Modeling of the P2M Past Fuel Melting Experiments with the FEMAXI-8 Code

References

• M. N. F. SUZUKI, H. SAITOU, and Y. UDAGAWA, “Input/Output Manual of Light Water Reactor Fuel Analysis Code FEMAXI-7 and Its Related Code,” JAEA-Data/Code 2013-009October, Japan Atomic Energy Agency (2013).
   Google Scholar
• Y. UDAGAWA et al., “Development of Fuel Performance Code FEMAXI-8; Model Improvements for Light Water Reactor Fuel Analysis and Systematic Validation,” JAEA-Data/Code 2018-016January, Japan Atomic Energy Agency (2018); https://doi.org/10.11484/jaea-data-code-2018-016.
   Google Scholar
• Y. UDAGAWA and M. AMAYA, “Model Updates and Performance Evaluations on Fuel Performance Code FEMAXI-8 for Light Water Reactor Fuel Analysis,” J. Nucl. Sci. Technol., 56, 6, 461 (2019); https://doi.org/10.1080/00223131.2019.1595766.
   Web of Science ®Google Scholar
• Y. UDAGAWA et al., “The Effect of Base Irradiation on Failure Behaviors of UO2 and Chromia-Alumina Additive Fuels Under Simulated Reactivity-Initiated Accidents: A Comparative Analysis with FEMAXI-8,” Ann. Nucl. Energy, 139, 107268 (2020); https://doi.org/10.1016/j.anucene.2019.107268.
   Web of Science ®Google Scholar
• Y. UDAGAWA and Y. TASAKI, “Improvement of Intragranular Fission Gas Behavior Model for Fuel Performance Code FEMAXI-8,” JAEA-Data/Code 2021-007July, Japan Atomic Energy Agency (2021); https://doi.org/10.11484/jaea-data-code-2021-007 (in Japanese).
   Google Scholar
• Y. TANIGUCHI, Y. UDAGAWA, and M. AMAYA, “Analytical Study of SPERT-CDC Test 859 Using Fuel Performance Codes FEMAXI-8 and RANNS,” Ann. Nucl. Energy, 139, 107188 (2020); https://doi.org/10.1016/j.anucene.2019.107188.
   Web of Science ®Google Scholar
• V. D’AMBROSI et al., “Presentation of the xM3 Test Case of the P2M Simulation Exercise and Modeling with the Fuel Performance Code ALCYONE,” Nucl. Technol., 210, 2, 285 (2023); https://doi.org/10.1080/00295450.2023.2253660.
   Google Scholar
• G. BONNY et al., “Presentation of the HBC4 Test Case of the P2M Simulation Exercise and Modeling with the Fuel Performance Code TRANSURANUS,” Nucl. Technol., 210, 2, 216 (2023); https://doi.org/10.1080/00295450.2023.2264505.
   Google Scholar
• M. G. ADAMSON, E. A. AITKEN, and R. W. CAPUTI, “Experimental and Thermodynamic Evaluation of the Melting Behavior of Irradiated Oxide Fuels,” J. Nucl. Mater., 130, 349 (1985); https://doi.org/10.1016/0022-3115(85)90323-X.
   Web of Science ®Google Scholar
• P. E. MACDONALD and L. B. THOMPSON, “MATPRO: Version 09. A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior,” TREE NUREG-1005, Idaho National Engineer4ing Laboratory (1876).
   Google Scholar
• J. R. MATTHEWS and M. H. WOOD, “A Simple Operational Gas Release and Swelling Model. II. Grain Boundary Gas,” J. Nucl. Mater, 91, 2–3, 241 (1980); https://doi.org/10.1016/0022-3115(80)90224-X.
   Web of Science ®Google Scholar
• R. J. WHITE, “The Development of Grain-Face Porosity in Irradiated Oxide Fuel,” J. Nucl. Mater., 325, 1, 61 (2004); https://doi.org/10.1016/j.jnucmat.2003.10.008.
   Web of Science ®Google Scholar
• A. B. J. A. CHRISTENSEN and R. J. ALLIO, “Melting Point of Irradiated Uranium Dioxide,” Trans. Am. Nucl. Soc., 7, 2, 390 (1964).
   Google Scholar
• G. R. D. HAGRMAN, “MATPRO. Version 11: A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, Idaho National Laboratory (1979).
   Google Scholar
• K. OHIRA, “Thermal Conductivity Measurements of High Burnup UO2 Pellet and a Benchmark Calculation of Fuel Center Temperature,” Proc. 1997 Int. Topl. Mtg. Light Water React Fuel Performance (1997).
   Google Scholar
• F. W. DITTUS and L. M. K. BOELTER, “Heat Transfer in Automobile Radiators of the Tubular Type,” Univ. Calif. Pubs. Eng., 2, 443 (1930).
   Google Scholar
• J. CHEN, “A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow,” Process Des. Dev., 5, 322 (1966); https://doi.org/10.1021/i260019a023.
   Web of Science ®Google Scholar