TENDL-2011 gen. purp. file: deuteron + Li- 7 1 MeV - 200 MeV 99 0 0 0 3007.00000 6.95573400 -1 0 0 1 328 1451 1 0.0 0.0 0 0 0 6 328 1451 2 1.99626000 200000000. 1 0 10020 0 328 1451 3 0.0 0.0 1 0 271 6 328 1451 4 3-Li- 7 NRG EVAL-NOV11 A.J. Koning 328 1451 5 TENDL-2011 DIST- REV1- 328 1451 6 ----TENDL-2011 Material 328 REVISION 1 328 1451 7 -----Incident deuteron data 328 1451 8 ------ENDF-6 Format 328 1451 9 328 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 328 1451 11 328 1451 12 Charged particle transport library made by TALYS 328 1451 13 328 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 328 1451 15 328 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 328 1451 17 328 1451 18 This evaluated data file is based primarily on the nuclear model 328 1451 19 code TALYS [kon07], version 1.4. 328 1451 20 It is part of a large collection of isotopic 328 1451 21 evaluations, all created by running TALYS with either default or 328 1451 22 adjusted input parameters. This means that the data in this 328 1451 23 evaluation have been tested in detail for some isotopes against 328 1451 24 individual experimental data, while for other isotopes it is 328 1451 25 only as good as the global quality of TALYS at the present moment. 328 1451 26 The mutual quality of all individual evaluations is however 328 1451 27 consistent: The same set of nuclear models is used and, equally 328 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 328 1451 29 The resulting data file provides a complete representation of 328 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 328 1451 31 and shielding applications over the incident projectile energy 328 1451 32 range from 1.0E-11 to 200 MeV. 328 1451 33 328 1451 34 All transport data for neutrons, other particles, photons and 328 1451 35 residual nuclides are filed using a combination of MF1,3, 328 1451 36 and MF6. This includes cross sections, angular distributions, 328 1451 37 double-differential spectra, production cross sections, 328 1451 38 residual production (activation) cross sections and recoils. 328 1451 39 This evaluation can thus be used as both transport and activation 328 1451 40 library. The data file has been created automatically using the 328 1451 41 ENDF-6 format generator TEFAL. 328 1451 42 328 1451 43 ##### ORIGIN 328 1451 44 328 1451 45 All data < 200 MeV : Produced with TALYS code 328 1451 46 328 1451 47 *************************** T H E O R Y ************************** 328 1451 48 328 1451 49 TALYS is a computer code system for the prediction and analysis 328 1451 50 of nuclear reactions. TALYS simulates reactions that involve 328 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 328 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 328 1451 53 target nuclides of mass 12 and heavier. This is achieved by 328 1451 54 implementing a suite of nuclear reaction models into a single 328 1451 55 code system. It enables to evaluate nuclear reactions from 328 1451 56 the unresolved resonance region up to intermediate energies. This 328 1451 57 evaluation is based on a theoretical analysis that utilizes the 328 1451 58 optical model, compound nucleus statistical theory, fission, 328 1451 59 direct reactions and pre-equilibrium processes, in combination 328 1451 60 with databases and models for nuclear structure. The following 328 1451 61 output of TALYS is stored in this data file: 328 1451 62 328 1451 63 - Elastic and non-elastic cross sections 328 1451 64 - Elastic scattering angular distributions 328 1451 65 - Residual production cross sections 328 1451 66 - Recoil data 328 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 328 1451 68 - Total particle energy spectra 328 1451 69 - Total particle double-differential spectra 328 1451 70 328 1451 71 Here follows a short description of the used nuclear models: 328 1451 72 328 1451 73 ##### OPTICAL MODEL 328 1451 74 328 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 328 1451 76 in TALYS used as a subroutine. The default optical model 328 1451 77 potentials (OMP) used are the local and global parameterizations 328 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 328 1451 79 for neutrons and protons which in principle are valid over the 328 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 328 1451 81 validity may differ from nucleus to nucleus. For neutrons and 328 1451 82 protons, the used parameterization is given in Eq. (7) of 328 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 328 1451 84 used in the case of deformed nuclides. 