TENDL-2011 gen. purp. file: proton + B - 10 1 MeV - 200 MeV 99 0 0 0 5010.00000 9.92692100 -1 0 0 1 525 1451 1 0.0 0.0 0 0 0 6 525 1451 2 .998620000 200000000. 1 0 10010 0 525 1451 3 0.0 0.0 1 0 271 6 525 1451 4 5-B - 10 NRG EVAL-OCT11 A.J. Koning 525 1451 5 TENDL-2011 DIST- REV1- 525 1451 6 ----TENDL-2011 Material 525 REVISION 1 525 1451 7 -----Incident proton data 525 1451 8 ------ENDF-6 Format 525 1451 9 525 1451 10 TENDL-2011 (TALYS Evaluated Nuclear Data Library) 525 1451 11 525 1451 12 Charged particle transport library made by TALYS 525 1451 13 525 1451 14 Author: A.J. Koning, NRG Petten, The Netherlands 525 1451 15 525 1451 16 ************** G E N E R A L I N F O R M A T I O N ************* 525 1451 17 525 1451 18 This evaluated data file is based primarily on the nuclear model 525 1451 19 code TALYS [kon07], version 1.4. 525 1451 20 It is part of a large collection of isotopic 525 1451 21 evaluations, all created by running TALYS with either default or 525 1451 22 adjusted input parameters. This means that the data in this 525 1451 23 evaluation have been tested in detail for some isotopes against 525 1451 24 individual experimental data, while for other isotopes it is 525 1451 25 only as good as the global quality of TALYS at the present moment. 525 1451 26 The mutual quality of all individual evaluations is however 525 1451 27 consistent: The same set of nuclear models is used and, equally 525 1451 28 important, the same ENDF-6 formatting procedures for each isotope. 525 1451 29 The resulting data file provides a complete representation of 525 1451 30 nuclear data needed for transport, damage, heating, radioactivity, 525 1451 31 and shielding applications over the incident projectile energy 525 1451 32 range from 1.0E-11 to 200 MeV. 525 1451 33 525 1451 34 All transport data for neutrons, other particles, photons and 525 1451 35 residual nuclides are filed using a combination of MF1,3, 525 1451 36 and MF6. This includes cross sections, angular distributions, 525 1451 37 double-differential spectra, production cross sections, 525 1451 38 residual production (activation) cross sections and recoils. 525 1451 39 This evaluation can thus be used as both transport and activation 525 1451 40 library. The data file has been created automatically using the 525 1451 41 ENDF-6 format generator TEFAL. 525 1451 42 525 1451 43 ##### ORIGIN 525 1451 44 525 1451 45 All data < 200 MeV : Produced with TALYS code 525 1451 46 525 1451 47 *************************** T H E O R Y ************************** 525 1451 48 525 1451 49 TALYS is a computer code system for the prediction and analysis 525 1451 50 of nuclear reactions. TALYS simulates reactions that involve 525 1451 51 neutrons, gamma-rays, protons, deuterons, tritons, helions and 525 1451 52 alpha-particles, in the 1 keV - 200 MeV energy range and for 525 1451 53 target nuclides of mass 12 and heavier. This is achieved by 525 1451 54 implementing a suite of nuclear reaction models into a single 525 1451 55 code system. It enables to evaluate nuclear reactions from 525 1451 56 the unresolved resonance region up to intermediate energies. This 525 1451 57 evaluation is based on a theoretical analysis that utilizes the 525 1451 58 optical model, compound nucleus statistical theory, fission, 525 1451 59 direct reactions and pre-equilibrium processes, in combination 525 1451 60 with databases and models for nuclear structure. The following 525 1451 61 output of TALYS is stored in this data file: 525 1451 62 525 1451 63 - Elastic and non-elastic cross sections 525 1451 64 - Elastic scattering angular distributions 525 1451 65 - Residual production cross sections 525 1451 66 - Recoil data 525 1451 67 - Total particle cross sections, e.g. (p,xn), (p,xp),.. 