TENDL-2011 gen. purp. file: deuteron + Li- 6 1 MeV - 200 MeV 99 0 0 0 3006.00000 5.96344900 -1 0 0 1 325 1451 1 0.0 0.0 0 0 0 6 325 1451 2 1.99626000 200000000. 1 0 10020 0 325 1451 3 0.0 0.0 1 0 271 6 325 1451 4 3-Li- 6 NRG EVAL-NOV11 A.J. Koning 325 1451 5 TENDL-2011 DIST- REV1- 325 1451 6 ----TENDL-2011 Material 325 REVISION 1 325 1451 7 -----Incident deuteron data 325 1451 8 ------ENDF-6 Format 325 1451 9 325 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 325 1451 11 325 1451 12 Charged particle transport library made by TALYS 325 1451 13 325 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 325 1451 15 325 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 325 1451 17 325 1451 18 This evaluated data file is based primarily on the nuclear model 325 1451 19 code TALYS [kon07], version 1.4. 325 1451 20 It is part of a large collection of isotopic 325 1451 21 evaluations, all created by running TALYS with either default or 325 1451 22 adjusted input parameters. This means that the data in this 325 1451 23 evaluation have been tested in detail for some isotopes against 325 1451 24 individual experimental data, while for other isotopes it is 325 1451 25 only as good as the global quality of TALYS at the present moment. 325 1451 26 The mutual quality of all individual evaluations is however 325 1451 27 consistent: The same set of nuclear models is used and, equally 325 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 325 1451 29 The resulting data file provides a complete representation of 325 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 325 1451 31 and shielding applications over the incident projectile energy 325 1451 32 range from 1.0E-11 to 200 MeV. 325 1451 33 325 1451 34 All transport data for neutrons, other particles, photons and 325 1451 35 residual nuclides are filed using a combination of MF1,3, 325 1451 36 and MF6. This includes cross sections, angular distributions, 325 1451 37 double-differential spectra, production cross sections, 325 1451 38 residual production (activation) cross sections and recoils. 325 1451 39 This evaluation can thus be used as both transport and activation 325 1451 40 library. The data file has been created automatically using the 325 1451 41 ENDF-6 format generator TEFAL. 325 1451 42 325 1451 43 ##### ORIGIN 325 1451 44 325 1451 45 All data < 200 MeV : Produced with TALYS code 325 1451 46 325 1451 47 *************************** T H E O R Y ************************** 325 1451 48 325 1451 49 TALYS is a computer code system for the prediction and analysis 325 1451 50 of nuclear reactions. TALYS simulates reactions that involve 325 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 325 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 325 1451 53 target nuclides of mass 12 and heavier. This is achieved by 325 1451 54 implementing a suite of nuclear reaction models into a single 325 1451 55 code system. It enables to evaluate nuclear reactions from 325 1451 56 the unresolved resonance region up to intermediate energies. This 325 1451 57 evaluation is based on a theoretical analysis that utilizes the 325 1451 58 optical model, compound nucleus statistical theory, fission, 325 1451 59 direct reactions and pre-equilibrium processes, in combination 325 1451 60 with databases and models for nuclear structure. The following 325 1451 61 output of TALYS is stored in this data file: 325 1451 62 325 1451 63 - Elastic and non-elastic cross sections 325 1451 64 - Elastic scattering angular distributions 325 1451 65 - Residual production cross sections 325 1451 66 - Recoil data 325 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 325 1451 68 - Total particle energy spectra 325 1451 69 - Total particle double-differential spectra 325 1451 70 325 1451 71 Here follows a short description of the used nuclear models: 325 1451 72 325 1451 73 ##### OPTICAL MODEL 325 1451 74 325 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 325 1451 76 in TALYS used as a subroutine. The default optical model 325 1451 77 potentials (OMP) used are the local and global parameterizations 325 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 325 1451 79 for neutrons and protons which in principle are valid over the 325 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 325 1451 81 validity may differ from nucleus to nucleus. For neutrons and 325 1451 82 protons, the used parameterization is given in Eq. (7) of 325 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 325 1451 84 used in the case of deformed nuclides. 