Aid in redesign of the calibration facility in order to: 1. Reduce the rate of exposure for calibration personnel below 1 mr/hr; 2. Reduce the rate of exposure outside of the 3745 Building; 3. Reduce the risk of breaking radium sources; 4. Improve the efficiency of calibration procedures by: a. Allowing calibration to be done faster; b. Providing several calibration ranges which can operate simultaneously in the present space allowed; c. Allowing instrument survey work to proceed while calibration work is done. A well-type calibration installation is proposed and a brief description of the required shielding is included.

Report discussing the development of Hanford Laboratories' nondestructive testing programs and a summary of their results. Programs include AlSi fuel testing, N-Reactor fuel testing, and reactor research and development.

Report describing a series of tests performed to review the safety procedures and systems of Hanford Laboratories' N-Reactor.

Report describing the development and installation of an automated conveyor-type laundry system for the purpose of monitoring laboratory clothing for beta-gamma contamination.

Report regarding the study of the calculated costs of fabricating plutonium-enriched fuel elements for Hanford Laboratories' Plutonium Recycle Program. Topics include descriptions of the process, design study of plants that use U235 and plutonium enriched fuel, and estimates of costs.

Report containing the results of a two year program wherein the 85 zircaloy-2 pressure tubes in Hanford Laboratories' Plutonium Recycle Test Reactor (PRTR) were monitored. Experimental procedures and results are described.

Report discussing the dissolution of simulated fuel cores. Equipment and procedures are discussed along with significant characteristics of the core materials.

Report discussing the "equilibrium uranium distribution between an aqueous nitric acid solution and a 30 per cent by volume solution of tributyl phosphate in a hydrocarbon diluent" (p. 2). This includes the necessary equations.

Report that summarizes the potential hazards of the Hanford Critical Mass Laboratory and presents operation procedures which will limit the probability of a nuclear accident. The report also outlines the planned experimental program, descriptions of the facility, and the administration of nuclear safety in the laboratory.

The relative neutron capture gamma production in several prospective iron thermal shielding materials for the New Production Reactor was measured to determine the merit of adding boron to the metal. It was found that for the beam geometry the used addition of 1 1/2 percent boron to the iron before casting reduced the soft gamma production by a factor of 6.5 and the hard gamma production by a factor of 10. No attempt was made to measure gamma or neutron transmissions.

Since the start of E-metal processing in December, 1958, critical mass control in the Redox dissolvers has been provided by a combination of administrative controls, revised dissolver charging procedures, and new dissolving processes and operation control procedures. These were all based on a maximum allowable charge of E-Metal to the dissolvers which was equivalent to 45 percent of the minimum critical mass.

In the field of nondestructive testing the ultrasonic test method is used extensively. In this method ultrasonic energy generated by a transducer is caused to enter the test specimen. The effect on the transducer is cause to enter the test specimen. The effect on the transmission or reflection of the ultrasonic energy by differing test specimen structure is used as an indirect measure of test specimen characteristics.

The referenced memo asked several questions on the operation of the Purex Plant which are of interest to the Eurochemic personnel. These questions are answered below. 1. A qualitative description of the nitric acid recovery systems is desired -- flow of materials, number of pieces of equipment, etc.

A multi-purpose program for processing exponential data has been prepared for the 709 computer. The main purposes of the program is to compute the material buckling from raw data (given counts, time, and counter information) or from previously calculated Athermal's. It is also possible to compute only CeCh (end and harmonic corrections) for a given B11 or series if B11's no counting data being entered. In every case, pile measurements must be submitted as input for corrections.

Disposal on a specific retention basic of certain radioactive liquid waste solutions emanating from separations from separations plants has been practiced at Hanford since 1944. As used at Hanford, the term "specific retention" is defined as that volume of waste liquids that may be disposed to the soil* and be held against the force of gravity by the molecular attraction between sand grains and the surface tension of the water, when expressed as percent of packed soil volume. In practice it represents the volume of liquid that may be discharged to a disposal pit of known dimensions without leakage to the ground, water, expressed as percent of the total volume of a column of soil with the same cross section as the pit, and extending from the bottom of the pit to the water table. It is recognized that some degree of lateral spreading will occur which has the effect of enlarging the volume of soil contacted by the liquid.

A series of experiments has been conducted during the past eighteen months with the objective of degerming the accuracy with which precision parts can be duplicated using a castable tooling plastic as a test shape to be reproduced. A precision ground ground 1/4 inch-thick, 90° steel sector of 4'' radius was chosen. The plastics used in the experiments were the die surfacer mixes of the Devon Cooperation and Ren Plastics, Inc. These were selected for use in the test because of their dimensional stability and wear resistance. Ordinary paste type floor wax was used as a mold release. The first series of tests were primarily intended to gain facility in handling this material. After achieving some familiarity with the use of these castable plastics a technique was devised which seemed to hold promise of precision duplication.

Basic Studies. Sinterability studies on the isomorphous system UO2-PuO2 have continued. One-half inch diameter X 3/8'' long compacts containing physical mixtures of the two components have been heated in hydrogen. Sintered density as a function of time and temperature has been determined for times of one and eight hours up to 1600 C. In general, there seems to be minima on isothermal plots of density versus composition in the intermediate range 20-40 w/o PuO2. At concentrations greater than 40 w/o PuO2, density increases rapidly with increasing PuO2. In every case pure PuO2, (produced by decomposing plutonium oxalate at at 300 C) sintered to a greater density that ball ball milled PWR grade UO2. The data, for a one hour soak time, and tabulated in percent of the theoretical crystallographic density, can be seen below:

Facilities shall be provided in the 221-U Building for removing fuel elements from the casks as placed in the railroad tunnel from the transfer facility and moving the elements to modified existing 10' X 16' X 14' storage tanks in ten existing cells where they will be stored, until scheduled for processing.

A number of Pu-Al and UO2-PuO2 Zircaloy clad capsules have been fabricated for irradiation in the MTR. In addition, a four rod cluster containing Al 8 w/o Pu and Al 12 w/o Si 8 w/o Pu cores has been successfully irradiated and discharged from Loop 3 of the KER. A second four rod cluster is awaiting irradiation and design and fabrication of a seven rod cluster test element is underway.

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 the decladding of stainless steel clad fuel elements with sulfuric acid (the Sulfex process), the loss of core material to the decladding solution is determined by the rate of reaction of the core material with the solution and the time of exposure if the core to the solution. Since very little was known about the effects of irradiation history on the rates at which core materials dissolve in sulfuric acid, the present study was undertaken in an effort to determine the magnitude of such effects. This report summarizes the results obtained to date. Further work is planned when materials of higher burn-up are available.

A program has been prepared for the 709 computer which will determine extrapolation length from horizontal traverses measurements taken in an exponential pile. The flux across the pile is assured to fit a cosine function of the form y = A cos B (X-C)

The small amount of irradiation tenting experience on the plutonium-aluminum type elements planned for the Plutonium Recycle Test Reactor (PRTR) has made such testing of great importance. The high temperature pressurized recirculating water 1706 KER facility is one possible place for conducting investigations of the irradiation behavior of this type fuel element. To obtain the maximum information from the in-reactor testing and to detect possible problems, out-of-reactor test both at room and anticipated operating temperatures must be made. Room temperature pressure drop measurements and high temperatures performance of two prototypical fuel element designs proposed for KER testing are reported in this document.