| 315 |
ENDF-379 |
Dermott E. Cullen |
A Survey of ENDF/B-VIII Resonance Parameters (MF=2) |
Nov. 2020 |
INDC(NDS)-0819 |
| 314 |
ENDF-378 |
CSEWG Covariance Committee |
Guidance on Generating Neutron Reaction Data Covariances for the ENDF/B Library |
2011 |
|
| 313 |
ENDF-377 |
Donald L. Smith |
Quality Assurance Requirements for ENDF/B-VII.1 Covariances |
2011 |
|
| 312 |
ENDF-376 |
Russell D. Mosteller |
An Expanded Criticality Validation Suite for MCNP |
2010 |
LA-UR-10-06230 |
| 311 |
ENDF-375 |
V. Sobes, L. Leal, G. Arbanas |
User's Guide To SAMINT: A Code For Nuclear Data Adjustment With SAMMY Based On Integral Experiments |
Aug 2014 |
ORNL/TM-2014/245 |
| 310 |
ENDF-374 |
D.E. Cullen |
How Accurate Are Our Processed ENDF Cross Sections? |
May 2014 |
INDC(NDS)-0666 |
| 309 |
ENDF-373 |
D.E. Cullen |
PROGRAM ENDF2C: Convert ENDF Data to Standard FORTRAN, X and C++ Format (Version 2014-1)
(The code is available here)
|
Apr 2014 |
IAEA-NDS-217 |
| 308 |
ENDF-372 |
D.E. Cullen |
ENDF/X: an Extended ENDF Format (Evolution, not Revolution) |
Dec 2012 |
INDC(NDS)-0659 |
| 307 |
ENDF-371 |
D.E. Cullen |
ENDF/B-VII.1 vesus ENDF/B-VII.0: What's Different? |
Mar 2012 |
LLNL-TR-548633 |
| 306 |
ENDF-370 |
D.E. Cullen |
Doppler Broadening Update: Broadening near the Unresolved Resonance Region |
Jan 2012 |
LLNL-TR-534931 |
| 305 |
ENDF-369 |
D.E. Cullen |
A Short History of ENDF/B Unresolved Resonance Parameters |
Oct 2010 |
LLNL-TR-461199 |
| 304 |
ENDF-368 |
D.E. Cullen |
ENDF Cross Sections are not Uniquely Defined |
June 2010 |
LLNL-TR-446331 |
| 303 |
ENDF-367 |
N.M. Larson |
SAMMY User Guidance for ENDF Formats |
Mar-07 |
ORNL/TM-2007/23 |
| 302 |
ENDF-366 |
F.B.Guimaraes, C.Y.Fu, L.C.Leal |
Nuclear cross-section calculations in the 1 MeV to 5 GeV range with combined semi-classical and quantum mechanical models |
Feb-2002 |
ORNL/TM-2001/191 |
| 301 |
ENDF-365 |
L.C.Leal, H.Derrien, J.A.Harvey, K.H.Guber, N.M.Larson, R.R.Spencer |
R-Matrix resonance analysis and statistical properties of the resonance parameters of 233U in the neutron energy range from thermal to 600 eV |
Mar-2001 |
ORNL/TM-2000/372 |
| 300 |
ENDF-364-R1 |
N.M. Larson |
Updated Users' Guide for SAMMY: Multilevel R-matrix ... |
Sep-06 |
ORNL/TM-9179/R7 |
| 299 |
ENDF-363 |
O. Bouland, R. Babut, N.M. Larson |
SAMQUA - A program for Generating All Possible Combinations of Quantum Numbers Leading to the Same Compound Nucleus State in the Framework of the R-Matrix Code SAMMY |
Apr-02 |
JEF-DOC 929 OECD/NEA Publications |
| 298 |
ENDF-362 |
Soo-Youl Oh, Jonghwa Chang, S. Mughabghab |
Neutron Cross Section Evaluations of Fission Products Below the FastEnergy Region |
Apr-00 |
BNL-NCS-67469 |
| 297 |
ENDF-358 |
W.P. Poenitz, S.E. Aumeier |
The Simultaneous Evaluation of the Standards and Other Cross Sections ofImportance for Technology |
Sep-97 |
ANL/NDM-139 |
| 296 |
ENDF-XXX |
R.E. Miller, D.L. Smith |
A Compilation of Information on the 32S(p,g)33Cl Reaction and Propertiesof Excited Levels in 33CL |
Jul-97 |
ANL/NDM-143 |
| 295 |
ENDF-356 |
R.E. MacFarlane |
New Thermal Neutron Scattering Files for ENDF/B-VI Release 2 |
Mar-94 |
LA-12639-MS |
| 294 |
ENDF-355 |
M.C. Moxon |
Comments on the ENDF/B-VI Evaluation for 235-U in the Neutron EnergyRegion from 1 to 20 eV |
Feb-93 |
ORNL/TM-12304 |
| 293 |
ENDF-354 |
L.W.Weston, D.C.Larson |
Compilation of Requests for Nuclear Data |
Jan-93 |
ORNL/TM-12291 |
| 292 |
ENDF-353 |
E.J.Axton |
An Evaluation of Kerma in Carbon and the Carbon Cross Sections |
Feb-92 |
NISTIR 4838 |
| 291 |
ENDF-352 |
J. Katakura, T.R. England |
Augmentation of ENDF/B Fission Product Gamma-Ray Spectra by Calculated Spectra |
Nov-91 |
LA-12125-MS |
| 290 |
ENDF-351 |
A.D. Carlson, W.P. Poenitz, G.M. Hale, et al. |
The ENDF/B-VI Neutron Cross Section Measurements Standards |
May-93 |
NISTIR 5177 |
| 289 |
ENDF-350 |
D.M. Hetrick, D.C. Larson, C.Y. Fu |
Generation of Covariance Files for the Isotopes of Cr, Fe, Ni, Cu, and Pb in ENDF/B-VI |
Feb-91 |
ORNL/TM-11763 |
| 289 |
ENDF-349 |
T.R.England, B.F.Rider |
Evaluation and Compilation of Fission Product Yields |
Oct-94 |
LA-UR-94-3106 |
| 288 |
ENDF-347 |
C.M. Perey, F.G. Perey, J.A. Harvey, et al. |
58Ni+n Transmission, Differential Elastic Scattering and Capture Measurements and Analysis from 5 to 813 KeV |
Sep-88 |
ORNL/TM-10841 |
| 287 |
ENDF-346 |
D.M. Hetrick, C.Y. Fu, D.C. Larson |
Calculated Neutron-Induced Cross Sections for 52-Cr from 1 to 20MeV andComparisons with Experiments |
Sep-87 |
ORNL/TM-10417 |
| 286 |
ENDF-345 |
K. Shibata, D.M. Hetrick |
Calculated Neutron-Induced Cross Sections for 53-Cr from 1 to 20 MeV |
May-87 |
ORNL/TM-10381 |
| 285 |
ENDF-344 |
D.M. Hetrick, C.Y. Fu, D.C. Larson |
Calculated Neutron-Induced Cross Sections for 58,60-Ni from 1 to 20 MeV and Comparisons with Experiments |
Jun-87 |
ORNL/TM-10219 |
| 284 |
ENDF-343 |
L.W. Weston, E.D. Arthur |
Evaluation of the Neutron Cross Sections for PU-240 |
Apr-87 |
ORNL/TM-10386 |
| 283 |
ENDF-342 |
M.S. Milgram, S. Thompson, R. Paulson |
Few Group Cross Sections for 274 Nuclides Based on ENDF/B-V |
Feb-87 |
CRNL-2916 |
| 282 |
ENDF-341 |
C.Y. Fu, D.M. Hetrick |
Update of ENDF/B-V Mod-3 Iron: Neutron Producting Reaction Cross Sections and Energy-Angle Correlations |
Jul-86 |
ORNL/TM-9964 |
| 281 |
ENDF-340 |
D.W. Muir |
Analysis of Central Worths and Other Integral Data from the Los Alamos Benchmark Assemblies |
Oct-84 |
LA-10230-MS |
| 280 |
ENDF-339 |
N.M. Larson |
Updated Users' Guide for SAMMY: Multilevel R-Matrix Fits to Neutron... |
Jun-76 |
ORNL/TM-9179/R1 |
| 279 |
ENDF-338 |
D.K. Olsen |
Report to the 238U Discrepancy Task Force on SIOB Fits to the ORNL, CBNM,and JAERI Transmission Data |
