The subject of this report is the generalized computer program written to solve problems involving steady fluid flow through heterogeneous, partially-saturated porous media.

Report containing the procedures and results of tests made to determine the corrosion resistance of materials in a simulation of Hanford Laboratories' N-Reactor secondary system water environment.

Report reviewing literature that concerns "the compatibility of stainless steels and nickel-base alloys with various gases which might be present in or around a high temperature nuclear reactor" (p. 2).

This report describes the final design of the Fuel Element Rupture Testing Facility and summarizes the safeguards analyses of the design and plans for operation.

Report that "describes the equipment, technologies of use, and reliability of an ultrasonic tester - designated UT-4 - capable of detecting very small defects in the brazed closures of reactor fuel elements" (p. 2).

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....

A brief series of experiments was performed to study the high temperature reaction of the Zircaloy-uranium-aluminum system at the request of Reactor Engineering Operation, Irradiation Processing Department. Knowledge of the high temperature behavior of this combination of materials is important to the evaluation of reactor behavior during possible accidental over-heating of a fuel channel.

Graphs have been made which show the nuclearly safe parameters for uranium rods in light water with uranium enrichments up to five weight percent U-235. These data were to serve as a guide to those persons who may be involved with the maintenance of nuclear safety in handling and processing operations with slightly enriched uranium fuel elements. The data are applicable to fuel element fabrication and processing operations, and in general to those operations involving the handling and storage of fuel elements apart from reactors.

The evaluation of the inductance of an electrical component is often important in understanding and operating an electrical circuit. Inductance is defined by several equations, the most common being e= -L d1/dt . This equation defines inductance as the constant which relates the back emf produced in an element to the time-rate-of-change of current.

This report summarizes the research and development work carried out during December, 1959, and January and February, 1960, for Budget Activity 2790 - Separations Development for Non-Production Reactors. The major effort on Activity 2790 has been completed. Current efforts on the remaining problem areas will enable Hanford to begin reprocessing fuel elements from power reactors which employ depleted or slightly enriched uranium fuels in July, 1962.

At the present time, cerium-144 and premtheium-147 are the two fission product rare earths that appear most promising for use as the heat source in isotopic power units. Under proper conditions, cerium and the trivalent rare earths can be extracted from the Purex fission product waste stream as an insoluble sodium-rare earth double sulfate. A reprecipitation as the double sulfate, dissolution of the hydroxide, serves to give almost complete separation from the corrosion products, inert constituents of the waste, and from most of the fission products. The cerium and the trivalent rare earths must then be separated from each other. In the case of cerium recovery, it is necessary to remove the trivalent rare earths in order to maximize the specific activity of the cerium. If promethium is the desired product, a preliminary cerium separation is desirable to protect the ion-exchange resin (used for separating promethium from its adjacent rare earths) from the intense high-energy radiation from cerium.

Critical mass experiments to be conducted at HAPO will require dilute and concentrated plutonium solutions. Since very high plutonium concentrations (~2,000 g/l) cannot be attained in pumpable solutions, solid plutonium compounds will be utilized for this phase of the work. Mixtures of plutonium oxide and polyethylene, methyl methacrylate or paraffin compacted in containers have been considers for use in highly-concentrated plutonium systems of known hydrogen to plutonium atomic ratio (H/Pu).

The measurements reported here are companion measurements to those reported earlier in HW-63585. The only significant difference between the measurements is that 1.8 w/o enrichment UO2 fuel was used for the first set, and 2.6 w/o enrichment UO2 fuel was used for the measurements described in this report. The new results will be presented graphically, and for completeness, the details of the measurement will be included here as well as in HW-63585.

Four capsules containing Al-1.65 w/o Pu and Al-12 w/o Si-1.65 w/o Pu were charged into the MTR. These capsules will be irradiated to a burnup of 80-100% of the plutonium atoms to determine the stability of the material at high exposures. An additional sixteen capsules containing 5 to 20 w/o Pu in Al and Al-Si have been prepared. Eight are awaiting reactor space and should be charged soon. The remaining eight are being recanned to increase the end gap between the core and the can. Also four capsules containing high density UO2-PuO2 pellets canned in Zircaloy are awaiting assignment of reactor space.

