TENDL-2011 gen. purp. file: proton + 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 .998620000 200000000. 1 0 10010 0 325 1451 3 0.0 0.0 1 0 271 6 325 1451 4 3-Li- 6 NRG EVAL-OCT11 A.J. Koning 325 1451 5 TENDL-2011 DIST- REV1- 325 1451 6 ----TENDL-2011 Material 325 REVISION 1 325 1451 7 -----Incident proton 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 8900 1 325 1451 280 8 5 9 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 98 325 3 2 285 98 1 325 3 2 286 2000000.00-.033702149 2125000.00-.025685905 2250000.00-.017669911 325 3 2 287 2375000.00-.009654129 2500000.00-.001638527 2625000.00 .006376919 325 3 2 288 2750000.00 .014392230 2875000.00 .022407423 3000000.00 .028182379 325 3 2 289 3125000.00 .031685818 3250000.00 .035189221 3375000.00 .038692593 325 3 2 290 3500000.00 .042195937 3625000.00 .045699256 3750000.00 .049202551 325 3 2 291 3875000.00 .052705827 4000000.00 .055554838 4200000.00 .057749614 325 3 2 292 4400000.00 .059944352 4600000.00 .062139058 4800000.00 .064333735 325 3 2 293 5000000.00 .065833323 5200000.00 .066628505 5400000.00 .067423680 325 3 2 294 5600000.00 .068218849 5800000.00 .069014012 6000000.00 .069553533 325 3 2 295 6200000.00 .069834560 6400000.00 .070115584 6600000.00 .070396608 325 3 2 296 6800000.00 .070677630 7000000.00 .070883773 7200000.00 .071014319 325 3 2 297 7400000.00 .071144865 7600000.00 .071275410 7800000.00 .071405956 325 3 2 298 8000000.00 .071524393 8200000.00 .071630623 8400000.00 .071736852 325 3 2 299 8600000.00 .071843080 8800000.00 .071949309 9000000.00 .072055496 325 3 2 300 9200000.00 .072161640 9400000.00 .072267784 9600000.00 .072373928 325 3 2 301 9800000.00 .072480071 10000000.0 .072766428 11000000.0 .073169273 325 3 2 302 12000000.0 .073398365 13000000.0 .073411987 14000000.0 .073199825 325 3 2 303 15000000.0 .072774184 16000000.0 .072159683 17000000.0 .071385746 325 3 2 304 18000000.0 .070481884 19000000.0 .069474977 20000000.0 .067816306 325 3 2 305 22000000.0 .065447032 24000000.0 .062967055 26000000.0 .060445766 325 3 2 306 28000000.0 .057927398 30000000.0 .053714846 35000000.0 .047882860 325 3 2 307 40000000.0 .042544485 45000000.0 .037739411 50000000.0 .033850813 325 3 2 308 54000000.0 .031824064 55000000.0 .029676109 60000000.0 .024934708 325 3 2 309 70000000.0 .019756138 80000000.0 .015834595 90000000.0 .012764789 325 3 2 310 100000000. .009541943 120000000. .006678818 140000000. .004967402 325 3 2 311 160000000. .003960821 180000000. .003327374 200000000. .003085352 325 3 2 312 240000000. .003085352 280000000. .003085352 320000000. .003085352 325 3 2 313 360000000. .003085352 400000000. .003085352 440000000. .003085352 325 3 2 314 480000000. .003085352 520000000. .003085352 560000000. .003085352 325 3 2 315 600000000. .003085352 640000000. .003085352 680000000. .003085352 325 3 2 316 720000000. .003085352 760000000. .003085352 800000000. .003085352 325 3 2 317 840000000. .003085352 880000000. .003085352 920000000. .003085352 325 3 2 318 960000000. .003085352 1.00000E+9 0.0 325 3 2 319 325 3 0 320 3006.00000 5.96344900 0 0 0 0 325 3 5 321 0.0 0.0 0 0 1 106 325 3 5 322 106 1 325 3 5 323 1000000.00 .528316282 1125000.00 .532903523 1250000.00 .537489859 325 3 5 324 1375000.00 .542075527 1500000.00 .546660687 1625000.00 .551245453 325 3 5 325 1750000.00 .555829905 1875000.00 .560414106 2000000.00 .558474682 325 3 5 326 2125000.00 .549874699 2250000.00 .541274985 2375000.00 .532675498 325 3 5 327 2500000.00 .524076205 2625000.00 .515477078 2750000.00 .506878097 325 3 5 328 2875000.00 .498279241 3000000.00 .491770554 3125000.00 .487381219 325 3 5 329 3250000.00 .482991927 3375000.00 .478602675 3500000.00 .474213459 325 3 5 330 3625000.00 .469824274 3750000.00 .465435117 3875000.00 .461045986 325 3 5 331 4000000.00 .458777512 4200000.00 .458651488 4400000.00 .458525467 325 3 5 332 4600000.00 .458399447 4800000.00 .458273429 5000000.00 .459388248 325 3 5 333 5200000.00 .461760540 5400000.00 .464132811 5600000.00 .466505064 325 3 5 334 5800000.00 .468877299 6000000.00 .471317147 6200000.00 .473825365 325 