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.

The purpose of this document is to describe the design and construction of a full-scale, electrically-heated model of a 19-rod PRTR fuel element for use in heat transfer experiments.

This is the third of a series of reports describing the phases of a study of the wear and galling of autoclavedZircaloy-2 by various bearing surface materials. The first report (1) showed that increasing contact area between the autoclaved Zircaloy-2 surfaces will not prevent scratching. It also showed that metals softer than Zircaloy-2 in general do not scratch it. The second report (2) described how eight metals and alloys were wear-tested on autoclaved Zircaloy-2 as prospects for fuel element support-bearing surface materials.

This document authorizes the performance in accordance with the specifications noted, the PRTR Critical Tests described herein. The experiments described have the following objectives:

This development work was undertaken to investigate and standardize scintillation detector and transistorized circuitry techniques in instruments used for detecting and measuring alpha and beta particles, gamma photons, and both slow and fast neutrons. At the present time, the standard radiation detection instrumentation used at Hanford employ vacuum-tube circuitry with typical detectors being ion chambers, G-M tubes, HF3 tubes, etc. The vacuum tubes require excessive battery power, and some of the circuits are adversely affected by humidity and temperature conditions.

In the Sulfex process, stainless steel cladding, generally 304-L, is removed from fuel elements by dissolution in boiling four to six molar sulfuric acid. When cladding removal is complete, the decladding solution is removed is removed to waste. The exposed fuel cores are rinsed and dissolved in nitric acid.

A prototype instrument has been designed and built which demonstrates an ultrasonic pulse echo technique of measuring metal thickness in the range .010 to .080 inches. Extension of the range of measurement on the thick side appears straightforward, however, transducer frequency response considerations may limit the extension of the method for very thin sections. The instrument is based upon the principle that sections, when driven by pulses of vibratory energy, resonate at frequencies dependent upon the thickness of the section.

A reliable spline disposal system is necessary for full utilization of the Poison Spline Control System. The spline choppers were not reliable enough so a spline coiler was developed and designed to replace the choppers. A Development Test, IP-306-X (HW-63820), was issued authorizing on-pile testing of the spline coiler at X area. This report closes the Development Test and discusses the use of spline coilers.

At present, emphasis is being given the application of nondestructive testing to the inspection of nuclear fuel in conjunction with complete product specifications. The objective is to quantitatively specify fuel quality parameters, and to measure the entire fuel production against these specifications nondestructively. Under this concept the acceptability of a given fuel element becomes a well-defined factor readily determined by appropriate nondestructive tests.

It is the purpose of this report to explain the bands and uses of a digital computer program for the calculation of steady state steam-water pressure drop and critical flow in pipes. The program has been coded for the IBM 709 computer. While so attempt has been made to obtain improved models of two-phase flow, it is believed that the numerical integration technique used in the code will permit more accurate calculation where pressure drops are large compared to the upstream pressure. The technique should lend itself readily to any improved correlations which appear in the future.

Two types of enclosed hot junction arrangements are possible with sheathed mineral insulated thermocouples. They are the integral junction and he isolated junction. The integral junction is formed by welding, soldering, brazing or forging the two wires into the end of the sheath. The process forms a good thermal and electrical contact between the three materials.

In answer to the question, "How well do we know the composition of the fuel material for the PRTR start-up tests?", the analytical data on the PRTR fuel elements and other fuel elements which were fabricated by similar processes was gathered and analyzed. The results of this analysis are presented.