PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT FOR THE PERIOD FEBRUARY 24, 1958 TO APRIL 23, 1958

Technical Report · Fri Oct 31 04:00:00 EST 1958

OSTI ID:4344066

The power and temperature controller was successfully tested under varying load changes and xenon transient conditions. Adjustment of the intermediate range of the nuclear instrumentation by modifying the detectors was completed with ion chambers now producing approximately 110 mu a at full power. The cause of failure uted to stress corrosion of unknown origin but originating from the secondary side. The Seed 2 fuel loading, fuel plates and subassembly configuration were established and the fuel plates were ordered. A design inof adopting the present seed exit water thermocouple instrumentatlon concept for use with the Seed 2 metal and inlet water temperature instrumentation. The loading for Seed 2 was established at 90 kg of U/sup 235/. The boron loading in the outer three fuel plates on both sides of the subassembly was established at 0.039 wt.% of the fuel alloy. The test analysis for the plastic 1/4scale model of the tube design bottom support for the 20-control-rod core was completed. Thermal performance calculations were made on the 20-control-rod core throughout lifetlme of the core. An' fective'' thermal conductivity for a UO/sup 2/ rod of Core 1 cross section under irradiation was determined to be 1.0 plus or minus 0.2 Btu/ hr-ft-/sup deg / between 300 and 1000 deg F. An irradiated, Zircaloy-2 base, 10 wt.% U samples clad with Zircaloy-2 swelled 213% as a result of a thermal treatment of 30 min at 760 No swelling was observed after a treatment of 30 min at 700 Four similar samples moderately swelled as a result of a 6- to 10- sec treatment at 760 deg C. Post irradiation metallographic examination determined that two of the CR-V-j fuel elements had been defected during a pre- irradiation corrosion test. The defects are attributed to the corrosion of stringers in the receptacle plates of one fuel element and to hydride embrittlement in other regions of the second fuel element. Excessive hydrogen pickup of high-nickel Zircaloy-2 was confirmed by out-ofpile tests; the same tests show that high-copper or high-iron contents do not cause excessive hydrogen pickup. The creep properties of Ag-base control rods can be improved by three metallurgical expedients: using of oxidized power for control rod fabrications coarsening of grain sizes increasing alloy composition to obtain a two-phase microstructure. A method for synthesizing three-dimensional power distributions from results of one- and two-dimensional calculations was developed to provide data for a re-evaluatlon of Core 1 thermal performance. The quantity and distribution of U/sup 235/ and boron in a Seed 2 subassembly were examined, and the distribution of materials which minimizes the water channel peaking factors was found. A study of the relationship between blanket power fraction and mechanical design of Core 2 blanket fuel elements was completed. (For preceding period see WAPD-MRP-72.) (auth)

• Research Organization: Westinghouse Electric Corp. Bettis Plant, Pittsburgh
• DOE Contract Number: AT(11-1)-GEN-14
• NSA Number: NSA-12-010082
• OSTI ID: 4344066
• Report Number(s): WAPD-MRP-73
• Country of Publication: United States
• Language: English

Related Subjects

ABSORPTION-- ADSORPTION-- COPPER-- HYDROGEN-- IRON-- POWER PLANTS-- PRESSURE-- PWR-- QUANTITY RATIO-- REACTORS-- TESTING-- WATER COOLANT-- WATER MODERATOR
ABSORPTION-- ADSORPTION-- HYDROGEN-- NICKEL-- POWER PLANTS-- PRESSURE-- PWR-- QUANTITY RATIO-- REACTORS-- TESTING-- WATER COOLANT-- WATER MODERATOR-- ZIRCALOY
ALLOYS-- BORON-- FUEL ELEMENTS-- FUELS-- MASS-- PLATES- - POWER PLANTS-- PRESSURE-- PWR-- REACTOR FUELING-- REACTORS-- WATER COOLANT-- WATER MODERATOR-- ZONES
ANALOG SYSTEMS-- CONTROL ELEMENTS-- PLANNING-- POLYMERS-- POWER PLANTS-- PRESSURE-- PWR-- REACTORS-- RODS-- TESTING-- TUBES-- WATER COOLANT-- WATER MODERATOR
BLANKETS-- FUEL
PHYSICS