Thermo-mechanical Behavior of Pressure Tube of Indian PHWR at 20 bar Pressure
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.
[1] S. K. Gupta, B. K. Dutta, V. Venkatraj, A. Kakodkar, "A study of Indian
PHWR reactor channel under prolonged deteriorated flow conditions", in
Proc. IAEA TCM on advances in heavy water reactor, Bhabha Atomic
Research Centre, India, 1996, pp. 1-20.
[2] P. D. Thomption, E. Kohn, "Fuel and fuel channel behavior in accident
without the availability of the emergency coolant injection system",
Specialist meeting on water reactor fuel safety and fission product release
in off Normal Reactor Accidents, IWGFPT/16 (May 1983).
[3] K. E. Locke, J. C. Luxat, A. P. Majumdar, C. B. So, R. G. Moyer, D. G.
Litke, "Progress on SMARTT simulation of pressure tube circumferential
temperature distribution experiments test 1 to 4", in Proc. CNS 8th Annual
Conference, Saint John, June 16 - 17, 1987.
[4] C. B. So, G. E. Gillespie, R. G. Moyer, D. G. Litke, "The experimental
determination of circumferential temperature distributions developed in
pressure tube during slow coolant boildown", in Proc. CNS 8th Annual
Conference, Saint John, 1987, pp. 241-248.
[5] P. S. Yuen, C. B. So, R. G. Moyer, D. G. Litke, "The experimental
measurement of circumferential temperature distributions developed on
pressure tubes under stratified two-phase of conditions", in Proc. CNS 9th
Annual Conference, Winnipeg, Manitoba, 1988, pp. 120-126.
[6] P. S. Yuen, K. A. Haugen, D. G. Litke, R. G. Moyer, H. E. Rosinger, The
experimental measurements of circumferential temperature distributions
developed on pressure tubes under stratified two-phase flow conditions:
Tests 1 to 5, in Proc. CNS 10th Annual Conference, AECL, Pinawa, 1989,
pp. 9-18.
[7] G. E. Gillespie, An experimental investigation of heat transfer from a
reactor fuel channel to surrounding water, in Proc. CNS 2nd Annual
Conference, Ottawa, Toronto, 1981, pp. 157-163.
[8] R. S. W. Shewfelt, L. W. Layall, D. P. Godin, "High temperature creep
model for Zr-2.5 wt % Nb pressure tubes", Journal of Nuclear materials
vol. 125, pp. 228-235, 1984.
[9] Gopal Nandan, P. K. Sahoo, Ravi Kumar, B Chatterjee, D.
Mukhopadhyay, H. G. Lele, "Experimental investigation of heat transfer
during LOCA with failure of emergency cooling system", in Proc. 5th
International Conference on Heat Transfer, Fluid Mechanics and
Thermodynamics (HEFAT2007), Sun City, South Africa,1 - 4 July 2007
[1] S. K. Gupta, B. K. Dutta, V. Venkatraj, A. Kakodkar, "A study of Indian
PHWR reactor channel under prolonged deteriorated flow conditions", in
Proc. IAEA TCM on advances in heavy water reactor, Bhabha Atomic
Research Centre, India, 1996, pp. 1-20.
[2] P. D. Thomption, E. Kohn, "Fuel and fuel channel behavior in accident
without the availability of the emergency coolant injection system",
Specialist meeting on water reactor fuel safety and fission product release
in off Normal Reactor Accidents, IWGFPT/16 (May 1983).
[3] K. E. Locke, J. C. Luxat, A. P. Majumdar, C. B. So, R. G. Moyer, D. G.
Litke, "Progress on SMARTT simulation of pressure tube circumferential
temperature distribution experiments test 1 to 4", in Proc. CNS 8th Annual
Conference, Saint John, June 16 - 17, 1987.
[4] C. B. So, G. E. Gillespie, R. G. Moyer, D. G. Litke, "The experimental
determination of circumferential temperature distributions developed in
pressure tube during slow coolant boildown", in Proc. CNS 8th Annual
Conference, Saint John, 1987, pp. 241-248.
[5] P. S. Yuen, C. B. So, R. G. Moyer, D. G. Litke, "The experimental
measurement of circumferential temperature distributions developed on
pressure tubes under stratified two-phase of conditions", in Proc. CNS 9th
Annual Conference, Winnipeg, Manitoba, 1988, pp. 120-126.
[6] P. S. Yuen, K. A. Haugen, D. G. Litke, R. G. Moyer, H. E. Rosinger, The
experimental measurements of circumferential temperature distributions
developed on pressure tubes under stratified two-phase flow conditions:
Tests 1 to 5, in Proc. CNS 10th Annual Conference, AECL, Pinawa, 1989,
pp. 9-18.
[7] G. E. Gillespie, An experimental investigation of heat transfer from a
reactor fuel channel to surrounding water, in Proc. CNS 2nd Annual
Conference, Ottawa, Toronto, 1981, pp. 157-163.
[8] R. S. W. Shewfelt, L. W. Layall, D. P. Godin, "High temperature creep
model for Zr-2.5 wt % Nb pressure tubes", Journal of Nuclear materials
vol. 125, pp. 228-235, 1984.
[9] Gopal Nandan, P. K. Sahoo, Ravi Kumar, B Chatterjee, D.
Mukhopadhyay, H. G. Lele, "Experimental investigation of heat transfer
during LOCA with failure of emergency cooling system", in Proc. 5th
International Conference on Heat Transfer, Fluid Mechanics and
Thermodynamics (HEFAT2007), Sun City, South Africa,1 - 4 July 2007
@article{"International Journal of Mechanical, Industrial and Aerospace Sciences:56218", author = "Gopal Nandan and P. K. Sahooa and Ravi Kumara and B Chatterjeeb and D. Mukhopadhyayb and H. G. Leleb", title = "Thermo-mechanical Behavior of Pressure Tube of Indian PHWR at 20 bar Pressure", 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-mechanicaldeformation, Boiling heat transfer, Reactor safety", volume = "4", number = "1", pages = "29-9", }