Nuclear Engineering and Technology

  • 제51권2호
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  • Pages.356-368
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  • 2019
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Verification and validation of STREAM/RAST-K for PWR analysis

  • Choe, Jiwon (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Choi, Sooyoung (Research Division of Mechanical, Aerospace and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Zhang, Peng (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Park, Jinsu (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Kim, Wonkyeong (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Shin, Ho Cheol (Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI)) ;
  • Lee, Hwan Soo (Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI)) ;
  • Jung, Ji-Eun (Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI)) ;
  • Lee, Deokjung (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST))
  • 투고 : 2018.05.20
  • 심사 : 2018.10.04
  • 발행 : 2019.04.25
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초록

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

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참고문헌

  1. SSP-09/442-U, CASMO-4E: Extended Capability CASMO-4 User's Manual, Studsvik Scandpower, 2009.
  2. SSP-09/447-U, SIMULATE-3: Advanced Three-Dimensional Two-Group Reactor Analysis Code User's Manual, Studsvik Scandpower, 2009.
  3. T. Downar, Y. Xu, V. Seker, PARCS v3.0 U.S.NRC Core Neutronics Simulator User Manual, UM-NERS-09-0001, University of Michigan, 2013.
  4. R.J.J. Stamml'er, HELIOS Methods, Studsvik Scandpower, 2002.
  5. Jin Young Cho, Jae Seung Song, Kyung Hoon Lee, Three dimensional nuclear analysis system DeCART/CHORUS/MASTER, in: ANS Annual Meeting, Atlanta, June 16-20, 2013.
  6. Chang Ho Lee, Byung Oh Cho, Jae Seung Song, Jae Seong Kim, Ha Yong Kim, Sung Quun Zee, Hyung Kook Joo, Verification of Extended Nuclide Chain of MASTER with CASMO-3 and HELIOS, KAERI/TR-947/98, Korea Atomic Energy Research Institute, 1998.
  7. R.A. Loretz, et al., User's Manual for DIT: Discrete Integral Transport Assembly Design Code, CE-CES-11 REV 4-P, April 1994.
  8. R.A. Loretz, et al., User's Manual for ROCS: Coarse and Fine Mesh Advanced Diffusion Theory Code for Reactor Core Analysis, CE-CES-4-P Rev 15, January 2003.
  9. TR-KHNP-0008, Rev. 1, Qualification of PARAGON/ANC Code System for PWR Applications, May 2007.
  10. WCAP-10965-P-A, ANC: A Westinghouse Advanced Nodal Computer Code, September 1986.
  11. T.Q. Nguyen, et al., Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11596-P-A, June, 1988.
  12. Tae Young Han, Joo Il Yoon, Jae Hee Kim, Chang Kyu Lee, Beom Jin Cho, Benchmark verification of the KARMA/ASTRA Code with OECD/NEA and USNRC PWR MOX/UO2 transient problem, in: Proceedings of ICAPP 2011, France, 2011.
  13. Jiwon Choe, Sooyoung Choi, Minyong Park, Peng Zhang, Ho Cheol Shin, Hwan Soo Lee, Deokjung Lee, Validation of the UNIST STREAM/RAST-K Code System with OPR -1000 Multi-cycle Operation, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  14. Sooyoung Choi, Jiwon Choe, Jaerim Jang, Deokjung Lee, Extension of PSM for Ring-type Burnable Absorber Containing Resonant Nuclides, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  15. Sooyung Choi, Minyong Park, Youqi Zheng, Chidong Kong, Jiwon Choe, Hanjoo Kim, Kiho Kim, Ho Cheol Shin, Deokjung Lee, Development status of reactor physics code suite in UNIST, in: 11st International Conference of the Croatian Nuclear Society, Zadar, Croatia, June 5-8, 2016, Croatian Nuclear Society, 2016.
  16. Hanjoo Kim, Jinsu Park, Jiwon Choe, Jiankai Yu, Deokjung Lee, Multi-physics Coupled Reactor Core Analysis System of RAST-K2.0 with CTF and FRAPCON, in: KNS Spring Meeting, Jeju, Korea, May 16-18, 2018.
  17. Sooyung Choi, Chang Ho Lee, DeokjunG Lee, Resonance treatment using pin-based pointwise energy slowing-down method, J. Comp. Phys. 330 (2017) 134-155. https://doi.org/10.1016/j.jcp.2016.11.007
  18. Maria Pusa, Rational approximation to the matrix exponential in burnup calculations, Nucl. Sci.. Eng. 169 (2011) 155-167. https://doi.org/10.13182/NSE10-81
