IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 22047.0000 46.5484000 0 0 41 12228 1451 1 0.0 0.0 0 0 0 62228 1451 2 1.00000000 60000000.0 0 0 10 22228 1451 3 300.000000 0.0 1 0 258 72228 1451 4 22-Ti- 47 FEI EVAL-Dec03 K.I.Zolotarev 2228 1451 5 DIST-Dec03 20031206 2228 1451 6 ----BROND-2 MATERIAL 2228 2228 1451 7 -----INCIDENT NEUTRON DATA 2228 1451 8 ------ENDF-6 FORMAT 2228 1451 9 ******************************************************************2228 1451 10 IAEA, July 2012 (A. Trkov) 2228 1451 11 Extended cross sections and covariances from 20 to 60 MeV 2228 1451 12 by TENDL-2011, renormalised for continuity. 2228 1451 13 Points were added at threshold to fix the cross section for Q 2228 1451 14 ******************************************************************2228 1451 15 *** Add MF2 ***************************************************** 2228 1451 16 ------Russian Reactor Dosimetry File RRDF-2002 2228 1451 17 ***************************************************************** 2228 1451 18 Author of evaluation: K.I.Zolotarev 2228 1451 19 ***************************************************************** 2228 1451 20 MF=3 2228 1451 21 MT=103 - (n,p) cross section data 2228 1451 22 -------------------------------------- 2228 1451 23 Microscopic experimental data [1-32] were analyzed in the 2228 1451 24 process of preparation of input data base for the evaluation of 2228 1451 25 cross sections and their uncertainty for the Ti-47(n,p)Sc-47 2228 1451 26 reaction. During this procedure all experimental data if it was 2228 1451 27 possible were corrected to the new recommended cross section data 2228 1451 28 for monitor reactions used in the measurements and to the new re- 2228 1451 29 commended decay data from ref. [33]. 2228 1451 30 Excitation function for the Ti-47(n,p)Sc-27 reaction in the 2228 1451 31 energy region from threshold to 20.0 MeV was evaluated by means 2228 1451 32 of statistical analysis of experimental cross section data [1-22].2228 1451 33 Special correction was done with experimental data [5], [9], 2228 1451 34 [10], [11], [15], [21] and [22]. 2228 1451 35 Experimental data of D.L.Smith and J.W.Meadows [5], H.Husain 2228 1451 36 and S.Hunt [9] and Lu Hanlin et al. [22] were renormalized to the 2228 1451 37 results of W.Mannhart et al. measurements [16] in the overlapping 2228 1451 38 energy intervals. Cross section data from ref.[5] measured in the 2228 1451 39 energy range 4.014 - 5.937 MeV with using Li-7(p,n)Be-7 neutron 2228 1451 40 source were multiplied to the factor Fc=0.80464 . The results of 2228 1451 41 D.L.Smith and J.W.Meadows measurements [5] obtained in the energy 2228 1451 42 range 5.954 - 9.950 MeV with using neutrons from D(d,n)He-3 reac- 2228 1451 43 tion were corrected to the factor Fc=0.87505 . 2228 1451 44 Experimental data of H.A.Husain and S.E.Hunt [9] in the neut- 2228 1451 45 ron energy range 2.351 - 4.288 MeV and Lu Hanlin et al. data [22] 2228 1451 46 for the incident neutron energies 6 - 11.4 MeV were renormalized 2228 1451 47 to the factors Fc=0.80529 and Fc=1.02248, respectively. 2228 1451 48 The Ti47(n,p)Sc47 reaction cross sections measured by D.Smith 2228 1451 49 and J.Meadows [5] in the energy range 0.914 - 3.958 MeV relative 2228 1451 50 U-235 fission cross section were renormalized to the ENDF/B-VI 2228 1451 51 standard for monitor reaction. This data were taken into account 2228 1451 52 in the evaluation only in the energy interval 0.914 - 2.754 MeV 2228 1451 53 there they are agree well with experimental data of W.Mannhart et 2228 1451 54 al. [16]. 