TENDL-2011 gen. purp. file: proton + O - 16 1 MeV - 200 MeV 99 0 0 0 8016.00000 15.8575100 -1 0 0 1 825 1451 1 0.0 0.0 0 0 0 6 825 1451 2 .998620000 200000000. 1 0 10010 0 825 1451 3 0.0 0.0 1 0 271 6 825 1451 4 8-O - 16 NRG EVAL-OCT11 A.J. Koning 825 1451 5 TENDL-2011 DIST- REV1- 825 1451 6 ----TENDL-2011 Material 825 REVISION 1 825 1451 7 -----Incident proton data 825 1451 8 ------ENDF-6 Format 825 1451 9 825 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 825 1451 11 825 1451 12 Charged particle transport library made by TALYS 825 1451 13 825 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 825 1451 15 825 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 825 1451 17 825 1451 18 This evaluated data file is based primarily on the nuclear model 825 1451 19 code TALYS [kon07], version 1.4. 825 1451 20 It is part of a large collection of isotopic 825 1451 21 evaluations, all created by running TALYS with either default or 825 1451 22 adjusted input parameters. This means that the data in this 825 1451 23 evaluation have been tested in detail for some isotopes against 825 1451 24 individual experimental data, while for other isotopes it is 825 1451 25 only as good as the global quality of TALYS at the present moment. 825 1451 26 The mutual quality of all individual evaluations is however 825 1451 27 consistent: The same set of nuclear models is used and, equally 825 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 825 1451 29 The resulting data file provides a complete representation of 825 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 825 1451 31 and shielding applications over the incident projectile energy 825 1451 32 range from 1.0E-11 to 200 MeV. 825 1451 33 825 1451 34 All transport data for neutrons, other particles, photons and 825 1451 35 residual nuclides are filed using a combination of MF1,3, 825 1451 36 and MF6. This includes cross sections, angular distributions, 825 1451 37 double-differential spectra, production cross sections, 825 1451 38 residual production (activation) cross sections and recoils. 825 1451 39 This evaluation can thus be used as both transport and activation 825 1451 40 library. The data file has been created automatically using the 825 1451 41 ENDF-6 format generator TEFAL. 825 1451 42 825 1451 43 ##### ORIGIN 825 1451 44 825 1451 45 All data < 200 MeV : Produced with TALYS code 825 1451 46 825 1451 47 *************************** T H E O R Y ************************** 825 1451 48 825 1451 49 TALYS is a computer code system for the prediction and analysis 825 1451 50 of nuclear reactions. TALYS simulates reactions that involve 825 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 825 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 825 1451 53 target nuclides of mass 12 and heavier. This is achieved by 825 1451 54 implementing a suite of nuclear reaction models into a single 825 1451 55 code system. It enables to evaluate nuclear reactions from 825 1451 56 the unresolved resonance region up to intermediate energies. This 825 1451 57 evaluation is based on a theoretical analysis that utilizes the 825 1451 58 optical model, compound nucleus statistical theory, fission, 825 1451 59 direct reactions and pre-equilibrium processes, in combination 825 1451 60 with databases and models for nuclear structure. The following 825 1451 61 output of TALYS is stored in this data file: 825 1451 62 825 1451 63 - Elastic and non-elastic cross sections 825 1451 64 - Elastic scattering angular distributions 825 1451 65 - Residual production cross sections 825 1451 66 - Recoil data 825 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 825 1451 68 - Total particle energy spectra 825 1451 69 - Total particle double-differential spectra 825 1451 70 825 1451 71 Here follows a short description of the used nuclear models: 825 1451 72 825 1451 73 ##### OPTICAL MODEL 825 1451 74 825 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 825 1451 76 in TALYS used as a subroutine. The default optical model 825 1451 77 potentials (OMP) used are the local and global parameterizations 825 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 825 1451 79 for neutrons and protons which in principle are valid over the 825 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 825 1451 81 validity may differ from nucleus to nucleus. For neutrons and 825 1451 82 protons, the used parameterization is given in Eq. (7) of 825 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 825 1451 84 used in the case of deformed nuclides. 