Fuel performance analysis for PWR cores

Nuclear Engineering and Design

Volumn 239, Issue 8, 2009, Pages 1481-1488

  Romano, Antonino ^a   Shuffler, Carter A ^b   Garkisch, Hans D ^c   Olander, Donald R ^d   Todreas, Neil E ^b  

^a PAUL SCHERRER INSTITUTE   (Switzerland)

^b Massachusetts Institute of Technology Cambridge   (United States)

^c WESTINGHOUSE ELECTRIC COMPANY   (United States)

^d Etcheverry Hall MCI 730   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

BURN UP; BURNUPS; COOLANT MASS FLUX; CORE POWER DENSITY; DISCHARGE BURNUPS; FISSION GAS RELEASE; FUEL DESIGN; FUEL PERFORMANCE; FUEL RODS; FUEL TYPES; HEAT RATE; HYDRIDE FUEL; INTERNAL PRESSURE; LARGE DIAMETER; MECHANICAL CONSTRAINTS; PITCH-TO-DIAMETER RATIOS; SMALL-DIAMETER;

ARCHITECTURAL DESIGN; FISSION PRODUCTS; FUELS; LIQUID METALS; PRESSURIZED WATER REACTORS; SMALL NUCLEAR REACTORS;

LEAKAGE (FLUID);

EID: 67349228492     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2008.11.022     Document Type: Article

References (20)

1. Cox B. Hydrogen uptake during oxidation of zirconium alloys. J. Alloys Compd. 256 (1997) 244-246
2. Forsberg K., and Massih A.R. Fission-gas release under time-varying conditions. J. Nucl. Mater. 135 (1985) 140-148
3. Garkisch Hans D., Petrovic B., 2003. Reference data and constraints for uranium-zirconium-hydride and uranium-thorium-hydride fuels for light water reactors. Westinghouse Internal Report prepared under the NERI project DE-FG03-02SF22615.
4. Huang J., Tsuchiya B., Konashi K., and Yamawaki M. Estimation of hydrogen redistribution in zirconium hydride under temperature gradient. J. Nucl. Sci. Technol. 37 (2000) 887-892
5. Kramman, M.A., Freeburn, H.R. (Eds.), February 1987. ESCORE-the EPRI Steady-State Core Reload Evaluator Code: General Description. Technical Report EPRI NP-5100, Electric Power Research Institute, Palo Alto, California.
6. Lassmann K. TRANSURANUS - A fuel rod analysis code ready for use. J. Nucl. Mater. 188 (1992) 295-302
7. McMahon, M.V., Handwerk, C.S., MacLean, H.J., Driscoll, M.J., Todreas, N.E. Modeling and Design of Reload LWR Cores for an Ultra-Long Operating Cycle. MIT Dept. of Nuclear Engineering MIT-NFC-TR-004 Rev. 1, Cambridge, MA, Sept. 1997, pp. 111-112.
8. NEA, March 2003. Fuel safety criteria in NEA member countries. Technical Report NEA/CSNI/R, p. 10.
9. Olander D.R., Garkisch H., Petrovich B., and Greenspan E. Material aspects of U-Zr hydride fuel. Nucl. Eng. Des. 239 (2009) 1442-1460
10. Olander D.R., and Ng M. Hydride fuel behavior in LWRs. J. Nucl. Mater. 346 (2005) 98-108
11. Rashid, Y.R., Dunham, R.S., Montgomery, R.O., July 2004. FALCON MOD01: fuel analysis and licensing code. New, Technical Report ANA-04-0666, vol. 1, ANATECH Corp.
12. Romano, A., Wallin, H., October 2005. Application of the FALCON and TRANSURANUS fuel performance codes to predict measured fission gas releases from the KKL reactor. PSI Technical Report TM-41-05-18.
13. Romano A., Wallin H., Zimmermann M.A., and Rakesh C. Modelling the CABRI high-burnup RIA test CIP0-1 using an extended version of the FALCON code. Nucl. Eng. Des. 236 (2006) 284-294
14. Shuffler C.A., Trant J.M., Malen J., and Todreas N.E. Thermal hydraulic analysis for grid supported pressurized water reactor cores. Nucl. Eng. Des. 239 (2009) 1406-1424
15. Simnad M.T. The U-ZrHx alloy: its properties and use in TRIGA fuel. Nucl. Eng. Des. 64 (1981) 402-422
16. Srikantiah G.S. VIPRE - A reactor core thermal-hydraulics analysis code for utility applications. Nucl. Technol. 100 (1992) 216-227
17. Todreas N.E., MacDonald P.E., Buongiorno J., and Loewen E.P. Medium-power lead alloy reactors: missions for this reactor technology. Nucl. Technol. 147 (2004) 305-320
18. 2 fuel. Trans. Am. Nucl. Soc. 31 (1979) 163-165
19. Wallin, H., Aounallah, Y., Romano, A., February 2006. Modelling the IFA650.3 Halden LOCA test coupling TRACE and FALCON. PSI Technical Report TM-41-06-01.
20. Wongawaeng D.A., and Olander D.R. Effect of replacing helium with a liquid metal in the fuel-clad gap on fission gas release. Nucl. Technol. 146 (2004) 211-220