Nuclear Engineering and Technology

  • 제51권6호
  • /
  • Pages.1596-1609
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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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Validation of the fuel rod performance analysis code FRIPAC

  • Deng, Yong-Jun (China Nuclear Power Technology Research Institute Co. Ltd) ;
  • Wei, Jun (China Nuclear Power Technology Research Institute Co. Ltd) ;
  • Wang, Yang (China Nuclear Power Technology Research Institute Co. Ltd) ;
  • Zhang, Bin (China Nuclear Power Technology Research Institute Co. Ltd)
  • 투고 : 2019.02.14
  • 심사 : 2019.05.02
  • 발행 : 2019.09.25
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초록

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

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참고문헌

  1. D.G. Cacuci, Handbook of Nuclear Engineering [M], Springer, New York, 2010.
  2. K.J. Geelhood, W.G. Luscher, FRAPCON-4.0: Integral Assessment, PNNL, U.S. Department of Energy, 2015. PNNL-19418.
  3. J.L. Jacoud, Ph Vesco, Despription and Qualification of the COPERNIC/TRANSURANS (Update of May 2000) Fuel Rod Design Code, Framatome Nuclear Fuel, 2000, p. TFJCDC1556.
  4. K. Lassmann, H. Blank, Modeling of fuel rod behavior and recent advances of the TRANSURANUS code, Nucl. Eng. Des. 106 (1988) 291-313. https://doi.org/10.1016/0029-5493(88)90292-0
  5. P.V. Uffelen, A. Schubert, J. van de Laar, et al., Verification of the TRANSURANUS fuel performance code - an overview, in: 7th International Conference on WWER Fuel Performance, Albena, Bulgaria, 2007.
  6. F.W. Dittus, L.M.K. Boelter, Heat transfer in automobile radiators of the tubular type, Int. Commun. Heat Mass Transf. 12 (1985) 3-22. https://doi.org/10.1016/0735-1933(85)90003-X
  7. W.H. Jens, P.A. Lottes, Analysis of Heat Transfer, Burnout, Pressure Drop and Density Data for High-Pressure Water, Argonne National Laboratory, 1951. ANL-4627.
  8. L.J. Siefken, E.W. Coryell, E.A. Harvego, J.K. Hohorst, SCDAP/RELAP5/MOD 3.1 Code Manual: MATPRO, A Library of Materials Properties for Light-Water- Reactor Accident Analysis, vol. 4, Idaho National Engineering and Environmental Laboratory, 2001. INEL-96/0422, NUREG/CR-6150.
  9. F. Kreith, R.M. Manglik, M. Bohn, Principles of Heat Transfer, seventh ed., Cengage Learning, 2011.
  10. C.E. Beyer, C.R. Hann, D.D. Lanning, F.E. Panisko, L.J. Parchen, GAPCONTHERMAL- 2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod, Battelle Northwest Laboratory, 1975. BNWL-1898.
  11. G. Jacobs, N. Todreas, Thermal contact conductance in reactor fuel elements, Nucl. Sci. Eng. 50 (1973) 283-290. https://doi.org/10.13182/NSE73-A28981
  12. D.D. Lanning, C.E. Beyer, K.J. Geelhood, FRAPCON-3 Updates, Including Mixed- Oxide Fuel Properties, vol. 4, Pacific Northwest National Laboratory, 2005. PNNL-11513, NUREG/CR-6534.
  13. P.G. Lucuta, H.S. Matzke, I.J. Hastings, A pragmatic approach to modeling thermal conductivity of irradiated UO2 fuel: review and recommendations, J. Nucl. Mater. 232 (1996) 166-180. https://doi.org/10.1016/S0022-3115(96)00404-7
  14. U.S. Nuclear Regulatory Commission, An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification, Office of Standards Development, 1978, p. 126. Regulatory Guide 1.
  15. 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, Pacific Northwest National Laboratory, 2015. PNNL-19418, Vol.1, Rev.2.
