Preliminary Analysis of the Cladding Mechanical Behavior of a Nuclear Superheat Boiling Water Reactor

In the present study, investigation of mechanical behaviour of the fuel cladding material for a nuclear superheat Boiling Water Reactor with annular fuel rods, is carried out. In this design, each annular fuel element is cooled internally by steam and externally by water. For the fuel cladding material, radiation embitterment and irradiation-assisted stress corrosion cracking (IASCC) are the most important issues that have to be taken into account. Hence, for cladding, two materials are considered. Preliminary thermal expansion and stress analysis have been done for a fresh (begin of cycle) ASBWR (Annular-fuelled Superheat Boiling Water Reactor) fuel element. The purpose of these analysis is to investigate the stress distribution and thermal expansion of the cladding in the initial phase of operation. The results show that there is a noticeable difference in the axial expansion between the inner and outer claddings. For T91 (modified 9Cr-1Mo steel) cladding, the maximum axial thermal growth of the inner cladding is 22.12 mm, which is about 9.7 mm more than the outer cladding. For Inconel 718 cladding, the results are 27.8 mm and 13.4 mm, respectively.