It is desirable that a reactor exhibit a self-regulating effect. If this were not true any disturbance to the reactor would result in a continual increase in the magnitude of the disturbance and the reactor would be unstable. In this investigation the reactor is considered to have two reactivity feed-backs: metal temperature and water temperature reactivity effects. These two variables through a metal temperature coefficient and water temperature coefficient determine not only the reactor stability but also determine many operational characteristics.

Zirflex cladding waste is to be neutralized to pH 10 before transfer to waste storage tanks. This treatment causes the precipitation of zirconium oxide or hydroxide, which may lead to flow difficulties during transfer. The purpose of this investigation was to determine the physical properties and flow characteristics of the neutralized slurry to assist in the selectin of satisfactory transfer equipment and storage conditions.

The purpose of the Analytical Laboratory Manual is to assemble the basic procedures to be used for the analyses of materials employed within the Fuels Preparation Department. The methods appear in detailed steps suitable for laboratory use. This document replaces the "Essential Material Analytical Manual, " HW-25375 and "Metal Preparation Analytical Manual," HW-30862.

This document describes required laboratory space and lists major equipment items necessary for a routine water quality laboratory in the P. R. T. R. Building. During discussions with R. D. Widrig and V. L. Rooney about the analytical sample program for the Plutonium Recycle Test Reactor, the author was asked to summarize equipment and space needs for a water control laboratory to provide routine analytical coverage on some of the water systems. Based upon 1706-KE-KER experience, some operating personnel may be used to provide analytical coverage on those routine analyses that are needed on around-the-clock basis with a savings of both time and money.

One of the UO2 fuel element designs proposed for use in the PTRT is the nested tubular concepts. This configuration compromises a central fuel todo surrounded by two concentric tubes of fuel (see sketch, appenx I.) . These UO2 shapes are to be jacketed in the Zircaloy and must be separated from each other and the procuresses tyvm vt annular spaces for the passage of coolant. The annuli are established and and maintained by the longitudinal ribs on the outer surface of all three jacketed fuel elements.

Two water quality control labs are being provided for the the NPR. One, a "cold" lab, is located in the183 Building adjacent to the control room. Its primary purpose is to provide facilities for quality control of the output of the filter plant and the demineralizer plant. The other, a "hot" lab, is located in the 190 Building. Its primary purpose is to provide facilities for the quality control of the primary and secondary coolants, and the moderator coolant.

Early in the U.S. gas-cooled reactor program, it was recognized that a considerable research and development effort would be needed to utilize the full potential of gas cooling. Hanford Laboratories contribution to this program began in August, 1958 when a graphite studies program was initiated. The primary objectives of the Hanford program are: (1) to secure improved knowledge of the reaction of gas coolants with protected and unprotected graphites over a range of conditions of practical interest for gas cooled power reactors with tolerable damage to the graphite moderator.

Photographic observations of the behavior of uranium dioxide at high temperatures are of great value in designing and evaluating fuel elements. This paper reports the growth of uranium dioxide crystals by vapor deposition during out-of-reactor and in-reactor experiments.

It is well known that neutron damage generally causes increases in the yield and ultimate strength and a decrease in ductility of a metal. There is a continuing program at HAPO to determine the extent of these changes in Zircaloy-2 as functions of integrated neutron exposure, irradiation temperature, and reactor atmosphere. Three investigations from this program will be described and the results summarized. The first investigation deals with both annealed and cold worked Zircaloy-2 irradiated at approximately 50 C. and the other two investigations deal with annealed Zircaloy-2 irradiated at approximately 100 and 280 C respectively. In each investigation tensile testing was performed at room temperature.

The purpose of this program is to determine the effects of neutron radiation and reactor gaseous atmospheres on the structural integrity of candidate materials for the PRTR gas-cooled loop. An exploratory investigation is being conducted to select a high temperature resistant alloy for the fabrication of tubes and supporting appliances for the PRTR gas-cooled loop. In addition to a low neutron cross section, the material must be able to withstand the operating conditions outlined in Table 1.

Disposal on a specific retention basic of certain radioactive liquid waste solutions emanating from separations from separations plants has been practiced at Hanford since 1944. As used at Hanford, the term "specific retention" is defined as that volume of waste liquids that may be disposed to the soil* and be held against the force of gravity by the molecular attraction between sand grains and the surface tension of the water, when expressed as percent of packed soil volume. In practice it represents the volume of liquid that may be discharged to a disposal pit of known dimensions without leakage to the ground, water, expressed as percent of the total volume of a column of soil with the same cross section as the pit, and extending from the bottom of the pit to the water table. It is recognized that some degree of lateral spreading will occur which has the effect of enlarging the volume of soil contacted by the liquid.

The measurement of a alpha-particle energy has been used by many radiochemical laboratories for the identification and analysis of alpha-active radio nuclides. The use of the total-ionization method for alpha-active radio-nuclides. The use of the total-ionization method for alpha energy in ionization chamber in which the alpha particle loses all its energy in ionization of the chamber gas. Collection of the electrons thus formed generates a voltage pulse across the chamber capacity which is proportional to the alpha particle energy. This pulse is then amplified using a suitable linear amplifier and fed to a pulses as to amplitude; the information is then recorded or stored. Since the pulse amplitude is proportional to the alpha energy lost to the chamber gas, the pulse height analysis can be used to estimate the energy of the alpha particles and in the case of several alpha emitters of different energies, the relative abundance of the alpha emitters can be determined. An alpha energy analyzer system using the ion collection method has been fabricated for use in radiochemical laboratories required to perform a large number of alpha energy determinations. This report describes the operation, maintenance, and application of this alpha energy analyzer system.

