Experimental and numerical study of phwr specific suspended debris

• Published in: Nuclear engineering and design Vol. 330; pp. 344 - 355
• Main Authors: Dutt, Nitesh, Sahoo, Pradeep Kumar
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 15.04.2018; Elsevier BV
• Subjects: Bundling; Channels; Computer simulation; Cooling rate; Debris; Decay; Design engineering; Detritus; Enthalpy; Heat exchangers; Heat flux; Heat sinks; Heat transfer; Heat transfer coefficients; Heavy water; IPHWR; LOCA; Loss of coolant accidents; Mathematical models; Nuclear engineering; Nuclear fuels; Nuclear safety; Numerical analysis; Numerical prediction; PHWR; Pressure; Pressurized heavy water; Radiation; Reactors; Safety engineering; Simulation; Steam electric power generation; Submerged fuel channel; Temperature; Water; LOCA; IPHWR; Submerged fuel channel
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2018.02.013

•CT outer surface obtained maximum temperature at the contact point of CT-PT.•PT outer surface obtained minimum temperature at the contact point of CT-PT.•Heat flux and associated heat transfer coefficient is maximum at the bottom part of the CT.•Maximum temperature of the fuel pin is 804 °C, which is well below the melting point of the cladding material. 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.