JENDL-5 (Neutron sublibrary (activation cs)) Fm-255 0 0 0 0 1.002550+5 2.528990+2 0 0 6 19936 1451 1 0.000000+0 1.000000+0 0 0 0 69936 1451 2 1.000000+0 2.000000+7 0 0 10 59936 1451 3 2.936000+2 0.000000+0 4 0 341 149936 1451 4 100-Fm-255 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al. 9936 1451 5 DIST-DEC21 20100304 9936 1451 6 ----JENDL-5 MATERIAL 9936 9936 1451 7 -----INCIDENT NEUTRON DATA 9936 1451 8 ------ENDF-6 FORMAT 9936 1451 9 9936 1451 10 History 9936 1451 11 07-09 Theoretical calculation was performed with CCONE code. 9936 1451 12 07-11 Data were compiled as JENDL/AC-2008/1/ 9936 1451 13 09-08 (MF1,MT458) was evaluated. 9936 1451 14 10-01 Data of prompt gamma rays due to fission were given. 9936 1451 15 10-03 Covariance data were given. 9936 1451 16 9936 1451 17 21-11 revised by O.Iwamoto 9936 1451 18 (MF3/MT19-21,38) deleted 9936 1451 19 (MF8/MT4,16-18,37,102) added 9936 1451 20 9936 1451 21 MF=1 General information 9936 1451 22 MT=452 Number of Neutrons per fission 9936 1451 23 Sum of MT's=455 and 456. 9936 1451 24 9936 1451 25 MT=455 Delayed neutron data 9936 1451 26 (Same as JENDL-3.3) 9936 1451 27 Estimated from systematics by Tuttle/2/, Benedetti et al./3/ 9936 1451 28 and Waldo et al./4/ 9936 1451 29 Decay constants were evaluated by Brady and England/5/. 9936 1451 30 9936 1451 31 MT=456 Number of prompt neutrons per fission 9936 1451 32 A value of 4.0 measured by Flynn et al./6/ was adopted at 9936 1451 33 the thermal energy. A coefficient of energy dependent term 9936 1451 34 was estimated from Howerton's systematics/7/. 9936 1451 35 9936 1451 36 MT=458 Components of energy release due to fission 9936 1451 37 Total energy and prompt energy were calculated from mass 9936 1451 38 balance using JENDL-4 fission yields data and mass excess 9936 1451 39 evaluation. Mass excess values were from Audi's 2009 9936 1451 40 evaluation/8/. Delayed energy values were calculated from 9936 1451 41 the energy release for infinite irradiation using JENDL FP 9936 1451 42 Decay Data File 2000 and JENDL-4 yields data. For delayed 9936 1451 43 neutron energy, as the JENDL FP Decay Data File 2000/9/ does 9936 1451 44 not include average neutron energy values, the average values 9936 1451 45 were calculated using the formula shown in the report by 9936 1451 46 T.R. England/10/. The fractions of prompt energy were 9936 1451 47 calculated using the fractions of Sher's evaluation/11/ when 9936 1451 48 they were provided. When the fractions were not given by Sher,9936 1451 49 averaged fractions were used. 9936 1451 50 9936 1451 51 9936 1451 52 MF= 2 Resonance parameters 9936 1451 53 MT=151 9936 1451 54 No resonance parameters are given. 9936 1451 55 9936 1451 56 9936 1451 57 Thermal cross sections and resonance integrals (at 300K) 9936 1451 58 ------------------------------------------------------- 9936 1451 59 0.0253 eV reson. integ.(*) 9936 1451 60 (barns) (barns) 9936 1451 61 ------------------------------------------------------- 9936 1451 62 total 3641.4 9936 1451 63 elastic 9.319 9936 1451 64 fission 3361.6 1460 9936 1451 65 capture 270.1 116 9936 1451 66 ------------------------------------------------------- 9936 1451 67 (*) In the energy range from 0.5 eV to 10 MeV. 9936 1451 68 9936 1451 69 9936 1451 70 MF= 3 Neutron cross sections 9936 1451 71 Below 0.45 eV: 9936 1451 72 * Elastic scattering cross section is 9.3 b calculated from 9936 1451 73 scattering radius of 8.62 fm/12/. 