May-84 |
ORNL/TM-9023 |
| 278 |
ENDF-337 |
D.M. Hetrick, C.Y. Fu, D.C. Larson |
Calculated Neutron-Induced Cross Sections for 63,65Cu from 1 to 20 MeV and Comparisons with Experiments |
Aug-84 |
ORNL/TM-9083 |
| 277 |
ENDF-336 |
E.D. Arthur, P.G. Young, D.G. Madland, R.E. MacFarlane |
Evaluation of n + 239Pu Nuclear Data for Revision 2 of ENDF/B-V |
Oct-83 |
LA-9873-MS |
| 276 |
ENDF-335 |
R.W. Roussin, J.R. Knight, J.H. Hubbell, R.J. Howerton |
Description of the DLC-99/HUGO Package of Photon Interaction Data in ENDF/B-V Format |
Dec-83 |
ORNL/RSIC-46 |
| 275 |
ENDF-334 |
D.C. Larson, N.M. Larson, J.A. Harvey |
ORELA Flight Path 1: Determinations of Its Effective Length vs Energy., Experimental Energies, and Energy Resolution Function and Their Uncertainties |
Jun-84 |
ORNL/TM-8880 |
| 274 |
ENDF-333 |
D.C. Larson, N.M. Larson, J.A. Harvey, et al. |
Application of New Techniques to ORELA Neutron Transmission Measurementsand their Uncertainty Analysis: The Case of Natural Nickel ... |
Oct-83 |
ORNL/TM-8203 |
| 273 |
ENDF-332 |
T.R. England |
ENDF/B-V Summary Data for Fission and Actinides |
|
Not published |
| 272 |
ENDF-331 |
R.W. Peelle, T.W. Burrows |
An Annotated Bibliography Covering Generation and Use of Evalusted Cross Section Uncertainty Files |
Mar-83 |
BNL-NCL-51684 |
| 271 |
ENDF-330 |
C.M. Perey, J.A. Harvey, R.L. Macklin, et al. |
Neutron Transmission and Capture Measurements and Analysis of 60Ni from 1 to 450 keV |
Nov-82 |
ORNL-5893 |
| 270 |
ENDF-329 |
P.F. Rose |
Symposium Proceedings: Thermal Reactor Benchmark Calculation... |
|
|
| 269 |
ENDF-328 |
B.A. Magurno, R.R. Kinsey, F.M. Scheffel |
Guidebook for the ENDF/B-V Nuclear Data Files |
Jul-82 |
EPRI NP-2510 |
| 268 |
ENDF-327 |
C.R. Weisbin, D. Gilai, G. deSaussure, R.T. Santoro |
Meeting Cross Section Requirements for Nuclear Energy Design |
Jul-82 |
ORNL/TM-8220 |
| 267 |
ENDF-326 |
D.M. Hetrick, C.Y. Fu, D.C. Larson |
Evaluated Neutron-Induced Cross Sections for 40-Ca from 20 to 40 MeV |
Sep-82 |
ORNL/TM-8290 |
| 266 |
ENDF-325 |
C.Y. Fu |
Summary of ENDF/B-V Evaluations for Carbon, Calcium, Iron, Copper, and Lead and ENDF/B-V Rev.2 for Calcium and Iron |
Sep-82 |
ORNL/TM-8283 |
| 265 |
ENDF-324 Vol.4 |
D.W. Muir, R.E. MacFarlane |
The NJOY Nuclear Data Processing System, Volume IV: The ERRORR and COVR Modules |
Dec-85 |
LA-9303-M Vol.4 |
| 264 |
ENDF-324 Vol.3 |
R.E. MacFarlane, D.W. Muir |
The NJOY Nuclear Data Processing System, Volume III: The GROUPR, GAMINR, and MODER Modules |
Oct-85 |
LA-9303-M Vol.3 |
| 263 |
ENDF-324 Vol.2 |
R.E. MacFarlane, D.W. Muir, R.M. Boicourt |
The NJOY Nuclear Data Processing System, Volume II: The NJOY, RECONR,BROADR, HEATR, AND THERMR Modules |
May-82 |
LA-9303-M Vol.2 |
| 262 |
ENDF-324 Vol.1 |
R.E. MacFarlane, D.W. Muir, R.M. Boicourt |
The NJOY Nuclear Data Processing System, Volume 1: User's Manual |
May-82 |
LA-9303-M Vol.1 |
| 261 |
ENDF-323 |
N.M. Larson |
User's Guide for BAYES: A General-Purpose Computer Code for Fitting a Functional Form to Experimental Data |
Aug-82 |
ORNL/TM-8185 |
| 260 |
ENDF-322 #2 |
B.F. Rider |
Compilation of Fission Products Yields (Microfiche Only) |
Sep-80 |
NEDO 12154-3(C) |
| 259 |
ENDF-322 #1 |
T.R. England, W.B. Wilson, R.E. Schenter, F.M. Mann |
Summary of ENDF/B-V Data for Fission Products and Actinides |
Dec-84 |
EPRI NP 3787
LA-UR 83-1285
|
| 258 |
ENDF-321 |
D.G. Madland |
New Fission Neutron Spectrum Representation for ENDF |
Apr-82 |
LA-9285-MS |
| 257 |
ENDF-320 |
R.J. LaBauve, T.R. England, D.C. George |
Integral Data Testing of ENDF/B Fission Product Data and Comparisons of ENDF/B with other Fission Product Data Files |
Nov-81 |
LA-9090-MS |
| 256 |
ENDF-319 |
D.K. Olsen |
An Evaluation of the Resolved-Resonance-Region Cross Sections of 232Th |
Mar-82 |
ORNL/TM-8056 |
| 255 |
ENDF-318 |
R.B. Kidman |
Los Alamos Benchmarks: Calculations Based on ENDF/B-V Data |
Nov-81 |
LA-9037-MS |
| 254 |
ENDF-317 |
M.A. Bjerke, C.C. Webster |
Neutron Cross Section Libraries in the AMPX Master Interface Format for Thermal and Fast Reactors |
Dec-81 |
ORNL/CSD/TM-164 |
| 253 |
ENDF-316 |
F.M. Mann |
FTR Set 500, A Multigroup Cross-Section Set for FTR Analysis |
Feb-82 |
HEDL-TME81-31 |
| 252 |
ENDF-315 |
R. Gwin, R.R. Spencer, R.W. Ingle |
Measurement of the Average Number of Prompt Neutrons Emitted per Fissionof 233-U Relative to 252-Cf ... |
Nov-81 |
ORNL/TM-7988 |
| 251 |
ENDF-314 |
R.B. Kidman |
ENDF/B-V, LIB-V, CSEWG Benchmarks |
Aug-81 |
LA-8950-MS |
| 250 |
ENDF-313 |
CSEWG Data Testing Committee |
Benchmark Testing of ENDF/B Data for Thermal Reactors |
Jul-81 |
BNL-NCS-29891 |
| 249 |
ENDF-312 |
G. De Saussure |
Representation of the Neutron Cross Sections of Several Fertile and... |
Sep-81 |
ORNL/TM-7945 |
| 248 |
ENDF-311 |
C.R. Weisbin, R.D. McKnight, J. Hardy Jr., et al. |
Benchmark Data Testing of ENDF/B-V |
Aug-82 |
BNL-NCS-31531 |
| 247 |
ENDF-310 |
J.G. Munoz-Cobos |
PAPIN. A Fortran-IV Program to Calculate Cross Sections Probability... |
Aug-81 |
ORNL/TM-7883 |
| 246 |
ENDF-309 |
J.M. Kallfelz |
Preliminary Analysis and Sensitivity Study of Phenix... |
Sep-81 |
ORNL/TM-7505 |
| 245 |
ENDF-308 |
D.M. Hetrick, C.Y. Fu |
A Calculation of Neutron and Gamma-Ray Production Cross Sections for Calcium from 8 to 20 MeV |
Jun-81 |
ORNL/TM-7752 |
| 244 |
ENDF-307 |
D.K. Olsen |
Measurement of Neutron Transmission Spectra Through 232-Th from 8 meV... |
Apr-81 |
ORNL/TM-7661 |
| 243 |
ENDF-306 |
D.W. Muir, R.J. LaBauve |
COVFILS: A 30-Group Covariance Library Based on ENDF/B-V |
Mar-81 |
LA-8733-MS |
| 242 |
ENDF-305 |
J.D. Smith III, B.L. Broadhead |
