TENDL-2011 gen. purp. file: deuteron + C - 12 1 MeV - 200 MeV 99 0 0 0 6012.00000 11.8969200 -1 0 0 1 625 1451 1 0.0 0.0 0 0 0 6 625 1451 2 1.99626000 200000000. 1 0 10020 0 625 1451 3 0.0 0.0 1 0 271 6 625 1451 4 6-C - 12 NRG EVAL-NOV11 A.J. Koning 625 1451 5 TENDL-2011 DIST- REV1- 625 1451 6 ----TENDL-2011 Material 625 REVISION 1 625 1451 7 -----Incident deuteron data 625 1451 8 ------ENDF-6 Format 625 1451 9 625 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 625 1451 11 625 1451 12 Charged particle transport library made by TALYS 625 1451 13 625 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 625 1451 15 625 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 625 1451 17 625 1451 18 This evaluated data file is based primarily on the nuclear model 625 1451 19 code TALYS [kon07], version 1.4. 625 1451 20 It is part of a large collection of isotopic 625 1451 21 evaluations, all created by running TALYS with either default or 625 1451 22 adjusted input parameters. This means that the data in this 625 1451 23 evaluation have been tested in detail for some isotopes against 625 1451 24 individual experimental data, while for other isotopes it is 625 1451 25 only as good as the global quality of TALYS at the present moment. 625 1451 26 The mutual quality of all individual evaluations is however 625 1451 27 consistent: The same set of nuclear models is used and, equally 625 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 625 1451 29 The resulting data file provides a complete representation of 625 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 625 1451 31 and shielding applications over the incident projectile energy 625 1451 32 range from 1.0E-11 to 200 MeV. 625 1451 33 625 1451 34 All transport data for neutrons, other particles, photons and 625 1451 35 residual nuclides are filed using a combination of MF1,3, 625 1451 36 and MF6. This includes cross sections, angular distributions, 625 1451 37 double-differential spectra, production cross sections, 625 1451 38 residual production (activation) cross sections and recoils. 625 1451 39 This evaluation can thus be used as both transport and activation 625 1451 40 library. The data file has been created automatically using the 625 1451 41 ENDF-6 format generator TEFAL. 625 1451 42 625 1451 43 ##### ORIGIN 625 1451 44 625 1451 45 All data < 200 MeV : Produced with TALYS code 625 1451 46 625 1451 47 *************************** T H E O R Y ************************** 625 1451 48 625 1451 49 TALYS is a computer code system for the prediction and analysis 625 1451 50 of nuclear reactions. TALYS simulates reactions that involve 625 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 625 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 625 1451 53 target nuclides of mass 12 and heavier. This is achieved by 625 1451 54 implementing a suite of nuclear reaction models into a single 625 1451 55 code system. It enables to evaluate nuclear reactions from 625 1451 56 the unresolved resonance region up to intermediate energies. This 625 1451 57 evaluation is based on a theoretical analysis that utilizes the 625 1451 58 optical model, compound nucleus statistical theory, fission, 625 1451 59 direct reactions and pre-equilibrium processes, in combination 625 1451 60 with databases and models for nuclear structure. The following 625 1451 61 output of TALYS is stored in this data file: 625 1451 62 625 1451 63 - Elastic and non-elastic cross sections 625 1451 64 - Elastic scattering angular distributions 625 1451 65 - Residual production cross sections 625 1451 66 - Recoil data 625 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 625 1451 68 - Total particle energy spectra 625 1451 69 - Total particle double-differential spectra 625 1451 70 625 1451 71 Here follows a short description of the used nuclear models: 625 1451 72 625 1451 73 ##### OPTICAL MODEL 625 1451 74 625 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 625 1451 76 in TALYS used as a subroutine. The default optical model 625 1451 77 potentials (OMP) used are the local and global parameterizations 625 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 625 1451 79 for neutrons and protons which in principle are valid over the 625 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 625 1451 81 validity may differ from nucleus to nucleus. For neutrons and 625 1451 82 protons, the used parameterization is given in Eq. (7) of 625 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 625 1451 84 used in the case of deformed nuclides. 