Hanford Laboratories Monthly Activities Report June, 1963

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This'report was prepared os an account of Government sponsored work. Neither the United States, nor the Commission, nor ony person acting on beholf of the Commission: Makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, o r usefulness of the information contained in this report, or thot the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report. As used in the above, "person acting on beholf of the Commission' includes any employee or contractor of the Commission, or employee of such controctor. to the extent that such employee or contractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with

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P. R. No. Location Route Date Signature and Date This document consists -,/ MONTHLY ACTIVITIES REPORT &J-856Y 2 HANFORD ATOMIC PRODUCTS OPERATION This report was prepared only for use within General Electric Company in 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 . Acting Manager, M. Lewis Radiation Protection Operation . Finance and Administration Operation . Test Reactor and Auxiliaries Operation

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BUDGET AND COST SUMMARY Operating cost figures for June and for the completed fiscal year are not yet available and will therefore be included in the July report available funds within designated budgetary controls. It is believed these figures will reflect effective utilization of RESEARCH AND DEVELOPMENT 1. Reactor and Fuels A test lot of NIE elements with self - b razed closures was completed for irradiation testing. on NOE fuel elements. Development work has been started for this closure Development of the hot-headed projection welded closure has included evaluations of: (1) increasing the width of projection welds, (2) varying postheat time and current, and (3) varying the amount of copper in the joint The performance of N-Reactor fuel components irradiated in the KER loops continues satisfactory, as shown by post-irradiation examinations. Both bonded and unbonded N-inner fuel elements with tapered rather than flat end caps were satisfactorily irradiated to 1900 Mwd/T in KER with this end cap design in either the bonded o r unbonded condition. No evidence of clad shearing at the cap - core interface was observed Fluted N-Reactor size single-tube fuel elements continued to operate satisfactorily under N- Reactor conditions. Precipitation of hydride platelets in the N-inner fuel supports during autoclaving has resulted in cracking of supports during sizing. solution anneal followed by quenching will restore the normal ductility of the

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Irradiation of Zircaloy- 2 clad lithium- aluminum target alloys in KER Loops 1 and 2 was started. the performance of the target material under N- Reactor conditions. These tests were designed t o establish The conditions resulting in recrystallization of cold and warm extruded aluminum-lithium material were determined. Irradiated test specimens having intentional striations in the Zircaloy-2 cladding show a high incidence of localized clad thinning at an Initial study of sulfur impurity in uranium indicates some solubility of sulfur in t h e gamma uranium phase. Uranium and uranium-zirconium alloys containing aluminum, iron, and phosphorus are being studied as dispersed phase fuel materials. In tests up to 390 hours, Zircaloy-2 containing up to1.0 wt 70 uranium exhibits corrosion behavior similar t o Zircaloy-2 in 360 C water. Zirconium containing 250-2500 ppm uranium with 1, 2, 3, 4, a n d 5 wt 70 beryllium shows decreasing corrosion resistance with increasing beryllium after 100 hours in 360 C water; above 3 wt 70 Be the samples developed white oxide in these tests. The major cause of N-fuel self-support flattening has been tentatively traced to using a charging machine magazine which is slightly larger in inside diameter than the process tube. magazine, no significant support flattening occurred at column loads up to 1500 pounds, approximately twice t h e anticipated load at N-Reactor. In tests employing a process tube as the N- Reactor Z irc alo y process tube material is remarkably resistant to brittle fracture originating froma 0.360-inch long crack introduced while the tube is a t elevated pressure. At room temperature the tube must be stressed tothe yield point (about 9870of ultimate) to cause such a crack to propagate in a brittle manner. At 150 C (302 F)t he crack will not prophgate even with the