325 1451 85 To calculate the transmission coefficients and reaction 325 1451 86 cross sections for deuterons, tritons, helions and alpha 325 1451 87 particles, we use OMPs that are directly derived from our nucleon 325 1451 88 potentials using Watanabe's folding approach. 325 1451 89 325 1451 90 ##### DIRECT REACTIONS 325 1451 91 325 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 325 1451 93 calculations for rotational or vibrational (or a combination of 325 1451 94 these) nuclides, depending on the information present in the 325 1451 95 nuclear structure database of TALYS. 325 1451 96 In addition, a macroscopic, phenomenological model to describe 325 1451 97 giant resonances in the inelastic channel is used. For each 325 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 325 1451 99 calculation with ECIS-06 is performed for each giant resonance 325 1451 100 state. The cross section is then spread over the continuum with a 325 1451 101 Gaussian distribution. 325 1451 102 325 1451 103 ##### COMPOUND NUCLEUS 325 1451 104 325 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 325 1451 106 model. The transmission coefficients have been generated 325 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 325 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 325 1451 109 For each nucleus that can be reached through a binary reaction, 325 1451 110 several discrete levels and a continuum described by level 325 1451 111 densities are included simultaneously as competing channels. 325 1451 112 325 1451 113 For each residual nucleus several discrete states 325 1451 114 are included as well as a continuum described by level densities. 325 1451 115 Multiple compound emission is continued until all reaction 325 1451 116 channels are closed and the population distribution of all 325 1451 117 residual nuclides is depleted, through gamma decay, until they 325 1451 118 end up in the ground state or in an isomer. 325 1451 119 325 1451 120 For the level density, we take the Constant Temperature Model 325 1451 121 with level density parameters as given in [kon08]. This employs 325 1451 122 the temperature law at low energies and a Fermi gas expression at 325 1451 123 high energies, and takes into account the damping of shell 325 1451 124 effects at high excitation energy. We have obtained the level 325 1451 125 density parameters from a simultaneous fit to all experimental 325 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 325 1451 127 325 1451 128 Gamma-ray transmission coefficients are generated with the 325 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 325 1451 130 giant dipole resonance parameters taken from the RIPL library 325 1451 131 [rip09], and normalized with experimental radiative widths. 325 1451 132 325 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 325 1451 134 325 1451 135 For pre-equilibrium reactions, which become important for 325 1451 136 incident energies above about 10 MeV, we use the two-component 325 1451 137 exciton model [kon04], in which the neutron or proton types of 325 1451 138 particles and holes are followed throughout the reaction. For 325 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 325 1451 140 any order of particle emission was included in the calculations. 325 1451 141 A parameterization for the squared matrix element is used that is 325 1451 142 valid for the whole energy range of this evaluation. 325 1451 143 325 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 325 1451 145 contribution was added from the pick/up and knock-out reaction 325 1451 146 model by Kalbach [kal05]. 325 1451 147 325 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 325 1451 149 applied, to simulate the direct and semi-direct capture 325 1451 150 processes. 325 1451 151 325 1451 152 The angular distribution systematics by Kalbach [kal88] were used 325 1451 153 to describe the angular distributions for all continuum 325 1451 154 particles. An isotopic distribution for photons was adopted. 325 1451 155 325 1451 156 ***************** F I L E I N F O R M A T I O N **************** 325 1451 157 325 1451 158 ##### MF1: GENERAL INFORMATION 325 1451 159 325 1451 160 - MT451 : Descriptive data and directory 325 1451 161 325 1451 162 This text and the full directory of used MF/MT sections. 