Multigroup Covariance Matrices for Fast Reactor Studies |
Apr-81 |
ORNL/TM-7389 |
| 241 |
ENDF-304 |
E.D. Arthur, P.G. Young |
Evaluated Neutron-Induced Cross Sections for 54 and 56 Fe to 40 MeV |
Dec-80 |
LA-8626-MS |
| 240 |
ENDF-303 |
D.M. Hetrick, C.Y. Fu |
GLUCS: A Generalized Least-Squares Program for Updating Cross-Section Evaluations with Correlated Data Sets |
Oct-80 |
ORNL/TM-7341 |
| 239 |
ENDF-302 |
C.Y. Fu, F.G. Perey |
Evaluation of Neutron and Gamma-Ray Production Cross Section for Natural Iron (ENDF/B-V MAT 1326) |
Nov-80 |
ORNL/TM-7523 |
| 238 |
ENDF-301 |
A.D. Carlson, M.R. Bhat |
ENDF/B-V Cross Section Measurement Standards |
Oct-82 |
BNL-NCS-51619 |
| 237 |
ENDF-300 |
M.R. Bhat |
Standard Reference and Other Important Nuclear Data |
May-84 |
BNL-NCS-51123 5/84 |
| 236 |
ENDF-300 |
M.R. Bhat |
Standard Reference and Other Important Nuclear Data |
Feb-82 |
BNL-NCS-51123 2/82 |
| 235 |
ENDF-300 |
M.R. Bhat |
Standard Reference and Other Important Nuclear Data |
Mar-81 |
BNL-NCS-51123 5/81 |
| 234 |
ENDF-300 |
M.R. Bhat |
Standard Reference and Other Important Nuclear Data by the CSEWG |
Dec-79 |
BNL-NCS-51123 |
| 233 |
ENDF-299 |
D.C. Larson |
An Evaluation Cross Sections for Neutron-Induced Reactions in Sodium |
Sep-80 |
ORNL-5662 |
| 232 |
ENDF-298 |
C.M. Perey, F.G. Perey |
Evaluation of Resonance Parameters for Neutron Interaction with Iron Isotopes for Energies up to 400 keV |
Sep-80 |
ORNL/TM-6405 |
| 231 |
ENDF-297 |
N.M. Larson, F.G. Perey, J.A. Harvey |
User's Guide for SAMMY: A Computer Model for Multilevel R-Matrix Fits to Neutron Data Using Bayes' Equstions |
Nov-80 |
ORNL/TM-7485 |
| 230 |
ENDF-296 |
R.W. Roussin, C.R. Weisbin, J.E. White, et al. |
VITAMIN C: The CTR Processed Multigroup Cross Section Library for Neutronics Studies |
Jul-80 |
ORNL/RSIC-37 |
| 229 |
ENDF-295 |
J.D. Smith III |
Processing ENDF/B-V Uncertainty Data into Multigroup Covariance Matrices |
Jun-80 |
ORNL/TM-7221 |
| 228 |
ENDF-294 |
M. Divadeenam |
Ni Elemental Neutron Induced Reaction Cross Section Evaluation |
Mar-79 |
BNL-NCS-51346 |
| 227 |
ENDF-293 |
T. Burrows |
ENDF/B-V Actinide Decay Data |
|
|
| 226 |
ENDF-292 |
B.F. Rider |
Compilation Efficient Products Yield |
Sep-80 |
NEDO 12154-3(B) |
| 225 |
ENDF-291 #2 |
J.D. Drischler |
The COVERX Service Module of the FORSS System |
Apr-80 |
ORNL/TM-7181 |
| 224 |
ENDF-291 #1 |
J.L. Lucius, C.R. Weisbin, J.H. Marable, et al. |
A Users Manual for the FORSS Sensitivity and Uncertainty Analysis Code System |
Jan-84 |
ORNL-5316 |
| 223 |
ENDF-290 |
N.M. Larson, D.K. Olsen |
Preliminary Study of Pseudorandom Binary Sequence Pulsing of ORELA |
Mar-80 |
ORNL/TM-6632 |
| 222 |
ENDF-289 |
R. Gwin, R.R. Spencer, R.W. Ingle, et al. |
Measurement of the Average Number of Prompt Neutrons Emitted per Fissionof 235U Relative to 252Cf ... |
Jan-80 |
ORNL/TM-7148 |
| 221 |
ENDF-288 |
R.J. Barrett, W.E. Ford III, Y. Gohar, et al. |
Comparison of Photon-Production Processing Codes LAPHNGAS, MACK-IV, and NJOY |
Nov-79 |
LA-8100-MS |
| 220 |
ENDF-287 |
W.E. Ford III, C.C. Webster, B.R. Diggs, et al. |
FCXSEC:Multigroup Cross-Section Libraries for Nuclear Fuel CycleShielding Calculations |
May-80 |
ORNL/TM-7038 |
| 219 |
ENDF-286 |
A. Prince, T.W. Burrows |
Evaluation of Natural Chromium Neutron Cross Sections for ENDF/B-V |
Feb-79 |
BNL-NCS-51152 |
| 218 |
ENDF-285 |
R.R. Spencer, R. Gwin, R. Ingle, H. Weaver |
Interim Report on the ORNL Absolute Measurements of nu p for 252Cf |
Sep-79 |
ORNL/TM-6805 |
| 217 |
ENDF-284 |
G.L. Morgan, G.T. Chapman |
The O(n,x gamma) Reaction Cross Section for Incident Neutron EnergiesBetween 6.5 and 20.0 MeV |
Sep-79 |
ORNL-5575 |
| 216 |
ENDF-283 |
P.G. Young, L. Stewart |
Evaluated Data for n+Berylium 9 Reactions |
Jul-79 |
LA-7932-MS |
| 215 |
ENDF-282 |
G.L. Morgan |
The Th(n,x gamma) Reaction Cross Section for Incident Neutron EnergiesBetween 0.3 and 20.0 Mev |
Aug-79 |
ORNL/TM-6758 |
| 214 |
ENDF-281 |
D.K. Olsen, G.L. Morgan, J.W. McConnell |
Measurement of 238-U(n,n'gamma) and Li-7(n,n'gamma) Gamma-Ray Production Cross Sections |
May-79 |
ORNL/TM-6832 |
| 213 |
ENDF-280 |
D.M. Hetrick, D.C. Larson, C.Y. Fu |
Status of ENDF/B-V Neutron Emission Spectra Induced by 14 MeV Neutrons |
Apr-79 |
ORNL/TM-6637 |
| 212 |
ENDF-279 |
A. Prince |
ENDF/B-V Neutron Cross Section Evaluation for the Krypton Isotopes |
Jan-79 |
BNL-NCS-51028 |
| 211 |
ENDF-278 |
L. Stewart |
Summary of Fission Spectrum Workshop Held at NNCSC |
Oct-78 |
LA-7739-C |
| 210 |
ENDF-277 |
F.C. Difilippo, R.B. Perez, G. deSaussure, et al. |
The U-238 Neutron Induced Fission Cross Section for Incident NeutronEnergies Between 5 eV and 3.5 MeV |
Feb-79 |
ORNL/TM-6788 |
| 209 |
ENDF-276 |
C.R. Weisbin |
Specifications for Adjusted Cross Sections and Covariance Libraries based... |
Feb-79 |
ORNL-5517 |
| 208 |
ENDF-275 |
J.L. Lucius, E.M. Oblow, G.W. Cunningham,III |
A Users Guide for the JULIET Module of the FORSS Sensitivity andUncertainty Analysis Code System |
Feb-79 |
ORNL/TM-6594 |
| 207 |
ENDF-274 |
C.R. Weisbin, R.W. Roussin, J.J. Wagschal, et al. |
VITAMIN-E: An ENDF/B-V Multigroup Cross Section Library for LMFBR Core and Shield, LWR Shield,... |
Dec-78 |
ORNL-5505 |
| 206 |
ENDF-273 |
G.L. Morgan |
Cross Sections for the Cu(n,xn) and Cu(n,x gamma) Reactions Between 1 and 20 Mev |
Feb-79 |
ORNL-5499 |
| 205 |
ENDF-272 |
R.E. MacFarlane, R.J. Barrett, D.W. Muir, R.M. Boicourt |
The NJOY Nuclear Data Processing System: Users Manual |
Dec-78 |
LA-7584-M |
| 204 |
ENDF-271 |
F.M. Mann |
HEDL Evaluation of Thorium Cycle Cross Sections for ENDF/B-V |
Nov-78 |
HEDL-TME-78-100 |
| 203 |
ENDF-270 |
P.F. Rose, S. Pearlstein, O. Ozer |