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tube stressed to the yield point. room temperature crack propagation tests show that the moving crack tip travels at about 2500 f t / sec,or1 / 6the speed of sound in Zircaloy. High speed photographs of one of t h e In 360 C, pH 1 0 (LiOH) deoxygenated water,, pre transition corrosion product hydrogen pickup is greater for Z ircd oy - 2samples quenched from the beta phase than for alpha heat treated samples. unaffected by these heat treatments. The inhibition of gas phase hydriding of Zircaloy-2 at 375 and 425 C in 2 % H2, 1% COZYhelium gas mixtures :s spparently the result of water formation by the carbon dioxide-hydrogen reaction. The effect of a 100-fold variation in t h e carbon monoxide concent ration of a typical N-Reactor gas atmosphere was found to be very small in reducing the equilibrium water concentration. Irradiations of N- Reactor graphite are proceeding satisfactorily in Two predefected target elements tested in an out - o f - reactor loop at 520 Ffor 200 hours showed no swelling. coolant increased rapidly during the first48 hours and then decreased, indicating plugging of the holes with corrosion product. The aluminum concentration in t h e Tests to determine effects of residual decontaminants in crevices of stressed stainless steel samples showed no cracking after three months of exposure, but some inter granular attack was noted; pits were as deep as 8 mils. Lithiated water at pH 10 and 530 F flashed across the face of a Zircaloy-2 specimen, to simulate a leak from the N-Reactor primary system showed a white deposit after three weeks, but no accelerated corrosion or hydriding was noted.

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Irradiation of f o u r i&E fuel elements to a maximum exposure above 3000 Mwd/T at high specific power h a s been completed to evaluate t h e susceptibility of this fuel to grain boundary tearing. Examination of corr oso meter probes placed in the inlet piping and storage tanks of present production reactors to measure corrosion rate of carbon steel in process wafer shows that at low velocities neither pH ( 6 . 6 o r 7. O ? nor dichromate concentration (1. 0 o r 1. 8 ppm) affects t h e At high velocity (afps), an increase 111 corrosion rate was An unautoclaved, Zircaloy- clad fuel element is being irradiated in routine basis is dependent on mamtaining clean metal surfaces during fabrication, delivery, and charging of the fuel elements. Potential incentives include increased fuel element life resulting from reduced initlal hydrogen concentration in t h e cladding. and approximately a 5% reduction in direct fuel fabrication costs. I The feasibility of eliminating pre ir ra & atlon autoclaving on a Inspection of irradiated PR TR fuel elements continues t o reveal excellent external appearance with the exception of an occasional damaged w i r e rod o r Remote repairs will b e made before returning such elements to Postirr;Ldiation examinations of four failed, vibrationally compacted PRTR fuel rods in one fuel assembly revealed internal cladding corrosion at end caps distant from the failures. in the archive specimen of the fuel used in three of the rods is currently suspected to be the cause of failure, rather than poor weld quality. Moisture and / o r high chloride impurities A maximum fuel thickness variatjon of 1. 6 mils in the outer fuel annulus of aPRTR tubular element had no detrimental effect on the irradiation

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After 106 days of exposure of an experimental P R T R Mark I fuel element in an out-of-reactor fretting corrosion test loop, only super - ficial wear was observed on the ring contact s u r f a c e of the bottom end Fuel for an In-Reactor Enrichment Test Element was fabricated. Controlled release of the U02enrichment particles from the fuel reservoir was demonstrated with a mockup of the feed section. Preliminary analyses of micro drill samples indicate that the Of t h e fission products analyzed (Ce plutonium distribution is relatively constar-t in a (U, Pu)O pellet irradiated Zrg5-Nbg5, and Cs137) only Ru106 shows a very marked radial variation Four short, prototypical EBWR fuel rods containing vibrationallya f t e r reaching approximately 1850 Mwd/ Tfuel at anticipated EBWR operating Preliminary examination of two irradiated specimens of vibrationally performance at high heat ratings, similar to that of pneumatically impacted Analyses of micro drill samples of irradiated, high-heat rated MgO-Pu02 pellets confirmed significant radial fission product concent ration variations that resulted from migration of fission products o r plutonium in an im miscible fuel system. P u O was densified to 9570 of theoretical density by pneumatic steel cer mets for irradiation tests. pellets can be simultaneously impacted. A novel die was designed in which six

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was increased from 0.48 wt% to 1.0 wt%. The loss is apparently a preferential mechanical effect due t ofirmed. APRTR fuel element (vibrated for two weeks) with full-circle end 15 mils at four locations on a. rod wire wrap near the center of the fuel element. However, after 87 days of exposure there was no apparent fretting at t h e end brackets. Analysis of data from experiments with an electrically heated 19-rod bundle test section revealed that the pressure drop was about what was expected from calculations, but surface temperatures during local boiling were somewhat higher than would b e predicted using standard correlations. The delay i n the Doppler effect due tothe lag time in transferring heat from Pu02toU02 in mixed oxide fuels is being investigated. lation of this problem has been completed in preparation for analog computer Review of cyclic corrosion tests show that repeated use of APOX for decontamination should cause no serious corrosion problems. Elevated temperature burst tests of pressurized irradiated PR TR pressure tubes show that increasing the test temperature 100 F decreases t h e ultimate strength by about 10%. ,Three coextruded Zircaloy clad thorium-uranium fuel elements attained an exposure of 1680 Mwd/T. This irradiation will continue in the E TR Loop.

