Study of ballooning of a completely voided pressure tube of Indian PHWR under heat up condition
Creators
- 1. Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee (India)
- 2. Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India)
Description
Highlights: ► Simulating of LOCA at 4 and 6 MPa internal pressures for IPHWR was carried out at IIT Roorkee, India. ► The sagging of PT takes place at 500 °C before the commencement of ballooning. Maximum sagging is less than 2 mm. ► Once PT touches the CT, there is no further rise in PT temperature. ► The maximum PT temperature was below 700 °C. ► The transverse creep correlation for CANDU material also holds good for Indian PT material for the temperature range expected during LOCA. - 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 case of LOCA with/without failure of emergency core cooling system in a Pressurized Heavy Water Reactor, the pressure tube (PT) temperature can rise significantly due to fuel heat up and gross mismatch of 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 using voided PTs at 4 and 6 MPa internal pressure at different heating rates. It is observed that at 4 MPa and 6 MPa internal pressure, the PT sagged at about 500 °C before the ballooning initiation. The ballooning initiates at a temperature around 625 °C and contact between PT and Calandria Tube (CT) occurs at around 680 °C, respectively, for 4 MPa and the same was at 550 °C and 640 °C for 6 MPa. The structural integrity of PT is retained (no breach) for all the experiments. The PT heat up is found to be arrested after the contact between PT and CT, thus establishing the moderator acting as an efficient heat sink for IPHWRs. A thermal creep model 'PTCREEP' has been developed to predict creep behaviour of the PT of IPHWR. It is found that the contact time predicted by PTCREEP is very close to the experimental result. Hence, PTCREEP can be used for the prediction of the ballooning behaviour of the PT for IPHWR in case of LOCA for the operating temperature and pressure range.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.nucengdes.2011.11.007Additional details
Identifiers
- DOI
- 10.1016/j.nucengdes.2011.11.007;
- PII
- S0029-5493(11)00980-0;
Publishing Information
- Journal Title
- Nuclear Engineering and Design
- Journal Volume
- 243
- Journal Page Range
- p. 301-310
- ISSN
- 0029-5493
- CODEN
- NEDEAU
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 43074177
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Descriptors DEI
- BALLOONING INSTABILITY; CALANDRIAS; CANDU TYPE REACTORS; COOLING; CREEP; ECCS; HEAT PIPES; HEAT SINKS; HEAT TRANSFER; HEATING RATE; LOSS OF COOLANT; PHWR TYPE REACTORS; PRESSURE RANGE MEGA PA; PRESSURE TUBES; REACTOR SAFETY; RUPTURES; TEMPERATURE RANGE
- Descriptors DEC
- ACCIDENTS; CONTAINERS; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; FAILURES; HEAVY WATER COOLED REACTORS; HEAVY WATER MODERATED REACTORS; INSTABILITY; MECHANICAL PROPERTIES; PLASMA INSTABILITY; PLASMA MACROINSTABILITIES; POWER REACTORS; PRESSURE RANGE; PRESSURE TUBE REACTORS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; SAFETY; SINKS; THERMAL REACTORS; TUBES
Optional Information
- Copyright
- Copyright (c) 2011 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.