Published November 2002 | Version v1
Book

Fracture integrity assessment of bi-metallic pipe joints in nuclear reactors

  • 1. Reactor Safety Div., Bhabha Atomic Research Centre, Mumbai (India)

Description

Full text: The current leak-before-break (LBB) methodology is based on evaluating the integrity of cracked nuclear piping, having mismatch in material properties, by considering the strength properties of the base material and fracture properties of weld material. The flaws are generally postulated in the welds. The crack driving force e.g. J-integral and limit-load solution are evaluated using standard formulae derived from homogeneous models [3]. This method generally yields safe result but it is highly conservative. Fig. 5 clearly represents the conservatism involved in the traditional approaches. In order to carry out a precise assessment of integrity of a cracked welded structure, knowledge of measurement of the fracture toughness of the welded joint, determination of limit load and estimation of crack driving force is needed. Moreover when fracture testing is done, generally the weld is at the center of the base material and hence the results obtained would invariably include the effects of ductility of the base material. Thus the strength of this composite/ welded junction not only depends on the properties of base and weld material but also on weld geometry. At present, FEM is considered as the best method to evaluate limit load and crack driving force for these mismatched structures because it can simulate various weld geometries and mismatched variables. However it is always useful to have some simplified estimation methods for engineering calculations. With this view in mind an estimation method has been developed for evaluating crack driving force and limit-load for mismatched specimen like center crack plate (CCP) and three-point bend specimen (TPB). The approach is quite general and can be extended for the case of pipes. The crack driving force can be evaluated from the existing estimation techniques like R-6 and A-16, by using the concept of equivalent stress plastic-strain relations, determined from limit-load solution

Part of:
First national conference on nuclear reactor technology

Additional details

Publishing Information

Publisher
Bhabha Atomic Research Centre
Imprint Place
Mumbai (India)
Imprint Title
First national conference on nuclear reactor technology
Imprint Pagination
421 p.
Journal Page Range
p. 354-355

Conference

Title
1. nuclear reactor safety
Acronym
NRT-1
Dates
25-27 Nov 2002
Place
Mumbai (India)

INIS

Country of Publication
India
Country of Input or Organization
India
INIS RN
34011439
Subject category
S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
CRACKS; FAILURES; FRACTURE MECHANICS; LEAKS; NUCLEAR POWER PLANTS; STRAINS
Descriptors DEC
MECHANICS; NUCLEAR FACILITIES; POWER PLANTS; THERMAL POWER PLANTS

Optional Information