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The corrosion of thoria dispersion oxygen fits an equation based on parabolic hardened nickel in 600 t o 1000 C film growth plus a linear phase rapidly than TD nickel, by a generally parabolic Process. Wrought alloys, 99. 97 and 99.99770 Ni, corrode more h creasing the b T in the nonisothermal loop from8C to 55 C increases aluminum corrosion rates by a factor of 1.5 to 2 at 330C sample rates by a factor of 1. 5 to 2 . Increasing water pH from 5. 4to6.4 also increases corrosion Time-lapse photographlc recording of tensile tests on irradiated molybdenum single crystals has revealed that for nearly ai! orientations tested, an offset occurs in the crystal after initial yielding. proceeds, the offset is gradually reduced, but rupture occurs at the point of the offset. As plastic flow for high quality plutonium (less than 300 ppm total impurities). with previous T - T - T curves, these are shifted in position due to the higher purity of the metal used. The physical damage associated with the delta-to-gamma and g a m m a - to-beta thermal cycling of unalloyed plutonium was much less than the physical damage associated with beta-to- alpha thermal cycling. The density decreases were less than 0.01 g / cmper cycle for the former two, whereas the density decrease for beta-to-alpha cycling varied with cycling rates from 0.03 t o 0.4g / cmper cycle for metal of the same quality. Direct experimental evidence that recovery processes are responsible for tearing in irradiated uranium has been obtained. These specimens represent the first refractory alloys to be irradiated under the Irradiation Damage t o Reactor Metals Program at Hanford.

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Investigation of the effect of loading rate on 22% cold-worked Zircaloy-2 has shown that Charpy specimens undergo a ductile to brittle transition at -175 C in slow bending as compared to 300 C in impact. A 30 C increase in the transition temperature is realized fora 40-fold increase in loading rate of fatigue-cracked notch beams. These are significant rate effects and are attributed to a deformation mode ora combination of inter - acting modes, which become operative in the high stress field in the immediate region of the notch. High frequency induction heating and infrared radiztion have both been successfully used as alternate heat sources for controlled-atmosphere tensile testing of refractory metal alloys at temperatures of 1000-1500 C. Two additional i n - reactor creep tests have been started, making three capsules in simultaneous operation. worked Zircaloy-2 at stresses of 25, 000,a d 35, 000 p s i at 350 C 1662 F) and The two new tests are on 20% coldInitial creep rates in these tests are There was no apparent reaction between molten stainless steel and U 0 2 after 90 minutes at 1600 C. A PU ~0170P U N cermet was heated for two hours at about 1000 C in vacuum and showed no indicaTion of melting, vaporization, o r reaction with t h e alumina crucible. Specimens of tungsten carbide, zirconium carbide, tantalum carbide, hafnium carbide and a mixture of two parts hafnium carbide with one part tantalum carbide were densified by pneumatic impaction at 1200 C and A sample of plutonium carbide being irradiated at the MTR has reached an estimated exposure of 6 x 1020 fission s / cc. The melting points of polyphase PUB, PUB*, PuB4, and PuB6were determined to be 2030, 1825, 2050, and 2100 C, respectively.

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Square honeycomb cer met grids of godd structural quality were fabricated by pneumatic impaction. A fiber reinforced cermet disc was formed by pneumatic impaction of a mixed bed of tungsten fibers and tungsten-20 ~ 0 1 %U 02cer met powder. Two stror,g infrared absorption bands, at 5.5 and 1 0 microns, were found i n sir-gle crystal U02 and appear to b e independent of oxygen-toAn apparatus tc mez sure the electrical coriductivity of U02 at tem - Electrolysis of liquid U 0 2p era ure s up to 3000 Crap put into operation. was observed during +he first run. Vacuum insulated capsules containing Mo-U02 and Mo-UN cermets cry st hl clad in WPere successfully irradiated in the ETR. temperatures were 2al:uia:ed …