TENDL-2011 gen. purp. file: deuteron + Be- 9 1 MeV - 200 MeV 99 0 0 0 4009.00000 8.93476400 -1 0 0 1 425 1451 1 0.0 0.0 0 0 0 6 425 1451 2 1.99626000 200000000. 1 0 10020 0 425 1451 3 0.0 0.0 1 0 271 6 425 1451 4 4-Be- 9 NRG EVAL-NOV11 A.J. Koning 425 1451 5 TENDL-2011 DIST- REV1- 425 1451 6 ----TENDL-2011 Material 425 REVISION 1 425 1451 7 -----Incident deuteron data 425 1451 8 ------ENDF-6 Format 425 1451 9 425 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 425 1451 11 425 1451 12 Charged particle transport library made by TALYS 425 1451 13 425 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 425 1451 15 425 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 425 1451 17 425 1451 18 This evaluated data file is based primarily on the nuclear model 425 1451 19 code TALYS [kon07], version 1.4. 425 1451 20 It is part of a large collection of isotopic 425 1451 21 evaluations, all created by running TALYS with either default or 425 1451 22 adjusted input parameters. This means that the data in this 425 1451 23 evaluation have been tested in detail for some isotopes against 425 1451 24 individual experimental data, while for other isotopes it is 425 1451 25 only as good as the global quality of TALYS at the present moment. 425 1451 26 The mutual quality of all individual evaluations is however 425 1451 27 consistent: The same set of nuclear models is used and, equally 425 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 425 1451 29 The resulting data file provides a complete representation of 425 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 425 1451 31 and shielding applications over the incident projectile energy 425 1451 32 range from 1.0E-11 to 200 MeV. 425 1451 33 425 1451 34 All transport data for neutrons, other particles, photons and 425 1451 35 residual nuclides are filed using a combination of MF1,3, 425 1451 36 and MF6. This includes cross sections, angular distributions, 425 1451 37 double-differential spectra, production cross sections, 425 1451 38 residual production (activation) cross sections and recoils. 425 1451 39 This evaluation can thus be used as both transport and activation 425 1451 40 library. The data file has been created automatically using the 425 1451 41 ENDF-6 format generator TEFAL. 425 1451 42 425 1451 43 ##### ORIGIN 425 1451 44 425 1451 45 All data < 200 MeV : Produced with TALYS code 425 1451 46 425 1451 47 *************************** T H E O R Y ************************** 425 1451 48 425 1451 49 TALYS is a computer code system for the prediction and analysis 425 1451 50 of nuclear reactions. TALYS simulates reactions that involve 425 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 425 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 425 1451 53 target nuclides of mass 12 and heavier. This is achieved by 425 1451 54 implementing a suite of nuclear reaction models into a single 425 1451 55 code system. It enables to evaluate nuclear reactions from 425 1451 56 the unresolved resonance region up to intermediate energies. This 425 1451 57 evaluation is based on a theoretical analysis that utilizes the 425 1451 58 optical model, compound nucleus statistical theory, fission, 425 1451 59 direct reactions and pre-equilibrium processes, in combination 425 1451 60 with databases and models for nuclear structure. The following 425 1451 61 output of TALYS is stored in this data file: 425 1451 62 425 1451 63 - Elastic and non-elastic cross sections 425 1451 64 - Elastic scattering angular distributions 425 1451 65 - Residual production cross sections 425 1451 66 - Recoil data 425 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 425 1451 68 - Total particle energy spectra 425 1451 69 - Total particle double-differential spectra 425 1451 70 425 1451 71 Here follows a short description of the used nuclear models: 425 1451 72 425 1451 73 ##### OPTICAL MODEL 425 1451 74 425 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 425 1451 76 in TALYS used as a subroutine. The default optical model 425 1451 77 potentials (OMP) used are the local and global parameterizations 425 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 425 1451 79 for neutrons and protons which in principle are valid over the 425 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 425 1451 81 validity may differ from nucleus to nucleus. For neutrons and 425 1451 82 protons, the used parameterization is given in Eq. (7) of 425 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 425 1451 84 used in the case of deformed nuclides. 