Space has been reserved near the southeast corner of the 100-N Area for the 122-N Decontamination Facility. Previous correspondence between Burns and Roe, Inc and General Electric bae discussed various facilities which might be needed in the building. The concepts of the decontamination processes are under active development by research groups at Hanford. At present, there are several workable processes known; each one has one or more fairly serious drawbacks.

Water treatment systems for preparing and maintaining high purity water in out-of-reactor or in-reactor test oops are becoming increasingly important. In out0of-reactor experiments the presence of ionic impurities in the water has a marked influence on film formation and corrosion rates. It is therefore , imperative that these impurities be maintained at the lower practical concentration.

From introduction: "This report provides background information regarding the study results presented in the first volume of this report and provides information resulting from the studies made on the gas cooled reactor concept."

Introduction: In anticipation of the design of equipment for heat transfer tests, a series of short exposure corrosion tests was initiated in an effort to determine the order of magnitude of the corrosion involved. Information was required on the effect on corrosion rates of applying a potential between two concentric tubes, the annulus between them containing condensate water.

Late in 1958 it became apparent that some fuel elements were failing in the Hanford reactors as a result of water entering through the weld. The mode of entry appeared to be first through a void in the weld, then through a non-wet area or a train of voids in the braze, and finally to the uranium core. Defective closures of a similar nature were also typical of many fuel elements which have failed in the autoclaving operation as shown in Figure 1.

From summary: "Uranium billets deficient in U235 were subjected to the normal successive steps in the slug fabrication process; viz., rolling, machining, and canning. No unusual behavior was observed at any point in the process, thus indicating that the depleted metal from the standpoint of mechanical properties can be considered equivalent to normal uranium."

The following report provides data from completed tests for corrosion testing of stellite-T347 stainless steel test cylinders.

From summary: "On the basis of the tests described herein, the corrosion characteristics of T-347 welded stainless steel are found not to be affected by the magnetic properties of the welds even though these properties are likely the result of the presence of the ferritic structure."

From summary: The object of the study was to determine if the heavy metal ions within the concentration ranges normally found in pile cooling water were responsible for corrosion of aluminum.

Introduction: "The need for a method of acid determination which could be used for small samples and easily adapted to remote control led to the investigation of the use of electrolytically generated base for the titration of acid in certain solutions. The use of electrolytically generated base for the titration of acetic acid in the presence of nitric acid has been reported (1)."

In many cases of analysis of two-phase flow in systems, considerable computation or program time could be saved if the critical pressures ratio were known. If a reservoir or plenum pressure is fixed, the usual computational procedure involves the assumption of several critical pressures and the generation of several momentum terms to find the applicable critical pressure ratio and thereby the critical flow. The formulation of an equation of state make it possible to compute critical pressure ratios directly.

The work described was undertaken to provide data for nuclear safety studies concerning NPF reprocessing equipment. The original objective was to determine the uranium density and water (hydrogen) content of UO2-H2O mixtures ranging from compact centrifuge cakes to dilute slurries. The scope was later expanded to include mixtures of UO2 with hydrocarbon oil and mixtures of UO3-H2O.

This report summarizes experimental results obtained for the General Electric Company under terms of Design, Development and Research Contrast DDR-33. A discussion of the results, the manner in which they were obtained and their probable accuracy are included. Suggested improvements or modifications of properties of this nature be contemplated.

The following report describes methods used to determine tributylphosphate in a deodorized shell base. An infrared absorption method is developed to emphasize toward determination of TBP in RCW streams, the organic waste streams.

The components for the present aluminum clad, Al-Si bonded, internally and externally cooled (I & E), uranium fuel elements are composed of impact extruded cans and spire caps as shown in Figure 1. This type of component requires two impact extrusions; however, in December, 1957, J. E. Ruffin proposed another design of component in which there was only one impact extrusion. For this component, Figure 2, the spire was impact extruded as a part of the can.

From introduction: "The following report presents a body of information obtained in the Health Instrument Development Laboratory from September, 1948, to the present, and is intended to be a preliminary review of [the study of electrodeposition of plutonium]. Many improvements in technique and methods and investigations along lines not explored are suggested by the present work. These will be undertaken as time permits, and reported at a later time."

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.

Primary control of the PRTR is achieved by regulating the level of the heavy water moderator which is held in the reactor vessel by a helium gas balance system. Emergency shutdown is effected by a gas-balanced moderator dump system which drain the moderator from the calandria at a rapid rate. This report presents a quantitative appraisal of the reactivity effects due to moderator level changes in controlling or scramming the reactor. In conjunction with the reactivity calculations, solutions were obtained which yield an evaluation of vertical flux or power distributions for any positioning of the moderator level. Coupled with the radial distributions for a given fuel loading, this information is useful in obtaining the value of the maximum specific power associated with a given power of operation and moderator height. The calculations were made using VALPROD, a one dimensional, multigroup diffusion theory reactor code programmed for the IBM-650 computer.

Recently proposals for justifiable additions to the Plutonium Recycle Test Reactor Complex were presented to the Atomic Energy Commission at their request. In addition to a critical reactivity measuring facility in the fuel element storage basin, the following in-reactor loops were proposed: 1. A high pressure, H2O cooled fuel test loop. 2. A rupture loop to investigate fuel element failures. 3. One or more materials testing loops.

Study of factors, such as the average particle size and particle-size distribution, which affect the fluidization properties of uranium trioxide powders.

Introduction: "The problem of determining the properties and necessary controls on the TBP Plant waste disposal to eliminate transport line plugging and to permit higher concentration prior to return of this to the tank farms was studied. A report of preliminary findings was made to L. W. Finsh on January 31, 1953. This report compiles information in the aforementioned report and findings since its issuance. Previous studies related to this problem are summarized in the U. R. Technical Manual, HW-19140."

Report describing the performance, fission product tolerance, design, and costs of the DR-1 Gas Loop, which is an in-reactor test facility.

Report that summarizes out-of-reactor tests undertaken to determine operating conditions, estimate damage from and extent of fretting corrosion, and describe continuing tests planned to monitor and eliminate this corrosion.

This report covers an investigation made to develop a gasometric determination that can determine 1.2 x 10-(-5) g. of nitrite with a precision (99% limits) of +/- 8%, and discusses how this procedure works.