Hanford Laboratories Monthly Activities Report March, 1963

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P. R. No. Location Route Date Signature and Date This document consists of 216 pages. Copy No. of copies. Series -iHANFORD 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(45-11-1350. 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 Auxiliaries Operation

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BUDGET AND COST SUMMARY March operating costs totaled $2,400,000, a decrease of $200,000 from the previous month; fiscal year-to-date c o s t s are $21,156,000, o r 6970 of the $30,857,000 control budget. The control budget includes an increase of $525,000 received in a recent RLOO-AEC financial plan. However, a reduction of $235,000 in the Gas-Cooled Power Reactor Program was not made because the AEC has indicated intent to restore a major portion of these funds. Hanford Laboratories' research and development costs for March compared with last month and the control budget a r e shown below: Month Month To Date Budget Spent RESEARCH AND DEVELOPMENT 1. Reactor and Fuels Radiometallurgical examination of prototype N-Reactor fuel elements irradiated t o a n average exposure of 1250 Mwd/ton continues to reveal no

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Two fluted single tube N-Reactor fuel elements successfully completed a second cycle of irradiation in the ETR. Exposure on these elements is now 400 Mwd/ton. T w o irradiation tests of fluted N-inner components continued with the exposure in these two tests now being 1300 Mwdhon. nents has been completed. These elements are being evaluated for i r r a d i a - tion test purposes. Analysis of N-Reactor fuel closures indicates a factor of three t o four reduction in stress level is possible by proper shaping of the closure. Tests simulating the wear of N-fuel supports on process tubes continued to show a minimal w e a r after four series of tests. Other charging tests of N-Reactor fuel elements w e r e performed to evaluate the effect of the magazine-to-nozzle transition and the nozzle-to-pressure tube r o l l joint on the suitcase handle supports. Gouging noted on the shoes of two charges was attributed t o the soft iron shims used between the supports and support shoes t o provide the proper support height. A program was initiated this month t o study the irradiation performance of tritium target materials under N-Reactor conditions. of double clad target rods are being fabricated with a 0. 91 and 2. 10 w t 70 cans. Aluminum end caps must now be welded in place and the rods a s s e m - bled into a second can of Zircaloy-4. Billets have been prepared, fabricated, and sized into A1 Phosphorus i n quantities up to 400-500 ppm has been introduced into the enriched fuel stream for the N-fuel and for the present reactor I&E fuel as a result of scrap recovery process changes at NLO. Evaluation.to date indicates that phosphorus forms a eutectic between U P and U. This behavior is quite unlike that of other nonmetallic impurities such as C, 02, and N2. Careful evaluation of effects of phosphorus throughout the fabrication process will be required t o establish a meaningful specification.

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,Evaluation of M 4 0 H as a substitute f o r LiOH to adjust pH in N-Reactor w a s delayed by operating difficulties of KER- 1. While attempting to establish a satisfactory initial base operation before proceeding with the change from LiOH to NH40H, unexplained high crud levels w e r e observed. Corrosion penetration of aluminum coupon samples was less i n quachrom glucosate inhibited water than i n dichromate inhibited process water based on tests in the single pass KE tubes. The distribution of radioisotopes deposited on fuel elements in a typical KE tube was determined f o r the first time. The activity distribution corresponded to the corrosion intensity. A procedure t o monitor sodium concentration and thus detect leakage Zircaloy-2 coupons with thick oxide films w e r e found to hydride in 1 month in a He-H2(37o)-CO (2%)-H20 ( 0 . 4 m m ) gas mixture a t both 375 and 425 C, w h e r e a s thinly filmed samples did not hydride. This result indicates H 2 0 inhibition of gas phase hydriding of Zircaloy-2 process tubes may fail if extensive corrosion produces a thick film. In studying the nickel plating of uranium, it has been discovered that gas pitting of uranium during the preplating electrolytic etch is reduced by ultrasonic agitation of the piece at 30 kc. Pits, penetrating one-third of the pipe wall, have been observed radiographically i n static sections of PRTR softened water and PRTR process water systema. No pitting is observed in sections of these s y s t e m s where flow is maintained. Intital measurements of the kinetics of the water gas shift reaction An expression describing the burnout profile in D-Reactor was derived from rate expressions, temperature dependences, and reactor tern -

