Hanford Laboratories Monthly Activities Report February 1964

Page 1

MONTHLY ACTIVITIES REPORT THIS DOCUMENT HAS BEEN SCANNED AND IS STORED ON THE OPTICAL DISK DRIVE 7'1.119 DOCUMENT IS PU ELI C LY HANFORD ATOMIC PRODUCTS OPERATION

Page 2

this mport was proparod as an account of Government sponsored work. Neither the United States, nor the Commission, IW any porson acting on behalf of the Commission: A. M o b m y warranty or reprosontation, exprossod or implied, with respect to the accuracy, compl.hnoss, or uufulnoss of the information contained in this report, or that the use of any information, apparatus mothod, or procass disclosod in this report may not infringe privately owned rights; or Assuma any liabilitir with rospoct to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report. A, u d in tho above, .parson acting on behalf of the Commission' includes any omployee or . contmctor of the Commission, or employee of such contractor, to the extent that such employee or cantractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant (0 his omployment or contract with the Commission, or his employment with

Page 3

P. R. No. Location Route Date Signature and Date This document consists MONTHLY ACTIVITIES REPORT 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(45-11-1350. Any views o r opinions expressed in the report a r e those of the author only. GROUP 1 - Excluded from automatic downgrading and declassification

Page 4

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

Page 5

TABLE OF CONTENTS Reactor and Fuels Laboratory Physics and Instruments Laboratory Applied Mathematics Operation Radiation Pr ot ection Operation Finance and Administration Operation Test Reactor and Auxiliaries Operation

Page 7

BUDGET AND COST SUMMARY .-. February operating cost totaled $ 2 , 4 8 0 , 0 0 0 , a decrease of $413,000 from the previous month; fiscal year-to-date costs are $ 2 1 , 5 0 7 , 0 0 0 o r 667'0 of the $32 , 723,000 control budget. Hanford Laboratories' research and development costs for February, compared with last month and the current control budget, a r e shown below: (Dollars in Cur rent Previous thou sands) Month Month To Date RESEARCH AND DEVELOPMENT 1. Reactor and Fuels Laboratory Swelling incurred in prototype N-Reactor fuel elements irradiated to tube average exposure of 1600 and 2000 Mwd/ton has been determined. Maximum swelling observed in the 1600 exposure test was 1 . 3 3 vol70 in an inner component and 3.44 vol7'0 in an adjacent outer component .' In support of N-Reactor hazard studies the effect of temperature transients has been investigated by annealing an N-Reactor inner fuel component previously irradiated to 2500 Mwd/ton. Portions of the element w e r e induction heated to 980 C for hold times of 0 , 5 , and 30 min. The resulting incremental volume increases were5,8, and 1370.

Page 8

Postirradiation of eight target elements irradiated 124 days in a KER loop show good irradiation performance. Additional studies of aluminum extrusion have shown good surfaces produced by the use of an aquadag lubricant on a 400 F billet combined with the use of a solid film lubricant on the die. In the study of alternate candidate target materials, lithium aluminate has been compacted by swaging to 9470 TD. irradiated to 0.25 at. % burnup at temperatures below 600 C show extensive grain boundary tearing. A thorium fueled thermocouple element has been charged in a KER loop and is operating normally. Damage to a connection between thermocouple and recorder has caused a loss of readings from the thermocouple. Two N-Reactor inner tube fuel elements of the fluted type have been irradiated to 1800 Mwd/ton and show a volume increase of 0.9%; irradiation Continued destructive examination of clad uranium rods containing intentional clad striations is forming a progressive picture of the deformation of the cladding from the first indications of instability to ultimate rupture. F o r example, on one specimen,in which 50% of the total cladding strain occurred in the striation, necking progressed from both inner and outer surfaces. The Zircaloy-2 in the region of maximum deformation showed severe grain elongation and some void formation. In the development of a second generation N-Reactor outer support, T section supports a r e being hot forged from Zircaloy wire. Metallographic examination shows the support is being formed in the beta phase, followed by an extremely rapid die quench. Corrosion studies are continuing on brazing alloys of varying beryllium content, with and without uranium contamination. Almost every

