Aid in redesign of the calibration facility in order to: 1. Reduce the rate of exposure for calibration personnel below 1 mr/hr; 2. Reduce the rate of exposure outside of the 3745 Building; 3. Reduce the risk of breaking radium sources; 4. Improve the efficiency of calibration procedures by: a. Allowing calibration to be done faster; b. Providing several calibration ranges which can operate simultaneously in the present space allowed; c. Allowing instrument survey work to proceed while calibration work is done. A well-type calibration installation is proposed and a brief description of the required shielding is included.

Porous alundum filter plate samples were received for testing to determine the effects of Redox process solutions and conditions on bend and impact properties. Alundum has been suggested as a barrier material for use in the Redox Silica Gel Absorption Column T-101. Consequently, data on possible engineering property changes caused by exposure to separation process solutions and radiation are needed to evaluate the applicability of alundum as a barrier material.

In the following report the effective radii of infinitely long cylindrical control rods of various radii and blacknesses have been calculated. The results are given in a graph at the end of the report.

The preparation, purification and chemical properties of TBP have been reviewed with emphasis on the hydrolytic reactions. The latter is supplemented by data on other alkyl phosphates.

One of the problems associated with evaluating the hazards resulting from the discharge of radioactive pile coolant water into the Columbia River results from the subsequent consumption of the river water by cities downstream of Hanford. Determination of the travel time of the river water from the reactor cooling water discharge points to the first downstream cities to consume the river water (Pasco and Kennevick) is a necessary part of the hazard evaluation. Knowledge of the travel time is required to establish decay correction factors for river samples collected in the Regional Monitoring Program and to determine the correct sampling time at downstream locations for samples that may represent specific Hanford operating conditions.

In the Purex process, nitric acid is recovered from aqueous wastes by concentration and rectification. The concentrator bottoms are relatively corrosive to stainless steels, particularly on heat-exchange surfaces. This report summarizes studies aimed at determining corrosion rates for certain candidate materials of construction (304-L, 347, 312 stain steels and A-70 titanium) for the aqueous waste concentrators. Synthetic Purex Process 2WW, representing bottoms from the low level waste concentrator, was used as the corroding medium. Purex Process flow sheet changes made since this work was started have resulted in some changes in the compositions of the aqueous wastes. Consequently, data reported here are representative of, but perhaps not precisely, those which would be obtained using the present 2WW composition.

The special x-ray photometer was designed to measure various characteristics of high-absorption materials by comparing the attenuation of an x- ray beam in passing through a sample of known size with the attenuation through a standard. A general description, discussion of elements, safety features and cautions alignment and adjustment, and operation and maintenance are presented.

Dimensional instability was one of the first difficulties encountered with uranium fuel elements. Early in the operation of the HAPO reactors it was found that uranium fuel elements changed shape as a result of having been irradiated. This problem of dimensional instability has been vigorously pursued up to the present day.

The scintillation alpha counter was further developed to produce better long-term reliability and ease of maintenance.

A new method to process uranyl nitrate solutions for the production of useable compounds for subsequent reduction to metal is suggested.

In the Columbia River downstream from Richland, Washington there is a gradual decrease in the concentration of radioactive isotopes. At any one location there are fluctuations due to factors such as ditution and decay time. During this period at all locations the concentration of alpha particle emitters in water averaged below the detection limit of 3 x 10⁻⁹ μ c/gm. Beta particle emitters in water upstream of McNary Dam increased by a factor of 1.5 to 2, with averages at various locations ranging from (3 to 40) x 10⁻⁷, but in river mud the average of 2 x 10⁻⁵ μc/gm was about 2/3 that during the previous quarter. Downstream from McNary Dam, beta particle emitter average concentrations in water were about the same as those during the past three quarters.

As a result of the interest in fission products for use as heat sources and/or radiation sources, in early 1960. Hanford was authorized to supply a semi-refined cerium-144 rare earth product to Oak Ridge National Laboratories. Facilities were installed for collecting this product and a cask war designed for shipping this product as a stabilized, dry powder. Since that time, numerous improvements have been made in the design of the HAPO I cask as described in Revision O of this document. In addition, the development of strontium-90 as a major product requires increased flexibility in the utilization of these casks.

Entrained solids have in the past been one of the major factors limiting solvent extraction fission product decontamination. These solids are known to be surface active, to absorb fission products, and to form interfacial films or "cruds". A major step in eliminating these cruds was taken when the Purex columns were modified to operate with the interfaces at the waste ends of the columns rather than the product ends. This type of operation tended to flush the solids out with the waste rather than the product streams.

