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
Files
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--.