The objective of this supplement to Production Test IP-584-D is to authorize low exposure (5--8 GVR) irradiation tests of the following lithium-containing target materials: Magnesium -- 14 percent lithium alloy, Lithium-aluminate (LiA1O2), Lithium-silicate (Li2SiO), Lithium-aluminate -- aluminum cermet, Lithium-silicate -- aluminum cermet. Lithium target materials will be contained in 4.50 inch long aluminum cans which are clad with 35 mil Zircaloy-2 alloy. The target elements will be contained in 1.9 inch OD, 1.5 inch ID Zircaloy-2 flow distributing sleeves. The target element assemblies and N Reactor inner-fuel elements (NIE-1) in Zircaloy-2 sleeves, will be irradiated in KER-1 or KER-2 at operating conditions approximating N Reactor operation.

This invention report suggests a method for drastically increasing the amount of Pu-241 in isotopic mixtures of plutonium. It is felt that with process experience, as much as 70 percent or more of the potential Pu-241 atoms can be concentrated in the fuel at one time. Such a concentration step would double the amount of production obtainable from Pu-241 by nuclear decay. The process to concentrate the Pu-241 consists of two basic steps: 1. Irradiate plutonium consisting largely of Pu-239 isotope until the Pu-239 has largely been converted to Pu-240 and 241 by thermal neutron fission and absorption events. Most of the fuel value has been taken advantage of at this point. The isotopic mixture will be largely Pu-240 but contain smaller percentages of 241 and 239. The irradiation is terminated at this point and the products are separated. 2. The depleted plutonium isotopes from the first irradiation are refabricated for a second irradiation with a thermal neutron absorber surrounding the depleted plutonium isotope. This element is again irradiated, preferably in an epithermal neutron flux with a peak energy slightly above 1 ev. The Pu-240 has a huge resonant cross section at the 1 ev level and will be most …

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locations in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.

This report provides the Hanford Atomic Products Operation plant acquisition and construction budget for fiscal year 1966 and the revision of budget for fiscal year 1965.

This monthly technical report details activities of the Reactor Physics Research and Engineering staff of the N-Reactor Department for the month of February 1964.

The horizontal control rods presently installed in the older Hanford Reactors have inadequate heat transfer characteristics for present and predicted future operation of the reactors. Continued graphite stack distortion, coupled with higher graphite temperatures, has resulted in ECR failure during reactor operation, such as swelling of the outer aluminum sheaths to the extent that rod movement in the graphite channel is severely restricted. Continued graphite stack distortion will tend to further aggravate the problem of rod operation. A new HCR design concept,was developed by P. H. Hutton of Reactor Design, IPD, to alleviate some of the pressing operational problems. Prior to the acceptance of such a design, the important rod operating parameters should be known to some degree of accuracy. This study was conducted to detexmine, by calculational methods, the temperature distributions that could be expected to occur in such an HCR when used at the present operating power levels and at 120% of the present power levels.

The Hanford reactors contain a very large inventory of fission products and it is very important that this inventory be contained. In recognition of this, rather stringent reactor cooling requirements have been established for the Hanford reactors. This includes the requirement of three independent cooling systems, including the system for full level operation. The second and last-ditch systems both are usable only after a reactor scram and take-over automatically from the flywheel decay of the process pumps. In 1962, two projects were completed that affected the last-ditch cooling adequacy of the K reactors, One project increased the reactor pumping capacity and, therefore, its power level; and the other considerably modified the emergency cooling systems. Since this time, three tests have been conducted to determine this last-ditch cooing adequacy. In order to provide a uniform approach to this adequacy determination, the method employed by Reactor Engineering is presented herein. The report presents a method for determining the maximum reactor power level for which measured last-ditch flows will provide adequate emergency cooling, A number of pertinent charts are included, A method is also presented for determining the last-ditch flow that should be measured during a test for any desired level of normal …

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A number of documents dealing with the Hallam moderator element development and design were critically reviewed. Comments clarifying some specific points of the reviewed documents were listed. Critical areas needing further investigation were depicted and a brief outline of the required analysis given.

The torquemotor, potentiometer, and piston-seals used in the first Generation Control Actuator were tested operationally to 10/sup/16 nvt at approximately 560°R.

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A Third Party inspection of the reactor work platforms was conducted by representatives of the Travelers Insurance Company in 1958. An inspection report submitted by these representatives described hazardous conditions noted and presented a series of recommendations to improve the operational safety of the systems. Project CGI-960, ``C`` & ``D`` Work Platform Safety Improvements -- All Reactors, vas initiated to modify the platforms in compliance with the Third Party recommendations. The American Standard Safety Code for Elevators Dumbwaiters and Escalators (A-17.1) is used as a guide by the Third Party in formulating their recommendations. This code is used because there is no other applicable code for this type of equipment. While the work platforms do not and in some cases can not comply with this code because of operational use, every effort is made to comply with the intent of the code.

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A model is proposed which depicts the sources of elutriatable fines in the fluidized bed calcination process as being in two major groups, spray drying mechanisms and attrition mechanisms. Based on this model, equations are derived which express the rate of change of the concentration of a chemical tracer material in the elutriated fines, following introduction of the tracer into the feed and following its removal from the feed. This system has been simulated on an analog computer, and by matching the computer simulation to results from an actual calciner run, the rates of generation of fines by each of the two groups of mechanisms has been determined; the same technique results in an estimate of the amount of these fines remaining in the fluidized bed. Agreement between postulated results and results of actual experimental tests lends credence to the usefulness of this analytical technique. (auth)

Isostatic-pressing techniques using reversible gels as the pressing medium were studied for improving the uniformity in density and structure of UC-- ZrC bodies. Control of powder-size fraction was studied as a means for controlling the pore distribution. Control of the carbon content by thermal treatment in a mixture of H/sub 2/ and hydrocarbon was also studied. Near stoichiometric 30 UC--70 ZrC powder was prepared by gas-metal reaction. Assembly of an apparatus for study of the thermochemical vapor-deposition of tungsten is near completion. The vaporization and fission product release rates of a hot- pressed high-density 30 UC -70 ZrC sample were measured from 1800 to 2000 deg C. A low-pressure gas adsorption apparatus was set up for measuring the true surface area of UC-ZrC samples. The cell used for the study of fission product diffusion through tungsten was fabricated. Samples are being prepared for fuel-clad compatibility and refractory-metal interdiffusion studies. Diffusion-emission studies were made on a rhenium-clad UC sample at 1800 deg C. The molybdenum pedestal of the loading device of a high-temperature mechanical testing furnace was modified. Thermionic emission microscopy showed that areas of high work function can co-exist with betteremitting UO/sub 2/ dispersions on the surface of a W-UO/sub …

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