9936 1451 74 * Fission cross section is in the 1/v shape. 9936 1451 75 3360+-150 b at 0.0253 eV/13,6/. 9936 1451 76 * Capture cross section is in the 1/v shape. 9936 1451 77 270 b at 0.0253 eV which was estimated from the ratio of 9936 1451 78 capture and fission cross sections calculated with CCONE 9936 1451 79 code/12/ at 1 eV and the thermal fission cross section. 9936 1451 80 9936 1451 81 Above 0.45 eV: 9936 1451 82 Cross sections were calculated with CCONE code/12/. 9936 1451 83 9936 1451 84 MT= 1 Total cross section 9936 1451 85 The cross section was calculated with CC OMP of Soukhovitskii 9936 1451 86 et al./14/ 9936 1451 87 9936 1451 88 9936 1451 89 MF= 4 Angular distributions of secondary neutrons 9936 1451 90 MT=2 Elastic scattering 9936 1451 91 Calculated with CCONE code/12/. 9936 1451 92 9936 1451 93 MT=18 Fission 9936 1451 94 Isotropic distributions in the laboratory system were assumed.9936 1451 95 9936 1451 96 9936 1451 97 MF= 5 Energy distributions of secondary neutrons 9936 1451 98 MT=18 Prompt netrons 9936 1451 99 Calculated with CCONE code/12/. 9936 1451 100 9936 1451 101 MT=455 Delayed neutrons 9936 1451 102 Calculated by Brady and England/5/. 9936 1451 103 9936 1451 104 9936 1451 105 MF= 6 Energy-angle distributions 9936 1451 106 Calculated with CCONE code/12/. 9936 1451 107 Distributions from fission (MT=18) are not included. 9936 1451 108 9936 1451 109 9936 1451 110 MF=12 Photon production multiplicities 9936 1451 111 MT=18 Fission 9936 1451 112 Calculated from the total energy released by the prompt 9936 1451 113 gamma-rays due to fission given in MF=1/MT=458 and the 9936 1451 114 average energy of gamma-rays. 9936 1451 115 9936 1451 116 9936 1451 117 MF=14 Photon angular distributions 9936 1451 118 MT=18 Fission 9936 1451 119 Isotoropic distributions were assumed. 9936 1451 120 9936 1451 121 9936 1451 122 MF=15 Continuous photon energy spectra 9936 1451 123 MT=18 Fission 9936 1451 124 Experimental data measured by Verbinski et al./15/ for 9936 1451 125 Pu-239 thermal fission were adopted. 9936 1451 126 9936 1451 127 9936 1451 128 MF=31 Covariances of average number of neutrons per fission 9936 1451 129 MT=452 Number of neutrons per fission 9936 1451 130 Sum of covariances for MT=455 and MT=456. 9936 1451 131 9936 1451 132 MT=455 9936 1451 133 Error of 15% was assumed. 9936 1451 134 9936 1451 135 MT=456 9936 1451 136 Covariance was obtained by fitting a linear function to the 9936 1451 137 at 0.0 and 5.0 MeV with uncertainties of 12.5% and 10%, 9936 1451 138 respectively. 9936 1451 139 9936 1451 140 9936 1451 141 MF=33 Covariances of neutron cross sections 9936 1451 142 Covariances were given to all the cross sections by using 9936 1451 143 KALMAN code/16/ and the covariances of model parameters 9936 1451 144 used in the cross-section calculations. 9936 1451 145 9936 1451 146 For the following cross sections, standard deviations in the 9936 1451 147 energy region below 0.45 eV were assumed as follows: 9936 1451 148 9936 1451 149 Total 46 % 9936 1451 150 Elastic scattering 50 % 9936 1451 151 Fission 50 % estimated from experimental data 9936 1451 152 Capture 50 % 9936 1451 153 9936 1451 154 9936 1451 155 MF=34 Covariances for Angular Distributions 9936 1451 156 MT=2 Elastic scattering 9936 1451 157 Covariances were given only to P1 components. 