Browse > Article

FUEL BEHAVIOR UNDER LOSS-OF-COOLANT ACCIDENT SITUATIONS  

CHUNG HEE M. (Consultant Temecula, California, USA)
Publication Information
Nuclear Engineering and Technology / v.37, no.4, 2005 , pp. 327-362 More about this Journal
Abstract
The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.
Keywords
Fuel Behavior; Loss-of-coolant Accident; Cladding Embrittlement; Peak Temperature Limit; Advanced Cladding Alloys; High Burnup;
Citations & Related Records
연도 인용수 순위
  • Reference
1 H. M. Chung, J.-H. Park, J. E. Sanecki, N. J. Zaluzec, T. T. Yang, and M. S. Yu, 'Cracking Mechanism of Type 304L Stainless Steel Core Shroud Welds,' Proc. 9th Intnl. Conf. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Eds. S. Bruemmer, P. Ford, and G. Was, The Minerals, Metals, and Materials Society, Warrendale, PA, 1999, pp. 973-984
2 R. F. Welton, private communication, November 2003; also see: 'Target Material Characterization Apparatus and Measurement Techniques,' presentation slides available in website of Holifield Radioactive Ion Beam Facility, Beam Development - Target Material Characterization Laboratory, Physics Division, Oak Ridge National Laboratory, 2003
3 The Handbook of Binary Phase Diagrams, W. G. Moffatt, Genium Publishing Corporation, Schenectady, NY, 1984
4 C. T. Ward, D. L. Mathis, and R. W. Staehle, Corrosion 25 (1969) 394   DOI
5 W. G. Dobson and R. R. Biederman, 'ZORO-1: A Finite Difference Computer Model for Zircaloy Oxidation in Steam,' EPRI-NP-347, Electric Power Research Institute, December 1976
6 K. Shimizu, K. Kobayashi, G. E. Thompson, P. Skeldon, and G. C. Wood, J. Electrochem. Soc. 144 (1997) 418   DOI
7 S. Malang, 'SIMTRAN-1: A Computer Code for Simultaneous Calculation of Oxygen Distribution and Temperature Profiles in Zircaloy during Exposure to High Temperature Oxidising Environment,' ORNL-5083, Oak Ridge National Laboratory, November 1975
8 L. Portier, T. Bredel, J. C. Brachet, V. Maillot, J. P. Mardon, and A. Lesbos, 'Influence of Long Service Exposures on the Thermal-Mechanical Behavior of Zry-4 and M5TM Alloys in LOCA Conditions,' presented at the 14th International Symposium on Zirconium in the Nuclear Industry, Stockholm, Sweden, June 13-17, 2004, to be published in meeting proceedings
9 M. C. Billone, 'LOCA Embrittlement Correlation,' Argonne National Laboratory Interim Report, April 8, 2005
10 P. E. C. Bryant and P. R. Habicht, Combustion Engineering Internal Report TIS-5065, also in Proc. IAEA Workshop on Stress Corrosion Cracking, March 29-31,1976
11 N. P. Sajin and E. A. Pepelyaeva, 'Separation of Hafnium from Zirconium and Production of Pure Zirconium Dioxide,' Proc. Intl. Conf. on Peaceful Uses of Atomic Energy, Vol. 8, Presentation P/634, pp.559, August 8-20, 1955, Geneva, Switzerland
12 Y. Yan, T. Burtseva, and M. Billone, 'LOCA Results for Advanced-Alloy and High-Burnup Zircaloy Cladding,' Proc. Nuclear Safety Research Conference, October 20-22, 2003, Washington, D.C.