A 19-rod Zircaloy-clad half-length PRTR spike element successfully irradiated to high plutonium burnout at full power in the ETR shoved partial bonding of the core and cladding. Unsatisfactory autoclave films on the Zircaloy cladding of the PRTR Al-Pu fuel elements are delaying final assembly of the first 30 clusters. Further conditioning of the autoclaves and the availability of permanent etching facilities are expected to correct the present difficulties.

Investigation of aqueous and steam corrosion at elevated temperature has been carried on at Hanford for several tears. A number of reports have been published on various phases of the corrosion program but, except for monthly reports, little has been written on the progress of the Corrosion and Coatings Operation alloy development program. This interim report will outline the more important current phases of the aluminum alloy program.

In the continuing search for better methods of effecting the secondary fusion welded closure in Alsi bonded aluminum clad uranium metal fuel element, an article on "The Twin-Argon Welding Process" by J. A. Donellon was recently found in the September 1954 issue if the British Welding Journal. From a review of the article, the process appeared to have enough possibilities to warrant exploring it. Communication were established with the General Electric Company, Ltd..

The development of a miniature probe was desired for measuring approximate single nuclide beta dose rate in solution and in various animal organs. This probe designed for biological experiments, was to have maximum possible sensitivity to detect low levels of nuclide concentrations. The desired dimensions of the light pipe were to be approximately one-fourth-inch diameter with lengths of three to twelve inches.

In order to prevent dangerous concentrations of combustible gases such as hydrogen and ammonia in Redox dissolvers, a continuous analysis of the off-gases is needed. Hydrogen is evolved from most fuel element dissolution processes now used or planned for use at Redox. Ammonia is also released from ammonium fluoride dissolution of zirconium and caustic dissolution of aluminum cladding. Oxides of nitrogen and hydrogen are formed during nitric acid dissolution of uranium or plutonium-aluminum alloys. Since no fixed ratio exists between the concentrations of any two of these gases, the determination of combustible gases in complex mixtures is not possible with a single bulk property analyzer.

In early July, 1958, it appeared that Columbia River temperature at HAPO would be near 24.5ºC by the end of August. River temperature were averaging 4º to 5º above 1957 figures and were 3º to 4º above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent.

Hastelloy F is a prime contender as a material of construction for plant dissolvers in the power fuel reprocessing program. Consequently, the fabrication and installation of dissolver was undertaken to delineate any unknown problems associated with the use of Hastelloy F; and, at the same time, to provide a vessel for development studies on the Niflex or the Sulfex processes. The purpose of this report is to describe the actual basis for design as well as to present the problems encountered during the fabrication of the vessels.

In considering the problems associated with the measurement of SSNM content of Non-Production Fuels, two distinct problems are apparent. The first is the technical problem of obtaining highly accurate measurements in a complicated chemical and physical system. The second is the administrative problem of choosing a measurement system which provides a basic of financial settlement acceptable to both the seller (reactor operator) and the buyer (Commission).

The spatial distribution of thermal neutrons in a reactor lattices cell is of fundamental importance for many reasons. First of all, this information allows the determination of the relative absorption rates in each component of the cell which in turn enables a determination of the thermal utilization, and the diffusion length of the lattice. In addition, the observed spatial distributions of thermal neutrons in the lattice cell is of major interest in testing various approximations to the solution of the transport equation such as the P1 and P3 solutions.

The major portion of transfers and inventory in the Chemical Processing Department are dependent on volume measurement. Good volume measurements are therefore essential to good SS material control. Considerable attention has been given to the volume problem especially on primary transfer points. Document HW-57888 discusses the biasing effect pf temperature on manometer oil. HW- 57866 given the procedure for vessel calibration. Other undocumented correspondence has discussed recent calibrations and certain aspects of temperature considerations in the volume determinations. However, there is a need to document for future reference the consideration given temperature and provide at the same time information which covers the temperature angle on all volume methods used in the Chemical Processing Department.