PSI - Issue 71

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ScienceDirect

Procedia Structural Integrity 71 (2025) 271–278

5 th International Structural Integrity Conference & Exhibition (SICE 2024) Study of Severely Hydrided Zr-4 PHWR Fuel Clad Using Ring Tensile Test Priti Kotak Shah* a , Prabhjot Kaur Bhatia b , A. Sushibine a J. S. Dubey a , P. P. Nanekar a a Post Irradiation Examination Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400085, India b Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Trombay, Mumbai-400085, India

Abstract Zircaloy (earlier Zr-2 and now Zr-4) is widely used as cladding alloy for nuclear fuel elements in pressurised heavy water reactor. High thermal stresses, neutron irradiation and corrosion in the reactor change the material properties of the clad. One such important factor for changing the mechanical properties of clad is hydrogen pickup by clad tubes. The stress imposed on the cladding is mostly hoop stress on the circumference of the tube and therefore the data on the circumferential strength and ductility of the nuclear fuel cladding are required. Hydrogen content has an effect on mechanical properties of clad. In this study an attempt has been made to study the effect hydrogen content (~45-50 ppm in as-fabricated and hydride clads had around 392, 780, and 1200 wppm hydrogen content in homogenised samples and hydrogen contents 393 wppm and having hydride layer thickness 11.8 microns and 480 wppm and having hydride layer thickness 16.2 microns for hydride rim samples) on the transverse mechanical properties of the Indian PHWR Zircaloy clads. It was observed that there was no change in mechanical properties for the clad tubes with rimmed hydrides. In case of uniformly distributed hydrides, upto 400 wppm hydrogen content, there is not much change in elongation. Thereafter, break strain decreases as we increase the hydrogen content of the uniformly hydrided clad tubes. © 2025 The Authors. Published by ELSEVIER B.V. This is an open access article under the CC BY-NC-ND license (https://creativecommons.org/licenses/by-nc-nd/4.0) Peer-review under responsibility of SICE 2024 organizers

Keywords: Zircaloy; Clad; Hydrides; Ring Tensile Test; Ring Compression Test; 1. Introduction

During reactor operation, waterside corrosion of the cladding results in oxide formation and generation of nascent hydrogen. The zircaloy cladding absorbs a proportion of hydrogen produced in the corrosion reaction and when the terminal solid solubility for hydrogen is exceeded, zirconium hydride is precipitated in the cladding. This causes embrittlement of clad tube which reduces the ductility of the cladding and thereby reduces the margin foreseen in clad tube. Circumferential ductility of the clad is very important for in-reactor operation. Mechanical properties of cladding can be estimated by several methods including standard tensile test, burst test and non-standard testing methods such as

Ring Tensile Test (RTT), Ring Compression Test (RCT) etc. ∗ Corresponding author. Tel.: + 91-22-25594043; fax: + 9-22-25505151. E-mail address: pritik@barc.gov.in

2452-3216 © 2025 The Authors. Published by ELSEVIER B.V. This is an open access article under the CC BY-NC-ND license (https://creativecommons.org/licenses/by-nc-nd/4.0) Peer-review under responsibility of SICE 2024 organizers 10.1016/j.prostr.2025.08.037