Nuclear Engineering and Technology

  • 제52권7호
  • /
  • Pages.1417-1428
  • /
  • 2020
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지
DOI QR코드

DOI QR Code

Power upgrading of WWR-S research reactor using plate-type fuel elements part I: Steady-state thermal-hydraulic analysis (forced convection cooling mode)

  • Alyan, Adel (Reactors Department, Nuclear Research Center, Atomic Energy Authority) ;
  • El-Koliel, Moustafa S. (Reactors Department, Nuclear Research Center, Atomic Energy Authority)
  • 투고 : 2019.05.19
  • 심사 : 2019.12.21
  • 발행 : 2020.07.25
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MWth. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MWth.

키워드

참고문헌

  1. Safety Analysis Report (SAR) of Egypt's First Research Reactor (ETRR-1), 2001.
  2. Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels Guidebook, 1980. IAEA-TECDOC-233.
  3. I.H. Bokhari, steady-state thermal-hydraulic and safety analysis of a proposed Mixed Fuel (HEU/LEU) Core for Pakistan research R actor, Ann. Nucl. Energy 31 (11) (2004) 1265-1273. https://doi.org/10.1016/j.anucene.2004.03.002
  4. Safety Related Benchmark Calculations for MTR Type Reactor with HEU, MEU and LEU Fuels, TECDOC-643, Vol. vol. 4.
  5. John R. Lamarch, Introduction to Nuclear Engineering, second ed., Addison-Wesley Publishing Series in Nuclear Science and Engineering, 1983.
  6. F.M. Bsebsu, G. Bede, Nuclear reactor channel modeling using THMOD2 code, Kerntechnik 64/5-6 (1999) 269-273. https://doi.org/10.1515/kern-1999-645-611
  7. Carlos Alberto de Oliveria, Miguel Mattar Neto, Flow velocity calculation to avoid instability in a typical research reactor core, in: Int. Nuclear Atlantic Conference - INAC-2011, Belo Horizonte, MG, Brazil 24-28, 2011.
  8. M.M. El-Wakil, Nuclear Heat Transport, International Text book Company, Cranton, Pa., USA, 1971.
  9. Y. Sudo, Study on critical heat flux in rectangular channels heated from one or both sides at pressures ranging from 0.1 to 14 MPa, ASME J. Heat Transfer 118 (Aug) (1996), 680-68. https://doi.org/10.1115/1.2822686
  10. INVAP MPR Document Number 0767-0750-3TPTH-359-1O, Core Thermo Hydraulics- Commissioning Stage, 17 March 1998.
  11. J.E. Matos, E.M. Pennington, K.E. Freese, W.L. Woodruff, Safety-Related Benchmark Calculations for MTR-type Reactors with HEU, MEU and LEU Fuels, vol. 3, Appendix G, IAEA, Vienna, 1992. IAEA-TECDOC-643.
  12. Prasuhn, A. L., Fundamentals of Fluid Mechanics, Prentice-Hall, Inc., Englewood Cliffs, N. J., USA.
  13. R.H. Whittlem, R. Forgan, A correlation for the minima in the pressure drop versus flow rate curves for sub-cooled water flowing in narrow heated channels, Nucl. Eng. Des. 6 (1967).
  14. I.H. Bokhari, M. Israr, S. Pervez, Thermal hydraulic and safety analyses for Pakistan research reactor-1, in: Proceedings of 22nd International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Budapest, Hungary, 1999.
  15. H. El-Sawy Temraz, N. Ashoub, A. Fathallah, Power up-grading study for the first Egyptian research reactor, Nucl. Eng. Des. 208 (2001) 245-254. https://doi.org/10.1016/S0029-5493(01)00359-4
  16. M. Aziz, Neutronic calculations for the new fuel configuration of the ETRR-1 research reactor, Kerntechnik 70 (5-6) (2005) 286-290. https://doi.org/10.3139/124.100257