Nuclear Engineering and Technology

  • 제37권4호
  • /
  • Pages.363-374
  • /
  • 2005
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지

CORE DESIGN FOR HETEROGENEOUS THORIUM FUEL ASSEMBLIES FOR PWR (II) - THERMAL HYDRAULIC ANALYSIS AND SPENT FUEL CHARACTERISTICS

  • BAE KANG-MOK (Korea Atomic Energy Research Institute) ;
  • HAN KYU-HYUN (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology) ;
  • KIM MYUNG-HYUN (Department of Nuclear Engineering, Kyung Hee University) ;
  • CHANG SOON-HEUNG (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology)
  • 발행 : 2005.08.01
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

키워드

참고문헌

  1. A. Galperin, E. Shwageraus and M. Todosow, 'Thorium Fuel cycles for Light Water Reactors: Homogeneous Design,' Trans. Am. Nucl. Soc., 83, November (2000)
  2. D. Wang, M. J. Driscoll and M. S. Kazimi, 'Design and Performance Assessment of a PWR Whole-Assembly Seed and Blanket Thorium Based Fuel cycle,' MIT-NFC-TR-026, MIT Nuclear Engineering Department, September (2000)
  3. K.H. Kim, K.M. Bae and M.H. Kim, 'Optimization of Thorium-Based Seed and Blanket Fuel Assembly Design for PWR,' Trans. Am. Nucl. Soc., Vol. 86, pp. 302-303, June (2002)
  4. K.M. Bae and M.H. Kim, 'Core Design for Heterogeneous Thorium Fuel Assemblies for PWR(I)-Nuclear Design and Fuel Cycle Economy,' Nuclear Engineering and Technology, 37, 91 (2005)
  5. B.M. Ma, 'Nuclear Reactor Materials and Application,' Van Nostrand Reinhold Company Inc. (1983)
  6. Y.J. Yoo and D.H. Hwang, 'Development of a sub-channel Analysis Code MATRA,' KAERI/TR-1088/98 (1998)
  7. W.J. Lee et al., 'Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification,' KAERI/TR-1108/98 (1998)
  8. RELAP5-3D Code Manual Volume 1: Code Structure, System Models and Solution Methods, INEEL-EXT-98-00834, Revision 2.2, October (2003)
  9. A. G. Croff, 'A User's Manual for the ORIGEN-2 Computer Code,' ORNL TM-7175, Oak Ridge National Laboratory, July (1980)
  10. D. Wang, 'Optimization of Seed and Blanket Thorium-Uranium Fuel Cycle for Pressurized Water Reactors,' Ph.D. Thesis, MIT (2003)
  11. C.W. Forsberg, C.M. Hopper and H.C. Vantine, 'What is Nonweapons-Usable U-233?' Trans. Am. Nucl. Soc., 81, November (1999)