TENDL-2011 gen. purp. file: deuteron + Be- 7 1 MeV - 200 MeV 99 0 0 0 4007.00000 6.95665100 -1 0 0 1 419 1451 1 0.0 0.0 0 0 0 6 419 1451 2 1.99626000 200000000. 1 0 10020 0 419 1451 3 0.0 0.0 1 0 271 6 419 1451 4 4-Be- 7 NRG EVAL-NOV11 A.J. Koning 419 1451 5 TENDL-2011 DIST- REV1- 419 1451 6 ----TENDL-2011 Material 419 REVISION 1 419 1451 7 -----Incident deuteron data 419 1451 8 ------ENDF-6 Format 419 1451 9 419 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 419 1451 11 419 1451 12 Charged particle transport library made by TALYS 419 1451 13 419 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 419 1451 15 419 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 419 1451 17 419 1451 18 This evaluated data file is based primarily on the nuclear model 419 1451 19 code TALYS [kon07], version 1.4. 419 1451 20 It is part of a large collection of isotopic 419 1451 21 evaluations, all created by running TALYS with either default or 419 1451 22 adjusted input parameters. This means that the data in this 419 1451 23 evaluation have been tested in detail for some isotopes against 419 1451 24 individual experimental data, while for other isotopes it is 419 1451 25 only as good as the global quality of TALYS at the present moment. 419 1451 26 The mutual quality of all individual evaluations is however 419 1451 27 consistent: The same set of nuclear models is used and, equally 419 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 419 1451 29 The resulting data file provides a complete representation of 419 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 419 1451 31 and shielding applications over the incident projectile energy 419 1451 32 range from 1.0E-11 to 200 MeV. 419 1451 33 419 1451 34 All transport data for neutrons, other particles, photons and 419 1451 35 residual nuclides are filed using a combination of MF1,3, 419 1451 36 and MF6. This includes cross sections, angular distributions, 419 1451 37 double-differential spectra, production cross sections, 419 1451 38 residual production (activation) cross sections and recoils. 419 1451 39 This evaluation can thus be used as both transport and activation 419 1451 40 library. The data file has been created automatically using the 419 1451 41 ENDF-6 format generator TEFAL. 419 1451 42 419 1451 43 ##### ORIGIN 419 1451 44 419 1451 45 All data < 200 MeV : Produced with TALYS code 419 1451 46 419 1451 47 *************************** T H E O R Y ************************** 419 1451 48 419 1451 49 TALYS is a computer code system for the prediction and analysis 419 1451 50 of nuclear reactions. TALYS simulates reactions that involve 419 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 419 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 419 1451 53 target nuclides of mass 12 and heavier. This is achieved by 419 1451 54 implementing a suite of nuclear reaction models into a single 419 1451 55 code system. It enables to evaluate nuclear reactions from 419 1451 56 the unresolved resonance region up to intermediate energies. This 419 1451 57 evaluation is based on a theoretical analysis that utilizes the 419 1451 58 optical model, compound nucleus statistical theory, fission, 419 1451 59 direct reactions and pre-equilibrium processes, in combination 419 1451 60 with databases and models for nuclear structure. The following 419 1451 61 output of TALYS is stored in this data file: 419 1451 62 419 1451 63 - Elastic and non-elastic cross sections 419 1451 64 - Elastic scattering angular distributions 419 1451 65 - Residual production cross sections 419 1451 66 - Recoil data 419 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 419 1451 68 - Total particle energy spectra 419 1451 69 - Total particle double-differential spectra 419 1451 70 419 1451 71 Here follows a short description of the used nuclear models: 419 1451 72 419 1451 73 ##### OPTICAL MODEL 419 1451 74 419 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 419 1451 76 in TALYS used as a subroutine. The default optical model 419 1451 77 potentials (OMP) used are the local and global parameterizations 419 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 419 1451 79 for neutrons and protons which in principle are valid over the 419 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 419 1451 81 validity may differ from nucleus to nucleus. For neutrons and 419 1451 82 protons, the used parameterization is given in Eq. (7) of 419 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 419 1451 84 used in the case of deformed nuclides. 