Publications Area
The NJOY Nuclear Data Processing System, Version 2012
Los Alamos National Laboratory report LA-UR-12-27079 (Dec 2012). Original Author: R. E. MacFarlane, Contributing Authors: D. W. Muir, R. M. Boicourt, A. C. Kahler, Current Editor: A. C. Kahler.
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data
Los Alamos National Laboratory report LA-UR-11-05121 (2011). This paper by M. B. Chadwick, M. Herman, P. Oblozinsky, M. E. Dunn, Y. Danon, A. C. Kahler, D. L. Smith, B. Pritychenko, G. Arbanas, R. Arcilla, R. Brewer, D. A. Brown, R. Capote, A. D. Carlson, Y. S. Cho, H. Derrien, K. Guber, G. M. Hale, S. Hoblit, S. Holloway, T. D. Johnson, T. Kawano, B. C. Kiedrowski, H. Kim, S. Kunieda, N. M. Larson, L. Leal, J. P. Lestone, R. C. Little, E. A. McCutchan, R. E. MacFarlane, M. Macinnes, C. M. Matton, R. D. McKnight, S. F. Mughabghab, G. P. A. Nobre, G. Palmiotti, A. Palumbo, M. T. Pigni, V. G. Pronyaev, R. O. Sayer, A. A. Sonzogni, N. C. Summers, P. Talou, I. J. Thompson, A. Trkov, R. L. Vogt, S. C. van der Marck, A. Wallner, M. C. White, D. Wiarda, and P. G. Young summarizes the new release of ENDF/B-VII evaluated nuclear data. Fifteen different institutions are represented in this work. It was published in Nuclear Data Sheets 12, December 2011, 2987-2996 and is provided here for scholarly purposes with the permission of Elsevier. Not for commercial use. The offical published paper is available at http://dx.doi.org/10.1016/j.nds.2011.11.002.
ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmark and Reactor Experiments
Los Alamos National Laboratory report LA-UR-11-11271(2011). This paper by A. C. Kahler, R. E. MacFarlane, R. D. Mosteller, B. C. Kiedrowski, S. C. Frankle, M. B. Chadwick, R. D. McKnight, R. M. Lell, G. Palmiotti, H. Hiruta, M. Herman, R. Arcilla, S. F. Mughabghab, J. C. Sublet, A. Trkov, T. H. Trumbell, and M. Dunn summarizes the testing of ENDF/B-VII Release 1 against a variety of benchmark experiments. It was published in Nuclear Data Sheets 12, December 2011, 2997-3036 and is provided here for scholarly purposes with the permission of Elsevier. Not for commercial use. The official published paper is available on line at http://dx.doi.org/10.1016/j.nds.2011.11.003.
Fission Product Yields for 14 MeV Neutrons on 235U, 238U, and 239Pu
Los Alamos National Laboratory report LA-UR-11-03445(2011). This report by M. Macinnes, M. B. Chadwick, and T. Kawano describes measurements of cumulative fission product yields for three important isotopes measured at Los Alamos. It was published in Nuclear Data Sheets 12, December 2011, 3135-3152 and is provided here for scholarly purposes with the permission of Elsevier. Not for commercial use. The official published paper is available at http://dx.doi.org/10.1016/j.nds.2011.11.009.
Quantification of Uncertainties for Evaluated Neutron-Induced Reactions on Actinides in the Fast Energy Range
Los Alamos National Laboratory report LA-UR-11-11629(2011). This paper by P. Talou, P. G. Young, T. Kawano, M. Rising, and M. B. Chadwick provides a discussion of the methods used to generate covariance data in the fast energy range for actinide isotopes. It is the basis for the covariance data included in the recent ENDF/B-VII.1 library of evaluated nuclear data. It was published in Nuclear Data Sheets 12, December 2011, 3054-3074 and is provided here for scholarly purposes with the permission of Elsevier. Not for commercial use. The official published paper is available on line at http://dx.doi.org/10.1016/j.nds.2011.11.005.
Data Testing for ENDF/B-VII.1beta2
A new version of the ENDF/B-VII library is undergoing beta testing. This paper shows some early testing results for beta2. LA-UR-11-10211 (March 2011).
Methods for Processing ENDF/B-VII with NJOY
This paper by R. E. MacFarlane and A. C. Kahler, Los Alamos National Laboratory report LA-UR-10-04652 (July 2010), provides a review of the methods using in the NJOY Nuclear Data Processing System with examples drawn from the processing of the ENDF/B-VII files of evaluated nuclear data. Published in Nuclear Data Sheets 111 2739-2890, and provided here for scholarly purposes. Not for commercial use. The published version of this paper is available at http://www.sciencedirect.com/science/issue/6972-2010-998889987-267536.
Unresolved Resonance Range Cross Section Probability and Self Shielding Factors
This paper by Jean Christophe Sublet, Roger N. Blomquist, Sedat Goluoglu, and Robert E. MacFarlane, CEA-R-6227 (July 2009) compares results and methods for a number of different codes and labs when computing average cross sections and self-shielding factors in the unresolved resonance range.
A Code Comparison Study for the Bigten Critical Assembly
This study compares results from several labs for a simplified model of the Bigten critical assembly with special emphasis on the unresolved resonance range for U-238. Agreement is generally good, but several interesting differences and data effects are observed. LA-UR-08-4668 (July 16, 2008).
Criticality Calculations Using LANL and LLNL Neutron Transport Codes
Dermott E. Cullen, Peter Brown, Edward Lent, Robert MacFarlane and Scott McKinly, Lawrence Livermore National Laboratory report UCRL-TR-237333 (November 22, 2007). Several LANL and LLNL neutron transport codes were used to calculate criticality for three systems: a simple U-235 sphere (Godiva), a simple Pu-239 sphere (Jezebel), and a simple U-233 sphere (Jezebel23). ENDF/B-VII.0 cross sections were used in all cases, and both multigroup SN and Monte Carlo methods were used. Good agreement between the codes was seen, and the agreement with experiment was also good.
