TENDL-2011 gen. purp. file: proton + Be- 10 1 MeV - 200 MeV 99 0 0 0 4010.00000 9.92751200 -1 0 0 1 428 1451 1 0.0 0.0 0 0 0 6 428 1451 2 .998620000 200000000. 1 0 10010 0 428 1451 3 0.0 0.0 1 0 268 6 428 1451 4 4-Be- 10 NRG EVAL-OCT11 A.J. Koning 428 1451 5 TENDL-2011 DIST- REV1- 428 1451 6 ----TENDL-2011 Material 428 REVISION 1 428 1451 7 -----Incident proton data 428 1451 8 ------ENDF-6 Format 428 1451 9 428 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 428 1451 11 428 1451 12 Charged particle transport library made by TALYS 428 1451 13 428 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 428 1451 15 428 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 428 1451 17 428 1451 18 This evaluated data file is based primarily on the nuclear model 428 1451 19 code TALYS [kon07], version 1.4. 428 1451 20 It is part of a large collection of isotopic 428 1451 21 evaluations, all created by running TALYS with either default or 428 1451 22 adjusted input parameters. This means that the data in this 428 1451 23 evaluation have been tested in detail for some isotopes against 428 1451 24 individual experimental data, while for other isotopes it is 428 1451 25 only as good as the global quality of TALYS at the present moment. 428 1451 26 The mutual quality of all individual evaluations is however 428 1451 27 consistent: The same set of nuclear models is used and, equally 428 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 428 1451 29 The resulting data file provides a complete representation of 428 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 428 1451 31 and shielding applications over the incident projectile energy 428 1451 32 range from 1.0E-11 to 200 MeV. 428 1451 33 428 1451 34 All transport data for neutrons, other particles, photons and 428 1451 35 residual nuclides are filed using a combination of MF1,3, 428 1451 36 and MF6. This includes cross sections, angular distributions, 428 1451 37 double-differential spectra, production cross sections, 428 1451 38 residual production (activation) cross sections and recoils. 428 1451 39 This evaluation can thus be used as both transport and activation 428 1451 40 library. The data file has been created automatically using the 428 1451 41 ENDF-6 format generator TEFAL. 428 1451 42 428 1451 43 ##### ORIGIN 428 1451 44 428 1451 45 All data < 200 MeV : Produced with TALYS code 428 1451 46 428 1451 47 *************************** T H E O R Y ************************** 428 1451 48 428 1451 49 TALYS is a computer code system for the prediction and analysis 428 1451 50 of nuclear reactions. TALYS simulates reactions that involve 428 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 428 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 428 1451 53 target nuclides of mass 12 and heavier. This is achieved by 428 1451 54 implementing a suite of nuclear reaction models into a single 428 1451 55 code system. It enables to evaluate nuclear reactions from 428 1451 56 the unresolved resonance region up to intermediate energies. This 428 1451 57 evaluation is based on a theoretical analysis that utilizes the 428 1451 58 optical model, compound nucleus statistical theory, fission, 428 1451 59 direct reactions and pre-equilibrium processes, in combination 428 1451 60 with databases and models for nuclear structure. The following 428 1451 61 output of TALYS is stored in this data file: 428 1451 62 428 1451 63 - Elastic and non-elastic cross sections 428 1451 64 - Elastic scattering angular distributions 428 1451 65 - Residual production cross sections 428 1451 66 - Recoil data 428 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 428 1451 68 - Total particle energy spectra 428 1451 69 - Total particle double-differential spectra 428 1451 70 428 1451 71 Here follows a short description of the used nuclear models: 428 1451 72 428 1451 73 ##### OPTICAL MODEL 428 1451 74 428 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 428 1451 76 in TALYS used as a subroutine. The default optical model 428 1451 77 potentials (OMP) used are the local and global parameterizations 428 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 428 1451 79 for neutrons and protons which in principle are valid over the 428 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 428 1451 81 validity may differ from nucleus to nucleus. For neutrons and 428 1451 82 protons, the used parameterization is given in Eq. (7) of 428 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 428 1451 84 used in the case of deformed nuclides. 