TENDL-2011 gen. purp. file: proton + C - 13 1 MeV - 200 MeV 99 0 0 0 6013.00000 12.8916500 -1 0 0 1 628 1451 1 0.0 0.0 0 0 0 6 628 1451 2 .998620000 200000000. 1 0 10010 0 628 1451 3 0.0 0.0 1 0 271 6 628 1451 4 6-C - 13 NRG EVAL-OCT11 A.J. Koning 628 1451 5 TENDL-2011 DIST- REV1- 628 1451 6 ----TENDL-2011 Material 628 REVISION 1 628 1451 7 -----Incident proton data 628 1451 8 ------ENDF-6 Format 628 1451 9 628 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 628 1451 11 628 1451 12 Charged particle transport library made by TALYS 628 1451 13 628 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 628 1451 15 628 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 628 1451 17 628 1451 18 This evaluated data file is based primarily on the nuclear model 628 1451 19 code TALYS [kon07], version 1.4. 628 1451 20 It is part of a large collection of isotopic 628 1451 21 evaluations, all created by running TALYS with either default or 628 1451 22 adjusted input parameters. This means that the data in this 628 1451 23 evaluation have been tested in detail for some isotopes against 628 1451 24 individual experimental data, while for other isotopes it is 628 1451 25 only as good as the global quality of TALYS at the present moment. 628 1451 26 The mutual quality of all individual evaluations is however 628 1451 27 consistent: The same set of nuclear models is used and, equally 628 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 628 1451 29 The resulting data file provides a complete representation of 628 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 628 1451 31 and shielding applications over the incident projectile energy 628 1451 32 range from 1.0E-11 to 200 MeV. 628 1451 33 628 1451 34 All transport data for neutrons, other particles, photons and 628 1451 35 residual nuclides are filed using a combination of MF1,3, 628 1451 36 and MF6. This includes cross sections, angular distributions, 628 1451 37 double-differential spectra, production cross sections, 628 1451 38 residual production (activation) cross sections and recoils. 628 1451 39 This evaluation can thus be used as both transport and activation 628 1451 40 library. The data file has been created automatically using the 628 1451 41 ENDF-6 format generator TEFAL. 628 1451 42 628 1451 43 ##### ORIGIN 628 1451 44 628 1451 45 All data < 200 MeV : Produced with TALYS code 628 1451 46 628 1451 47 *************************** T H E O R Y ************************** 628 1451 48 628 1451 49 TALYS is a computer code system for the prediction and analysis 628 1451 50 of nuclear reactions. TALYS simulates reactions that involve 628 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 628 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 628 1451 53 target nuclides of mass 12 and heavier. This is achieved by 628 1451 54 implementing a suite of nuclear reaction models into a single 628 1451 55 code system. It enables to evaluate nuclear reactions from 628 1451 56 the unresolved resonance region up to intermediate energies. This 628 1451 57 evaluation is based on a theoretical analysis that utilizes the 628 1451 58 optical model, compound nucleus statistical theory, fission, 628 1451 59 direct reactions and pre-equilibrium processes, in combination 628 1451 60 with databases and models for nuclear structure. The following 628 1451 61 output of TALYS is stored in this data file: 628 1451 62 628 1451 63 - Elastic and non-elastic cross sections 628 1451 64 - Elastic scattering angular distributions 628 1451 65 - Residual production cross sections 628 1451 66 - Recoil data 628 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 628 1451 68 - Total particle energy spectra 628 1451 69 - Total particle double-differential spectra 628 1451 70 628 1451 71 Here follows a short description of the used nuclear models: 628 1451 72 628 1451 73 ##### OPTICAL MODEL 628 1451 74 628 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 628 1451 76 in TALYS used as a subroutine. The default optical model 628 1451 77 potentials (OMP) used are the local and global parameterizations 628 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 628 1451 79 for neutrons and protons which in principle are valid over the 628 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 628 1451 81 validity may differ from nucleus to nucleus. For neutrons and 628 1451 82 protons, the used parameterization is given in Eq. (7) of 628 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 628 1451 84 used in the case of deformed nuclides. 