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 0 271 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 ***************** Program GROUPIE (VERSION 2011-1) ************** 428 1451 273 Unshielded Group Averages Using 162 Groups 428 1451 274 Weighting Spectrum: 1/E Spectrum 428 1451 275 1 451 275 1 428 1451 276 3 2 18 1 428 1451 277 3 5 18 1 428 1451 278 6 2 1401 1 428 1451 279 6 5 11807 1 428 1451 280 8 5 19 1 428 1451 281 428 1 0 282 428 0 0 283 4010.00000 9.92751200 0 0 0 0 428 3 2 284 0.0 0.0 0 0 1 72 428 3 2 285 72 1 428 3 2 286 6000000.00-.003120018 6200000.00 .001058333 6400000.00 .005236663 428 3 2 287 6600000.00 .009414972 6800000.00 .013593263 7000000.00 .017275793 428 3 2 288 7200000.00 .020457821 7400000.00 .023639839 7600000.00 .026821847 428 3 2 289 7800000.00 .030003845 8000000.00 .032780169 8200000.00 .035147427 428 3 2 290 8400000.00 .037514678 8600000.00 .039881925 8800000.00 .042249167 428 3 2 291 9000000.00 .044312479 9200000.00 .046069602 9400000.00 .047826722 428 3 2 292 9600000.00 .049583840 9800000.00 .051340955 10000000.0 .055480759 428 3 2 293 11000000.0 .061368845 12000000.0 .065980731 13000000.0 .069692938 428 3 2 294 14000000.0 .072743635 15000000.0 .075277123 16000000.0 .077382647 428 3 2 295 17000000.0 .079117300 18000000.0 .080519342 19000000.0 .081620168 428 3 2 296 20000000.0 .082671961 22000000.0 .083414364 24000000.0 .083396771 428 3 2 297 26000000.0 .082779176 28000000.0 .081694114 30000000.0 .078879779 428 3 2 298 35000000.0 .073956663 40000000.0 .068354188 45000000.0 .062567430 428 3 2 299 50000000.0 .057433252 54000000.0 .054628382 55000000.0 .051503104 428 3 2 300 60000000.0 .044274879 70000000.0 .035880483 80000000.0 .029165204 428 3 2 301 90000000.0 .023738540 100000000. .017857759 120000000. .012484580 428 3 2 302 140000000. .009207987 160000000. .007261314 180000000. .006031514 428 3 2 303 200000000. .005561803 240000000. .005561803 280000000. .005561803 428 3 2 304 320000000. .005561803 360000000. .005561803 400000000. .005561803 428 3 2 305 440000000. .005561803 480000000. .005561803 520000000. .005561803 428 3 2 306 560000000. .005561803 600000000. .005561803 640000000. .005561803 428 3 2 307 680000000. .005561803 720000000. .005561803 760000000. .005561803 428 3 2 308 800000000. .005561803 840000000. .005561803 880000000. .005561803 428 3 2 309 920000000. .005561803 960000000. .005561803 1.00000E+9 0.0 428 3 2 310 428 3 0 311 4010.00000 9.92751200 0 0 0 0 428 3 5 312 0.0 0.0 0 0 1 106 428 3 5 313 106 1 428 3 5 314 1000000.00 .543677071 1125000.00 .545059750 1250000.00 .546442156 428 3 5 315 1375000.00 .547824361 1500000.00 .549206413 1625000.00 .550588346 428 3 5 316 1750000.00 .551970184 1875000.00 .553351947 2000000.00 .557151470 428 3 5 317 2125000.00 .563419529 2250000.00 .569687393 2375000.00 .575955091 428 3 5 318 2500000.00 .582222648 2625000.00 .588490084 2750000.00 .594757414 428 3 5 319 2875000.00 .601024652 3000000.00 .607880278 3125000.00 .615332522 428 3 5 320 3250000.00 .622784691 3375000.00 .630236793 3500000.00 .637688836 428 3 5 321 3625000.00 .645140825 3750000.00 .652592765 3875000.00 .660044661 428 3 5 322 4000000.00 .668963597 4200000.00 .679397590 4400000.00 .689831408 428 3 5 323 4600000.00 .700265073 4800000.00 .710698603 5000000.00 .718394010 428 3 5 324 5200000.00 .723314597 5400000.00 .728235140 5600000.00 .733155643 428 3 5 325 5800000.00 .738076111 6000000.00 .740784710 6200000.00 .741256751 428 3 5 326 6400000.00 .741728790 6600000.00 .742200826 6800000.00 .742672860 428 3 5 327 7000000.00 .742018207 7200000.00 .740226092 7400000.00 .738433982 428 3 5 328 7600000.00 .736641879 7800000.00 .734849781 8000000.00 .732492025 428 3 5 329 8200000.00 .729563880 8400000.00 .726635741 8600000.00 .723707609 428 3 5 330 8800000.00 .720779483 9000000.00 .717803539 9200000.00 .714779420 428 3 5 331 9400000.00 .711755305 9600000.00 .708731196 9800000.00 .705707091 428 3 5 332 10000000.0 .697296258 11000000.0 .684143813 12000000.0 .672587231 428 3 5 333 13000000.0 .662211133 14000000.0 .652539213 15000000.0 .643170014 428 3 5 334 16000000.0 .633852133 17000000.0 .624464962 18000000.0 .614982891 428 3 5 335 19000000.0 .605440862 20000000.0 .591300408 22000000.0 .572588763 428 3 5 336 24000000.0 .554527650 26000000.0 .537224606 28000000.0 .520692558 428 3 5 337 30000000.0 .494637156 35000000.0 .459964871 40000000.0 .429474376 428 3 5 338 45000000.0 .402645075 50000000.0 .381159429 54000000.0 .369997762 428 3 5 339 55000000.0 .358161521 60000000.0 .331884347 70000000.0 .302664669 428 3 5 340 80000000.0 .279752147 90000000.0 .261085897 100000000. .240053549 428 3 5 341 120000000. .219224939 140000000. .204545277 160000000. .193861850 428 3 5 342 180000000. .185445136 200000000. .181670000 240000000. .181670000 428 3 5 343 280000000. .181670000 320000000. .181670000 360000000. .181670000 428 3 5 344 400000000. .181670000 440000000. .181670000 480000000. .181670000 428 3 5 345 520000000. .181670000 560000000. .181670000 600000000. .181670000 428 3 5 346 640000000. .181670000 680000000. .181670000 720000000. .181670000 428 3 5 347 760000000. .181670000 800000000. .181670000 840000000. .181670000 428 3 5 348 880000000. .181670000 920000000. .181670000 960000000. .181670000 428 3 5 349 1.00000E+9 0.0 428 3 5 350 428 3 0 351 428 0 0 352 4010.00000 9.92751200 0 0 19 0 42810 5 353 0.0 0.0 1 0 1 106 42810 5 354 106 1 42810 5 355 1000000.00 .007025204 1125000.00 .021371711 1250000.00 .035715389 42810 5 356 1375000.00 .050056977 1500000.00 .064396977 1625000.00 .078735744 42810 5 357 1750000.00 .093073532 1875000.00 .107410531 2000000.00 .120786373 42810 5 358 2125000.00 .133181002 2250000.00 .145575243 2375000.00 .157969157 42810 5 359 2500000.00 .170362791 2625000.00 .182756184 2750000.00 .195149370 42810 5 360 2875000.00 .207542373 3000000.00 .216478337 3125000.00 .221908991 42810 5 361 3250000.00 .227339590 3375000.00 .232770141 3500000.00 .238200648 42810 5 362 3625000.00 .243631116 3750000.00 .249061549 3875000.00 .254491949 42810 5 363 4000000.00 .259755115 4200000.00 .264865297 4400000.00 .269975392 42810 5 364 4600000.00 .275085413 4800000.00 .280195368 5000000.00 .285402354 42810 5 365 5200000.00 .290707679 5400000.00 .296012956 5600000.00 .301318190 42810 5 366 5800000.00 .306623387 6000000.00 .309978559 6200000.00 .311361940 42810 5 367 6400000.00 .312745313 6600000.00 .314128680 6800000.00 .315512040 42810 5 368 7000000.00 .317870217 7200000.00 .321212533 7400000.00 .324554838 42810 5 369 7600000.00 .327897132 7800000.00 .331239416 8000000.00 .334253798 42810 5 370 8200000.00 .336937535 8400000.00 .339621265 8600000.00 .342304990 42810 5 371 8800000.00 .344988709 9000000.00 .346980447 9200000.00 .348275059 42810 5 372 9400000.00 .349569669 9600000.00 .350864278 9800000.00 .352158884 42810 5 373 10000000.0 .360190396 11000000.0 .376389894 12000000.0 .395724464 42810 5 374 13000000.0 .416541191 14000000.0 .436300023 15000000.0 .454484947 42810 5 375 16000000.0 .470383823 17000000.0 .483769517 18000000.0 .494430301 42810 5 376 19000000.0 .501476026 20000000.0 .490510533 22000000.0 .460752927 42810 5 377 24000000.0 .418915254 26000000.0 .398302345 28000000.0 .404322841 42810 5 378 30000000.0 .412431106 35000000.0 .431587953 40000000.0 .450745832 42810 5 379 45000000.0 .462175564 50000000.0 .464861285 54000000.0 .463748875 42810 5 380 55000000.0 .459369484 60000000.0 .444837264 70000000.0 .421971300 42810 5 381 80000000.0 .400838997 90000000.0 .382346044 100000000. .363304971 42810 5 382 120000000. .340709108 140000000. .323179587 160000000. .310784994 42810 5 383 180000000. .302178849 200000000. .298343924 240000000. .298343924 42810 5 384 280000000. .298343924 320000000. .298343924 360000000. .298343924 42810 5 385 400000000. .298343924 440000000. .298343924 480000000. .298343924 42810 5 386 520000000. .298343924 560000000. .298343924 600000000. .298343924 42810 5 387 640000000. .298343924 680000000. .298343924 720000000. .298343924 42810 5 388 760000000. .298343924 800000000. .298343924 840000000. .298343924 42810 5 389 880000000. 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