328 1451 85 To calculate the transmission coefficients and reaction 328 1451 86 cross sections for deuterons, tritons, helions and alpha 328 1451 87 particles, we use OMPs that are directly derived from our nucleon 328 1451 88 potentials using Watanabe's folding approach. 328 1451 89 328 1451 90 ##### DIRECT REACTIONS 328 1451 91 328 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 328 1451 93 calculations for rotational or vibrational (or a combination of 328 1451 94 these) nuclides, depending on the information present in the 328 1451 95 nuclear structure database of TALYS. 328 1451 96 In addition, a macroscopic, phenomenological model to describe 328 1451 97 giant resonances in the inelastic channel is used. For each 328 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 328 1451 99 calculation with ECIS-06 is performed for each giant resonance 328 1451 100 state. The cross section is then spread over the continuum with a 328 1451 101 Gaussian distribution. 328 1451 102 328 1451 103 ##### COMPOUND NUCLEUS 328 1451 104 328 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 328 1451 106 model. The transmission coefficients have been generated 328 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 328 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 328 1451 109 For each nucleus that can be reached through a binary reaction, 328 1451 110 several discrete levels and a continuum described by level 328 1451 111 densities are included simultaneously as competing channels. 328 1451 112 328 1451 113 For each residual nucleus several discrete states 328 1451 114 are included as well as a continuum described by level densities. 328 1451 115 Multiple compound emission is continued until all reaction 328 1451 116 channels are closed and the population distribution of all 328 1451 117 residual nuclides is depleted, through gamma decay, until they 328 1451 118 end up in the ground state or in an isomer. 328 1451 119 328 1451 120 For the level density, we take the Constant Temperature Model 328 1451 121 with level density parameters as given in [kon08]. This employs 328 1451 122 the temperature law at low energies and a Fermi gas expression at 328 1451 123 high energies, and takes into account the damping of shell 328 1451 124 effects at high excitation energy. We have obtained the level 328 1451 125 density parameters from a simultaneous fit to all experimental 328 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 328 1451 127 328 1451 128 Gamma-ray transmission coefficients are generated with the 328 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 328 1451 130 giant dipole resonance parameters taken from the RIPL library 328 1451 131 [rip09], and normalized with experimental radiative widths. 328 1451 132 328 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 328 1451 134 328 1451 135 For pre-equilibrium reactions, which become important for 328 1451 136 incident energies above about 10 MeV, we use the two-component 328 1451 137 exciton model [kon04], in which the neutron or proton types of 328 1451 138 particles and holes are followed throughout the reaction. For 328 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 328 1451 140 any order of particle emission was included in the calculations. 328 1451 141 A parameterization for the squared matrix element is used that is 328 1451 142 valid for the whole energy range of this evaluation. 328 1451 143 328 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 328 1451 145 contribution was added from the pick/up and knock-out reaction 328 1451 146 model by Kalbach [kal05]. 328 1451 147 328 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 328 1451 149 applied, to simulate the direct and semi-direct capture 328 1451 150 processes. 