Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

  • Received : 2017.11.20
  • Accepted : 2018.05.26
  • Published : 2018.08.25
Download PDF
⟨ Previous Next ⟩

Abstract

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

Keywords

References

  1. R. Tregoning, L. Abramson, P. Scott, A. Csontos, Estimating Loss-of-coolant Accident (LOCA) Frequencies through the Elicitation Process, USNRC Report NUREG-1829, U.S. Nuclear Regulatory Commission, Washington, DC, 2008.
  2. The ROSA-V Group, ROSA-v Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies, JAERI-tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.
  3. Y. Kukita, K. Tasaka, H. Asaka, T. Yonomoto, H. Nakamura, The effects of break locaion on PWR small break LOCA: experimental study at the ROSA-IV LSTF, Nucl. Eng. Des. 122 (1990) 255-262. https://doi.org/10.1016/0029-5493(90)90210-O
  4. USNRC Nuclear Safety Analysis Division, RELAP5/MOD3.3 Code Manual, NUREG/CR-5535/Rev 1, Information Systems Laboratories, Inc., 2001.
  5. S. Abe, A. Satou, T. Takeda, H. Nakamura, RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project, J. Nucl. Sci. Technol. 51 (2014) 1164-1176. https://doi.org/10.1080/00223131.2014.911710
  6. T. Takeda, Y. Maruyama, T. Watanabe, H. Nakamura, RELAP5 analyses of OECD/NEA ROSA-2 Project experiments on intermediate-break LOCAs at hot leg or cold leg, J. Power Energy Syst. 6 (2012) 87-98. https://doi.org/10.1299/jpes.6.87
  7. G.E. Wilson, B.E. Boyack, The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis, Nucl. Eng. Des. 186 (1998) 23-37. https://doi.org/10.1016/S0029-5493(98)00216-7
  8. J. Freixa, T.-W. Kim, A. Manera, Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation, Nucl. Eng. Des. 264 (2013) 153-160. https://doi.org/10.1016/j.nucengdes.2013.02.023
  9. F. Mascari, H. Nakamura, K. Umminger, F. De Rosa, F. D'Auria, Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool, in: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), Chicago, IL, American Nuclear Society (ANS), IL, USA, 2015.
  10. Nuclear Energy Agency, Scaling in System Thermal-Hydraulics Applications to Nuclear Reactor Safety: a State-of-the-art Report, NEA/CSNI/R, vol. 14, 2016, p. 2017.
  11. K. Umminger, L. Dennhardt, S. Schollenberger, B. Schoen, Integral test facility PKL: experimental PWR accident investigation, Sci. Technol. Nucl. Installations 2012 (2012) 1-16. Article ID 891056.
  12. T. Takeda, A. Ohwada, H. Nakamura, Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA Project and RELAP5 post-test analysis, Exp. Therm. Fluid Sci. 51 (2013) 112-121. https://doi.org/10.1016/j.expthermflusci.2013.07.007
  13. T. Takeda, RELAP5 analyses of ROSA/LSTF experiments on AM measures during PWR vessel bottom small-break LOCAs with gas inflow, Int. J. Nucl. Energy 2014 (2014) 1-17. Article ID 803470.
  14. N. Zuber, Problems in Modeling Small Break LOCA, USNRC Report NUREG- 0724, U.S. Nuclear Regulatory Commission, Washington, DC, 1980.
  15. H. Kumamaru, K. Tasaka, Recalculation of Simulated Post-scram Core Power Decay Curve for Use in ROSA-IV/LSTF Experiments on PWR Small-break LOCAs and Transients, JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990.
  16. V.H. Ransom, J.A. Trapp, The RELAP5 choked flow model and application to a large scale flow test, in: Proceedings of the ANS/ASME/NRC International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Saratoga Springs, New York, USA, American Nuclear Society (ANS), IL, USA, 1980.
  17. G.B. Wallis, One-dimensional Two-phase Flow, McGraw-Hill Book, New York, USA, 1969.
  18. B.E. Boyack, L.W. Ward, Validation test matrix for the consolidated TRAC (TRAC-M) code, in: Proceedings of International Meeting on Best Estimate Methods in Nuclear Installation Safety Analysis (BE '00), Washington, DC, Los Alamos National Laboratory, NM, USA, 2000.
  19. M.J. Griffiths, J.P. Schlegel, T. Hibiki, M. Ishii, I. Kinoshita, Y. Yoshida, Phenomena identification and ranking table for thermal-hydraulic phenomena during a small-break LOCA with loss of high pressure injection, Prog. Nucl. Energy 73 (2014) 51-63. https://doi.org/10.1016/j.pnucene.2014.01.008
  20. M. Ishii, K. Mishima, Study of Two-fluid Model and Interfacial Area, NUREG/ CR-1873, Argonne National Laboratory, Lemont, IL, 1980.
  21. H. Kumamaru, Y. Kukita, H. Asaka, M. Wang, E. Ohtani, RELAP5/MOD3 code analyses of LSTF experiments on intentional primary-side depressurization following SBLOCAs with totally failed HPI, Nucl. Technol. 126 (1999) 331-339. https://doi.org/10.13182/NT99-A2978
  22. F.W. Dittus, L.M.K. Boelter, Heat transfer in automobile radiators of the tubular type, Int. Comm. Heat Mass. Transfer 12 (1985) 3-22. https://doi.org/10.1016/0735-1933(85)90003-X
  23. J.R. Sellars, M. Tribus, J.S. Klein, Heat transfer to laminar flows in a round tube or flat conduit: the Graetz problem extended, Trans. ASME 78 (1956) 441-448.
  24. S.W. Churchill, H.H.S. Chu, Correlating equations for laminar and turbulent free convection from a vertical plate, Int. J. Heat Mass. Transfer 18 (1975) 1323-1329. https://doi.org/10.1016/0017-9310(75)90243-4
  25. A. Guba, M. Makai, L. Pal, Statistical aspects of best estimate methodeI, Reliability Eng. Syst. Saf. 80 (2003) 217-232. https://doi.org/10.1016/S0951-8320(03)00022-X
  26. A. de Crecy, P. Bazin, H. Glaeser, et al., Uncertainty and sensitivity analysis of the LOFT L2-5 test: results of the BEMUSE programme, Nucl. Eng. Des. 238 (2008) 3561-3578. https://doi.org/10.1016/j.nucengdes.2008.06.004
  27. W.W. Daniel, Spearman rank correlation coefficient, in: Applied Nonparametric Statistics, second ed., PWS-Kent Publishing, Boston, MA, 1990.
  28. Nuclear Energy Agency, NEA Annual Report 2016, 2017, pp. 1-68. NEA No. 7349.

Cited by

  1. Analysis of nodalization uncertainty for nuclear system analysis code with Lax-Wendroff numerical scheme vol.167, pp.None, 2018, https://doi.org/10.1016/j.anucene.2021.108853