Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)
- Oak Ridge National Lab., TN (United States)
The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the primary acceptance criterion'' in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering; Oak Ridge National Lab., TN (United States)
- Sponsoring Organization:
- USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5584342
- Report Number(s):
- NUREG/CR-5799; ORNL/TM-11982; ON: TI92010943
- Country of Publication:
- United States
- Language:
- English
Similar Records
Review of recent ORNL specific-plant analyses
Review of recent ORNL specific-plant analyses
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
ROWE YANKEE REACTOR
REACTOR VESSELS
BATTELLE PACIFIC NORTHWEST LABORATORIES
EVALUATION
FAILURES
FRACTURE MECHANICS
FRACTURE PROPERTIES
FRACTURES
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
O CODES
ORNL
REACTOR SAFETY
THERMAL SHOCK
TRANSIENTS
V CODES
ACCIDENTS
COMPUTER CODES
CONTAINERS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICAL PROPERTIES
MECHANICS
NATIONAL ORGANIZATIONS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
SAFETY
THERMAL REACTORS
US AEC
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
220900 - Nuclear Reactor Technology- Reactor Safety