본문 바로가기
KERD
논문KCI 등재

Comparative study of constitutive relations implemented in RELAP5 and TRACE – Part I: Methodology & wall friction

0

조회

0

다운로드

1

인용

0

참고문헌

키워드

주제분류원자력공학

초록

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

Abstract

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

참고문헌 (64)

  1. W. Li, "An investigation of numerical performance enhancement of RELAP5: numerical stability, higher resolution and an alternative constitutive relation", Nucl. Eng. Des, 328, pp.309~320 (2018)
  2. W. Kolar, "Programming for Software Sharing", pp.265~274 (1983), Springer
  3. D. Bestion, "System Code Models and Capabilities", ~ (2008)
  4. A. E. Bergles, "Anticipated and abnormal plant transients in light water reactorsdvolumes 1 and 2", Nucl. Technol, 81(1), pp.126~127 (1988)
  5. 하태욱, "Modification of the fast fourier transform-based method by signal mirroring for accuracy quantifi cation of thermal-hydraulic system code", Nuclear Engineering and Technology, 49(5), pp.1100~1108 (2017), 한국원자력학회
  6. 김연식, "FIRST ATLAS DOMESTIC STANDARD PROBLEM (DSP-01)FOR THE CODE ASSESSMENT", Nuclear Engineering and Technology, 43(1), pp.25~44 (2011), 한국원자력학회
  7. C. Song, "TRACE code calculations of MIT pressurizer tests for review of SPACE code capability", Transactions, 111(1), pp.1570~1571 (2014)
  8. A. Shin, "Assessment of TRACE 4.050 using UPTF bypass tests", Trans. Am. Nucl. Soc, 93, pp.538~538 (2005)
  9. Y. S. Bang, "Multidimensional prediction of the effect of flow blockage in SCTF S1-01 experiment using MARS-KS code", Ann. Nucl. Energy, 152, pp.108028~ (2021)
  10. V. Ransom, "RELAP5", MOD2 code manual, 1987
  11. U. Nrc, "TRACE V5. 0 Assessment ManualeMain Report(Draft)", US Nuclear Regulatory Commission, ~ (2005)
  12. Code Manual Volume IV: Developmental Assessment Report, "M. Kaeri", ~, pp.KAERI (2009), TR-3042
  13. S. Ahn, "FONESYS : the FOrum & NEtwork of SYStem thermal-hydraulic codes in nuclear reactor thermal-hydraulics", Nucl. Eng. Des, 281, pp.103~113 (2015)
  14. S. Ahn, "Hyperbolicity and numerics in SYS-TH codes : the FONESYS point of view", Nucl. Eng. Des, 322, pp.227~239 (2017)
  15. RELAP5, "MOD1: Code Manual Vol. 1: System Models and Numerical Methods", V. Ranson, 1981, NUREG[DOI]
  16. TRACE V5. 0 Theory ManualeField Equations, Solution Methods and Physical Models, "U. Nrc", United States Nucl. Regul. Comm, ~ (2010)
  17. G. Mesina, "A history of RELAP computer codes", Nucl. Sci. Eng, 182(1) (2016)
  18. Development of Transient-Reactor Analysis Code (TRAC) for Real-Time Applications, "G. F. Niederauer", Los Alamos National Lab, ~ (1997)
  19. RELAP5, "MOD3. 3 code manual Volume IV:models and correlations", V. H. Ransom, 2001
  20. Y. -G. Lee, "Implementation of a New Empirical Model of Steam Condensation for the Passive Containment Cooling System into MARS-KS Code: Application to Containment Transient Analysis, Nuclear Engineering and Technology", ~ (2021)
  21. P. Wu, "Improvement and validation of the wall heat transfer package of RELAP5", MOD3. 3, 2016, pp.310, 418~428
  22. 이일석, "Effect of critical flow model in MARS-KS code on uncertainty quantifi cation of large break Loss of coolant accident (LBLOCA)", Nuclear Engineering and Technology, 52(4), pp.755~763 (2020), 한국원자력학회
  23. YOUNG-SUK BANG, "Improvements of Condensation Heat Transfer Models in MARS Code for Laminar Flow in Presence of Noncondensable Gas", Nuclear Engineering and Technology, 41(8), pp.1015~1024 (2009), 한국원자력학회
  24. J. J. Jeong, "Improvement of the subchannel flow mixing model of the MARS code", Nucl. Technol, 156(3), pp.360~368 (2006)
  25. Anis Bousbia Salah, "CATHARE SIMULATION RESULTS OF NATURAL CIRCULATION CHARACTERISATION TEST OF THE PKL TEST FACILITY", Nuclear Engineering and Technology, 53(5), pp.1446~1453 (2021), 한국원자력학회[DOI]
  26. Yoon Ho Joon, "Validation of RELAP5 MOD3.3 code for Hybrid-SIT against SET and IET experimental data", Nuclear Engineering and Technology, 52(9), pp.1926~1938 (2020), 한국원자력학회[DOI]
  27. 김연식, "Second ATLAS Domestic Standard Problem (DSP-02) for a Code Assessment", Nuclear Engineering and Technology, 45(7), pp.871~894 (2013), 한국원자력학회
  28. T. S. Choi, "Improvement of the reflood model of RELAP5", MOD3. 3based on the assessments against FLECHT-SEASET tests, 2010, pp.240(4), 832~841
  29. T. S. Choi, "An improved RELAP5", MOD3. 3 reflood model considering the effect of spacer grids, 2012, pp.250, 613~625
  30. T. W. Ha, "Development of an empirical correlation for the onset of flow instability in narrow rectangular channels", Nucl. Eng. Des, 375, pp.111090~ (2021)
  31. S. Bae, "An implementation and assessment of the viscous stress model in the MARS multi-dimensional component", Intern. Conf. Nucl. Eng, 4689, pp.673~679 (2004)
