Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

  • Received : 2016.08.09
  • Accepted : 2016.08.17
  • Published : 2016.10.25
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Abstract

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

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References

  1. D. Hahn, Y.-I. Kim, C.B. Lee, S.-O. Kim, J.-H. Lee, Y.-B. Lee, B.-H. Kim, H.-Y. Jeong, Conceptual design of the sodium-cooled fast reactor KALIMER-600, Nucl. Eng. Technol. 39 (2007) 193-206. https://doi.org/10.5516/NET.2007.39.3.193
  2. D. Hahn, J.-W. Chang, Y.I. Kim, Y.-B. Lee, Advanced SFR design concepts and R&D activities, Nucl. Eng. Technol. 41 (2009) 427-446. https://doi.org/10.5516/NET.2009.41.4.427
  3. J. Yoo, J. Chang, H.-K. Joo, Overall reactor system description of prototype Gen-IV sodium cooled fast reactor, Trans. Am. Nucl. Soc. 114 (2016) 694.
  4. System Description of Prototype Gen-IV Sodium-cooled Fast Reactor, SFR-000-SP-403-001, Korea Atomic Energy Research Institute, 2015.
  5. Preliminary Safety Information Document of PGSFR, Korea Atomic Energy Research Institute, 2015.
  6. T. Sofu, A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents, Nucl. Eng. Technol. 47 (2015) 227-239. https://doi.org/10.1016/j.net.2015.03.004
  7. W.R. Robinson, T.H. Bauer, A.E. Wright, E.A. Rhodes, G.S. Stanford, A.E. Klickman, First TREAT transient overpower tests on U-Pu-Zr Fuel: M5 and M6, Trans. Am. Nucl. Soc. 55 (1987) 418.
  8. J.D. Gabor, R.T. Purviance, R.W. Aeschlimann, B.W. Spencer, Characterization of IFR metal fuel fragmentation, Trans. Am. Nucl. Soc. 54 (1987) 251.
  9. J.D. Gabor, R.T. Purviance, R.W. Aeschlimann, B.W. Spencer, Pour stream breakup of molten IFR metal fuel in sodium," ANS Topical Meeting, Safety of Next Generation Power Reactors, Seattle, WA, May 1-5, 1988.
  10. H.P. Planchon, J.I. Sackett, G.H. Golden, R.H. Sevy, Implications of the EBR-II inherent safety demonstration test, Nucl. Eng. Des. 101 (1987) 75-90. https://doi.org/10.1016/0029-5493(87)90152-X
  11. S.W. Lee, W.-P. Chang, K.-S. Ha, S.J. Ahn, S.H. Kang, C.-W. Choi, K.L. Lee, J.-H. Jeong, J.S. Kim, T. Jeong, Evaluation on invessel source term in PGSFR (2015 results), Trans. of the KNS spring meeting, Jeju, Korea, May 12-13, 2016.

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