419 1451 85 To calculate the transmission coefficients and reaction 419 1451 86 cross sections for deuterons, tritons, helions and alpha 419 1451 87 particles, we use OMPs that are directly derived from our nucleon 419 1451 88 potentials using Watanabe's folding approach. 419 1451 89 419 1451 90 ##### DIRECT REACTIONS 419 1451 91 419 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 419 1451 93 calculations for rotational or vibrational (or a combination of 419 1451 94 these) nuclides, depending on the information present in the 419 1451 95 nuclear structure database of TALYS. 419 1451 96 In addition, a macroscopic, phenomenological model to describe 419 1451 97 giant resonances in the inelastic channel is used. For each 419 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 419 1451 99 calculation with ECIS-06 is performed for each giant resonance 419 1451 100 state. The cross section is then spread over the continuum with a 419 1451 101 Gaussian distribution. 419 1451 102 419 1451 103 ##### COMPOUND NUCLEUS 419 1451 104 419 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 419 1451 106 model. The transmission coefficients have been generated 419 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 419 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 419 1451 109 For each nucleus that can be reached through a binary reaction, 419 1451 110 several discrete levels and a continuum described by level 419 1451 111 densities are included simultaneously as competing channels. 419 1451 112 419 1451 113 For each residual nucleus several discrete states 419 1451 114 are included as well as a continuum described by level densities. 419 1451 115 Multiple compound emission is continued until all reaction 419 1451 116 channels are closed and the population distribution of all 419 1451 117 residual nuclides is depleted, through gamma decay, until they 419 1451 118 end up in the ground state or in an isomer. 419 1451 119 419 1451 120 For the level density, we take the Constant Temperature Model 419 1451 121 with level density parameters as given in [kon08]. This employs 419 1451 122 the temperature law at low energies and a Fermi gas expression at 419 1451 123 high energies, and takes into account the damping of shell 419 1451 124 effects at high excitation energy. We have obtained the level 419 1451 125 density parameters from a simultaneous fit to all experimental 419 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 419 1451 127 419 1451 128 Gamma-ray transmission coefficients are generated with the 419 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 419 1451 130 giant dipole resonance parameters taken from the RIPL library 419 1451 131 [rip09], and normalized with experimental radiative widths. 419 1451 132 419 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 419 1451 134 419 1451 135 For pre-equilibrium reactions, which become important for 419 1451 136 incident energies above about 10 MeV, we use the two-component 419 1451 137 exciton model [kon04], in which the neutron or proton types of 419 1451 138 particles and holes are followed throughout the reaction. For 419 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 419 1451 140 any order of particle emission was included in the calculations. 419 1451 141 A parameterization for the squared matrix element is used that is 419 1451 142 valid for the whole energy range of this evaluation. 419 1451 143 419 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 419 1451 145 contribution was added from the pick/up and knock-out reaction 419 1451 146 model by Kalbach [kal05]. 419 1451 147 419 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 419 1451 149 applied, to simulate the direct and semi-direct capture 419 1451 150 processes. 419 1451 151 419 1451 152 The angular distribution systematics by Kalbach [kal88] were used 419 1451 153 to describe the angular distributions for all continuum 419 1451 154 particles. An isotopic distribution for photons was adopted. 419 1451 155 419 1451 156 ***************** F I L E I N F O R M A T I O N **************** 419 1451 157 419 1451 158 ##### MF1: GENERAL INFORMATION 419 1451 159 419 1451 160 - MT451 : Descriptive data and directory 419 1451 161 419 1451 162 This text and the full directory of used MF/MT sections. 