User's manual for the ORIGEN2 computer code
TLDR
This report describes how to use a revised version of the ORIGEN computer code, designated ORIGEN2, and a description of the input data, input deck organization, and sample input and output.Abstract:
This report describes how to use a revised version of the ORIGEN computer code, designated ORIGEN2. Included are a description of the input data, input deck organization, and sample input and output. ORIGEN2 can be obtained from the Radiation Shielding Information Center at ORNL.read more
Citations
More filters
Journal ArticleDOI
ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials
TL;DR: ORIGEN2 as discussed by the authors is a point-depletion and radioactive decay computer code for nuclear fuel cycles and calculating the nuclide compositions and characteristics of materials conta...
ReportDOI
ORIGEN2: a revised and updated version of the Oak Ridge isotope generation and depletion code
TL;DR: ORIGEN2 as discussed by the authors is a revised and updated version of the original ORIGEN computer code, which has been designated ORIGEN2 and is a versatile point depletion and decay computer code for use in simulating nuclear fuel cycles.
Journal ArticleDOI
General solution of Bateman equations for nuclear transmutations
TL;DR: In this paper, the linear chain method of solving Bateman equations for nuclear transmutation in derivation of the general solution for linear chain with repeated transitions and thus elimination of existing numerical problems is discussed.
Development of a New Monte Carlo reactor physics code
TL;DR: An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes, used in coupled LWR full-core analyses and typically based on fewgroup nodal diffusion methods.
Journal ArticleDOI
Mccard : monte carlo code for advanced reactor design and analysis
TL;DR: McCARD as discussed by the authors is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design of nuclear reactors and fuel systems and is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis.
Related Papers (5)
ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials
Japanese evaluated nuclear data library version 3 revision-3: JENDL-3.3
Keiichi Shibata,Toshihiko Kawano,Tsuneo Nakagawa,Osamu Iwamoto,Jun Ichi Katakura,Tokio Fukahori,Satoshi Chiba,Akira Hasegawa,Toru Murata,Hiroyuki Matsunobu,Takaaki Ohsawa,Yutaka Nakajima,Tadashi Yoshida,Atsushi Zukeran,Masayoshi Kawai,Mamoru Baba,Makoto Ishikawa,Tetsuo Asami,Takashi Watanabe,Yukinobu Watanabe,Masayuki Igashira,Nobuhiro Yamamuro,Hideo Kitazawa,Naoki Yamano,Hideki Takano +24 more
ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology
Mark B. Chadwick,P. Obložinský,M.W. Herman,N.M. Greene,R.D. McKnight,David Livingstone Smith,Phillip G. Young,R. E. MacFarlane,Gerald M. Hale,S.C. Frankle,A.C. Kahler,Toshihiko Kawano,R. C. Little,D. G. Madland,Peter Möller,Russell D. Mosteller,Philip R. Page,Patrick Talou,Holly R. Trellue,Morgan C. White,William B. Wilson,R. Arcilla,C.L. Dunford,Said F. Mughabghab,Boris Pritychenko,D. Rochman,Alejandro Sonzogni,C.R. Lubitz,T.H. Trumbull,J.P. Weinman,David Brown,D.E. Cullen,David P. Heinrichs,D. P. McNabb,Herve Derrien,Michael E Dunn,Nancy M. Larson,Luiz C Leal,Allan D. Carlson,R.C. Block,J.B. Briggs,E.T. Cheng,H.C. Huria,M.L. Zerkle,K.S. Kozier,A. Courcelle,V.G. Pronyaev,S. C. van der Marck +47 more