Hanford Laboratories Operation Monthly Activities Report September 1956 (Deleted)

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Work performed under Contract No. U-3L-109-ENG-52 between the Atomic Energy Commission and General Electrio Company HAS EEE N SCANNED HE OPTiCAL DISK DRIVE

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Force Report and Persumel Status Changes Research and Development Activities Metallurgy and Reactor Fuels Physics and Instrumentation Chemistry and Separations Processes Operations Research and Syatbsis Visits and Visitors Invent ion Re port

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Manager, Hanford Laborsto................................................ M. Parker Manager, Chemical Research and ~~elo~ent.........................V. R. Cooper Manager, Operations Reseesch..... ..................................C. A . Bennett Manager, Physics and Instrument Research and development...........^. F. Gast Manager, Reactor and Fuels Research and Developent. ...............F. W. Albaugh

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Uranium Me tallurffg sleeve and canned ina IkK-filled Zircaloy-2 capsule, was examined after irradiation t o determine the woperties of material exposed a t beta or gamma phase temperatures. T h specimen was exposed s a days i n the MFR for an estimated exposure of 0.02 percent burnup (185 MWD/T) a t aa estimated temperature of 800C. The sample appearance a f t e r exposure was excellent and there was no evidence of surface irregularity or bumping. The 0.250-inch diameter of the t e s t specimen had increased approximately following table with data for the same material exposed t o 0.018 percent burnug a t 150C and an unirradiated control sample. A natural uranium tensile sample enclosed with a uranium split The tensile data obtained from a room temperature test are give= i n the Tensile Properties of Irradiated Uranium Percent burnu2 Temperature 1000 psi 1000 psi $ in 1 in. Fracture of the sample contained bright arear3 typical of large grained cleavage. The lar ultimate strength is probably due t o some extent t o the large grain size produced by heating i n t o the gamma phase but the data indicate that high temperature irradiation does not reduce the embrittlement suffered by uranium during irradiation. beta phase on the activation energy required f o r alpha phase recrystallization of urani’m has been completed. specimen cooling r a t e s of 4.4, 9.4, 250 and ’jOOC/second were establishe for temperatures of 580, 596, 620 and 640~. Utilizing the Arrhenius r a t e equation.(Ae-QbT) the activat i o n energies ( Q ) and A values were calculated and are shown in the followiLg table. A study initiated to determine the effect of cooling rate from the Time-temperature relationships for recrystallization at Jody Cooling Rate Act iv at ion Energy The activation energy decreased with increased cooling rate. the accepted theory that the activation energy should remain relatively constant as a function of strain. This is a departure from

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Investigated to provide a basis for interpretation of the results from irradiated couples. Final results from seven couples are shm in the following table. Maximum Diffusion of Uranium into AlSi and AlSi into Uranium From the Original U/ALSi Interface Each couple was cross sectioned six times and the maximum diffusion into the planes thus produced was recordsd. the maximum penetration that could be expected in each couple. were annealed 300 days at 5OOC under a pFessure of 10,000 psi were cross secticned three times in planes perpendicular to the diffusion interface. examination failed to reveal any points at which diffusion occurred. These couples will be cross sectioned another seven times. The data were analyzed statisticallg in order to determine Two V/Zr couples that fuels has been developed. radial dependent coefficient of viscosity. A method of evaluating the thermal stresses In cylindrical reactor This method utilizes a viscoelastic model with a time and had been exposed to approximstely 150, 300, and 600 W/T at HAPO. determinations were 0.18 percent burnup of the total atoms (1% MWD/T>, 0.031 percent Samples for burnup determination were cut from uranium tensile specimens which The results of these The responsibility for Project Whitney was transferred from the Chemical Processing Department to Word Laboratories Operation. ' Data for the diffusion of Xenon io. silver have been completed and a r e shown in the