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EGCR graphite samples have been irradiated t o 1 . 4 x nvt ( E > 0. 18 Mev). Examination is not yet complete. Graphite samples compression-loaded t o 300 psi were found to have contracted to a greater extent than unstressed samples; however, the change was less than i n the first irradiation period. The gamma induced water-graphite reaction rate was found t o be proportional t o dose rate. The effect of vacuum degassing in lowering the initial r a t e of oxidation of graphite by C 0 2 w a s explained on the basis of removal of C 0 2 f r o m the active sites and slow re-establishing of the steady state condition. Rate expressions applicable to the nonsteady state period w e r e derived. disclosed s t e e p flux gradients over irradiation positions that could lead to serious errors in interpretation if monitors are improperly used. Flow characteristics w e r e determined for aluminum support pieces t o be placed in empty tubes at K-Reactor to decrease the flow rate and thus prevent cavitation in the inlet fittings. The boiling burnout data from a 0.050-inch spaced 19-rod bundle which had small "warts" to maintain spacing between rods were compared with previous data obtained with atest section using wire wraps on 12 of the rods to maintain spacing. very closely and indicated no effect of the wire wraps on boiling burnout. It was found that the burnout heat fluxes coincided Approximately 1990 two-phase, pressure - drop data points were obtained for typical reactor piping and fittings at flow rates between 500,000 and 4,000,000 lb/hr-ft and at coolant conditions between all liquid and 25% steam by weight. The General Electric Technological Hazards Council has endorsed increase of the PRTR heat transfer flux limit from 400,000 to 650,000 Btu/(hr)(ft 1, and of the tube power limit from 1200 t o 1800 kw (AEC approval

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valves open properly and correct operator action is taken at approximately the right times. The Council concluded that boiling convection cooling i n PRTR t is adequate in event of a total power failure provided the pressurize r relief AEC approval has been received t o operate the PRTR Fuel Element Rupture Testing Facility. Design and fabrication of the second generation mechanical shim r o d for PRTR continued with fabrication of the drive head approximately 7070 complete. A first generation shim rod will be tested in a shim rod environmental test facility recently completed. Experiments were continued to develop a procedure for monitoring zirconium concentrations in the PRTR coolant as a n indication of fretting corrosion. Duplicate analyses of one s a m p l e analyzed at an interval of 4 days differed by a factor of 5 0 , indicating that the zirconium tends t o precipitate o r adsorb on container surfaces. As a result, the validity of previous analyses is questioned. A radiation resistant television camera has been considered for Measurement of electronic properties of a special radiation resistant Vidicon showed no change after 1.1 x 10 R accumulated gamma irradiation. This substantiates previous results in which no degradation i n visual television picture quality was noted after this irradiation. successful irradiation of a television Vidicon tube. This is the highest known A &all hole through the cladding of a vibrationally compacted U 0 2 - PuOz fuel rod was found ina PRTR Mk-I fuel element that was suspected to be a leaker on the basis of stagnant water tests. No evidence of waterlogging was seen, and slight fuel washout occurred. Other suspected leaky elements were examined without discovering cladding defects.

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Center void and columnar grain formation, found in incrementally loaded, swaged U02 - Pu02 fuel rods for the first time, was accompanied by normal cladding behavior. Other out-of -reactor experiments confirmed that low temperature irradiation sintering observed earlier in swaged U02 PRTR fuel rods was not a gross thermal effect. element after irradiation to 1360 Mwd/to% revealed that interpal pressures of released fission gases and desorbed impurity gases w e r e less than anticA PRTR fuel element equipped with modified 360' contact-type end brackets and exposed for 2 weeks under PRTR conditions in an out-ofreactor loop, produced negligible fretting corrosion on the pressure tube. After 60 days of cooling, the rejuvenation t e s t fuel element (GEA 4-81) is being returned t o HL from NRTS for "recharging" with enriched material. Equipment for remotely disassembling and overhauling damaged PRTR Operational testing of the new manipulator system in the remote fuel fuel elements w a s fabricated and successfully tested in a simulated storage fabrication studies facility was nearly completed. Operation of the system is Ideal particle s i z e distribution for vibrational compaction of fuel in Very high density (9370TD) w a s achieved, a n annular test element was calculated from a model developed by the Applied Mathematics Operation. indicating the applicability of the model to novel fuel geometries. High densities (99.070TD) w e r e achieved in high energy rate compaction of U02 sintered scrap using higher impact pressures than previously believed possible with tool steel punches. Atest program for prototype EBWR U02 - Pu02 fuel rods was prepared. The first irradiation test capsules containing impacted U02 - Pu02were assembled and shipped t o NRTS for irradiation.

