EPJ Web Conf.
Volume 146, 2017ND 2016: International Conference on Nuclear Data for Science and Technology
|Number of page(s)||4|
|Section||Nuclear Data for Applications|
|Published online||13 September 2017|
- GFR 2400 MWth pin core at start of GoFastR (GoFastR, 2009)
- Š. Čerba et al., Development of multigroup neutron cross section library for fast reactor calculations (Proceedings of the 22nd international conference on applied physichs of condensed matter, pp. 29–36, ISBN: 978-961-6702-39-3 (2016)
- M.B Chadwick et al., ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data, Nuclear Data Sheets, 112(2), 2887–2996 (2011) [CrossRef]
- R.E. MacFarlane et al., NJOY99: Data Processing System of Evaluated Nuclear Data Files ENDF Format (Los Alamos National Laboratory, 2000)
- LANL, MCNP - A General N - Particle Transport Code (LANL, 2003)
- R.E. MacFarlane, TRANSX-CTR: A Code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis (LANL, 1984)
- ORNL, PARTISN: Multi-Dimensional, Time-Independent or Time-Dependent, Multigroup, Discrete Ordinates Transport Code System (RSIC, 2009)
- ORNL, DIF3D: Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems (RSIC, 2011)
- D.H. Kim, C.S. Gil, Y.O. Lee, ZZ KAFAX-E70, 150 and 12 Groups Cross Section Library in MATXS Format based on ENDF/B-VII.0 for Fast Reactors (KAERI, Daejeon, 2008)
- OECD NEA, International Handbook of Evaluated Criticality Safety Benchmark Experiments (OECD, Paris, 2007)
- OECD NEA, Methods and Issues for the Combined Use of Integral Experiments and Covariance Data (OECD, Paris, 2013)
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