Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

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

  • Shin, Sung Gil (Department of Nuclear and Quantum Engineering Korea Advanced Institute of Science and Technology (KAIST)) ;
  • Lee, Jeong Ik (Department of Nuclear and Quantum Engineering Korea Advanced Institute of Science and Technology (KAIST))
  • Received : 2021.07.22
  • Accepted : 2022.03.26
  • Published : 2022.09.25
Download PDF
⟨ Previous Next ⟩

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.

Keywords

Acknowledgement

This work was supported by the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KOFONS), granted financial resource from the Nuclear Safety and Security Commission (NSSC), Republic of Korea (No. 1805004-0522-SB120).

References

  1. W. Li, X. Wu, K. Shirvan, G. Su, An investigation of numerical performance enhancement of RELAP5: numerical stability, higher resolution and an alternative constitutive relation, Nucl. Eng. Des. 328 (2018) 309-320. https://doi.org/10.1016/j.nucengdes.2017.12.033
  2. W. Kolar, S. Duffield, R. Nijsing, G. de Vries, Experience with the RELAP code, in: Programming for Software Sharing, Springer, 1983, pp. 265-274.
  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) (1988) 126-127. https://doi.org/10.13182/NT88-A34085
  5. T.W. Ha, J.J. Jeong, K.Y. Choi, Modification of the fast fourier transform-based method by signal mirroring for accuracy quantification of thermal-hydraulic system code, Nucl. Eng. Technol. 49 (5) (2017) 1100-1108. https://doi.org/10.1016/j.net.2017.03.010
  6. Y.-S. Kim, et al., First ATLAS domestic standard problem (DSP-01) for the code assessment, Nucl. Eng. Technol. 43 (1) (2011) 25-44. https://doi.org/10.5516/NET.2011.43.1.025
  7. C. Song, Y.S. Bang, A. Shin, S.-W. Woo, TRACE code calculations of MIT pressurizer tests for review of SPACE code capability, Transactions 111 (1) (2014) 1570-1571.
  8. A. Shin, C. Huh, I. Kim, S.M. Bajorek, Assessment of TRACE 4.050 using UPTF bypass tests, Trans. Am. Nucl. Soc. 93 (2005), 538-538.
  9. Y.S. Bang, J. Lee, A. Shin, D.-Y. Oh, Multidimensional prediction of the effect of flow blockage in SCTF S1-01 experiment using MARS-KS code, Ann. Nucl. Energy 152 (2021) 108028. https://doi.org/10.1016/j.anucene.2020.108028
  10. V. Ransom, R. Wagner, G. Johnsen, RELAP5/MOD2 code manual, in: Developmental Assessment Problems, vol. 3, 1987. EGG-TFM-7952.
  11. U. Nrc, TRACE V5. 0 Assessment ManualeMain Report(Draft), US Nuclear Regulatory Commission, 2005.
  12. M. Kaeri IV, Code Manual Volume IV: Developmental Assessment Report, 2009. KAERI/TR-3042.
  13. S. Ahn, et al., FONESYS: the FOrum & NEtwork of SYStem thermal-hydraulic codes in nuclear reactor thermal-hydraulics, Nucl. Eng. Des. 281 (2015) 103-113. https://doi.org/10.1016/j.nucengdes.2014.12.001
  14. S. Ahn, et al., Hyperbolicity and numerics in SYS-TH codes: the FONESYS point of view, Nucl. Eng. Des. 322 (2017) 227-239. https://doi.org/10.1016/j.nucengdes.2017.06.033
  15. V. Ranson, RELAP5/MOD1: Code Manual Vol. 1: System Models and Numerical Methods, 1981. NUREG/CR-1826, EGG-2070 DRAFT.
  16. U. Nrc, TRACE V5. 0 Theory ManualeField Equations, Solution Methods and Physical Models, United States Nucl. Regul. Comm, 2010.
  17. G. Mesina, A history of RELAP computer codes, Nucl. Sci. Eng. 182 (1) (2016) veix.
  18. G.F. Niederauer, P.T. Giguere, J.F. Lime, T.D. Knight, O. Ashy, R. Fakory, Development of Transient-Reactor Analysis Code (TRAC) for Real-Time Applications, Los Alamos National Lab., NM (United States), 1997.
  19. V.H. Ransom, J. Trapp, R. Wagner, RELAP5/MOD3. 3 code manual Volume IV: models and correlations, in: Information Systems Laboratories, IR, Maryland Idaho Falls, Idaho, 2001.