428 1451 85 To calculate the transmission coefficients and reaction 428 1451 86 cross sections for deuterons, tritons, helions and alpha 428 1451 87 particles, we use OMPs that are directly derived from our nucleon 428 1451 88 potentials using Watanabe's folding approach. 428 1451 89 428 1451 90 ##### DIRECT REACTIONS 428 1451 91 428 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 428 1451 93 calculations for rotational or vibrational (or a combination of 428 1451 94 these) nuclides, depending on the information present in the 428 1451 95 nuclear structure database of TALYS. 428 1451 96 In addition, a macroscopic, phenomenological model to describe 428 1451 97 giant resonances in the inelastic channel is used. For each 428 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 428 1451 99 calculation with ECIS-06 is performed for each giant resonance 428 1451 100 state. The cross section is then spread over the continuum with a 428 1451 101 Gaussian distribution. 428 1451 102 428 1451 103 ##### COMPOUND NUCLEUS 428 1451 104 428 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 428 1451 106 model. The transmission coefficients have been generated 428 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 428 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 428 1451 109 For each nucleus that can be reached through a binary reaction, 428 1451 110 several discrete levels and a continuum described by level 428 1451 111 densities are included simultaneously as competing channels. 428 1451 112 428 1451 113 For each residual nucleus several discrete states 428 1451 114 are included as well as a continuum described by level densities. 428 1451 115 Multiple compound emission is continued until all reaction 428 1451 116 channels are closed and the population distribution of all 428 1451 117 residual nuclides is depleted, through gamma decay, until they 428 1451 118 end up in the ground state or in an isomer. 428 1451 119 428 1451 120 For the level density, we take the Constant Temperature Model 428 1451 121 with level density parameters as given in [kon08]. This employs 428 1451 122 the temperature law at low energies and a Fermi gas expression at 428 1451 123 high energies, and takes into account the damping of shell 428 1451 124 effects at high excitation energy. We have obtained the level 428 1451 125 density parameters from a simultaneous fit to all experimental 428 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 428 1451 127 428 1451 128 Gamma-ray transmission coefficients are generated with the 428 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 428 1451 130 giant dipole resonance parameters taken from the RIPL library 428 1451 131 [rip09], and normalized with experimental radiative widths. 428 1451 132 428 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 428 1451 134 428 1451 135 For pre-equilibrium reactions, which become important for 428 1451 136 incident energies above about 10 MeV, we use the two-component 428 1451 137 exciton model [kon04], in which the neutron or proton types of 428 1451 138 particles and holes are followed throughout the reaction. For 428 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 428 1451 140 any order of particle emission was included in the calculations. 428 1451 141 A parameterization for the squared matrix element is used that is 428 1451 142 valid for the whole energy range of this evaluation. 428 1451 143 428 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 428 1451 145 contribution was added from the pick/up and knock-out reaction 428 1451 146 model by Kalbach [kal05]. 428 1451 147 428 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 428 1451 149 applied, to simulate the direct and semi-direct capture 428 1451 150 processes. 428 1451 151 428 1451 152 The angular distribution systematics by Kalbach [kal88] were used 428 1451 153 to describe the angular distributions for all continuum 428 1451 154 particles. An isotopic distribution for photons was adopted. 