525 1451 68 - Total particle energy spectra 525 1451 69 - Total particle double-differential spectra 525 1451 70 525 1451 71 Here follows a short description of the used nuclear models: 525 1451 72 525 1451 73 ##### OPTICAL MODEL 525 1451 74 525 1451 75 All optical model calculations are performed by ECIS-06 [ray94], 525 1451 76 in TALYS used as a subroutine. The default optical model 525 1451 77 potentials (OMP) used are the local and global parameterizations 525 1451 78 of Koning and Delaroche [kon03]. These are phenomenological OMPs 525 1451 79 for neutrons and protons which in principle are valid over the 525 1451 80 1 keV - 200 MeV energy range, though the low energy boundary of 525 1451 81 validity may differ from nucleus to nucleus. For neutrons and 525 1451 82 protons, the used parameterization is given in Eq. (7) of 525 1451 83 [kon03]. Revised versions of these potentials (unpublished) are 525 1451 84 used in the case of deformed nuclides. 525 1451 85 To calculate the transmission coefficients and reaction 525 1451 86 cross sections for deuterons, tritons, helions and alpha 525 1451 87 particles, we use OMPs that are directly derived from our nucleon 525 1451 88 potentials using Watanabe's folding approach. 525 1451 89 525 1451 90 ##### DIRECT REACTIONS 525 1451 91 525 1451 92 The built-in ECIS-06 is used for DWBA or coupled-channels 525 1451 93 calculations for rotational or vibrational (or a combination of 525 1451 94 these) nuclides, depending on the information present in the 525 1451 95 nuclear structure database of TALYS. 525 1451 96 In addition, a macroscopic, phenomenological model to describe 525 1451 97 giant resonances in the inelastic channel is used. For each 525 1451 98 multipolarity an energy weighted sum rule applies and a DWBA 525 1451 99 calculation with ECIS-06 is performed for each giant resonance 525 1451 100 state. The cross section is then spread over the continuum with a 525 1451 101 Gaussian distribution. 525 1451 102 525 1451 103 ##### COMPOUND NUCLEUS 525 1451 104 525 1451 105 For compound nucleus reactions we use the the Hauser-Feshbach 525 1451 106 model. The transmission coefficients have been generated 525 1451 107 with the aforementioned OMPs and the full j,l-dependence of the 525 1451 108 transmission coefficients in the Hauser-Feshbach model is used. 525 1451 109 For each nucleus that can be reached through a binary reaction, 525 1451 110 several discrete levels and a continuum described by level 525 1451 111 densities are included simultaneously as competing channels. 525 1451 112 525 1451 113 For each residual nucleus several discrete states 525 1451 114 are included as well as a continuum described by level densities. 525 1451 115 Multiple compound emission is continued until all reaction 525 1451 116 channels are closed and the population distribution of all 525 1451 117 residual nuclides is depleted, through gamma decay, until they 525 1451 118 end up in the ground state or in an isomer. 525 1451 119 525 1451 120 For the level density, we take the Constant Temperature Model 525 1451 121 with level density parameters as given in [kon08]. This employs 525 1451 122 the temperature law at low energies and a Fermi gas expression at 525 1451 123 high energies, and takes into account the damping of shell 525 1451 124 effects at high excitation energy. We have obtained the level 525 1451 125 density parameters from a simultaneous fit to all experimental 525 1451 126 discrete levels and D0 values from the RIPL library [rip09]. 525 1451 127 525 1451 128 Gamma-ray transmission coefficients are generated with the 525 1451 129 Kopecky-Uhl generalized Lorentzian for strength functions, with 525 1451 130 giant dipole resonance parameters taken from the RIPL library 525 1451 131 [rip09], and normalized with experimental radiative widths. 