Nuclear Engineering and Technology

  • 제41권3호
  • /
  • Pages.307-318
  • /
  • 2009
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지
DOI QR코드

DOI QR Code

THE BENCHMARK CALCULATIONS OF THE GAMMA+ CODE WITH THE HTR-10 SAFETY DEMONSTRATION EXPERIMENTS

  • 발행 : 2009.04.30
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

키워드

참고문헌

  1. Jonghwa Chang, et al, “A Study of a Nuclear Hydrogen Production Demonstration Plant”, Nuclear Engineering and Technology, Vol. 39 No.2, 2007.4 https://doi.org/10.5516/NET.2007.39.2.111
  2. Hong Sik Lim, Hee Cheon No, “GAMMA Multidimensional Multicomponent Mixture Analysis to Predict Air Ingress Phenomena in an HTGR”, Nuclear Science and Engineering 152 (2006) 1-11, 2006
  3. James J. Duderstadt, Louis J. Hamilton, “Nuclear Reactor Analysis”, John Wiley & Sons, Inc, 1976
  4. INET, “HTR-10 Experiments Report - Loss of Helium Flow without Reactor Scram”, NH-JRC-P1-R06-009, KAERI/INET Internal Report, 2006.5
  5. INET, "HTR-10 Experiments Report - Reactivity Insertion without Reactor Scram", NH-JRC-P1-R06-010, KAERI/INET Internal Report, 2006.5
  6. Zongxin Wu, Dengcai Lini, Daxin Zhong, “The Design Features of the HTR-10”, Nuclear Engineering and Design 218 (2002) 25-32, 2002 https://doi.org/10.1016/S0029-5493(02)00182-6
  7. Ji Su Jun, HongSik Lim, Won-Jae Lee, “The Preliminary GAMMA Code Thermal Hydraulic Analysis of the HTR- 10 Steady State Initial Core”, 2006 KNS Spring Conference, 2006. 5
  8. Ji Su Jun, Hong Sik Lim, Won-Jae Lee, “Final Report of IAEA CRP5 Benchmark Problem of HTR-10 Steady State Temperature Distribution (Rev. 1)”, NHDD-KA06- RD011-01, KAERI Internal Report, 2006.11
  9. Ji Su Jun, Hong Sik Lim, Won-Jae Lee, “The GAMMA Code Assessment of the HTR-10 Safety Demonstration Experiments - LOFC ATWS”, 2007 KNS autumn conference, 2007.10
  10. Ji Su Jun, Hong Sik Lim, Jae Man Noh, Won-Jae Lee, “The GAMMA Code Assessment of the HTR-10 Safety Demonstration Experiments - CRW ATWS”, 2008 KNS spring conference, 2008.5
  11. Hyun Chul Lee, Seung Wook Lee, Jae Man Noh, Won Jae Lee, “MARS-GCR/CAPP Coupled Multi-Physics Calculation for the OECD/NEA PBMR-400 Benchmark Problem”, KNS Spring Meeting, Gyeongju, Korea, May 29-30, 2008
  12. IAEA, “Benchmark Problems of the HTR-10 ; Prepared for the IAEA CRP-5 by INET”, IAEA-TECDOC-xxx, on going version, 2005. 12
  13. D.A. Nield, A. Bejan, "Convection in Porous Media", Springer-Verlag, New York, 1999
  14. “Numerical Recipes: The Art of Scientific Computing”, Cambridge University Press, 2007
  15. J.P. Holaman, “Heat Transfer”, McGraw-Hill, 1986
  16. Shouyin Hu, Ruipian Wang, Zuying Gao, “Transient Tests on Blower Trip and Rod Removal at the HTR-10 ”, Nuclear Engineering and Design 236 (2006) 677-680, 2006 https://doi.org/10.1016/j.nucengdes.2005.11.015
  17. Baumer R., et al, “AVR: Experimental High Temperature Reactor ; 21 Years of Successful Operation for a Future Energy Technology/Association of German Engineers (VDI)”, The Society for Energy Technologies (Publ.), VDIVerl, 1990

피인용 문헌

  1. A Practical Method for Whole-Core Thermal Analysis of a Prismatic Gas-Cooled Reactor vol.177, pp.3, 2012, https://doi.org/10.13182/NT12-A13480