Open Access
| Issue |
EPJ Web Conf.
Volume 247, 2021
PHYSOR2020 – International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future
|
|
|---|---|---|
| Article Number | 11003 | |
| Number of page(s) | 10 | |
| Section | Reactor Operations | |
| DOI | https://doi.org/10.1051/epjconf/202124711003 | |
| Published online | 22 February 2021 | |
- Bhardwaj, S.A., “The future 700 MWe pressurized heavy water reactor.” Nuclear Engineering and Design 236, 861–871 (2006). [Google Scholar]
- Bajaj, S.S., Gore, A.R., “The Indian PHWR.” Nuclear Engineering and Design 236, 701–722 (2006). [Google Scholar]
- Gupta, S.K., Dutta, B.K., Venkat Raj, V., Kokodkar, A., “A Study of the Indian PHWR Reactor Channel Under Prolonged Deteriorated Flow Conditions.” International Atomic Energy Agency, Vienna (Austria), pp. 331–359 475 (1997). [Google Scholar]
- Mukhopadhyay, D., Majumdar, P., Behera, G., Gupta, S.K., Raj, V.V., “Thermal analysis of severe channel damage caused by a stagnation channel break in PHWR.” J Pressure Vessel Technol., 124, 161–167 (2002). [Google Scholar]
- Ajay, Ketan, Kumar, Ravi, Mukhopadhyay, Deb., Gupta, Akhilesh, Gokhale, Onkar, Das, Arup K., “Experimental investigation of radiation heat transfer in coolant channel under impaired cooling scenario for Indian PHWR.” Nucl. Eng. Des., 347, 45-52 (2019). [Google Scholar]
- D. Bowslaugh, “CHAN-IIA MOD 2.0: Prediction of CANDU Fuel Channel Behaviour Under Prolonged Low Flows—Program Description,” TTR-490, Atomic Energy of Canada Limited. (1993). [Google Scholar]
- B. N. Hanna,. “CATHENA :A thermalhydraulic Code for CANDU analysis.” Nucl. Eng. Des., 180[2], 113–131 (1998). [Google Scholar]
- Kim, Hyoung Tae, Rhee, Bo Wook, Park, Joo Hwan, “Benchmark calculations of a radiation heat tansfer for a CANDU fuel channel analysis using the CFD Code.” Journal of Nuclear Science and Technology, 43:11, 1422-1430 (2012). [Google Scholar]
- Nandan, Gopal, Sahoo, P.K., Kumar, Ravi, Chatterjee, B., Mukhopadhyay, D., Lele, H.G., “Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition.” Nucl. Eng. Des., 240, 3504–3512 (2010). [Google Scholar]
- Yadav, Ashwini K., Majumdar, P., Kumar, Ravi, Chatterjee, B., Gupta, Akhilesh, “Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR.” Nucl. Eng. Des., 255, 138–145 (2013). [Google Scholar]
- Niteanandan, T., July-04, “Test result for the IAEA International Collaborative Standard Problem on Heavy Water Reactor Moderator Subcooling Requirements.” AECL (2014). [Google Scholar]
- Ajay, Ketan, Kumar, Ravi, Mukhopadhyay, Deb., Gokhle, Onkar, Gupta, Akhilesh, & Das, Arup K., “Influence of PT-CT contact on PHWR fuel channel thermal behaviour under accident condition–An experimental study.” Nuclear Engineering and Design, 361, 110543 (2020). [Google Scholar]
- Ajay, Ketan, Kumar, Ravi, & Gupta, Akhilesh, “Thermal analysis of single channel for pressurized heavy water reactor during LOCA-A numerical approach.” In Proceedings of the 25th National and 3rd International ISHMT-ASTFE Heat and Mass Transfer Conference (IHMTC-2019). Begel House Inc., (2019). [Google Scholar]
Current usage metrics show cumulative count of Article Views (full-text article views including HTML views, PDF and ePub downloads, according to the available data) and Abstracts Views on Vision4Press platform.
Data correspond to usage on the plateform after 2015. The current usage metrics is available 48-96 hours after online publication and is updated daily on week days.
Initial download of the metrics may take a while.