Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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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))
  • Received : 2018.05.20
  • Accepted : 2018.10.04
  • Published : 2019.04.25
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Abstract

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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References

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