CVTR PROJECT: CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC. MONTHLY PROGRESS REPORT, MARCH 1961

Technical Report · Mon Oct 30 23:00:00 EST 1961

OSTI ID:4059168

Thermal bowing of the Phase II pressure tube was measured. A temperature difference of 55 deg F between the inlet and outlet legs produced a movement of the bottom of the pressure tube of 0.308'', which is within the calculated value. Phase II heat leakage tests prior to the endurance run demonstrated that the ball joint seal effectively reduced coolant leakage between the baffles and the pressure tube. In addition, the evacuated shroud acted as an effective insulator. The prototype cooling fixture, and the jumper con- nector and port test fitting were connected to water supplies and operation was checked out. Tests were continued on the control rod drive traln components using a gearmotor drive. The 84 full-height pressure-tube core in the critical experiment was reduced to a 72-tube core by removing two outside rows; 2.0% fuel was substituted for 1.1% fuel which was previously located in the 2.0% enrichment region. The net effect was a very small loss of reactivity. WTR resumed operation on Cycle H12 on March 26, 1961. Capsules A-2 and A-4 were relocated to obtain more satisfactory flux conditions. The M-1 and M-3 capsules remalned under irradiation having experienced approximately lO/sup 2//sup 1/ fast nvt. The dynamic loop experiment continues in operation having experienced 73 days of irradiation. Results of the feasibility study on ultrasonic inspection of integrally finned tubing indicate that such inspection can be satisfactorily performed. A recommendation was made to discontinue work on CVTR fuel tubes with welded fins. End closure weld development tests resulted in achieving a yield of better than 90% including second pass repairs. A study of the consequences of an auxiliary power failure indicated that boiling in the moderator tank would occur, and cause boiling of heavy water 12 to 22 1/2 minutes after the power outage, depending on the mechanism of the heavy-water steam escape. The time required for boiling to lower the moderator level to the top of the fuel would be in excess of two hours. No fuel damage would occur during this period. Development of codes for the loss-of-coolant accident analysis was continued. A convergence problem was encountered during the final phase of debugging the steady state portion of the ALPS Code. This instability was sufficiently reduced to use the steady state routine in the transient portion of the ALPS Code. (auth)

• Research Organization: Westinghouse Electric Co. Atomic Power Dept., Pittsburgh
• NSA Number: NSA-15-021845
• OSTI ID: 4059168
• Report Number(s): WCAP-4053
• Country of Publication: United States
• Language: English

Related Subjects

ACCIDENTS
BOILING
CONTROL ELEMENTS
CVTR
DEFORMATION
DISTRIBUTION
ENRICHMENT
FAILURES
FINS
HEAVY WATER MODERATOR
INSPECTION
IRRADIATION
JOINTS
LEAKS
MACHINE PARTS
MEASURED VALUES
MODERATORS
MOTION
MOTORS
NEUTRON FLUX
POWER PLANTS
PRESSURE
PROGRAMMING
REACTIVITY
REACTOR CORE
REACTOR SAFETY
REACTOR TECHNOLOGY
REACTORS
STABILITY
STEAM
TESTING
TUBES
ULTRASONICS
VESSELS
WELDING