References
-
Y.H.Koo, B.H. Lee, D.S. Sohn, "COSMOS: a Computer Code to Analyze LWR
$UO_{2}$ and MOX Fuel up to High Burnup", Annals of Nuclear Energy, 26, 47 (1999). https://doi.org/10.1016/S0306-4549(98)00033-4 - H. Zimmermann, "Investigations on Swelling and Fission Gas Behavior in Uranium Oxide", J. Nucl. Mater. 75 (1978) 154. https://doi.org/10.1016/0022-3115(78)90039-9
- T. Kogai, K. Ito and Y. Iwano, "The Effect of Cladding Restraint on Fission Gas Release Behaviors", J. Nucl. Mater. 158, 64 (1988). https://doi.org/10.1016/0022-3115(88)90155-9
-
Y.H. Koo, et al "Analysis of Fission gas release and gaseous Swelling in
$UO_{2}$ Fuel under the Effect of External Restraint", J. Nucl. Mater. 280 86 (2000). https://doi.org/10.1016/S0022-3115(00)00034-9 - K. Une and S. Kashibe, J. Nucl Sci. Tech. 27, 1002 (1990). https://doi.org/10.1080/18811248.1990.9731285
- Y.H. Koo, B.H. Lee, J.S. Cheon, D.S. Sohn, "Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel", Annals of Nuclear Energy, 29, 271 (2002). https://doi.org/10.1016/S0306-4549(01)00035-4
- W. Wiesenack, T. Tverberg, "Thermal Performance of High Burnup Fue˜l In-pile Temperature Data and Analysis", 2000 International Topical Meeting on LWR Fuel Performance, Park City, Utah, April, 2000.
- H. Kampf and G. Karsten, "Effect of Different Types of Void Volumes on the Radial Temperature Distribution of Fuel Pins", Nucl. Appl. Tech., 9, 288 (1970). https://doi.org/10.13182/NT70-A28783
-
Y. Philipponneau, "Thermal Conductivity of (U,Pu)
$O_{2-x}$ Mixed Oxide Fuel", J. Nucl. Maters., 188, 194 (1992). https://doi.org/10.1016/0022-3115(92)90470-6 - G. Gates, et al., "Thermal Performance Modeling with the ENIGMA Code", Proc. of Seminar on Thermal Performance of High Burnup LWR Fuel, France, p.301 (1998).
- L. Heins, H. Landskron, "Fuel Rod Design by Statistical Methods for MOX Fuel", Int. Symposium on MOX Fuel Cycle Technologies for Medium and Long-term Deployment, IAEA-SM-358/21, Austria (1999).
-
M. Lippens, et al., "Comparative Thermal Behavior of MOX Fuel and
$O_{2-x}$ Fuels", Proc. of the Seminar on Thermal Performance of High Burnup LWR Fuel, France, p.243 (1998). - B.H. Lee, Y.H. Koo, J.Y. Oh, D.S. Sohn, "Zircaloy-4 cladding corrosion model covering a wide range of PWR experiences", 378, 127 (2008). https://doi.org/10.1016/j.jnucmat.2008.04.019
- B.H. Lee, Y.H. Koo, Cheon, D.S. Sohn, "Modeling of creep behavior of Zircaloy-4 by considering metallurgical effect", Annals of Nuclear Energy 29, 1 (2002). https://doi.org/10.1016/S0306-4549(01)00030-5
- Y. Kosaka, "Thermal Conductivity Degradation Analysis of the Ultra High Burnup Experiment (IFA-562)", HWR- 341, Halden Reactor Project.
- L.J. Ott, T. Tverberg, E. Sartori, Annals of Nuclear Energy, 36, 375 (2009). https://doi.org/10.1016/j.anucene.2008.12.019
- L.J. Ott, NEA/NSC/DOC (2009)
- M. Lipppens, FIGARO International Programme, 2000
Cited by
- Implementation of effective-stress-function algorithm for nuclear fuel performance code vol.14, pp.5, 2013, https://doi.org/10.1007/s12541-013-0103-1
- Abnormal Prediction of Dense Crowd Videos by a Purpose–Driven Lattice Boltzmann Model vol.27, pp.1, 2017, https://doi.org/10.1515/amcs-2017-0013