RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT FOR THE PERIOD JULY 1, 1959 TO SEPTEMBER 30, 1959

Technical Report · Fri Nov 20 00:00:00 EST 1959

OSTI ID:4203060

Improvements were made to the methods used to calculate the resonance escape probability in heavy water moderated, clustered rod, UO/sub 2/ fueled reactors. A preliminary analog computer investigation of the externally controlled CVTR system was completed. The sensitivity of the controllers to variations in reactivity coefficients and the size of the pressurizer were determined. Proposals for prototype seal welding and cutting equipment were received from several companies. On the basis of these proposals, the vendor has been selected. The designs of several of the pressure tube components, including the seal weld lips, the connector body, and the neutron streaming plug, were modified. A Plexiglas model of the connector body and an aluminum streaming plug for use in hydraulic tests are being fabricated. The expensive circumferential corrugations in the thermal baffles were found to be unnecessary since stresses due to differential thermal expansion are low. Dimples on the baffles will be used to provide spacing between baffles. Alternative orifice designs were studied These include single plate, multiple plate, and stacked rotating plate designs. The pressure drop tests using fullscale models of the fuel assemblies were initiated in Loop E. A study was undertaken to determine the advantages to be gained by using river water, directly, as cooling water rather than an intermediate cooling system. There are possible savings in heavy water inventory and equipment but these must be reviwed considering safety and reliability aspects. Effort was devoted toward revising the auxiliary system designs prepared for Reference Design II. Alternative designs were investigated in an effort to improve system operation or to affect cost savings. The gamma activity contributed to the primary system by a failed fuel element and the background activity in the primary system were estimated. These data will be applied to the study of the sensitivity of the failed fuel element detection system. Information was developed for the Preliminary Hazards Summary Report which was published by CVNPA and submitted to the AEC. As a result of reevaluation of the steam generator design in which a change in the average wall thickness of the Inconel tubing was made (from 18 BWG-0.049 in. to 19 BWG-0.042 in.), a reduction of 14% in the holdupvolume was achieved. An analysis of primary coolant pressure distribution was performed which established the reactor core pressure drop as 111.3 psi. Preparation of a preliminary Equipment Specification for the design, fabrication and test of a bottom-mounted, upward-scram control rod drive mechanism was initiated. Preliminary calculations for sizing the pressurizer were completed which established the total liquid-vapor volume to be 90 cubic feet (50 cubic feet-vapor; 40 cubic feet liquid). A study was initiated of an alternate loop neutron shield design of heavy water and stainless steel. The design would use an elevated moderator level in combination with parallel, separated steel plates to form the shield. It would avoid the difficulty of sealing a light water shield so as to avoid degradation of heavy water. A prototype of a Zircaloy-2 fuel tube support grid was successfully brazed using a Zr-5Be-15Cu brazing alloy, 1000 deg C temperature for one hour in a vacuum of 0.05 microns. Results of preliminary studies and consultation with vendors indicate that it will probably be necessary to machine the U-bend sections of the pressure tubes from a solid U-bend shape. (auth)

• Research Organization: Westinghouse Electric Corp. Atomic Power Dept., Pittsburgh
• DOE Contract Number: AT(30-1)-2289, SUBCONTRACT NO. 1
• NSA Number: NSA-14-004941
• OSTI ID: 4203060
• Report Number(s): CVNA-36
• Resource Relation: Other Information: For Carolinas Virginia Nuclear Power Associates, InC. Orig. Receipt Date: 31-DEC-60
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ANALOG SYSTEMS
BERYLLIUM ALLOYS
CLOSURES
CONTROL ELEMENTS
CONTROL SYSTEMS
COOLANT LOOPS
COPPER ALLOYS
DETECTION
ECONOMICS
FAILURES
FUEL ELEMENTS
HEAVY WATER
HEAVY WATER MODERATOR
HIGH TEMPERATURE
HYDRAULICS
MECHANICS
MOCKUP
NEUTRONS
PLANNING
POWER PLANTS
PRESSURE
PRESSURE VESSELS
REACTIVITY
REACTOR SAFETY
RESONANCE ESCAPE PROBABILITY
SEALS
SENSITIVITY
SHIELDING
STAINLESS STEELS
STEAM
TESTING
URANIUM OXIDES
VARIATIONS
WATER COOLANT
WELDING
ZIRCALOY
ZIRCONIUM ALLOYS