328 1451 151 328 1451 152 The angular distribution systematics by Kalbach [kal88] were used 328 1451 153 to describe the angular distributions for all continuum 328 1451 154 particles. An isotopic distribution for photons was adopted. 328 1451 155 328 1451 156 ***************** F I L E I N F O R M A T I O N **************** 328 1451 157 328 1451 158 ##### MF1: GENERAL INFORMATION 328 1451 159 328 1451 160 - MT451 : Descriptive data and directory 328 1451 161 328 1451 162 This text and the full directory of used MF/MT sections. 328 1451 163 328 1451 164 ##### MF3: REACTION CROSS SECTIONS 328 1451 165 328 1451 166 All the data present in the following MT-sections have been 328 1451 167 calculated with TALYS. If the maximal cross section in an 328 1451 168 excitation function over the whole energy range does not exceed 328 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 328 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 328 1451 171 are set to zero. The minor reaction channels are also present 328 1451 172 but they are stored in MT5 and can be reproduced in combination 328 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 328 1451 174 The following reaction channels/MT numbers are generally included: 328 1451 175 328 1451 176 - MT2 : Elastic scattering cross section: nuclear + 328 1451 177 interference terms 328 1451 178 328 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 328 1451 180 distributions of MF=6. Note that because of the interference 328 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 328 1451 182 some energies and angles. 328 1451 183 328 1451 184 - MT5 : (p,anything) cross section 328 1451 185 328 1451 186 MT5 contains the reactions which can not be stored in any other 328 1451 187 MT-number. As the incident energy increases, the cross section 328 1451 188 of MT5 increases as more reaction channels are no longer stored 328 1451 189 under a specific MT number. The information of MF3/MT5 can be 328 1451 190 combined with MF6/MT5 to obtain residual production cross 328 1451 191 sections particle production cross sections and 328 1451 192 (double-)differential cross sections. 328 1451 193 328 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 328 1451 195 328 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 328 1451 197 distributions, as well as all residual and photon production 328 1451 198 cross sections. All data are generated with TALYS. 328 1451 199 328 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 328 1451 201 interference terms 328 1451 202 328 1451 203 Relative angular distributions are tabulated on an angular grid. 328 1451 204 They are obtained by using the "nuclear-plus-interference" option 328 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 328 1451 206 integrated cross section is stored in MF=3. Note that because 328 1451 207 of the interference effect, the tabulations in both MF=6 and 328 1451 208 MF=3 can be negative at some energies and angles. 328 1451 209 328 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 328 1451 211 328 1451 212 MT5 contains the production yields of particles and residual 328 1451 213 products. It also contains the secondary energy-angle 328 1451 214 distributions for all particles and photons. First, the yields 328 1451 215 for neutrons are given for the whole energy range. Next, on a 328 1451 216 secondary energy grid the relative emission spectra are given 328 1451 217 together with the parameters for the Kalbach systematics for 328 1451 218 angular distributions. Inelastic scattering cross sections for 328 1451 219 discrete states have been broadened and added to the continuum 328 1451 220 spectra. This procedure is repeated for protons, deuterons, 328 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 328 1451 222 residual production yields are given per final product. All these 328 1451 223 yields and relative distributions can be multiplied with the 328 1451 224 cross sections given in MF3/MT5 to get the production cross 328 1451 225 sections and (double-)differential cross sections. 