Symposium Proceedings: Nuclear Data Problems for Thermal ReactorApplications |
Jun-79 |
BNL-NCS-25047 |
| 202 |
ENDF-269 |
J.U. Koppel, D.H. Houston |
Reference Manual for ENDF Thermal Neutron Scattering Data |
Jul-78 |
GA-8774 Revised |
| 201 |
ENDF-268 |
M.R. Bhat |
Evaluation of Th-232 for ENDF/B-V |
Feb-81 |
BNL-NCS-51360 |
| 200 |
ENDF-267 |
C.Y. Fu |
Evaluation for Th-233(n,n')(n,2n) and (n,3n) Cross Section |
May-78 |
ORNL/TM-6316 |
| 199 |
ENDF-266 |
Y.D. Harker, J.W. Rogers, D.A. Millsap |
Fission Product and Reactor Dosimetry Studies at Coupled Fast Reactivity Measurements Facility |
Mar-78 |
TREE-1259 |
| 198 |
ENDF-265 |
C.R. Weisbin, J.H. Marable, J. Hardy Jr., R.D. McKnight |
Sensitivity Coefficient Compilation for CSEWG Data Testing Benchmarks |
Aug-78 |
BNL-NCS-24853 |
| 197 |
ENDF-264 |
R. Gwin, R.R. Spencer, R.W. Ingle, et al. |
Measurements of the Average Number of Prompt Neutrons Emitted per Fission of 239Pu and 235U |
May-78 |
ORNL/TM-6246 |
| 196 |
ENDF-263 |
M.L. Williams, C.R. Weisbin |
Sensitivity and Uncertainity Analysis for Functionals of the Time-Dependent Nuclide Density Field |
Apr-78 |
ORNL-5393 |
| 195 |
ENDF-262 |
R. Gwin |
Review and Combination of Experimental Results for Neutron-Emission per Fission of 232Th |
May-78 |
ORNL/TM-6245 |
| 194 |
ENDF-261 |
G. De Saussure, D.K. Olsen, R.B. Perey |
SIOB: A Fortran Code for Least Squares Shape Fitting Several Neutron Transmission Measurements Using ... |
May-78 |
ORNL/TM-6286 |
| 193 |
ENDF-260 |
G. De Saussure, R.L. Macklin |
Evaluation of the Th-232 Neutron Capture Cross Section above 3 keV |
Feb-78 |
ORNL/TM-6161 |
| 192 |
ENDF-259 |
F.G. Perey |
Contributions to Few-Channel Spectrum Unfolding |
Feb-78 |
ORNL/TM-6267 |
| 191 |
ENDF-258 |
R.E. Maerker, F.J. Muckenthaler, C.E. Clifford |
SB4. Measurements and Calculations of the ORNL CRBR Upper Axial Shield Experiment |
Jun-77 |
ORNL-5259 |
| 190 |
ENDF-257 |
G. De Saussure, D.K. Olsen, R.B. Perey, F.C. Difilippo |
Evaluation of the U-238 Neutron Cross Sections for Incident Neutron Energies up to 4 Kev |
Jan-78 |
ORNL/TM-6152 |
| 189 |
ENDF-256 |
J.D. Drischler, J.H. Marable, C.R. Weisbin |
COVERT and CAVALIER: Two Computer Codes for Estimating Uncertainties of Calculated Neutronics Parameters ... |
Aug-78 |
ORNL/TM-6078 |
| 188 |
ENDF-255 |
D.K. Olsen, G. deSaussure, R.B. Perey, et al. |
150-m Measurement of 0.880- to 100.0-keV Neutron Transmissions ThroughFour Samples of 238U |
Oct-77 |
ORNL/TM-5915 |
| 187 |
ENDF-254 |
F.G. Perey |
Least-Squares Dosimetry Unfolding: the Program STAY'SL |
Oct-77 |
ORNL/TM-6062 |
| 186 |
ENDF-253 |
E.T. Tomlinson, J.L. Lucius, J.D. Drischler |
A Compendium of Energy-Dependent Sensitivity Profiles for TRX-2 Thermal Lattice |
Mar-78 |
ORNL-5336 |
| 185 |
ENDF-252 |
E.T. Tomlinson, D. deSaussure, C.R. Weisbin |
Sensitivity Analysis of TRX-2 Lattice Parameters with Emphasis on Epithermal 238-U Capture |
Mar-77 |
EPRI-NP-346 |
| 184 |
ENDF-251 |
F.M. Mann |
HEDL Evaluation of Actinide Cross Section for ENDF/B-V |
Jun-77 |
HEDL-TME-77-54 |
| 183 |
ENDF-250 |
J.H. Marable, J.D. Drischler, C.R. Weisbin |
SENDIN and SENTINEL: Two Computer Codes to Assess the Effects of Nuclear Data Changes |
Jul-77 |
ORNL/TM-5946 |
| 182 |
ENDF-249 |
F.G. Perey |
Data Covariance Files for ENDF/B-V |
Jul-77 |
ORNL/TM-5938 |
| 181 |
ENDF-248 |
M.R. Bhat |
Evaluation of 235-U Neutron Cross Section and Gamma Ray Production Data for ENDF/B-V |
Mar-80 |
BNL-NCS-51184 |
| 180 |
ENDF-247 |
D.G. Madland, L. Stewart |
Light Ternary Fission Products: Probabilities and Charge Distributions |
Apr-77 |
LA-6783-MS |
| 179 |
ENDF-246 |
A. Prince |
Evaluation of Chromium Neutron and Gamma Production Cross Sections for ENDF/IV |
Aug-76 |
BNL-NCS-50593 |
| 178 |
ENDF-245 |
F.M. Mann |
HAUSER*4: A Computer Code to Calculate Nuclear Cross Sections |
Sep-76 |
HEDL-TME-76-80 |
| 177 |
ENDF-244 |
G.M. Hale, L. Stewart, P.G. Young |
Light Element Standard Cross Standards for ENDF/B-IV |
Oct-76 |
LA-6518-MS |
| 176 |
ENDF-243 Vol.II |
P.F. Rose, T.W. Burrows |
ENDF/B Fission Product Decay Data |
Aug-76 |
BNL-NCS-50545 Vol.II |
| 175 |
ENDF-243 Vol.I |
P.F. Rose, T.W. Burrows |
ENDF/B Fission Product Decay Data |
Aug-76 |
BNL-NCS-50545 Vol.I |
| 174 |
ENDF-242 |
M. Stamatelatos, T.R. England |
Beta-Energy Averaging and Beta Spectra |
Aug-76 |
LA-6445-MS |
| 173 |
ENDF-241 |
D.G. Madland, T.R. England |
Distribution of Independent Fission-Product Yields to Isomeric States |
Nov-76 |
LA-6595-MS |
| 172 |
ENDF-240 |
D.G. Madland, T.R. England |
The Influence of Pairing on the Distribution of Independent Yield Strengths in Neutron-induced Fission |
Jul-76 |
LA-6430-MS |
| 171 |
ENDF-239 |
H. Henryson II |
MC2-2: A Code to Calculate Fast Neutron Spectra and Multigroup CrossSections |
Jun-76 |
ANL-8144 |
| 170 |
ENDF-238 |
R.A. Grimesey |
ETOP 14: A Fortran Code to Process ENDF/B Data into the 68-Group PHROG... |
Jul-76 |
ANCR-1322 |
| 169 |
ENDF-237 |
C.R. Weisbin, P.D. Soran, R.E. MacFarlane, et al. |
MINX: A Multigroup Interpretation of Nuclear X-Sections from ENDF/B |
Sep-76 |
LA-6486-MS |
| 168 |
ENDF-236 |
C.R. Weisbin |
Application of FORSS Sensitivity and Uncertainty Methodology to Fast |
Dec-76 |
ORNL/TM-5563 |
| 167 |
ENDF-235 |
J.D. Drischler, C.R. Weisbin |
Compilation of Multigroup Cross-Section Covariance Matrices for Several Important Reactor Materials |
Oct-77 |
ORNL-5318 |
| 166 |
ENDF-234 |
J.H. Marable, J.L. Lucius, C.R. Weibin |
Compilation of Sensitivity Profiles for Several CSEWG Fast Reactor Benchmarks |
Mar-77 |
ORNL-5262 |
| 165 |
ENDF-233 |
R.W. Peelle |
An Evaluation for ENDF/B-IV of the Neutron Cross Sections for U-235 from 82 eV to 25 keV |
Jun-76 |
