Report summarizing corrosion tests meant to determine the corrosion rates of bronze, mild steel, T-304 stainless steel and 2S aluminum in 10% potassium tetraborate solution. Tests were completed under the boiling point of each solution except for 2S aluminum, which was tested at 40C, the approximate temperature in a pile.

The following remarks are intended primarily to assist in the orientation of the PRP reactor control development program in a direction consistent with the basic objectives of the program as a whole, and to discuss certain aspects of particular control systems which have a bearing on the capabilities of the reactor with respect to these objectives.

It was shown that without further major development the detectors proposed for the Body Monitor can be expected to detect 0.02 microcurie of plutonium at the center of the chest cavity. This is the permissible body burden of insoluble forms of plutonium. The detection level can be decreased by decreasing the background counting rate. Decreases in the background can reasonably be expected; it should be possible to detect a small fraction of the permissible body burden. If it can be shown, as suggested by other work, that americium will remain with plutonium, then within at least a few months after entry into the body it should be possible to measure the plutonium by detection of rays from Am²⁴¹.

An ex-reactor test stimulating the charging of a tubular fuel element in a Zircaloy-2 process tube showed a problem of severe process tube scratching and galling may occur if metal sliding contact fuel element supports are used. Both the fuel supports and the process tubes were autoclaved. Although the conditions of this test were severe in that the sharp edges rather than a flat surface of the supports were in contact with the process tube, it pointed out the potential problem of galling by point on small area contact supports on heavy fuel elements. This difficulty had not been observed with the lighter tubular or cluster fuel elements used in KER irradiation. In addition to the loss of metal and thinning of the process tube after a number of charge-discharge operations, these scratches may lead to localized corrosion or act as points of stress concentration.

Experimental studies on continuous denitration performed in 321 Building resulted in the development of an agitated-through-type calciner. The performance of a paddle-agitated 4-in.-diameter by 44-in.-long reactor was sufficiently promising to warrant construction of a 16-in.-diameter by 8 ft.-long semi-plant scale model. This progress report summarizes the results of studies to date employing the semi-plant scale calciner.

Production Test 313-105-13-M authorized the irradiation of a uranium - 0.4 atomic per cent chromium alloy to determine the relative stability of this material during irradiation. Five tubes, each containing six four-inch alloy slugs, were irradiated to exposures ranging from 156 to 834 MWD/T. This report describes the results of the visual examination of the bare uranium alloy slugs and the results of the test to determine slug breaking strength. The breaking strength of both the U-Cr alloy slugs and the normal uranium control slugs was reduced by exposure in the pile. The breaking strength of those U-Cr alloy pieces not cracked or broken in the jackets before examination was consistently greater than the breaking strength of the normal uranium at all exposures.

The purpose of this report is to provide data pertaining to the corrosion damage incurred in the use of oxalic acid as a cleaning agent for process equipment.

This report describes the work carried out in 321 Building semiworks equipment, to define the factors contributing to high corrosion-product contamination and presents recommendations for reducing the impurity level to meet current specifications (maximum of 200 parts total metals per million parts U).

The effects of oxidation and irradiation on the microsurface structure of pile graphite have been investigated by measuring the surface area and pore size distribution of several samples. The results obtained for both oxidized and irradiated graphite samples indicate that changes in surface characteristics which occur are determined by the flux, temperature of irradiation, and gaseous atmosphere in which the radiation takes place.

Unanticipated radioactive contamination of the 200 Areas environs was discovered in the latter part of 1947, about 2.5 years after the startup of the Bismuth Phosphate Separation Plants. It was subsequently established that this contamination was due to radioactive particles emanating from the Separation Plants' stacks, and the need for exhaust ventilation air filtration equipment was recognized. Sand filters were installed in the fall of 1948 in the ventilation systems of the B and T Bismuth Phosphate Plants. These filter units essentially eliminated this contamination problem and have performed satisfactorily to the present time.

