Report describing the Hanford Atomic Works' Plutonium Recycle Test Reactor, its siting concerns, operation, and safety concerns.

Series of isothermal transformation studies for improvement in fabrication and reactor performance of uranium fuel elements.

Experimental combination hand and shoe counter redesign as a compact, simplified, register readout instrument no larger than the present Five-Fold called Model 2.

A test program conducted to develop high integrity piping and seal components for Plutonium Recycle Test Reactor installation. Methods of preventing leakage of costly D2O coolant and maintaining operational safety were investigated.

Decontamination of stainless steel and carbon steel used in high temperature recirculation systems is currently being studied to obtain an effective and economical decontamination process for use in these systems. This report presents the preliminary investigation process which has demonstrated very effective decontamination and is low in cost.

In formulating this general diffusion theory expression which represents the neutron balance in a nuclear chain reactor the following assumptions were made : (1) the medium through which the neutrons are diffusing has a low neutron capture cross section, (2) the region in which the flux distribute is being described is two or three mean free paths from strong sources and sinks or from a boundary. Certainly, is going to the lattice cell of a receptor, both of the above conditions are violated; fuel elements have a large absorption cross section and most lattice cells are only two or three mean free paths to is over-all sites

Plutonium Fuels Development Basic studies. Experiments to determine the effect of plutonium dioxide additions on the sinterability of UO2 have continued. PuO2 has been added to ball milled PWR grade UO2 as a physical mixture, and in the form of the mixed crystal oxide.

The transfer or circulation of raw dissolver solutions containing gross particulate matter presents many problems not easily overcome by standard pumping equipment. In April of 1956 the HAPO concept of a modified Archimedes screw pump was developed. Two basic models, externally powered and driven through extended shafts, were constructed and tested. Operation of these preliminary models was so satisfactory that a third unit, integrally formed into drive motor, was built and placed in extended life test. This report describes construction and testing of the third and final model.

The methods for the separation of the elements from thorium to americium having wide use are those employing solvent extraction techniques (1) (2). During recent years the behavior of these elements on anion exchangers has been studied, resulting in the wide application of these resins to the separation of the actinides (3) (4) (5) (6) (7).

The instability of the steel tank bottom of 113SX waste storage tank was postulated to have been caused by a pressure underneath the steel liner which was in excess of the hydrostatic liquid load of the waste resting on the steel bottom.

In HW-56161(1), the preliminary background, bases, and advantages which could be visualized in the study and establishment of reactor concepts utilizing boiling mercury- 204 were presented. The attractive chemical and metallurgical properties of mercury which make it particularly suitable for use in special non-rigid fuel systems as well as its potential for heat transfer applications were considered to be of significant interest to the Plutonium Recycle Program also, since the issuance of the original document, continued study of the potentials for an economical isotope separations process for mercury has shown attractive possibilities for a method based on photochemical activation (HW-59329). (2) The prospectas(3) for this process appear so favorable that earlier opinions and expectations seem to have been conservative. Although the desired level of initial effort in related studies has not yet been realized, particularly in the area of chemical engineering, reactor engineering, and economics, it appears advisable to revise the earlier document to recognize the impact of the favorable outlook for economical mercury isotope production, to present other related information which has been developed, and to recommend the beginning of a research and development program.

The processing of spent fuels from nuclear power and propulsion reactors is planned by Hanford Atomic Products Operation as part of the Atomic Energy Commission's interim processing scheme. The spent nuclear fuels have cores of UO2, U, or alloys of U-Mb, U-Al, or U-Zr clad in either stainless steel, aluminum, or Zircaloy. This report discusses only the Zircaloy-clad fuels and the applications of the Zirflex Process for dissolution. Zirflex chemical flow sheets are presented as developed by pilot plant operations.

This note describes an analysis of the primary creep rates in thin-walled pressure tubing. The analysis accounts for the increase in tube radius and decrease of tube wall thickness. Both of these effects increase the stress; hence, they can affect the creep rate and cause unstable creep conditions.

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.

A new package type facility for corrosion decontamination studies has been designed and fabricated. This report gives a description, drawings, and comments on its operation. It has been in operation since December, 1958.

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.

Following is a description of what may be an invention in the art of compacting uranium dioxide (UO2) and other ceramic or cermet nuclear fuel materials for ultimate use as reactor fuel elements. Vibration of ceramic fuel materials at the resonant frequency of the metal encasing material, using high rates of acceleration, produces uniformly compacted high-density fuel elements.

It was suggested to the writer by K.A. Eschbach that liquid magnesium might offer certain advantages as a reactor coolant. As a result of this suggestion, a preliminary investigation of the possibilities of this material was made. Definite advantages for a restricted class of applications were found, but a detailed evaluation would seem to require further basic experimental research on the heat transfer, corrosion and flux-mechanical properties of the substance.

This examination was conducted primarily to determine the extent of corrosion existing between the silver bearing alloy and the type 416 stainless steel nut of K reactor units subjected to service conditions and also to establish irrevocably that metallurgical defects exist in the brazes.

Graphite suffers radiation damage when exposed to high energy particles. Many of its physical and chemical properties are changed. One of the effects which is easily measured is an increase in the interlayer or C₀ distance of the crystal lattice. This has been attributed to the displacement of atoms or groups of atoms into interstitial positions of the lattice forcing the planes apart. The C₀ expansion may be annealed out by heating the damaged crystal to high temperatures. This paper deals with the rate of annealing of C₀ damage at several constant temperatures.

The primary objective of this work is to add to the understanding of irradiated graphite. This report describes an experimental technique whereby changes in crystal lattice parameter may be studied as a function of time at constant temperature. The results given here are limited to changes in the C₀ spacing for a typical sample of irradiated graphite in the temperature range 100 to 750 C.

A corrosion testing program has been initiated in Purex and Redox storage tanks to obtain corrosion data on SAE 1020 carbon steel, two alloy-high strength steels, and an alloy steel in neutralized process wastes.

The purpose of this report is to develop an economic selection of conventional reinforced concrete structures of various shapes to establish a design criteria for intermediate level radioactive liquid wastes.

Reactivity measurements conducted in the Hanford Test Pile are interpreted to yield resonance integrals for uranium, uranium oxide, thorium and thorium oxide. The experiments were made using cadmium filters and are discussed in terms of the multiplication factor of the Hanford Test Pile.