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ENDF Reports 

The NNDC is responsible for assigning ENDF reference numbers for all formal laboratory reports associated with the ENDF system.
Please contact the CSEWG chair or the ENDF library manager for an ENDF report number.

# Report Number Authors Title Date Cross Reference
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 ORELA Flight Path 1: Determinations of Its Effective Length vs Energy... 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 Neutron Cross Section Libraries in the AMPX Master Interface... 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-SectionEvaluations 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 ReactivityMeasurements 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 NeutronCross Sections for U-235 from... 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 V(n,x gamma) Reaction Cross Section for Incident Neutron Energies... 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 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... Oct-72 ORNL/TM-3957
89 ENDF-175 P.G. Young A Preliminary Evaluation of the Neutron and Photon-Production Cross... Dec-72 LA-4726
88 ENDF-174 D.G. Foster A Preliminary Evaluation of the Neutron and Photon Production Cross... 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 (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

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