Page 9

has gone into breakaway corrosion after 3500 h r in pH 7 , 360 C same alloy series in a pH 10 environment has accumulated 1400 h r of exposure and also shows breakaway corrosion. An irradiated N-Reactor inner fuel piece w i t h an unbonded tapered end cap w a s ruptured in the KRP at 300 C. The current N-Reactor shutdown cooling program w a s followed after a 62 min incubation period. Examination of the fuel element showed that the end cap had come loose, and the rupture extended 2 in. back from the end cap over the entire An electrically heated zirconium-clad device with N-Reactor fuel supports welded t o it w a s exposed 19 days to 288 C, pH 10 (NH40H) water at a heat flux of 600,000 Btu/(hr)(sq f t ) . Virtually no corrusion w a s found in the crevice under the support. similar tests with LiOH. Severe attack had been encountered in Crud deposition was monitored in KER- 1 Loop during operation with N H 4 0 H at pH 10 following a decontamination which produced freshly exposed surfaces of carbon steel piping. good, solids concentrations in w a t e r very low , and no crud deposition has been detected with thermocouple elements. Thus far, water quality has been Analysis w a s continued of laboratory data obtained in heat transfer and fluid flow experiments with the full scale, electrically heated model of the downstream half of an N-Reactor fuel column and outlet fittings. It was found that an outlet riser pressure of 1200 psig, a flow reduction caused by plugging upstream of the fuel column, would not produce flow instability in a maximum power tube. An in-reactor test w a s started to evaluate a mixture of sodium nitrite and sodium silicate as a corrosion inhibitor to eliminate C r activity from the once-through cooled reactors. Laboratory experiments and data analysis were concluded for the tests with a full-scale, electrically heated model of a charge of kv

Page 10

self-supported fuel elements in a smooth-bore K-Reactor Zircaloy tube. It was concluded that the kv fuel elements will require modified heat transfer limits like those specified for the self-supported fuel elements in the overbore tubes at C-Reactor. very low burnout rates. The activation energy for the reaction of C 0 2 with H2 was determined a s 5 2 . 5 kcal/mole from 762 to 975 C. Specific rate constants were also determined. Dependence of the reaction kinetics on H2 concentration seemed anomalous in the low pressure region. The r a t e of oxidation of TSX by water vapor in helium can be expressed by a function containing a half-power dependence on water vapor Swaged U02 fuel elements as well as swaged and vibrationally compacted U02-0. 5 wt% P u 0 2 elements have exceeded 5000 Mwd/ton exposure in PRTR. Continuing FERT F tests of a deliberately defected swaged U02 fuel assembly show no fuel loss o r dimensional changes. Two U 0 2 PRTR fuel elements (1034 and 1099) with broken wire wraps were successfully repaired in the PRTR basin. Irradiation tests of prototypic EBWR fuel are continuing satisfactorily. Reconditioning and inspection of EBWR fuel rod cladding increased yields by approximately 25%. Four PRTR elements containing impacted mixed oxides were fabricated by vibrational compaction. Gamma scanning and powder outgassing A stainless steel clad, 80 vol70 U02-stainless steel, 4-rod cluster, complete with stainless steel end caps and spacing brackets, w a s fabricated by a "one-shot" pneumatic impaction process.

Page 11

PRTR fuel element design studies were directed toward fuel element Out-of-reactor loop t e s t s of an eddy current instru-rejuvenation concepts. mented, PRTR Mark I element demonstrated its pendulum behavior rather Irradiated Th02-8. 8 w t % P u 0 2 pellets revealed an unexpectedly low thermal conductivity, based on irradiation conditions requested from Examination of a section of PRTR pressure tube in TF-7 disclosed that 1/16 x 1 / 2 in. end bracket supports of a previously tested PRTR fuel element had caused fretting corrosion 10 mils deep. A 360° ring support and 1/4 x 1 / 2 in. supports had produced no measurable depth of penetration under similar conditions. Preliminary data from laboratory tests indicate that sulfamic acid solutions have low effectiveness in dissolving Pu02. A preliminary study of increasing the specific power in the PRTR by reducing the number of tubes a n d / o r the fuel length was completed. Specific power would be increased by a factor of up to 3 . 6 and enrichment up to about 2.1%. Although control system strength would be changed, inherent reactor safety would not be appreciably affected. Assembly of the second generation PRTR shim control rod driving head w a s completed and the unit is operating satisfactorily in the environmental test facility. Vibration measurements from the instrumented 19-rod bundle fuel element assembly exposed to prototypical PRTR coolant conditions in the EDEL-I Loop indicate that the fuel element vibrates about 1 mil at a frequency of less than 1 cps. Pressure tube vibration is recorded at 0.2 to 0.5 m i l at 20 to 25 cps frequency. tude of 8 m i l s at 25 to 30 cps on the inlet connections to Tube 1348 in the This compares to a vibration ampli-