13 J.-C. Brachet, J. Pelchat, D. Harmon, R. Maury, P. Jacques, and J.-P. Mardon, 'Mechanical Behavior at Room Temperature and Metallurgical Study of Low-Tin Zry-4 and M5 Alloys after Oxidation at $1100^{\circ}C$ and Quenching,' Proc. IAEA Technical Committee Mtg. on Fuel Behavior under Transient and LOCA Conditions, September 10-14, 2001, Halden, Norway
14 V. Vrtilkova, L. Novotny, R. Doucha, and J. Vesely, 'An Approach to the Alternative LOCA Embrittlement Criterion,' Proc. of Meeting of OECD Nuclear Energy Agency Committee on the Safety of Nuclear Installations Special Expert Group on Fuel Safety Margins (SEGFSM), May 25-26, 2004, Argonne, Illinois, USA
15 M. Takemoto, T. Shonohara, M. Shirai, and T. Shinogaya, Mater. Performance 24 (1985) 26
16 H. M. Chung, W. E. Ruther, J.-H. Park, J. E. Sanecki, and N. J. Zaluzec, Paper #443, Corrosion '99, San Antonio, Texas, April 25-30, 1999
17 M. Takahashi, H. Miyazaki, and Y. Katoh, 'New Solvent Extraction Process for Zirconium and Hafnium,' Zirconium in the Nuclear Industry: Sixth International Conference, ASTM STP 824, D. G. Franklin and R. B. Adams, Eds., American Society for Testing and Materials, 1984, pp. 45-56
18 P. Pint and S. N. Flengas, 'Production of Zirconium Metal by Fused-Salt Electrolysis,' Transactions, Institution of Mining and Metallurgy, London, Section C, Vol. 87, March 1978, pp. C29-C49
19 K. Kiukkola and C. Wagner, J. Electrochem. Soc., 104 (1957) 379   DOI
20 W. J. Leach, 'Ductility Testing of Zircaloy-4 and Zirlo Cladding after High-Temperature Oxidation in Steam,' Proc. OECD Topical Mtg. on LOCA Fuel Safety Criteria, March 22-23, 2001, Aix-en-Provence, France
21 V. Vrtilkova, M. Valach, and L. Molin, 'Oxidation and Hydriding Properties of Zr-lNb Cladding Material in Comparison with Zircaloys,' Proc. IAEA Technical Committee Mtg. on Influence of Water Chemistry on Fuel Clad Behavior, October 4-8, 1993, Rez, Czech Republic; also in IAEA TECDOC-927, Vienna, 1997, pp. 227-251
22 V. Asmolov, L. Yegorova, K. Lioutov, A. Konoveyev, V. Smirnov, A. Goryachev, V. Chesanov, and V. Prokhorov, 'Understanding LOCA-Related Ductility in E110 Cladding,' Proc. Nuclear Safety Research Conference, October 28-30, 2002, Washington, D.C.
23 L. Yegorova and K. Lioutov, 'LOCA Behavior of E110 Alloy,' Proc. Nuclear Safety Research Conference, October 20-22, 2003, Washington, D.C.
