JENDL-5 (Neutron sublibrary (activation cs)) Es-254 0 0 0 0 9.925400+4 2.519050+2 0 0 6 19914 1451 1 0.000000+0 1.000000+0 0 0 0 69914 1451 2 1.000000+0 2.000000+7 0 0 10 59914 1451 3 2.936000+2 0.000000+0 4 0 350 149914 1451 4 99-Es-254 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al. 9914 1451 5 DIST-DEC21 20100304 9914 1451 6 ----JENDL-5 MATERIAL 9914 9914 1451 7 -----INCIDENT NEUTRON DATA 9914 1451 8 ------ENDF-6 FORMAT 9914 1451 9 9914 1451 10 History 9914 1451 11 07-09 Theoretical calculation was performed with CCONE code. 9914 1451 12 07-10 Theoretical calculation was performed with CCONE code. 9914 1451 13 07-11 Data were compiled as JENDL/AC-2008/1/ 9914 1451 14 09-08 (MF1,MT458) was evaluated. 9914 1451 15 10-01 Data of prompt gamma rays due to fission were given. 9914 1451 16 10-03 Covariance data were given. 9914 1451 17 9914 1451 18 21-11 revised by O.Iwamoto 9914 1451 19 (MF3/MT19-21,38) deleted 9914 1451 20 (MF8/MT16-18,37,102) JENDL/AD-2017 adopted 9914 1451 21 (MF8/MT4) added 9914 1451 22 9914 1451 23 MF=1 General information 9914 1451 24 MT=452 Number of Neutrons per fission 9914 1451 25 Sum of MT's=455 and 456. 9914 1451 26 9914 1451 27 MT=455 Delayed neutron data 9914 1451 28 (Same as JENDL-3.3) 9914 1451 29 Estimated from systematics by Tuttle/2/, Benedetti et al./3/ 9914 1451 30 and Waldo et al./4/ 9914 1451 31 Decay constants were taken from Ref./5/ 9914 1451 32 9914 1451 33 MT=456 Number of prompt neutrons per fission 9914 1451 34 (Same as JENDL-3.3) 9914 1451 35 Estimated from Howerton's sytematics/6/. 9914 1451 36 9914 1451 37 MT=458 Components of energy release due to fission 9914 1451 38 Total energy and prompt energy were calculated from mass 9914 1451 39 balance using JENDL-4 fission yields data and mass excess 9914 1451 40 evaluation. Mass excess values were from Audi's 2009 9914 1451 41 evaluation/7/. Delayed energy values were calculated from 9914 1451 42 the energy release for infinite irradiation using JENDL FP 9914 1451 43 Decay Data File 2000 and JENDL-4 yields data. For delayed 9914 1451 44 neutron energy, as the JENDL FP Decay Data File 2000/8/ does 9914 1451 45 not include average neutron energy values, the average values 9914 1451 46 were calculated using the formula shown in the report by 9914 1451 47 T.R. England/9/. The fractions of prompt energy were 9914 1451 48 calculated using the fractions of Sher's evaluation/10/ when 9914 1451 49 they were provided. When the fractions were not given by Sher,9914 1451 50 averaged fractions were used. 9914 1451 51 9914 1451 52 9914 1451 53 MF= 2 Resonance parameters 9914 1451 54 MT=151 9914 1451 55 No resonance parameters are given. 9914 1451 56 9914 1451 57 9914 1451 58 Thermal cross sections and resonance integrals (at 300K) 9914 1451 59 ------------------------------------------------------- 9914 1451 60 0.0253 eV reson. integ.(*) 9914 1451 61 (barns) (barns) 9914 1451 62 ------------------------------------------------------- 9914 1451 63 total 2168.1 9914 1451 64 elastic 10.22 9914 1451 65 fission 2129.2 1120 9914 1451 66 capture 28.31 683 9914 1451 67 ------------------------------------------------------- 9914 1451 68 (*) In the energy range from 0.5 eV to 10 MeV. 9914 1451 69 9914 1451 70 9914 1451 71 MF= 3 Neutron cross sections 9914 1451 72 Below 0.5 eV: 9914 1451 73 * Elastic scattering cross section is 10.2 b calculated from 9914 1451 74 scattering radius of 9.001 fm/11/. 9914 1451 75 * Fission cross section is in the 1/v shape, and 2128 b at 9914 1451 76 0.0253 eV was estimated from experimental data/12,13,14,15/.9914 1451 77 * Capture cross section is in the 1/v shape, and 28.3 b at 9914 1451 78 0.0253 eV/14/ 9914 1451 79 9914 1451 80 Above 0.5 eV: 9914 1451 81 Cross sections were calculated with CCONE code/11/. 9914 1451 82 9914 1451 83 MT= 1 Total cross section 9914 1451 84 The cross section was calculated with CC OMP of Soukhovitskii 9914 1451 85 et al./16/ 9914 1451 86 9914 1451 87 MT=18 Fission cross section 9914 1451 88 The experimental data of Danon et al./15,17/ were used to 9914 1451 89 determine the parameters in the CCONE calculation. 