A theoretical expression for the stripping differential cross section is derived and prepared for computer evaluation. Input information is provided for the code B53DWC, which calculates this cross section, and the codes B53GOR, B53SIX, and B53RIM, which evaluate 3-J, 6-J, and 9-J coefficients. (RWR)

One source of the parent materials which are converted into the radioisotopes in reactor effluent is the impurity content of Columbia River water. It has been demonstrated that as the efficiency of removal of the radioisotope parent materials increases, the amount of radioisotopes in the reactor effluent decreases. An example of this is the activity reduction which results from the use of high alum feed. In order to determine whether effluent activity would be influenced by the length of filter run, a half-plant test was initiated at DR Reactor on September 28, 1964, under the authorization of Production Test IP-719. This report presents the results of the test. The water treatment plant at DR Reactor was operated for a three-month period with half the filters on a seven-hour filter cycle and the other half on a 5.25 hour filter cycle. The shorter filter cycle did not significantly reduce either the filtered water turbidity or the P{sup 32} and As{sup 76} concentration in the reactor effluent. It can be concluded that in the range investigated effluent activity is not influenced by the length of filter cycle as long as breakthrough is avoided.

This report gives the exposure data for the C Reactor at the Hanford Reservation February 1965.

This report details activities of the Chemical Processing Department during the month of January 1965.

Metallographic examination has been completed on samples from the Nichrome V wire mesh furnace belt in the 232-Z building incinerator. During service the furnace belt had broken in several places where the wire was welded together. A grain boundary phase was found in weld metal during metallographic examination of the as received material. It is probable to conclude that the failure during service was caused by accelerated corrosion of the weld and that the accelerated corrosion in that area was related to the structure of the weld metal. It would be informative to examine welds from the furnace belts that have not failed to further study the cause for the failure.

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The use of uranium oxide fuel and the dispensing with metallic fuel elements in the Hanford production reactors have been contemplated in the past but have not been examined in any detail. This report reviews the technical feasibility of using oxide fuel and discusses the reactor limitations and changes in plutonium and isotope production. These results are based on preliminary calculations and a research and development program would be required to establish the validity of the calculated values. The research and development program is estimated to require $3.6 million over a three-year period. Technical aspects of using oxide are discussed in general for N, K, and the smaller reactors. The variations in uranium-to-graphite weight ratio and operating conditions vary the limitations encountered with each reactor type with an oxide fuel cycle. Plutonium separation and uranium oxide recycle on site have been examined, and a technical review of the limitations and the additional process requirements are presented.

This document provides a discussion of Reactor Branch activities during the months of January 1965 through December 1965. (FI)

Pressure drop tests were performed in the 189-D-Hydraulics Laboratory to determine the number of K-Reactor solid aluminum pieces (SAs) required to throttle the tube flow of a 46-piece KVE and a 38-piece KVN arch-rail I&E fuel charge to that of a corresponding bridge-rail charge. In addition, the effects on Panellit pressure and tube flow were determined for several cases.

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Process and Reactor Development progress includes depleted uranium irradiation in the B and KE reactors and thoria irradiation in the B, C, D, KE, and KW reactors. Reactor Engineering reports on higher graphite temperature tests, graphite strength tests, heat removal requirements following reactor shutdown, and reactor deactivation. Reactor Physics topics include: E-Q loading, test facility loss of coolant protection, ball 3X replacement, and in-core flux monitors. Radiological Engineering reports on radiation control experience, classification, radiation occurrences, effluent activity data, radiation standards and controls and procedures, and dose rates in the DR reactor block. Also listed are many revised standards issued during the report period. Operational Physics Operations report on pile physics plant assistance for the B, C, D, DR, F, H, KE, AND KW reactors. Process Physics studies include reactivity and control studies and product related studies.

This report shows the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials.

A two dimensional photoelastic study of two designs of clamps for the Phoebus 2 main closure is detailed. Predicted prototype stress data is given in the form of a stress parameter which includes the axial load and a representative clamp radius.

This report provides the minutes of the Graphite Distortion, VSR, HCR Serviceability Program meeting on February 4, 1965.

A program is being conducted to develop a process for cladding tungsten and tungsten cermet fuels with tungsten deposited from the vapor state by the hydrogen reduction of tungsten hexafluoride. Early work was performed using recrystallized, high purity, commercial tungsten as the substrate material. Temperatures in the range 660 to 12950F (350 to 1700°C) and pressures from 10 to 350 mm Hg were investigated. Hydrogen to WF 6 ratios of 10: 1 to 150: 1 were utilized. Efforts were directed toward optimizing deposition process parameters to attain control of qualities such as coating thickness, uniformity, density, impurity content, and surface quality. Substrate penetration methods have been investigated in the interest of completely eliminating the interface between the fueled substrate and cladding. In addition, the effects of process parameters and post-cladding heat treatments on the fuel retention properties of clad composites at 4500 degrees F (2480 degrees C) in hydrogen for 2 hours have been evaluated. As a result of work performed during the first phase of the program it has been shown that the rate of deposition of tungsten from WF 6 and the uniformity of the deposit can be varied in a predictable and reproducible manner by exercising control …

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This report addresses the calculational models for beta dose from spherical particles.

This report addresses criticality hazards control procedure high temperature test room engineering mechanics laboratory.

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Declassified 26 Nov 1973. Development of an extrusion-based process for producing beryllia fuel elements for the Pluto nuciear powered ramjet is described. The goal was to extrude and fire hexagonal tubes to close dimensional tolerances without grinding. Sinterable beryllia powder was fueled by precipitating uranium and fuel additives from a single aqueous solution in which the beryllia had been slurried. After sintering, the fuel phase consisted of a phasestabilized solid solution of urania (UO/sub 2/) with yttria and zirconia. Principal operations consisted of slurrying and precipitation, filtration, drying and calcination, vehicle incorporation and mixing, extrusion, drying and vehicle burnout, and hydrogen sintering (including straightening). Directions for continued development toward the no grinding'' goal are pointed out. (auth)

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