  19. Jinsu Park, Wonkyeong Kim, Sooyoung Choi, Hyunsuk Lee, Deokjung Lee, Comparative analysis of VERA depletion problems, in: KNS Fall Meeting, Gyeongju, Korea, October 26-28, 2016.
  20. Sooyoung Choi, Pin-based Pointwise Energy Slowing-down Method for Resonance Self-Shielding Calculation, Doctoral Thesis, UNIST, 2017.
  21. Bamidele Ebiwonjumi, Sooyoung Choi, Matthieu Lemaire, Deokjung Lee, Ho Cheol Shin, Experimental Validation of STREAM for Spent Nuclear Fuel Applications, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  22. Hyun Chul Lee, Unified nodal method for static and Transient Analysis of Power Reactor, Thesis (doctoral), Seoul National University, 2001, http://hdl.handle.net/10371/34547.
  23. Z. Weiss, A consistent definition of the number density of pseudo-isotopes, Ann. Nucl. Energy. 17 (3) (1990) 153. https://doi.org/10.1016/0306-4549(90)90093-S
  24. T. Bahadir, S.O. Lindahl, S.P. Palmtag, SIMULATE-4 Multigroup Nodal Code With Microscopic Depletion Model, Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, Palais des Papes, Avignon, France, 2005. Sept 12-15, 2005, on CD-ROM, ANS LaGrange Park, IL.
  25. Jinsu Park, Minyong Park, Jiwon Choe, Peng Zhang, Jaerim Jang, Deokjung Lee, Development Status of Dynamic Reactor Nodal Computational Code RAST-K v2.0, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  26. Y.S. Jung, C.B. Shim, C.H. Lim, H.G. Joo, Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers, Ann. Nucl. Energ. 62 (2013) 357-374. https://www.sciencedirect.com/science/article/pii/S0306454913003344. https://doi.org/10.1016/j.anucene.2013.06.031
  27. KNF-S1ICD-10004 Rev. 0, The Nuclear Design Report for Shin-Kori Nuclear Power Plant Unit 1 Cycle 1, Korea Nuclear Fuel Company, Ltd, February 2010.
  28. KNF-U4C7-06017 Rev. 0, The Nuclear Design Report for Ulchin Nuclear Power Plant Unit 4 Cycle 7, Korea Nuclear Fuel Company, Ltd, June 2006.
  29. KNF-S3ICD-12034 Rev. 0, The Nuclear Design Report for Shin-Kori Nuclear Power Plant Unit 3 Cycle 1, KEPCO Nuclear Fuel Company, Ltd., October 2012.
  30. KNF-K3C19-08015, The Nuclear Design and Core Physics Characteristics of the Kori Nuclear Power Plant Unit 3 Cycle 19, Korea Nuclear Fuel Company, Ltd., May 2008.
  31. The Nuclear Design Report for Yonggwang Nuclear Power Plant Unit 1 Cycle 19, KNF-Y1C19-09008, Korea Nuclear Fuel Company, Ltd.
  32. Woonghee Lee, Sooyoung Choi, Bamidele Ebiwonjumi, Matthieu Lemaire, Deokjung Lee, Implementation of On-The-Fly Energy Release per Fission Model in STREAM, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  33. Deokjung Lee, Joel Rhodes, Kord Smith, Quadratic depletion model for gadolinium isotopes in CASMO-5, Nucl. Sci.. Eng. 174 (2013) 79-86. https://doi.org/10.13182/NSE12-20.
  34. Deokjung Lee, Kord Smith, Joel Rhodes, The impact of U-238 resonance elastic scattering approximations on thermal reactor doppler reactivity, Annal. Nucl. Energy 36 (3) (2009) 274-280. https://doi.org/10.1016/j.anucene.2008.11.026.
  35. Kord S. Smith, Nodal Diffusion Methods: Understing Numerous Unpublished Details, PHYSOR2016, Sun Valley, ID, USA, May 1-5, 2016.
  36. K.J. Geelhood, W.G. Luscher, P.A. Raynaud, I.E. Porter, FRAPCON-4.0: a Computer Code for the Calculation of Steady-state, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup, Vol. 1, Pacific Northwest National Laboratory, Richland, WA, 2015. PNNL-19417, Rev. 2.
  37. K.J. Geelhood, W.G. Luscher, C.E. Beyer, FRAPCON-4.0: Integral assessment, Vol. 2, Pacific Northwest National Laboratory, 2015. PNNL-19418, Rev. 2.
  38. Tung Dong, Cao Nguyen, Hyunsuk Lee, Jiwon Choe, Ho Cheol Shin, Hwan Soo Lee, Deokjung Lee, LPPT Analysis of APR1400 Reactor Core by UNIST Monte Carlo Code MCS, RPHA17, Chengdu, Sichuan, China, August 24-25, 2017.
  39. Pre-Operational Inspection Report of Shin-Kori Nuclear Power Plant Unit 3 (Initial Fuel Load and Startup Test, Vol. 2, Korea Institute of Nuclear Safety, 2016. KINS/AR-1008, http://nsic.nssc.go.kr/dta/reguResultView.do?seq=882.

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