French verification of PWR vessel integrity

Technical Report · Thu Feb 01 00:00:00 EST 1990

OSTI ID:7257735

Tanon, A P; Grandemange, J; Houssin, B; Buchalet, C ^[1]

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1. Framatome, 92 - Paris La Defense (France)

This report presents the fracture mechanics methods and criteria developed in France for fast fracture risk analysis of PWR vessel belt lines. Regulatory requirements and design code (RCC-M) provisions are presented, as well as the new set of criteria developed for a generic reevaluation of the margins with respect to fast fracture of the 900 MWe serie vessels. Emphasis is placed on features which have direct interest for the effort undertaken in the United States on improving the criteria for low temperature operation of LWR plants and evaluation of PTS conditions. The following topics are more particularly developed: the reference defects, which are defined in relation to manufacturing aspects and non destructive examination capacity; the fracture mechanics models and plasticity corrections, with comparison to elastoplastic analysis involving the criteria of the local approach of fracture; the safety margins to be met, which depend on the combined probability of the transient and reference defect class under consideration; and the French experience on defining trend curves to predict irradiation embrittlement. The case of the CHOOZ A vessel illustrates how the new set of criteria has been combined with the surveillance program and in-service inspection results for proving the vessel reliability, and allowing plant restart. 37 refs., 63 figs., 24 tabs.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (USA); Framatome, 92 - Paris La Defense (France)
• Sponsoring Organization: EPRI
• OSTI ID: 7257735
• Report Number(s): EPRI-NP-6713; TRN: 90-007916
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
PRESSURE VESSELS
PERFORMANCE TESTING
PWR TYPE REACTORS
CRACKS
DAMAGE
DATA PROCESSING
DEFECTS
EMBRITTLEMENT
FAILURE MODE ANALYSIS
FAULT TREE ANALYSIS
FINITE ELEMENT METHOD
FRACTURE MECHANICS
FRANCE
IMPACT SHOCK
IRRADIATION
MATERIALS TESTING
MECHANICAL PROPERTIES
PROBABILISTIC ESTIMATION
PROGRESS REPORT
QUALITY ASSURANCE
RELIABILITY
RISK ASSESSMENT
STRESS ANALYSIS
STRUCTURAL MODELS
VERIFICATION
WELDED JOINTS
CONTAINERS
DOCUMENT TYPES
EUROPE
JOINTS
MECHANICS
NUMERICAL SOLUTION
PROCESSING
REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
WESTERN EUROPE
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
420500 - Engineering- Materials Testing