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obey the following formula for difflxsivity versus temperature: graphite irradiated in the MIB showed a physical contraction in the sample 0.04 percent. It is estimated that the samples received a Hanford equivalent exposure of 0000 MD/AT. Measurenents completed on the weight losses experienced by graphite samples exyosed to high temperature water and steam shoved significant attack on the samples at steam Samples of plastic materials irradiated in the gamma facility at the MFR exhibited an absence of a rate effect. can be performed at high fluxes and the results applied at the lower fluxes encountered If this phenomena is correct, rapid material evaluation tests Examination of corrosion data for slumlnum in low flow rate deionized water at 350C has shownthat corrosion is dependent on the square r o o t of time for a longer period than literature sources indicate. A transition to a linear dependence on time appears to occur 2 to 5 months depending on the alloy. Corrosion products formed on alumiaum in water at high temperatures consists mostly Of boehmite which progresses from an amorphous structure on the "metalside" to a randomly oriented crystalline state and finally to a h i g h l y oriented crystalline state on the "water-side" of the film. A study is being made to determine which component of the reactor coolant affects the corrosion of aluminum. The strongest effect WELB that of pH. at a pH of 5 to 7 and five times this a m o m t at pH 4 and 8.5. ing effect in the region pH 4.3 to 6.6. creasing corrosion about ten-fold. Tests have been run in several of the EIMO loops to determine the effects of phosphate ion, aluminum ion and low pH 011 aludnum corrosion. A corrosion rate of 0.02 mils/ month has been observed after an exposure of nine weeks to water of pH 4.5 at 250C. T h i s rate is lower by a factor of 50 than rates observed in dynamic systems using neutral pure water. No inhibiting effect of the aluminum ion has been noted at a concenCorrosion was at a minkux Phosphate had an inhibitCitrate had a definite corrosive effect, inEmi nation of several aluminum alloys discharged from the out-of-pile H Loop indicates a general decrease in corrosion rates after 70 days exposure. The equilibrium corr os ~ o l

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rates of the coupons were 0.20 t o 0.38 mils/month o r approximately one-half t o onethird of the initial rates. Additional tests w i l l be run to determine whether inreactor rates will decrease after 60 t o 70 days. Samples of CommerciaUy pure titanium (A-55) were enclosed i n capsules containing 40 and 70 weight percent nitric acid and held at 145C fora period of two months. samples, following the exposure, showed no ignition sensitivity when checked by abrasion and impact. Corrosion rates observed were 0.0003 inches penetration per lconth in 70 weight percent HNO3 and l e s s than O.OOOO5 in 40 weight percent acid. Commercially pure titanium was exposed t o a concentrated solution of boiling UNH. significant attack occurred nor bas pyrophoricity been noted. less than 0.0001 inches penetration per month. The corrosion rate was Tensile Properties of Zircalog-2 Post irradiation emt ion was started on Zircaloy-2 tensile specimens exposed in the T h e estimated neutron slposure of the material was 1.4 x 1@* nvt (>I Mev) and The data obtained a r e shuwn in the following table along with Tensile Properties of Irradiated Zircalog-2 The changes observed compare favorably with the maximum c Zirconium irradiated at Word to ap exposure of 2.4 x 13'n* (thermal). e determined for annealed Standard Uranium Fuel Elements between reactor failure rate and inspection.reject r a t e was issued. sents strong evidence that bond and braze porosity due princip ll y to gas evolution from the uranium core i s a hsavily contributing cauae of in-reactor failure. Orre of the standard fuel elements being corrosion t e s t e d in a 1706 XE in-pile tube had attained en exposure of 479 MWDh. Xx ami nation of two split type failures from XU Reactor were completed. The failure OCCming i n tube 0986-KU had attained an exposure of 668 WD/T and that i n 0990-rn A report (RU-45211) demonstrating and discussing the excellent correlation It was a core failure of the longitudinal s p l l t variety. This fuel element There was no indication of excessive attack. Both failures resulted f r o m an unusually large number of 2 t o 8-mil