9914 1451 90 9914 1451 91 9914 1451 92 MF= 4 Angular distributions of secondary neutrons 9914 1451 93 MT=2 Elastic scattering 9914 1451 94 Calculated with CCONE code/11/. 9914 1451 95 9914 1451 96 MT=18 Fission 9914 1451 97 Isotropic distributions in the laboratory system were assumed.9914 1451 98 9914 1451 99 9914 1451 100 MF= 5 Energy distributions of secondary neutrons 9914 1451 101 MT=18 Prompt netrons 9914 1451 102 Calculated with CCONE code/11/. 9914 1451 103 9914 1451 104 MT=455 Delayed neutrons 9914 1451 105 Calculated by Brady and England/5/. 9914 1451 106 9914 1451 107 9914 1451 108 MF= 6 Energy-angle distributions 9914 1451 109 Calculated with CCONE code/11/. 9914 1451 110 Distributions from fission (MT=18) are not included. 9914 1451 111 9914 1451 112 9914 1451 113 MF=12 Photon production multiplicities 9914 1451 114 MT=18 Fission 9914 1451 115 Calculated from the total energy released by the prompt 9914 1451 116 gamma-rays due to fission given in MF=1/MT=458 and the 9914 1451 117 average energy of gamma-rays. 9914 1451 118 9914 1451 119 9914 1451 120 MF=14 Photon angular distributions 9914 1451 121 MT=18 Fission 9914 1451 122 Isotoropic distributions were assumed. 9914 1451 123 9914 1451 124 9914 1451 125 MF=15 Continuous photon energy spectra 9914 1451 126 MT=18 Fission 9914 1451 127 Experimental data measured by Verbinski et al./18/ for 9914 1451 128 Pu-239 thermal fission were adopted. 9914 1451 129 9914 1451 130 9914 1451 131 MF=31 Covariances of average number of neutrons per fission 9914 1451 132 MT=452 Number of neutrons per fission 9914 1451 133 Sum of covariances for MT=455 and MT=456. 9914 1451 134 9914 1451 135 MT=455 9914 1451 136 Error of 15% was assumed. 9914 1451 137 9914 1451 138 MT=456 9914 1451 139 Covariance was obtained by fitting a linear function to the 9914 1451 140 at 0.0 and 5.0 MeV with an uncertainty of 10%. 9914 1451 141 9914 1451 142 9914 1451 143 MF=33 Covariances of neutron cross sections 9914 1451 144 Covariances were given to all the cross sections by using 9914 1451 145 KALMAN code/19/ and the covariances of model parameters 9914 1451 146 used in the cross-section calculations. 9914 1451 147 9914 1451 148 Covariances of the fission cross section were determined from 9914 1451 149 experimental data. 9914 1451 150 9914 1451 151 For the following cross sections, standard deviations in the 9914 1451 152 energy region below 0.5 eV were assumed as follows: 9914 1451 153 9914 1451 154 Total 6.7 % 9914 1451 155 Elastic scattering 50 % 9914 1451 156 Fission 6.8 % estimated from experimental data 9914 1451 157 Capture 11 % estimated from experimental data 9914 1451 158 9914 1451 159 9914 1451 160 MF=34 Covariances for Angular Distributions 9914 1451 161 MT=2 Elastic scattering 9914 1451 162 Covariances were given only to P1 components. 9914 1451 163 9914 1451 164 9914 1451 165 MF=35 Covariances for Energy Distributions 9914 1451 166 MT=18 Fission spectra 9914 1451 167 Estimated with CCONE and KALMAN codes. 9914 1451 168 9914 1451 169 9914 1451 170 ***************************************************************** 9914 1451 171 Calculation with CCONE code 9914 1451 172 ***************************************************************** 9914 1451 173 9914 1451 174 Models and parameters used in the CCONE/11/ calculation 9914 1451 175 1) Coupled channel optical model 9914 1451 176 Levels in the rotational band were included. Optical model 9914 1451 177 potential and coupled levels are shown in Table 1. 9914 1451 178 9914 1451 179 2) Two-component exciton model/20/ 9914 1451 180 * Global parametrization of Koning-Duijvestijn/21/ 9914 1451 181 was used. 9914 1451 182 * Gamma emission channel/22/ was added to simulate direct 9914 1451 183 and semi-direct capture reaction. 9914 1451 184 9914 1451 185 3) Hauser-Feshbach statistical model 9914 1451 186 * Moldauer width fluctuation correction/23/ was included. 