PSI - Issue 60

Rishi K Sharma et al. / Procedia Structural Integrity 60 (2024) 264–276 Rishi K. Sharma/ Structural Integrity Procedia 00 (2019) 000 – 000

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1. Introduction Pressurized Heavy Water Reactors (PHWRs) are the backbone of India’s closed fuel cycle three stage nuclear power program and serve as the first stage of nuclear power generation in the country (Bajaj & Gore, 2006) (Muktibodh, Dixit, Ingole, & Prakash, 2016). Unlike conventional reactors, PHWRs employ a multitude of pressure tubes instead of a singular pressure vessel. These pressure tubes serve the crucial function of transporting heavy water (D 2 O) as a coolant over the fuel elements within the reactor core. The design of PHWRs centers around the use of natural uranium dioxide as fuel, with heavy water serving as both a neutron moderator and a high-pressure, high temperature coolant in a separate circuit. A hallmark of PHWRs is their horizontal pressure tube configuration, wherein natural uranium dioxide fuel is utilized alongside heavy water (D 2 O) for cooling and neutron moderation. Zirconium alloys, specifically Zr-2.5%Nb, have been chosen as the preferred structural materials within the nuclear core, particularly for pressure tube construction. These alloys have been favored not only for their neutron economy but also for their remarkable corrosion resistance and excellent structural properties, both of which are paramount for their intended application. Ensuring the structural integrity of the pressure tubes is essential during reactor operation to ensure effective cooling of fuel bundles. These pressure tubes serve as the primary pressure boundary for the hot heavy water coolant within PHWRs, effectively functioning as miniature pressure vessels operating under an internal pressure of ~10 MPa and temperatures ranging from 250 to 300 °C. Furthermore, their mechanical and fracture properties change during reactor life due to a fast neutron fluence (flux × time) and also undergo hydride embrittlement. The effect of fast neutron fluence saturates at around 10 25 n/m 2 (Rodgers, et al., 2008). The saturation fluence for pressure tubes is achieved within 1-5 effective full power years of reactor operation. After attaining the saturation fluence, the properties are mainly governed by the hydrogen embrittlement. Due to its susceptibility to hydrogen embrittlement, pressure tubes are manufactured with very low initial hydrogen concentration. Over a period of time, hydrogen is released as a result of the corrosion reaction between the coolant and the hot zirconium alloy. A portion of this released hydrogen (~10%) is absorbed by the pressure tube material (Motta, et al., 2019). Initially, the hydrogen is present in the dissolved condition in α -Zr phase as interstitial solutes which then precipitates out within the α -Zr matrix mainly as δ -hydrides (Coleman & Hardie, 1966) (Singh, Mukherjee, Gupta, & Banerjee, 2005) (Sharma, Tewari, Singh, & Kashyap, 2018). The desirable texture and microstructure of pressure tubes are deliberately imparted during their fabrication to minimize the diametrical creep during reactor operation (Tewari, Srivastava, Dey, Chakravarty, & Banerjee, 2008) (Singh, Mukherjee, Kishore, & Kashyap, 2005). Owing to strong texture, either circumferential or radial hydrides are formed. Circumferential hydrides are predominantly formed under unstressed conditions, guided by the microstructure of the pressure tube material. However, the radial hydrides are formed when pressure tubes having hydrogen are cooled from the solutionizing temperature under tensile hoop stress exceeding a certain threshold value required for hydride reorientation (Singh, et al., 2006) (Singh, Kishore, Singh, Sinha, & Kashyap, 2004). The morphology of these hydrides, encompassing their size, shape, distribution, and orientation, plays a critical role in determining the material's fracture characteristics (Sharma, et al., 2018) (Sharma, et al., 2017) (Gopalan, et al., 2021) (Bind, Avinash, Sunil, Sharma, & Singh, 2023). Change in fracture properties due to varying hydride morphology have been reported in the recent literature. However, burst behavior of these tubes under varying hydride morphology is not available in the open literature. As such, understanding the influence of hydride morphology on the fracture behavior of Zr-2.5%Nb is essential for evaluating the safe operating pressure-temperature domain for the nuclear reactors. An experimental setup was meticulously designed and fabricated to form radial hydrides in Zr-2.5%Nb alloy pressure tube spools. The design of the setup revolved around achieving a hoop stress exceeding the threshold stress within the spool, required for hydride reorientation in this specific alloy. The radial hydrides can be formed in the Zr 2.5%Nb pressure tube material due to various construction and operating fallacies viz., material variability, improper rolling, blister formation (Singh R. N., Kishore, Sinha, Banerjee, & Kashyap, 2003), presence of flaw or discontinuity, induced stresses due to nodular corrosion (Murty, Singh, & Ståhle, 2019) etc. The present study brings out burst behavior of pressure tubes in the presence of varying hydride morphologies. This investigation is crucial for evaluating the service life of pressure tubes and their qualification in Leak-Before-Break (LBB) scenarios (Moan, Coleman, Price,