ORNL-4955 |
| 164 |
ENDF-232 |
A. Prince |
Evaluation of Neutron Cross Sections For the Krypton Isotopes |
Aug-74 |
BNL-NCS-50503 |
| 163 |
ENDF-231 |
Void See ENDF-265 |
|
|
|
| 162 |
ENDF-230 Vol.II |
E.M. Bohn, R. Maerker, B.A. Magurno, et al. |
Benchmark Testing of ENDF/B-IV |
Mar-76 |
BNL-NCS-21118 Vol.II |
| 161 |
ENDF-230 Vol.I |
E.M. Bohn, R. Maerker, B.A. Magurno, et al. |
Benchmark Testing of ENDF/B-IV |
Mar-76 |
BNL-NCS-21118 Vol.I |
| 160 |
ENDF-229 |
S.F. Mughabghab, T.J. Krieger |
Neutron Cross Sections of 59Co Below 100 keV |
Apr-75 |
BNL-NCS-50468 |
| 159 |
ENDF-228 |
R.E. Maerker |
SB3. Experiment on Secondary Gamma-Ray Production Cross Sections Averaged ... |
Jan-76 |
ORNL/TM-5204 |
| 158 |
ENDF-227 |
R.E. Maerker |
SB2. Experiment on Secondary Gamma-Ray Production Cross Sections Arisingfrom Thermal-Neutron Capture ... |
Jan-76 |
ORNL-TM-5203 |
| 157 |
ENDF-226 |
R.E. Maerker |
Subject: Benchmark |
|
ORNL/TM-5202 |
| 156 |
ENDF-225 |
B.A. Magurno |
ENDF/B-IV Cross Section Measurement Standards |
Aug-75 |
BNL-NCS-50464 |
| 155 |
ENDF-224 |
C.R. Weisbin |
Specification for Pseudo-Composition-Independent Fine-Group and... |
Dec-75 |
ORNL/TM-5142 |
| 154 |
ENDF-223 |
T.R. England, R.E. Schenter |
ENDF/B-IV Fission-Product Files: Summary of Major Nuclide Data |
Oct-75 |
LA-6116-MS |
| 153 |
ENDF-222 |
G.L. Morgan |
Cr(n,x gamma)Reaction Cross Section for Incident Neutron Energies Between ... |
Jan-76 |
ORNL/TM-5098 |
| 152 |
ENDF-221 |
E. Newman, G.L. Morgan |
V(n,x gamma) Reaction Cross Section for Incident Neutron Energies Between 0.2 and 20.0 MeV |
Apr-76 |
ORNL/TM-5299 |
| 151 |
ENDF-220 |
G.L. Morgan, E. Newman |
The Mo(n,x gamma) Reaction Cross Section for Incident Neutron Energies Between 0.2 and 20.0 MeV |
Dec-75 |
ORNL-TM-5097 |
| 150 |
ENDF-219 |
J.K. Dickens, G.L. Morgan, E. Newman |
The Nb(n,x gamma) Reaction Cross Section for Incident Neutron Energies Between 0.65 and 20.0 MeV |
Sep-75 |
ORNL-TM-4972 |
| 149 |
ENDF-218 |
C.R. Weisbin, E.M. Oblow, J. Ching, et al. |
Cross Section and Method Uncertainties:the Applicati on of SensitivityAnalysis ... |
Aug-75 |
ORNL-TM-4847 |
| 148 |
ENDF-217 |
S. Pearlstein |
Seminar on 238-U Resonance Capture |
Mar-75 |
BNL-NCS-50451 |
| 147 |
ENDF-216 |
B.A. Magurno |
ENDF/B-IV Dosimetry File |
Apr-75 |
BNL-NCS-50446 |
| 146 |
ENDF-215 |
S.F. Mughabghab, A. Prince, M.D. Goldberg, et al. |
Evaluated Neutron Cross Sections of Au-197 |
Oct-74 |
BNL-50439 |
| 145 |
ENDF-214 |
H. Takahashi |
Evaluation of the Neutron Cross Sections for Eu-152 and Eu-154 |
Nov-74 |
BNL-19456 |
| 144 |
ENDF-213 |
H.Takahashi |
Evaluation of the Neutron and Gamma-Ray Production Cross Sections of Eu-151 and Eu-153 |
Nov-74 |
BNL-19455 |
| 143 |
ENDF-212 |
M.R. Bhat, B.A. Magurno, S. Pearlstein, F.M. Scheffel |
Nuclear Data for CTR Related Projects |
Oct-74 |
BNL-19344 |
| 142 |
ENDF-211 |
Void-Not Used |
|
|
|
| 141 |
ENDF-210 |
C.W. Reich, R.G. Helmer, M.H. Putnam |
Radioactive-Nuclide Decay Data for ENDF/B |
Aug-74 |
ANCR-1157 |
| 140 |
ENDF-209 |
Void See ENDF-246 |
|
|
|
| 139 |
ENDF-208 |
H. Takahashi |
Evaluation of the Neutron and Gamma-Ray Production Cross Sections for 55Mn |
Nov-74 |
BNL-50442 |
| 138 |
ENDF-207 |
M.R. Bhat |
Neutron and Gamma-Ray Production Cross Sections for Nickel |
Oct-74 |
BNL-50435 |
| 137 |
ENDF-206 |
W.E. Kinney, F.G. Perey |
Pb-206, Pb-207, and Pb-208 Neutron Elastic and Inelastic Scattering Cross From 5.50 To 8.50 MeV |
Jun-74 |
ORNL-4909 |
| 136 |
ENDF-205 |
F.G. Perey |
Nitrogen Neutron Elastic and Inelastic Scattering Cross Sections From 4.34... |
Mar-74 |
ORNL-4905 |
| 135 |
ENDF-204 |
W.E. Kinney, F.G. Perey |
Cu-63 and Cu-65 Neutron Elastic and Inelastic Scattering Cross Sections From 5.50 To 8.50 MeV |
Mar-74 |
ORNL-4908 |
| 134 |
ENDF-203 |
W.E. Kinney, F.G. Perey |
Fe-54 Neutron Elastic and Inelastic Scattering Cross Sections From 5.50 To 8.50 MeV |
Mar-74 |
ORNL-4907 |
| 133 |
ENDF-202 1991 |
R. McKnight |
Cross Sections Evaluation Working Group Benchmark Specification |
Sep-91 |
BNL-19302 Upd. 9/91 |
| 132 |
ENDF-202 1983 Vol.2 Suppl. |
P.F. Rose |
Cross Section Evaluation Working Group Benchmark Specifications |
Sep-86 |
BNL-19302 Vol.2 Suppl. |
| 131 |
ENDF-202 1983 Vol.2 |
P.F. Rose |
Cross Sections Evaluation Working Group Benchmark Specifications |
Dec-83 |
BNL-19302 Vol.2 |
| 130 |
ENDF-202 1982 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Sep-82 |
BNL-19302 Upd. 9/82 |
| 129 |
ENDF-202 1981-2 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Nov-81 |
BNL-19302 Upd.11/81 |
| 128 |
ENDF-202 1981-1 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
May-81 |
BNL-19302 Upd. 5/81 |
| 127 |
ENDF-202 1978-2 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Oct-78 |
BNL-19302 Upd. 10/78 |
| 126 |
ENDF-202 1978-1 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Feb-78 |
BNL-19302 Upd. 2/78 |
| 125 |
ENDF-202 1976 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Oct-76 |
BNL-19302 Upd. 10/76 |
| 124 |
ENDF-202 1975 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
May-75 |
BNL-19302 Upd. 5/75 |
| 123 |
ENDF-202 1974 Vol.2 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Nov-74 |
BNL-19302 Vol. 2 |
| 122 |
ENDF-202 1974 Vol.1 |
H. Alter |
Cross Sections Evaluation Working Group Benchmark Specification |
Nov-74 |
BNL-19302 |
| 121 |
ENDF-201 4th Ed. Suppl. 1 |
V. McLane |
ENDF/B-VI Summary Documentation Supplement 1, ENDF/HE-VI Summary Documentation |
Dec-96 |
BNL-NCS-17541 4th Ed. Suppl.1 |
| 120 |
ENDF-201 4th Ed. |
P.F. Rose |
ENDF/B-VI Summary Documentation |
Oct-91 |
BNL-NCS-17541 4th Ed. |
| 119 |
ENDF-201 1985 |
B.A. Magurno. P.G. Young |