628 1451 85 To calculate the transmission coefficients and reaction 628 1451 86 cross sections for deuterons, tritons, helions and alpha 628 1451 87 particles, we use OMPs that are directly derived from our nucleon 628 1451 88 potentials using Watanabe's folding approach. 628 1451 89 628 1451 90 ##### DIRECT REACTIONS 628 1451 91 628 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 628 1451 93 calculations for rotational or vibrational (or a combination of 628 1451 94 these) nuclides, depending on the information present in the 628 1451 95 nuclear structure database of TALYS. 628 1451 96 In addition, a macroscopic, phenomenological model to describe 628 1451 97 giant resonances in the inelastic channel is used. For each 628 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 628 1451 99 calculation with ECIS-06 is performed for each giant resonance 628 1451 100 state. The cross section is then spread over the continuum with a 628 1451 101 Gaussian distribution. 628 1451 102 628 1451 103 ##### COMPOUND NUCLEUS 628 1451 104 628 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 628 1451 106 model. The transmission coefficients have been generated 628 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 628 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 628 1451 109 For each nucleus that can be reached through a binary reaction, 628 1451 110 several discrete levels and a continuum described by level 628 1451 111 densities are included simultaneously as competing channels. 628 1451 112 628 1451 113 For each residual nucleus several discrete states 628 1451 114 are included as well as a continuum described by level densities. 628 1451 115 Multiple compound emission is continued until all reaction 628 1451 116 channels are closed and the population distribution of all 628 1451 117 residual nuclides is depleted, through gamma decay, until they 628 1451 118 end up in the ground state or in an isomer. 628 1451 119 628 1451 120 For the level density, we take the Constant Temperature Model 628 1451 121 with level density parameters as given in [kon08]. This employs 628 1451 122 the temperature law at low energies and a Fermi gas expression at 628 1451 123 high energies, and takes into account the damping of shell 628 1451 124 effects at high excitation energy. We have obtained the level 628 1451 125 density parameters from a simultaneous fit to all experimental 628 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 628 1451 127 628 1451 128 Gamma-ray transmission coefficients are generated with the 628 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 628 1451 130 giant dipole resonance parameters taken from the RIPL library 628 1451 131 [rip09], and normalized with experimental radiative widths. 628 1451 132 628 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 628 1451 134 628 1451 135 For pre-equilibrium reactions, which become important for 628 1451 136 incident energies above about 10 MeV, we use the two-component 628 1451 137 exciton model [kon04], in which the neutron or proton types of 628 1451 138 particles and holes are followed throughout the reaction. For 628 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 628 1451 140 any order of particle emission was included in the calculations. 628 1451 141 A parameterization for the squared matrix element is used that is 628 1451 142 valid for the whole energy range of this evaluation. 628 1451 143 628 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 628 1451 145 contribution was added from the pick/up and knock-out reaction 628 1451 146 model by Kalbach [kal05]. 628 1451 147 628 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 628 1451 149 applied, to simulate the direct and semi-direct capture 628 1451 150 processes. 628 1451 151 628 1451 152 The angular distribution systematics by Kalbach [kal88] were used 628 1451 153 to describe the angular distributions for all continuum 628 1451 154 particles. An isotopic distribution for photons was adopted. 628 1451 155 628 1451 156 ***************** F I L E I N F O R M A T I O N **************** 628 1451 157 628 1451 158 ##### MF1: GENERAL INFORMATION 628 1451 159 628 1451 160 - MT451 : Descriptive data and directory 628 1451 161 628 1451 162 This text and the full directory of used MF/MT sections. 