Hanford Laboratories Monthly Activities Report May, 1963

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Location Route Date Signature and Date This document consists MONTHLY ACTIVITIES REPORT HAN. 8.5454 C om piled by HANFORD ATOMIC PRODUCTS OPERATION This report was prepared only for use within General Electric Company i n the course of work under Atomic Energy Commission Contract AT(45-1)-1350. Any views o r opinions expressed in the report are those of the author only.

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Atomic Energy Commission, Richland Operations Office F. P. Baranowski, Director, Division of Production Savannah River Operations Office, Aiken, South Carolina

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.Force Report and Personnel Status Changes Reactor and Fuels Laboratory Physics and Instruments Laboratory Applied Mathematics Operation Radiation Protection Operation Finance and Administration Operation Test Reactor and Auziliaries Operation

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. BUDGET AND COST SUMMARY May operating costs totaled $2, 689,000, a n increase of $136,000 over the previous month; fiscal year-to-date costs are $26, 398,000, o r 86% of the $30,536,000 control budget. Hanford Laboratories' research and development c o s t s for May compared with last month and the control budget are shown below: Month To Date Budget RESEARCH AND DEVELOPMENT 1. Reactor and Fuels N o detrimental effects from phosphorus impurity in uranium have been observed in clad-to-core bonds o r braze-to-core bonds i n N fuel. Phosphorus lowers the room-temperature ductility and ultimate strength of uranium, but tests at temperatures of 100 C and higher show no differences due to phosphorus. A study of the effect of sulfur impurity in uranium has - Outer components forN fuel continued t o appear satisfactory i n performance and behavior as shown by the postirradiation examination of three N outer fuel element components.

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The high temperature swelling performance of uranium fuel elements has tended t o verify that fuel swelling is ultimately limited by coolant pressure. Fluted, single tube fuel elements (N -Reactor size) operating under N-Reactor conditions continue to perform satisfactorily. A companion element failed during irradiation ina cold water loop. Clad thinning, with intentional internal striations, has been exper - imentally produced during irradiation of experimental capsules. Uranium samples containing dispersions of UA12, U6Fe, uranium carbide, and uranium phosphide are being studied as alternate core materials. Only slight differences have been observed between Zircaloy-2 and Zircaloy-2 containing up to 1 wt% uranium in corrosion tests in 360 C w a t e r Damage through wear on N-Reactor process tubes caused by fuel element supports after s i x simulated charging operations, with intervening autoclaving between each charging operation, has scratched the process tube t o a maximum depth of 2 . 2 mils. This rate of w e a r is five times that of tests made on a tube without intermediate autoclaving. A theoretical study has been completed on the effect of process tube vibrations and their relation t o fuel element support design. Multiproduct fuel element development is proceeding through a study of fabrication methods, recrystallization temperatures, diffusion of lithium in aluminum, and out-of-reactor target element defect tests. Metallographic examination of N-Reactor self-supports on a Zircaloy-2 clad heater rod exposed f o r 146 days at 315 C in lithiated water at pH 10, and at a heat flux of 280,000 Btu/ft h r , revealed accelerated corrosion at each of the crevices (eight) between the self-supports and the jacket. Maximum penetrations were about 6 mils.

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Significant fuel element self-support flattening was encountered in N-Reactor charging machine tests when fuel columns were charged with 200 to 500 pounds back-pressure, but when the process tube w a s loaded with fuel elements to provide prototypical back-pressure the self -support flattening w a s reduced from about 25 mils t o about 5 mils. Equations have been developed for application of laboratory boiling Based on test operation of K-1 Loop f o r about 30 days with the pH controlled with NH40H, ammoniated systems control crud better than lithiated systems at low temperatures, probably because of hydrazine formation which getters oxygen in the system. The K-1 and K-2 Loops w e r e decontaminated sufficiently to allow maintenance operations during tube replacement in 105 KE. (stainless steel) was treated with alkaline permanganate followed by sulfamic acid; the K - 1 Loop (carbon steel) was decontaminated using a single-pass procedure with inhibited ammonium citrate solution. An unautoclaved section of 3070 cold worked N-Reactor pressure tubing exhibits a primary creep rate about five to six times that of autoclaved tubing, and a much higher secondary creep rate. effect of autoclave treatment was previously noted on annealed Zircaloy-2 strip material tested under uniaxial creep loading conditions. Results of the N-Reactor atmosphere computer program have been correlated by means of a n empirical expression which relates maximum graphite burnout rate, maximum graphite temperature, initial water vapor partial pressure, and initial hydrogen partial pressure. t o be significant above 500 C during experiments i n which He, CO, H20 gas mixtures were allowed t o diffuse through hot graphite thimbles. experiments in which the gas mixtures flowed past a graphite b a r little