428 1451 155 428 1451 156 ***************** F I L E I N F O R M A T I O N **************** 428 1451 157 428 1451 158 ##### MF1: GENERAL INFORMATION 428 1451 159 428 1451 160 - MT451 : Descriptive data and directory 428 1451 161 428 1451 162 This text and the full directory of used MF/MT sections. 428 1451 163 428 1451 164 ##### MF3: REACTION CROSS SECTIONS 428 1451 165 428 1451 166 All the data present in the following MT-sections have been 428 1451 167 calculated with TALYS. If the maximal cross section in an 428 1451 168 excitation function over the whole energy range does not exceed 428 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 428 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 428 1451 171 are set to zero. The minor reaction channels are also present 428 1451 172 but they are stored in MT5 and can be reproduced in combination 428 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 428 1451 174 The following reaction channels/MT numbers are generally included: 428 1451 175 428 1451 176 - MT2 : Elastic scattering cross section: nuclear + 428 1451 177 interference terms 428 1451 178 428 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 428 1451 180 distributions of MF=6. Note that because of the interference 428 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 428 1451 182 some energies and angles. 428 1451 183 428 1451 184 - MT5 : (p,anything) cross section 428 1451 185 428 1451 186 MT5 contains the reactions which can not be stored in any other 428 1451 187 MT-number. As the incident energy increases, the cross section 428 1451 188 of MT5 increases as more reaction channels are no longer stored 428 1451 189 under a specific MT number. The information of MF3/MT5 can be 428 1451 190 combined with MF6/MT5 to obtain residual production cross 428 1451 191 sections particle production cross sections and 428 1451 192 (double-)differential cross sections. 428 1451 193 428 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 428 1451 195 428 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 428 1451 197 distributions, as well as all residual and photon production 428 1451 198 cross sections. All data are generated with TALYS. 428 1451 199 428 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 428 1451 201 interference terms 428 1451 202 428 1451 203 Relative angular distributions are tabulated on an angular grid. 428 1451 204 They are obtained by using the "nuclear-plus-interference" option 428 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 428 1451 206 integrated cross section is stored in MF=3. Note that because 428 1451 207 of the interference effect, the tabulations in both MF=6 and 428 1451 208 MF=3 can be negative at some energies and angles. 428 1451 209 428 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 428 1451 211 428 1451 212 MT5 contains the production yields of particles and residual 428 1451 213 products. It also contains the secondary energy-angle 428 1451 214 distributions for all particles and photons. First, the yields 428 1451 215 for neutrons are given for the whole energy range. Next, on a 428 1451 216 secondary energy grid the relative emission spectra are given 428 1451 217 together with the parameters for the Kalbach systematics for 428 1451 218 angular distributions. Inelastic scattering cross sections for 428 1451 219 discrete states have been broadened and added to the continuum 428 1451 220 spectra. This procedure is repeated for protons, deuterons, 428 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 428 1451 222 residual production yields are given per final product. All these 428 1451 223 yields and relative distributions can be multiplied with the 428 1451 224 cross sections given in MF3/MT5 to get the production cross 428 1451 225 sections and (double-)differential cross sections. 428 1451 226 428 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 **** 428 1451 228 428 1451 229 This file has been checked successfully by the BNL checking 428 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 428 1451 231 been processed successfully into an MCNP library by the 428 1451 232 processing code NJOY99.90 [mac00]. 428 1451 233 428 1451 234 *********************** R E F E R E N C E S ********************** 428 1451 235 428 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 428 1451 237 (1985). 