825 1451 85 To calculate the transmission coefficients and reaction 825 1451 86 cross sections for deuterons, tritons, helions and alpha 825 1451 87 particles, we use OMPs that are directly derived from our nucleon 825 1451 88 potentials using Watanabe's folding approach. 825 1451 89 825 1451 90 ##### DIRECT REACTIONS 825 1451 91 825 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 825 1451 93 calculations for rotational or vibrational (or a combination of 825 1451 94 these) nuclides, depending on the information present in the 825 1451 95 nuclear structure database of TALYS. 825 1451 96 In addition, a macroscopic, phenomenological model to describe 825 1451 97 giant resonances in the inelastic channel is used. For each 825 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 825 1451 99 calculation with ECIS-06 is performed for each giant resonance 825 1451 100 state. The cross section is then spread over the continuum with a 825 1451 101 Gaussian distribution. 825 1451 102 825 1451 103 ##### COMPOUND NUCLEUS 825 1451 104 825 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 825 1451 106 model. The transmission coefficients have been generated 825 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 825 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 825 1451 109 For each nucleus that can be reached through a binary reaction, 825 1451 110 several discrete levels and a continuum described by level 825 1451 111 densities are included simultaneously as competing channels. 825 1451 112 825 1451 113 For each residual nucleus several discrete states 825 1451 114 are included as well as a continuum described by level densities. 825 1451 115 Multiple compound emission is continued until all reaction 825 1451 116 channels are closed and the population distribution of all 825 1451 117 residual nuclides is depleted, through gamma decay, until they 825 1451 118 end up in the ground state or in an isomer. 825 1451 119 825 1451 120 For the level density, we take the Constant Temperature Model 825 1451 121 with level density parameters as given in [kon08]. This employs 825 1451 122 the temperature law at low energies and a Fermi gas expression at 825 1451 123 high energies, and takes into account the damping of shell 825 1451 124 effects at high excitation energy. We have obtained the level 825 1451 125 density parameters from a simultaneous fit to all experimental 825 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 825 1451 127 825 1451 128 Gamma-ray transmission coefficients are generated with the 825 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 825 1451 130 giant dipole resonance parameters taken from the RIPL library 825 1451 131 [rip09], and normalized with experimental radiative widths. 825 1451 132 825 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 825 1451 134 825 1451 135 For pre-equilibrium reactions, which become important for 825 1451 136 incident energies above about 10 MeV, we use the two-component 825 1451 137 exciton model [kon04], in which the neutron or proton types of 825 1451 138 particles and holes are followed throughout the reaction. For 825 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 825 1451 140 any order of particle emission was included in the calculations. 825 1451 141 A parameterization for the squared matrix element is used that is 825 1451 142 valid for the whole energy range of this evaluation. 825 1451 143 825 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 825 1451 145 contribution was added from the pick/up and knock-out reaction 825 1451 146 model by Kalbach [kal05]. 825 1451 147 825 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 825 1451 149 applied, to simulate the direct and semi-direct capture 825 1451 150 processes. 825 1451 151 825 1451 152 The angular distribution systematics by Kalbach [kal88] were used 825 1451 153 to describe the angular distributions for all continuum 825 1451 154 particles. An isotopic distribution for photons was adopted. 825 1451 155 825 1451 156 ***************** F I L E I N F O R M A T I O N **************** 825 1451 157 825 1451 158 ##### MF1: GENERAL INFORMATION 825 1451 159 825 1451 160 - MT451 : Descriptive data and directory 825 1451 161 825 1451 162 This text and the full directory of used MF/MT sections. 