325 1451 163 325 1451 164 ##### MF3: REACTION CROSS SECTIONS 325 1451 165 325 1451 166 All the data present in the following MT-sections have been 325 1451 167 calculated with TALYS. If the maximal cross section in an 325 1451 168 excitation function over the whole energy range does not exceed 325 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 325 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 325 1451 171 are set to zero. The minor reaction channels are also present 325 1451 172 but they are stored in MT5 and can be reproduced in combination 325 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 325 1451 174 The following reaction channels/MT numbers are generally included: 325 1451 175 325 1451 176 - MT2 : Elastic scattering cross section: nuclear + 325 1451 177 interference terms 325 1451 178 325 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 325 1451 180 distributions of MF=6. Note that because of the interference 325 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 325 1451 182 some energies and angles. 325 1451 183 325 1451 184 - MT5 : (p,anything) cross section 325 1451 185 325 1451 186 MT5 contains the reactions which can not be stored in any other 325 1451 187 MT-number. As the incident energy increases, the cross section 325 1451 188 of MT5 increases as more reaction channels are no longer stored 325 1451 189 under a specific MT number. The information of MF3/MT5 can be 325 1451 190 combined with MF6/MT5 to obtain residual production cross 325 1451 191 sections particle production cross sections and 325 1451 192 (double-)differential cross sections. 325 1451 193 325 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 325 1451 195 325 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 325 1451 197 distributions, as well as all residual and photon production 325 1451 198 cross sections. All data are generated with TALYS. 325 1451 199 325 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 325 1451 201 interference terms 325 1451 202 325 1451 203 Relative angular distributions are tabulated on an angular grid. 325 1451 204 They are obtained by using the "nuclear-plus-interference" option 325 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 325 1451 206 integrated cross section is stored in MF=3. Note that because 325 1451 207 of the interference effect, the tabulations in both MF=6 and 325 1451 208 MF=3 can be negative at some energies and angles. 325 1451 209 325 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 325 1451 211 325 1451 212 MT5 contains the production yields of particles and residual 325 1451 213 products. It also contains the secondary energy-angle 325 1451 214 distributions for all particles and photons. First, the yields 325 1451 215 for neutrons are given for the whole energy range. Next, on a 325 1451 216 secondary energy grid the relative emission spectra are given 325 1451 217 together with the parameters for the Kalbach systematics for 325 1451 218 angular distributions. Inelastic scattering cross sections for 325 1451 219 discrete states have been broadened and added to the continuum 325 1451 220 spectra. This procedure is repeated for protons, deuterons, 325 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 325 1451 222 residual production yields are given per final product. All these 325 1451 223 yields and relative distributions can be multiplied with the 325 1451 224 cross sections given in MF3/MT5 to get the production cross 325 1451 225 sections and (double-)differential cross sections. 325 1451 226 325 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 **** 325 1451 228 325 1451 229 This file has been checked successfully by the BNL checking 325 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 325 1451 231 been processed successfully into an MCNP library by the 325 1451 232 processing code NJOY99.90 [mac00]. 325 1451 233 325 1451 234 *********************** R E F E R E N C E S ********************** 325 1451 235 325 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 325 1451 237 (1985). 325 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 325 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 325 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 325 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 325 1451 242 (2003). 325 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 325 1451 244 (2004). 325 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 325 1451 246 Proceedings of the International Conference on Nuclear 325 1451 247 Data for Science and Technology - ND-2007, 325 1451 248 April 22-27, 2007, Nice, France 325 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 325 1451 250 local level density models'', Nucl Phys. A810, 325 1451 251 13-76 (2008). 