This report is the final in a series of preliminary reports that follow the studies of scavenging systems related to radioactive fallout. The project consisted of two phases: preliminary experiments to relate the sizes of particles in air to specific radioisotopes, and preliminary laboratory studies of scavenging of particles by liquid drops, including studies of sticking probability and effects of Brownian motion and water vapor diffusion.

Abstract: "The niobium-oxygen equilibrium has been determined by metallographic examination of arc-cast alloys made of electron-gun-refined niobium metal and special purity niobium pentoxide. Two intermediate oxides, NbO and NbO2, melt without decomposition at 1945 C and 1915 C, respectively. Eutectic reactions exist between niobium and NbO and 1915 C and between NbO and NbO2 at 1810 C. Experimental supports a peritectic reaction between NbO2 and Nb2O5 1510 C. The maximum solid solubility of oxygen in niobium metal is 0.72 weight per cent."

This is the final report on the progress, "A Wide-Range Neutron Detector and Monitoring Instrument," ARF 1118-6. This report covers activities from May 8, 1958 to May 8, 1959. This report describes the basic design principles of a proportional counter that has been constructed whose cathode diameter increases towards one end.

"This is a selected bibliography of books, journal articles and reports published on creep and relaxation in metals and ceramics from 1950 to March 1959. Particular attention has been paid to studies on or applicable to the consideration of creep and relaxation in beryllium oxide. Emphasis has alro been placed on studies of phenomenological materials behavior and high temperature studies."

The apparatus and procedures that are to be used to fill the liquid metal system of the Pratt and Whitney Forced Convection Liquid Metal Inpile Loop are described. The liquid metal to be used is a mixture containing 56% Na and 44% K, which is a liquid at room temperature. In order to simplify the filling procedure at the reactor site, two containers, each of which contains exactly enough liquid metal to fill the experiment to the prescribed level, a fill dolly incorporating a purified helium system, and an evacuation system will be provided at the reactor site. After completion of the liquid metal transfer, the liquid metal fill tube will be crimped, cut, and seal welded. The liquid metal and helium systems will then be pressurized through the helium fill tubes, which will also be crimped, cut, and seal welded. Each tube seal weld will be inspected after completion of the weld by mass spectrometry leak check and dye penetrant fault detection

Technical report. From Abstract : "Predictions of the behavior of technetium in the Purex process and details of the chemistry of technetium for Purex process conditions are given."

Introduction: Seldom does the actual arrangement of control elements in a nuclear reactor confers to the ideal and convenient mathematical array. In order to achieve shim control. safety and regulation, it is desirable to design with rods of different sizes and materials. With given fuel element arrangement, typically in square or hexagonal lattice spacing, there will be rods located at different distances form the center of the core and from each other. As the reactor operates, absorbers will be withdrawn, leaving further asymmetries in the location of those remaining. It is the purpose of this report to develop in detail a two-group diffusion theory with as complete generality as possible. The method is as yet restricted to the unreflected core, or to the reflected core by use of reflector savings and bare equivalent geometries.

A method is presented for measuring the average energy of the neutrons from a source. The attenuation of the neutrons by polyethylene is measured by the use of a long counter in good geometry. The attenuation length is a sensitive function of the neutron energy. The average neutron energies from several (o, n) sources have been measured and agree well with values obtained by other techniques.

With the increasing availability of accelerators capable of producing beams of heavy ions it has become useful to know the cross section for a heavy ion to form a compound nucleus by interaction with a target nucleus. The results of calculations based on two simple models are presented here in order to give some idea of the magnitude of this cross section as a function of the energy of the bombarding particle.

This report was prepared at the request of the Subcommittee on Radiochemistry of the Committee on Nuclear Science of the National Research Council as a contribution to a proposed master file on the radiochemistry of all the elements. All known material on the radiochemistry of thorium is included in this report.

The build up of contaminated film on the internal surfaces of high temperature in-reactor recirculating water loops has created serious radiation exposure problems to operational and maintenance personnel. A considerable amount of work has been applied to develop an effective decontamination process for the decontamination of these loops and their components.

The Atomic Energy Commission has recently developed an enlarged program for the study of graphite moderated, gas cooled power reactor systems. It has been recognized, however, that understanding of radiation damage and radiation induced chemical reactions of graphite at the proposed high moderator temperatures is inadequate and that improved understanding is essential if the design of such reactors is to be optimized. Accordingly, the Atomic Energy Commission requested Hanford to organize a modest research and development program directed toward the study of these graphite problems.

