Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

Nuclear Engineering and Technology

Volumn 48, Issue 1, 2016, Pages 1-15

  Kim, Hyun Gil ^a   Yang, Jae Ho ^a   Kim, Weon Ju ^a   Koo, Yang Hyun ^a  

^a Korea Atomic Energy Research Institute   (South Korea)

Author keywords

Accident tolerant fuel; Fukushima accident; Microcell fuel pellet; SiC composite cladding; Surface modified cladding

Indexed keywords

EID: 84958747018     PISSN: 17385733     EISSN: 2234358X     Source Type: Journal    
DOI: 10.1016/j.net.2015.11.011     Document Type: Article

References (51)

1. Koo Y.H., Yang J.H., Park J.Y., Kim K.S., Kim H.G., Kim D.J., Jung Y.I., Song K.W. KAERI's development of LWR accident-tolerant fuel. Nucl. Technol. 2014, 186:295-304.
2. J. Carmack, F. Goldner, S.M. Bragg-Sitton, L.L. Snead, Overview of the U.S. DOE accident tolerant fuel development program, TopFuel 2013, Charlotte, North Carolina, Sep. 15-19, 2013.
3. B. Chang, Y.J. Kim, P. Chou, J. Deshon, Development of Mo-alloy for LWR fuel cladding to enhance fuel tolerance to severe accidents, TopFuel 2013, Charlotte, NC, Sep. 15-19, 2013.
4. I. Idarraga-Trujillo, M. Le Flem, J-C. Brachet, M. Le Saux, D. Hamon, S. Muller, V. Vandenberghe, M. Tupin, E. Papin, E. Monsifrot, A. Billard, F. Schuster, Assessment at CEA of coated nuclear fuel cladding for LWRs with increased margins in LOCA and beyond LOCA conditions, TopFuel 2013, Charlotte, NC, Sep. 15-19, 2013.
5. Kim H.G., Kim I.H., Park J.Y., Koo Y.H. Application of Coating Technology on Zr-based Alloy to Decrease High-temperature Oxidation, Zirconium in the Nuclear Industry STP 1543 2013, 10.1520/STP154320120161.
6. Pint B.A., Terrani K.A., Brady M.P., Cheng T., Keiser J.R. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments. J. Nucl. Mater. 2013, 440:420-427.
7. Terrani K.A., Zinkel S.J., Snead L.L. Advanced oxidation-resistance iron-based alloys for LWR fuel cladding. J. Nucl. Mater. 2013, 448:420-435.
8. Stempien J.D., Carpenter D.M., Kohse G., Kazimi M.S. Characteristics of composite silicon carbide fuel cladding after irradiation under simulated PWR conditions. Nucl. Technol. 2013, 183:13-29.
9. R. Montgomery, E. Mader, N. Domenico, R. Fawcett, J. Guerci, E. Lahoda, B. Minnick, P. Murray, S. Nesbit, M. Meyer, S.M. Bragg-Sitton, Industry-valued design objectives for advanced LWR fuels and concept screening results, TopFuel 2013, Charlotte, NC, Sep. 15-19, 2013.
10. Zinkle S.J., Terrani K.A., Gehin J.C., Ott L.J., Snead L.L. Accident tolerant fuels for LWRs: a perspective. J. Nucl. Mater. 2014, 448:374-379.
11. Bischoff J., Blanpain P., Strumpell J. Development of Fuels with Enhanced Accident Tolerance, IAEA Technical Meeting on Accident Tolerant Fuel Concepts for LWRs Oct. 13-14, 2015, Oak Ridge National Lab, USA.
12. A.M. Savchenko, V.B. Ivanov, V.V. Novikov, M.V. Skupov, G.V. Kulakov, V.K. Orlov, O.I. Uferov, Y.V. Konovalov, Review of A.A. BOCHVAR Institute activities in developing potentially accident tolerant fuel for LWRs, TopFuel 2015, American Nuclear Society, Zurich(Switzerland) Sep. 13-17, 2015.