Total Prompt Energy Release in the Neutron-Induced Fission of U-235, U-238, and Pu-239
This study from 2006 addresses, for the first time, the total prompt energy release and its components for the fission of U-235, U-238, and Pu-239 as a function of the kinetic energy of the neutron inducing the fission. The components are extracted from experimental measurements, where they exist, together with model-dependent calculation, interpolation, and extrapolation. While the components display clear dependencies upon the incident neutron energy, their sums display only weak, yet definite, energy dependencies. Also addressed is the total prompt energy deposition in fission for the same thr ee systems. Results are presented in equation form. New measurements are recommended as a consequence of this study.
Testing New Actinide Cross Sections Proposed for ENDF/B-VII
This paper from the International Conference on Nuclear Data for Science and Technology, Santa Fe, September 2004, presents predictions of the improvements in calculating a range of criticality benchmark experiments expected from the new ENDF/B-VII library of evaluated nuclear cross sections.
How Accurately Can We Calculate Neutrons Slowing Down in Water?
This paper by Dermot E. Cullen, Roger N. Blomquist, Maurice Greene, Edward Lent, Robert MacFarlane, Scott McKinley, Ernest F. Plechaty, and Jean Christophe Sublet, published as Lawrence Livermore National Laboratory report UCRL-TR-220605 (April 1, 2006), discusses results from a number of codes and laboratory for the apparantly simple problem of neutrons slowing down in water. Interesting differences are discussed.
Prompt Fission Neutron Spectrum Matrices from the Los Alamos Model
The prompt fission neutron spectrum matrices for the n + 235-U system
(u5matrx03.dat,
December 1999), the n + 238-U system
(u8matrx02.dat,
December 2002), and the n + 239-Pu system
(pu9matrx01.dat, September
2001) were later transformed into ENDF format and were then installed
into ENDF/B-VII. A single exception exists, however, for the thermal
spectrum in the n + 235-U system. Here, the thermal spectrum from
ENDF/B-VI (an earlier calculation using the
Los Alamos model) has been
re-installed in ENDF/B-VII. This is due to the fact that the two most
recent differential measurements of the thermal spectrum and the most
accepted set of integral cross section measurements in the thermal
spectrum constitute three mutually incompatible experimental results.
Whereas the ENDF/B-VI thermal spectrum reproduces the accepted set of
integral cross section measurements reasonably well, which are very
important to the applied (reactor) community, it is too hard a spectrum
to reproduce well the two differential spectrum measurements. The
thermal spectrum in the file u5matrx03.dat is a compromise spectrum for
the three measurements and is described in the attached reference.
Reference:
"Fission Neutron Spectra of Uranium-235,"
D. G. Madland, OECD
NEA NSC International Evaluation Cooperation, Vol. 9, ISBN-92-64-02134-5,
NEA/WPEC-9, OECD, Paris, FR, 2003.
Fission Product Yields
This section contains the paper T. R. England and B. F. Rider, "Evaluation and Compilation of Fission Product Yields," Los Alamos National Laboratory report LA-UR-94-3106 (Oct. 1994). This is a very large piece of work, and we break it up into parts here. The General Discussion and Bibliography are given as a Postscript file (1.75 MB). Click here for a PDF version. The recommended yields, original data, and reference sources are presented in a series of appendices in the form of ordinary computer text files. In these yield-set names, S stands for spontaneous fission, T for thermal energies, F for fission spectrum energies, and H or HE for high energy (14 MeV). For more information on the yield sets, see the yld.txt file.
- Appendix A (3.24 MB): U-235F, U-235HE, U-238F, U-238HE, Pu-239T,Pu-239F, Pu-241T, U-233T, and Th-232F.
- Appendix B (1.49 MB): U-233F, U-233HE, U-236F, Pu-239H, Pu-240F, Pu-241F, Th-232H, Np-237F, and Csf-252S.
- Appendix C (1.35 MB): U-234F, U-237F, Pu-240H, U-234HE, U-236HE, Pu-238F, Am-241F, Am-243F, Np-238F, Cm-244F.
- Appendix D (1.18 MB): Th-227T, Th-229T, Pa-231F, Am-241T, Am-241H, Am-242MT, Cm-245T, Cf-249T, Cf-251T, Es-254T.
- Appendix E (1.06 MB): Cf-250S, Cm-244S, Cm-248S, Es-253S, Fm-254S, Fm-255T, Fm-256S, Np-237H, U-232T, U-238S.
- Appendix F (0.93 MB): Cm-243T, Cm-246S, Cm-243F, Cm-244F, Cm-246F, Cm-248F, Pu-242H, Np-237T, Pu-240T, and Pu-242T.
New Thermal Evaluations
This report "New Thermal Neutron Scattering Files for ENDF/B-VI Release 2," Los Alamos National Laboratory report LA-12639-MS (ENDF 356) (March 1994) gives details on the LEAPR module for NJOY, and it describes how LEAPR was used to prepare thermal scattering law files in ENDF-6 format. Actually these new evaluations were included in Release 3 of ENDF/B-VI, not Release 2.
TRANSX
TRANSX is a code for preparing data tables for multigroupnuclear transport codes from nuclear data in MATXS format: R. E. MacFarlane, "TRANSX 2: A Code for Interfacing MATXS Cross-Section Libraries to Nuclear Transport Codes," Los Alamos National Laboratory report LA-12312-MS (July 1992). Information on the code will be found in the Codes section of the T-2 Nuclear Information Service. A hypertext verion of the TRANSX report is also available.
