Experimental and numerical study of phwr specific suspended debris

N. Dutt, P.K. Sahoo

Research output: Contribution to journalArticle

1 Citation (Scopus)

Abstract

In 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR), under a postulated scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident (LOCA) along with the failure of all heat sinks, the fuel channels are likely to heat up due to un-availability of coolant. These events are called Beyond Design Basis Events (BDBEs) where designed engineered safety feature to be not available. The accident sequence leads to moderator boil-off in Calandria. The postulated boil-off leads to slow un-covering of the channels leading to fuel bundle heat up. The exposed channels in the steam environment are known as suspended debris. The present investigation aims to study the heat up behaviour of a single fuel channel of the Calandria for the 220 MWe PHWR. During moderator boil-off phase, it is postulated that decay heat of submerged fuel channels produce steam and acts as a coolant for the suspended channels forming debris like configuration. Experimental study and numerical simulation have been carried out for one-meter, fuel channel at 0.25%–1% of the decay heat. Numerically predicted temperature profile of the outer surface of CT, PT and fuel pins are validated with experimental results. Numerical results are obtained for total heat flux, radiation heat transfer and heat transfer coefficient of the outer surface of CT. Radiation is found to be the dominant mode of heat transfer and for 0.25–1.0% it contributes between 72 and 79% of total heat transfer. From the numerical study, it is concluded that the fuel bundles of the channel are heated maximum up to 804 °C if the decay heat 1% of the rated power. Radiation and convective steam cooling helps to limit the temperature rise of the bundle. © 2018 Elsevier B.V.
Original languageEnglish
Pages (from-to)344-355
Number of pages12
JournalNuclear Engineering and Design
Volume330
DOIs
Publication statusPublished - 2018

Fingerprint

debris
Debris
heat
Steam
heat transfer
steam
bundles
Moderators
moderators
coolants
Coolants
accidents
Enthalpy
accident
decay
Heavy water reactors
radiation
Heat transfer
heavy water reactors
Radiation

Cite this

@article{f5134175c52f41f2b97d1442977b0aa4,
title = "Experimental and numerical study of phwr specific suspended debris",
abstract = "In 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR), under a postulated scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident (LOCA) along with the failure of all heat sinks, the fuel channels are likely to heat up due to un-availability of coolant. These events are called Beyond Design Basis Events (BDBEs) where designed engineered safety feature to be not available. The accident sequence leads to moderator boil-off in Calandria. The postulated boil-off leads to slow un-covering of the channels leading to fuel bundle heat up. The exposed channels in the steam environment are known as suspended debris. The present investigation aims to study the heat up behaviour of a single fuel channel of the Calandria for the 220 MWe PHWR. During moderator boil-off phase, it is postulated that decay heat of submerged fuel channels produce steam and acts as a coolant for the suspended channels forming debris like configuration. Experimental study and numerical simulation have been carried out for one-meter, fuel channel at 0.25{\%}–1{\%} of the decay heat. Numerically predicted temperature profile of the outer surface of CT, PT and fuel pins are validated with experimental results. Numerical results are obtained for total heat flux, radiation heat transfer and heat transfer coefficient of the outer surface of CT. Radiation is found to be the dominant mode of heat transfer and for 0.25–1.0{\%} it contributes between 72 and 79{\%} of total heat transfer. From the numerical study, it is concluded that the fuel bundles of the channel are heated maximum up to 804 °C if the decay heat 1{\%} of the rated power. Radiation and convective steam cooling helps to limit the temperature rise of the bundle. {\circledC} 2018 Elsevier B.V.",
author = "N. Dutt and P.K. Sahoo",
note = "Export Date: 19 June 2018",
year = "2018",
doi = "10.1016/j.nucengdes.2018.02.013",
language = "English",
volume = "330",
pages = "344--355",
journal = "Nuclear Engineering and Design",
issn = "0029-5493",
publisher = "Elsevier Ltd",

}

Experimental and numerical study of phwr specific suspended debris. / Dutt, N.; Sahoo, P.K.

In: Nuclear Engineering and Design, Vol. 330, 2018, p. 344-355.

Research output: Contribution to journalArticle

TY - JOUR

T1 - Experimental and numerical study of phwr specific suspended debris

AU - Dutt, N.

AU - Sahoo, P.K.

N1 - Export Date: 19 June 2018

PY - 2018

Y1 - 2018

N2 - In 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR), under a postulated scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident (LOCA) along with the failure of all heat sinks, the fuel channels are likely to heat up due to un-availability of coolant. These events are called Beyond Design Basis Events (BDBEs) where designed engineered safety feature to be not available. The accident sequence leads to moderator boil-off in Calandria. The postulated boil-off leads to slow un-covering of the channels leading to fuel bundle heat up. The exposed channels in the steam environment are known as suspended debris. The present investigation aims to study the heat up behaviour of a single fuel channel of the Calandria for the 220 MWe PHWR. During moderator boil-off phase, it is postulated that decay heat of submerged fuel channels produce steam and acts as a coolant for the suspended channels forming debris like configuration. Experimental study and numerical simulation have been carried out for one-meter, fuel channel at 0.25%–1% of the decay heat. Numerically predicted temperature profile of the outer surface of CT, PT and fuel pins are validated with experimental results. Numerical results are obtained for total heat flux, radiation heat transfer and heat transfer coefficient of the outer surface of CT. Radiation is found to be the dominant mode of heat transfer and for 0.25–1.0% it contributes between 72 and 79% of total heat transfer. From the numerical study, it is concluded that the fuel bundles of the channel are heated maximum up to 804 °C if the decay heat 1% of the rated power. Radiation and convective steam cooling helps to limit the temperature rise of the bundle. © 2018 Elsevier B.V.

AB - In 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR), under a postulated scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident (LOCA) along with the failure of all heat sinks, the fuel channels are likely to heat up due to un-availability of coolant. These events are called Beyond Design Basis Events (BDBEs) where designed engineered safety feature to be not available. The accident sequence leads to moderator boil-off in Calandria. The postulated boil-off leads to slow un-covering of the channels leading to fuel bundle heat up. The exposed channels in the steam environment are known as suspended debris. The present investigation aims to study the heat up behaviour of a single fuel channel of the Calandria for the 220 MWe PHWR. During moderator boil-off phase, it is postulated that decay heat of submerged fuel channels produce steam and acts as a coolant for the suspended channels forming debris like configuration. Experimental study and numerical simulation have been carried out for one-meter, fuel channel at 0.25%–1% of the decay heat. Numerically predicted temperature profile of the outer surface of CT, PT and fuel pins are validated with experimental results. Numerical results are obtained for total heat flux, radiation heat transfer and heat transfer coefficient of the outer surface of CT. Radiation is found to be the dominant mode of heat transfer and for 0.25–1.0% it contributes between 72 and 79% of total heat transfer. From the numerical study, it is concluded that the fuel bundles of the channel are heated maximum up to 804 °C if the decay heat 1% of the rated power. Radiation and convective steam cooling helps to limit the temperature rise of the bundle. © 2018 Elsevier B.V.

U2 - 10.1016/j.nucengdes.2018.02.013

DO - 10.1016/j.nucengdes.2018.02.013

M3 - Article

VL - 330

SP - 344

EP - 355

JO - Nuclear Engineering and Design

JF - Nuclear Engineering and Design

SN - 0029-5493

ER -