425 1451 85 To calculate the transmission coefficients and reaction 425 1451 86 cross sections for deuterons, tritons, helions and alpha 425 1451 87 particles, we use OMPs that are directly derived from our nucleon 425 1451 88 potentials using Watanabe's folding approach. 425 1451 89 425 1451 90 ##### DIRECT REACTIONS 425 1451 91 425 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 425 1451 93 calculations for rotational or vibrational (or a combination of 425 1451 94 these) nuclides, depending on the information present in the 425 1451 95 nuclear structure database of TALYS. 425 1451 96 In addition, a macroscopic, phenomenological model to describe 425 1451 97 giant resonances in the inelastic channel is used. For each 425 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 425 1451 99 calculation with ECIS-06 is performed for each giant resonance 425 1451 100 state. The cross section is then spread over the continuum with a 425 1451 101 Gaussian distribution. 425 1451 102 425 1451 103 ##### COMPOUND NUCLEUS 425 1451 104 425 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 425 1451 106 model. The transmission coefficients have been generated 425 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 425 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 425 1451 109 For each nucleus that can be reached through a binary reaction, 425 1451 110 several discrete levels and a continuum described by level 425 1451 111 densities are included simultaneously as competing channels. 425 1451 112 425 1451 113 For each residual nucleus several discrete states 425 1451 114 are included as well as a continuum described by level densities. 425 1451 115 Multiple compound emission is continued until all reaction 425 1451 116 channels are closed and the population distribution of all 425 1451 117 residual nuclides is depleted, through gamma decay, until they 425 1451 118 end up in the ground state or in an isomer. 425 1451 119 425 1451 120 For the level density, we take the Constant Temperature Model 425 1451 121 with level density parameters as given in [kon08]. This employs 425 1451 122 the temperature law at low energies and a Fermi gas expression at 425 1451 123 high energies, and takes into account the damping of shell 425 1451 124 effects at high excitation energy. We have obtained the level 425 1451 125 density parameters from a simultaneous fit to all experimental 425 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 425 1451 127 425 1451 128 Gamma-ray transmission coefficients are generated with the 425 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 425 1451 130 giant dipole resonance parameters taken from the RIPL library 425 1451 131 [rip09], and normalized with experimental radiative widths. 425 1451 132 425 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 425 1451 134 425 1451 135 For pre-equilibrium reactions, which become important for 425 1451 136 incident energies above about 10 MeV, we use the two-component 425 1451 137 exciton model [kon04], in which the neutron or proton types of 425 1451 138 particles and holes are followed throughout the reaction. For 425 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 425 1451 140 any order of particle emission was included in the calculations. 425 1451 141 A parameterization for the squared matrix element is used that is 425 1451 142 valid for the whole energy range of this evaluation. 425 1451 143 425 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 425 1451 145 contribution was added from the pick/up and knock-out reaction 425 1451 146 model by Kalbach [kal05]. 425 1451 147 425 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 425 1451 149 applied, to simulate the direct and semi-direct capture 425 1451 150 processes. 425 1451 151 425 1451 152 The angular distribution systematics by Kalbach [kal88] were used 425 1451 153 to describe the angular distributions for all continuum 425 1451 154 particles. An isotopic distribution for photons was adopted. 