2228 1451 55 Experimental data [10], [11], [15] and [21] were corrected 2228 1451 56 for the contribution from Ti-48(n,x)Sc-46 reaction. After applied 2228 1451 57 correction data of S.Firkin [10] were renormalized to the experi- 2228 1451 58 mental data of Y.Ikeda et al. at 14.1 MeV [13] (Fc=0.87036). Data 2228 1451 59 of W.V.Hecker et al. [15] were also renormalized to the result of 2228 1451 60 Y.Ikeda et al. measurements at 14.91 MeV [13] (Fc=0.85784). 2228 1451 61 Experimental data for the Ti-47(n,p)Sc-47 reaction obtained 2228 1451 62 by S.K.Ghorai et al. [3] were used partially. Cross section data 2228 1451 63 for 5.0 and 6.1 MeV were rejected due to their big inconsistency 2228 1451 64 with W.Mannhart, D.L.Smith and J.W.Meadows experimental data [16].2228 1451 65 Experimental cross section data [23-32] were also rejected 2228 1451 66 due to their big discrepancy with the main bulk of experimental 2228 1451 67 data [1-22]. In the rejected experiments [23-25], [28-32] cross 2228 1451 68 section values were measured only in a one energy point in the 2228 1451 69 interval 14 - 15 MeV. 2228 1451 70 Statistical analysis of input cross section data was carried 2228 1451 71 out by means of PADE-2 code [34]. Rational function was used as 2228 1451 72 the model function [35]. 2228 1451 73 Evaluated excitation function for the reaction Ti47(n,p)Sc47 2228 1451 74 was tested with using integral experimental data [36-39] for 2228 1451 75 U-235 thermal fission neutron spectrum and evaluated integral ex- 2228 1451 76 perimental data [39] for Cf-252 spontaneous fission neutron spec- 2228 1451 77 trum. Calculated and measured average cross section values for 2228 1451 78 U-235 thermal fission neutron spectrum [40] and Cf-252 sponta- 2228 1451 79 neous fission neutron spectrum [41] are given in the table 1. 2228 1451 80 Table 1 2228 1451 81 ================================================================= 2228 1451 82 TYPE OF SPECTRUM ,mb (calc.) , mb (measured) 2228 1451 83 ----------------------------------------------------------------- 2228 1451 84 U-235 neutron fission 18.145 20.14 +- 0.66 [36] 2228 1451 85 18.01 +- 0.82 [37] 2228 1451 86 17.90 +- 0.36 [38] 2228 1451 87 17.84 +- 0.36 [39] 2228 1451 88 ----------------------------------------------------------------- 2228 1451 89 CF-252 spont. fission 19.534 19.27 +- 0.32 [39] 2228 1451 90 ================================================================= 2228 1451 91 2228 1451 92 MF=33 2228 1451 93 MT=103 - (n,p) cross section cov. matrix 2228 1451 94 --------------------------------------------- 2228 1451 95 Uncertainties in the evaluated excitation function for the 2228 1451 96 Ti-47(n,p)Sc-47 reaction are given in the form of relative cova- 2228 1451 97 riance matrix for the 38-neutron energy groups (LB=5). Covariance 2228 1451 98 matrix of uncertainties was calculated simultaneously with recom- 2228 1451 99 mended cross section data by means of PADE-2 code. 2228 1451 100 Eigenvalues of the 6-th digits relative covariance matrix 2228 1451 101 given in the 33-file are the following: 2228 1451 102 2228 1451 103 3.54548E-07 3.58900E-07 3.66204E-07 3.75883E-07 2228 1451 104 3.90527E-07 4.05190E-07 4.27795E-07 4.47637E-07 2228 1451 105 4.79974E-07 5.04967E-07 5.48618E-07 5.79467E-07 2228 1451 106 6.34694E-07 6.75119E-07 7.41245E-07 7.96018E-07 2228 1451 107 8.71170E-07 9.48743E-07 1.03206E-06 1.14158E-06 2228 1451 108 1.23977E-06 