Page 12

The first corrosion test of aluminum-clad fuel elements in C - 1 Loop, conducted at 290 C outlet temperature, resulted in a very rapid rise in cladding temperature (after a 50 hr induction period) from crud deposition. A second test is being run at 260 C. The cladding temperature has risen slowly for 10 days, but has leveled out since. solid solution are highly volatile in atorr vacuum at 1800 C. In contrast, ThN, TiN, and ZrN a r e very stable under these conditions. Three capsules containing UN-20 wt% PUN were charged into the MTR this month. Samples of P u 0 2 and PUN with burnups of approximately 20 x lo2' fissions/cm a r e being examined in the radiometallurgy facility. Significant corrosion apparently did not occur during irradiation of a deliberately defected, water cooled Tho2 fuel element. Gas analyses from nondefected, irradiated elements revealed a high concentration of hydrogen at approximately 3 atm pressure (STP). High temperature transmission microscopy of U 0 2 at 500 C in H2 shows fracture of U 0 2 single crystals resulting from the hydriding of A 10 1 zone of porosity w a s observed at the surface of a U 0 2 single crystal irradiated to an exposure of 5 x lo2' fissions/cm 3 . Work has progressed in the development of two new methods of testing-low cycle fatigue of reactor structural materials and a cleavage technique for studying the fracture behavior of materials. The construction and assembly of a dynamic loading system for fracture studies was completed. This system, which is capable of producing loading r a t e s a s high as five million pounds per second, is presently being checked out for pe.rformance. Two creep capsules containing annealed AIS1 304 SS have been shipped to the reactor for charging. To obtain control data two

Page 13

out-of-reactor creep tests have been started at 30,000 psi, one at 500 C Button-head specimens of Inconel X-750, representing two test heat treatments, a r e being irradiated in the G - 6 position of the ETR during Cycle 61, andothers representing a third test heat treatment a r e being prepared for shipment for Cycle 62. Recent tensile tests performed by extensometry at 300 C (572 F) on stainless steel specimens yielded results indicating that this high temperature method may be successfully applied to remotely operated elevated temperature tests. The determination of the functional dependence of strain rate and effective stress and the quantitative calculation of effective stress (the rate controlling s t r e s s ) is the most important single obstacle to the formulaation of a rate theory of deformation. Data from stress relaxation experiments w i l l provide information about this functional relation. of the effective stress at specific strain rates agree well with estimates made by other methods. The fast neutron spectrum distribution in N-Reactor was measured using a Li6 detector. Agreement of the measured spectrum and previously calculated spectrum was very good. Various gaseous mixtures made up from hydrogen, methane, carbon monoxide, o r nitrogen were examined After irradiation in a Hanford reactor at 550 C o r in the Co60 facility at 30 C. The long-term EGCR graphite irradiations continue successfully. A capsule was discharged during the report period. Evidence of a strain rupture limit was revealed in studies of the effects of tensile loading on the irradiation-induced contraction of The thermal conductivity of several graphites has been determined over he temperature range -195 to 525 C. A maximum is observed at -10 C.

Page 14

The irradiation of boronated graphite samples continues satisfactorily; all temperatures a r e within acceptable limits. Three prototype Zircaloy-2 clad Th-U fuel elements have been irradiated to 4600 Mwd/ton at fuel temperatures of 540 C and higher. The elements show a volume increase of only 0.570;irradiation is continuing. in 300 C water have shown a tendency to longitudinal splitting of the cladding in as-extruded rods. Specimens heat treated at 750 C for 1 h r o r 800 C for 3 h r have now been defect tested. No clad splitting occurred in the A method has been developed for evaluation of Th-U bond strengths by measuring the force required to peel the cladding in a circumferential direction. Forces required to peel the cladding were measured on a s - extruded rods. Heat treatment of rods at 800 C for 3 h r strengthened the bond to the point that cladding failure occurred before bond peeling started. An irradiation test is in progress to investigate the ability of a submicron dispersion of the uranium carbide to reduce swelling. One capsule has achieved an exposure of 0 . 2 at. 70 with discharging targeted for 0 . 3 at. %. Two additional capsules await charging. Preliminary results from the U 235:U238 burnup control cases for the Fuel Reuse study show only a 0.2 to 0.3 mill fuel cycle cost variation in the control c a s e s …