425 1451 155 425 1451 156 ***************** F I L E I N F O R M A T I O N **************** 425 1451 157 425 1451 158 ##### MF1: GENERAL INFORMATION 425 1451 159 425 1451 160 - MT451 : Descriptive data and directory 425 1451 161 425 1451 162 This text and the full directory of used MF/MT sections. 425 1451 163 425 1451 164 ##### MF3: REACTION CROSS SECTIONS 425 1451 165 425 1451 166 All the data present in the following MT-sections have been 425 1451 167 calculated with TALYS. If the maximal cross section in an 425 1451 168 excitation function over the whole energy range does not exceed 425 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 425 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 425 1451 171 are set to zero. The minor reaction channels are also present 425 1451 172 but they are stored in MT5 and can be reproduced in combination 425 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 425 1451 174 The following reaction channels/MT numbers are generally included: 425 1451 175 425 1451 176 - MT2 : Elastic scattering cross section: nuclear + 425 1451 177 interference terms 425 1451 178 425 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 425 1451 180 distributions of MF=6. Note that because of the interference 425 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 425 1451 182 some energies and angles. 425 1451 183 425 1451 184 - MT5 : (p,anything) cross section 425 1451 185 425 1451 186 MT5 contains the reactions which can not be stored in any other 425 1451 187 MT-number. As the incident energy increases, the cross section 425 1451 188 of MT5 increases as more reaction channels are no longer stored 425 1451 189 under a specific MT number. The information of MF3/MT5 can be 425 1451 190 combined with MF6/MT5 to obtain residual production cross 425 1451 191 sections particle production cross sections and 425 1451 192 (double-)differential cross sections. 425 1451 193 425 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 425 1451 195 425 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 425 1451 197 distributions, as well as all residual and photon production 425 1451 198 cross sections. All data are generated with TALYS. 425 1451 199 425 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 425 1451 201 interference terms 425 1451 202 425 1451 203 Relative angular distributions are tabulated on an angular grid. 425 1451 204 They are obtained by using the "nuclear-plus-interference" option 425 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 425 1451 206 integrated cross section is stored in MF=3. Note that because 425 1451 207 of the interference effect, the tabulations in both MF=6 and 425 1451 208 MF=3 can be negative at some energies and angles. 425 1451 209 425 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 425 1451 211 425 1451 212 MT5 contains the production yields of particles and residual 425 1451 213 products. It also contains the secondary energy-angle 425 1451 214 distributions for all particles and photons. First, the yields 425 1451 215 for neutrons are given for the whole energy range. Next, on a 425 1451 216 secondary energy grid the relative emission spectra are given 425 1451 217 together with the parameters for the Kalbach systematics for 425 1451 218 angular distributions. Inelastic scattering cross sections for 425 1451 219 discrete states have been broadened and added to the continuum 425 1451 220 spectra. This procedure is repeated for protons, deuterons, 425 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 425 1451 222 residual production yields are given per final product. All these 425 1451 223 yields and relative distributions can be multiplied with the 425 1451 224 cross sections given in MF3/MT5 to get the production cross 425 1451 225 sections and (double-)differential cross sections. 425 1451 226 425 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 **** 425 1451 228 425 1451 229 This file has been checked successfully by the BNL checking 425 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 425 1451 231 been processed successfully into an MCNP library by the 425 1451 232 processing code NJOY99.90 [mac00]. 425 1451 233 425 1451 234 *********************** R E F E R E N C E S ********************** 425 1451 235 425 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 425 1451 237 (1985). 425 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 425 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 425 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 425 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 425 1451 242 (2003). 