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Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat

[+] Author Affiliations
Yifeng Zhou, Paul Ponomaryov, Cristina Mazza, Igor Pioro

University of Ontario Institute of Technology, Oshawa, ON, Canada

Paper No. ICONE24-60442, pp. V005T15A028; 10 pages
  • 2016 24th International Conference on Nuclear Engineering
  • Volume 5: Student Paper Competition
  • Charlotte, North Carolina, USA, June 26–30, 2016
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5005-3
  • Copyright © 2016 by ASME


Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%).

Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature.

As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors.

Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.

Copyright © 2016 by ASME



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