328 1451 226 328 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 **** 328 1451 228 328 1451 229 This file has been checked successfully by the BNL checking 328 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 328 1451 231 been processed successfully into an MCNP library by the 328 1451 232 processing code NJOY99.90 [mac00]. 328 1451 233 328 1451 234 *********************** R E F E R E N C E S ********************** 328 1451 235 328 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 328 1451 237 (1985). 328 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 328 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 328 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 328 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 328 1451 242 (2003). 328 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 328 1451 244 (2004). 328 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 328 1451 246 Proceedings of the International Conference on Nuclear 328 1451 247 Data for Science and Technology - ND-2007, 328 1451 248 April 22-27, 2007, Nice, France 328 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 328 1451 250 local level density models'', Nucl Phys. A810, 328 1451 251 13-76 (2008). 328 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 328 1451 253 pointwise and multigroup neutron and photon cross 328 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 328 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 328 1451 256 No. CEA-N-2772, 1994. 328 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 328 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 328 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 328 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 328 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 328 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 328 1451 263 RIPL - Reference Input Parameter Library for calculation 328 1451 264 of nuclear reactions and nuclear data evaluation, 328 1451 265 Nucl. Data Sheets 110, 3107 (2009). 328 1451 266 328 1451 267 ************************* C O N T E N T S ************************ 328 1451 268 328 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 328 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 328 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 328 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 328 1451 273 Unshielded Group Averages Using 162 Groups 328 1451 274 Weighting Spectrum: 1/E Spectrum 328 1451 275 1 451 275 1 328 1451 276 3 2 18 1 328 1451 277 3 5 18 1 328 1451 278 6 2 1401 1 328 1451 279 6 5 11469 1 328 1451 280 8 5 13 1 328 1451 281 328 1 0 282 328 0 0 283 3007.00000 6.95573400 0 0 0 0 328 3 2 284 0.0 0.0 0 0 1 67 328 3 2 285 67 1 328 3 2 286 7000000.00-.006529579 7200000.00-.003776814 7400000.00-.001024059 328 3 2 287 7600000.00 .001728688 7800000.00 .004481427 8000000.00 .006825646 328 3 2 288 8200000.00 .008757931 8400000.00 .010690210 8600000.00 .012622486 328 3 2 289 8800000.00 .014554757 9000000.00 .016229839 9200000.00 .017645818 328 3 2 290 9400000.00 .019061796 9600000.00 .020477771 9800000.00 .021893744 328 3 2 291 10000000.0 .025308115 11000000.0 .030290112 12000000.0 .034339301 328 3 2 292 13000000.0 .037649320 14000000.0 .040356250 15000000.0 .042589767 328 3 2 293 16000000.0 .044478371 17000000.0 .046134987 18000000.0 .047645180 328 3 2 294 19000000.0 .049063762 20000000.0 .051043881 22000000.0 .053553371 328 3 2 295 24000000.0 .055864087 26000000.0 .057998619 28000000.0 .059995440 328 3 2 296 30000000.0 .063142285 35000000.0 .067305487 40000000.0 .070827051 328 3 2 297 45000000.0 .073659435 50000000.0 .075625799 54000000.0 .076529616 328 3 2 298 55000000.0 .077268683 60000000.0 .078217785 70000000.0 .077664444 328 3 2 299 80000000.0 .075182028 90000000.0 .071401724 100000000. .064004677 328 3 2 300 120000000. .052486818 140000000. .040932748 160000000. .031019289 328 3 2 301 180000000. .023498463 200000000. .020423030 240000000. .020423030 328 3 2 302 280000000. .020423030 320000000. .020423030 360000000. .020423030 328 3 2 303 400000000. .020423030 440000000. .020423030 480000000. .020423030 328 3 2 304 520000000. .020423030 560000000. .020423030 600000000. .020423030 328 3 