628 1451 163 628 1451 164 ##### MF3: REACTION CROSS SECTIONS 628 1451 165 628 1451 166 All the data present in the following MT-sections have been 628 1451 167 calculated with TALYS. If the maximal cross section in an 628 1451 168 excitation function over the whole energy range does not exceed 628 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 628 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 628 1451 171 are set to zero. The minor reaction channels are also present 628 1451 172 but they are stored in MT5 and can be reproduced in combination 628 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 628 1451 174 The following reaction channels/MT numbers are generally included: 628 1451 175 628 1451 176 - MT2 : Elastic scattering cross section: nuclear + 628 1451 177 interference terms 628 1451 178 628 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 628 1451 180 distributions of MF=6. Note that because of the interference 628 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 628 1451 182 some energies and angles. 628 1451 183 628 1451 184 - MT5 : (p,anything) cross section 628 1451 185 628 1451 186 MT5 contains the reactions which can not be stored in any other 628 1451 187 MT-number. As the incident energy increases, the cross section 628 1451 188 of MT5 increases as more reaction channels are no longer stored 628 1451 189 under a specific MT number. The information of MF3/MT5 can be 628 1451 190 combined with MF6/MT5 to obtain residual production cross 628 1451 191 sections particle production cross sections and 628 1451 192 (double-)differential cross sections. 628 1451 193 628 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 628 1451 195 628 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 628 1451 197 distributions, as well as all residual and photon production 628 1451 198 cross sections. All data are generated with TALYS. 628 1451 199 628 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 628 1451 201 interference terms 628 1451 202 628 1451 203 Relative angular distributions are tabulated on an angular grid. 628 1451 204 They are obtained by using the "nuclear-plus-interference" option 628 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 628 1451 206 integrated cross section is stored in MF=3. Note that because 628 1451 207 of the interference effect, the tabulations in both MF=6 and 628 1451 208 MF=3 can be negative at some energies and angles. 628 1451 209 628 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 628 1451 211 628 1451 212 MT5 contains the production yields of particles and residual 628 1451 213 products. It also contains the secondary energy-angle 628 1451 214 distributions for all particles and photons. First, the yields 628 1451 215 for neutrons are given for the whole energy range. Next, on a 628 1451 216 secondary energy grid the relative emission spectra are given 628 1451 217 together with the parameters for the Kalbach systematics for 628 1451 218 angular distributions. Inelastic scattering cross sections for 628 1451 219 discrete states have been broadened and added to the continuum 628 1451 220 spectra. This procedure is repeated for protons, deuterons, 628 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 628 1451 222 residual production yields are given per final product. All these 628 1451 223 yields and relative distributions can be multiplied with the 628 1451 224 cross sections given in MF3/MT5 to get the production cross 628 1451 225 sections and (double-)differential cross sections. 628 1451 226 628 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 **** 628 1451 228 628 1451 229 This file has been checked successfully by the BNL checking 628 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 628 1451 231 been processed successfully into an MCNP library by the 628 1451 232 processing code NJOY99.90 [mac00]. 628 1451 233 628 1451 234 *********************** R E F E R E N C E S ********************** 628 1451 235 628 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 628 1451 237 (1985). 628 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 628 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 628 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 628 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 628 1451 242 (2003). 628 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 628 1451 244 (2004). 628 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 628 1451 246 Proceedings of the International Conference on Nuclear 628 1451 247 Data for Science and Technology - ND-2007, 628 1451 248 April 22-27, 2007, Nice, France 628 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 628 1451 250 local level density models'', Nucl Phys. A810, 628 1451 251 13-76 (2008). 628 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 628 1451 253 pointwise and multigroup neutron and photon cross 628 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 628 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 628 1451 256 No. CEA-N-2772, 1994. 