Introduction: "This is the seventh letter report on ARF Project C 127, entitled "Preliminary Studies of Scavenging Systems Related to Radioactive Fallout." This report covers the period from April 1 to May 31, 1959."

The following report discusses direct recording of nuclear pulse height data on magnetic tape that represents an inexpensive method of data storage where some degradation of the original energy resolution can be tolerated.

"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."

From abstract: "The toxicological properties of beryllium and compounds of beryllium are briefly reviewed, together with the historical development of the recommendations for maximum permissible beryllium air concentrations. The application of the enclosure technique presently in use at this laboratory for the control of beryllium hazards is described... Monitoring procedures used to evaluate the performance of enclosures are discussed."

Abstract: A magnetic-jack control-safety rod is under development for the 45.5 thermal megawatt Organic Moderated Reactor. The rod is "unitized," i.e., the poison element, drive, position indicator, and shock absorber are contained in a compact assembly which is inserted in a regular fuel channel opening in the core. Tests to develop components capable of operating under these conditions are described and results are reported.

Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.

Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).

Introduction: A low-capacity, low-decontamination plant can be built, as part of a power reactor complex, to avoid long distance transfer of fuel to a high-capacity aqueous processing plant. Activation of such a complex, with the processing plant adjacent to the reactor it serves, could decrease the cost of the integrated fuel cycle. The study of this concept is a major objective of the Processing Refabrication Experiment (PRE).

Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.

This report documents the study of the recovery of thorium by solvent extraction in pilot plant pulse columns, using a filtered liquor from nitric acid digestions of the raffinate cake produced by the ethyl ether extraction of uranium from pitchblende.

Wind tunnel testing of swept wing fighter aircraft model for determining drag and static longitudinal and lateral stability and control characteristics. Results regarding performance, longitudinal stability, lateral stability, and strakes are provided.

Preliminary experiments have been conducted to evaluate the use of the nuclear reactions Li6 (n,α)H3 and O16(H13,n)F18 to determine the thickness of oxide films on metals. Sheets of thin paper and of aluminum, imbedded in powdered LiF, were irradiated with pile neutrons at a flux of 6 x 10^11 n/cm^2/sec and counted with an end-window proportional counter. A saturation activity of 1.87 hr F18 of 150 dis/min per microgram of oxygen was observed in the paper, but radioactivity due to impurities masked F18 in the aluminum. It is concluded that a 1 A (0.01 μgm/cm^2) oxide film thickness may be measured by a neutron irradiation at a flux of 10^14 n/cm^2/sec but chemical separation of induced radioactivity from the bulk metal is essential.

The "absolute" containment of ions in the DCX magnetic mirror field resulting from the cylindrical symmetry of the field is discussed. The regions of confinement in space and momentum are plotted for 300 Kev deuterons.

For many years effort has been directed toward the development of means for the practical utilization of the heat evolved from the radioactive decay of certain fission and irradiation products. Fission products, in view of their availability in the plentiful, high-level wastes resulting from the processing of irradiated, nuclear reactor fuels have been most intensively studied for applications where their heat of decay might be converted into useful energy such as electricity for the operations of certain devices in place of chemical batteries . In addition other materials having desired radioactive properties may be produced by neutron bombardment of readily available elements.

"The objective of project 32.3 was to evaluate some operational characteristics of a radiological shelter and to determine values for some countermeasures-system parameters."

Sinterability studies on mixtures of UO2 and PuO2 powders have been continued. All of the censities from previous sintering experiments have been recalculated utilizing a somewhat different water immersion technique. Since densities of pellete obtained by physical dimensions are inaccurate due to chipping, and water immersion does not account for the open porosity, the pellets were coated with clear krylon before being immersed in water. This effectively sealed the open pores. Plots of density versus composition still indicate a drastic reduction in UO2 sinterability with even the slightest additions of PuO2 and a rapid increase with PuO2 contents greater than 20 weight percent. The following table compares densities of pure UO2 and PuO2 after an eight hour soak time:

Conceptual design studies conducted at HAPO indicate that the improved thermal efficiencies which can be obtained in a high pressure, high temperature, nuclear-electric power plant might reduce unit power costs substantially. The Plutonium Recycle Test Reactor (PRTR) is designed primarily to investigate various facets of plutonium recycle operation. In order to increases the usefulness of the PRTR as an experimental tool, it is desirable to include facilities for testing materials and fuel concepts under a variety of conditions, including the high temperatures and pressures which may be encountered in future water-cooled power reactors.

Abstract: The caustic permanganate-rinse decontamination treatment was investigated. Loop and metallurgical studies were performed to determine optimum operating conditions as well as the metallurgical effects of the treatment. A treatment with 10 percent sodium hydroxide and 5 percent potassium permanganate solution followed by a rinse with a 5 percent ammonium citrate, 2 percent citric acid and 1/2 percent Versene solution was chosen for the decontamination of a stainless steel steam generator. Decontamination factors of greater than 50 were obtained in loop tests using the above treatment. Corrosion and metallurgical results indicated a total penetration of less than 0.01 mil on annealed Type 304 stainless steel with no evidence of any deleterious effects.

Report describing a the design requirements of a proposed nuclear power plant for use on an ice cap.

Plutonium Oxide Fuels. Mixtures of PWR grade UO2 containing 10, 20, 30, 40, 50, and 60 w/o PuO2 were sintered in hydrogen for 44 hours at 1600 C to get additional data on solubility in this system. Densities of all the pieces were low, approximately 80 percent of theoretical; however, solid solution formation was complete in every case. The low density material should not affect lattice parameter values, but it did slightly reduce the intensity of the reflections.

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 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.

Two Zircaloy-clad capsules (GKH-14-19,20) containing two compacts each of high density PuO2-UO2 mixed crystal oxides were shipped to the MTR in December 2, 1958. The compacts contain 0.026 a/o PuO2, have densities of 91 percent of the theoretical value, and will generate the same specific power as an Al-1.8w/o Pu alloy rod of the same diameter would produce. Two capsules (GKH-14-21,22) have been prepared and contain three compacts each of low density, about 65 percent of the theoretical value, PuO2-UO2 mixed crystal oxides. It is tentatively planned to ship the last two capsules during January 1959.