425 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 425 1451 244 (2004). 425 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 425 1451 246 Proceedings of the International Conference on Nuclear 425 1451 247 Data for Science and Technology - ND-2007, 425 1451 248 April 22-27, 2007, Nice, France 425 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 425 1451 250 local level density models'', Nucl Phys. A810, 425 1451 251 13-76 (2008). 425 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 425 1451 253 pointwise and multigroup neutron and photon cross 425 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 425 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 425 1451 256 No. CEA-N-2772, 1994. 425 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 425 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 425 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 425 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 425 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 425 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 425 1451 263 RIPL - Reference Input Parameter Library for calculation 425 1451 264 of nuclear reactions and nuclear data evaluation, 425 1451 265 Nucl. Data Sheets 110, 3107 (2009). 425 1451 266 425 1451 267 ************************* C O N T E N T S ************************ 425 1451 268 425 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 425 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 425 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 425 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 425 1451 273 Unshielded Group Averages Using 162 Groups 425 1451 274 Weighting Spectrum: 1/E Spectrum 425 1451 275 1 451 275 1 425 1451 276 3 2 18 1 425 1451 277 3 5 18 1 425 1451 278 6 2 1401 1 425 1451 279 6 5 12619 1 425 1451 280 8 5 19 1 425 1451 281 425 1 0 282 425 0 0 283 4009.00000 8.93476400 0 0 0 0 425 3 2 284 0.0 0.0 0 0 1 52 425 3 2 285 52 1 425 3 2 286 10000000.0-.001597021 11000000.0 .008606958 12000000.0 .016567794 425 3 2 287 13000000.0 .022848647 14000000.0 .027899184 15000000.0 .032042253 425 3 2 288 16000000.0 .035487598 17000000.0 .038370189 18000000.0 .040792842 425 3 2 289 19000000.0 .042841668 20000000.0 .045278992 22000000.0 .048013803 425 3 2 290 24000000.0 .050282877 26000000.0 .052282669 28000000.0 .054101624 425 3 2 291 30000000.0 .056903388 35000000.0 .060699371 40000000.0 .064193195 425 3 2 292 45000000.0 .067361438 50000000.0 .069880023 54000000.0 .071162524 425 3 2 293 55000000.0 .072426945 60000000.0 .074828247 70000000.0 .076335642 425 3 2 294 80000000.0 .075753599 90000000.0 .073586441 100000000. .067842835 425 3 2 295 120000000. .057323424 140000000. .045805096 160000000. .035357871 425 3 2 296 180000000. .027124116 200000000. .023671210 240000000. .023671210 425 3 2 297 280000000. .023671210 320000000. .023671210 360000000. .023671210 425 3 2 298 400000000. .023671210 440000000. .023671210 480000000. .023671210 425 3 2 299 520000000. .023671210 560000000. .023671210 600000000. .023671210 425 3 2 300 640000000. .023671210 680000000. .023671210 720000000. .023671210 425 3 2 301 760000000. .023671210 800000000. .023671210 840000000. .023671210 425 3 2 302 880000000. .023671210 920000000. .023671210 960000000. .023671210 425 3 2 303 1.00000E+9 0.0 425 3 2 304 425 3 0 305 4009.00000 8.93476400 0 0 0 0 425 3 5 306 0.0 0.0 0 0 1 106 425 3 5 307 106 1 425 3 5 308 1000000.00 .614279300 1125000.00 .651649184 1250000.00 .689011697 425 3 5 309 1375000.00 .726368766 1500000.00 .763721700 1625000.00 .801071420 425 3 5 310 1750000.00 .838418591 1875000.00 .875763707 2000000.00 .900683311 425 3 5 311 2125000.00 .912916831 2250000.00 .925149968 2375000.00 .937382782 425 3 5 312 2500000.00 .949615320 2625000.00 .961847621 2750000.00 .974079717 425 3 5 313 2875000.00 .986311632 3000000.00 .993297457 3125000.00 .994963929 425 3 5 314 3250000.00 .996630385 3375000.00 .998296825 3500000.00 .999963253 425 3 5 315 3625000.00 1.00162967 3750000.00 1.00329607 3875000.00 1.00496247 425 3 5 316 4000000.00 1.00560163 4200000.00 1.00520155 4400000.00 1.00480148 425 3 5 317 4600000.00 1.00440142 4800000.00 1.00400136 5000000.00 1.00313835 425 3 5 318 5200000.00 1.00180619 