ENDF/B-V.2 Summary Documentation |
Jan-85 |
BNL-NCS-17541 3rd Ed. Suppl.1 |
| 118 |
ENDF-201 1979 |
R.R. Kinsey |
ENDF/B Summary Documentation |
Jul-79 |
BNL-NCS-17541 3rd Ed. |
| 117 |
ENDF-201 1975 |
D.I. Garber |
ENDF/B Summary Documentation |
Oct-75 |
BNL-NCS-17541 2nd Ed. |
| 116 |
ENDF-201 1973 |
O. Ozer, D. Garber |
ENDF/B Summary Documentation |
May-73 |
BNL-NCS-17541 1st Ed. |
| 115 |
ENDF-200 2nd Edition |
D.I. Garber, C. Brewster |
ENDF/B Cross Sections |
Oct-75 |
BNL-17100 2nd Ed. |
| 114 |
ENDF-200 |
D.E. Cullen, P.J. Hlavac |
ENDF/B Cross Sections |
Nov-72 |
BNL-17100 |
| 113 |
ENDF-199 |
B. Hutchins |
Subject: Pu-239 |
|
|
| 112 |
ENDF-198 |
W.E. Kinney, F.G. Perey |
Natural Chromium and Cr-52 Neutron Elastic and Inelastic Scattering Cross Sections from 4.07 to 8.56 MeV |
Jan-74 |
ORNL-4806 |
| 111 |
ENDF-197 |
W.E. Kinney, F.G. Perey |
Natural Nickel and Ni-60 Neutron Elastic and Inelastic Scattering Cross Sections from 4.07 to 8.56 MeV |
Jan-74 |
ORNL-4807 |
| 110 |
ENDF-196 |
D.R. Finch |
Standard Thermal Energy Group Structure for Generation of Thermal Group Constants from ENDF/B Data |
Mar-74 |
DP-1346 |
| 109 |
ENDF-195 |
F. Schmittroth |
Neutron Capture Calculations for En=3D100 keV to 4 MeV |
Nov-73 |
HEDL-TME-73-79 |
| 108 |
ENDF-194 |
F. Schmittroth, R.E. Schenter |
Fast Neutron Capture Cross Section for Fission Product Isotopes |
Aug-73 |
HEDL-TME-73-63 |
| 107 |
ENDF-193 |
Void-Not Used |
|
|
|
| 106 |
ENDF-192 |
C.R. Weisbin |
Specification of a Generally Useful Multigroup Structure for NeutronTranspor |
May-73 |
LA-5277-MS |
| 105 |
ENDF-191 |
R.Q. Wright |
ADLER-III: A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters |
Jun-73 |
ORNL-TM-4257 |
| 104 |
ENDF-190 |
A. Prince, M.K. Drake, P. Hlavac |
An Analysis of the Pu-239 Neutron Cross Sections from 20 keV to 20 MeV |
Apr-73 |
BNL-50388 |
| 103 |
ENDF-189 |
R.E. Maerker |
SDT12. The ORNL Benchmark Experiment for Neutron Transport Through Sodium |
Sep-74 |
ORNL-TM-4223 |
| 102 |
ENDF-188 |
R.E. Maerker |
SDT11. The ORNL Benchmark Experiment for Neutron Transport Through Iron and Stainless Steel, Part 1 |
Sep-74 |
ORNL-TM-4222 |
| 101 |
ENDF-187 |
M.R. Bhat |
Multi-Level Effects in Reactor Calculations and the Probability TableMethod |
Apr-73 |
BNL-50387 |
| 100 |
ENDF-186 |
M.R. Bhat, M.D. Goldberg, R.R. Kinsey, et al. |
Neutron and Gamma Ray Production Cross Sections for Silicon |
Mar-73 |
BNL-50379 |
| 99 |
ENDF-185 |
M.R. Bhat, S.F. Mughabghab |
Evaluated Neutron Cross Sections for the Stable Isotopes of Xenon |
Feb-73 |
BNL-50374 |
| 98 |
ENDF-184 |
R.Q. Wright |
A Comparison of the Group Constants Generated by the ENDF/B Processing... |
Apr-73 |
ORNL/TM-4041 |
| 97 |
ENDF-183 |
H. Alter |
Report to the Cross Section Evaluation Working Group GEDANKEN Calculations |
Nov-72 |
BNL-17510 |
| 96 |
ENDF-182 |
D.J. Dudziak, G.E. Bosler |
LAPHAN: A Code to Compute the P0 to P4 Multigroup Photon-Production Matrices |
Jan-73 |
LA-4963 |
| 95 |
ENDF-181 |
P.F. Rose, H. Alter, R.K. Paschall, A.W. Thiele |
SDT9. CSEWG Shielding Benchmark Specifications Neutron AttenuationMeasurements in a Mockup of the FFTF Radial Shield |
Jan-73 |
AI-AEC-13048 |
| 94 |
ENDF-180 |
W.E. Ford III |
The Testing of Photon Production Data from ENDF/B-III Material 1135(Al) |
Jan-73 |
ORNL-TM-4032 |
| 93 |
ENDF-179 |
M.K. Drake |
ENDF/B-III Cross Section Measurement Standards |
Jul-72 |
BNL-17188 |
| 92 |
ENDF-178 |
F.G. Perey |
A Test of Neutron Total Cross Section Evaluations from 0.2 to 20 MeV for ... |
Dec-72 |
ORNL-4823 |
| 91 |
ENDF-177 |
R.E. Maerker |
SDT7. Experiment on Secondary Gamma-ray Production Cross Sections Arising ... |
Oct-72 |
ORNL/TM-3974 |
| 90 |
ENDF-176 |
R.E. Maerker |
SDT6. Experiment on Secondary Gamma-Ray Production Cross Sections Arising from Thermal-Neutron Capture in Iron, Stainless Steel, Nitrogen, and Sodium |
Oct-72 |
ORNL/TM-3957 |
| 89 |
ENDF-175 |
P.G. Young |
A Preliminary Evaluation of the Neutron and Photon-Production Cross Sections for Aluminum |
Dec-72 |
LA-4726 |
| 88 |
ENDF-174 |
D.G. Foster |
A Preliminary Evaluation of the Neutron and Photon Production Cross Sections of Oxygen |
Aug-72 |
LA-4780 |
| 87 |
ENDF-173 |
P.G. Young |
Evaluation of the Neutron and Gamma Ray Production Cross Sections for Nitrogen |
Sep-72 |
LA-4725 |
| 86 |
ENDF-172 |
W.E. Ford III |
Comparison of (n th,gamma) Yields from the Current ENDF/B-III Data with Published Data |
Aug-72 |
ORNL-TM-3910 |
| 85 |
ENDF-171 |
F. Schmittroth |
Neutron Resonance Spacings for Spherical Nuclei |
Jan-73 |
HEDL-TME-73-30 |
| 84 |
ENDF-170 |
R.E. Maerker |
SDT5. Stainless-Steel Broomstick Experiment |
Jul-72 |
ORNL/TM-3871 |
| 83 |
ENDF-169 |
R.E. Maerker |
SDT4. Sodium Broomstick Experiment |
Jul-72 |
ORNL/TM-3870 |
| 82 |
ENDF-168 |
R.E. Maerker |
SDT3. Nitrogen Broomstick Experiment |
Jul-72 |
ORNL/TM-3869 |
| 81 |
ENDF-167 |
R.E. Maerker |
SDT2. Oxygen Broomstick Experiment |
Jul-72 |
ORNL/TM-3868 |
| 80 |
ENDF-166 |
R.E. Maerker |
SDT1. Iron Broomstick Experiment |
Jul-72 |
ORNL/TM-3867 |
| 79 |
ENDF-165 |
Void-Not Used |
|
|
|
| 78 |
ENDF-164 |
R.E. Schenter |
FTR Set 300, Multigroup Cross Sections for FTR Design |
Oct-71 |
HEDL-TME-71-153 |
| 77 |
ENDF-163 |
M.R. Bhat, A. Prince |
Evaluated Neutron Cross Sections for Ag-107, Ag-109 and Cs-133 |
Apr-73 |
BNL-50383 |
| 76 |
ENDF-162 |
O.D. Simpson, F.B. Simpson |
Evaluation of the Pu-239 Cross Sections in the Resonance Region for the ENDF/B-III Data File |
Dec-71 |
ANCR-1045 |
| 75 |
ENDF-161 |
J.R. Smith, R.C. Young |
U-235 Resolved Resonance Parameters for ENDF/B-III |
Dec-71 |
ANCR-1044 |
| 74 |
ENDF-160 |
A.D. MacKellar, R.E. Schenter |
