Modeling of the P2M Past Fuel Melting Experiments with the Fuel Performance Code CYRANO3

References

• R. LARGENTON and G. THOUVENIN, “CYRANO3: The EDF Fuel Performance Code—Global Overview and Recent Developments on Fission Gas Modelling,” Proc. WRFPM 2014, Paper No. 100032, Sendai, Japan, September 14–17, 2014.
   Google Scholar
• D. BARON et al., “CYRANO3: The EDF Fuel Performance Code Especially Designed for Engineering Applications,” Proc. WRFPM, Paper No. 8032, Seoul, Korea, October 19–23, 2008.
   Google Scholar
• A. CHAIEB et al., “Computational Assessment of LOCA Simulation Tests on High Burnup Fuel Rods in Halden and Studsvik Using Cyrano3 Code,” Proc. Top Fuel, Santander, Spain (Oct. 2021).
   Google Scholar
• V. D’AMBROSI et al., “P2M Simulation Exercise on Past Fuel Melting Irradiation Experiments,” Nucl. Technol., 210, 2, 189 (2023); https://doi.org/10.1080/00295450.2023.2194270.
   Web of Science ®Google Scholar
• A. BOULORE et al., “Approach to Better Assess Fission Gas Behaviors, Applicable to Fuels with Complex Microstructures,” presented at the Water Reactor Fuel Performance Mtg., Jeju Island, South Korea, September 2017.
   Google Scholar
• G. JOMARD et al., “CARACAS: An Industrial Model for the Description of Fission Gas Behavior in LWR-UO2 Fuel,” Proc. WRFPM 2014, Paper No. 100154, September 14–17, 2014.
   Google Scholar
• A. MAGNI et al., “Modelling and Assessment of Thermal Conductivity and Melting Behaviour of MOX Fuel for Fast Reactor Applications,” J. Nucl. Mater., 541, 152410 (Dec. 1, 2020); https://doi.org/10.1016/j.jnucmat.2020.152410.
   Web of Science ®Google Scholar
• J. J. CARBAJO et al., “A Review of the Thermophysical Properties of MOX and UO2 Fuels,” J. Nucl. Mater., 299, 181 (Dec. 12, 2001); https://doi.org/10.1016/S0022-3115(01)00692-4.
   Web of Science ®Google Scholar
• D. R. OLANDER, “Fundamental Aspects of Nuclear Reactor Fuel Elements,” TID-26711-P1–TRN: 08-019392, Department of Nuclear Engineering, University of California, Berkeley (Jan. 1, 1976).
   Google Scholar
• J. REST et al., “Fission Gas Release from UO2 Nuclear Fuel: A Review,” J. Nucl. Mater., 513, 310 (Jan. 1, 2019); https://doi.org/10.1016/j.jnucmat.2018.08.019.
   Web of Science ®Google Scholar
• L. C. BERNARD, J. L. JACOUD, and P. VESCO, “An Efficient Model for the Analysis of Fission Gas Release,” J. Nucl. Mater., 302, 125 (2002); https://doi.org/10.1016/S0022-3115(02)00793-6.
   Web of Science ®Google Scholar
• A. H. BOOTH, “A Method of Calculating Fission Gas Diffusion from UO2 Fuel and Its Application to the X-2-f Loop Test,” AECL-496, Atomic Energy of Canada Limited (Sep. 1957).
   Google Scholar
• J. A. TURNBULL et al., “The Diffusion Coefficients of Gaseous and Volatile Species During the Irradiation of Uranium Dioxide,” J. Nucl. Mater., 107, 168 (June 1, 1982); https://doi.org/10.1016/0022-3115(82)90419-6.
   Web of Science ®Google Scholar
• B. BOER and M. VERWERFT, “Qualification of the New Pressurized Water Capsule (PWC) for Fuel Testing at BR2,” Proc. RRFM 2021, Helsinki, Finland, September 26–30, 2021.
   Google Scholar
• B. BOER et al., “Transient Tests in BR2 with the Pressurized Water Capsule—Performance and Uncertainty,” Proc. Top Fuel 2021, Santander, Spain, October 2021.
   Google Scholar
• M. GROUNES et al., “Fuel R&D at Studsvik I. Introduction and Experimental Facilities,” Nucl. Eng. Des., 168, 1–3, 129 (1997); https://doi.org/10.1016/S0029-5493(96)01367-2.
   Web of Science ®Google Scholar
• V. I. ARIMESCU et al., “Third SCIP Modeling Workshop: Beneficial Impact of Slow Power Ramp on PCI Performance,” Proc. WRFPM 2014, Paper No. 100045, Sendai, Japan, September 14–17, 2014.
   Google Scholar
• I. VALLEJO, L. E. HERRANZ, and I. ARIMESCU, “Compilation of Preliminary Contributions to the 3rd SCIP Benchmark,” presented at the SCIP II PRG, Studsvik, June 12, 2012.
   Google Scholar
• B. SUNDMAN et al., “OpenCalphad—A Free Thermodynamic Software,” Integr. Mater. Manuf. Innov., 4, 1 (Dec. 1, 2015); https://doi.org/10.1186/s40192-014-0029-1.
   Web of Science ®Google Scholar
• C. GUÉNEAU et al., “TAF-ID: An International Thermodynamic Database for Nuclear Fuels Applications,” Calphad, 72, 102212 (Mar. 2021); https://doi.org/10.1016/j.calphad.2020.102212.
   Google Scholar