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reaction occurred. It would appear that the water-gas shift reaction could produce significant amounts of hydrogen in the N-Reactor tube block, but probably not i n the gas channels. The in-reactor creep rate for 2070 cold worked Zircaloy-2 after 1700 hours at’20,OOO psi stress and 350 C was determined t o be 1.8 x in. /In. /hr. The rate during r e a c t o r outage was essentially ‘the same. An error in the out-of-reactor rate measurement has been corrected and a rate of 0.5 x in. /in. / h r at 2000 hours has been determined. In-reactor c r e e p capsules have been shown capable of operation at 900 Cina laboratory mockup. test stainless steels. Capsules with this capability will be used t o Zircaloy-2 bend-test specimens with 23% cold work have a ductile-tofracture surfaces by electron microscopy reveal similar crystallographic -0I.& features for both transverse and longitudinal specimens. In-reactor weight gain data have been obtained for Zircaloy- 2 13 samples after 125 days in 540 F (282 C) water and a fast flux of 7.6x 10 nv. The average weight gain of 1 2 1 mg/dm compares with 9 mg/drn2 for reference coupons exposed in the out-of-reactor loop at similar conditions. A proposal was prepared for experimental evaluation of N-Reactor shielding during plant s t a r t u p using the top shield plug facility of the N-Reactor. The MAC shielding code has been modified slightly to eliminate a n e r r o r in the gamma calculation, and the BARNS code has been debugged to calculate unit weighted shield material cross sections. The gamma- induced oxidation of graphite by low concentrations of water in flowing helium was found to be independent of flow rate. In studies of the inhibition of oxidation of graphite by halo-methanes, the rate of oxidation could be related to surface areas as measured by the B E T method.

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DECLASSI FI ED The flash thermal diffusivity technique was successfully used t o determine thermal conductivity of porous carbon having a value only about 1 / 3 0 of that of CSF graphite. The use of a lead sulfide detector t o monitor the rear surface temperature rise has not been wholly successful. In a half-reactor test the corrosion rates of reactor materials appear t o be about the s a m e in low alum (7.5 ppm) as in high alum (15 ppm) water, except for type 1100 aluminum (tube material). slightly more in the higher alum concentration. This aluminum corroded The General Electric Technological Hazards Council accepted the final response to the audit of PRTR operation, approved the discharge method for the PRTR Fuel Element Rupture Test Facility, and approved modification of the PRCF for light-water-moderated critical tests. The AEC approved a proposed increased of limits on PRTR maximum heat transfer flux and tube power.+ Revised PRTR aqueous effluent activity limits have been developed -63 which will become effective when the new water treatment plant at Richland By means of a new recommended procedure, ..recovery of the PRTR primary coolant system from boiling convection cooling could be accomplished without further l o s s o r degradation of D20 and with minimal thermal shock Detailed design of the second-generation shim rod assembly for the PRTR is 90% complete, and fabrication of the two identical prototype units is

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'The entire PRTR fuel lsading was examined in the basin, and ten selected elements were examined in the Fuel Elcment Examination Facility. In general, the fuel elements w e r e in good ccnditfcn. Defects included: scratches on exterior surfaces of many of the elements, displacement of some circumferential bend-type bundle wraps, and loosening of rod wire wraps on s o m e of the A1-Pu spikes. The clip-on wear pads on the V02-P-.i02 e Le m ents are perf o r ming very satisfactorily , Individual fuel rods from a vibrationally ccmpazted U 0 2 - P u 0 2 , PRTR Mark-Z element that failed in the PRTR were exsmined i n detail. were detected in the heat-affected zone of the top end-csp welds of three rods. Likeiy causes of this locaiized corrosicn are eitner contamination of the ZI rca hy dcring welding o r the presence of some impurity In the ?del materia! which caused severe attack from the inside. fuel of the type under consideration for m e in the EBWR w e r e chzrged into the ETR cn May 13. w a s completed; the first cluster WGS charged Intq the PRTR. Gamma scEn measurements cf Pa conczntraticn correlated i 4 ith calibrated V02 - Pu02rcds in the range 0.1 to 10% Pu. The rejwenatlon fuel element r'GEH-4-81) is ready for return to the MTR. Rejuvenation was accomplished by adding enriched U 0 2 to a fuel rod from the fuel assembly which had been irrzdlated in Jangary in the MTR. A uniformity of Pu distribution of 3. 1 wtyo was cbtained In "bott2e