Optical Model Studies for Fast Neutron Capture Cross Section Calculations |
Aug-72 |
HEDL-TME-71-154 |
| 73 |
ENDF-159 |
R.E. Schenter |
Cross Section Evaluations of Twenty-Seven Fission Product Isotopes for... |
Oct-71 |
HEDL-TME-71-143 |
| 72 |
ENDF-158 |
F.J.McCrosson, D.R.Finch, E.C.Olson |
Testing of ENDF/B-Thermos Cross Sections for H2O, D2O, C, ZrH2, (C2H4)x, Be, Be0, C6H6, and U02 |
Oct-71 |
DP-1276 |
| 71 |
ENDF-157 |
M. Raymund |
Subject:PSYCHE Code |
|
|
| 70 |
ENDF-156 |
D.J. Dudziak |
LAPHANO: P0 Multigroup Photon Production Matrix and Source Code for ENDF |
Jan-72 |
LA-4750-MS |
| 69 |
ENDF-155 |
D.J. Dudziak, J.M. Romero |
VIXEN: A Code to Check Physical Consistency of Photon-Production Data in Rived ENDF Format |
Oct-71 |
LA-4739 |
| 68 |
ENDF-154 |
Void-Not Used |
|
|
|
| 67 |
ENDF-153 |
B.R. Leonard Jr. |
Thermal Cross Sections of the Fissile and Fertile Nuclei for ENDF/B-II |
Jun-71 |
BNWL-1586 |
| 66 |
ENDF-152 |
H.C. Honeck, D.R. Finch |
FLANGE II(Version 71-1). A Code to Process Thermal Neutron Data from an ENDF/B tape |
Oct-71 |
DP-1278 |
| 65 |
ENDF-151 |
R.W. Roussin |
Preparation of Data Sets in ENDF Format for Na, Mg, Cl, and K for Use in ... |
May-71 |
ORNL/TM-3429 |
| 64 |
ENDF-150 |
E.H. Ottewitte, J.M. Otter, P.F. Rose, C.L. Dunford |
An Evaluation of Ta-181 and Ta-182 for the ENDF/B Data File |
Sep-71 |
AI-AEC-12990 |
| 63 |
ENDF-149 |
H. Alter |
Evaluation of Several ENDF/B-II Cross-Section Sets Using Monte Carlo... |
Jun-71 |
AI-AEC-13001 |
| 62 |
ENDF-148 |
M.R. Bhat |
ENDF/B Processing Codes for the Resonance Region |
Jun-71 |
BNL-50296 |
| 61 |
ENDF-147 |
H. Alter, R.S. Hubner |
Status of Fast Neutron Cross Section Data Testing using ENDF/B-II DataFiles |
May-71 |
AI-AEC-12999 |
| 60 |
ENDF-146 Suppl. |
M. Raymund |
ETOT, A Fortran-IV Program to Process Data from the ENDF/B File to Thermal Library Format |
Nov-73 |
WCAP-7363 |
| 59 |
ENDF-146 |
C.L. Beard, R.A. Dannels |
ETOT: A Fortran IV Program to Process Data from the ENDF/B File to Thermal Library Format |
Mar-71 |
WCAP-7363 |
| 58 |
ENDF-145 |
B.A. Hutchins, C.L. Cowen, M.D. Kelley, J.E. Turner |
ENDRUN-II: A Computer Code to Generate a Generalized Multigroup Data File from ENDF/B |
Mar-71 |
GEAP-13704 |
| 57 |
ENDF-144 |
A.Z. Livolsi |
Evaluation of Tc-99 and Rh-103 Neutron Cross Sections for ENDF/B-III |
Nov-71 |
BAW-1367 |
| 56 |
ENDF-143 |
R.B. Kidman, R.E. Schenter |
Group Constants for Fast Reactor Calculations |
Mar-71 |
HEDL-TME-71-36 |
| 55 |
ENDF-142 |
C.L. Thompson, J.R. Stockton, L.M. Petrie, et al. |
EDITOR, A Processing code for ENDF/B Format Data |
Feb-71 |
ORNL-TM-3266 |
| 54 |
ENDF-141 |
L. Stewart |
Evaluated Nuclear Data for Hydrogen in the ENDF/B-II Format |
Feb-71 |
LA-4574 |
| 53 |
ENDF-140 |
D.J. Dudziak |
PHOXE: A Fortran-IV Code to Check Format Syntax, Consistency, and Physical Realism of ENDF/B Photon Production Data |
Sep-70 |
LA-4506-MS |
| 52 |
ENDF-139 |
S.K. Penny |
A Re-Evaluation of Natural Iron Neutron and Gamma-Ray Production Cross... |
Apr-71 |
ORNL-4617 |
| 51 |
ENDF-138 |
D.C. Irving |
Evaluation of the Cross Sections of Iron: ENDF/B MAT=1101 |
Sep-70 |
ORNL/TM-2891 |
| 50 |
ENDF-137 |
D.C. Irving |
LEGCK: A Subroutine to Analyze Legendre Coefficients for Negativity in the... |
Sep-70 |
ORNL/TM-2903 |
| 49 |
ENDF-136 |
T.A. Pitterle |
Evaluation of U-238 Neutron Cross Sections for the ENDF/B Version II File |
Mar-71 |
WARD-4181-1 |
| 48 |
ENDF-135 |
J.T. Reynolds |
Evaluated Neutron Cross Sections for the Zirconium Isotopes |
Mar-70 |
KAPL-M-7078 (Restrict) |
| 47 |
ENDF-134R |
R.Q. Wright, S.N. Cramer, D.C. Irving |
UKE-III: A Computer Program for Translating Neutron Cross Section Data From the UKAEA Nuclear Data Library ... |
Oct-73 |
ORNL-TM-2880 REV |
| 46 |
ENDF-134 |
R.Q. Wright, S.N. Cramer, D.C. Irving |
UKE-III: A Computer Program for Translating Neutron Cross Section ... |
Mar-70 |
ORNL-TM-2880 |
| 45 |
ENDF-133 |
S. Kellman |
Description of the Generation of Data Decks by ETOG-1 for Use in Creating... |
Jan-70 |
WCAP-3845-2 |
| 44 |
ENDF-132 |
D.J. Dudziak, A.H. Marshall, R.E. Seamon |
LAPH: A Multigroup Photon Production Matrix and Source Vector Code forENDF/B |
May-70 |
LA-4337 |
| 43 |
ENDF-131 |
N.M. Green |
An Evaluation and Compilation of the Fission and Capture Cross Sections of... |
Feb-70 |
ORNL/TM-2797 |
| 42 |
ENDF-130 |
D.J. Dudziak |
Translation to ENDF/B and "Physics" Checking of Cross Sections for Shielding |
Nov-69 |
DASA-2379 |
| 41 |
ENDF-129 |
N. Azziz |
Iron, Nickel, and Chromium Neutron Cross Sections from 0-15 MeV |
Aug-69 |
WCAP-7281 |
| 40 |
ENDF-128 |
D.J. Dudziak, J.M. Cook |
LUTE and LATEX, Special-Purpose Codes to Translate from Modified UK to ENDF/B Format |
Aug-69 |
NE-3383-102-69U |
| 39 |
ENDF-127 |
R.E. Schenter |
ETOX-A. Code to Calculate Group Constants for Nuclear Reactor Calculations |
May-69 |
BNWL-1002 |
| 38 |
ENDF-126 |
C.L. Dunford ,R.F. Berland, R.S. Hubner, R.J. Creasy |
SCORE II-An Interactive Neutron Evaluation System |
Mar-69 |
AI-AEC-12757 |
| 37 |
ENDF-125 |
E.M. Pennington |
ENDF/B Neutron Cross Section Data for Natural Helium |
Oct-68 |
ANL-7462 |
| 36 |
ENDF-124 |
J.M. Otter, R.S. Hubner, R.W. Campbell, et al. |
Evaluated Neutron Cross Sections forCu-63, Cu-65, and Natural Cu |
Dec-68 |
AI-AEC-12741 |
| 35 |
ENDF-123 |
J.T. Reynolds |
Evaluated Neutron Cross Sections for the Gadolinium Isotopes |
May-68 |
KAPL-3416 (Restrict) |
| 34 |
ENDF-122 |
T.A. Pitterle, M. Yamamoto |
Evaluated Neutron Cross Sections of Pu-240 for the ENDF/B File |
Jun-68 |
APDA-218 |
| 33 |
ENDF-121 |
T.A. Pitterle |
Evaluated Neutron Cross Sections of Sodium-23 for the ENDF/B File |
Jun-68 |
APDA-217 |
| 32 |