9936 1451 158 9936 1451 159 9936 1451 160 MF=35 Covariances for Energy Distributions 9936 1451 161 MT=18 Fission spectra 9936 1451 162 Estimated with CCONE and KALMAN codes. 9936 1451 163 9936 1451 164 9936 1451 165 ***************************************************************** 9936 1451 166 Calculation with CCONE code 9936 1451 167 ***************************************************************** 9936 1451 168 9936 1451 169 Models and parameters used in the CCONE/12/ calculation 9936 1451 170 1) Coupled channel optical model 9936 1451 171 Levels in the rotational band were included. Optical model 9936 1451 172 potential and coupled levels are shown in Table 1. 9936 1451 173 9936 1451 174 2) Two-component exciton model/17/ 9936 1451 175 * Global parametrization of Koning-Duijvestijn/18/ 9936 1451 176 was used. 9936 1451 177 * Gamma emission channel/19/ was added to simulate direct 9936 1451 178 and semi-direct capture reaction. 9936 1451 179 9936 1451 180 3) Hauser-Feshbach statistical model 9936 1451 181 * Moldauer width fluctuation correction/20/ was included. 9936 1451 182 * Neutron, gamma and fission decay channel were included. 9936 1451 183 * Transmission coefficients of neutrons were taken from 9936 1451 184 coupled channel calculation in Table 1. 9936 1451 185 * The level scheme of the target is shown in Table 2. 9936 1451 186 * Level density formula of constant temperature and Fermi-gas 9936 1451 187 model were used with shell energy correction and collective 9936 1451 188 enhancement factor. Parameters are shown in Table 3. 9936 1451 189 * Fission channel: 9936 1451 190 Double humped fission barriers were assumed. 9936 1451 191 Fission barrier penetrabilities were calculated with 9936 1451 192 Hill-Wheler formula/21/. Fission barrier parameters were 9936 1451 193 shown in Table 4. Transition state model was used and 9936 1451 194 continuum levels are assumed above the saddles. The level 9936 1451 195 density parameters for inner and outer saddles are shown in 9936 1451 196 Tables 5 and 6, respectively. 9936 1451 197 * Gamma-ray strength function of Kopecky et al/22/,/23/ 9936 1451 198 was used. The prameters are shown in Table 7. 9936 1451 199 9936 1451 200 9936 1451 201 ------------------------------------------------------------------9936 1451 202 Tables 9936 1451 203 ------------------------------------------------------------------9936 1451 204 9936 1451 205 Table 1. Coupled channel calculation 9936 1451 206 -------------------------------------------------- 9936 1451 207 * rigid rotor model was applied 9936 1451 208 * coupled levels = 0,1,3 (see Table 2) 9936 1451 209 * optical potential parameters /14/ 9936 1451 210 Volume: 9936 1451 211 V_0 = 49.97 MeV 9936 1451 212 lambda_HF = 0.01004 1/MeV 9936 1451 213 C_viso = 15.9 MeV 9936 1451 214 A_v = 12.04 MeV 9936 1451 215 B_v = 81.36 MeV 9936 1451 216 E_a = 385 MeV 9936 1451 217 r_v = 1.2568 fm 9936 1451 218 a_v = 0.633 fm 9936 1451 219 Surface: 9936 1451 220 W_0 = 17.2 MeV 9936 1451 221 B_s = 11.19 MeV 9936 1451 222 C_s = 0.01361 1/MeV 9936 1451 223 C_wiso = 23.5 MeV 9936 1451 224 r_s = 1.1803 fm 9936 1451 225 a_s = 0.601 fm 9936 1451 226 Spin-orbit: 9936 1451 227 V_so = 5.75 MeV 9936 1451 228 lambda_so = 0.005 1/MeV 9936 1451 229 W_so = -3.1 MeV 9936 1451 230 B_so = 160 MeV 9936 1451 231 r_so = 1.1214 fm 9936 1451 232 a_so = 0.59 fm 9936 1451 233 Coulomb: 9936 1451 234 C_coul = 1.3 9936 1451 235 r_c = 1.2452 fm 9936 1451 236 a_c = 0.545 fm 9936 1451 237 Deformation: 9936 1451 238 beta_2 = 0.213 9936 1451 239 beta_4 = 0.066 9936 1451 240 beta_6 = 0.0015 9936 1451 241 9936 1451 242 * Calculated strength function 9936 1451 243 S0= 1.34e-4 S1= 3.34e-4 R'= 8.62 fm (En=1 keV) 9936 1451 244 -------------------------------------------------- 9936 1451 245 9936 1451 246 Table 2. Level Scheme of Fm-255 9936 1451 247 ------------------- 9936 1451 248 No. Ex(MeV) J PI 9936 1451 249 ------------------- 9936 1451 250 0 0.00000 7/2 + * 9936 1451 251 1 0.06000 9/2 + * 9936 1451 252 2 0.08400 11/2 - 9936 1451 253 3 0.13300 11/2 + * 9936 1451 254 4 0.16500 13/2 - 9936 1451 255 5 0.23800 15/2 - 9936 1451 256 6 0.27200 9/2 + 9936 1451 257 7 0.31800 11/2 + 9936 1451 258 ------------------- 9936 1451 259 *) Coupled levels in CC calculation 9936 1451 260 9936 1451 261 Table 3. Level density parameters 9936 1451 262 -------------------------------------------------------- 9936 1451 263 Nuclide a* Pair Eshell T E0 Ematch 9936 1451 264 1/MeV MeV MeV MeV MeV MeV 9936 1451 265 -------------------------------------------------------- 9936 1451 266 Fm-256 19.5580 1.5000 1.5985 0.3543 0.1323 3.3878 9936 1451 267 Fm-255 19.4922 0.7515 1.2018 0.3532 -0.5536 2.5647 9936 1451 268 Fm-254 19.4263 1.5059 0.8560 0.3723 0.0580 3.5059 9936 1451 269 Fm-253 19.3604 0.7544 0.3502 0.3909 -0.8123 2.9141 9936 1451 270 Fm-252 19.2945 1.5119 0.1288 0.3822 0.0732 3.5119 9936 1451 271 -------------------------------------------------------- 9936 1451 272 9936 1451 273 Table 4. Fission barrier parameters 9936 1451 274 ---------------------------------------- 9936 1451 275 Nuclide V_A hw_A V_B hw_B 9936 1451 276 MeV MeV MeV MeV 9936 1451 277 ---------------------------------------- 9936 1451 278 Fm-256 6.000 1.040 5.000 0.600 9936 1451 279 Fm-255 6.500 1.000 5.500 0.800 9936 1451 280 Fm-254 6.500 1.040 5.500 0.600 9936 1451 281 Fm-253 6.500 1.000 5.500 0.800 9936 1451 282 Fm-252 6.500 1.040 5.500 0.600 9936 1451 283 ---------------------------------------- 9936 1451 284 9936 1451 285 Table 5. Level density above inner saddle 9936 1451 286 -------------------------------------------------------- 9936 1451 287 Nuclide a* Pair Eshell T E0 Ematch 9936 1451 288 1/MeV MeV MeV MeV MeV MeV 9936 1451 289 -------------------------------------------------------- 9936 1451 290 Fm-256 21.9050 1.7500 2.3000 0.3196 -0.6668 3.7500 9936 1451 291 Fm-255 21.8312 0.8767 2.3000 0.3201 -1.5401 2.8767 9936 1451 292 Fm-254 21.7575 1.7569 2.3000 0.3207 -0.6600 3.7569 9936 1451 293 Fm-253 21.6837 0.8802 2.3000 0.3213 -1.5368 2.8802 9936 1451 294 Fm-252 21.6098 1.7638 2.3000 0.3219 -0.6531 3.7638 9936 1451 295 -------------------------------------------------------- 9936 1451 296 9936 1451 297 Table 6. Level density above outer saddle 9936 1451 298 -------------------------------------------------------- 9936 1451 299 Nuclide a* Pair Eshell T E0 Ematch 9936 1451 300 1/MeV MeV MeV MeV MeV MeV 9936 1451 301 -------------------------------------------------------- 9936 1451 302 Fm-256 21.9050 1.7500 1.1200 0.3442 0.0370 3.7500 9936 1451 303 Fm-255 21.8312 0.8767 1.0800 0.3453 -0.8358 2.8767 9936 1451 304 Fm-254 21.7575 1.7569 1.0400 0.3464 0.0450 3.7569 9936 1451 305 Fm-253 21.6837 0.8802 1.0000 0.3474 -0.8312 2.8802 9936 1451 306 Fm-252 21.6098 1.7638 0.9600 0.3485 0.0531 3.7638 9936 1451 307 -------------------------------------------------------- 9936 1451 308 9936 1451 309 Table 7. Gamma-ray strength function for Fm-256 9936 1451 310 -------------------------------------------------------- 9936 1451 311 K0 = 1.500 E0 = 4.500 (MeV) 9936 1451 312 * E1: ER = 11.36 (MeV) EG = 2.70 (MeV) SIG = 263.70 (mb) 9936 1451 313 ER = 14.15 (MeV) EG = 4.10 (MeV) SIG = 527.39 (mb) 9936 1451 314 * M1: ER = 6.46 (MeV) EG = 4.00 (MeV) SIG = 1.89 (mb) 9936 1451 315 * E2: ER = 9.92 (MeV) EG = 3.04 (MeV) SIG = 7.65 (mb) 9936 1451 316 -------------------------------------------------------- 9936 1451 317 9936 1451 318 9936 1451 319 References 9936 1451 320 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 9936 1451 321 2) R.J.Tuttld: INDC(NDS)-107/G+Special, p.29 (1979). 9936 1451 322 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 9936 1451 323 4) R.Waldo et al.: Phys. Rev., C23, 1113 (1981). 9936 1451 324 5) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989). 9936 1451 325 6) K.F.Flynn et al.: Phys. Rev., C11, 1676 (1975).. 9936 1451 326 7) R.J.Howerton: Nucl. Sci. Eng., 62, 438 (1977). 9936 1451 327 8) G.Audi: Private communication (April 2009). 9936 1451 328 9) J.Katakura et al.: JAERI 1343 (2001). 9936 1451 329 10) T.R.England et al.: LA-11151-MS (1988). 9936 1451 330 11) R.Sher, C.Beck: EPRI NP-1771 (1981). 9936 1451 331 12) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 9936 1451 332 13) R.C.Ragaini et al.: Phys. Rev., C9. 399 (1974). 9936 1451 333 14) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005). 9936 1451 334 15) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973). 9936 1451 335 16) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in 9936 1451 336 Japanese. 9936 1451 337 17) C.Kalbach: Phys. Rev. C33, 818 (1986). 9936 1451 338 18) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 9936 1451 339 19) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 9936 1451 340 20) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 9936 1451 341 21) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 9936 1451 342 22) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 9936 1451 343 23) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). 9936 1451 344 9936 1451 345 1 451 359 19936 1451 346 2 151 4 19936 1451 347 3 4 27 19936 1451 348 3 16 14 19936 1451 349 3 17 9 19936 1451 350 3 18 116 19936 1451 351 3 37 5 19936 1451 352 3 102 145 19936 1451 353 8 4 2 19936 1451 354 8 16 2 19936 1451 355 8 17 2 19936 1451 356 8 18 2 19936 1451 357 8 37 2 19936 1451 358 8 102 2 19936 1451 359 9936 1 099999 9936 0 0 0 1.002550+5 2.528990+2 0 0 1 09936 2151 1 1.002550+5 1.000000+0 0 0 1 09936 2151 2 1.000000-5 4.500000-1 0 0 0 09936 2151 3 3.500000+0 8.620000-1 0 0 0 09936 2151 4 9936 2 099999 9936 0 0 0 1.002550+5 2.528990+2 0 0 0 09936 3 4 1 0.000000+0-6.000000+4 0 0 1 729936 3 4 2 72 2 9936 3 4 3 6.023720+4 0.000000+0 7.000000+4 6.087240-2 8.433210+4 1.134250-19936 