Hanford Laboratories Monthly Activities Report April, 1963

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Route To: P. R. No. Location Route Date Signature and Date This document consists MONTHLY ACTIVITIES REPORT HANFORD ATOMIC PRODUCTS OPERATION PRELIMINARY RE PORT 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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TABLE OF CONTENTS - Force Report and Personnel Status Changes .iv Reactor and Fuels Laboratory Physics and Instruments Laboratory Applied Mathematics Operation Radiation Protection Operation Finance and Administration Ope rat ion Test Reactor and Auxiliaries Operation

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April operating costs totaled $2,553,000, an increase of $73,000 over the previous month; fiscal year-to-date costs are $23,709,000 o r 7770 of the $30,857,000 control budget. Hanford Laboratories' research and development costs for April compared with last month and the control budget are shown below: RESEARCH AND DEVELOPMENT 1. Reactor and Fuels The effects of several hundred ppm phosphorus impurity in uranium on the uranium structure, clad to core bonds, b raze to core bonds, and fabrication.have been experimentally studied. No detrimental effects have been observed t o date. Volume increase resulting from grain boundary tearing in uranium under irradiation is being studied in Hanford I&E fuel elements. T e s t elements irradiated to 1000 Mwd/ton with a maximum fuel temperature of 400 C have shown no visual changes; irradiation is being continued.

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modate substantial swelling of uranium are now in their third cycle of irradiation i n the ETR. An N inner-size fluted element has sustained a 3% volume increase without difficulty and is being further irradiated. .d Clad thinning is being studied i n a series of Zircaloy-2 clad uranium - r o d s being irradiated i n NaK capsules. Nine capsules have been discharged at 1100 Mwd/ton for examination t o determine irradiation goals for the Design studies of shaped end caps f o r N-Reactor fuel element closures indicate that a major reduction in stress levels can be obtained with properly contoured end caps. The exposed braze layer of a self-brazed closure has withstood 100 days of exposure in a high temperature test loop without apparent attack. T w o lots of elements with self-brazed closures for a n N-Reactor production test are in the final stages of preparation. Further development of the hot-headed closure has virtually eliminated the previously existing joint crevice; the immersion copper plating procedure has been improved; and the four ring projection weld c a p has produced sound welds with bonded interfacial areas. Damage to N-Reactor process tubes by wear of fuel element supports a f t e r five simulated charging operations (with intervening autoclaving between each charging operation) has scratched tubes t o a maximum depth of 1 . 7 mils. Average wear for the last scratching cycle was 0.3 mil. Wear occurred only after deliberately starting the scratching process. Multiproduct fuel element development is proceeding with the fabrication of 2 1 experimental lithium-aluminum target cores jacketed i n aluminum. Eleven of these are being directly canned in Zircaloy-4 and the remaining 10 will be anodized before Zircaloy-4 canning.

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Teat fabrication of N-Reactor fuel element supports continued; substantial amounts of this work will be transferred to NRD in the near future. A high temperature induction furnace is being operated at temperatures in excess of 3000 C using a newly designed magnetic flux concentrator. A Zircaloy-2 clad heater rod assembly exposed t o simulated N-Reactor coolant at a heat flux of 280,000 Btu/(hr)(ft ) for 143 days showed extensive corrosion and oxide s c a l e formation in crevices. The first instances of significant crud deposition on Zircaloy-clad fuel elements operating in high pH water were noted in recent KER t e s t s recently cleaned loops. These depositions may be related to off-standard water quality o r No effect of heat treatment on the corrosion of Zircaloy-2 exposed t o 360 C, pH 10 (LiOH) deoxygenated water w a s observed after 2 weeks of fraction w a s higher for beta than alpha heat treated samples. However, it was noted that the pretransition hydrogen pickup continued. In out -of-reactor tests, corrosion rates of N-Reactor coolant system structural materials were no greater in NH40H treated water than in Studies to evaluate fretting tendencies of N-Reactor fuel elements continued; different spring supports and external vibration frequencies were Evaluation of the Hays dissolved oxygen analyzer with influent streams containing dissolved oxygen and hydrazine demonstrated that the analyzer can measure oxygen concentration accurately when hydrazine is present. This instrument will be used on N-Reactor water. A newly developed decontamination procedure, the use of alkaline permanganate followed by sulfamic acid, was recommended for decontamination of an ETR loop and performed well.