13. S. Ray, P. Xu, E. Lahoda, L. Hallstadius, F. Boylan, S. Johnson, Westinghouse accident tolerant fuel program-current results & future plans, TopFuel 2015, American Nuclear Society, Zurich(Switzerland) Sep. 13-17, 2015.
14. T. Liu, Update of accident tolerant fuel R&D status in China, 4th Meeting of the Expert Group on Accident Tolerant Fuels for Light Water Reactors, PSI, Villingen(Switzerland) Sep. 17-18. 2015.
15. Kim W.-J., Kim D., Park J.Y. Fabrication and material issues for the application of SiC composites to LWR fuel cladding. Nucl. Eng. Technol. 2013, 45:565-572.
16. S. Ray, S.C. Johnson, E.J. Lahoda, Preliminary assessment of the performance of SiC based accident tolerant fuel in commercial LWR systems, TopFuel 2013, American Nuclear Society, Charlotte, USA, Sep. 15-19, 2013.
17. J.H. Yang, K.S. Kim, D.J. Kim, J.H. Kim, J.S. Oh, Y.W. Rhee, Y.H. Koo, Micro-cell UO2 pellets for enhanced accident tolerant fuel, TopFuel 2013, American Nuclear Society, Charlotte, USA, Sep. 15-19, 2013.
18. 2-Mo pellet with enhanced thermal conductivity. J. Nucl. Mater. 2015, 462:289-295.
19. 2-UN composites with enhanced uranium density and thermal conductivity. J. Nucl. Mater. 2015, 465:509-515.
20. Ott L.J., Robb K.R., Wang D. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions. J. Nucl. Mater. 2014, 448:520-533.
21. Terrani K.A., Wang D., Ott L.J., Montgomery R.O. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents,. J. Nucl. Mater. 2014, 448:512-519.
22. Brillant G., Gupta F., Pasturel A. Fission products stability in uranium dioxide. J. Nucl. Mater. 2011, 412:170-176.
23. D. Jadernas, F. Corleoni, A. Puranen, P. Tejland, M. Granfors, PCI mitigation using fuel additives, TopFuel 2015, American Nuclear Society, Zurich(Switzerland) Sep. 13-17, 2015.
24. H.S. Lee, D.J. Kim, S.W. Kim, J.H. Yang, Y.H. Koo, D.R. Kim, Thermal conductivity of metallic micro-cell fuel pellet with different unit cell geometry, Spring Meeting 2015, Korean Nuclear Society, Jeju(Korea) May 7-8, 2015.
25. J.H. Yang, D.J. Kim, K.S. Kim, Y.H. Koo, Thermo-physical property of micro-cell UO2 pellets and high density composite pellets for accident tolerant fuel, IAEA Technical Meeting on Accident Tolerant Fuel Concepts for LWRs, Oak Ridge National Lab., USA, Oct. 13-14, 2015.
26. D.J. Kim, J.H. Yang, J.H. Kim, K.S. Kim, Y.W. Rhee, J.S. Oh, Y.H. Koo, Metal network containing micro-cell UO2 pellets for accident tolerant fuel, 11th Pacific Rim Conference of Ceramic Societies, Jeju(Korea) Aug. 30-Sept. 4, 2015.
27. Lewis B.J., Thompson W.T., Akbari F., Thompson D., Thurgood C., Higgs J. Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel. J. Nucl. Mater. 2004, 328:180-196.
28. Spino J., Peerani P. Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: review of data and alternative interpretation of recently published results. J. Nucl. Mater. 2008, 375:8-25.
29. D.H. Hwang, S.G. Hong, W.K. In, Evaluation of physical characteristics of PWR cores with accident tolerant fuels, Autumn Meeting 2015, Korean Nuclear Society, Gyeongju(Korea), Oct. 29-30, 2015.
30. Youinou G.J., Sen R.S. Impact of accident-tolerant fuels and claddings on the overall fuel cycle: a preliminary systems analysis. Nucl. Technol. 2014, 188:123-128.