24 R. H. Meservey and R. Henzel, 'Embrittlement Behavior of Zircaloy in an Emergency Core Cooling Environment,' IN-1389, Idaho Nuclear Corporation, September, 1970
25 L. Maroti, 'Ring-Compression Test Results and Experiments Supporting LOCA PCT, Oxidation and Channel Blockage Criteria,' Proc. OECD Topical Mtg. on LOCA Fuel Safety Criteria, March 22-23, 2001, Aix-en-Provence, France
26 Yu. K. Bibilashvili, N. B. Sokolov, L. N. Andreeva-Andrievskaya, V. Yu. Tonkov, A. V. Salatov, A. M. Morosov, and V. P. Smirnov, 'Thermomechanical Properties of Oxidized Zirconium-Based Alloy Claddings in Loss of Coolant Accident Conditions,' Proc. OECD Topical Mtg. on LOCA Fuel Safety Criteria, March 22-23, 2001, Aix-en-Provence, France
27 R. A. Lorenz, 'Fuel Rod Failure under Loss-of-Coolant Conditions in TREAT,' Nucl. Tech. 11 (1971) 502-520   DOI
28 F. M. Haggag, 'Zircaloy Cladding Embrittlement Criteria: Comparison of In-Pile and Out-of-Pile. Results,' NUREG/CR-2757, July 1982
29 G. J. Scatena, 'Fuel Cladding Embrittlement during a Loss of Coolant Accident,' NEDO-10674, General Electric Company, October 1972
30 H. Uetsuka et al., 'Failure-Bearing Capability of Oxidized Zircaloy-4 Cladding under Simulated Loss-of-Coolant Condition,' J. Nucl. Sci. Tech. 20 (1983) 941-950   DOI
31 M. Reocreux and E. Scott de Martinville, 'A Study of Fuel Behavior in PWR Design Basis Accident: An Analysis of Results from the PHEBUS and EDGAR Experiments,' Nucl. Eng. Design 124 (1990) 363-378   DOI   ScienceOn
32 M. Suzuki and S. Kawasaki, 'Development of Computer Code PRECIP-II for Calculation of Zr-Steam Reaction,' J. Nucl. Sci. Tech. 17 (1980) 291   DOI
33 C. Grandjean et al., 'High Burnup UO2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation after Burst,' ANL Program Review Meeting, Rockville (USA), October 22, 1999
34 P. Hofmann et al., 'PECLOX: A Computer Model for the Calculation of the Internal and the External Zircaloy Cladding Oxidation,' KFK-4422 Part 2, October 1988
35 D. O. Hobson, 'Ductile-brittle behavior of Zircaloy fuel cladding,' Proc. ANS Topical Mtg. on Water Reactor Safety, Salt Lake City, March 26, 1973, pp. 274-288
36 General Design Criteria for Nuclear Power Plants, U.S. Code of Federal Regulations, Title 10, Part 50, Appendix A, February 20, 1971, amended
37 E. Kolstad, W. Wiesenack, and V. Grismanovs, 'LOCA Testing at Halden,' Proc. Nuclear Safety Research Conference, October 25-27, 2004, Washington, DC
38 L. Yegorova, K. Lioutov, N. Jouravkova, A. Konobeev, V. Smirnov, V. Chesanov, and A. Goryachev, 'Experimental Study of Embrittlement of Zr-1%Nb VVER Cladding under LOCA-Relevant Conditions,' NUREG/IA-0211, IRSN-194, NSI RRC KI 3188, U.S. Nuclear Regulatory Commission, March 2005
39 Report of Advisory Task Force on Power Reactor Emergency Cooling, TID-24226, 1967
40 H. M. Chung, 'The Effects of Aliovalent Elements on Nodular Oxidation of Zr-Base Alloys,' Proc. Nuclear Safety Research Conference, October 21-24, 2003, Washington, DC.