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To quicken rupture, stress rupture tests on N-Reactor process tube sections have recently been conducted at temperatures up t o 400 C. The Larson-Miller parameter was then used t o estimate time t o rupture at lower temperatures for the s a m e stress level. By this means, a rupture time of 20 years may be predicted for 3070 cold-worked N-Reactor tubing at 300 C and a hoop stress of 48,000 psi. Operating N-Reactor loop stress is only about 12,000 psi. The effects of high flow rates of water on aluminum have been s i m - ulated by spinning aluminum discs i n water at velocities up t o 7 2 fps. Substantial erosion-corrosion attack is achieved at the outer perimeter of observed on introduction of dichromate. Definite changes in the type of attack are Monitoring of graphite i n KW-Reactor from December 1962 t o April 1963 disclosed the very low average burnout rate of 1.570 p e r 1000 operating days at positions in the reactor normally associated with peak burnout rates. Preliminary information has been obtained from analog and digital computer programs which shows the expected gas composition for are actor traverse as a function of entry composition. The calculated values Qf initial water content for burnout rates of 2, 3, and 470 per 1000 operating days were, respectively, 0.47, 0.71, and 0.94 mm. Capsule H-5-2 of the N-Reactor graphite long t e r m irradiation series was removed from the GETR after 5 cycles of irradiation. Cumulative exposure, estimated at 5 x 1021 nvt (E > 0.18 Mev), has been obtained on some of the samples. Sample H-4-3 was installed in the reactor and sample H-6-2 continues to operate satisfactorily. Flow tests to determine the effect of a broken spline lodged in a K-Reactor outlet nozzle were performed. the perforated plate t o cover the collection channel leading to the nozzle elbow would decrease the pressure drop across the nozzle if a broken spline becomes lodged over the outlet channel. Tests indicated that the u s e of

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Empirical correlations which describe much of the boiling burnout experiments on the N-Reactor fuel element obtained in the Heat Transfer Laboratory have been formulated. The accuracy of these correlations is generally & 10% over the range of conditions considered. Improvement of the MAC computer code for solving shielding attenuation problems has begun. and ultimately simplify input-output so that the code may be m o r e widely This will correct minor discrepancies Design and fabrication of the PRTR second generation mechanical shim rod continued. A new first-generation PRTR shim rod has been operated in the environmental test facility for 11 days. single control element has been observed during testing. Coasting of the Corrosion coupons placed in the secondary side of the main steam. generator (HX-1)of PRTR in June 1961 have been periodically examined since their installation. There are no indications of cracks or local corro - sion on AIS1 304 SS specimens. Carbon s t e e l coupons corrode initially in a relatively rapid manner, but subsequent corrosion is minor. 8 m i l s deep has been observed on carbon steel. Pitting up to The first U02 -PuOz fuel cluster containing recycled PRTR plutonium w a s completed and another is in process. Conditions were determined f o r producing uniform phys ic a1 mixtures of U 0 2 - P u 0 2 in fuel rods by bottle loading. Good linear correlation w a s found between a gamma s c a n of a U 0 2 - P u 0 2 fuel rod and known plutonium concentrations in the rod. Specifications were written for off-site fabrication of the top and bottom end brackets of the PRTR 19-rod clusters. Destructive examination of a coqcentric tubular PRTR fuel element after 1360 Mwd/tonU indicated that internal gas pressure should cause no problem on exposure to design exposure level.

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Destructive examination of two vibrationally compacted UO, - P u 0 2 , PRTR Mark I elements that failed during operation revealed brittle Zircaloy-2 fracture and loss of a small plug of Zircaloy-2 in each case. The failures seem associated with cladding faults rather than with fuel cladding -interactions, plutonium segregation, o r local overheating of the cladding. Routine examination of PRTR test element 1082 (hot swaged U 0 2 ) in the FEEF disclosed a severe longitudinal scratch on an outer fuel rod. The element will be refurbished by removing the outer wire wrap, replacing the scratched rod, and installing circumferential w i r e bands. Inspection and ring gaging of fuel elements i n the PRTR basin indicates 10 t o 20% of the elements need some type of maintenance. Bent pins and white corrosion spots are the most common difficulties. Work started on installation of additional vibrational compaction facilities for fabrication of EBWR fuel. Performance of the double can container assembly being developed for impaction of U 0 2 - 2 . 5 wt% P u 0 2 for EBWR was improved by inserting a mild steel oxygen getter between the inner and outer cans. U 0 2 - P u 0 2 w e r e prepared for irradiation at power generations prototypic of EBWR fuel rods. impacted U 0 2 - P u 0 2 w e r e successfully irradiated in the MTR to cy 1000 Test capsules containing high-energy-rate impacted T w o capsules containing the first high-energy-rate Continued out-of - reactor tests (47 days at PRTR coolant conditions) of a ring-contact end fixture on a PRTR Mark I fuel element caused no wear Essentially no grain growth was observed in magnetic force resistance welds on beryllium supplied by France and Oak Ridge. pressed pellets under 1atm of argon.