31. Uno M., Nishi T., Takano M. Thermodynamic and thermophysical properties of the actinide nitrides. Comprehensive Nuclear Materials 2012, 61-85. Elsevier.
32. Harp J.M., Lessing P.A., Hoggan R.E. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation. J. Nucl. Mater. 2015, 466:728-738.
33. 2 fuel composites. J. Nucl. Mate.r 2015, 466:745-754.
34. Lessing P.A. INL/EXT-12-24974, Oxidation Protection of Uranium Nitride Fuel Using Liquid Phase Sintering 2012, Idaho National Laboratory.
35. Rama Rao G.A., Mukerjee S.K., Vaidya V.N., Venugopal V., Sood D.D. Oxidation and hydrolysis kinetic studies on UN. J. Nucl. Mater. 1991, 185:231-241.
36. Kim K.H., Lee D.B., Kim C.K., Hofman G.E., Paik K.W. Characterization U-2wt% Mo and U-10 wt% Mo alloy powders prepared by centrifugal atomization. J. Nucl. Mater. 1997, 245:179-184.
37. Kim H.G., Kim I.H., Jung Y.I., Park D.J., Park J.Y., Koo Y.H. Microstructure and mechanical strength of surface ODS treated Zircaloy-4 sheet using laser beam scanning. Nucl. Eng. Technol. 2014, 46(4):521-528.
38. Kim H.G., Kim I.H., Jung Y.I., Park D.J., Park J.Y., Koo Y.H. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating. J. Nucl. Mater. 2015, 465:531-539.
39. Snead L.L., Nozawa T., Katoh Y., Byun T.S., Kondo S., Petti D.A. Handbook on SiC properties for fuel performance modeling. J. Nucl. Mater. 2007, 371:329-377.
40. Yueh K., Carpenter D., Feinroth H. Clad in clay. Nucl. Eng. Int. 2010, 55:14-16.
41. Terrani K.A., Pint B.A., Parish C.M., Silva C.M., Snead L.L., Katoh Y. Silicon carbide oxidation in steam up to 2 MPa. J. Am. Ceram. Soc. 2014, 97:2331-2352.
42. Carpenter D. An Assessment of Silicon Carbides as a Cladding Material for Light Water Reactors 2010, Ph.D. thesis, Massachusetts Institute of Technology.
43. Katoh Y., Snead L.L., Szlufarska I., Weber W.J. Radiation effects in SiC for nuclear structural applications. Curr. Opin. Solid State Mater. Sci. 2012, 16:143-152.
44. Kim D., Lee H.-G., Park J.Y., Kim W.-J. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications. J. Nucl. Mater. 2015, 458:29-36.
45. f/SiC composite tubes. J. Korean Ceram. Soc. 2013, 50:359-363.
46. Davies D.G.S. The statistical approach to engineering design in ceramics,. Proc. Br. Ceram. Soc. 1973, 22:429-452.
47. Kim W.-J., Hwang H.S., Park J.Y., Ryu W.-S. Corrosion behaviors of sintered and chemically vapor deposited silicon carbide ceramics in water at 360°C. J. Mater. Sci. Lett. 2003, 22:581-584.
48. Tan L., Allen T.R., Barringer E. Effect of microstructure on the corrosion of CVD-SiC exposed to supercritical water. J. Nucl. Mater. 2009, 394:95-101.
49. Park J.-Y., Kim I.-H., Jung Y.-I., Kim H.-G., Park D.-J., Kim W.-J. Long-term corrosion behavior of CVD SiC in 360°C water and 400°C steam. J. Nucl. Mater. 2013, 433:603-607.
50. Henager C.H., Schemer-Kohrn A.L., Pitman S.G., Senor D.J., Geelhood K.J., Painter C.L. Pitting corrosion in CVD SiC at 300°C in deoxygenated high-purity water. J. Nucl. Mater. 2008, 378:9-16.
51. Kim D., Lee H.-G., Park J.Y., Park J.-Y., Kim W.-J. Effect of dissolved hydrogen on the corrosion behavior of chemically vapor deposited SiC in a simulated pressurized water reactor environment. Corros. Sci. 2015, 98:304-309.