Material development for India’s nuclear power programme

Abstract

The area of materials research has registered a phenomenal growth in the recent years, assiduously accepting and assimilating ideas, concepts and analytical as well as experimental methodologies and techniques form almost all scientific disciplines, thereby demonstrating its remarkably multidisciplinary and interdisciplinary character. The focus of the materials programme of this centre is to provide materials, processes and processing solutions to the emerging needs of evolving indigenous nuclear energy systems by proactive research and development on a continuing basis. The initial stage of our activities was formulated around three stage Indian nuclear power programme. In stage I, material issues related to in-core materials with emphasis on development of fabrication routes of zirconium alloys for structural application were addressed. Subsequently the thrust areas were development and characterization of mixed oxide fuel, advanced zirconium alloys, structural steels, superalloys, neutron absorber materials based on boron carbides and borides, and shape memory alloys. The research was useful for in-service performance evaluation, safety assessment, residual life estimation and life extension of nuclear reactors built during stage I i.e., PHWRs and BWRs. It also included developments which would permit rapid expansion of nuclear power initially through fast breeder reactor based on mixed oxide fuel and later based on metallic fuels. For the 3rd stage, multilayer coatings, graphite coolant tube, BeO, refractory metals and alloys, heat-treated zirconium alloys are being developed for CHTR, ADSS and AHWR. The materials being developed for fusion programme are low Z and high Z material for plasma facing application, Cu-alloys for heat sink, austenitic steels, RAFMS and ODS for structurals and NbTi, Nb 3 Sn and Nb 3 Al superconductors, lithium titanate, lithium silicate breeders, and Pb-Bi coolant. A brief overview of the materials research activities currently being pursued at Bhabha Atomic Research Centre is presented in this article.

VVER-1200 Reaktörünün Yapısı ve İşleyişi Üzerine Çalışma

“…Nükleer reaktörleri farklı teknik özellikleri bakımından Basınçlı Su Reaktörü (PWR: Pressurised Water Reactor), Kaynar Su Reaktörü (BWR: Boiling Water Reactor), Basınçlı Ağır Su Reaktörü (PHWR: Pressurised Heavy Water Reactor), Gaz Soğutmalı Reaktör (GCR: Gas Cooled Reactor), Hafif Su Soğutmalı Grafit Yavaşlatıcılı Reaktör (LWGR: Light Water Cooled Graphite Moderated Reactor) ve Hızlı Üretken Reaktör (FBR: Fast Breeder Reactor) olarak sınıflandırabiliriz. Bu reaktörler türlerinden PWR, BWR ve PHWR yaygın olarak kullanılmaktadır [16][17][18][19][20]. Dünya geneline bakıldığında kullanılan reaktörlerin % 80'ninden fazlası PWR ve BWR türleridir [21].…”

Delayed hydride crack growth study on irradiated Zr–2.5Nb pressure tube

“…Pressure tubes are primary pressure boundary components in Pressurised Heavy Water Reactors (PHWRs) and hence their structural integrity has to be assured by satisfying leak before break (LBB) criterion [1]. Operating conditions in the reactor such as high temperature, stress, corrosive media and neutron irradiation enhance susceptibility to delayed hydride cracking (DHC) and increases DHC velocity (DHCV) [1,2].…”

“…Operating conditions in the reactor such as high temperature, stress, corrosive media and neutron irradiation enhance susceptibility to delayed hydride cracking (DHC) and increases DHC velocity (DHCV) [1,2]. DHCV is required for estimating the response time before any detected leaking crack becomes unstable [3].…”