PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT FOR THE PERIOD OCTOBER 24, 1960 TO DECEMBER 23, 1960

Technical Report · Tue Oct 31 00:00:00 EST 1961

OSTI ID:4098466

: ; : 8 ; : : = 8 7 4 = 7 Approximately 400,000 blanket fuel wafers were compacted, and 200,000 were sintered. Approximately 75,000 seed fuel wafers were compacted and Zircaloy-4 ingots for blanket and seed receptacle plates were rolled to final thickness. Analyses and tests, which confirm the adequacy of the mechanical design, were completed. FEDAL sampling rake testing was started and a mechanical alignment study of the core was completed considering worst cases of clearances, manufacturing tolerances, and deflections and their effect on the reactivity control system. Designs of the supportflange instrumentation from the support tubes in the bottom support to the lateral support were approved. Air-flow studies in a 0.4 scale model of PWR-2 were completed. No gross changes in core-flow distribution or areas of flow instability were detected. Studies are being conducted to investigate methods of increasing the allowable core power 1evel at 9,100 to lO,OOO EFPH by control-rod programming schemes. The program for irradiation of fuel and poison materials for the seed of Core 2 produced measurements of swelling after irradiations approximating the maximum expected during the lifetime of the seed. Fuel platelets containing ZrO/sub 2/ + 34 wt.% UO/sub 2/ and ZrO/sub 2/ + 46 wt.% UO/sub 2/ were irradiated to 31.2 x 10/sup 20/ and 32.3 x 10/sup 20/ fissions/cc, respectively; and they show maximum increases in thickness of l5.7 and 25.7%. One compartment of a sample containing ZrO/sub 2/ + 46 wt.% UO/sub 2/, irradiated to 24.2 x 10/sup 20/ fissions/cc failed; the unfailed five remaining compartments showed a maximum increase in thickness of 22%. One sample of 30 wt.% B/sub 4/C + SiC which contained an intentionally defected compartment showed abnormal swelling of this compartment after about 30% burnup of the B/sup 10/ atoms (9.5 mills max). An adjacent nondefected compartment showed no significant dimensional change. A test of 48 previously irradiated ceramic fuel specimens and 5 burnable poison specimens, gave an indication of high specimen temperatures after one day of operation. Examination after shutdown revealed an unidentified deposit on the samples. It was determined that B/sub 4/C--SiC mixtures free from localized large B/sub 4/C-SiC agglomerates could be prepared by properly blending B/sub 4/C powder of 2- mu (average) particle diameter with SiC powder of 5- mu (average) particle diameter. Corrosion testing of hot-pressed B/sub 4/C-SiC compacts in 680 deg F water showed that the only significant factor affecting corrosion resistance is the amount of open porosity. Reduction of the stress-relieving temperature from 1150 to 700 deg F of Zircaloy cladding components, which had been cold rolled to a 25% reduction prior to isostatic pressure bonding, effectively eliminated the grain-growth problem. Pressure fatigue results on compartments in isostatic- pressurebonded plates beta -quenched from 1850 deg F continue to indicate properties that compare favorably with those previously determined on as-bonded plates. The effectiveness of pyrolytic carbon vs. sprayed graphite coatings on fuel wafers for minimizing uranium penetration into the cladding was further confirmed. After 3500 EFPH of operation of PWR Core l Seed 2 at Shippingport, the observed critical position of the controlling group of rods during full power operation is substantially below the previously predicted value; however it is believed that the lifetime prediction of 6500 EFPH is not grossly in error. Comparisons of the measured and calculated critical conditions, excess reactivities, and onerod-withdrawn shutdown margins at various temperatures in cores containing the PWR Core 2 UO/sub 2/ mock-up fuel were completed. Measurements and calculations were made at 470 deg F of the U/sup 235/ activation distributions in the seed and blanket regions of a 5 x 4-cluster UO/sub 2/ slab assembly in PWR Core 2 geometry. Examination of an Inconel holding spring, removed from the blanket at the end of the first seed lifetime, revealed essentially no change in properties

• Research Organization: Westinghouse Electric Corp. Bettis Atomic Power Lab., Pittsburgh
• DOE Contract Number: AT-11-1-GEN-14
• NSA Number: NSA-15-008354
• OSTI ID: 4098466
• Report Number(s): WAPD-MRP-89
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
AIR
BLANKETS
BORON 10
BORON CARBIDES
BURNUP
CANNING
CARBON
CERAMICS
COATING
COMPACTS
CONTROL ELEMENTS
CONTROL SYSTEMS
CORROSION
DEPOSITS
DIFFUSION
FABRICATION
FATIGUE
FISSION PRODUCTS
FUEL CANS
FUEL ELEMENTS
FUELS
GAS FLOW
GRAPHITE
INSTRUMENTS
IRRADIATION
LIFETIME
PLATES
POROSITY
POWER
PRESSURE
PROGRAMMING
PWR
RADIATION EFFECTS
REACTIVITY
REACTOR CORE
REACTORS
ROLLING
SEEDS
SILICON CARBIDES
SINTERING
STABILITY
TEMPERATURE
THERMAL STRESSES
THICKNESS
URANIUM
URANIUM DIOXIDE
WATER
ZIRCALOY
ZIRCONIUM OXIDES