Skip to main content
U.S. Department of Energy
Office of Scientific and Technical Information

CONTROL ROD WITHDRAWAL TRANSIENTS. Test Evaluation, DLCS 2760101 DLCS 2760201

Technical Report ·
OSTI ID:4804846
Criticality and the average reactor coolant temperature of the Shippingport Atomic Power Station's nuclear reactor is controlled by hafnium control rods. Continuous withdrawal of the control rods, due to equipment malfunction or operator error. can be a potential danger to the reactor core. The reactor protection system provides automatic shutdown of the reactor core and thus limits the extent of accidental control rod withdrawal transients to protect the core. Analog simulation was employed to calculate the reactor plant response to control rod withdrawal accidents and thus determine the reactor shutdown setpoints of the protection system. The adequacy and/or degree of conservatism of the analog simulation in predicting plant response is of considerable importance. Testing was performed at the Shippingport Atomic Power Station to determine the actual plant response to mild control rod withdrawals under controlled conditions. The results of these tests were compared with calculations performed by analog simulation. The comparison of test data with calculations illustrated fairly good agreement or conservative calculations, i.e., calculations indicated greater transient excursions than did plant test data. (auth)
Research Organization:
Westinghouse Electric Corp. Bettis Atomic Power Lab., Pittsburgh
DOE Contract Number:
AT(11-1)-GEN-14
NSA Number:
NSA-16-025000
OSTI ID:
4804846
Report Number(s):
WAPD-PWR-TE-99
Country of Publication:
United States
Language:
English