1.38260E-06 1.52693E-06 1.68260E-06 2228 1451 109 1.98527E-06 2.08478E-06 3.01383E-06 7.61925E-05 2228 1451 110 6.94258E-04 1.02297E-03 1.19552E-03 2.90196E-03 2228 1451 111 5.65030E-03 7.92796E-03 8.69316E-03 1.42193E-02 2228 1451 112 2.05677E-02 9.36503E-02 2228 1451 113 2228 1451 114 References : 2228 1451 115 1. W.G.Cross, H.L.Pai Progress Report EANDC(CAN)-16, p.1, 2228 1451 116 January 1963 2228 1451 117 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2228 1451 118 3. S.K.Ghorai et al. J. Nucl. Energy, v.25, p.319, August 1971 2228 1451 119 4. F.Foroughi, J.Rossel Helvetica Physica Acta (Switzerland), 2228 1451 120 v.45, p.439, August 1972 2228 1451 121 5. D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.58, p.314, 1975 2228 1451 122 6. R.Spangler et al. J. Trans. Amer. Nucl. Soc., v.22, p.818, 2228 1451 123 November 1975 2228 1451 124 7. H.Gotoh et al. Prog. Report, JAERY-6320, p.165, November 1975 2228 1451 125 8. S.M.Qaim, N.I.Molla Nucl. Phys.,v.A283, p.269, June 1977 2228 1451 126 9. H.A.Husain, S.E.Hunt Int. J. of Applied and Isotopes, v.34, 2228 1451 127 no.4, p.731, 1983 2228 1451 128 10. S.Firkin Report AERE-M-3350, Harwell, September 1983 2228 1451 129 11. R.Pepelink et al. Progress Report, NEANDC(E)-262U,(5), p.32, 2228 1451 130 June 1985 2228 1451 131 12. Hoang Dac Luc et al. Prog. Report INDC(VN)-5, September 1986 2228 1451 132 13. Y.Ikeda et al. Report JAERI-1312, March 1988 2228 1451 133 14. K.Kobayashi, I.Kimura Proc. of the Intern. Conf. on Nuclear 2228 1451 134 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2228 1451 135 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 2228 1451 136 15. W.V.Hecker et al. Nucl. Instr. Meth., v.B40/41, p.478, 2228 1451 137 April 1989 2228 1451 138 16. W.Mannhart, D.L.Smith, J.W.Meadows Prog. Report, NEANDC-259, 2228 1451 139 p.121, September 1989 2228 1451 140 17. Lu Hanlin et al. Report, INDC(CPR)-16, IAEA, Vienna, 2228 1451 141 September 1989 2228 1451 142 18. Y.Ikeda, C.Konno, K.Kosako, K.Oishi Progress Report, 2228 1451 143 INDC(JPN)-142, p.11, IAEA, Vienna, August 1990 2228 1451 144 19. M.Viennot et al. Sci. Eng., v.108, p.289, July 1991 2228 1451 145 20. S.M.Qaim et al. Proc. of an Int. Conf. on Nuclear Data for 2228 1451 146 Science and Technology, Julich, FRG, 13-17 May 1991, Springer 2228 1451 147 Verlag, Berlin - Heidelberg, 1992, p.297-300 2228 1451 148 21. Y.Ikeda et al. Proc. of an Int. Conf. on Nuclear Data for 2228 1451 149 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 2228 1451 150 Verlag, Berlin - Heidelberg, 1992, p.294-296 2228 1451 151 22. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2228 1451 152 23. A.Poularikas, R.W.Fink Phys. Rev., v.115, p.989, 1959 2228 1451 153 24. D.L.Allan Nucl. Phys., v.24, p.274, April 1961 2228 1451 154 25. M.Hillman Nucl. Phys, v.37, p.78, 1962 2228 1451 155 26. L.Gonzalez et al. Phys. Rev., v.126, p.271, April 1962 2228 1451 156 27. F.G.Armitage EXFOR 30045.002 2228 1451 157 28. V.N.Levkovskiy et al. Jadernaja Fizika, v.10, n.1, p.44, 2228 1451 158 July 1969 2228 1451 159 29. V.K.Tikku et al. Proc. of Nucl. Phys. and Solid State Phys. 2228 1451 160 Symp., Chandigarh, 28 Dec 1972 - 1 Jan 1973, Vol. 15B, p.115 2228 1451 161 30. I.Ribansky, S.Gmuca J. of Physics, pt.G, v.9, p.1537, 2228 1451 162 December 1983 2228 1451 163 31. N.I.Molla et al. Report INDC(BAN)-003, September 1986 2228 1451 164 32. K.T.Osman, F.I.Habbani Report, INDC(SUD)-001, IAEA, Vienna, 2228 1451 165 33. O.Bersillon Decay Data and Isotopic Abundances for Dosimetry 2228 1451 166 Applications, Report INDC(NDS)-448, p.95, IAEA, Vienna, 2228 1451 167 October 2003 2228 1451 168 34. