TENDL-2011 gen. purp. file: deuteron + 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 1.99626000 200000000. 1 0 10020 0 428 1451 3 0.0 0.0 1 0 271 6 428 1451 4 4-Be- 10 NRG EVAL-NOV11 A.J. Koning 428 1451 5 TENDL-2011 DIST- REV1- 428 1451 6 ----TENDL-2011 Material 428 REVISION 1 428 1451 7 -----Incident deuteron 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 12669 1 428 1451 280 8 5 22 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 52 428 3 2 285 52 1 428 3 2 286 10000000.0 .004690313 11000000.0 .014652787 12000000.0 .022407922 428 3 2 287 13000000.0 .028575746 14000000.0 .033583199 15000000.0 .037692011 428 3 2 288 16000000.0 .041067038 17000000.0 .043837674 18000000.0 .046121790 428 3 2 289 19000000.0 .048030274 20000000.0 .050315570 22000000.0 .052921665 428 3 2 290 24000000.0 .055111672 26000000.0 .057028354 28000000.0 .058750566 428 3 2 291 30000000.0 .061408158 35000000.0 .065040503 40000000.0 .068396070 428 3 2 292 45000000.0 .071415757 50000000.0 .073786082 54000000.0 .074980088 428 3 2 293 55000000.0 .076128085 60000000.0 .078203208 70000000.0 .079201454 428 3 2 294 80000000.0 .078059119 90000000.0 .075325078 100000000. .068729296 428 3 2 295 120000000. .057348821 140000000. .045398928 160000000. .034988582 428 3 2 296 180000000. .027190032 200000000. .024052340 240000000. .024052340 428 3 2 297 280000000. .024052340 320000000. .024052340 360000000. .024052340 428 3 2 298 400000000. .024052340 440000000. .024052340 480000000. .024052340 428 3 2 299 520000000. .024052340 560000000. .024052340 600000000. .024052340 428 3 2 300 640000000. .024052340 680000000. .024052340 720000000. .024052340 428 3 2 301 760000000. .024052340 800000000. .024052340 840000000. .024052340 428 3 2 302 880000000. .024052340 920000000. .024052340 960000000. .024052340 428 3 2 303 1.00000E+9 0.0 428 3 2 304 428 3 0 305 4010.00000 9.92751200 0 0 0 0 428 3 5 306 0.0 0.0 0 0 1 106 428 3 5 307 106 1 428 3 5 308 1000000.00 .649192810 1125000.00 .693472323 1250000.00 .737743101 428 3 5 309 1375000.00 .782007429 1500000.00 .826266857 1625000.00 .870522477 428 3 5 310 1750000.00 .914775077 1875000.00 .959025243 2000000.00 .987982610 428 3 5 311 2125000.00 1.00132646 2250000.00 1.01466990 2375000.00 1.02801298 428 3 5 312 2500000.00 1.04135576 2625000.00 1.05469829 2750000.00 1.06804059 428 3 5 313 2875000.00 1.08138269 3000000.00 1.08850349 3125000.00 1.08931609 428 3 5 314 3250000.00 1.09012869 3375000.00 1.09094129 3500000.00 1.09175387 428 3 5 315 3625000.00 1.09256645 3750000.00 1.09337903 3875000.00 1.09419160 428 3 5 316 4000000.00 1.09387594 4200000.00 1.09241566 4400000.00 1.09095541 428 3 5 317 4600000.00 1.08949518 4800000.00 1.08803497 5000000.00 1.08633634 428 3 5 318 5200000.00 1.08439610 5400000.00 1.08245588 5600000.00 1.08051567 428 3 5 319 5800000.00 1.07857548 6000000.00 1.07643639 6200000.00 1.07409619 428 3 5 320 6400000.00 1.07175600 6600000.00 1.06941582 6800000.00 1.06707565 428 3 5 321 7000000.00 1.06467577 7200000.00 1.06221562 7400000.00 1.05975547 428 3 5 322 7600000.00 1.05729532 7800000.00 1.05483519 8000000.00 1.05230535 428 3 5 323 8200000.00 1.04970522 8400000.00 1.04710510 8600000.00 1.04450498 428 3 5 324 8800000.00 1.04190487 9000000.00 1.03918520 9200000.00 1.03634509 428 3 5 325 9400000.00 1.03350498 9600000.00 1.03066488 9800000.00 1.02782478 428 3 5 326 10000000.0 1.01931435 11000000.0 1.00531338 12000000.0 .992362084 428 3 5 327 13000000.0 .980535064 14000000.0 .969373174 15000000.0 .958364318 428 3 5 328 16000000.0 .947266427 17000000.0 .936128016 18000000.0 .925148938 428 3 5 329 19000000.0 .914539374 20000000.0 .899887076 22000000.0 .881605910 428 3 5 330 24000000.0 .865148495 26000000.0 .849635714 28000000.0 .834633245 428 3 5 331 30000000.0 .811258105 35000000.0 .780431932 40000000.0 .753561404 428 3 5 332 45000000.0 .729751724 50000000.0 .710365733 54000000.0 .700158184 428 3 5 333 55000000.0 .689038763 60000000.0 .663428157 70000000.0 .632767900 428 3 5 334 80000000.0 .605860869 90000000.0 .581360027 