Published April 1997 | Version v1
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A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

  • 1. Bhabha Atomic Research Centre, Bombay (India)

Description

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

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Additional details

Publishing Information

Imprint Title
Proceedings of the seminar on leak before break in reactor piping and vessels
Imprint Pagination
773 p.
Journal Page Range
p. 159-169.
Report number
NUREG/CP--0155

Conference

Title
Seminar on leak before break in reactor piping and vessels.
Dates
9-11 Oct 1995.
Place
Lyon (France).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
28076190
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S36: MATERIALS SCIENCE;
Resource subtype / Literary indicator
Conference
Descriptors DEI
CARBON STEELS; CRACK PROPAGATION; FATIGUE; HEAVY WATER COOLED REACTORS; HEAVY WATER MODERATED REACTORS; LEAKS; PIPES; SAFETY ANALYSIS
Descriptors DEC
ALLOYS; CARBON ADDITIONS; IRON ALLOYS; IRON BASE ALLOYS; MECHANICAL PROPERTIES; REACTORS; STEELS; TRANSITION ELEMENT ALLOYS

Optional Information

Secondary number(s)
CONF-9510432--.