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inclusions i n the normal reactor grade uranium. removing possible copper contamination from canning sleeves prior to use. attractive reagent was a mirture of chromic acid and sulfuric acid. Laboratory tests were made t o evaluate various etchants for the purpose of ina strong magnetic field, has demonstrated that it should be suitable t o automatic welding applications because the mrt t o be welded i s heated uniformly and no crater weld is formed. The Roto-Arc welding process, consisting of a cylindrical electrode located I and E Fuel Elements Forty vacuum canned I & E slugs were evaluated t o establish a standkrd cansing procedure T k s e slugs showed good quality with respect t o (1)residual can wall (2)'camplete wet$- ing and (3) absence of voids or inclusions. Several assemblies exhibited excessive san wall penetration and por bonds. One hundred slugs will be examined i n the near futuro The occurrence of small black surface inclusions i n the weld beads of I & E slugs which were canned i n a, manner similar t o the "F" process were partly responsible for ar, abnormally high weld rejection rate f o r these slugs. Experimental tests t o establish instability characteristics of an I & Efuel columc were initiated on the full scale, low pressure electrical prototype mockup. simuleted active zone consisted of a 1.43" O.D. cupro-nickel rod with a 3/8" hole drilled along the axis. The heater rod was located inside an electrically insulated B-D-F type process tube. and a hole flow of 18 gpn.. A t 750 kw tube power, 753' i n l e t water temperature sod 50 gpm total flow, 695 of the total heat generated was found t o be d is sim ted to the t o the tube was reduced stepwise toa flow of 3lto 32 gpm. remained at initial values of 1.83 and 2.2. ratios to decrease. These results show that for 730 kw total heat generation and vith 400 psig pump discharge pressure the system is stable t o a t l e a s t 18.9 gpn in spi5e cf the fact that boiling was occurring i n the annulus. Flows were adjusted t o give an initial annulus flow of 3 gpm To determine pressure-flow characteristics for the entire tube, the total flok Flm and heat removal ratic Further reduction in flow caused both Using information gained from the in-pile irradiation of I & E f u e l elements, a study was made of the relationship between the high temperature a t the top of the amulus flow channel and the bulk temperature a t the outlet of the process tube. that the ratio of the two temperatures is as high as 1.6 i n some cases asld that the ratio increases slightly with time. eccentric position of the f u e l element in the process tube. The high temperatures are probably due t o the Enriched Fuel Elements Ten "E" slugs were irradiated i n the in-pile locp a t H Reactor f o r 47 days, using water with pH 5.4 a t an average outlet temperature of 195C. Corrosior: data obtained oa these slugs a t effective surface temperatures of 225C indicated aluminum corrosio?l rates of

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Preliminary evaluation of a l o t of vacuum-hot-pressed solid fuel elements from Sylvania has been completed. than the HAP0 hot-pressing process. However, the slugs from Splvania are inferior ia (1) closure quality (2) surface quality and (3) conformance Yith dimensiw specificaThe SEP hot-pressing process appears t o yield a better prodcct Insulated Fuel Elements Examinatioa -das continued on an insulated cored urallium themocmFle slug which was irradiated in the KTR. The material exhibited some inclusioas and b l o v - ? d e s . %is indicates that recrystallization had occurred after the gases causiag the blox - bles had been driven t o the grain boundaries. Projection Fuel Elements A test to aid i n the corrosion evaluation of projectior: fuel elernentc hes c:xdays. No l x aliz ad attack was spparent. Film patterns downstream of the projections indicate the pclssibilitg of s p t s oa the surfaces of the heat generating f u e l elements. Such hot spots can be n source of excessive corrosion leading t o jacket failures. Tests hare also been made'+,o determine the rupture characteristics of projection fuel elements. Two faili;ires hue been induced i n the Flow Iaboratory using defective fuel elements i n 145C process water. Both failures resulted i n corresponding failure of the process tube. ing of the process tube occurred adjacent t o the f u e l element projection,s. The fuel elements were exposed in process water a t 125C f o r peric;da up t o 94 A corrosion rate of 0.9 mils/month was observed. The greatest swell Tests made an AlSi canned fuel elements with ribbed supports attached by resistance spot welding showed that this method of attacbment offers several sdvazltages over other Three powder metallurgical U-Mg samples from K4PL have been rettlved from the MTR. Two samples were irradiated t o 1000 MWDD and the third to about ?KO p/IwD/b. porosity appears t c have developed i n the U-Mg material, pEsi ' c 1 ; ~ 6 sirteriry ef fast. The U-Mg material was stuck t o the aluminum capsule. This could te dtie to iz ter - diffusion of A1-U or A1-Mg t o form a bond. Experiments are being prformed t o determine whethe? U-big matrix f selma t < erFa lCZC he

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Lattice Testing Reactor Operation of the Lattice Testing Reactor continued routinely during the month. were three unscheduled shutdowns and several improvement items were completed. Experiments were performed t o t e s t the validity of methods used for determining k, i n the Lattice Testing Reactor when large lattices are used. 11-1/4 inches, 10-1/2 inches and 9-1/2 inches were employed. further analysis of these experiments arein progress. Lattice sp xing s of Thermal Test Reactor Preliminary calibrations of the control rods i n the T h e m 1 Test Reactor showed a value of 60 cents each. These calibrations show maximum rates of addition of reactivity of 3.4 x 10-5Akk/sec for the individual control rods and 2.3 x 10'4Ak/sec for the final safety sheet. Safety rods are valued a t 53.5 cents each. resulting from the f u e l pieces and water reaction was determined t o be …