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*- High-energy-rate impaction techniques were used t o impact spec - imens of PUN, PuN-W cer met, and Pu02. This is the first time it has been - Approximately 60% loss of material occurred during heating of PuN- * A1N pellets lost about 2570 of their m a s s after heating in helium for 8 hours Selective loss of plutonium from molten UC-2 wt% PuC w a s observed. The electrical resistivity of 37.2 at. 70 C, arc - cast PuC abruptly decreases at temperatures of 500 to 550 C. Four test capsules containing various geometrical combinations of 1/2-inch diameter single and polycrystalline U 0 2 pellets w e r e discharged after irradiation t o - 1000 Mwd/tonfuel at surface heat fluxes of -800, 000 Single crystal U02 was found t o be significantly more transparent than polycrystalline U02 in the infrared at wavelengths greater than 1.0 p. U02 single crystal spheres and 30 kg of high-energy-rate impacted U02were prepared and characterized for shipment t o Euratom. I Uranium metal shot and phosphorous were reacted to produce uranium phosphide. Preliminary data indicate a melting point of 1900 i 50 C. Dense aluminum nitride, AlN, was formed by high-energy-rate impaction. Four -inch diameter, solid cer met discs approximately 1 inch thick were'fabricated for u s e as mechanical property samples. A U 0 2 - tungsten cer met with only a very small amount of the metal phase was formed by impaction of tungsten coated U 0 2 particles. A technique for the encapsulation of cer mets simultaneously with the impaction of the cer met was developed. Six c e r m e t fuel specimens, two Mo-U02 and two Mo-UN cer mets clad in Mo, and one W-UOz and one W-Un cer met clad in W, were fabricated f o r irradiation at NRTS.

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Work continued on four different plutonium fuels f o r special test pur ppe s. Over 100 fuel rods were made involving extrusion, coextrusion, and pelletizing fabrication processes. An 800 psi tensile loaded graphite creep sample was installed i n Oxidation studies of Hastelloy-C i n 1000 C air and 25 mm O2 indicate the rate of weight gain is initially pressure dependent but becomes independent The low exposure irradiation testing at 320 and 540 C of graphite containing boron carbide showed damage greatly exceeding that expected for unboronated graphite, a n effect undoubtedly arising from irradiation effects of the n, alpha reaction on boron. The irradiation induced decomposition of carbon monoxide was studied at 550 C and under pressures ranging from 1to 18 atm. The overall stoichiometry was 5 C O + 2 C 0 2 + solids which are believed t o be carbon Electron microscopic examination of graphite, made f r o m natural flake filler and coal tar pitch binder, before and after irradiation disclosed a structure not observed in coke graphite. F o r the first time i n Hanford tests, aluminum coupons exposed at high temperature conditions (H-1 Loop at 290 C, pH 7.9) f o r 45 days showed about 5070 lower in-reactor corrosion rates than the out-of-reactor control pieces, probably due t o the deposited crud film. The corrosion rates of the out -of-reactor samples appeared t o be velocity-dependent and were high (1-3 mils/month) as would be expected in water at this pH. An increase in the ATof deionized water, from 8.to 61 C, as it circulates through a n out-of-reactor nonisothermal loop has increased aluminum corrosion up t o a factor of 2. The maximum effect was observed with upstream samples; downstream samples in the specimen holder w e r e 'little affected by the temperature drop across the system.

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- - Pipe sections in the as -received, strained, and ruptured conditions were heated t o 2100 F with an alternating current through the pipe length to determine the effects of strained conditions on temperature uniformity. Initial test results indicate that temperatures were uniform in all three conditions. Pipe resistance t o the flow of electrical current may prove to 4 be a quick, simple, and relatively inexpensive way t o heat pipe for high temperature burst test work. Fracture studies of notched Zircaloy-2 specimens containing 2370 cold work and tested in slow bending showed that a transition from ductile to brittle fracture occurs at -80 C. Electron microscopy.of fracture surfaces has revealed a distinctive …