  16. W.G. Luscher, K.J. Geelhood, Material Property Correlations: Comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO, Pacific Northwest National Laboratory, 2011. PNNL-19417, NUREG/CR-7024.
  17. P.V. Uffelen, Contribution to the Modelling of Fission Gas Release in Light Water Reactor Fuel, PhD Thesis, Nuclear Engineering, University of Liege, 2002.
  18. A.H. Booth, A Method of Calculating Fission Gas Diffusion from UO2 Fuel and its Application to the X-2 Loop Test, Atomic Energy of Canada, Ltd., Chalk River Project, 1957. CRDC-721.
  19. K. Forsberg, A.R. Massih, Fission gas release under time-varying conditions, J. Nucl. Mater. 127 (1985) 141-145. https://doi.org/10.1016/0022-3115(85)90348-4
  20. K. Forsberg, A.R. Massih, Diffusion theory of fission gas migration in irradiated nuclear fuel UO2, J. Nucl. Mater. 135 (1985) 140-148. https://doi.org/10.1016/0022-3115(85)90071-6
  21. K. Lassmann, H. Benk, Numerical algorithms for intragranular fission gas release, J. Nucl. Mater. 280 (2000) 127-135. https://doi.org/10.1016/S0022-3115(00)00044-1
  22. R. Delorme, Ch Valot, L. Fayaette, X. Pujol, I. Aubrun, J. Lamontagne, T. Blay, B. Pasquet, P. Bienvenu, I. Roure, C. Pozo, G. Carlot, C. Sabathier, P. Martin, G. Trillon, V. Auret, S. Bouffard, Study of fission gas behaviour and fuel restructuration in irradited (U,Gd)$O_2$ fuel, in: Transactions of the TopFuel 2012 Reactor Fuel Performance Conference, Manchester, UK, September 2-6, 2012.
  23. I. Arana, C. Munoz-Reja, F. Culbebras, Post-irradiation examination of high burnup fuel rods from Vandellos II, in: Transactions of the TopFuel 2012 Reactor Fuel Performance Conference, Manchester, UK, September 2-6, 2012.
  24. D. Peng, D.B. Robinson, A new two-constant equation of state, Ind. Eng. Chem. Fundam. 15 (1975) 59-64. https://doi.org/10.1021/i160057a011
  25. C. Nealley, D.D. Lanning, M.E. Cunningham, C.R. Hann, Post-irradiation Data Analysis for NRC/PNL Halden Assembly IFA-431, Pacific Northwest Laboratory, 1979. NCREG/CR-0797, PNL-2975.
  26. D.D. Lanning, Irradiation History and Final Post-irradiation Data for IFA-432, Pacific Northwest Laboratory, 1986. NCREG/CR-4717, PNL-5971.
  27. J.A. Turnbull, R.J. White, The Thermal Performance of the Gas Flow Rigs: A Review of Experiments and Their Analyses, OECD Halden Reactor Project, 2002. HWR-715.
  28. I. Matsson, J.A. Turnbull, The Integral Fuel Rod Behaviour Test IFA-597.3: Analysis of the Measurements, OECD Halden Reactor Project, 1998. HWR-543.
  29. P.M. Chantoin, E. Sartori, J.A. Turnbull, The compilation of a public domain database on nuclear fuel performance for the purpose of code development and validation, in: Proceedings of the ANS/ENS International Topical Meeting on Light Water Reactor Fuel Performance, Portland, Oregon, March 2-6, 1997.
  30. C. Bagger, H. Carlson, P. Knudson, Details of Design Irradiation and Fission Gas Release for the Danish $UO_2-Zr$ Irradiation Test 022, Riso National Laboratory, 1978. RISO-M-2152.
  31. T. Tverberg, M. Amaya, Study of Thermal Behavior of $UO_2$ and (U,Gd)$O_2$ to High Burnup (IFA-515), OECD Halden Reactor Project, 2001. HWR-671.
  32. H. Ruhamann, W. Beere, B.H. Lee, Irradiation Performance of Commercial UO2 and (U,Gd)O2 Fuel; Update of the Test IFA-636.1, OECD Halden Reactor Project, 2001. HWR-672.