419 1451 163 419 1451 164 ##### MF3: REACTION CROSS SECTIONS 419 1451 165 419 1451 166 All the data present in the following MT-sections have been 419 1451 167 calculated with TALYS. If the maximal cross section in an 419 1451 168 excitation function over the whole energy range does not exceed 419 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 419 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 419 1451 171 are set to zero. The minor reaction channels are also present 419 1451 172 but they are stored in MT5 and can be reproduced in combination 419 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 419 1451 174 The following reaction channels/MT numbers are generally included: 419 1451 175 419 1451 176 - MT2 : Elastic scattering cross section: nuclear + 419 1451 177 interference terms 419 1451 178 419 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 419 1451 180 distributions of MF=6. Note that because of the interference 419 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 419 1451 182 some energies and angles. 419 1451 183 419 1451 184 - MT5 : (p,anything) cross section 419 1451 185 419 1451 186 MT5 contains the reactions which can not be stored in any other 419 1451 187 MT-number. As the incident energy increases, the cross section 419 1451 188 of MT5 increases as more reaction channels are no longer stored 419 1451 189 under a specific MT number. The information of MF3/MT5 can be 419 1451 190 combined with MF6/MT5 to obtain residual production cross 419 1451 191 sections particle production cross sections and 419 1451 192 (double-)differential cross sections. 419 1451 193 419 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 419 1451 195 419 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 419 1451 197 distributions, as well as all residual and photon production 419 1451 198 cross sections. All data are generated with TALYS. 419 1451 199 419 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 419 1451 201 interference terms 419 1451 202 419 1451 203 Relative angular distributions are tabulated on an angular grid. 419 1451 204 They are obtained by using the "nuclear-plus-interference" option 419 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 419 1451 206 integrated cross section is stored in MF=3. Note that because 419 1451 207 of the interference effect, the tabulations in both MF=6 and 419 1451 208 MF=3 can be negative at some energies and angles. 419 1451 209 419 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 419 1451 211 419 1451 212 MT5 contains the production yields of particles and residual 419 1451 213 products. It also contains the secondary energy-angle 419 1451 214 distributions for all particles and photons. First, the yields 419 1451 215 for neutrons are given for the whole energy range. Next, on a 419 1451 216 secondary energy grid the relative emission spectra are given 419 1451 217 together with the parameters for the Kalbach systematics for 419 1451 218 angular distributions. Inelastic scattering cross sections for 419 1451 219 discrete states have been broadened and added to the continuum 419 1451 220 spectra. This procedure is repeated for protons, deuterons, 419 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 419 1451 222 residual production yields are given per final product. All these 419 1451 223 yields and relative distributions can be multiplied with the 419 1451 224 cross sections given in MF3/MT5 to get the production cross 419 1451 225 sections and (double-)differential cross sections. 419 1451 226 419 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 **** 419 1451 228 419 1451 229 This file has been checked successfully by the BNL checking 419 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 419 1451 231 been processed successfully into an MCNP library by the 419 1451 232 processing code NJOY99.90 [mac00]. 419 1451 233 419 1451 234 *********************** R E F E R E N C E S ********************** 419 1451 235 419 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 419 1451 237 (1985). 419 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 419 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 419 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 419 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 419 1451 242 (2003). 419 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 419 1451 244 (2004). 419 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 419 1451 246 Proceedings of the International Conference on Nuclear 419 1451 247 Data for Science and Technology - ND-2007, 419 1451 248 April 22-27, 2007, Nice, France 419 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 419 1451 250 local level density models'', Nucl Phys. A810, 419 1451 251 13-76 (2008). 