The temperature characteristics associated with the handling of self-boiling radioactive wastes from the separations extraction processes in the Chemical Processing Department have necessitated several heat transfer studies. Earlier studies 1,2,3 defined the feasibility of self-concentration in existing waste storage facilities by determining the rate of heat generation from the decay of stored fission products and by defining the rate of heat loss from existing storage tanks to the surrounding soil.

Laboratory studies involving fluid flow through porous media require use of fluids having low dissolved gas content. Water is the major fluid used in various and box model and soil permeability studies carried out by the Geochemical and Geophysical Research group. Tap water supplied to the 222-U Bldg. contains a large amount of dissolved air. Under the reduced pressure encountered during model studies, the air is released from solution and gradually clogs the pores of the sand or other porous material. This, of course. leads to anomalous results and cannot be tolerated in precious studies. A system was required to effectively remove the air and make available a continuous supply of desired water for the model studies.

"The PRTR Pressured Gas-Cooled Loop Facility, or Gas Loop, is an experimental facility to be installed in the Plutonium Recycle Test Reactor for use in studies contributing to advancement of the technology of gas-cooled reactors. The facility will provide an in-reactor loop for studying phenomena occurring under conditions likely to exist in gas-cooled reactors.

Several hollow uranium cores of "C" size I & E diameters were fabricated by cold extrusion (550 to 750 F) at Hunter Douglas Aluminum Corporation. Results show diameter control and reproducibility are excellent. Preferred orientation induced by this process is completely removed by a single standard beta heat treatment.

This report was written to describe briefly several possible variations of such systems along with probable cost estimates. Previous work has been shown that the application of scintillation detectors is the simplest and most reliable means for such monitoring. By proper detector selection and measuring methods, the system can be of an approximate gamma dose-rate with gamma energy independence above about 100 Kev. The same system using detector change only is directly applicable to beta, gamma, slow neutron and fast neutron detection. Any type of alarming device desired can be incorporated.

At the request of 234-5 Development, studies of a system for controlling plutonium powder transfer were undertaken by Process Equipment Development. This report presents details of equipment designed to fulfill requirements and proposes equipment for installation on additional Hot Button Line prototypes.

In conjunction with the fuel element development program at Hanford, it is desired to determine the effects of cladding and core temperatures, cladding thickness, and exposure upon the swelling behavior of unalloyed uranium. To obtain this information, it is proposed to irradiate several fuel rods, clad by coextrusion with Zr-2, in NeK filled stainless steel capsules. The central uranium temperatures are to be monitored by axial thermocouples. Irradiation tests in the MTR and ETR using capsules of similar design have been and are now being conducted. GKH 3-31, a fuel rod, clad with 0.030" Zr-2, operated in the MTR at an average center fuel temperature of 425 C to an exposure of 2100 MWD/T. GKH 3-57, 3-58, and 3-59 are presently undergoing irradiation in the ETR.

A previous study has been completed on the buildup and decay of xenon poisoning in the PRTR by simulating the reactor as a single region. The results of this study indicated that a study using a more refined model of the reactor would be valuable.

The ambient temperature of the Columbia River is of interest to the Aquatic Biology Operation since it provides a basis for controlling the temperatures in various experimental tanks which contain Columbia River organisms and since it provides some indication of whether conditions are favorable for the valuable species of fish living in the river. Since the start-up of the plant, temperatures of the river water have been taken in the several water treatment plants.

Basic Studies. It has been reported previously that mixed crystals of PuO2 and UO2 have a higher sintering rate than UO2 alone. However, results to the contrary were obtained on addition to 1/4, 1/2, 1, 2, 5, and 10 w/o PuO2 to PWR grade UO2. Pellets of the above concentrations were heated for one hour in hydrogen at 100 C intervals from 1000 to 1600C.

Crib evaluation studies showed the soil uptake of plutonium from the Recuplex process (CAW) waste to be low. Preliminary studies of the low soil adsorption of plutonium revealed the presence of a plutonium nitrate anion complex which could be removed by adsorption on a strong base anion exchange resin.

Thermal Resistance of the contact between to core and the jackets or unbonded fuel elements may easily be the largest source of error in core temperature predictions. The object of this work is to improve contact conductance predictions by measuring conductance of the joint between reactor fuel and cladding materials at joint pressures, temperatures and thermal flux levels approaching reactor service conditions.