A research and development program aimed at devising processes for the economical recovery of the potentially valuable long-lived fission products from Purex waste has been under wat at Hanford for several years. When this work has begun, the concentrated waste was primarily a nitric acid solution (6 to 10 M HNO3) containing the fission products and relatively small concentrations of iron, sulfate, and other corrosion products. Flowsheets based on classical separation schemes and rather similar to processes used by the Isotopes Division at the AEC's Oak Ridge operation served to separate the desired fission products from one another and from the corrosion products (1,2,3).These separation schemes employed careful step-wise pH adjustment to precipitate first the iron and then to separate the desired fission products from one another. The flowsheets were demonstrated on a pilot-plant scale with full-level plant waste. However, since the earlier work was complete, plant operations have been modified....

Startup and testing of the PRTR has been divided into three parts, Design Tests, Critical Tests and Power Tests. This document is the third of the series and describes the performance tests which will be made under nuclear power and the general schedule of operation. Responsibility for the startup of the reactor has been assigned to the PRTR Startup Council. The Power Test Sub-Council, which has done the planning of the initial power operation, was established by the Startup Council. Operation of the reactor will be the responsibility of the PRTR Operation (PRTRO).

This article presents data on the effects of gamma radiation on the physical properties of a number of different classes of plastics materials of construction which were screened to determine their capabilities and limitations for service in applications where ionizing radiation exists. The irradiations were carried out in air at ambient temperature at normal atmospheric pressure to various gamma radiation doses up to 1*10^9 r depending upon the stability of the materials to radiation exposure. The different classes of materials and materials within a class are compared.

In the last two or three years there has been a notable improvement in the knowledge and understanding of reactor lattice physics. During this interval, measurements of parameters have been improved and the methods of analysis have been greatly strengthened. It has become apparent that the classical ways for computing the lattice parameters are not sufficient. It is the purpose of this report to publish under one cover as much of the most recent information as possible.

A flexible foil data processing program is described. Raw data on foil radioactivity are the basic input information required. Output may consists of relative activities, saturated activities, and/or cadmium ratio and flux spectrum data. A statistical analysis of the data is executed with the direct calculation, and errors estimated for the output data.

Facilities are to be designed for installation at Hanford for the reprocessing of irradiated, low enrichment fuels from non-production reactors, see Reference 1 and 2. The initial design of the processing facilities is to be based on processing the fuels discharged from the Dresden, Yankee, Piqua, Pathfinder and Shippingport (blanket elements only) reactors. Properties of the fuels and cladding make it impossible to completely process them in existing equipment, although the separation and decontamination can be performed in existing facilities once the fueks have been placed in solution form.

The Hanford water treatment plants were operated for a number of years with alum as a coagulant and activated silica as a coagulant and a polyacrylamide as a filter conditioning agent. (other polyelectrolytes may be used.) Sufficient time has elapsed to make a comparison of the two methods of operation valid and useful. Such a comparison is the purpose of this paper.

This paper describes the measurement of control rod strength in the PCTR. The first section outlines the theory and method of such a measurement, and the second section presents the results of the measurement of the Experimental Gas Cooled Reactor (EGCR) control rod strength. The reactor parameter actually inferred is the difference in the infinitive multiplication factor for a control rod supercell with and without the control rod in place in the supercell. These measurements are, therefore, very closely related to the measurement of k∞ for an ordinary lattice cell.

In recognition of the need for an integrated system plan for future development requirements for radiological instrumentation, this information report was written to provide the initial step concerning such a planning. The report is to be considered a basic information analysis of the present in-use plant radiological instrumentation, of the instrumentation now undergoing active development, and of the future development work contemplated.

Applications of sheathed (metal) thermocouples may require in many instances, the insertion of the thermocouple thermal junction into an area that must be hermetically sealed to the outside. This same area usually contains highly reactive heat transfer medium such as an eutectic alloy of sodium and potassium of sodium alone. When fissionable materials are being irradiated, fission products that are driven off would also be present and must be contained under varying temperatures and pressures.

The Plutonium Recycle Test Reactor is heavy water moderated with a heavy or light water reflector contained by a complex aluminum vessel called a Calandria. ( See Figure 1). Construction of this vessel started in August, 1958, at a large West Coast vendor's plant and was completed at Hanford in December, 1959. The fabrication problems associated with a high integrity welded aluminum vessel were generally unrealized prior to this period. This report covers the fabrication of the Calandria and lists recommendations for improving the design and reducing the cost.