Judging scratch depth or surface roughness by unaided visual inspection under controlled conditions, while rapid and popular, is not quantitative. Comparison methods improve reproducibility but are generally not applicable to evaluation of depths of single widely spaced scratches. Stylus-type contour recorders yield valuable scratch contour data but may themselves plow through soft materials and fine details. Depth measuring microscopes are particularly applicable to measurement of pinhole depth but do not graphically reveal profiles and provide only a small field of view. The comparatively large field of view and graphic display of contour provided by profile microscopes make them particularly suitable for evaluation scratch depth as well as surface roughness. A HAPO-constructed instrument has demonstrated an accuracy of +/- 50 micro inches in the range of 50 to 15,000 micro-inches scratch depth. It is a pocket-sized, portable, and can be used on horizontal and vertical surfaces by untrained persons with only brief instruction.

At the request of the Hanford Operations Office of the Atomic Energy Commission, a criticality instrumentation meeting was held at Hanford on August 17 and 18, 1959. The purposes of the meeting were: (a) to review types of criticality control and/or alarm instrumentation at each site, their problems, shortcomings and potential obsolescence; (b) to discuss instrumentation in the area if approaching criticality; and (c) to consider for material balance purposes, the feasibility of devising instruments to detect the buildup of plutonium and enriched uranium bearing residues in process systems.

The region in which Hanford's chemical processing plant areas (separation areas) are located in the central portion of the Hanford Works area on a plateau, shaped like a giant bar, of glaciofluviatile materials. The general locations of these areas is shown in Figure 1. The region studied rises from 200 to 300 feet above the surrounding area and encompasses approximately 40 square miles or somewhat less than 8% of the Hanford reservation (Figure 2). Within the region are located the chemical separation plants and most of the major waste disposal sites.

In solvent extraction purification processes such as are used at Hanford, the fuel elements or "slugs" from the reactor containing uranium, plutonium, and fission products are dissolved in nitric acid, adjusted to the required feed composition, and pumped to the solvent extraction columns. Figure 1 in a schematic diagram of such a solvent extraction process. In the A column, the uranium and the plutonium are extracted into an organic phase while the bulk of the fission products remain in the aqueous phase and leave as waste with the column raffinate.

In HAPO solvent extraction processes there are two independent aspects of nitrous acid chemistry. One concern the decomposition of the solvent through nitration reactions and the attendant problems. These reactions are autocatalytic in the presence of nitric acid and have threshold values for both temperature and nitric acid concentration for a given solvent below which nitrous acid disappears and above which it is generated with continuous destruction of the solvent. These reactions are identical to those found in the prior study of the hexone system.

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 purpose of this report is to show the impedance of an approximate equivalent circuit on a complex plan for various constant frequencies and different metals. The equivalent circuit represents a testing coil driven with an AC voltage and a network to represent the impedance of the metal.

Extraction systems involving chelating agents are widely used in radiochemical separations on a control basis due to the economy in time and equipment. In the case of waste streams, the element of interest is, by definition, at a very low concentration in the presence of gross fission product activities. The resultant large sample size necessary for a reliable analytical results creates a marked radiation hazard to laboratory personnel.

Aluminum nitrate solutions are used as salting agents in the redox process and frequently in organic-aqueous extractions in the laboratory. Determination of specific gravity is a convenient means of estimating the concentration of such solutions. However, where an aluminum nitrate solution has been prepared by dissolving aluminum hydrate in nitric acid, the ratio of nitrate to aluminum may intentionally be other than stoichiometric, the solutions being described as acid, neutral, or acid deficient.

In support of the Plutonium Recycle Program at Hanford, the properties of PuO2 are being investigated. This paper reports sintering studies on PuO2 and PuO2-UO2 mixtures and solid solutions; melting studies on PuO2 and UO2; thermal expansion data to 1000 C; and irradiation data on PuO2-UO2 mixtures. The existence of a continuous series of solid solutions formed during sintering has been established for the system UO2-PuO2. A linear relation between lattice parameter and composition exists. Extensive sinterability data on the isomorphous compound UO2 and PuO2 and their intermediate compositions has been obtained. The effect on sintered density of temperature, time, and concentration for both physical additions of PuO2 and additions of UO2-PuO2 solid solutions to UO2 have been studied. Evaluation data are supported by microscopic examination and X-ray diffraction.

Whenever fissile materials are handled in significant quantities such as in fuel element fabrication, separation processes, or in exponential and/or critical experiments a potential criticality hazard exists. The usual procedure which is followed by those persons conducting critical mass experiments is to either place the potential reactor in a heavily shielded cell or to conduct the experiments remotely in which case distance provides a measure of safety in the event of an unscheduled radiation outburst. In considering potential critically incidents, especially for the personnel not specifically engaged in critical mass studies, it is very likely that at the time of the incident neither the conditions of shielding nor distance will prevail for the personnel involved.

The processing of irradiated fuels from power and propulsion reactors is planned by Hanford Atomic Products Operation as part of the Atomic Energy Commission's interim reprocessing scheme. The several chemical processes proposed for the reprocessing of these fuels may be divided into two categories: first, total dissolution processes which dissolve both fuel element cladding and core material and, second, selective dissolution processes which dissolve either cladding or core material. The Niflex process uses a nitric and hydrofluoric acid solution for the total dissolution of stainless steel clad uranium fuel elements.