428 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 428 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 428 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 428 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 428 1451 242 (2003). 428 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 428 1451 244 (2004). 428 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 428 1451 246 Proceedings of the International Conference on Nuclear 428 1451 247 Data for Science and Technology - ND-2007, 428 1451 248 April 22-27, 2007, Nice, France 428 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 428 1451 250 local level density models'', Nucl Phys. A810, 428 1451 251 13-76 (2008). 428 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 428 1451 253 pointwise and multigroup neutron and photon cross 428 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 428 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 428 1451 256 No. CEA-N-2772, 1994. 428 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 428 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 428 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 428 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 428 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 428 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 428 1451 263 RIPL - Reference Input Parameter Library for calculation 428 1451 264 of nuclear reactions and nuclear data evaluation, 428 1451 265 Nucl. Data Sheets 110, 3107 (2009). 428 1451 266 428 1451 267 ************************* C O N T E N T S ************************ 428 1451 268 428 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 428 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 428 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 428 1451 272 1 451 275 1 428 1451 273 3 2 18 1 428 1451 274 3 5 18 1 428 1451 275 6 2 1401 1 428 1451 276 6 5 11807 1 428 1451 277 8 5 19 1 428 1451 278 428 1 0 279 428 0 0 280 4010.00000 9.92751200 0 0 0 0 428 3 2 281 0.000000+0 0.000000+0 0 0 1 45 428 3 2 282 45 2 428 3 2 283 1.000000+6-2.141663-1 2.000000+6-1.759580-1 3.000000+6-1.059698-1 428 3 2 284 4.000000+6-6.346307-2 5.000000+6-3.145667-2 6.000000+6-5.197597-3 428 3 2 285 7.000000+6 1.569235-2 8.000000+6 3.160147-2 9.000000+6 4.343717-2 428 3 2 286 1.000000+7 5.222244-2 1.100000+7 5.884425-2 1.200000+7 6.396773-2 428 3 2 287 1.300000+7 6.804816-2 1.400000+7 7.137885-2 1.500000+7 7.414017-2 428 3 2 288 1.600000+7 7.643880-2 1.700000+7 7.834576-2 1.800000+7 7.990368-2 428 3 2 289 1.900000+7 8.114620-2 2.000000+7 8.210231-2 2.200000+7 8.326000-2 428 3 2 290 2.400000+7 8.357327-2 2.600000+7 8.321550-2 2.800000+7 8.233194-2 428 3 2 291 3.000000+7 8.104145-2 3.500000+7 7.660415-2 4.000000+7 7.118869-2 428 3 2 292 4.500000+7 6.540620-2 5.000000+7 5.962720-2 5.500000+7 5.407107-2 428 3 2 293 6.000000+7 4.885957-2 6.500000+7 4.405157-2 7.000000+7 3.966408-2 428 3 2 294 7.500000+7 3.569670-2 8.000000+7 3.212886-2 9.000000+7 2.608289-2 428 3 2 295 1.000000+8 2.131040-2 1.100000+8 1.755232-2 1.200000+8 1.460775-2 428 3 2 296 1.300000+8 1.230570-2 1.400000+8 1.050810-2 1.500000+8 9.102316-3 428 3 2 297 1.600000+8 8.007726-3 1.800000+8 6.485017-3 2.000000+8 5.561803-3 428 3 2 298 428 3 0 299 4.010000+3 9.927512+0 0 0 0 0 428 3 5 300 0.000000+0 0.000000+0 0 0 1 45 428 3 5 301 45 2 428 3 5 302 1.000000+6 5.430000-1 2.000000+6 5.540500-1 3.000000+6 6.041800-1 428 3 5 303 4.000000+6 6.637900-1 5.000000+6 7.159501-1 6.000000+6 7.405500-1 428 3 5 304 7.000000+6 7.429100-1 8.000000+6 7.339500-1 9.000000+6 7.193100-1 428 3 5 305 1.000000+7 7.041900-1 1.100000+7 6.901800-1 1.200000+7 6.779300-1 428 3 5 306 1.300000+7 6.671000-1 1.400000+7 6.572000-1 1.500000+7 6.477700-1 428 3 5 307 1.600000+7 6.384700-1 1.700000+7 6.291400-1 1.800000+7 6.197000-1 428 3 5 308 1.900000+7 6.101800-1 2.000000+7 6.006200-1 2.200000+7 5.816800-1 428 3 5 309 2.400000+7 5.632300-1 2.600000+7 5.455900-1 2.800000+7 5.286500-1 428 3 5 310 3.000000+7 5.125500-1 3.500000+7 4.757800-1 4.000000+7 4.434300-1 428 3 5 311 4.500000+7 4.149600-1 5.000000+7 3.898900-1 5.500000+7 3.677800-1 428 3 5 312 6.000000+7 3.482600-1 6.500000+7 3.309800-1 7.000000+7 3.156400-1 428 3 5 313 7.500000+7 3.020100-1 8.000000+7 2.898500-1 9.000000+7 2.692500-1 428 3 5 314 1.000000+8 2.526300-1 1.100000+8 2.390800-1 1.200000+8 2.279200-1 428 3 5 315 1.300000+8 2.186400-1 1.400000+8 2.108200-1 1.500000+8 2.041800-1 428 3 5 316 1.600000+8 