ENDF-120 |
D.M. Green, T.A. Pitterle |
ETOE-A Program for ENDF/B to MC2 Data Conversion |
Jun-68 |
APDA-219 |
| 31 |
ENDF-119 |
Void See ENDF-133 |
|
|
|
| 30 |
ENDF-118 |
Void See ENDF-133 |
|
|
|
| 29 |
ENDF-117 |
J.R. Smith |
Subject:Am-241,Am-243 |
|
|
| 28 |
ENDF-116 |
J.R. Smith, R.A. Grimesey |
An Evaluation and Compilation of Np-237 Cross Section Data for the ENDF/B File |
May-69 |
IN-1182 |
| 27 |
ENDF-115 |
W.B. Henderson |
Evaluation of Re-185 and Re-187 Neutron Cross Sections for ENDF/B |
Mar-68 |
GEMP-587 |
| 26 |
ENDF-114 Suppl. |
M. Raymund |
ETOG-1-A Fortran IV Program to Process Data from the ENDF/B File to the MUFT, GAM and ANISN Formats |
Aug-73 |
WCAP-3845-1 Suppl.1 |
| 25 |
ENDF-114 |
D.E. Kusner, S. Kellman |
ETOG-1-A Fortran IV Program to Process Data from the ENDF/B File to the MUFT,GAM and ANISN Formats |
Dec-69 |
WCAP-3845-1 |
| 24 |
ENDF-113 |
R.A. Dannels, D.E. Kusner |
ETOM-1-A Fortran IV Program to Process Data from the ENDF/B File to the MUFT Format |
May-68 |
WCAP-3688-1 |
| 23 |
ENDF-112 |
J.T. Reynolds, C.R. Lubitz |
Evaluated Cross Sections for the Hafnium Isotopes |
Aug-67 |
KAPL-3327 (Restrict) |
| 22 |
ENDF-111 |
D.J. Dudziak |
ENDF/B Format Requirements for Shielding Applications |
Apr-67 |
LA-3801 |
| 21 |
ENDF-110 |
O. Ozer |
Description of the ENDF/B Processing Codes and Retrieval Subroutines |
Jun-71 |
BNL-50300 |
| 20 |
ENDF-109 |
D.C. Irving |
Evaluation of Neutron Cross Sections for Boron-10 |
Oct-67 |
ORNL/TM-1872 |
| 19 |
ENDF-108 |
M.K. Drake |
Handwritten Notes of Be, U-234, U-236, Pu-241 for ENDF/B |
|
|
| 18 |
ENDF-107 |
Void-Not Used |
|
|
|
| 17 |
ENDF-106 |
C.L. Dunford, R.F. Berland, R.J. Creasy |
SCORE - An Automated Cross Section Evaluation System |
Jan-68 |
NAA-SR-MEMO-12529 |
| 16 |
ENDF-105 |
R.S. Hubner, B.J. Lemke |
EDIT-A Fortran IV level H Program to Punch, Print, and Plot Selected Portions of an ENDF/B Data Tape |
Nov-67 |
NAA-SR-12525 |
| 15 |
ENDF-104 |
W.A. Wittkopf |
Th-232 Neutron Cross Section Data for the ENDF/B |
|
|
| 14 |
ENDF-103 |
W.A. Wittkopf, D.H. Roy, A.Z. Livolsi |
U-238 Neutron Cross-Section Data for the ENDF/B |
May-67 |
BAW-316 |
| 13 (2023) |
ENDF-102 2023 |
Written by the Members of the Cross Sections Evaluation Working Group |
ENDF 102 Data Formats and Procedures for the Evaluated Nuclear Data Files ENDF/B-VI, ENDF/B-VII and ENDF/B-VIII |
Sep-28 |
BNL-224854-2023-INRE, Git Revision SHA1: 3576914 |
| 13 (2018) |
ENDF-102 2018 |
Written by the Members of the Cross Sections Evaluation Working Group |
ENDF 102 Data Formats and Procedures for the Evaluated Nuclear Data Files ENDF/B-VI, ENDF/B-VII and ENDF/B-VIII |
Feb-01 |
BNL-203218-2018-INRE, SVN Commit: Revision 215 |
| 13 (2012) |
ENDF-102 2012 |
A. Trkov, M. Herman, D.A. Brown, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF-VII |
Oct-12 |
BNL-90365-2009 Rev.2, SVN Commit 85 |
| 13 (2011) |
ENDF-102 2011 |
A. Trkov, M. Herman, D.A. Brown, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF-VII |
Dec-11 |
BNL-90365-2009 Rev.2 |
| 13 (2010) |
ENDF-102 2010 |
M. Herman, A. Trkov, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF-VII |
Jul-10 |
BNL-90365-2009 Rev.1 |
| 13 (2009) |
ENDF-102 2009 |
M. Herman, A. Trkov, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF-VII |
Jun-09 |
BNL-90365-2009 |
| 13 (2005) |
ENDF-102 2005 |
M. Herman, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-VII |
Jun-05 |
BNL-NCS-44945-05/06-Rev. |
| 13 (2004) |
ENDF-102 2004 |
V. McLane, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Apr-04 |
BNL-NCS-44945-04/04-Rev. |
| 13 (2001) |
ENDF-102 2001 |
V. McLane, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Apr-01 |
BNL-NCS-44945-01/04-Rev. |
| 13 (1998) |
ENDF-102 1998 |
V. McLane, Ed. |
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
May-98 |
BNL-NCS-44945-98/05-Rev. |
| 12 |
ENDF-102 1997 |
V. McLane, C.L. Dunford, P.F. Rose |
Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Feb-97 |
BNL-NCS-44945 REV.2/97 |
| 11 |
ENDF-102 1995 |
V. McLane, C.L. Dunford, P.F. Rose |
Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Nov-95 |
BNL-NCS-44945 REV.11/95 |
| 10 |
ENDF-102 1991 |
P.F. Rose, C.L. Dunford |
Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Oct-91 |
BNL-NCS-44945 REV |
| 9 |
ENDF-102 1990 |
P.F. Rose, C.L. Dunford |
Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6 |
Jul-90 |
BNL-NCS-44945 |
| 8 |
ENDF-102 1983 |
B.A. Magurno |
Data Formats and Procedures for the Evaluated Nuclear Data File,ENDF.V-V |
Nov-83 |
BNL-NCS-50496 2 Ed. Rev. |
| 7 |
ENDF-102 1980 |
S. Pearlstein |
Supp. to the ENDF/B-V Formats and Procedures Manual for Using ENDF/B-IV ... |
Nov-80 |
BNL-NCS-28949 2 Ed. Suppl. |
| 6 |
ENDF-102 1979 |
R.R. Kinsey |
Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF/B-V |
Oct-79 |
BNL-NCS-50496 3 Ed. |
| 5 |
ENDF-102 1975 |
D.I. Garber |
Data Formats and Procedures for Evaluated Nuclear Data File |
Oct-75 |
BNL-NCS-50496 2 Ed. |
| 4 |
ENDF-102 1970 Vol.2 |
D.J. Dudziak |
ENDF Formats and Procedures for Photon Production and Interaction Data |
Jul-71 |
LA-4549 1st Ed. Vol.2 |
| 3 |
ENDF-102 1970 Vol.1 |
M.K. Drake |
Data Formats and Procedures for the ENDF Neutron Cross Section Library |
Oct-70 |
BNL-50274 1st Ed. Vol.1 |
| 2 |
ENDF-102 1966 |
H.C. Honeck |
Specifications for an Evaluated Nuclear Data File for REACTOR APPLICATION... |
May-66 |
BNL-50066 1st Ed. |
| 1 |
ENDF-101 |
T.E. Stephenson, A. Prince, S. Pearlstein |
Evaluation of the Neutron Cross Section of Manganese for the ENDF/B Library |
Jun-67 |
BNL-50060 |
| 0 |
BNL-8381 |
H.C. Honeck |
ENDF: Evaluated Nuclear Data File Description and Specifications |
Jan-65 |
BNL-8381 |