628 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 628 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 628 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 628 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 628 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 628 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 628 1451 263 RIPL - Reference Input Parameter Library for calculation 628 1451 264 of nuclear reactions and nuclear data evaluation, 628 1451 265 Nucl. Data Sheets 110, 3107 (2009). 628 1451 266 628 1451 267 ************************* C O N T E N T S ************************ 628 1451 268 628 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 628 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 628 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 628 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 628 1451 273 Unshielded Group Averages Using 162 Groups 628 1451 274 Weighting Spectrum: 1/E Spectrum 628 1451 275 1 451 275 1 628 1451 276 3 2 18 1 628 1451 277 3 5 18 1 628 1451 278 6 2 1401 1 628 1451 279 6 5 12403 1 628 1451 280 8 5 32 1 628 1451 281 628 1 0 282 628 0 0 283 6013.00000 12.8916500 0 0 0 0 628 3 2 284 0.0 0.0 0 0 1 52 628 3 2 285 52 1 628 3 2 286 10000000.0 .004714347 11000000.0 .015872510 12000000.0 .025391409 628 3 2 287 13000000.0 .033685704 14000000.0 .040966730 15000000.0 .047341731 628 3 2 288 16000000.0 .052881993 17000000.0 .057651825 18000000.0 .061718972 628 3 2 289 19000000.0 .065154393 20000000.0 .069076363 22000000.0 .073041535 628 3 2 290 24000000.0 .075576669 26000000.0 .077014026 28000000.0 .077597558 628 3 2 291 30000000.0 .076818363 35000000.0 .073986170 40000000.0 .069572999 628 3 2 292 45000000.0 .064418945 50000000.0 .059562083 54000000.0 .056839752 628 3 2 293 55000000.0 .053736708 60000000.0 .046429414 70000000.0 .037782973 628 3 2 294 80000000.0 .030785486 90000000.0 .025103740 100000000. .018936468 628 3 2 295 120000000. .013317530 140000000. .009915877 160000000. .007912360 628 3 2 296 180000000. .006656700 200000000. .006178382 240000000. .006178382 628 3 2 297 280000000. .006178382 320000000. .006178382 360000000. .006178382 628 3 2 298 400000000. .006178382 440000000. .006178382 480000000. .006178382 628 3 2 299 520000000. .006178382 560000000. .006178382 600000000. .006178382 628 3 2 300 640000000. .006178382 680000000. .006178382 720000000. .006178382 628 3 2 301 760000000. .006178382 800000000. .006178382 840000000. .006178382 628 3 2 302 880000000. .006178382 920000000. .006178382 960000000. .006178382 628 3 2 303 1.00000E+9 0.0 628 3 2 304 628 3 0 305 6013.00000 12.8916500 0 0 0 0 628 3 5 306 0.0 0.0 0 0 1 106 628 3 5 307 106 1 628 3 5 308 1000000.00 1.32164E-6 1125000.00 1.34073E-6 1250000.00 1.35980E-6 628 3 5 309 1375000.00 1.37888E-6 1500000.00 1.39795E-6 1625000.00 1.41703E-6 628 3 5 310 1750000.00 1.43610E-6 1875000.00 1.45517E-6 2000000.00 1.52145E-6 628 3 5 311 2125000.00 1.63593E-6 2250000.00 1.75041E-6 2375000.00 1.86489E-6 628 3 5 312 2500000.00 1.97936E-6 2625000.00 2.09383E-6 2750000.00 2.20830E-6 628 3 5 313 2875000.00 2.32277E-6 3000000.00 .014333438 3125000.00 .043195741 628 3 5 314 3250000.00 .072057753 3375000.00 .100919506 3500000.00 .129781028 628 3 5 315 3625000.00 .158642342 3750000.00 .187503467 3875000.00 .216364424 628 3 5 316 4000000.00 .251644688 4200000.00 .293542612 4400000.00 .335439831 628 3 5 317 4600000.00 .377336436 4800000.00 .419232501 5000000.00 .450211925 628 3 5 318 5200000.00 .470128401 5400000.00 .490044699 5600000.00 .509960837 628 3 5 319 5800000.00 .529876831 6000000.00 .545941836 6200000.00 .558112872 628 3 5 320 6400000.00 .570283843 6600000.00 .582454755 6800000.00 .594625612 628 3 5 321 7000000.00 .604579886 7200000.00 .612296383 7400000.00 .620012853 628 3 5 322 7600000.00 .627729298 7800000.00 .635445721 8000000.00 .641184296 628 3 5 323 8200000.00 .644928482 8400000.00 .648672659 8600000.00 .652416827 628 3 5 324 8800000.00 .656160989 9000000.00 .658977563 9200000.00 .660859657 628 3 5 325 9400000.00 .662741748 9600000.00 .664623836 9800000.00 .666505922 628 3 5 326 10000000.0 .669900503 11000000.0 .674435492 12000000.0 .682005568 628 3 5 327 13000000.0 .686884744 14000000.0 .684276149 15000000.0 .679437478 628 3 5 328 16000000.0 .672544303 17000000.0 .664197533 18000000.0 .654987806 628 3 5 329 19000000.0 .645381596 20000000.0 .630983267 22000000.0 .612034561 628 3 5 330 24000000.0 .593917650 26000000.0 .576693668 28000000.0 .560301919 628 3 5 331 30000000.0 .534312077 35000000.0 .499576480 40000000.0 .468950651 628 3 5 332 45000000.0 .442015777 50000000.0 .420477583 54000000.0 .409300771 628 3 5 333 55000000.0 .397456593 60000000.0 .371186464 70000000.0 .341994446 628 3 5 334 80000000.0 .319077442 90000000.0 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