525 1451 132 525 1451 133 ##### PRE-EQUILIBRIUM REACTIONS 525 1451 134 525 1451 135 For pre-equilibrium reactions, which become important for 525 1451 136 incident energies above about 10 MeV, we use the two-component 525 1451 137 exciton model [kon04], in which the neutron or proton types of 525 1451 138 particles and holes are followed throughout the reaction. For 525 1451 139 energies above 20 MeV, multiple pre-equilibrium emission up to 525 1451 140 any order of particle emission was included in the calculations. 525 1451 141 A parameterization for the squared matrix element is used that is 525 1451 142 valid for the whole energy range of this evaluation. 525 1451 143 525 1451 144 For deuterons, tritons, helions and alpha-particles, an extra 525 1451 145 contribution was added from the pick/up and knock-out reaction 525 1451 146 model by Kalbach [kal05]. 525 1451 147 525 1451 148 For photons, the model of Akkermans and Gruppelaar [akk85] was 525 1451 149 applied, to simulate the direct and semi-direct capture 525 1451 150 processes. 525 1451 151 525 1451 152 The angular distribution systematics by Kalbach [kal88] were used 525 1451 153 to describe the angular distributions for all continuum 525 1451 154 particles. An isotopic distribution for photons was adopted. 525 1451 155 525 1451 156 ***************** F I L E I N F O R M A T I O N **************** 525 1451 157 525 1451 158 ##### MF1: GENERAL INFORMATION 525 1451 159 525 1451 160 - MT451 : Descriptive data and directory 525 1451 161 525 1451 162 This text and the full directory of used MF/MT sections. 525 1451 163 525 1451 164 ##### MF3: REACTION CROSS SECTIONS 525 1451 165 525 1451 166 All the data present in the following MT-sections have been 525 1451 167 calculated with TALYS. If the maximal cross section in an 525 1451 168 excitation function over the whole energy range does not exceed 525 1451 169 1.e-9 b, the MT-number is not included at all. Cross sections 525 1451 170 lower than 1.e-20 b are assumed to have no physical meaning and 525 1451 171 are set to zero. The minor reaction channels are also present 525 1451 172 but they are stored in MT5 and can be reproduced in combination 525 1451 173 with MF6. All MT numbers extend up to the highest incident energy. 525 1451 174 The following reaction channels/MT numbers are generally included: 525 1451 175 525 1451 176 - MT2 : Elastic scattering cross section: nuclear + 525 1451 177 interference terms 525 1451 178 525 1451 179 Obtained by integrating the "nuclear-plus-interference" angular 525 1451 180 distributions of MF=6. Note that because of the interference 525 1451 181 effect, the tabulations in both MF=6 and MF=3 can be negative at 525 1451 182 some energies and angles. 525 1451 183 525 1451 184 - MT5 : (p,anything) cross section 525 1451 185 525 1451 186 MT5 contains the reactions which can not be stored in any other 525 1451 187 MT-number. As the incident energy increases, the cross section 525 1451 188 of MT5 increases as more reaction channels are no longer stored 525 1451 189 under a specific MT number. The information of MF3/MT5 can be 525 1451 190 combined with MF6/MT5 to obtain residual production cross 525 1451 191 sections particle production cross sections and 525 1451 192 (double-)differential cross sections. 525 1451 193 525 1451 194 ##### MF6: PRODUCT ENERGY-ANGLE DISTRIBUTIONS 525 1451 195 525 1451 196 In MF6 we store all secondary energy, angle, and energy-angle 525 1451 197 distributions, as well as all residual and photon production 525 1451 198 cross sections. All data are generated with TALYS. 525 1451 199 525 1451 200 - MT2 : Elastic scattering angular distribution: nuclear + 525 1451 201 interference terms 525 1451 202 525 1451 203 Relative angular distributions are tabulated on an angular grid. 525 1451 204 They are obtained by using the "nuclear-plus-interference" option 525 1451 205 in MF=6, which corresponds to LAW=5, LTP=12, and the appropriate 525 1451 206 integrated cross section is stored in MF=3. Note that because 525 1451 207 of the interference effect, the tabulations in both MF=6 and 525 1451 208 MF=3 can be negative at some energies and angles. 