825 1451 163 825 1451 164 ##### MF3: REACTION CROSS SECTIONS 825 1451 165 825 1451 166 All the data present in the following MT-sections have been 825 1451 167 calculated with TALYS. If the maximal cross section in an 825 1451 168 excitation function over the whole energy range does not exceed 825 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 825 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 825 1451 171 are set to zero. The minor reaction channels are also present 825 1451 172 but they are stored in MT5 and can be reproduced in combination 825 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 825 1451 174 The following reaction channels/MT numbers are generally included: 825 1451 175 825 1451 176 - MT2 : Elastic scattering cross section: nuclear + 825 1451 177 interference terms 825 1451 178 825 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 825 1451 180 distributions of MF=6. Note that because of the interference 825 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 825 1451 182 some energies and angles. 825 1451 183 825 1451 184 - MT5 : (p,anything) cross section 825 1451 185 825 1451 186 MT5 contains the reactions which can not be stored in any other 825 1451 187 MT-number. As the incident energy increases, the cross section 825 1451 188 of MT5 increases as more reaction channels are no longer stored 825 1451 189 under a specific MT number. The information of MF3/MT5 can be 825 1451 190 combined with MF6/MT5 to obtain residual production cross 825 1451 191 sections particle production cross sections and 825 1451 192 (double-)differential cross sections. 825 1451 193 825 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 825 1451 195 825 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 825 1451 197 distributions, as well as all residual and photon production 825 1451 198 cross sections. All data are generated with TALYS. 825 1451 199 825 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 825 1451 201 interference terms 825 1451 202 825 1451 203 Relative angular distributions are tabulated on an angular grid. 825 1451 204 They are obtained by using the "nuclear-plus-interference" option 825 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 825 1451 206 integrated cross section is stored in MF=3. Note that because 825 1451 207 of the interference effect, the tabulations in both MF=6 and 825 1451 208 MF=3 can be negative at some energies and angles. 825 1451 209 825 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 825 1451 211 825 1451 212 MT5 contains the production yields of particles and residual 825 1451 213 products. It also contains the secondary energy-angle 825 1451 214 distributions for all particles and photons. First, the yields 825 1451 215 for neutrons are given for the whole energy range. Next, on a 825 1451 216 secondary energy grid the relative emission spectra are given 825 1451 217 together with the parameters for the Kalbach systematics for 825 1451 218 angular distributions. Inelastic scattering cross sections for 825 1451 219 discrete states have been broadened and added to the continuum 825 1451 220 spectra. This procedure is repeated for protons, deuterons, 825 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 825 1451 222 residual production yields are given per final product. All these 825 1451 223 yields and relative distributions can be multiplied with the 825 1451 224 cross sections given in MF3/MT5 to get the production cross 825 1451 225 sections and (double-)differential cross sections. 825 1451 226 825 1451 227 ***** F I L E C H E C K I N G A N D P R O C E S S I N G **** 825 1451 228 825 1451 229 This file has been checked successfully by the BNL checking 825 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 825 1451 231 been processed successfully into an MCNP library by the 825 1451 232 processing code NJOY99.90 [mac00]. 825 1451 233 825 1451 234 *********************** R E F E R E N C E S ********************** 825 1451 235 825 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 825 1451 237 (1985). 825 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 825 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 825 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 825 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 825 1451 242 (2003). 825 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 825 1451 244 (2004). 825 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 825 1451 246 Proceedings of the International Conference on Nuclear 825 1451 247 Data for Science and Technology - ND-2007, 825 1451 248 April 22-27, 2007, Nice, France 825 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 825 1451 250 local level density models'', Nucl Phys. A810, 825 1451 251 13-76 (2008). 825 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 825 1451 253 pointwise and multigroup neutron and photon cross 825 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 825 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 825 1451 256 No. CEA-N-2772, 1994. 