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2228 1451 169 35. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2228 1451 170 st's Meeting on Evaluation and Processing of Covariance Data, 2228 1451 171 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2228 1451 172 36. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2228 1451 173 Washington D.C., 25-28 April 1989, v.2, p.923 2228 1451 174 37. K.Kobayashi, T.Kobayashi Progress Report NEANDC(J)-155/U, 2228 1451 175 p.52, August 1990 2228 1451 176 38. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 2228 1451 177 39. W.Mannhart Validation of Differential Cross Sections with 2228 1451 178 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2228 1451 179 September 2002 2228 1451 180 40. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 2228 1451 181 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 2228 1451 182 41. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2228 1451 183 ***************************************************************** 2228 1451 184 2228 1451 185 ***************************************************************** 2228 1451 186 ******** Adopted from IRDF-2002 ******* 2228 1451 187 ***************************************************************** 2228 1451 188 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2228 1451 189 2228 1451 190 For this special purpose library it was decided the reaction 2228 1451 191 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2228 1451 192 This was done after processing through the codes. The 2228 1451 193 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2228 1451 194 ******************************************************************2228 1451 195 MF=3 2228 1451 196 MT= 28+104 -(n,np+pn+d) --> Sc-46(m+g) production 2228 1451 197 ------------------------------------------------- 2228 1451 198 In this section is given the sum of cross section of the reac-2228 1451 199 tions Ti47(n,np)Sc46m+g , Ti47(n,pn)Sc46m+g and Ti47(n,d)Sc46m+g. 2228 1451 200 Excitation function for the Ti47(n,x)Sc46m+g reaction in the 2228 1451 201 energy region from threshold to 20 MeV was evaluated by means of 2228 1451 202 statistical analysis of experimental cross section data [1-6] and 2228 1451 203 data from STAPRE [7] calculation. 2228 1451 204 All experimental data were renormalized to the new standards 2228 1451 205 for monitor reactions cross sections and decay data. 2228 1451 206 The final procedure of evaluation Ti47(n,x)Sc46m+g excitation 2228 1451 207 function from threshold to 20 MeV has been carried out within the 2228 1451 208 framework of generalized least squares method. Rational function 2228 1451 209 was used as model function [8]. Calculations was performed by 2228 1451 210 means of Pade-2 code [9]. 2228 1451 211 U-235 thermal fission [10] and Cf-252 spontaneous fission 2228 1451 212 neutron spectra [11] averaged cross-sections calculated from the 2228 1451 213 evaluated Ti47(n,x)Sc46m+g excitation function are the following: 2228 1451 214 2228 1451 215 -------------------------------------------- 2228 1451 216 TYPE OF SPECTRUM I , mb (calc.) 2228 1451 217 --------------------------I----------------- 2228 1451 218 U-235 neutron fission I 8.1158E-3 2228 1451 219 CF-252 spontan. fission I 1.9201E-2 2228 1451 220 2228 1451 221 MF=33 2228 1451 222 MT= 28 -(n,np+pn+d) cross section cov. matrix 2228 1451 223 --------------------------------------------- 2228 1451 224 Uncertainties in the evaluated excitation function for the 2228 1451 225 reaction Ti-47(n,x)Sc-46m+g are given in the form of relative 2228 1451 226 covariance matrix for the 17-neutron energy groups (LB=5). Cova- 2228 1451 227 riance matrix of uncertainties was calculated simultaneously with 2228 1451 228 recommended cross section data by means of PADE-2 code. 2228 1451 229 Eigenvalues of the 6-th digits relative covariance matrix 2228 1451 230 given in the 33-file are the following: 2228 1451 231 2228 1451 232 3.91044E-07 4.46920E-07 5.37497E-07 7.02044E-07 2228 1451 233 9.85160E-07 1.54860E-06 2.86775E-06 6.97782E-06 2228 1451 234 2.67223E-05 1.90995E-04 9.14779E-04 5.12707E-03 2228 1451 235 8.27566E-03 1.15029E-02 1.68918E-02 5.65241E-02 2228 1451 236 8.27240E-01 2228 1451 237 2228 1451 238 References : 2228 1451 239 1. W.G.Cross, H.L.Pai Progress Rep. EANDC(CAN)-16, p.1, Jan.1963 2228 1451 240 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2228 1451 241 3. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2228 1451 242 4. N.I.Molla et al. Report INDC(BAN)-003, September 1986 2228 1451 243 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2228 1451 244 6. Y.Uno et al. Report JAERI-M-93-046, p.247-256, 1993 2228 1451 245 7. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2228 1451 246 Activation Cross Section and Related Quantities, Report 2228 1451 247 IRK 76-01, Vienna, 1976 2228 1451 248 8. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2228 1451 249 st's Meeting on Evaluation and Processing of Covariance Data, 2228 1451 250 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2228 1451 251 9. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2228 1451 252 10. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2228 1451 253 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2228 1451 254 11. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2228 1451 255 ***************************************************************** 2228 1451 256 ***************** Program LINEAR (VERSION 2012-1) ***************2228 1451 257 For All Data Greater than 1.0000D-10 barns in Absolute Value 2228 1451 258 Data Linearized to Within an Accuracy of .100000000 per-cent 2228 1451 259 ***************** Program GROUPIE (VERSION 2012-1) **************2228 1451 260 Unshielded Group Averages Using 640 Groups 2228 1451 261 Weighting Spectrum: Flat (Constant) Spectrum 2228 1451 262 1 451 263 12228 1451 263 2 151 4 12228 1451 264 3 103 124 12228 1451 265 8 5 2 12228 1451 266 10 5 32 12228 1451 267 33 103 147 12228 1451 268 40 5 36 02228 1451 269 2228 1 0 270 2228 0 0 271 22047.0000 46.5484000 0 0 1 02228 2151 272 2.204700+4 1.000000+0 0 0 1 02228 2151 273 1.000000-5 1.000000+5 0 0 0 02228 2151 274 3.500000+0 5.733000-1 0 0 0 02228 2151 275 2228 2 0 276 2228 0 0 277 22047.0000 46.5484000 0 0 0 02228 3103 278 182300.000 182300.000 0 0 1 1962228 3103 279 196 1 2228 3103 280 800000.000 2.47807E-8 840000.000 1.62772E-6 880000.000 2.19919E-52228 3103 281 920000.000 1.29380E-4 960000.000 3.13100E-4 1000000.00 6.87935E-42228 3103 282 1100000.00 .001222038 1200000.00 .002735627 1300000.00 .0039225142228 3103 283 1400000.00 .002692951 1500000.00 .003595615 1600000.00 .0064833642228 3103 284 1700000.00 .009098379 1800000.00 .009378613 1900000.00 .0115424932228 3103 285 2000000.00 .016761920 2100000.00 .017809880 2200000.00 .0212343502228 3103 