  33. H. Koike, The MOX Fuel Behavior Test IFA-597.4/.5/.6/.7; Summary of In-Pile Fuel Thermal Temperature and Gas Release Data, OECD Halden Reactor Project, 2003. HWR-729.
  34. L. Mertens, M. Lippens, J. Alvis, The FIGARO Programme: the Behaviour of Irradiated MOX Fuel Tested in the IFA-606 Experiment, Description of Results and Comparison with COMETHE Calculation, 1998. HPR 349/30 Halden Reactor Project.
  35. R.J. White, The Re-irradiation of MIMAS MOX Fuel in IFA-629.1, OECD Halden Reactor Project, 1999. HWR-586.
  36. B. Petiprez, Ramp Tests with Two High Burnup MOX Fuel Rods in IFA-629.3, OECD Halden Reactor Project, 2002. HWR-714.
  37. P. Blair, J. Wright, The IMF/MOX Comparative Test, IFA-651.1: Result after Four Cycles of Irradiation, OECD Halden Reactor Project, 2004. HWR-763.
  38. K.R. Merckx, L.F. Van Swam, G.L. Ritter, The Third Risoe Fission Gas Project (RISOE-III) Database, Risoe National Laboratory, 1995. NEA-1493 IFPE/RISOEIII.
  39. S. Djurle, The Super-ramp Project, Final Report of the Super-ramp Project, Studsvik AB Atomenergi, 1984. STIR-32.
  40. W.F. Lyon, US-PWR 16x16 LTA Extended Burnup Demonstration Program Database, EPRI, 2005. NEA-1738 IFPE/US-PWR-16x16LTA.
  41. M. Boulanger, M. Lippens, Gadolinia Doped $UO_2$ Fuel Behaviour Experiment Database, Belgonucleaire, 2002. NEA-1625 IFPE/GAIN.
  42. M.G. Balfour, BR-3 High Burnup Fuel Rod Hot Cell Program, Final Report, Westinghouse Electric Corporation, 1982. WCAP 10238, DOE/ET/34073-1.
  43. L.W. Newman, The Hot Cell Examination of Oconee-1 Fuel Rods after Five Cycles of Irradiation, Babcock and Wilcox Company, 1986. DOE/ET/34212-50, BAW-1874.
  44. D.D. Lanning, M.E. Cunningham, J.O. Barner, E.R. Bradley, Qualification of Fission Gas Release Data from Task 2 Rods, HBEP 25, Final Report, Pacific Northwest Laboratory, 1987.
  45. T. Ozawa, Performance of ATR MOX fuel assemblies irradiated to 40 GWd/tU, in: Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22, 2004.
  46. D.A. Wesley, K. Mori, S. Inoue, Mark BEB ramp testing program, in: Proceedings of the 1994 ANS/ENS International Topical Meeting on Light Water Reactor Fuel Performance, American Nuclear Society, West Palm Beach, Florida, 1994.
  47. S. Beguin, The Lift-Off Experiment with MOX Fuel Rod in IFA-610.2 Initial Results, OECD Halden Reactor Project, 1999. HWR-603.
  48. E. de Meulemeester, N. Hoppe, G. de Contenson, M. Watteau, Review of work carried out by BELGONUCLEAIRE and CEA on the improvement and verification of the COMETHE computer code with the aid of in-pile experimental results, in: BNES International Conference on Nuclear Fuel Performance, Paper ECS-EEC-73-595, London, UK, 1973.
  49. R.J. White, S.B. Fisher, P.M.A. Cook, R. Stratton, C.T. Walker, I.D. Palmer, Measurement and analysis of fission gas release from BNFL's SBR MOX fuel, J. Nucl. Mater. 288 (2001) 43-56. https://doi.org/10.1016/S0022-3115(00)00591-2
  50. P. Cook, E. Matthews, M. Barker, R. Foster, A. Donaldson, C. Ott, D. Papaioannou, C. Walker, Post-irradiation examination and testing of BNFL SBR MOX fuel, In: Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22, 2004.
  51. IAEA, Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III): Report of a Coordinated Research Project 2008-2012, International Atomic Energy Agency, 2013. Technical Report IAEA-TECDOC-1697.