525 1451 209 525 1451 210 - MT5 : (p,anything) yields and energy-angle distributions 525 1451 211 525 1451 212 MT5 contains the production yields of particles and residual 525 1451 213 products. It also contains the secondary energy-angle 525 1451 214 distributions for all particles and photons. First, the yields 525 1451 215 for neutrons are given for the whole energy range. Next, on a 525 1451 216 secondary energy grid the relative emission spectra are given 525 1451 217 together with the parameters for the Kalbach systematics for 525 1451 218 angular distributions. Inelastic scattering cross sections for 525 1451 219 discrete states have been broadened and added to the continuum 525 1451 220 spectra. This procedure is repeated for protons, deuterons, 525 1451 221 tritons, Helium-3, alpha particles and photons. Finally, the 525 1451 222 residual production yields are given per final product. All these 525 1451 223 yields and relative distributions can be multiplied with the 525 1451 224 cross sections given in MF3/MT5 to get the production cross 525 1451 225 sections and (double-)differential cross sections. 525 1451 226 525 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 **** 525 1451 228 525 1451 229 This file has been checked successfully by the BNL checking 525 1451 230 codes CHECKR-8.0, FIZCON-8.0 and PSYCHE-8.0 [her08] and has 525 1451 231 been processed successfully into an MCNP library by the 525 1451 232 processing code NJOY99.90 [mac00]. 525 1451 233 525 1451 234 *********************** R E F E R E N C E S ********************** 525 1451 235 525 1451 236 [akk85] J.M. Akkermans and H. Gruppelaar, Phys. Lett. 157B, 95 525 1451 237 (1985). 525 1451 238 [her08] M. Herman, ENDF Utility Codes Release 8.0, (2008). 525 1451 239 [kal88] C. Kalbach, Phys. Rev. C37, 2350 (1988). 525 1451 240 [kal05] C. Kalbach, Phys. Rev. C71, 034606 (2005). 525 1451 241 [kon03] A.J. Koning and J.P. Delaroche, Nucl. Phys. A713, 231 525 1451 242 (2003). 525 1451 243 [kon04] A.J. Koning and M.C. Duijvestijn, Nucl. Phys. A744, 15 525 1451 244 (2004). 525 1451 245 [kon07] A.J. Koning, S. Hilaire and M.C. Duijvestijn, TALYS-1.0, 525 1451 246 Proceedings of the International Conference on Nuclear 525 1451 247 Data for Science and Technology - ND-2007, 525 1451 248 April 22-27, 2007, Nice, France 525 1451 249 [kon08] A.J. Koning, S. Hilaire and S. Goriely, ``Global and 525 1451 250 local level density models'', Nucl Phys. A810, 525 1451 251 13-76 (2008). 525 1451 252 [mac00] R.E. Macfarlane, NJOY99 - Code system for producing 525 1451 253 pointwise and multigroup neutron and photon cross 525 1451 254 sections from ENDF/B Data, RSIC PSR-480 (2000). 525 1451 255 [ray94] J. Raynal, Notes on ECIS94, CEA Saclay Report 525 1451 256 No. CEA-N-2772, 1994. 525 1451 257 [rip09] R. Capote, M. Herman, P. Oblozinsky, P.G. Young, 525 1451 258 S. Goriely, T. Belgya, A.V. Ignatyuk, A.J. Koning, 525 1451 259 S. Hilaire, V. Plujko, M. Avrigeanu, O. Bersillon, 525 1451 260 M.B. Chadwick, T. Fukahori, S. Kailas, J. Kopecky, 525 1451 261 V.M. Maslov, G. Reffo, M. Sin, E. Soukhovitskii, 525 1451 262 P. Talou, H. Yinlu, and G. Zhigang, 525 1451 263 RIPL - Reference Input Parameter Library for calculation 525 1451 264 of nuclear reactions and nuclear data evaluation, 525 1451 265 Nucl. Data Sheets 110, 3107 (2009). 