286 2300000.00 .023725200 2400000.00 .025511750 2500000.00 .0259080052228 3103 287 2600000.00 .028487155 2700000.00 .029799425 2800000.00 .0312817002228 3103 288 2900000.00 .032882200 3000000.00 .034578350 3100000.00 .0363369502228 3103 289 3200000.00 .038131050 3300000.00 .039939600 3400000.00 .0417463502228 3103 290 3500000.00 .043539000 3600000.00 .045288950 3700000.00 .0469970002228 3103 291 3800000.00 .048681300 3900000.00 .050334000 4000000.00 .0519521502228 3103 292 4100000.00 .053465750 4200000.00 .054870200 4300000.00 .0562714362228 3103 293 4400000.00 .057584188 4500000.00 .058884700 4600000.00 .0602633002228 3103 294 4700000.00 .061576087 4800000.00 .062854624 4900000.00 .0641337372228 3103 295 5000000.00 .065383350 5100000.00 .066606000 5200000.00 .0678057382228 3103 296 5300000.00 .069019025 5400000.00 .070214088 5500000.00 .0714672502228 3103 297 5600000.00 .072745400 5700000.00 .073999037 5800000.00 .0752464262228 3103 298 5900000.00 .076453888 6000000.00 .077620350 6100000.00 .0787266892228 3103 299 6200000.00 .079832578 6300000.00 .080939578 6400000.00 .0820493892228 3103 300 6500000.00 .083184550 6600000.00 .084346988 6700000.00 .0854960822228 3103 301 6800000.00 .086632537 6900000.00 .087778343 7000000.00 .0889343832228 3103 302 7100000.00 .090101522 7200000.00 .091280476 7300000.00 .0924717712228 3103 303 7400000.00 .093675933 7500000.00 .094893238 7600000.00 .0961238352228 3103 304 7700000.00 .097367885 7800000.00 .098625375 7900000.00 .0998962242228 3103 305 8000000.00 .101179987 8100000.00 .102475987 8200000.00 .1037844742228 3103 306 8300000.00 .105104474 8400000.00 .106434473 8500000.00 .1077744872228 3103 307 8600000.00 .109122986 8700000.00 .110478973 8800000.00 .1118409732228 3103 308 8900000.00 .113207473 9000000.00 .114576986 9100000.00 .1159474862228 3103 309 9200000.00 .117317473 9300000.00 .118684473 9400000.00 .1200464732228 3103 310 9500000.00 .121401473 9600000.00 .122746973 9700000.00 .1240799732228 3103 311 9800000.00 .125397474 9900000.00 .126697487 10000000.0 .1279770002228 3103 312 10100000.0 .129232000 10200000.0 .130460500 10300000.0 .1316590002228 3103 313 10400000.0 .132824000 10500000.0 .133953000 10600000.0 .1350420002228 3103 314 10700000.0 .136087500 10800000.0 .137086500 10900000.0 .1380360002228 3103 315 11000000.0 .138932500 11100000.0 .139773000 11200000.0 .1405545002228 3103 316 11300000.0 .141273500 11400000.0 .141928000 11500000.0 .1425150002228 3103 317 11600000.0 .143032000 11700000.0 .143477000 11800000.0 .1438480002228 3103 318 11900000.0 .144143000 12000000.0 .144360500 12100000.0 .1444995002228 3103 319 12200000.0 .144558500 12300000.0 .144537500 12400000.0 .1444360002228 3103 320 12500000.0 .144253500 12600000.0 .143990500 12700000.0 .1436480002228 3103 321 12800000.0 .143226000 12900000.0 .142726000 13000000.0 .1421500002228 3103 322 13100000.0 .141498500 13200000.0 .140774000 13300000.0 .1399790002228 3103 323 13400000.0 .139115500 13500000.0 .138186000 13600000.0 .1371930002228 3103 324 13700000.0 .136140000 13800000.0 .135030000 13900000.0 .1338655002228 3103 325 14000000.0 .132650000 14100000.0 .131387000 14200000.0 .1300800002228 3103 326 14300000.0 .128732000 14400000.0 .127346000 14500000.0 .1259260002228 3103 327 14600000.0 .124475500 14700000.0 .122997500 14800000.0 .1214950002228 3103 328 14900000.0 .119971000 15000000.0 .118429000 15100000.0 .1168720002228 3103 329 15200000.0 .115302500 15300000.0 .113723000 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