2 305 640000000. .020423030 680000000. .020423030 720000000. .020423030 328 3 2 306 760000000. .020423030 800000000. .020423030 840000000. .020423030 328 3 2 307 880000000. .020423030 920000000. .020423030 960000000. .020423030 328 3 2 308 1.00000E+9 0.0 328 3 2 309 328 3 0 310 3007.00000 6.95573400 0 0 0 0 328 3 5 311 0.0 0.0 0 0 1 106 328 3 5 312 106 1 328 3 5 313 1000000.00 .797785472 1125000.00 .811254405 1250000.00 .824720681 328 3 5 314 1375000.00 .838184994 1500000.00 .851647818 1625000.00 .865109483 328 3 5 315 1750000.00 .878570229 1875000.00 .892030235 2000000.00 .900320507 328 3 5 316 2125000.00 .903332614 2250000.00 .906344627 2375000.00 .909356560 328 3 5 317 2500000.00 .912368426 2625000.00 .915380233 2750000.00 .918391990 328 3 5 318 2875000.00 .921403702 3000000.00 .922615219 3125000.00 .922001400 328 3 5 319 3250000.00 .921387586 3375000.00 .920773779 3500000.00 .920159976 328 3 5 320 3625000.00 .919546178 3750000.00 .918932383 3875000.00 .918318593 328 3 5 321 4000000.00 .916749425 4200000.00 .914206924 4400000.00 .911664466 328 3 5 322 4600000.00 .909122045 4800000.00 .906579657 5000000.00 .905243373 328 3 5 323 5200000.00 .905129359 5400000.00 .905015345 5600000.00 .904901333 328 3 5 324 5800000.00 .904787322 6000000.00 .905922448 6200000.00 .908320656 328 3 5 325 6400000.00 .910718851 6600000.00 .913117034 6800000.00 .915515207 328 3 5 326 7000000.00 .916534873 7200000.00 .916162849 7400000.00 .915790827 328 3 5 327 7600000.00 .915418805 7800000.00 .915046785 8000000.00 .914108117 328 3 5 328 8200000.00 .912598062 8400000.00 .911088011 8600000.00 .909577963 328 3 5 329 8800000.00 .908067918 9000000.00 .906492097 9200000.00 .904850013 328 3 5 330 9400000.00 .903207931 9600000.00 .901565851 9800000.00 .899923774 328 3 5 331 10000000.0 .894735439 11000000.0 .885381340 12000000.0 .874968347 328 3 5 332 13000000.0 .863568028 14000000.0 .857337983 15000000.0 .850249317 328 3 5 333 16000000.0 .836929813 17000000.0 .824403301 18000000.0 .812784514 328 3 5 334 19000000.0 .802119153 20000000.0 .787585537 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920000000. .347940000 960000000. .347940000 328 3 5 348 1.00000E+9 0.0 328 3 5 349 328 3 0 350 328 0 0 351 3007.00000 6.95573400 0 0 13 0 32810 5 352 0.0 0.0 1 0 1 106 32810 5 353 106 1 32810 5 354 1000000.00 .020196042 1125000.00 .061439353 1250000.00 .102674529 32810 5 355 1375000.00 .143903696 1500000.00 .185128301 1625000.00 .226349357 32810 5 356 1750000.00 .267567602 1875000.00 .308783578 2000000.00 .329771519 32810 5 357 2125000.00 .330107060 2250000.00 .330442589 2375000.00 .330778110 32810 5 358 2500000.00 .331113624 2625000.00 .331449131 2750000.00 .331784632 32810 5 359 2875000.00 .332120129 3000000.00 .331885718 3125000.00 .331073441 32810 5 360 3250000.00 .330261171 3375000.00 .329448909 3500000.00 .328636653 32810 5 361 3625000.00 .327824404 3750000.00 .327012159 3875000.00 .326199919 32810 5 362 4000000.00 .324822494 4200000.00 .322867850 4400000.00 .320913239 32810 5 363 4600000.00 .318958656 4800000.00 .317004098 5000000.00 .322133707 32810 5 364 5200000.00 .334442462 5400000.00 .346751106 5600000.00 .359059651 32810 5 365 5800000.00 .371368108 6000000.00 .393927542 6200000.00 .426852408 32810 5 366 6400000.00 .459777098 6600000.00 .492701627 6800000.00 .525626011 32810 5 367 7000000.00 .552552678 7200000.00 .573424280 7400000.00 .594295810 32810 5 368 7600000.00 .615167274 7800000.00 .636038677 8000000.00 .652742534 32810 5 369 8200000.00 .665243988 8400000.00 .677745414 8600000.00 .690246813 32810 5 370 8800000.00 .702748186 9000000.00 .711983207 9200000.00 .717927602 32810 5 371 9400000.00 .723871987 9600000.00 .729816363 9800000.00 .735760730 32810 5 372 10000000.0 .747591485 11000000.0 .763919798 12000000.0 .776073115 32810 5 373 13000000.0 .781497039 14000000.0 .742723415 15000000.0 .698190912 32810 5 374 16000000.0 .686628156 17000000.0 .680155733 18000000.0 .672626647 32810 5 375 19000000.0 .660080059 20000000.0 .644434544 22000000.0 .573852092 32810 5 376 24000000.0 .491490966 26000000.0 .459290213 28000000.0 .442424365 32810 5 377 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