell and measurements of sub-cadmium and epi-cadmium U/sup 238/ absorption in the fuel rods are described. These quantities are required for determination of the thermal utilization and resonance escape probability of the fuel loading. (auth)

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.

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For additional details see report LA-2788. Thermal expansion characteristics for arc melted plutonium-carbon alloys containing 41.1, 43.2, 44.8, 45.4, 45.8 at.% carbon and for U/sub 0.87/Pu/sub 0.13/C have been determined between room temperature and 900 deg C. The average coefflcients of linear thermal expansion for a singie phase alloy conthining 45.8 at.% carbon and for single phase U/sub 0.87/Pu/sub 0.13/C are presented as well as results pertaining to the plutonium-carbon phase diagram. A mechanism has been suggested to explain the unusual contraction that was observed in the dilatometric curves for the alloys that contained 43.2, 44.9, and 45.4 at.% carbon. (auth)

This document presents the initial definition of a 2 Mwt, SNAP-50 Reactor Test and is intended and is intended to provide guidance to the Architect/Engineer in evaluating and costing various Nuclear Test Facility design concepts. The information is based on preliminary studies; and, consequently, the configuration and component sizes have not been firmly established.