525 1451 266 525 1451 267 ************************* C O N T E N T S ************************ 525 1451 268 525 1451 269 *************** PROGRAM ACTIVATE (VERSION 2010-1) *************** 525 1451 270 MF=10 Activation Cross Sections Defined by Combining MF=3 525 1451 271 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 525 1451 272 ***************** Program GROUPIE (VERSION 2011-1) ************** 525 1451 273 Unshielded Group Averages Using 162 Groups 525 1451 274 Weighting Spectrum: 1/E Spectrum 525 1451 275 1 451 275 1 525 1451 276 3 2 18 1 525 1451 277 3 5 18 1 525 1451 278 6 2 1401 1 525 1451 279 6 5 11296 1 525 1451 280 8 5 19 1 525 1451 281 525 1 0 282 525 0 0 283 5010.00000 9.92692100 0 0 0 0 525 3 2 284 0.0 0.0 0 0 1 67 525 3 2 285 67 1 525 3 2 286 7000000.00-.013303862 7200000.00-.009287049 7400000.00-.005270250 525 3 2 287 7600000.00-.001253464 7800000.00 .002763311 8000000.00 .006373657 525 3 2 288 8200000.00 .009574178 8400000.00 .012774691 8600000.00 .015975196 525 3 2 289 8800000.00 .019175696 9000000.00 .022004493 9200000.00 .024458826 525 3 2 290 9400000.00 .026913154 9600000.00 .029367479 9800000.00 .031821800 525 3 2 291 10000000.0 .037610645 11000000.0 .045761279 12000000.0 .051942457 525 3 2 292 13000000.0 .056692123 14000000.0 .060401600 15000000.0 .063340441 525 3 2 293 16000000.0 .065686962 17000000.0 .067559350 18000000.0 .069037027 525 3 2 294 19000000.0 .070175937 20000000.0 .071239701 22000000.0 .071965666 525 3 2 295 24000000.0 .071907960 26000000.0 .071243657 28000000.0 .070118349 525 3 2 296 30000000.0 .067299890 35000000.0 .062500970 40000000.0 .057214990 525 3 2 297 45000000.0 .051905532 50000000.0 .047299271 54000000.0 .044814116 525 3 2 298 55000000.0 .042086823 60000000.0 .035874674 70000000.0 .028819192 525 3 2 299 80000000.0 .023299816 90000000.0 .018906380 100000000. .014228103 525 3 2 300 120000000. .010023230 140000000. .007500384 160000000. .006021404 525 3 2 301 180000000. .005094333 200000000. .004741801 240000000. .004741801 525 3 2 302 280000000. .004741801 320000000. .004741801 360000000. .004741801 525 3 2 303 400000000. .004741801 440000000. .004741801 480000000. .004741801 525 3 2 304 520000000. .004741801 560000000. .004741801 600000000. .004741801 525 3 2 305 640000000. .004741801 680000000. .004741801 720000000. .004741801 525 3 2 306 760000000. .004741801 800000000. .004741801 840000000. .004741801 525 3 2 307 880000000. .004741801 920000000. .004741801 960000000. .004741801 525 3 2 308 1.00000E+9 0.0 525 3 2 309 525 3 0 310 5010.00000 9.92692100 0 0 0 0 525 3 5 311 0.0 0.0 0 0 1 106 525 3 5 312 106 1 525 3 5 313 1000000.00 .146216361 1125000.00 .158318869 1250000.00 .170418990 525 3 5 314 1375000.00 .182517347 1500000.00 .194614366 1625000.00 .206710343 525 3 5 315 1750000.00 .218805495 1875000.00 .230899982 2000000.00 .241170659 525 3 5 316 2125000.00 .249579337 2250000.00 .257987752 2375000.00 .266395945 525 3 5 317 2500000.00 .274803948 2625000.00 .283211788 2750000.00 .291619487 525 3 5 318 2875000.00 .300027062 3000000.00 .307657353 3125000.00 .314499516 525 3 5 319 3250000.00 .321341610 3375000.00 .328183642 3500000.00 .335025620 525 3 5 320 3625000.00 .341867548 3750000.00 .348709432 3875000.00 .355551275 525 3 5 321 4000000.00 .364354026 4200000.00 .375172071 4400000.00 .385989935 525 3 5 322 4600000.00 .396807641 4800000.00 .407625207 5000000.00 .420029211 525 3 5 323 5200000.00 .434040939 5400000.00 .448052542 5600000.00 .462064032 525 3 5 324 5800000.00 .476075421 6000000.00 .488561112 6200000.00 .499504081 525 3 5 325 6400000.00 .510446992 6600000.00 .521389849 6800000.00 .532332659 525 3 5 326 7000000.00 .544557356 7200000.00 .558076205 7400000.00 .571595009 525 3 5 327 7600000.00 .585113769 7800000.00 .598632490 8000000.00 .606827185 525 3 5 328 8200000.00 .609653325 8400000.00 .612479459 8600000.00 .615305586 525 3 5 329 8800000.00 .618131708 9000000.00 .620736647 9200000.00 .623118761 525 3 5 330 9400000.00 .625500871 9600000.00 .627882977 9800000.00 .630265080 525 3 5 331 10000000.0 .630033081 11000000.0 .624741188 12000000.0 .616493358 525 3 5 332 13000000.0 .607556317 14000000.0 .598486798 15000000.0 .590324254 525 3 5 333 16000000.0 .582205314 17000000.0 .573227819 18000000.0 .564255323 525 3 5 334 19000000.0 .555442058 20000000.0 .542414688 22000000.0 .525024411 525 3 5 335 24000000.0 .508101114 26000000.0 .491827937 28000000.0 .476391177 525 3 5 336 30000000.0 .452324533 35000000.0 .420675275 40000000.0 .393313563 525 3 5 337 45000000.0 .369588213 50000000.0 .350812449 54000000.0 .341124869 525 3 5 338 55000000.0 .330931656 60000000.0 .308474008 70000000.0 .283746861 525 3 5 339 80000000.0 .264499568 90000000.0 .248847283 100000000. .231167134 525 3 5 340 120000000. .213493281 140000000. .200807893 160000000. .191362233 525 3 5 341 180000000. .183754605 200000000. .180300000 240000000. .180300000 525 3 5 342 280000000. .180300000 320000000. .180300000 360000000. .180300000 525 3 5 343 400000000. .180300000 440000000. .180300000 480000000. .180300000 525 3 5 344 520000000. .180300000 560000000. .180300000 600000000. .180300000 525 3 5 345 640000000. .180300000 680000000. .180300000 720000000. .180300000 525 3 5 346 760000000. .180300000 800000000. .180300000 840000000. .180300000 525 3 5 347 880000000. .180300000 920000000. .180300000 960000000. .180300000 525 3 5 348 1.00000E+9 0.0 525 3 5 349 525 3 0 350 525 0 0 351 5010.00000 9.92692100 0 0 19 0 52510 5 352 0.0 0.0 1 0 1 82 52510 5 353 82 1 52510 5 354 4000000.00 6.38895E-5 4200000.00 1.92741E-4 4400000.00 3.21589E-4 52510 5 355 4600000.00 4.50436E-4 4800000.00 5.79281E-4 5000000.00 7.48788E-4 52510 5 356 5200000.00 9.59501E-4 5400000.00 .001170212 5600000.00 .001380921 52510 5 357 5800000.00 .001591629 6000000.00 .001804765 6200000.00 .002020354 52510 5 358 6400000.00 .002235943 6600000.00 .002451531 6800000.00 .002667118 52510 5 359 7000000.00 .015783247 7200000.00 .041923295 7400000.00 .068063254 52510 5 360 7600000.00 .094203131 7800000.00 .120342931 8000000.00 .134130775 52510 5 361 8200000.00 .135463353 8400000.00 .136795927 8600000.00 .138128499 52510 5 362 8800000.00 .139461068 9000000.00 .139776366 9200000.00 .139066833 52510 5 363 9400000.00 .138357301 9600000.00 .137647771 9800000.00 .136938241 52510 5 364 10000000.0 .141258127 11000000.0 .138676391 12000000.0 .130977778 52510 5 365 13000000.0 .129114252 14000000.0 .125908739 15000000.0 .123377391 52510 5 366 16000000.0 .121534614 17000000.0 .119932711 18000000.0 .118918605 52510 5 367 19000000.0 .118086743 20000000.0 .117099590 22000000.0 .119370537 52510 5 368 24000000.0 .122996466 26000000.0 .129368668 28000000.0 .137494060 52510 5 369 30000000.0 .140184648 35000000.0 .146861299 40000000.0 .154384338 52510 5 370 45000000.0 .158762575 50000000.0 .167682700 54000000.0 .173962023 52510 5 371 55000000.0 .174186605 60000000.0 .178418048 70000000.0 .179269756 52510 5 372 80000000.0 .180209402 90000000.0 .182933887 100000000. .185499307 52510 5 373 120000000. .186526253 140000000. .190733370 160000000. .193478508 52510 5 374 180000000. .195599513 200000000. .196871373 240000000. .196871373 52510 5 375 280000000. .196871373 320000000. .196871373 360000000. .196871373 52510 5 376 400000000. .196871373 440000000. .196871373 480000000. .196871373 52510 5 377 520000000. .196871373 560000000. .196871373 600000000. .196871373 52510 5 378 640000000. .196871373 680000000. .196871373 720000000. .196871373 52510 5 379 760000000. .196871373 800000000. .196871373 840000000. .196871373 52510 5 380 880000000. .196871373 920000000. .196871373 960000000. .196871373 52510 5 381 1.00000E+9 0.0 52510 5 382 0.0 0.0 1001 0 1 106 52510 5 383 106 1 52510 5 384 1000000.00 1.19629E-4 1125000.00 3.63928E-4 1250000.00 6.08180E-4 52510 5 385 1375000.00 8.52396E-4 1500000.00 .001096584 1625000.00 .001340752 52510 5 386 1750000.00 .001584903 1875000.00 .001829041 2000000.00 .002368467 52510 5 387 2125000.00 .003209384 2250000.00 .004050274 2375000.00 .004891143 52510 5 388 2500000.00 .005731992 2625000.00 .006572825 2750000.00 .007413644 52510 5 389 2875000.00 .008254451 3000000.00 .009447383 3125000.00 .010997362 52510 5 390 3250000.00 .012547325 3375000.00 .014097274 3500000.00 .015647211 52510 5 391 3625000.00 .017197136 3750000.00 .018747052 3875000.00 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