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Tungsten cladding components w e r e joined successfully by magnetic force welding without producing large recrystallized grains that normally appear in fusion zones. The first successful joining of W - U02cer met material was achieved by magnetic force welding. A technique was developed for fabricating 4-inch-diameter, solid cer met discs approximately 1 inch thick by impaction. W - U02cer mets completely encased in tungsten were formed by loading the cer met powders in the impaction container surrounded by a nonfueled layer of tungsten Both phases in UN-W cer mets reacted appreciably with oxygen at 1000 C, but little interaction occurred between phases. The melting point of UOS, prepared by deposition from molten salt, was determined as 1880 C. In cooperation with Chemical Laboratory personnel, simulated fission product oxides were impacted in both 2 . 5 - and 4-inch-diameter Self-damage of P u 0 2 and @-Pu2O3f r o m alpha radiation caused and 0.70 xlattice expansions of 0. 35 x Preliminary measurements of thermal conductivity were made on temperature range of previous measurements. As predicted, the thermal conductivity of the single crystal oxide was greater than that of the polyThe thermoelectric potential of U02 was successfully measured inThe polarity of the thermoelectric effect reversed from p- tor e actor. n-type as the temperature increased.

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U02 microcores, as small as 0.015-inch diameter, were extracted from U 0 2 samples by ultrasonic drilling. This new technique permits greater accuracy in studies of fission product relocation. Fabrication of 11 Zircaloy-2 clad thorium-uranium fuel elements tion testing in the ETR. These elements will be autoclaved i n preparation for ir radi a - An optical s t r a i n readout system has been developed which has no real limit t o its range, a characteristic that will be extremely useful in measuring the largest rains involved in stress rupture testing. Metallography samples from the AIS1 316 SS inner liner of heliumcooled DR-1 gas loop were examined. Samples from the high flux-high temperature region, 100-1200 F, showed a large amount of second phase precipitate at the grain boundaries. Samples from lower temperature, 800-1000 F, no flux region, were f r e e from this second phase. Oxidation tests of Hastelloy X, a heat resistant nickel-base alloy, in laboratory air and pure oxygen have been made at 1000 C on as-received and abraded sheet specimens. Weight gain curves for all tests deviate. no more than 15% from one another for the 48-hour duration of the tests. The occurrence of minimum hydrogen pickup for 10% cold-worked Zircaloy-2, in-reactor, has been further explored by x - r a y diffraction studies which showed a highly oriented texture in annealed material with the basal planes predominantly parallel to the rolling plane and a reduced intensity of orientation i n this mode. Two thorium specimens irradiated t o burnups of 0.18 and 0.92 at.% at e 3 0 0 C are being vacuum annealed at 7 5 0 C for 100 hours. hardness, and density will be determined. Notched beams of 23% cold-worked Zircaloy-2 containing a sharp It was observed that a sharp transition from ductile to brittle fracture fatigue crack w e r e tested in bending at temperature intervals from -196 to

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occurred from -60 to -80 C, and that the range w a s the s a m e for specimens cut i n either the longitudinal or transverse direction of rolled sheet. In an in-reactor creep test on 207'0 coldworked Zircaloy-2 at 20,000 psi stress and 350 C, the creep rate during a 10-day down-period which occurred shortly after the capsule was charged did not show the sharp increase typical of tests run previously at 30,000 psi. Tensile testing has been conducted on several samples of iron of varying impurity content to characterize the role of interstitial impurities in the deformation mechanisms of unirradiated iron. of these samples showed marked differences in behavior which were dependent on the purity. The yielding behavior Procurement of materials for the coordinated Irradiation Effects on Reactor Structural Materials Program is continuing. Plates 4 inches thick representing 18,000 pounds each of A-212B and A-302B pressure vessel alloys were received on site. Strip and rod of AIS1 406 SS have been ordered and bids have been received f o r sheet, plate, and rods of AB1 304 SS, X - r a y lattice parameter …