Nuclear data sensitivity, uncertainty and target accuracy assessment for future nuclear systems
Tài liệu tham khảo
Aliberti, 2004, Transmutation dedicated systems: An assessment of nuclear data uncertainty impact, Nucl. Sci. Eng., 146, 13, 10.13182/NSE02-94
Bell, 1973
Cecchini, G., Farinelli, U., Gandini, A., Salvatores, M., 1964. Analysis of integral data for few group parameter evaluation of fast reactors. A/CONF 28/P/627, Geneva, Switzerland.
Courcelle, A. et al., 2004. JEF2.2 Nuclear data statistical adjustment using post-irradiation experiments. In: Proceedings of International Conference on PHYSOR-2004, 25–29 April, Chicago, IL, USA.
Forsberg, 2005, Brayton power cycles and high-temperature salt-cooled reactors, Trans. Am. Nucl. Soc., 92, 231
Gandini, 1967, J. Nucl. Energy, 21, 755, 10.1016/0022-3107(67)90086-X
Gandini, 1988
Gandini, 1986, Equivalent generalized perturbation theory (EGPT), Ann. Nucl. Energy, 13, 109, 10.1016/0306-4549(86)90001-0
Gill, P.E., Murray, W., Saunders, M.A., 1997. SNOPT: An SQP algorithm for large-scale constrained programming. Technical Report SOL 97-3, Systems Optimization Laboratory, Department of Operations Research, Stanford University, Stanford, CA 94305-4022.
Gruppelaar, H.,1998. Status of pseudo-fission product cross-sections for fast reactors. Report by the WP on Int. Eval. Co-operation of the NEA Nucl. Sci. Committee and NEA and NEA/WPEC-17.
Kallfelz, 1977, Burn-up calculations with time-dependent generalized perturbation theory, Nucl. Sci. Eng., 62, 304, 10.13182/NSE77-A26966
Kawai, M., 2002. Evaluation method of inelastic scattering cross-section for weakly absorbing fission product nuclides. Report by the WP on Int. Eval. Co-operation of the NEA Nucl. Sc. Committee and NEA and NEA/WPEC-10.
Kim, 2005, Sensitivity study of design parameters for liquid-salt-cooled VHTR, Trans. Am. Nucl. Soc., 93
Kodeli, I., Sartori, E., ZZ-COV-15GROUP-2005, NEA-1730 Package. OECD/NEA Data Bank (in preparation).
Koning, A., 2004. Code TALYS Monte-Carlo and covariances. Annual Mtg of the CSEWG, National Nuclear Data Center, Brookhaven National Laboratory.
Lubitz, C.R., 2004. Epithermal capture cross-sections of U-235. Report by the WP on Int. Eval. Co-operation of the NEA Nucl. Sc. Committee, NEA/WPEC-18.
Marcian, R. et al., 2004. Assessment of CASMO-4 prediction of the isotopic inventory of high burn-up MOX fuel. In: Proceedings of International Conference PHYSOR-2004, 25–29 April, Chicago, IL, USA.
Obložinský, P., 2005. Assessment of neutron cross-sections evaluations for the bulk of fission products. Report by the WP on Int. Eval. Co-operation of the NEA Nucl. Sci. Committee, ISBN-92-64-01063-7 and NEA/WPEC-21.
OECD/NEA Data Bank, 2005. The JEFF-3.0 Nuclear Data Library. JEFF Report 19.
Palmiotti, 1984, Use of integral experiments in the assessment of large liquid-metal fast breeder reactor basic design parameters, Nucl. Sci. Eng., 87, 333, 10.13182/NSE87-333
Palmiotti, G., Salvatores, M., 1988. Multidimensional transport sensitivity for shielding analysis. In: Seventh International Conference on Reactor Shielding, Bournemouth, UK.
Palmiotti, G., Salvatores, M., 2005. Proposal for nuclear data covariance matrix. JEFDOC 1063 Rev. 1.
Palmiotti, 1990, BISTRO optimized two dimensional Sn transport code, Nucl. Sci. Eng., 104, 26, 10.13182/NSE90-1
Palmiotti, 1994, Sensitivity, uncertainty assessment, and target accuracies related to radiotoxicity evaluation, Nucl. Sci. Eng., 117, 239, 10.13182/NSE94-A21501
Palmiotti, G., Aliberti, G., Salvatores, M., Tommasi, J., 2004. Integral experiment analysis for validation and improvement of minor actinide data for transmutation needs. In: Proceedings of International Conference on ND-2004, September, Santa Fe, NM, USA.
Rimpault, G. et al., 2002. The ERANOS code and data system for fast reactor neutronic analyses, In: Proceedings of PHYSOR 2002 Conference, Seoul, Korea.
Santamarina, A., 2005. Private communication. See also JEFDOC 1008, 2004.
Smith, D.L., 2004. Covariance matrices for nuclear cross-sections derived from nuclear model calculations. Report ANL/NDM-159, Argonne National Laboratory.
Smith, D.L., 2005. Neutron reaction data for IFMIF: pointing the way forward. In: IAEA Technical Mtg “Nuclear Data for IFMIF”, 4–6 October, Karlsruhe, Germany.
Tommasi, J., Dupont, E., Marimbeau, P. Analysis of sample irradiation experiments in Phénix for JEFF-3.0 nuclear data validation. Nucl. Sci. Eng. (in press).
Trakas, C., Dandin, L., 2004. Benchmarking of MONTEBURNS against measurements on irradiated UOX and MOX fuels. In: Proceedings of International Conference PHYSOR-2004, 25–29 April, Chicago, IL, USA.
Usachev, L.M. et. al., 1973. IAEA Publication, Conference 730302, vol. 1, p. 129.
US Department of Energy Office of Nuclear Energy, Science, and Technology, 2005. Advanced Fuel Cycle Initiative: Objectives, Approach, and Technology Summary.
USDOE, 2002. Technology Roadmap for Generation IV Nuclear Energy Systems. GIF-002-00.