419 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 419 1451 253 pointwise and multigroup neutron and photon cross 419 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 419 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 419 1451 256 No. CEA-N-2772, 1994. 419 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 419 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 419 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 419 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 419 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 419 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 419 1451 263 RIPL - Reference Input Parameter Library for calculation 419 1451 264 of nuclear reactions and nuclear data evaluation, 419 1451 265 Nucl. Data Sheets 110, 3107 (2009). 419 1451 266 419 1451 267 ************************* C O N T E N T S ************************ 419 1451 268 419 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 419 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 419 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 419 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 419 1451 273 Unshielded Group Averages Using 162 Groups 419 1451 274 Weighting Spectrum: 1/E Spectrum 419 1451 275 1 451 275 1 419 1451 276 3 2 18 1 419 1451 277 3 5 18 1 419 1451 278 6 2 1401 1 419 1451 279 6 5 11466 1 419 1451 280 8 5 13 1 419 1451 281 419 1 0 282 419 0 0 283 4007.00000 6.95665100 0 0 0 0 419 3 2 284 0.0 0.0 0 0 1 51 419 3 2 285 51 1 419 3 2 286 11000000.0-.001988646 12000000.0 .006759405 13000000.0 .013909735 419 3 2 287 14000000.0 .019783797 15000000.0 .024641343 16000000.0 .028694768 419 3 2 288 17000000.0 .032112880 18000000.0 .035022437 19000000.0 .037516795 419 3 2 289 20000000.0 .040505097 22000000.0 .043771187 24000000.0 .046259503 419 3 2 290 26000000.0 .048204637 28000000.0 .049771019 30000000.0 .051775447 419 3 2 291 35000000.0 .054086077 40000000.0 .055857459 45000000.0 .057344458 419 3 2 292 50000000.0 .058526756 54000000.0 .059141153 55000000.0 .059787866 419 3 2 293 60000000.0 .061167358 70000000.0 .062439993 80000000.0 .062853941 419 3 2 294 90000000.0 .062461326 100000000. .060225557 120000000. .054697453 419 3 2 295 140000000. .046897223 160000000. .037720511 180000000. .029041849 419 3 2 296 200000000. .025088310 240000000. .025088310 280000000. .025088310 419 3 2 297 320000000. .025088310 360000000. .025088310 400000000. .025088310 419 3 2 298 440000000. .025088310 480000000. .025088310 520000000. .025088310 419 3 2 299 560000000. .025088310 600000000. .025088310 640000000. .025088310 419 3 2 300 680000000. .025088310 720000000. .025088310 760000000. .025088310 419 3 2 301 800000000. .025088310 840000000. .025088310 880000000. .025088310 419 3 2 302 920000000. .025088310 960000000. .025088310 1.00000E+9 0.0 419 3 2 303 419 3 0 304 4007.00000 6.95665100 0 0 0 0 419 3 5 305 0.0 0.0 0 0 1 106 419 3 5 306 106 1 419 3 5 307 1000000.00 .536800447 1125000.00 .571374929 1250000.00 .605942592 419 3 5 308 1375000.00 .640505218 1500000.00 .675064018 1625000.00 .709619845 419 3 5 309 1750000.00 .744173313 1875000.00 .778724881 2000000.00 .800888814 419 3 5 310 2125000.00 .810405312 2250000.00 .819921512 2375000.00 .829437461 419 3 5 311 2500000.00 .838953195 2625000.00 .848468745 2750000.00 .857984134 419 3 5 312 2875000.00 .867499384 3000000.00 .873460737 3125000.00 .875818564 419 3 5 313 3250000.00 .878176367 3375000.00 .880534149 3500000.00 .882891912 419 3 5 314 3625000.00 .885249659 3750000.00 .887607389 3875000.00 .889965107 419 3 5 315 4000000.00 .893474038 4200000.00 .898160905 4400000.00 .902847692 419 3 5 316 4600000.00 .907534411 4800000.00 .912221070 5000000.00 .916664286 419 3 5 317 5200000.00 .920860804 5400000.00 .925057284 5600000.00 .929253730 419 3 5 318 5800000.00 .933450146 6000000.00 .936336732 6200000.00 .937898867 419 3 5 319 6400000.00 .939460995 6600000.00 .941023114 6800000.00 .942585227 419 3 5 320 7000000.00 .943275446 7200000.00 .943085434 7400000.00 .942895422 419 3 5 321 7600000.00 .942705411 7800000.00 .942515401 8000000.00 .941917088 419 3 5 322 8200000.00 .940907058 8400000.00 .939897030 8600000.00 .938887005 419 3 5 323 8800000.00 .937876981 9000000.00 .936727453 9200000.00 .935437382 419 3 5 324 9400000.00 .934147313 9600000.00 .932857247 9800000.00 .931567182 419 3 5 325 10000000.0 .927327972 11000000.0 .919589305 12000000.0 .911103674 419 3 5 326 13000000.0 .902171108 14000000.0 .893107128 15000000.0 .884172429 419 3 5 327 16000000.0 .881427571 17000000.0 .878688154 