100000000. .550612697 428 3 5 335 120000000. .515263582 140000000. .484547986 160000000. .459252764 428 3 5 336 180000000. .436304003 200000000. .425060000 240000000. .425060000 428 3 5 337 280000000. .425060000 320000000. .425060000 360000000. .425060000 428 3 5 338 400000000. .425060000 440000000. .425060000 480000000. .425060000 428 3 5 339 520000000. .425060000 560000000. .425060000 600000000. .425060000 428 3 5 340 640000000. .425060000 680000000. .425060000 720000000. .425060000 428 3 5 341 760000000. .425060000 800000000. .425060000 840000000. .425060000 428 3 5 342 880000000. .425060000 920000000. .425060000 960000000. .425060000 428 3 5 343 1.00000E+9 0.0 428 3 5 344 428 3 0 345 428 0 0 346 4010.00000 9.92751200 0 0 22 0 42810 5 347 0.0 0.0 1 0 1 106 42810 5 348 106 1 42810 5 349 1000000.00 .045791835 1125000.00 .139305553 1250000.00 .232800825 42810 5 350 1375000.00 .326282473 1500000.00 .419753775 1625000.00 .513217034 42810 5 351 1750000.00 .606673915 1875000.00 .700125654 2000000.00 .750458558 42810 5 352 2125000.00 .756768081 2250000.00 .763077407 2375000.00 .769386567 42810 5 353 2500000.00 .775695584 2625000.00 .782004478 2750000.00 .788313267 42810 5 354 2875000.00 .794621963 3000000.00 .798316072 3125000.00 .799359083 42810 5 355 3250000.00 .800402083 3375000.00 .801445074 3500000.00 .802488057 42810 5 356 3625000.00 .803531032 3750000.00 .804574000 3875000.00 .805616962 42810 5 357 4000000.00 .804792698 4200000.00 .802073156 4400000.00 .799353659 42810 5 358 4600000.00 .796634203 4800000.00 .793914781 5000000.00 .790371901 42810 5 359 5200000.00 .785994519 5400000.00 .781617175 5600000.00 .777239867 42810 5 360 5800000.00 .772862590 6000000.00 .770011012 6200000.00 .768702164 42810 5 361 6400000.00 .767393322 6600000.00 .766084487 6800000.00 .764775657 42810 5 362 7000000.00 .763621696 7200000.00 .762624084 7400000.00 .761626476 42810 5 363 7600000.00 .760628871 7800000.00 .759631268 8000000.00 .758783411 42810 5 364 8200000.00 .758086550 8400000.00 .757389690 8600000.00 .756692833 42810 5 365 8800000.00 .755995976 9000000.00 .755618206 9200000.00 .755561895 42810 5 366 9400000.00 .755505583 9600000.00 .755449271 9800000.00 .755392960 42810 5 367 10000000.0 .757264034 11000000.0 .764000514 12000000.0 .773083330 42810 5 368 13000000.0 .782210261 14000000.0 .792874537 15000000.0 .804584013 42810 5 369 16000000.0 .816583535 17000000.0 .828659429 18000000.0 .840381539 42810 5 370 19000000.0 .851562550 20000000.0 .867416042 22000000.0 .887203507 42810 5 371 24000000.0 .927104829 26000000.0 .962766495 28000000.0 .942179510 42810 5 372 30000000.0 .951460429 35000000.0 .995086049 40000000.0 1.02461384 42810 5 373 45000000.0 1.04173730 50000000.0 1.04319880 54000000.0 1.05313128 42810 5 374 55000000.0 1.07746850 60000000.0 1.10473967 70000000.0 1.09910584 42810 5 375 80000000.0 1.12018798 90000000.0 1.13511332 100000000. 1.12252332 42810 5 376 120000000. 1.11592827 140000000. 1.09099089 160000000. 1.06653707 42810 5 377 180000000. 1.03179902 200000000. 1.01128150 240000000. 1.01128150 42810 5 378 280000000. 1.01128150 320000000. 1.01128150 360000000. 1.01128150 42810 5 379 400000000. 1.01128150 440000000. 1.01128150 480000000. 1.01128150 42810 5 380 520000000. 1.01128150 560000000. 1.01128150 600000000. 1.01128150 42810 5 381 640000000. 1.01128150 680000000. 1.01128150 720000000. 1.01128150 42810 5 382 760000000. 1.01128150 800000000. 1.01128150 840000000. 1.01128150 42810 5 383 880000000. 1.01128150 920000000. 1.01128150 960000000. 1.01128150 42810 5 384 1.00000E+9 0.0 42810 5 385 0.0 0.0 1001 0 1 98 42810 5 386 98 1 42810 5 387 2000000.00 5.69956E-4 2125000.00 .001721834 2250000.00 .002873676 42810 5 388 2375000.00 .004025488 2500000.00 .005177274 2625000.00 .006329038 42810 5 389 2750000.00 .007480782 2875000.00 .008632509 3000000.00 .009606989 42810 5 390 3125000.00 .010401748 3250000.00 .011196499 3375000.00 .011991243 42810 5 391 3500000.00 .012785981 3625000.00 .013580713 3750000.00 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