  32. Pengcheng Zhao, "Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor", Nuclear Engineering and Technology, 52(12), pp.2789~2802 (2020), 한국원자력학회[DOI]
  33. M. G. Kim, "Preliminary assessment of the nuclear thermalehydraulic system code MARS for the application to a refrigeration cycle", Nucl. Eng. Des, 367, pp.110798~ (2020)
  34. S. -J. Kang, "Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code", Ann. Nucl. Energy, 115, pp.138~153 (2018)
  35. J. H. Park, "Transient analysis and validation with experimental data of supercritical CO2 integral experiment loop by using MARS", Energy, 147, pp.1030~1043 (2018)
  36. F. D'Auria, "Design, V&V&C in Nuclear Reactor Thermal-Hydraulics", 110162~ (2019)
  37. F. D'Auria, "Moving from V&V to V&V&C in nuclear thermalhydraulics", 2018 26th International Conference on Nuclear Engineering, American Society of Mechanical Engineers Digital Collection (2018)
  38. G. Bandini, "Assessment of systems codes and their coupling with CFD codes in thermalehydraulic applications to innovative reactors", Nucl. Eng. Des, 281, pp.22~38 (2015)
  39. T. Skorek, "Quantification of the uncertainty of the physical models in the system thermal-hydraulic codesePREMIUM benchmark", Nucl. Eng. Des, 354, pp.110199~ (2019)
  40. G. A. Roth, "Theory and implementation of nuclear safety system codesePart I: conservation equations, flow regimes, numerics and significant assumptions", Prog. Nucl. Energy, 76, pp.160~182 (2014)
  41. G. A. Roth, "Theory and implementation of nuclear safety system codesePart II: system code closure relations, validation, and limitations", Prog. Nucl. Energy, 76, pp.55~72 (2014)
  42. W. Marcum, "Benchmark comparison of experimental data with thermal-hydraulic codes RELAP5-3D and TRACE for a RIA transient scenario", Nucl. Technol, 206(6), pp.911~923 (2020)
  43. Y. Lee, "Influence of two-phase crossflow for void prediction in bundles using thermal-hydraulic system codes", Energies, 13(14), pp.3686~ (2020)
  44. D. Griggs, "Comparison of TRAC and RELAP5 Reactor System Calculations for a DEGB LOCA in K-14. 1", Westinghouse Savannah River Co., ~ (1990)
  45. V. N. Blinkov, "Investigation on the interphase drag and wall friction in vertically oriented upward and downward two-phase flows under accident conditions in light water reactors", Nucl. Eng. Des, 389, pp.111666~ (2022)
  46. R. Lockhart, "Proposed correlation of data for isothermal twophase, two-component flow in pipes", Chem. Eng. Prog, 45(1), pp.39~48 (1949)
  47. M. Davis, "Wall friction for two-phase bubbly flow in rough and smooth tubes", Int. J. Multiphas. Flow, 16(5), pp.921~927 (1990)
  48. K. T. Claxton, "HTFS Correlations for Two-phase Pressure Drop and Void Fraction in Tubes", Heat Transfer and Fluid Flow Service (1972)
  49. D. Zigrang, "A review of explicit friction factor equations", J. Energy Resour. Technol, 107(2), pp.280~283 (1985)
  50. C. F. Colebrook, "Correspondence. Turbulent flow IN pipes, with particular reference to the transition region between the smooth and rough pipe LAWS.(INCLUDES plates)", J. Inst. Civil Eng, 12(8), pp.393~422 (1939)
  51. S. W. Churchill, "Friction-factor equation spans all fluid-flow regimes", Chem. Eng, 84(24), pp.91~92 (1977)
  52. G.B. Wallis, "One-dimensional Two-phase Flow", ~ (1969)
  53. G. Hewitt, "Annular Two-phase Flow" (2013), Elsevier
  54. S. E. Haaland, "Simple and explicit formulas for the friction factor in turbulent pipe flow", J. Fluid Eng, 105(1), pp.89~90 (1983)
  55. J.G. Collier, "Convective Boiling and Condensation" (1994), Clarendon Press
  56. A Study of Convection Boiling inside Channels, "J. Ferrell", MME, 42~ (1966)
  57. S. Levy, "Forced convection subcooled boilingdprediction of vapor volumetric fraction", Int. J. Heat Mass Tran, 10(7), pp.951~965 (1967)
  58. "BNPP 1&2 PSAR", ~ (2012)
  59. N. Zuber, "Average volumetric concentration in two-phase flow systems", J. Heat Tran, 87(4), pp.453~468 (1965)
  60. C. E. Brennen, "Cavitation and Bubble Dynamics" (2013), Cambridge University Press
  61. M. D. McKay, "Comparison of three methods for selecting values of input variables in the analysis of output from a computer code", Technometrics, 21(2), pp.239~245 (1979)
  62. S.G. Shin, "Analysis of Two-phase Constitutive Relation Models Implemented in Thermal-Hydraulic System Analysis Codes", ~ (2019)
  63. RELAP5 and TRACE Consitutive Relations Comparison(NUREG, "IA-0522)", S. G. Shin, 2020
  64. P. Vassallo, "Slug-to-annular regime transitions in R-134a flowing through a vertical duct", Int. J. Multiphas. Flow, 27(1), pp.119~145 (2001)

관련 논문AI추천

이 간행물의 다른 논문 (20건)