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Equipment has been developed and fabricated f o r rapidly positioning and tightening a cluster fuel element wire spacing member remotely, under approximately 10 f t of water. temperatures from 200 F (93 C) to 400 F (204 C))indicate that the burst strength decreases linearly with temperature over this temperature range. The ratio of thermal-plus -epithermal to fast neutrons w a s found constant over 7070 of the fuel length of the PRTR core. This ratio determined by nickel and cobalt wire monitors was 12.3 k 0.2. A Zircaloy-clad uranium honeycomb element has been successfully fabricated by coextrusion. T h r e e coextruded Zircaloy -clad thorium -uranium fuel elements have been irradiated to 1100 Mwd/ton with no appreciable swelling occurring. Irradiation will continue in the ETR loop. Autoclave tests show that thorium contamination of Zircaloy- 5 Be braze .0*..- has a definite detrimental effect on the corrosion resistance of the braze. Allowable levels of thorium contamination of brazing alloy are being established. In preparation f o r fabricating EBWR fuel, uranium dioxide was prepared and densified successfully by pneumatic impaction, and preparation of PuOz from 60 kg of P u metal was completed. The pneumatic impaction technique was used to densify 1100 g r a m s of PuO a t one time. Density achieved was 10.89 g / c c (95.170 TD). PUN pellets heated in flowing hydrogen for 12 hours at 1700 C slumped and flowed together, possibly due to the formation of free molten plutonium. Several plutonium -bearing borides were prepared by reacting Pu02, carbon, and boron at 1600 C i n vacuum. The density of pneumatically impacted PUN-50 vol70 W was 1 6 . 9 g / c m (10070TD). However, the P U N was not evenly distributed i n the tungsten.

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Thin wafers of P u - 1 5 did not diffuse into U 0 2 after heating f o r 12 hours at 1 7 0 0 C. X - ray diffraction data revealed no crystallographic changes i n Analyses of the first two samples of pneumatically compacted P u 0 2 revealed densities of 8 9 and 95% TD with oxygen-to-plutonium ratios of 1 . 9 4 AU02 - 5 0 wt% W cer met was examined by reflection electron microscopy during controlled heating to 1900 C. and one W-UN) w e r e charged into the ETR and are to be irradiated with fuel surface temperatures in excess of 2000 C. A tungsten-clad U 0 2 capsule designed to operate with a completely molten core was charged into the ETR on May 13. Chemical analysis of UOS powder prepared from a fused chloride melt revealed an uranium content of 8 4 . 7 wt% (vs theoretical of 8 3 . 2 wt70). The absorption coefficient for single crystal UO, was measured at wavelengths between 0 . 6 - 1 5 microns. T h e r e is a large optical window in the range 3 - 1 3 microns, with a minimum at 1 1 microns. The relative transmission of a laser beam (wavelength 0.69 p) was significantly greater i n single crystal U02 than in polycrystalline U02. Transmission at this wavelength decreased with increasing temperature to exhibits enhanced heat transmission, relative t o polycrystalline UOz, due to radiant heat transmission.

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A single crystal of U02was irradiated for 310 days, to an estimated Autoradiographs of fuel elements defected during irradiation revealed significant relocation of fission products in the immediate area of the defect. Laboratory data for air cooling of 19-rod fuel bundles in the F u e l Element Examination Facility w e r e correlated using the Colburn equation for forced convection heat transfer with a modified velocity term. Eighty-eight boiling burnout data points obtained with 19-rod bundle test sections w e r e correlated fairly well by a new empirical method appearing in the May, 1963, issue of Nucleonics. Data obtained with long (76-inch) test sections w a s not published. Employing fog cooling (steam -water mixtures) 42 experimental boiling points were determined with a n electrically heated 19-rod bundle test section. The burnout heat flux at low flow rates w a s quite sensitive t o the hydraulic stiffness (i. e. , availability of excess inlet pressure) of the system. X - ray examination of molybdenum foils annealed at 900 Cfor2 hours line widths have returned t o preirradiation values. 200 ppm carbon and 400 -500 ppm carbon have apparently recrystallized t o some extent, as ratios of peak intensities are different from the corresponding ratios in the unirradiated foils. The foils containing 100Defect structures have been observed i n irradiated molybdenum foils by transmission electron microscopy. causes dislocations t o move i n straight line channels. interact with the defect structures sweeping them from the channel. Deformation of the irradiated foils The moving dislocations To prepare foils for transmission electron microscopy, spark machining and a n electrolytic jet polisher have been successfully employed to cut wafers froma single crystal of molybdenum.

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A molybdenum single crystal annealed at 2000 Cfor1 hour was used …