PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT, APRIL 24, 1961 TO JUNE 23, 1961

Technical Report · Sun Jan 01 00:00:00 EST 1961

OSTI ID:4013980

9 8 7 4 7 5 < : < ; 7 8 < was completed. A measurement was made of the U/sup 235/ activation distribution in the PWR-2 mockup. Core power capability as a function of lifetime was re-evaluated using the latest physics and fuel zoning information. Tests were initiated to determine the effect of blanket fuel plate ribs on DNB and the effect of design changes in the flow baffle on core flow distribution. Calculations were initiated to establish core safety after loss of flow after the time when pump- forced circulation ceases. The effect of a revision in engineering hot channel factors applicable to compartmented plates was evaluated with results favorable to core power. The design of a secondary blanket variable orifice lcck was developed and tested satisfactorily. A series of simulated fuel growth tests on seed fuel plate specimens was completed. A program to revise core manufacturing handling techniques and equipment is in progress. The isotopic content variation in the powders problem was resolved by slurry blending with Oxylene. An interim salt bath facility was utilized to manufacture V.Q. and SOAP fuel plates. SOAP wafers were successfully graphite sprayed, baked, buffed, and delivered for assembly. Long-range programs were initiated to study hafnium properties. The equipment and procedures necessary to assure proper loading of seed zones were developed. Machining of the 1D loop main coolant pump volute for Core 2 operation was completed. In-pile failures of two U0/sub 2/ fuel plates, one containing 0.160-in. thick fuel and the other containing 0.100-in. thick fuel, operating at heat fluxes of 700,000 Btu/hr-ft/sup 2/, showed a relationship between fuel temperature and swelling burnup. Conversely, these results showed that fission rate is not a parameter influencing sample failure. A similar relationship was established for the ZrO/sub 2/ +34 and 46 wt% U0/sub 2/ fuel materials. Samples containing bulk U0/sub 2/ with about 15% internal void volume continued to show better in-pile behavior than similar samples containing fuel with essentially no void volume. Similar results were obtained on ZrO/sub 2/ + U0/sub 2/ fuel materials with about 6% internal void volume. In-pile thermal conductivity measurements on U0/sub 2/, U0/sub 2/ + 46 wt% U0/sub 2/, and U0/sub 2/ + 34 wt% U0/sub 2/ were obtained after exposures of 8 to 10 x 10/sup 20/ fiss/ cc in the temperature ranges 400 tc 550 deg C. The U0/sub 2/ fuel decreased in thermal conductivity about 24%, the ZrO/sub 2/ + 46 wt% U0/sub 2/ decreased about 10%, and the ZrO/sub 2/ + 34 wt% U0/sub 2/ showed no change from its unirradiated value. The quality of Zircaloy-to- Zircaloy bonds in the fuel-bearing areas of pack-bonded seed oxide plates after beta-quenching was excellent. Low density ( approximately 90%) oxide fuel did not exhibit any greater tendency to waterlog than does high density fuel. The reactivity at 470 deg F of a core which is essentially a half section of a PWR-2 mockup was measured. The difference between the reactivity of this core and a 5 x 4 slab of the same materials was found to be the same at 470 deg F as it is at 77 deg F. This indicated the validity of extrapolating available 535'F reactivity data for 5 x 4 slabs to full core conditions via differences measured at 77 deg F. Isotcpic analyses of ten U0/ sub 2/ blanket fuel rods removed after Seed 1 operation were completed and evaluated. A program for examination of core components removed during Seed 2- Seed 3 refueling was formulated. Measurements made at the Expended Core Facility on Seed 1 clusters indicated growths in the length direction of up to 1/10 in. Tests were initiated to determine the residual stresses in the cladding of irradiated fuel plates and to evaluate the effect of clad yielding on effective plate stiffness. Input information was prepared for a new study of the rod withdrawal accident which uses an axially sectionalized core model and accounts for the effects of temperature variation on all parameters. It was determined that postulated buckling of fuel plates in the depleted Seed 2 assemblies

• Research Organization: Westinghouse Electric Corp. Bettis Atomic Power Lab., Pittsburgh
• DOE Contract Number: AT-11-1-GEN-14
• NSA Number: NSA-15-025653
• OSTI ID: 4013980
• Report Number(s): WAPD-MRP-92
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
BLANKETS
BONDING
BURNUP
CANNING
CONFIGURATION
DEFORMATION
DELAYED NEUTRONS
DENSITY
DISTRIBUTION
EQUATIONS
FABRICATION
FAILURES
FINS
FUEL ELEMENTS
HAFNIUM
HIGH TEMPERATURE
IRRADIATION
LIFETIME
LIQUID FLOW
MACHINE PARTS
MEASURED VALUES
MECHANICAL PROPERTIES
MIXING
MOCKUP
NEUTRON FLUX
PERFORMANCE
PLANNING
PLATES
POWDERS
POWER
POWER PLANTS
PRESSURE
PUMPS
PWR
QUANTITY RATIO
REACTIVITY
REACTOR CORE
REACTOR SAFETY
REACTORS
STARTUP
TEMPERATURE
TESTING
THERMAL CONDUCTIVITY
THERMAL STRESSES
URANIUM DIOXIDE
VARIATIONS
WATER COOLANT
WATER MODERATOR
ZIRCALOY
ZIRCONIUM COMPOUNDS
ZIRCONIUM OXIDES
ZONES