825 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 825 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 825 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 825 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 825 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 825 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 825 1451 263 RIPL - Reference Input Parameter Library for calculation 825 1451 264 of nuclear reactions and nuclear data evaluation, 825 1451 265 Nucl. Data Sheets 110, 3107 (2009). 825 1451 266 825 1451 267 ************************* C O N T E N T S ************************ 825 1451 268 825 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 825 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 825 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 825 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 825 1451 273 Unshielded Group Averages Using 162 Groups 825 1451 274 Weighting Spectrum: 1/E Spectrum 825 1451 275 1 451 275 1 825 1451 276 3 2 18 1 825 1451 277 3 5 18 1 825 1451 278 6 2 1401 1 825 1451 279 6 5 12085 1 825 1451 280 8 5 47 1 825 1451 281 825 1 0 282 825 0 0 283 8016.00000 15.8575100 0 0 0 0 825 3 2 284 0.0 0.0 0 0 1 48 825 3 2 285 48 1 825 3 2 286 14000000.0-.003105701 15000000.0 .007840400 16000000.0 .017367721 825 3 2 287 17000000.0 .025601885 18000000.0 .032693806 19000000.0 .038794574 825 3 2 288 20000000.0 .046082319 22000000.0 .053965866 24000000.0 .059768882 825 3 2 289 26000000.0 .063952848 28000000.0 .066842081 30000000.0 .068997696 825 3 2 290 35000000.0 .069325425 40000000.0 .066907484 45000000.0 .063011645 825 3 2 291 50000000.0 .058882415 54000000.0 .056461660 55000000.0 .053595420 825 3 2 292 60000000.0 .046649209 70000000.0 .038176927 80000000.0 .031187435 825 3 2 293 90000000.0 .025470302 100000000. .019240118 120000000. .013574352 825 3 2 294 140000000. .010174450 160000000. .008197561 180000000. .006973173 825 3 2 295 200000000. .006511956 240000000. .006511956 280000000. .006511956 825 3 2 296 320000000. .006511956 360000000. .006511956 400000000. .006511956 825 3 2 297 440000000. .006511956 480000000. .006511956 520000000. .006511956 825 3 2 298 560000000. .006511956 600000000. .006511956 640000000. .006511956 825 3 2 299 680000000. .006511956 720000000. .006511956 760000000. .006511956 825 3 2 300 800000000. .006511956 840000000. .006511956 880000000. .006511956 825 3 2 301 920000000. .006511956 960000000. .006511956 1.00000E+9 0.0 825 3 2 302 825 3 0 303 8016.00000 15.8575100 0 0 0 0 825 3 5 304 0.0 0.0 0 0 1 106 825 3 5 305 106 1 825 3 5 306 1000000.00 1.86994E-9 1125000.00 5.68864E-9 1250000.00 9.50658E-9 825 3 5 307 1375000.00 1.33240E-8 1500000.00 1.71409E-8 1625000.00 2.09576E-8 825 3 5 308 1750000.00 2.47739E-8 1875000.00 2.85901E-8 2000000.00 3.24060E-8 825 3 5 309 2125000.00 3.62218E-8 2250000.00 4.00374E-8 2375000.00 4.38529E-8 825 3 5 310 2500000.00 4.76683E-8 2625000.00 5.14837E-8 2750000.00 5.52989E-8 825 3 5 311 2875000.00 5.91142E-8 3000000.00 6.48237E-8 3125000.00 7.24541E-8 825 3 5 312 3250000.00 8.00844E-8 3375000.00 8.77147E-8 3500000.00 9.53449E-8 825 3 5 313 3625000.00 1.02975E-7 3750000.00 1.10605E-7 3875000.00 1.18235E-7 825 3 5 314 4000000.00 1.46285E-7 4200000.00 1.95123E-7 4400000.00 2.43959E-7 825 3 5 315 4600000.00 2.92795E-7 4800000.00 3.41630E-7 5000000.00 3.96529E-7 825 3 5 316 5200000.00 4.57572E-7 5400000.00 5.18615E-7 5600000.00 5.79658E-7 825 3 5 317 5800000.00 6.40700E-7 6000000.00 .001543526 6200000.00 .004646461 825 3 5 318 6400000.00 .007749380 6600000.00 .010852283 6800000.00 .013955172 825 3 5 319 7000000.00 .038570836 7200000.00 .084905016 7400000.00 .131239037 825 3 5 320 7600000.00 .177572912 7800000.00 .223906651 8000000.00 .271600045 825 3 5 321 8200000.00 .320664478 8400000.00 .369728796 8600000.00 .418793008 825 3 5 322 8800000.00 .467857121 9000000.00 .500330138 9200000.00 .516088759 825 3 5 323 9400000.00 .531847354 9600000.00 .547605924 9800000.00 .563364471 825 3 5 324 10000000.0 .596118964 11000000.0 .640198849 12000000.0 .669806205 825 3 5 325 13000000.0 .685372003 14000000.0 .692062572 15000000.0 .695415050 825 3 5 326 16000000.0 .695607377 17000000.0 .691547149 18000000.0 .684993430 825 3 5 327 19000000.0 .677135476 20000000.0 .664491511 22000000.0 .647373396 825 3 5 328 24000000.0 .630555697 26000000.0 .614184655 28000000.0 .598284971 825 3 5 329 30000000.0 .572583818 35000000.0 .537976429 40000000.0 .507331044 825 3 5 330 45000000.0 .480351216 50000000.0 .458763891 54000000.0 .447556789 825 3 5 331 55000000.0 .435677028 60000000.0 .409327748 70000000.0 .380023391 825 3 5 332 80000000.0 .356954110 90000000.0 .338038531 100000000. .316434836 825 3 5 333 120000000. .294535590 140000000. .278569100 160000000. .266684018 825 3 5 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