Abstract—In a nuclear reactor Loss of Coolant accident (LOCA) considers wide range of postulated damage or rupture of pipe in the heat transport piping system. In the case of LOCA with/without failure of emergency core cooling system in a Pressurised Heavy water Reactor, the Pressure Tube (PT) temperature could rise significantly due to fuel heat up and gross mismatch of the heat generation and heat removal in the affected channel. The extent and nature of deformation is important from reactor safety point of view. Experimental set-ups have been designed and fabricated to simulate ballooning (radial deformation) of PT for 220 MWe IPHWRs. Experiments have been conducted by covering the CT by ceramic fibers and then by submerging CT in water of voided PTs. In both the experiments, it is observed that ballooning initiates at a temperature around 665�C and complete contact between PT and Caldaria Tube (CT) occurs at around 700 � C approximately. The strain rate is found to be 0.116% per second. The structural integrity of PT is retained (no breach) for all the experiments. The PT heatup is found to be arrested after the contact between PT and CT, thus establishing moderator acting as an efficient heat sink for IPHWRs. Keywords—Pressure Tube, Calandria Tube, Thermo-mechanical deformation, Boiling heat transfer, Reactor safety I. INTRODUCTION NDIAN PHWRs are of 220 MWe and 500 MWe capacity. The 220 MWe IPHWRs consists of a horizontal reactor core of 306 parallel reactor channels. The coolant flows through half of the channels in one direction and in the remaining 153 channels in the opposite direction. All the reactor channels are submerged in a pool of heavy water called moderator maintained at around 65 o C. The channels are housed in the Calandria Vessel. Each reactor channel consists of a PT of 90 mm outside diameter, which is concentrically placed in a CT of 110 mm outside diameter. Short fuel bundles are housed in PT. The gap between PT and Gopal Nandan is with Mechanical and Industrial Engineering Department, Indian Institute of Technology, Roorkee India, (email:gopalnandan@gmail.com) Pradeep K Sahoo is with Mechanical and Industrial Engineering Department, Indian Institute of Technology, Roorkee India, (e-mail: sahoofme@iitr.ernet.in). Ravi Kumar is with Mechanical and Industrial Engineering Department, Indian Institute of Technology, Roorkee India, (e-mail: ravikfme@iitr.ernet.in) Barun Chatterjee with Reactor Safety Division, Bhabha Atomic Research Centre, India (e-mail:barun@barc.gov.in) D. Mukhopadhyay is with with Reactor Safety Division, Bhabha Atomic Research Centre, India (e-mail: dmukho@barc.gov.in ) H. G. Lele is with the Reactor Safety Division, Bhabha Atomic Research Centre, India (e-mail: hglele@barc.gov.in) CT is 8.95 mm and is filled with CO 2 for thermal insulation. The PT is supported along its length through tight garter springs as shown in Fig. 1. INLET OUTLET PRESSURE TUBE CALANDRIA TUBE GARTER SPRING CALANDRIA TUBESHEET Fig. 1 Schematic of Indian PHWR Reactor Channel PT and CT are made of Zirconium 2.5 wt% Nb and Zircaloy-2 material respectively. Nuclear heat is removed from fuel bundles by heavy water coolant and transferred to the Steam Generators secondary side, where the secondary side water boil-off generate steam at 40 bar pressure for turbine. The coolant after releasing the heat in Steam Generator returns back to other half of the reactor channels through centrifugal pumps. The schematic of Primary Heat Transport System is shown in Fig. 2. During postulated low frequency events like LOCA along with the failure of the Emergency Core Cooling System (ECCS), the cooling environment for the bundles degrades that results heatup of the fuel bundles [1], and in turn heatup the PT through radiation heat transfer. The heat flux incident on the surface of PT during such event is equivalent to that of decay power (2% - 1% of nominal power) as the reactor will undergo shutdown during such situation. However, the temperature of the CT is not affected significantly as it is submerged in the low temperature moderator. In such event, CT will experience a high rate of heat transfer from its surface to bulk moderator by various mode of pool boiling heat transfer. The rise in temperature of the PT will lead to deterioration in its thermo- mechanical properties. The pressure inside the PT could be in the range of 0.1 to 9 MPa. If the internal pressure is lower than 1.0 MPa, the PT deforms (sags) due to high temperature creep and due to its own weight and weight of the fuel bundles. Ballooning deformation takes place when internal pressure is more than 1.0 MPa. The deformation of the PT leads to a physical contact between the PT and CT, thereby resulting in high heat transfer to the moderator. Enhanced heat transfer from PT to Ct arrests the rise in temperature of the fuel bundles and PT. It is an important Thermo-mechanical Behavior of Pressure Tube of Indian PHWR at 20 bar Pressure Gopal Nandan a , P. K. Sahoo a,* , Ravi Kumar a , B Chatterjee b , D. Mukhopadhyay b , H. G. Lele b I World Academy of Science, Engineering and Technology 61 2010 205