  20. Y.-G. Lee, S.G. Lim, 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, X. Xiong, J. Shan, J. Gou, B. Zhang, B. Zhang, Improvement and validation of the wall heat transfer package of RELAP5/MOD3. 3, Nucl. Eng. Des. 310 (2016) 418-428. https://doi.org/10.1016/j.nucengdes.2016.10.049
  22. I. Lee, D. Oh, Y. Bang, Y. Kim, Effect of critical flow model in MARS-KS code on uncertainty quantification of large break Loss of coolant accident (LBLOCA), Nucl. Eng. Technol. 52 (4) (2020) 755-763. https://doi.org/10.1016/j.net.2019.09.014
  23. Y.-S. Bang, J.-R. Chun, B.-D. Chung, G.-C. Park, Improvements of condensation heat transfer models in MARS code for laminar flow in presence of noncondensable gas, Nucl. Eng. Technol. 41 (8) (2009) 1015-1024. https://doi.org/10.5516/NET.2009.41.8.1015
  24. J.J. Jeong, D.H. Hwang, B.D. Chung, Improvement of the subchannel flow mixing model of the MARS code, Nucl. Technol. 156 (3) (2006) 360-368. https://doi.org/10.13182/NT06-A3797
  25. A.B. Salah, CATHARE simulation results of the natural circulation characterisation test of the PKL test facility, Nucl. Eng. Technol. (2020).
  26. H.J. Yoon, W. Al Naqbi, O.S. Al-Yahia, D. Jo, Validation of RELAP5 MOD3. 3 code for Hybrid-SIT against SET and IET experimental data, Nucl. Eng. Technol. 52 (9) (2020) 1926-1938. https://doi.org/10.1016/j.net.2020.02.007
  27. Y.-S. Kim, et al., Second ATLAS domestic standard problem (DSP-02) for a code assessment, Nucl. Eng. Technol. 45 (7) (2013) 871-894. https://doi.org/10.5516/NET.02.2013.009
  28. T.S. Choi, H.C. No, Improvement of the reflood model of RELAP5/MOD3. 3 based on the assessments against FLECHT-SEASET tests, Nucl. Eng. Des. 240 (4) (2010) 832-841. https://doi.org/10.1016/j.nucengdes.2009.11.043
  29. T.S. Choi, H.C. No, An improved RELAP5/MOD3. 3 reflood model considering the effect of spacer grids, Nucl. Eng. Des. 250 (2012) 613-625. https://doi.org/10.1016/j.nucengdes.2012.06.025
  30. T.W. Ha, S.W. Bae, J.-P. Park, D. Jo, B.-J. Yun, J.J. Jeong, Development of an empirical correlation for the onset of flow instability in narrow rectangular channels, Nucl. Eng. Des. 375 (2021) 111090. https://doi.org/10.1016/j.nucengdes.2021.111090
  31. S. Bae, J. Jeong, M. Hwang, B. Chung, An implementation and assessment of the viscous stress model in the MARS multi-dimensional component, Intern. Conf. Nucl. Eng. 4689 (2004) 673-679.
  32. P. Zhao, Z. Liu, T. Yu, J. Xie, Z. Chen, C. Shen, Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor, Nucl. Eng. Technol. 52 (12) (2020) 2789-2802. https://doi.org/10.1016/j.net.2020.05.023
  33. M.G. Kim, S.W. Bae, G.M. Choi, J.J. Jeong, Preliminary assessment of the nuclear thermalehydraulic system code MARS for the application to a refrigeration cycle, Nucl. Eng. Des. 367 (2020) 110798. https://doi.org/10.1016/j.nucengdes.2020.110798
  34. S.-J. Kang, H.-Y. Jeong, S.-W. Bae, C.-W. Choi, K.-S. Ha, J.-S. Suh, Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code, Ann. Nucl. Energy 115 (2018) 138-153. https://doi.org/10.1016/j.anucene.2018.01.033
  35. J.H. Park, S.W. Bae, H.S. Park, J.E. Cha, M.H. Kim, Transient analysis and validation with experimental data of supercritical CO2 integral experiment loop by using MARS, Energy 147 (2018) 1030-1043. https://doi.org/10.1016/j.energy.2017.12.092
  36. F. D'Auria, M.J.N.E. Lanfredini, Design, V&V&C in Nuclear Reactor ThermalHydraulics, 2019, p. 110162.
  37. F. D'Auria, M. Lanfredini, Moving from V&V to V&V&C in nuclear thermalhydraulics, in: 2018 26th International Conference on Nuclear Engineering, American Society of Mechanical Engineers Digital Collection, 2018.
  38. G. Bandini, et al., Assessment of systems codes and their coupling with CFD codes in thermalehydraulic applications to innovative reactors, Nucl. Eng. Des. 281 (2015) 22-38. https://doi.org/10.1016/j.nucengdes.2014.11.003
  39. T. Skorek, et al., Quantification of the uncertainty of the physical models in the system thermal-hydraulic codesePREMIUM benchmark, Nucl. Eng. Des. 354 (2019) 110199. https://doi.org/10.1016/j.nucengdes.2019.110199