41 L. Portier, T. Bredel, J.C. Brachet, V. Maillot, J.P. Mardon, and A. Lesbos, 'Influence of Long-Service Exposures on the Thermal-Mechanical Behavior of Zry-4 and M5 Alloys in LOCA Conditions,' Proc. 14th Intl. Symp. On Zirconium in the Nuclear Industry, June 13-17, 2004, Stockholm, Sweden
42 F. Nagase and T. Fuketa, 'Results from Studies on High Burnup Fuel Behavior under LOCA Conditions,' Proc. Nuclear Safety Research Conference, October 25-27, 2004, Washington, DC
43 Y. Yan, T. A. Burtseva, and M. C. Billone, 'LOCA Integral Test Results for High-Burnup BWR Fuel,' Proc. Nuclear Safety Research Conference, October 25-27, 2004, Washington, DC
44 M. C. Billone, Y. Yan, and T. A. Burtseva, 'Post-Quench Ductility of Advance Alloy Cladding,' Proc. Nuclear Safety Research Conference, October 25-27, 2004, Washington, DC
45 J. Boehmert, 'Embrittlement of Zr-Nbl at Room Temperature after High-Temperature Oxidation in Steam Atmosphere,' Kerntechnik 57 (1992) 55-58
46 W. W. Stephens, 'Extractive Metallurgy of Zirconioum - 1945 to the Present,' Zirconium in the Nuclear Industry: Sixth International Conference, ASTM STP 824, D. G. Franklin and R. B. Adams, Eds., American Society for Testing and Materials, 1984, pp. 5-36
47 L. Moulin, P. Thouvenin, and P. Brun, 'New Process for Zirconium and Hafnium Separation,' Zirconium in the Nuclear Industry: Sixth International Conference, ASTM STP 824, D. G. Franklin and R. B. Adams, Eds., American Society for Testing and Materials, 1984, pp. 37-44
48 C. Wagner, Naturwissenschaften, 31 (1943) 265   DOI
49 M. Billone, 'Overview of Advanced Alloy Post-Quench Ductility Program,' Review Mtg. of LOCA and Dry-Cask-Storage Programs, July 16-17, 2003, Argonne, Illinois
50 A. Griger, L. Maroti, L. Matus, and P. Windberg, 'Ambient and High-Temperature Mechanical Properties of ZrNbl Cladding with Different Oxygen and Hydrogen Content,' Proc. Enlarged Halden Program Group Meeting, May 24-29, 1999, Loen, Norway
51 J. Boehmert, M. Dietrich, and J. Linek, 'Comparative Studies on High-Temperature Corrosion of ZrNbl and Zircaloy-4,' Nucl. Engr. Design 147 (1993) 53-62   DOI   ScienceOn
52 A. Sawatzky, 'Proposed Criterion for the Oxygen Embrittlement of Zircaloy-4 Fuel Cladding,' Proc. 4th Symp. on Zirconium in the Nuclear Industry, Stratford-on-Avon, UK, June 27-29, 1978
53 H. Uetsuka et al., 'Zircaloy-4 Cladding Embrittlement due to Inner Surface Oxidation under Simulated Loss-of-Coolant Condition,' J. Nucl. Sci. Tech. 18 (1981) 705-717   DOI
54 H. Uetsuka, et al., 'Embrittlement of Zircaloy-4 due to Oxidation in Environment of Stagnant Steam,' J. Nucl. Sci. Tech. 19 (1982), 158-165   DOI
55 K. Komatsu, 'The Effects of Oxidation Temperature and Slow Cooldown on Ductile-Brittle Behavior of Zircaloy Fuel Cladding,' Proc. CSNI Specialists' Meeting on the Behavior of Water Reactor Fuel Elements under Accident Conditions, Spaatind, Norway, September 13-16, 1976
56 K. Komatsu et al., 'Load-Bearing Capability in Deformed and Oxidized Zircaloy Cladding,' Proc. CSNI Specialist Mtg. on Safety Aspects of Fuel Behavior in Off-Normal and Accident Conditions, Espoo, Finland, September 1-4, 1980
57 J. D. Duncan and J. E. Leonard, Thermal Response and Cladding Performance of Zircaloy Clad Simulated Fuel Bundles Under High Temperature Loss of Coolant Conditions, GEAP-13174, General Electric Company (May 1971)
58 L. Baker and L. C. Just, 'Studies of Metal-Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium-Water Reaction,' ANL-6548, Argonne National Laboratory, May 1962