625 1451 85 To calculate the transmission coefficients and reaction 625 1451 86 cross sections for deuterons, tritons, helions and alpha 625 1451 87 particles, we use OMPs that are directly derived from our nucleon 625 1451 88 potentials using Watanabe's folding approach. 625 1451 89 625 1451 90 ##### DIRECT REACTIONS 625 1451 91 625 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 625 1451 93 calculations for rotational or vibrational (or a combination of 625 1451 94 these) nuclides, depending on the information present in the 625 1451 95 nuclear structure database of TALYS. 625 1451 96 In addition, a macroscopic, phenomenological model to describe 625 1451 97 giant resonances in the inelastic channel is used. For each 625 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 625 1451 99 calculation with ECIS-06 is performed for each giant resonance 625 1451 100 state. The cross section is then spread over the continuum with a 625 1451 101 Gaussian distribution. 625 1451 102 625 1451 103 ##### COMPOUND NUCLEUS 625 1451 104 625 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 625 1451 106 model. The transmission coefficients have been generated 625 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 625 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 625 1451 109 For each nucleus that can be reached through a binary reaction, 625 1451 110 several discrete levels and a continuum described by level 625 1451 111 densities are included simultaneously as competing channels. 625 1451 112 625 1451 113 For each residual nucleus several discrete states 625 1451 114 are included as well as a continuum described by level densities. 625 1451 115 Multiple compound emission is continued until all reaction 625 1451 116 channels are closed and the population distribution of all 625 1451 117 residual nuclides is depleted, through gamma decay, until they 625 1451 118 end up in the ground state or in an isomer. 625 1451 119 625 1451 120 For the level density, we take the Constant Temperature Model 625 1451 121 with level density parameters as given in [kon08]. This employs 625 1451 122 the temperature law at low energies and a Fermi gas expression at 625 1451 123 high energies, and takes into account the damping of shell 625 1451 124 effects at high excitation energy. We have obtained the level 625 1451 125 density parameters from a simultaneous fit to all experimental 625 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 625 1451 127 625 1451 128 Gamma-ray transmission coefficients are generated with the 625 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 625 1451 130 giant dipole resonance parameters taken from the RIPL library 625 1451 131 [rip09], and normalized with experimental radiative widths. 625 1451 132 625 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 625 1451 134 625 1451 135 For pre-equilibrium reactions, which become important for 625 1451 136 incident energies above about 10 MeV, we use the two-component 625 1451 137 exciton model [kon04], in which the neutron or proton types of 625 1451 138 particles and holes are followed throughout the reaction. For 625 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 625 1451 140 any order of particle emission was included in the calculations. 625 1451 141 A parameterization for the squared matrix element is used that is 625 1451 142 valid for the whole energy range of this evaluation. 625 1451 143 625 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 625 1451 145 contribution was added from the pick/up and knock-out reaction 625 1451 146 model by Kalbach [kal05]. 625 1451 147 625 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 625 1451 149 applied, to simulate the direct and semi-direct capture 625 1451 150 processes. 