325 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 325 1451 253 pointwise and multigroup neutron and photon cross 325 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 325 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 325 1451 256 No. CEA-N-2772, 1994. 325 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 325 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 325 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 325 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 325 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 325 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 325 1451 263 RIPL - Reference Input Parameter Library for calculation 325 1451 264 of nuclear reactions and nuclear data evaluation, 325 1451 265 Nucl. Data Sheets 110, 3107 (2009). 325 1451 266 325 1451 267 ************************* C O N T E N T S ************************ 325 1451 268 325 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 325 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 325 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 325 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 325 1451 273 Unshielded Group Averages Using 162 Groups 325 1451 274 Weighting Spectrum: 1/E Spectrum 325 1451 275 1 451 275 1 325 1451 276 3 2 18 1 325 1451 277 3 5 18 1 325 1451 278 6 2 1401 1 325 1451 279 6 5 11330 1 325 1451 280 8 5 11 1 325 1451 281 325 1 0 282 325 0 0 283 3006.00000 5.96344900 0 0 0 0 325 3 2 284 0.0 0.0 0 0 1 62 325 3 2 285 62 1 325 3 2 286 8000000.00-7.19842E-4 8200000.00 .001420598 8400000.00 .003561034 325 3 2 287 8600000.00 .005701464 8800000.00 .007841891 9000000.00 .009707724 325 3 2 288 9200000.00 .011296923 9400000.00 .012886120 9600000.00 .014475314 325 3 2 289 9800000.00 .016064506 10000000.0 .019840393 11000000.0 .025271900 325 3 2 290 12000000.0 .029624376 13000000.0 .033208315 14000000.0 .036201476 325 3 2 291 15000000.0 .038715414 16000000.0 .040834112 17000000.0 .042633426 325 3 2 292 18000000.0 .044185568 19000000.0 .045552263 20000000.0 .047313437 325 3 2 293 22000000.0 .049440894 24000000.0 .051334422 26000000.0 .053052727 325 3 2 294 28000000.0 .054629269 30000000.0 .057051979 35000000.0 .060278437 325 3 2 295 40000000.0 .063140015 45000000.0 .065638833 50000000.0 .067570904 325 3 2 296 54000000.0 .068540772 55000000.0 .069485298 60000000.0 .071267882 325 3 2 297 70000000.0 .072384609 80000000.0 .071512480 90000000.0 .068797940 325 3 2 298 100000000. .062616049 120000000. .052446544 140000000. .041739892 325 3 2 299 160000000. .032008234 180000000. .024101927 200000000. .020669230 325 3 2 300 240000000. .020669230 280000000. .020669230 320000000. .020669230 325 3 2 301 360000000. .020669230 400000000. .020669230 440000000. .020669230 325 3 2 302 480000000. .020669230 520000000. .020669230 560000000. .020669230 325 3 2 303 600000000. .020669230 640000000. .020669230 680000000. .020669230 325 3 2 304 720000000. .020669230 760000000. .020669230 800000000. .020669230 325 3 2 305 840000000. .020669230 880000000. .020669230 920000000. .020669230 325 3 2 306 960000000. .020669230 1.00000E+9 0.0 325 3 2 307 325 3 0 308 3006.00000 5.96344900 0 0 0 0 325 3 5 309 0.0 0.0 0 0 1 106 325 3 5 310 106 1 325 3 5 311 1000000.00 .768036430 1125000.00 .786326582 1250000.00 .804613127 325 3 5 312 1375000.00 .822897008 1500000.00 .841178865 1625000.00 .859459148 325 3 5 313 1750000.00 .877738184 1875000.00 .896016215 2000000.00 .906307433 325 3 5 314 2125000.00 .908444300 2250000.00 .910581100 2375000.00 .912717844 325 3 5 315 2500000.00 .914854540 2625000.00 .916991194 2750000.00 .919127812 325 3 5 316 2875000.00 .921264399 3000000.00 .922558597 3125000.00 .922998643 325 3 5 317 3250000.00 .923438685 3375000.00 .923878723 3500000.00 .924318757 325 3 5 318 3625000.00 .924758788 3750000.00 .925198817 3875000.00 .925638842 325 3 5 319 4000000.00 .925080391 4200000.00 .923508094 4400000.00 .921935823 325 3 5 320 4600000.00 .920363575 4800000.00 .918791348 5000000.00 .917003556 325 3 5 321 5200000.00 .914997309 5400000.00 .912991079 5600000.00 .910984866 325 3 5 322 5800000.00 .908978667 6000000.00 .907081880 6200000.00 .905295725 325 3 5 323 6400000.00 .903509580 6600000.00 .901723443 6800000.00 .899937314 325 3 5 324 7000000.00 .898082517 7200000.00 .896158393 7400000.00 .894234276 325 3 5 325 7600000.00 .892310165 7800000.00 .890386059 8000000.00 .888359393 325 3 5 326 8200000.00 .886229307 8400000.00 .884099227 8600000.00 .881969151 325 3 5 327 8800000.00 .879839079 9000000.00 .877758808 9200000.00 .875728708 