3 5 335 6400000.00 .476333569 6600000.00 .478841761 6800000.00 .481349942 325 3 5 336 7000000.00 .483441080 7200000.00 .485111187 7400000.00 .486781289 325 3 5 337 7600000.00 .488451385 7800000.00 .490121477 8000000.00 .491213951 325 3 5 338 8200000.00 .491723976 8400000.00 .492234000 8600000.00 .492744023 325 3 5 339 8800000.00 .493254045 9000000.00 .493313722 9200000.00 .492919706 325 3 5 340 9400000.00 .492525691 9600000.00 .492131677 9800000.00 .491737663 325 3 5 341 10000000.0 .490309853 11000000.0 .487103493 12000000.0 .482164817 325 3 5 342 13000000.0 .475569646 14000000.0 .467931279 15000000.0 .459660658 325 3 5 343 16000000.0 .451098597 17000000.0 .442495818 18000000.0 .434037575 325 3 5 344 19000000.0 .425913767 20000000.0 .414701658 22000000.0 .401021495 325 3 5 345 24000000.0 .389469630 26000000.0 .378989900 28000000.0 .368173928 325 3 5 346 30000000.0 .349523363 35000000.0 .323656550 40000000.0 .300415881 325 3 5 347 45000000.0 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2.03595E-4 5200000.00 6.13516E-4 5400000.00 .001023433 32510 5 365 5600000.00 .001433346 5800000.00 .001843257 6000000.00 .002910873 32510 5 366 6200000.00 .004643536 6400000.00 .006376190 6600000.00 .008108836 32510 5 367 6800000.00 .009841473 7000000.00 .011545663 7200000.00 .013221132 32510 5 368 7400000.00 .014896595 7600000.00 .016572053 7800000.00 .018247506 32510 5 369 8000000.00 .020549829 8200000.00 .023484266 8400000.00 .026418697 32510 5 370 8600000.00 .029353121 8800000.00 .032287539 9000000.00 .034285613 32510 5 371 9200000.00 .035340385 9400000.00 .036395155 9600000.00 .037449923 32510 5 372 9800000.00 .038504689 10000000.0 .040047666 11000000.0 .041459226 32510 5 373 12000000.0 .042140098 13000000.0 .042507088 14000000.0 .042368764 32510 5 374 15000000.0 .041780910 16000000.0 .040846557 17000000.0 .039665516 32510 5 375 18000000.0 .038295307 19000000.0 .036655790 20000000.0 .033712692 32510 5 376 22000000.0 .028807709 24000000.0 .022083449 26000000.0 .014946420 32510 5 377 28000000.0 .009645469 30000000.0 .004954616 35000000.0 .009301287 32510 5 378 40000000.0 .023729801 45000000.0 .035993792 50000000.0 .043958278 32510 5 379 54000000.0 .047526477 55000000.0 .050156073 60000000.0 .054452797 32510 5 380 70000000.0 .056932617 80000000.0 .057222090 90000000.0 .056907882 32510 5 381 100000000. .055543295 120000000. .053338333 140000000. .052842566 32510 5 382 160000000. .052573818 180000000. .052255225 200000000. .052091132 32510 5 383 240000000. .052091132 280000000. .052091132 320000000. .052091132 32510 5 384 360000000. .052091132 400000000. .052091132 440000000. .052091132 32510 5 385 480000000. .052091132 520000000. .052091132 560000000. .052091132 32510 5 386 600000000. .052091132 640000000. .052091132 680000000. .052091132 32510 5 387 720000000. .052091132 760000000. .052091132 800000000. .052091132 32510 5 388 840000000. .052091132 880000000. .052091132 920000000. .052091132 32510 5 389 960000000. .052091132 1.00000E+9 0.0 32510 5 390 0.0 0.0 1001 0 1 98 32510 5 391 98 1 32510 5 392 2000000.00 4.33594E-4 2125000.00 .001309886 2250000.00 .002186150 32510 5 393 2375000.00 .003062392 2500000.00 .003938613 2625000.00 .004814818 32510 5 394 2750000.00 .005691008 2875000.00 .006567185 3000000.00 .009516129 32510 5 395 3125000.00 .014566798 3250000.00 .019617415 3375000.00 .024667987 32510 5 396 3500000.00 .029718518 3625000.00 .034769013 3750000.00 .039819476 32510 5 397 3875000.00 .044869908 4000000.00 .049335685 4200000.00 .053222860 32510 5 398 4400000.00 .057109970 4600000.00 .060997023 4800000.00 .064884026 32510 5 399 5000000.00 .069767626 5200000.00 .075661190 5400000.00 .081554700 32510 5 400 5600000.00 .087448164 5800000.00 .093341585 6000000.00 .098778351 32510 5 401 6200000.00 .103753369 6400000.00 .108728359 6600000.00 .113703326 32510 5 402 6800000.00 .118678271 7000000.00 .123766736 7200000.00 .128969811 32510 5 403 7400000.00 .134172868 7600000.00 .139375909 7800000.00 .144578934 32510 5 404 8000000.00 .147789809 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