59 R. E. Pawel, 'Oxygen Diffusion in Beta Zircaloy during Steam Oxidation,' J. Nucl. Mater. 50 (1974) 247-258   DOI   ScienceOn
60 J. V. Cathcart, R. E. Pawel, R. A. McKee, R. E. Durscel, G. J. Yurek, J. J. Campbell, and S. H. Jury, 'Zirconium Metal-Water Oxidation Kinetics IV, Reaction Rate Studies, ORNL/NUREG-17, August 1977
61 H. M. Chung and T. F. Kassner, 'Pseudobinary Zircaloy-Oxygen Phase Diagram,' J. Nucl. Mater. 84 (1979) 327-339   DOI   ScienceOn
62 H. M. Chung and T. F. Kassner, 'Embrittlement Criteria for Zircaloy Fuel Cladding Applicable to Accident Situations in Light-Water Reactors,' NUREG/CR-1344, ANL-79-48, Argonne National Laboratory, January 1980
63 Parsons, P.D. et al., 'The Deformation, Oxidation and Embrittlement of PWR Fuel Cladding in a Loss-of-Coolant Accident: A State-of-the-Art Report,' CSNI Report 129, December 1986
64 Zuzek, E. et al., 'The H-Zr (Hydrogen-Zirconium) System,' Bulletin of Alloy Phase Diagrams, 11 (1990) 385-395   DOI
65 Atomic Energy Commission Rule-Making Hearing, Supplemental Testimony of the Regulatory Staff Docket RM-50-1, October 26, 1972
66 Y. Yan, T. Burtseva, and M. C. Billone, 'LOCA Results for Advanced-Alloy and High-Burnup Zircaloy Cladding,' Proc. Nuclear Safety Research Conference, October 21-24, 2003, Washington, DC.
67 U. S. Code of Federal Regulations, Title 10, Energy, Parts 0 to 50, Revised January 1, 1997, U.S. Government Printing Office, Washington, DC, 1997
68 C. Grandjean et al., 'Oxidation and Quenching Experiments with High Burnup Cladding under LOCA Conditions,' Proc. 26th Water Reactor Safety Information Meeting, Bethesda, USA, October 26-28, 1998
69 F. Nagase et al., 'Experiments on High Burnup Fuel Behavior under LOCA Conditions at JAERI,' Proc. ANS Topical Meeting on LWR fuel Performance, Park City, USA, April 10-13, 2000
70 Atomic Energy Commission Rule-Making Hearing, Concluding Statement of the Regulatory Staff, Docket RM-50-1, April 16, 1973
71 D. O. Hobson and P. L. Rittenhouse, 'Embrittlement of Zircaloy Clad Fuel Rods by Steam During LOCA Transients,' ORNL-4758, Oak Ridge National Laboratory, January 1972
72 Interim Acceptance Criteria for Emergency Core-Cooling Systems for Light-Water Power Reactors, U.S. Federal Register 36 (125), pp. 12247-12250, June 29, 1971
73 W.B. Cottrell, 'ECCS Rule-Making Hearing,' Nucl. Safety 15 (1974) 30-55
74 New Acceptance Criteria for Emergency Core-Cooling Systems of Light-Water-Cooled Nuclear Power Reactors, Nucl. Safety 15 (1974) 173-184
75 Atomic Energy Commission Rule-Making Hearing, Opinion of the Commission, Docket RM-50-1, December 28, 1973
76 J. C. Hesson et al., 'Laboratory Simulations of Cladding-Steam Reactions Following Loss-of-Coolant Accidents in Water-Cooled Power Reactors,' ANL-7609, Argonne National Laboratory, January 1970
77 G. W. Parker et al., 'Release of Fission Products from Reactor Fuels during Transient Accidents Simulated in TREAT,' Proc. Intl. Symp. Fission Product Release and Transport under Accident Conditions, Oak Ridge, TN, April 5-7,1965
78 T. Fujishiro et al., 'Light Water Reactor Fuel Response during Reactivity Initiated Accident Experiments,' NUREG/CR-0269, August 1978
79 J.C. Brachet, L. Portier, V. Maillot, T. Forgeron, J.P. Mardon, P. Jacques, A. Lesbros, 'Overview of the CEA Data on the Influence of Hydrogen on the Metallurgical and Thermalmechanical Behavior of Zircaloy-4 and M5 Alloys under LOCA Conditions,' Proc. Nuclear Safety Research Conference, October 25-27, 2004, Washington, DC