625 1451 151 625 1451 152 The angular distribution systematics by Kalbach [kal88] were used 625 1451 153 to describe the angular distributions for all continuum 625 1451 154 particles. An isotopic distribution for photons was adopted. 625 1451 155 625 1451 156 ***************** F I L E I N F O R M A T I O N **************** 625 1451 157 625 1451 158 ##### MF1: GENERAL INFORMATION 625 1451 159 625 1451 160 - MT451 : Descriptive data and directory 625 1451 161 625 1451 162 This text and the full directory of used MF/MT sections. 625 1451 163 625 1451 164 ##### MF3: REACTION CROSS SECTIONS 625 1451 165 625 1451 166 All the data present in the following MT-sections have been 625 1451 167 calculated with TALYS. If the maximal cross section in an 625 1451 168 excitation function over the whole energy range does not exceed 625 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 625 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 625 1451 171 are set to zero. The minor reaction channels are also present 625 1451 172 but they are stored in MT5 and can be reproduced in combination 625 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 625 1451 174 The following reaction channels/MT numbers are generally included: 625 1451 175 625 1451 176 - MT2 : Elastic scattering cross section: nuclear + 625 1451 177 interference terms 625 1451 178 625 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 625 1451 180 distributions of MF=6. Note that because of the interference 625 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 625 1451 182 some energies and angles. 625 1451 183 625 1451 184 - MT5 : (p,anything) cross section 625 1451 185 625 1451 186 MT5 contains the reactions which can not be stored in any other 625 1451 187 MT-number. As the incident energy increases, the cross section 625 1451 188 of MT5 increases as more reaction channels are no longer stored 625 1451 189 under a specific MT number. The information of MF3/MT5 can be 625 1451 190 combined with MF6/MT5 to obtain residual production cross 625 1451 191 sections particle production cross sections and 625 1451 192 (double-)differential cross sections. 625 1451 193 625 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 625 1451 195 625 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 625 1451 197 distributions, as well as all residual and photon production 625 1451 198 cross sections. All data are generated with TALYS. 625 1451 199 625 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 625 1451 201 interference terms 625 1451 202 625 1451 203 Relative angular distributions are tabulated on an angular grid. 625 1451 204 They are obtained by using the "nuclear-plus-interference" option 625 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 625 1451 206 integrated cross section is stored in MF=3. Note that because 625 1451 207 of the interference effect, the tabulations in both MF=6 and 625 1451 208 MF=3 can be negative at some energies and angles. 625 1451 209 625 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 625 1451 211 625 1451 212 MT5 contains the production yields of particles and residual 625 1451 213 products. It also contains the secondary energy-angle 625 1451 214 distributions for all particles and photons. First, the yields 625 1451 215 for neutrons are given for the whole energy range. Next, on a 625 1451 216 secondary energy grid the relative emission spectra are given 625 1451 217 together with the parameters for the Kalbach systematics for 625 1451 218 angular distributions. Inelastic scattering cross sections for 625 1451 219 discrete states have been broadened and added to the continuum 625 1451 220 spectra. This procedure is repeated for protons, deuterons, 625 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 625 1451 222 residual production yields are given per final product. All these 625 1451 223 yields and relative distributions can be multiplied with the 625 1451 224 cross sections given in MF3/MT5 to get the production cross 625 1451 225 sections and (double-)differential cross sections. 