3 4 4 1.000000+5 2.004988-1 1.108720+5 2.437571-1 1.200000+5 2.800780-19936 3 4 5 1.335260+5 3.275102-1 1.400000+5 3.549482-1 1.656520+5 4.609186-19936 3 4 6 1.700000+5 4.774891-1 2.000000+5 5.866012-1 2.389410+5 6.876114-19936 3 4 7 2.500000+5 7.132257-1 2.730760+5 7.607331-1 3.000000+5 8.438183-19936 3 4 8 3.192570+5 8.807395-1 4.000000+5 1.014865+0 5.000000+5 1.159123+09936 3 4 9 6.000000+5 1.285832+0 7.000000+5 1.410896+0 8.000000+5 1.514482+09936 3 4 10 9.000000+5 1.602149+0 1.000000+6 1.687234+0 1.100000+6 1.753940+09936 3 4 11 1.200000+6 1.808042+0 1.300000+6 1.846404+0 1.400000+6 1.877140+09936 3 4 12 1.500000+6 1.890400+0 1.600000+6 1.898169+0 1.700000+6 1.890864+09936 3 4 13 1.800000+6 1.879867+0 1.900000+6 1.859095+0 2.000000+6 1.836109+09936 3 4 14 2.100000+6 1.808540+0 2.300000+6 1.750708+0 2.500000+6 1.695301+09936 3 4 15 2.700000+6 1.647591+0 3.000000+6 1.589619+0 3.500000+6 1.536353+09936 3 4 16 4.000000+6 1.491591+0 4.500000+6 1.484009+0 5.000000+6 1.479844+09936 3 4 17 5.500000+6 1.448834+0 6.000000+6 1.207271+0 6.500000+6 9.567108-19936 3 4 18 7.000000+6 7.760013-1 7.500000+6 6.490489-1 8.000000+6 5.684913-19936 3 4 19 8.500000+6 5.123004-1 9.000000+6 4.736323-1 9.500000+6 4.438948-19936 3 4 20 1.000000+7 4.205444-1 1.050000+7 4.025617-1 1.100000+7 3.828623-19936 3 4 21 1.150000+7 3.671405-1 1.200000+7 3.517769-1 1.250000+7 3.377346-19936 3 4 22 1.300000+7 3.250118-1 1.350000+7 3.112406-1 1.400000+7 2.994185-19936 3 4 23 1.450000+7 2.892703-1 1.500000+7 2.759883-1 1.550000+7 2.662976-19936 3 4 24 1.600000+7 2.566976-1 1.650000+7 2.478847-1 1.700000+7 2.390486-19936 3 4 25 1.750000+7 2.310493-1 1.800000+7 2.232615-1 1.850000+7 2.180213-19936 3 4 26 1.900000+7 2.105026-1 1.950000+7 2.040952-1 2.000000+7 1.998367-19936 3 4 27 9936 3 099999 1.002550+5 2.528990+2 0 0 0 09936 3 16 1 -5.176250+6-5.176250+6 0 0 1 319936 3 16 2 31 2 9936 3 16 3 5.196720+6 0.000000+0 5.500000+6 3.137810-2 6.000000+6 2.756300-19936 3 16 4 6.500000+6 4.959100-1 7.000000+6 5.964170-1 7.500000+6 6.145920-19936 3 16 5 8.000000+6 6.042310-1 8.500000+6 5.783610-1 9.000000+6 5.543080-19936 3 16 6 9.500000+6 5.308310-1 1.000000+7 5.142690-1 1.050000+7 4.981810-19936 3 16 7 1.100000+7 4.928040-1 1.150000+7 4.841800-1 1.200000+7 4.699050-19936 3 16 8 1.250000+7 4.444730-1 1.300000+7 4.064790-1 1.350000+7 3.676620-19936 3 16 9 1.400000+7 3.265300-1 1.450000+7 2.889180-1 1.500000+7 2.615090-19936 3 16 10 1.550000+7 2.363770-1 1.600000+7 2.178830-1 1.650000+7 2.035290-19936 3 16 11 1.700000+7 1.921000-1 1.750000+7 1.828850-1 1.800000+7 1.749700-19936 3 16 12 1.850000+7 1.653120-1 1.900000+7 1.600450-1 1.950000+7 1.554180-19936 3 16 13 2.000000+7 1.473250-1 9936 3 16 14 9936 3 099999 1.002550+5 2.528990+2 0 0 0 09936 3 17 1 -1.169290+7-1.169290+7 0 0 1 189936 3 17 2 18 2 9936 3 17 3 1.173920+7 0.000000+0 1.200000+7 2.625640-6 1.250000+7 1.154500-39936 3 17 4 1.300000+7 1.095020-2 1.350000+7 2.919700-2 1.400000+7 5.516660-29936 3 17 5 1.450000+7 8.496450-2 1.500000+7 1.140790-1 1.550000+7 1.385600-19936 3 17 6 1.600000+7 1.493200-1 1.650000+7 1.566040-1 1.700000+7 1.608620-19936 3 17 7 1.750000+7 1.632260-1 1.800000+7 1.640580-1 1.850000+7 1.638820-19936 3 17 8 1.900000+7 1.622970-1 1.950000+7 1.555430-1 2.000000+7 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