Degraded piping program -- Phase II: Summary of technical results and their significance to leak-before-break and in-service flaw acceptance criteria, March 1984--January 1989

Technical Report · Wed Mar 01 00:00:00 EST 1989

DOI:https://doi.org/10.2172/6377903· OSTI ID:6377903

Wilkowski, G. M.; Ahmad, J.; Barnes, C. R.; Brust, F.; Ghadiali, N.; Guerrieri, D.; Jones, D.; Kramer, G.; Landow, M.; Marschall, C. W.

The intent of this program is to experimentally validate and enhance available analytical methods for evaluating the mechanical behavior of nuclear power plant piping containing circumferentially oriented defects. This document describes the results of the approach taken by Battelle to provide state-of-the-art analysis methods to predict the behavior of degraded piping operating at LWR conditions. Section 1 gives a summary of the scope of the program. Section 2 of this report summarizes the technical efforts during the course of this program, discusses the material characterization efforts and describes the full-scale pipe fracture experiments and fracture mechanical analyses conducted. Section 3 discusses the significance of results to LBB and in-service flaw acceptance criteria, and examines the potential impact of characterization evaluations and unusual fracture modes observed in nuclear piping materials on evaluation procedures. Section 4 summarizes implications of the program findings and indicates areas for future study in the area of piping integrity research. 44 refs., 37 figs., 8 tabs.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Engineering; Battelle Columbus Div., OH (USA)
• OSTI ID: 6377903
• Report Number(s): NUREG/CR-4082-Vol.8; BMI-2120-Vol.8; ON: TI89009768
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
BWR TYPE REACTORS
REACTOR COOLING SYSTEMS
PIPES
IN-SERVICE INSPECTION
PWR TYPE REACTORS
CARBON STEELS
CRACKS
DATA BASE MANAGEMENT
EVALUATION
EXPERIMENTAL DATA
FERRITIC STEELS
FRACTURE MECHANICS
INTERGRANULAR CORROSION
MATERIALS TESTING
MECHANICAL PROPERTIES
PROGRESS REPORT
REACTOR SAFETY
STAINLESS STEELS
STRESS CORROSION
WELDED JOINTS
ALLOYS
CHEMICAL REACTIONS
COOLING SYSTEMS
CORROSION
DATA
DOCUMENT TYPES
ENERGY SYSTEMS
HIGH ALLOY STEELS
INFORMATION
INSPECTION
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
MANAGEMENT
MECHANICS
NUMERICAL DATA
REACTOR COMPONENTS
REACTORS
SAFETY
STEELS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
220900 - Nuclear Reactor Technology- Reactor Safety