5400000.00 1.00047404 5600000.00 .999141895 425 3 5 319 5800000.00 .997809763 6000000.00 .996233988 6200000.00 .994411850 425 3 5 320 6400000.00 .992589722 6600000.00 .990767603 6800000.00 .988945491 425 3 5 321 7000000.00 .986999949 7200000.00 .984929796 7400000.00 .982859650 425 3 5 322 7600000.00 .980789511 7800000.00 .978719377 8000000.00 .976643294 425 3 5 323 8200000.00 .974561211 8400000.00 .972479132 8600000.00 .970397059 425 3 5 324 8800000.00 .968314989 9000000.00 .966213973 9200000.00 .964093869 425 3 5 325 9400000.00 .961973769 9600000.00 .959853673 9800000.00 .957733579 425 3 5 326 10000000.0 .950908042 11000000.0 .938938646 12000000.0 .926894034 425 3 5 327 13000000.0 .915452988 14000000.0 .904838639 15000000.0 .894911842 425 3 5 328 16000000.0 .885313244 17000000.0 .875765438 18000000.0 .871400900 425 3 5 329 19000000.0 .866927529 20000000.0 .852846279 22000000.0 .835182285 425 3 5 330 24000000.0 .819165960 26000000.0 .804532131 28000000.0 .790771978 425 3 5 331 30000000.0 .769248820 35000000.0 .740701289 40000000.0 .715658890 425 3 5 332 45000000.0 .693577201 50000000.0 .675647842 54000000.0 .666222720 425 3 5 333 55000000.0 .655993010 60000000.0 .632509795 70000000.0 .604568548 425 3 5 334 80000000.0 .580255799 90000000.0 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.310196168 2000000.00 .334348855 42510 5 352 2125000.00 .340887828 2250000.00 .347426597 2375000.00 .353965193 42510 5 353 2500000.00 .360503642 2625000.00 .367041964 2750000.00 .373580176 42510 5 354 2875000.00 .380118291 3000000.00 .385503391 3125000.00 .389719379 42510 5 355 3250000.00 .393935324 3375000.00 .398151232 3500000.00 .402367106 42510 5 356 3625000.00 .406582949 3750000.00 .410798765 3875000.00 .415014557 42510 5 357 4000000.00 .418313819 4200000.00 .420694416 4400000.00 .423074973 42510 5 358 4600000.00 .425455496 4800000.00 .427835988 5000000.00 .430875455 42510 5 359 5200000.00 .434582738 5400000.00 .438289987 5600000.00 .441997206 42510 5 360 5800000.00 .445704398 6000000.00 .449087967 6200000.00 .452144301 42510 5 361 6400000.00 .455200619 6600000.00 .458256923 6800000.00 .461313212 42510 5 362 7000000.00 .464643760 7200000.00 .468251190 7400000.00 .471858607 42510 5 363 7600000.00 .475466013 7800000.00 .479073409 8000000.00 .482781865 42510 5 364 8200000.00 .486592227 8400000.00 .490402581 8600000.00 .494212926 42510 5 365 8800000.00 .498023264 9000000.00 .501300436 9200000.00 .504040480 42510 5 366 9400000.00 .506780519 9600000.00 .509520554 9800000.00 .512260585 42510 5 367 10000000.0 .520403249 11000000.0 .534461343 12000000.0 .548714630 42510 5 368 13000000.0 .562444012 14000000.0 .574665708 15000000.0 .583369362 42510 5 369 16000000.0 .587894664 17000000.0 .584914239 18000000.0 .525360358 42510 5 370 19000000.0 .478268034 20000000.0 .494085886 22000000.0 .510886495 42510 5 371 24000000.0 .527273607 26000000.0 .539845928 28000000.0 .555281376 42510 5 372 30000000.0 .593570326 35000000.0 .654914554 40000000.0 .704971927 42510 5 373 45000000.0 .744233086 50000000.0 .764922998 54000000.0 .763999897 42510 5 374 55000000.0 .756505093 60000000.0 .771409516 70000000.0 .791443578 42510 5 375 80000000.0 .801918602 90000000.0 .808267750 100000000. .795416368 42510 5 376 120000000. .796895667 140000000. .780017280 160000000. .776448870 42510 5 377 180000000. .758678511 200000000. .749010237 240000000. .749010237 42510 5 378 280000000. .749010237 320000000. .749010237 360000000. .749010237 42510 5 379 400000000. .749010237 440000000. .749010237 480000000. .749010237 42510 5 380 520000000. .749010237 560000000. .749010237 600000000. .749010237 42510 5 381 640000000. .749010237 680000000. .749010237 720000000. .749010237 42510 5 382 760000000. .749010237 800000000. .749010237 840000000. .749010237 42510 5 383 880000000. .749010237 920000000. .749010237 960000000. .749010237 42510 5 384 1.00000E+9 0.0 42510 5 385 0.0 0.0 1001 0 1 106 42510 5 386 106 1 42510 5 387 1000000.00 .003235733 1125000.00 .009843579 1250000.00 .016450122 42510 5 388 1375000.00 .023055702 1500000.00 .029660551 1625000.00 .036264832 42510 5 389 1750000.00 .042868662 1875000.00 .049472128 2000000.00 .054215939 42510 5 390 2125000.00 .057061106 2250000.00 .059906183 2375000.00 .062751185 42510 5 391 2500000.00 .065596123 2625000.00 .068441006 2750000.00 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