TENDL-2011 gen. purp. file: proton + 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 .998620000 200000000. 1 0 10010 0 425 1451 3 0.0 0.0 1 0 271 6 425 1451 4 4-Be- 9 NRG EVAL-OCT11 A.J. Koning 425 1451 5 TENDL-2011 DIST- REV1- 425 1451 6 ----TENDL-2011 Material 425 REVISION 1 425 1451 7 -----Incident proton 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 10882 1 425 1451 280 8 5 16 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 77 425 3 2 285 77 1 425 3 2 286 5000000.00-.015249062 5200000.00-.011269082 5400000.00-.007289137 425 3 2 287 5600000.00-.003309225 5800000.00 6.70659E-4 6000000.00 .004396305 425 3 2 288 6200000.00 .007864876 6400000.00 .011333429 6600000.00 .014801966 425 3 2 289 6800000.00 .018270487 7000000.00 .021451124 7200000.00 .024341126 425 3 2 290 7400000.00 .027231118 7600000.00 .030121101 7800000.00 .033011075 425 3 2 291 8000000.00 .035599181 8200000.00 .037882894 8400000.00 .040166602 425 3 2 292 8600000.00 .042450305 8800000.00 .044734003 9000000.00 .046746687 425 3 2 293 9200000.00 .048486341 9400000.00 .050225993 9600000.00 .051965642 425 3 2 294 9800000.00 .053705288 10000000.0 .057774832 11000000.0 .063450452 425 3 2 295 12000000.0 .067663983 13000000.0 .070804594 14000000.0 .073160032 425 3 2 296 15000000.0 .074931131 16000000.0 .076255157 17000000.0 .077224492 425 3 2 297 18000000.0 .077901943 19000000.0 .078331667 20000000.0 .078514166 425 3 2 298 22000000.0 .078251402 24000000.0 .077422868 26000000.0 .076154507 425 3 2 299 28000000.0 .074550663 30000000.0 .071152208 35000000.0 .065730409 425 3 2 300 40000000.0 .060024702 45000000.0 .054405096 50000000.0 .049572784 425 3 2 301 54000000.0 .046974195 55000000.0 .044127138 60000000.0 .037646515 425 3 2 302 70000000.0 .030277535 80000000.0 .024491521 90000000.0 .019868862 425 3 2 303 100000000. .014923468 120000000. .010448380 140000000. .007744026 425 3 2 304 160000000. .006144552 180000000. .005137684 200000000. .004753641 425 3 2 305 240000000. .004753641 280000000. .004753641 320000000. .004753641 425 3 2 306 360000000. .004753641 400000000. .004753641 440000000. .004753641 425 3 2 307 480000000. .004753641 520000000. .004753641 560000000. .004753641 425 3 2 308 600000000. .004753641 640000000. .004753641 680000000. .004753641 425 3 2 309 720000000. .004753641 760000000. .004753641 800000000. .004753641 425 3 2 310 840000000. .004753641 880000000. .004753641 920000000. .004753641 425 3 2 311 960000000. .004753641 1.00000E+9 0.0 425 3 2 312 425 3 0 313 4009.00000 8.93476400 0 0 0 0 425 3 5 314 0.0 0.0 0 0 1 106 425 3 5 315 106 1 425 3 5 316 1000000.00 .416960782 1125000.00 .422619130 1250000.00 .428276362 425 3 5 317 1375000.00 .433932769 1500000.00 .439588551 1625000.00 .445243845 425 3 5 318 1750000.00 .450898754 1875000.00 .456553352 2000000.00 .462098806 425 3 5 319 2125000.00 .467532875 2250000.00 .472966775 2375000.00 .478400530 425 3 5 320 2500000.00 .483834163 2625000.00 .489267691 2750000.00 .494701127 425 3 5 321 2875000.00 .500134483 3000000.00 .506199691 3125000.00 .512905586 425 3 5 322 3250000.00 .519611413 3375000.00 .526317180 3500000.00 .533022893 425 3 5 323 3625000.00 .539728558 3750000.00 .546434179 3875000.00 .553139761 425 3 5 324 4000000.00 .561608205 4200000.00 .571890149 4400000.00 .582171921 425 3 5 325 4600000.00 .592453542 4800000.00 .602735030 5000000.00 .611462625 425 3 5 326 5200000.00 .618615507 5400000.00 .625768325 5600000.00 .632921086 425 3 5 327 5800000.00 .640073795 6000000.00 .645145890 6200000.00 .648114148 425 3 5 328 6400000.00 .651082389 6600000.00 .654050616 6800000.00 .657018831 425 3 5 329 7000000.00 .658574695 7200000.00 .658704703 7400000.00 .658834711 425 3 5 330 7600000.00 .658964719 7800000.00 .659094726 8000000.00 .658306537 425 3 5 331 8200000.00 .656592472 8400000.00 .654878411 8600000.00 .653164354 425 3 5 332 8800000.00 .651450300 9000000.00 .649329724 9200000.00 .646799604 425 3 5 333 9400000.00 .644269489 9600000.00 .641739377 9800000.00 .639209269 425 3 5 334 10000000.0 .630972498 11000000.0 .617019521 12000000.0 .603957502 425 3 5 335 13000000.0 .592573974 14000000.0 .582931166 15000000.0 .574314744 425 3 5 336 16000000.0 .565947485 17000000.0 .557594485 18000000.0 .556467202 425 3 5 337 19000000.0 .555133340 20000000.0 .542254843 22000000.0 .525381366 425 3 5 338 24000000.0 .508715514 26000000.0 .492453419 28000000.0 .476867550 425 3 5 339 30000000.0 .452280191 35000000.0 .419726177 40000000.0 .391368474 425 3 5 340 45000000.0 .366636991 50000000.0 .346973374 54000000.0 .336801162 425 3 5 341 55000000.0 .326058326 60000000.0 .302327451 70000000.0 .276106779 425 3 5 342 80000000.0 .255665068 90000000.0 .239054657 100000000. .220373893 425 3 5 343 120000000. .201860758 140000000. .188768187 160000000. .179192625 425 3 5 344 180000000. .171603903 200000000. .168190000 240000000. .168190000 425 3 5 345 280000000. .168190000 320000000. .168190000 360000000. .168190000 425 3 5 346 400000000. .168190000 440000000. .168190000 480000000. .168190000 425 3 5 347 520000000. .168190000 560000000. .168190000 600000000. .168190000 425 3 5 348 640000000. .168190000 680000000. .168190000 720000000. .168190000 425 3 5 349 760000000. .168190000 800000000. .168190000 840000000. .168190000 425 3 5 350 880000000. .168190000 920000000. .168190000 960000000. .168190000 425 3 5 351 1.00000E+9 0.0 425 3 5 352 425 3 0 353 425 0 0 354 4009.00000 8.93476400 0 0 16 0 42510 5 355 0.0 0.0 1 0 1 98 42510 5 356 98 1 42510 5 357 2000000.00 .003939411 2125000.00 .011900947 2250000.00 .019862235 42510 5 358 2375000.00 .027823312 2500000.00 .035784210 2625000.00 .043744953 42510 5 359 2750000.00 .051705563 2875000.00 .059666055 3000000.00 .063668704 42510 5 360 3125000.00 .063658237 3250000.00 .063647769 3375000.00 .063637302 42510 5 361 3500000.00 .063626834 3625000.00 .063616367 3750000.00 .063605900 42510 5 362 3875000.00 .063595433 4000000.00 .065542182 4200000.00 .069478951 42510 5 363 4400000.00 .073415654 4600000.00 .077352300 4800000.00 .081288894 42510 5 364 5000000.00 .085979060 5200000.00 .091432906 5400000.00 .096886703 42510 5 365 5600000.00 .102340456 5800000.00 .107794170 6000000.00 .111306404 42510 5 366 6200000.00 .112855488 6400000.00 .114404563 6600000.00 .115953631 42510 5 367 6800000.00 .117502692 7000000.00 .119297489 7200000.00 .121340373 42510 5 368 7400000.00 .123383251 7600000.00 .125426122 7800000.00 .127468987 42510 5 369 8000000.00 .128837561 8200000.00 .129526205 8400000.00 .130214847 42510 5 370 8600000.00 .130903487 8800000.00 .131592127 9000000.00 .132525248 42510 5 371 9200000.00 .133704669 9400000.00 .134884087 9600000.00 .136063504 42510 5 372 9800000.00 .137242919 10000000.0 .138793572 11000000.0 .139606524 42510 5 373 12000000.0 .138412382 13000000.0 .134474665 14000000.0 .127332345 42510 5 374 15000000.0 .122364295 16000000.0 .119555847 17000000.0 .116083795 42510 5 375 18000000.0 .077556585 19000000.0 .042579961 20000000.0 .051634134 42510 5 376 22000000.0 .069193854 24000000.0 .095746484 26000000.0 .127126110 42510 5 377 28000000.0 .156477317 30000000.0 .193524442 35000000.0 .232420518 42510 5 378 40000000.0 .252924918 45000000.0 .259523348 50000000.0 .258231364 42510 5 379 54000000.0 .256100836 55000000.0 .253067321 60000000.0 .244680097 42510 5 380 70000000.0 .233337750 80000000.0 .224994260 90000000.0 .218190791 42510 5 381 100000000. .208308245 120000000. .199247370 140000000. .193881467 42510 5 382 160000000. .190671790 180000000. .187511964 200000000. .185816312 42510 5 383 240000000. .185816312 280000000. .185816312 320000000. .185816312 42510 5 384 360000000. .185816312 400000000. .185816312 440000000. .185816312 42510 5 385 480000000. .185816312 520000000. .185816312 560000000. .185816312 42510 5 386 600000000. .185816312 640000000. .185816312 680000000. .185816312 42510 5 387 720000000. .185816312 760000000. .185816312 800000000. .185816312 42510 5 388 840000000. .185816312 880000000. .185816312 920000000. .185816312 42510 5 389 960000000. .185816312 1.00000E+9 0.0 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