18000000.0 .869799387 419 3 5 328 19000000.0 .861280619 20000000.0 .849281079 22000000.0 .834838937 419 3 5 329 24000000.0 .822211952 26000000.0 .809639412 28000000.0 .797189102 419 3 5 330 30000000.0 .777786196 35000000.0 .751865654 40000000.0 .728773383 419 3 5 331 45000000.0 .708003161 50000000.0 .690811051 54000000.0 .681658634 419 3 5 332 55000000.0 .671542141 60000000.0 .647851286 70000000.0 .618653332 419 3 5 333 80000000.0 .592096362 90000000.0 .567191735 100000000. .532748753 419 3 5 334 120000000. .488535273 140000000. .446573010 160000000. .417882512 419 3 5 335 180000000. .401168523 200000000. .393140000 240000000. .393140000 419 3 5 336 280000000. .393140000 320000000. .393140000 360000000. .393140000 419 3 5 337 400000000. .393140000 440000000. .393140000 480000000. .393140000 419 3 5 338 520000000. .393140000 560000000. .393140000 600000000. .393140000 419 3 5 339 640000000. .393140000 680000000. .393140000 720000000. .393140000 419 3 5 340 760000000. .393140000 800000000. .393140000 840000000. .393140000 419 3 5 341 880000000. .393140000 920000000. .393140000 960000000. .393140000 419 3 5 342 1.00000E+9 0.0 419 3 5 343 419 3 0 344 419 0 0 345 4007.00000 6.95665100 0 0 13 0 41910 5 346 0.0 0.0 1 0 1 98 41910 5 347 98 1 41910 5 348 2000000.00 8.78760E-4 2125000.00 .002654732 2250000.00 .004430649 41910 5 349 2375000.00 .006206518 2500000.00 .007982348 2625000.00 .009758143 41910 5 350 2750000.00 .011533908 2875000.00 .013309647 3000000.00 .014955970 41910 5 351 3125000.00 .016471073 3250000.00 .017986161 3375000.00 .019501235 41910 5 352 3500000.00 .021016298 3625000.00 .022531349 3750000.00 .024046390 41910 5 353 3875000.00 .025561423 4000000.00 .030240970 4200000.00 .038142883 41910 5 354 4400000.00 .046044663 4600000.00 .053946328 4800000.00 .061847890 41910 5 355 5000000.00 .072019082 5200000.00 .084490343 5400000.00 .096961493 41910 5 356 5600000.00 .109432542 5800000.00 .121903501 6000000.00 .132389760 41910 5 357 6200000.00 .140869169 6400000.00 .149348532 6600000.00 .157827854 41910 5 358 6800000.00 .166307138 7000000.00 .173193600 7200000.00 .178472008 41910 5 359 7400000.00 .183750397 7600000.00 .189028770 7800000.00 .194307128 41910 5 360 8000000.00 .198369525 8200000.00 .201205792 8400000.00 .204042052 41910 5 361 8600000.00 .206878306 8800000.00 .209714554 9000000.00 .211574167 41910 5 362 9200000.00 .212449888 9400000.00 .213325607 9600000.00 .214201324 41910 5 363 9800000.00 .215077041 10000000.0 .215655076 11000000.0 .215277384 41910 5 364 12000000.0 .213353173 13000000.0 .209885159 14000000.0 .205383721 41910 5 365 15000000.0 .200170642 16000000.0 .172013041 17000000.0 .150107007 41910 5 366 18000000.0 .157487493 19000000.0 .164308394 20000000.0 .175069874 41910 5 367 22000000.0 .178448657 24000000.0 .173491218 26000000.0 .176999458 41910 5 368 28000000.0 .185764406 30000000.0 .201395864 35000000.0 .227119780 41910 5 369 40000000.0 .246753084 45000000.0 .250875367 50000000.0 .252309668 41910 5 370 54000000.0 .255491439 55000000.0 .262741083 60000000.0 .286142828 41910 5 371 70000000.0 .300584625 80000000.0 .322281083 90000000.0 .344021928 41910 5 372 100000000. .371403769 120000000. .382371602 140000000. .372857142 41910 5 373 160000000. .369831391 180000000. .373170535 200000000. .377985239 41910 5 374 240000000. .377985239 280000000. .377985239 320000000. .377985239 41910 5 375 360000000. .377985239 400000000. .377985239 440000000. .377985239 41910 5 376 480000000. .377985239 520000000. .377985239 560000000. .377985239 41910 5 377 600000000. .377985239 640000000. .377985239 680000000. .377985239 41910 5 378 720000000. .377985239 760000000. .377985239 800000000. .377985239 41910 5 379 840000000. .377985239 880000000. .377985239 920000000. .377985239 41910 5 380 960000000. .377985239 1.00000E+9 0.0 41910 5 381 0.0 0.0 1001 0 1 106 41910 5 382 106 1 41910 5 383 1000000.00 .020870507 1125000.00 .063491176 1250000.00 .106103439 41910 5 384 1375000.00 .148709492 1500000.00 .191310830 1625000.00 .233908502 41910 5 385 1750000.00 .276503267 1875000.00 .319095689 2000000.00 .341481339 41910 5 386 2125000.00 .343236311 2250000.00 .344991228 2375000.00 .346746098 41910 5 387 2500000.00 .348500929 2625000.00 .350255726 2750000.00 .352010494 41910 5 388 2875000.00 .353765235 3000000.00 .355142768 3125000.00 .356137825 41910 5 389 3250000.00 .357132872 3375000.00 .358127910 3500000.00 .359122940 41910 5 390 3625000.00 .360117963 3750000.00 .361112980 3875000.00 .362107991 41910 5 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