PRESSURIZED WATER REACTOR (PWR) PROJECT. Technical Progress Report, October 24, 1962 to December 23, 1962

Technical Report · Thu Oct 31 04:00:00 EST 1963

OSTI ID:4749285

: < < ; 8 9 C 7 6 6 6 3 9 8and design studies were completed. A study was initiated to determine the interaction of various parameters as analyzed by the XITE 1 code, in order to establish criteria as to which channels may be safely excluded from XITE 1 analysis, since it is no longer obvious from physics (input) data which channels will be thermally limiting. Also, a study was completed of the effect on power capability of variations in the thickness dimensions of blanket coolant channels, in order to provide a basis for evaluating manufacturing deviations of this dimension. Preliminary results of cold water accident studies for PWR Core 2 showed that all worst-case secondary plant incidents are tolerable. Initial results of the electron beam welding program indicate generally acceptable corrosion behavior, after shortterm testing, of annealed Zircaloy welds. The production autoradiographic technique for detecting mislocated seed fuel wafers was re-evaluated and found to be capable of clearly identifying all wafers of the various zones. The equipment specification for the design and fabrication of the PWR-2 axial flux measurement system was prepared. The design of the poison shipping rods for the Seed 1 fuel assemblies was established. Several modifications to the core instrumentation lock weld designs were considered. The test specification for the Seed 1 reactivity control system alignment test was prepared. Process development work has started on fabrication of Core 2 Seed 2 fuel. Reactor protection setpoints was established based on Core 1 Seed 4 thermal analysis and analog computer studies. The method for modifying the pressurizer surge nozzle is being re- evaluated. The modification was deferred until the plant modification period. Further investigation was carried out on the BEWI piping, removed from PWR during Seed 2-3 refueling, which was not effectively decontaminated by the APAC process. A program to verify the effectiveness of APAC in decontaminating PWR is presented. The 1A reactor coolant pump volute was successfully decontaminated in place by use of an electrolytic descaling technique. The design of the PWR Decontamination System was started. System demineralizers were placed on order; the remaining long-lead-time component specifications are being processed for order placement. A proposed control rod programming scheme was developed for use during Seed 4 lifetime if axial xenon oscillations develop. Core 2 studies were initiated in the development of a nuclear design model suitable for three- dimensional depletion studies for Core 2. Initial experiments with the pre-mockup of PWR-2 Seed 2 are in progress. A loss in control swing between one-rod- withdrawn reactivity and core excess reactivity of 1.7% DELTA k/k was measured at the estimated Seed 2 loading compared to the Seed 1 loading. This estimate is preliminary and does not include the effects of fuel zoning or cold-to-hot reactivity change. The reactivity associated with irradiation-induced changes in volume of samples used in the Irradiated Fuel Reactivity Experiments program was found to be small but not negligible for all samples. Metallographic examination of a fuel plate containing ZrO/sub 2/ + 46 wt% UO/sub 2/ and ZrO/sub 2/ + 36 wt% UO/sub 2/ fuel of nominal as-fabricated densities of 70, 80, and 90% of theoretical after 2.5 x 10/sup 20/ fissions/cc indicated the initial mode of densification of ZrO/sub 2/ + UO/sub 2/ in-pile. This was attributed to the plastic behavior of the ZrO/sub 2/ + UO/sub 2/ fuels when subjected to the forces transmitted by the cladding as it is deflected by the pressurized loop water. The experimental conditions were established for obtaining the required compartment diameter and minimum ligament thickness for MCO irradiation samples. Waterlogging tests of rod-type elements (5 in. long x 0.5 in. I.D.) containing 10 and 20% void volume and small (0.003 to 0.010 in.) diameter defects were continued through 19 cycles, after which a uniform swelling of 0.009 in. was observed on one specimen (20% void volume, 0.005

• Research Organization: Westinghouse Electric Corp. Bettis Atomic Power Lab., Pittsburgh
• DOE Contract Number: AT(11-1)-GEN-14
• NSA Number: NSA-17-015473
• OSTI ID: 4749285
• Report Number(s): WAPD-MRP-101
• Country of Publication: United States
• Language: English

Related Subjects

ACCIDENTS
ANALOG SYSTEMS
BLANKETS
BURNUP
CLEANING
CONTROL ELEMENTS
CONTROL SYSTEMS
CORROSION
DECONTAMINATION
DEFORMATION
DENSITY
DEPOSITS
DETECTION
DIFFERENTIAL EQUATIONS
DISTRIBUTION
ELECTROLYSIS
GENERATORS
INSTRUMENTS
ION EXCHANGE
LIQUID FLOW
LOW TEMPERATURE
MEASURED VALUES
MECHANICAL PROPERTIES
METALLOGRAPHY
MIXING
NEUTRON FLUX
NOZZLES
OSCILLATIONS
PLANNING
POISONING
POWER PLANTS
PRESSURE
PRESSURE VESSELS
PROGRAMMING
PUMPS
PWR
RADIATION EFFECTS
RADIOGRAPHY
REACTIVITY
REACTOR CORE
REACTOR SAFETY
REACTOR TECHNOLOGY
REACTORS
STANDARDS
STEAM
THERMODYNAMICS
THERMOMETERS
THICKNESS
URANIUM DIOXIDE
VARIATIONS
VOLUME