Introduction: Static, total immersion corrosion tests were carried out at room temperatures with Carboloy grades 44-A, 78, 907, X-3119-A and X-3119-B in Redox process solutions (ANL, June 1, 1949 Flowsheet) 1AX, 1AF, and 1AS.

A report which deals with the Redox IA Column studies made in small-scale columns packed with 3/16-in. Fenske helices and using NH4NO3 salting agent.

Abstract: "The critical mass both of an untamped and water-tamped sphere with parabolic radial variation of the intrinsic buckling is calculated by a perturbation method. The result is applied to finding the minimum critical mass of plutonium in water suspension with infinite water temper; the calculations show that the minimum critical mass at constant concentration may be reduced by an amount of the order of 7.6% when the concentration of plutonium is permitted to vary throughout the suspension."

Summary: "Criticality was considered impossible in any contingency encountered in practice under either of the following conditions which include no safety factor: (a) In any isolated cylinder not more than 5 in. in disruptor, if the total amount of U235 present does not exceed 7.5 Eg. (b) In an isolated vessel of any shape and size, if the total amount of U235 present does not exceed 880 g. An "isolated" vessel is one which does not "see" more than 0.1 [...] other vessels containing fissionable material."

The following report examines methods to determine uranium nitrate hexahydrate on pretreated hexone in the range of 0 to 40 g./l. by infrared absorption measurements.

Objective: "The observations summarized in the present report were made to determine the stability of ferrous and sulfamate ions and sulfamic acid in RA Column scrub solution (RAS) and in aqueous solution comparable to the RA Column feed tee composition (RAFS). Knowledge of the stability of the reducing agent (ferrous iron) is necessary to properly plan the method of preparation and introduction of this component into the RA system and to assure dependable and reproducible performance."

Introduction: "A laboratory investigation of the relative corrosion resistance of several austenitic stainless steels to the RAW stream of the Tributyl Phosphate Waste Metal Recovery Process was made during the first quarter of 1950. The results of this investigation are reported herein; all original data are recorded in laboratory notebook HW-3282-T."

Introduction: "A corrosion test program, to determine the suitability of LaBour B-55 and Worthite for use in the construction of circulating pumps for th UO3 process has been completed and the data are appended."

Introduction: A limited test progress on Alloghany-Ludlum type 347 stainless steel plate which failed the Huey test (ASTN A-262-44T) has been completed.

This report follows an investigation on how hydrolysis products of tributyl phosphate (such as dibutyl and monobutyl phosphoric acids, orthophosphoric acid and butanol) effect the tributyl phosphate process for uranium recovery.

From introduction: "Analytical procedures for the determination of plutonium samples from the extraction and concentration plants utilize the carrying of the hydroxylamine-reduced trivulent plutonium by coprecipitation on lanthanum fluoride. Most other plutonium procedures also take advantage of the efficient carrying on lanthanum fluoride."

At the request of 234-5 Development, studies of a system for controlling plutonium powder transfer were undertaken by Process Equipment Development. This report presents details of equipment designed to fulfill requirements and proposes equipment for installation on additional Hot Button Line prototypes.

In conjunction with the fuel element development program at Hanford, it is desired to determine the effects of cladding and core temperatures, cladding thickness, and exposure upon the swelling behavior of unalloyed uranium. To obtain this information, it is proposed to irradiate several fuel rods, clad by coextrusion with Zr-2, in NeK filled stainless steel capsules. The central uranium temperatures are to be monitored by axial thermocouples. Irradiation tests in the MTR and ETR using capsules of similar design have been and are now being conducted. GKH 3-31, a fuel rod, clad with 0.030" Zr-2, operated in the MTR at an average center fuel temperature of 425 C to an exposure of 2100 MWD/T. GKH 3-57, 3-58, and 3-59 are presently undergoing irradiation in the ETR.

Crib evaluation studies showed the soil uptake of plutonium from the Recuplex process (CAW) waste to be low. Preliminary studies of the low soil adsorption of plutonium revealed the presence of a plutonium nitrate anion complex which could be removed by adsorption on a strong base anion exchange resin.

Thermal Resistance of the contact between to core and the jackets or unbonded fuel elements may easily be the largest source of error in core temperature predictions. The object of this work is to improve contact conductance predictions by measuring conductance of the joint between reactor fuel and cladding materials at joint pressures, temperatures and thermal flux levels approaching reactor service conditions.

A previous study has been completed on the buildup and decay of xenon poisoning in the PRTR by simulating the reactor as a single region. The results of this study indicated that a study using a more refined model of the reactor would be valuable.

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.

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.

A series of tests conducted in an attempt to reduce the observed O2 content of primary recirculation loop water to a value of <0.14ppm by the current loop method which is a combination of pressurizer degasification and venting flow.

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.

Laboratory studies involving fluid flow through porous media require use of fluids having low dissolved gas content. Water is the major fluid used in various and box model and soil permeability studies carried out by the Geochemical and Geophysical Research group. Tap water supplied to the 222-U Bldg. contains a large amount of dissolved air. Under the reduced pressure encountered during model studies, the air is released from solution and gradually clogs the pores of the sand or other porous material. This, of course. leads to anomalous results and cannot be tolerated in precious studies. A system was required to effectively remove the air and make available a continuous supply of desired water for the model studies.

The build up of contaminated film on the internal surfaces of high temperature in-reactor recirculating water loops has created serious radiation exposure problems to operational and maintenance personnel. A considerable amount of work has been applied to develop an effective decontamination process for the decontamination of these loops and their components.

The Atomic Energy Commission has recently developed an enlarged program for the study of graphite moderated, gas cooled power reactor systems. It has been recognized, however, that understanding of radiation damage and radiation induced chemical reactions of graphite at the proposed high moderator temperatures is inadequate and that improved understanding is essential if the design of such reactors is to be optimized. Accordingly, the Atomic Energy Commission requested Hanford to organize a modest research and development program directed toward the study of these graphite problems.

This report was written to describe briefly several possible variations of such systems along with probable cost estimates. Previous work has been shown that the application of scintillation detectors is the simplest and most reliable means for such monitoring. By proper detector selection and measuring methods, the system can be of an approximate gamma dose-rate with gamma energy independence above about 100 Kev. The same system using detector change only is directly applicable to beta, gamma, slow neutron and fast neutron detection. Any type of alarming device desired can be incorporated.