68 Design of Indian Pressurized Heavy Water Reactor Components

At present, the Nuclear Power Program in India (Chapter 68, by H.S. Kushwaha, K.K. Vaze, and K.B. Dixit) is based mainly on a series of Pressurized Heavy Water Reactors (PHWRs). This chapter first provides a general overview of the Indian PHWR design and its evolution. The design approach, material selection, and fabrication practices are described for major components such as calandria, headers, steam generators, and piping.

In Indian PHWRs, the design, fabrication, testing, and inspection of all mechanical components basically follow the requirements of appropriate sections of the ASME Boiler & Pressure Vessel Code (ASME B&PV Code). In a few cases, where it was not possible to meet the code criteria, it is the intent of the code that is met.

Other international codes used are (1) Canadian Code CAN/CSA N285.4-05 and IAEA Safety Guide 50-SG-02 for ISI and (2) French Code RCC-G for containment design.

Details are provided of the development and the use of leak-before-break (LBB) criterion to eliminate the need for installation ASME_FM_Vol_I_pi-lxxxii.qxd 5/19/09 3:11 PM Page lxxii of pipe whip restraints. Results of experiments conducted to determine load-carrying capacity of cracked pipes and the results of fatigue crack growth rate tests in support of LBB criteria are discussed. As a further example of the research and development work conducted in India related to nuclear power plant applications, the development of a modified B2 stress index (used in NB-3600-type stress analyses) for pipe elbows and curved pipes and quantification of additional safety factors to account cyclic tearing in LBB assessment are discussed.