1.984500-1 1.800000+8 1.890900-1 2.000000+8 1.816700-1 428 3 5 317 428 3 0 318 428 0 0 319 4010.00000 9.92751200 0 0 19 0 42810 5 320 0.0 0.0 1 0 1 45 42810 5 321 45 2 42810 5 322 1.000000+6 0.0 2.000000+6 .114653445 3.000000+6 .213781843 42810 5 323 4.000000+6 .257221280 5.000000+6 .282767356 6.000000+6 .309290708 42810 5 324 7.000000+6 .316207012 8.000000+6 .332917518 9.000000+6 .346335537 42810 5 325 1.000000+7 .352808344 1.100000+7 .367810726 1.200000+7 .385221520 42810 5 326 1.300000+7 .406511394 1.400000+7 .426821826 1.500000+7 .445998714 42810 5 327 1.600000+7 .463155715 1.700000+7 .477759479 1.800000+7 .489895159 42810 5 328 1.900000+7 .499047917 2.000000+7 .503946007 2.200000+7 .476641389 42810 5 329 2.400000+7 .444396918 2.600000+7 .392744598 2.800000+7 .403999088 42810 5 330 3.000000+7 .404654125 3.500000+7 .420618067 4.000000+7 .443057075 42810 5 331 4.500000+7 .458742430 5.000000+7 .465731403 5.500000+7 .463527845 42810 5 332 6.000000+7 .455088755 6.500000+7 .445025779 7.000000+7 .433080175 42810 5 333 7.500000+7 .421584819 8.000000+7 .410636292 9.000000+7 .390649440 42810 5 334 1.000000+8 .373745875 1.100000+8 .363492450 1.200000+8 .351117598 42810 5 335 1.300000+8 .340713271 1.400000+8 .329165915 1.500000+8 .323723306 42810 5 336 1.600000+8 .315499779 1.800000+8 .305881439 2.000000+8 .298343924 42810 5 337 0.0 0.0 1001 0 1 45 42810 5 338 45 2 42810 5 339 1.000000+6 0.0 2.000000+6 1.3634E-10 3.000000+6 1.04637E-9 42810 5 340 4.000000+6 .002050102 5.000000+6 .033487062 6.000000+6 .039780791 42810 5 341 7.000000+6 .046850876 8.000000+6 .073074411 9.000000+6 .088840539 42810 5 342 1.000000+7 .098677441 1.100000+7 .107389247 1.200000+7 .116458205 42810 5 343 1.300000+7 .124242038 1.400000+7 .129906095 1.500000+7 .134666849 42810 5 344 1.600000+7 .138493082 1.700000+7 .141824514 1.800000+7 .144884001 42810 5 345 1.900000+7 .147988786 2.000000+7 .150891961 2.200000+7 .170965059 42810 5 346 2.400000+7 .189814706 2.600000+7 .174233621 2.800000+7 .181099631 42810 5 347 3.000000+7 .192748015 3.500000+7 .231344695 4.000000+7 .265187103 42810 5 348 4.500000+7 .285087479 5.000000+7 .295495682 5.500000+7 .298967626 42810 5 349 6.000000+7 .297911704 6.500000+7 .296311831 7.000000+7 .293129818 42810 5 350 7.500000+7 .289260346 8.000000+7 .285181966 9.000000+7 .277174028 42810 5 351 1.000000+8 .270324205 1.100000+8 .266863487 1.200000+8 .262983213 42810 5 352 1.300000+8 .260610134 1.400000+8 .257759073 1.500000+8 .257477105 42810 5 353 1.600000+8 .256032252 1.800000+8 .256182914 2.000000+8 .255889462 42810 5 354 0.0 0.0 1002 0 1 45 42810 5 355 45 2 42810 5 356 1.000000+6 0.0 2.000000+6 0.0 3.000000+6 0.0 42810 5 357 4.000000+6 0.0 5.000000+6 0.0 6.000000+6 .018888246 42810 5 358 7.000000+6 .047216982 8.000000+6 .064971382 9.000000+6 .101170232 42810 5 359 1.000000+7 .136099505 1.100000+7 .154567191 1.200000+7 .168216127 42810 5 360 1.300000+7 .178594011 1.400000+7 .184890733 1.500000+7 .188506900 42810 5 361 1.600000+7 .190030380 1.700000+7 .189965677 1.800000+7 .188672623 42810 5 362 1.900000+7 .186161647 2.000000+7 .182157835 2.200000+7 .162425996 42810 5 363 2.400000+7 .140051082 2.600000+7 .196835778 2.800000+7 .188614919 42810 5 364 3.000000+7 .184966994 3.500000+7 .187706153 4.000000+7 .180175364 42810 5 365 4.500000+7 .170122396 5.000000+7 .157537394 5.500000+7 .146963049 42810 5 366 6.000000+7 .137500013 6.500000+7 .129924544 7.000000+7 .124106176 42810 5 367 7.500000+7 .119683241 8.000000+7 .114897989 9.000000+7 .107017721 42810 5 368 1.000000+8 .100434572 1.100000+8 .094462182 1.200000+8 .089559569 42810 5 369 1.300000+8 .085306113 1.400000+8 .081936880 1.500000+8 .078830835 42810 5 370 1.600000+8 .076293507 1.800000+8 .071978621 2.000000+8 .069450079 42810 5 371 0.0 0.0 1003 0 1 45 42810 5 372 45 2 42810 5 373 1.000000+6 0.0 2.000000+6 .094525362 3.000000+6 .089263970 42810 5 374 4.000000+6 .099490837 5.000000+6 .138902911 6.000000+6 .144165090 42810 5 375 7.000000+6 .124664013 8.000000+6 .096521765 9.000000+6 .074539937 42810 5 376 1.000000+7 .057000730 1.100000+7 .044817528 1.200000+7 .035865412 42810 5 377 1.300000+7 .029406302 1.400000+7 .026847014 1.500000+7 .024836279 42810 5 378 1.600000+7 .022835263 1.700000+7 .021008746 1.800000+7 .019217021 42810 5 379 1.900000+7 .017689545 2.000000+7 .018102807 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