  40. G.A. Roth, F. Aydogan, Theory and implementation of nuclear safety system codesePart I: conservation equations, flow regimes, numerics and significant assumptions, Prog. Nucl. Energy 76 (2014) 160-182. https://doi.org/10.1016/j.pnucene.2014.05.011
  41. G.A. Roth, F. Aydogan, Theory and implementation of nuclear safety system codesePart II: system code closure relations, validation, and limitations, Prog. Nucl. Energy 76 (2014) 55-72. https://doi.org/10.1016/j.pnucene.2014.05.003
  42. W. Marcum, D. LaBrier, E. Brown, C. Jensen, Y.-J. Choi, Benchmark comparison of experimental data with thermal-hydraulic codes RELAP5-3D and TRACE for a RIA transient scenario, Nucl. Technol. 206 (6) (2020) 911-923. https://doi.org/10.1080/00295450.2020.1713673
  43. Y. Lee, T. Kim, Influence of two-phase crossflow for void prediction in bundles using thermal-hydraulic system codes, Energies 13 (14) (2020) 3686. https://doi.org/10.3390/en13143686
  44. D. Griggs, M. Liebmann, Comparison of TRAC and RELAP5 Reactor System Calculations for a DEGB LOCA in K-14. 1, Westinghouse Savannah River Co., Aiken, SC (United States), 1990.
  45. V.N. Blinkov, et al., 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 (2022) 111666. https://doi.org/10.1016/j.nucengdes.2022.111666
  46. R. Lockhart, R. Martinelli, Proposed correlation of data for isothermal twophase, two-component flow in pipes, Chem. Eng. Prog. 45 (1) (1949) 39-48.
  47. M. Davis, Wall friction for two-phase bubbly flow in rough and smooth tubes, Int. J. Multiphas. Flow 16 (5) (1990) 921-927. https://doi.org/10.1016/0301-9322(90)90013-9
  48. K.T. Claxton, J.G. Collier, J. Ward, HTFS Correlations for Two-phase Pressure Drop and Void Fraction in Tubes, Heat Transfer and Fluid Flow Service, 1972.
  49. D. Zigrang, N. Sylvester, A review of explicit friction factor equations, J. Energy Resour. Technol. 107 (2) (1985) 280-283. https://doi.org/10.1115/1.3231190
  50. C.F. Colebrook, et al., 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) (1939) 393-422. https://doi.org/10.1680/ijoti.1939.14509
  51. S.W. Churchill, Friction-factor equation spans all fluid-flow regimes, Chem. Eng. 84 (24) (1977) 91-92.
  52. G.B. Wallis, One-dimensional Two-phase Flow, 1969.
  53. G. Hewitt, Annular Two-phase Flow, Elsevier, 2013.
  54. S.E. Haaland, Simple and explicit formulas for the friction factor in turbulent pipe flow, J. Fluid Eng. 105 (1) (1983) 89-90. https://doi.org/10.1115/1.3240948
  55. J.G. Collier, J.R. Thome, Convective Boiling and Condensation, Clarendon Press, 1994.
  56. J. Ferrell, A Study of Convection Boiling inside Channels, MME, 1966, p. 42.
  57. S. Levy, Forced convection subcooled boilingdprediction of vapor volumetric fraction, Int. J. Heat Mass Tran. 10 (7) (1967) 951-965. https://doi.org/10.1016/0017-9310(67)90071-3
  58. BNPP 1&2 PSAR, 2012.
  59. N. Zuber, J. Findlay, Average volumetric concentration in two-phase flow systems, J. Heat Tran. 87 (4) (1965) 453-468. https://doi.org/10.1115/1.3689137
  60. C.E. Brennen, Cavitation and Bubble Dynamics, Cambridge University Press, 2013.
  61. M.D. McKay, R.J. Beckman, W.J. Conover, Comparison of three methods for selecting values of input variables in the analysis of output from a computer code, Technometrics 21 (2) (1979) 239-245.
  62. S.G. Shin, Analysis of Two-phase Constitutive Relation Models Implemented in Thermal-Hydraulic System Analysis Codes, 2019.
  63. S.G. Shin, RELAP5 and TRACE Consitutive Relations Comparison (NUREG/IA0522), US Nuclear Regulatory Commission, 2020.
  64. P. Vassallo, T. Trabold, R. Kumar, D. Considine, Slug-to-annular regime transitions in R-134a flowing through a vertical duct, Int. J. Multiphas. Flow 27 (1) (2001) 119-145. https://doi.org/10.1016/S0301-9322(00)00009-4