Many procedures for the chemical isolation of the microscopic angular inclusions in rolled uranium have been investigated. Four were selected as most satisfactory for this work and will be described in detail. The procedures are: (1) the copper displacement of the uranium, (2) the electrolytic displacement of the uranium, (3) the chemical dissolution of the uranium in an anhydrous solution of hydrogen chloride in methanol, and (4) the chemical dissolution of the uranium in an aqueous solution of hydrogen chloride and hydrogen peroxide.

The useful lifetimes of the phosphoric-hydrochloric anodic etching solution and the nitric acid stripping solutions were studied. Groups of uranium wafers of surface areas approximately 0.033 sq ft were anodically etched and stripped. From the losses in weight observed, it was concluded that at least three square feet of uranium may be anodized per liter of phosphoric-hydrochloric acid solution without regeneration. The nitric acid stripping solution was found to be satisfactory for at least one square foot of uranium per liter, but the reaction rate decreased rapidly after the treatment of two square feet per liter.

Equations for the quantity of thorium daughters on an air sampler after various times of collection and decay are derived and evaluated. The most limiting assumption is the constant concentration over the sampling period. Applications of these equations to measurement of decay curves, alpha energy analysis and the simultaneous solution involving measurements at the three decay times are discussed.

In support of an in-reactor thermocouple stability testing program now in progress, laboratory studies were made relative to the stability of thermocouple bundles similar to those projected for in-reactor use. Two sets of chromel-alumel, copper-constantan and iron-constantan thermocouples were held for extended periods at 300 C in successive atmospheres of 75% helium-25% carbon dioxide. 100% carbon dioxide and air. Thermocouple stability was verified by measuring emf's at the freezing point of lead. The results of these tests are to be used to evaluate data obtained from like thermocouples subjected to condition similar to the laboratory tests with the added facet of reactor irradiation.

Maximum safe cylinder diameters, slab thicknesses, and minimum critical masses in spherical geometries for fuel of < 5.0& U-235 enrichment have been evaluated. It is of interest to know the minimum critical masses as a function of cylinder diameter for vessels that are not safe by geometry. Curves will be presented of critical masses as a function of cylinder diameter and fuel rod size for 1.6%, 3.0%, and 5.0% U-235 enriched metal-water systems.

Neutron radiation can be greatly reduced by substituting plutonium trichloride for plutonium tetrafluoride in the dry chemistry step of metal preparation. Chloride processes are now common in the metal preparation industry, and the 234-5 Developmental Laboratory has shown plutonium trichloride can be satisfactorily prepared and reduced to the metal.

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.

The design of the Purex Off-Gas Treatment Facility requires the use of new dissolver off-gas jets which will be installed in series with present jets in the Purex in-canyon "donut" jumpers. When the new jets are placed in operation, the motive air and steam supply services to the in-canyon jets will no longer be required and the in-canyon jets will become flow restrictions in the dissolver off-gas train. Calculated values of pressure drop across these "dead" jets at maximum off-gas rate varied from about 18 to 7 inches of water depending on whether the dead jets were considered as sharp-edged orefices or venturi meters without downstream pressure recovery. In the preparation of process specifications for hets for the new facility, accurate values for the pressure drcp introduced by the dead jets were required to avoid oversizing the new jets with resultant waster of motive jet fluid.

A centerless-ground, 0.25-in.-diam. by 0.75-in.-long sapphire single crystal exposed to an integrated thermal neutron flux of 2 x 10 19 nvt at a temperature of 350 deg C (660 deg F) did not change its dimensions appreciably simple beam, with slightly less force than was required to break a similar unirradiated rod. In a darkened room the irradiated sapphire begins to thermoluminesce noticeably at 280 deg C (535 deg F) and most of the amber or brown coloration produced by the pile exposure is retained after annealing for 24 hr at the pile exposure temperature of 350 deg C (660 deg F). (auth)