325 3 5 328 9400000.00 .873698612 9600000.00 .871668518 9800000.00 .869638428 325 3 5 329 10000000.0 .864157028 11000000.0 .855322205 12000000.0 .846801407 325 3 5 330 13000000.0 .838273156 14000000.0 .829536043 15000000.0 .820529204 325 3 5 331 16000000.0 .811326779 17000000.0 .802093869 18000000.0 .793025284 325 3 5 332 19000000.0 .784266723 20000000.0 .772032717 22000000.0 .756518270 325 3 5 333 24000000.0 .741939824 26000000.0 .727984475 28000000.0 .714555656 325 3 5 334 30000000.0 .694235185 35000000.0 .666308143 40000000.0 .639963951 325 3 5 335 45000000.0 .616158968 50000000.0 .596377282 54000000.0 .585826402 325 3 5 336 55000000.0 .574072101 60000000.0 .546437574 70000000.0 .512239721 325 3 5 337 80000000.0 .484510404 90000000.0 .464164325 100000000. .440175888 325 3 5 338 120000000. .412108889 140000000. .388034112 160000000. .367798017 325 3 5 339 180000000. .350077123 200000000. .341840000 240000000. .341840000 325 3 5 340 280000000. .341840000 320000000. .341840000 360000000. .341840000 325 3 5 341 400000000. .341840000 440000000. .341840000 480000000. .341840000 325 3 5 342 520000000. .341840000 560000000. .341840000 600000000. .341840000 325 3 5 343 640000000. .341840000 680000000. .341840000 720000000. .341840000 325 3 5 344 760000000. .341840000 800000000. .341840000 840000000. .341840000 325 3 5 345 880000000. .341840000 920000000. .341840000 960000000. .341840000 325 3 5 346 1.00000E+9 0.0 325 3 5 347 325 3 0 348 325 0 0 349 3006.00000 5.96344900 0 0 11 0 32510 5 350 0.0 0.0 1 0 1 106 32510 5 351 106 1 32510 5 352 1000000.00 .011461852 1125000.00 .034868653 1250000.00 .058270837 32510 5 353 1375000.00 .081669611 1500000.00 .105065795 1625000.00 .128459966 32510 5 354 1750000.00 .151852540 1875000.00 .175243828 2000000.00 .186509748 32510 5 355 2125000.00 .185395911 2250000.00 .184282109 2375000.00 .183168336 32510 5 356 2500000.00 .182054588 2625000.00 .180940862 2750000.00 .179827155 32510 5 357 2875000.00 .178713464 3000000.00 .176857631 3125000.00 .174249289 32510 5 358 3250000.00 .171640972 3375000.00 .169032680 3500000.00 .166424408 32510 5 359 3625000.00 .163816154 3750000.00 .161207918 3875000.00 .158599697 32510 5 360 4000000.00 .156645091 4200000.00 .155346874 4400000.00 .154048679 32510 5 361 4600000.00 .152750503 4800000.00 .151452343 5000000.00 .151566067 32510 5 362 5200000.00 .153110602 5400000.00 .154655124 5600000.00 .156199633 32510 5 363 5800000.00 .157744131 6000000.00 .157711378 6200000.00 .156083768 32510 5 364 6400000.00 .154456166 6600000.00 .152828572 6800000.00 .151200986 32510 5 365 7000000.00 .149403602 7200000.00 .147434796 7400000.00 .145465996 32510 5 366 7600000.00 .143497203 7800000.00 .141528415 8000000.00 .140149518 32510 5 367 8200000.00 .139365445 8400000.00 .138581374 8600000.00 .137797304 32510 5 368 8800000.00 .137013236 9000000.00 .135865203 9200000.00 .134350499 32510 5 369 9400000.00 .132835798 9600000.00 .131321099 9800000.00 .129806402 32510 5 370 10000000.0 .124851018 11000000.0 .115247312 12000000.0 .104635797 32510 5 371 13000000.0 .094681813 14000000.0 .086062521 15000000.0 .078771909 32510 5 372 16000000.0 .072445290 17000000.0 .066875124 18000000.0 .062734498 32510 5 373 19000000.0 .060863186 20000000.0 .061457657 22000000.0 .064711031 32510 5 374 24000000.0 .070205129 26000000.0 .076941749 28000000.0 .084063384 32510 5 375 30000000.0 .142800286 35000000.0 .212909845 40000000.0 .231199166 32510 5 376 45000000.0 .238585271 50000000.0 .232509224 54000000.0 .222768633 32510 5 377 55000000.0 .220443217 60000000.0 .223016703 70000000.0 .228944064 32510 5 378 80000000.0 .226140820 90000000.0 .223411290 100000000. .233497074 32510 5 379 120000000. .241934777 140000000. .248052468 160000000. .251515134 32510 5 380 180000000. .255713114 200000000. .257314932 240000000. .257314932 32510 5 381 280000000. .257314932 320000000. .257314932 360000000. .257314932 32510 5 382 400000000. .257314932 440000000. .257314932 480000000. .257314932 32510 5 383 520000000. .257314932 560000000. .257314932 600000000. .257314932 32510 5 384 640000000. .257314932 680000000. .257314932 720000000. .257314932 32510 5 385 760000000. .257314932 800000000. .257314932 840000000. .257314932 32510 5 386 880000000. .257314932 920000000. .257314932 960000000. .257314932 32510 5 387 1.00000E+9 0.0 32510 5 388 0.0 0.0 1001 0 1 106 32510 5 389 106 1 32510 5 390 1000000.00 .013633802 1125000.00 .041476047 1250000.00 .069312799 32510 5 391 1375000.00 .097145496 1500000.00 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