625 1451 226 625 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 **** 625 1451 228 625 1451 229 This file has been checked successfully by the BNL checking 625 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 625 1451 231 been processed successfully into an MCNP library by the 625 1451 232 processing code NJOY99.90 [mac00]. 625 1451 233 625 1451 234 *********************** R E F E R E N C E S ********************** 625 1451 235 625 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 625 1451 237 (1985). 625 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 625 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 625 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 625 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 625 1451 242 (2003). 625 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 625 1451 244 (2004). 625 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 625 1451 246 Proceedings of the International Conference on Nuclear 625 1451 247 Data for Science and Technology - ND-2007, 625 1451 248 April 22-27, 2007, Nice, France 625 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 625 1451 250 local level density models'', Nucl Phys. A810, 625 1451 251 13-76 (2008). 625 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 625 1451 253 pointwise and multigroup neutron and photon cross 625 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 625 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 625 1451 256 No. CEA-N-2772, 1994. 625 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 625 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 625 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 625 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 625 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 625 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 625 1451 263 RIPL - Reference Input Parameter Library for calculation 625 1451 264 of nuclear reactions and nuclear data evaluation, 625 1451 265 Nucl. Data Sheets 110, 3107 (2009). 625 1451 266 625 1451 267 ************************* C O N T E N T S ************************ 625 1451 268 625 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 625 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 625 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 625 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 625 1451 273 Unshielded Group Averages Using 162 Groups 625 1451 274 Weighting Spectrum: 1/E Spectrum 625 1451 275 1 451 275 1 625 1451 276 3 2 18 1 625 1451 277 3 5 18 1 625 1451 278 6 2 1401 1 625 1451 279 6 5 12938 1 625 1451 280 8 5 32 1 625 1451 281 625 1 0 282 625 0 0 283 6012.00000 11.8969200 0 0 0 0 625 3 2 284 0.0 0.0 0 0 1 45 625 3 2 285 45 1 625 3 2 286 17000000.0 .003467267 18000000.0 .010203905 19000000.0 .015978704 625 3 2 287 20000000.0 .022856753 22000000.0 .030289384 24000000.0 .035813524 625 3 2 288 26000000.0 .039969741 28000000.0 .043140736 30000000.0 .046756652 625 3 2 289 35000000.0 .050494235 40000000.0 .053040035 45000000.0 .055181336 625 3 2 290 50000000.0 .057054058 54000000.0 .058110150 55000000.0 .059388006 625 3 2 291 60000000.0 .062636316 70000000.0 .066899871 80000000.0 .069948977 625 3 2 292 90000000.0 .070993489 100000000. .069006228 120000000. .061935861 625 3 2 293 140000000. .051737808 160000000. .041038160 180000000. .031719009 625 3 2 294 200000000. .027536520 240000000. .027536520 280000000. .027536520 625 3 2 295 320000000. .027536520 360000000. .027536520 400000000. .027536520 625 3 2 296 440000000. .027536520 480000000. .027536520 520000000. .027536520 625 3 2 297 560000000. .027536520 600000000. .027536520 640000000. .027536520 625 3 2 298 680000000. .027536520 720000000. .027536520 760000000. .027536520 625 3 2 299 800000000. .027536520 840000000. .027536520 880000000. .027536520 625 3 2 300 920000000. .027536520 960000000. .027536520 1.00000E+9 0.0 625 3 2 301 625 3 0 302 6012.00000 11.8969200 0 0 0 0 625 3 5 303 0.0 0.0 0 0 1 106 625 3 5 304 106 1 625 3 5 305 1000000.00 .238682407 1125000.00 .281409065 1250000.00 .324127295 625 3 5 306 1375000.00 .366839300 1500000.00 .409546578 1625000.00 .452250181 625 3 5 307 1750000.00 .494950870 1875000.00 .537649210 2000000.00 .568763221 625 3 5 308 2125000.00 .588050040 2250000.00 .607336256 2375000.00 .626621962 625 3 5 309 2500000.00 .645907233 2625000.00 .665192131 2750000.00 .684476704 625 3 5 310 2875000.00 .703760994 3000000.00 .720022614 3125000.00 .733219375 625 3 5 311 3250000.00 .746416002 3375000.00 .759612512 3500000.00 .772808915 625 3 5 312 3625000.00 .786005223 3750000.00 .799201446 3875000.00 .812397590 625 3 5 313 4000000.00 .827378558 4200000.00 .844215743 4400000.00 .861052644 625 3 5 314 4600000.00 .877889298 4800000.00 .894725735 5000000.00 .907745105 625 3 5 315 5200000.00 .916896254 5400000.00 .926047320 5600000.00 .935198313 625 3 5 316 5800000.00 .944349241 6000000.00 .952321564 6200000.00 .959102132 625 3 5 317 6400000.00 .965882664 6600000.00 .972663163 6800000.00 .979443632 625 3 5 318 7000000.00 .984324047 7200000.00 .987286237 7400000.00 .990248417 625 3 5 319 7600000.00 .993210588 7800000.00 .996172751 8000000.00 .998759457 625 3 5 320 8200000.00 1.00096757 8400000.00 1.00317567 8600000.00 1.00538377 625 3 5 321 8800000.00 1.00759187 9000000.00 1.00966645 9200000.00 1.01160652 625 3 5 322 9400000.00 1.01354660 9600000.00 1.01548667 9800000.00 1.01742673 625 3 5 323 10000000.0 1.02061426 11000000.0 1.02349130 12000000.0 1.02316267 625 3 5 324 13000000.0 1.02002717 14000000.0 1.01483449 15000000.0 1.00848603 625 3 5 325 16000000.0 1.00160077 17000000.0 .994538629 18000000.0 .988125951 625 3 5 326 19000000.0 .981675610 20000000.0 .970974565 22000000.0 .957017635 625 3 5 327 24000000.0 .943377157 26000000.0 .929940573 28000000.0 .917157019 625 3 5 328 30000000.0 .896478341 35000000.0 .868209466 40000000.0 .842244343 625 3 5 329 45000000.0 .818138972 50000000.0 .797680310 54000000.0 .786634701 625 3 5 330 55000000.0 .774195777 60000000.0 .744405438 70000000.0 .706358032 625 3 5 331 80000000.0 .674987684 90000000.0 .652492404 100000000. .626375488 625 3 5 332 120000000. .595784864 140000000. .569678919 160000000. .546829414 625 3 5 333 180000000. .526582752 200000000. .517150000 240000000. .517150000 625 3 5 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351 2875000.00 .108152004 3000000.00 .110611552 3125000.00 .113829103 62510 5 352 3250000.00 .117046622 3375000.00 .120264112 3500000.00 .123481576 62510 5 353 3625000.00 .126699017 3750000.00 .129916437 3875000.00 .133133839 62510 5 354 4000000.00 .140530724 4200000.00 .152187321 4400000.00 .163843721 62510 5 355 4600000.00 .175499951 4800000.00 .187156031 5000000.00 .194721064 62510 5 356 5200000.00 .198140218 5400000.00 .201559341 5600000.00 .204978436 62510 5 357 5800000.00 .208397507 6000000.00 .208892737 6200000.00 .206431485 62510 5 358 6400000.00 .203970246 6600000.00 .201509020 6800000.00 .199047804 62510 5 359 7000000.00 .197132412 7200000.00 .195768063 7400000.00 .194403719 62510 5 360 7600000.00 .193039380 7800000.00 .191675044 8000000.00 .194519341 62510 5 361 8200000.00 .201607474 8400000.00 .208695590 8600000.00 .215783691 62510 5 362 8800000.00 .222871777 9000000.00 .235531549 9200000.00 .253804417 62510 5 363 9400000.00 .272077255 9600000.00 .290350064 9800000.00 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.962591981 480000000. .962591981 62510 5 377 520000000. .962591981 560000000. .962591981 600000000. .962591981 62510 5 378 640000000. .962591981 680000000. .962591981 720000000. .962591981 62510 5 379 760000000. .962591981 800000000. .962591981 840000000. .962591981 62510 5 380 880000000. .962591981 920000000. .962591981 960000000. .962591981 62510 5 381 1.00000E+9 0.0 62510 5 382 0.0 0.0 1001 0 1 106 62510 5 383 106 1 62510 5 384 1000000.00 .019221256 1125000.00 .058473910 1250000.00 .097718822 62510 5 385 1375000.00 .136958016 1500000.00 .176192866 1625000.00 .215424340 62510 5 386 1750000.00 .254653137 1875000.00 .293879776 2000000.00 .319732738 62510 5 387 2125000.00 .331931560 2250000.00 .344130000 2375000.00 .356328118 62510 5 388 2500000.00 .368525961 2625000.00 .380723567 2750000.00 .392920968 62510 5 389 2875000.00 .405118190 3000000.00 .410927584 3125000.00 .410259938 62510 5 390 3250000.00 .409592299 3375000.00 .408924666 3500000.00 .408257039 62510 5 391 3625000.00 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