9914 1451 187 * Neutron, gamma and fission decay channel were included. 9914 1451 188 * Transmission coefficients of neutrons were taken from 9914 1451 189 coupled channel calculation in Table 1. 9914 1451 190 * The level scheme of the target is shown in Table 2. 9914 1451 191 * Level density formula of constant temperature and Fermi-gas 9914 1451 192 model were used with shell energy correction and collective 9914 1451 193 enhancement factor. Parameters are shown in Table 3. 9914 1451 194 * Fission channel: 9914 1451 195 Double humped fission barriers were assumed. 9914 1451 196 Fission barrier penetrabilities were calculated with 9914 1451 197 Hill-Wheler formula/24/. Fission barrier parameters were 9914 1451 198 shown in Table 4. Transition state model was used and 9914 1451 199 continuum levels are assumed above the saddles. The level 9914 1451 200 density parameters for inner and outer saddles are shown in 9914 1451 201 Tables 5 and 6, respectively. 9914 1451 202 * Gamma-ray strength function of Kopecky et al/25/,/26/ 9914 1451 203 was used. The prameters are shown in Table 7. 9914 1451 204 9914 1451 205 9914 1451 206 ------------------------------------------------------------------9914 1451 207 Tables 9914 1451 208 ------------------------------------------------------------------9914 1451 209 9914 1451 210 Table 1. Coupled channel calculation 9914 1451 211 -------------------------------------------------- 9914 1451 212 * rigid rotor model was applied 9914 1451 213 * coupled levels = 0,1,3,7 (see Table 2) 9914 1451 214 * optical potential parameters /16/ 9914 1451 215 Volume: 9914 1451 216 V_0 = 49.97 MeV 9914 1451 217 lambda_HF = 0.01004 1/MeV 9914 1451 218 C_viso = 15.9 MeV 9914 1451 219 A_v = 12.04 MeV 9914 1451 220 B_v = 81.36 MeV 9914 1451 221 E_a = 385 MeV 9914 1451 222 r_v = 1.2568 fm 9914 1451 223 a_v = 0.633 fm 9914 1451 224 Surface: 9914 1451 225 W_0 = 17.2 MeV 9914 1451 226 B_s = 11.19 MeV 9914 1451 227 C_s = 0.01361 1/MeV 9914 1451 228 C_wiso = 23.5 MeV 9914 1451 229 r_s = 1.1803 fm 9914 1451 230 a_s = 0.601 fm 9914 1451 231 Spin-orbit: 9914 1451 232 V_so = 5.75 MeV 9914 1451 233 lambda_so = 0.005 1/MeV 9914 1451 234 W_so = -3.1 MeV 9914 1451 235 B_so = 160 MeV 9914 1451 236 r_so = 1.1214 fm 9914 1451 237 a_so = 0.59 fm 9914 1451 238 Coulomb: 9914 1451 239 C_coul = 1.3 9914 1451 240 r_c = 1.2452 fm 9914 1451 241 a_c = 0.545 fm 9914 1451 242 Deformation: 9914 1451 243 beta_2 = 0.213 9914 1451 244 beta_4 = 0.066 9914 1451 245 beta_6 = 0.0015 9914 1451 246 9914 1451 247 * Calculated strength function 9914 1451 248 S0= 1.54e-4 S1= 3.21e-4 R'= 9.00 fm (En=1 keV) 9914 1451 249 -------------------------------------------------- 9914 1451 250 9914 1451 251 Table 2. Level Scheme of Es-254 9914 1451 252 ------------------- 9914 1451 253 No. Ex(MeV) J PI 9914 1451 254 ------------------- 9914 1451 255 0 0.00000 7 + * 9914 1451 256 1 0.08010 8 + * 9914 1451 257 2 0.08420 2 + 9914 1451 258 3 0.11420 3 + * 9914 1451 259 4 0.15420 4 + 9914 1451 260 5 0.17110 9 + 9914 1451 261 6 0.20420 5 + 9914 1451 262 7 0.21470 5 - * 9914 1451 263 ------------------- 9914 1451 264 *) Coupled levels in CC calculation 9914 1451 265 9914 1451 266 Table 3. Level density parameters 9914 1451 267 -------------------------------------------------------- 9914 1451 268 Nuclide a* Pair Eshell T E0 Ematch 9914 1451 269 1/MeV MeV MeV MeV MeV MeV 9914 1451 270 -------------------------------------------------------- 9914 1451 271 Es-255 19.4922 0.7515 2.2159 0.3615 -0.7550 2.8069 9914 1451 272 Es-254 19.4263 0.0000 1.6001 0.2773 -0.6899 1.0000 9914 1451 273 Es-253 19.3604 0.7544 1.3654 0.3757 -0.7841 2.8632 9914 1451 274 Es-252 19.2945 0.0000 0.8872 0.2854 -0.6910 1.0000 9914 1451 275 Es-251 19.2285 0.7574 0.6125 0.4218 -1.1719 3.3703 9914 1451 276 -------------------------------------------------------- 9914 1451 277 9914 1451 278 Table 4. Fission barrier parameters 9914 1451 279 ---------------------------------------- 9914 1451 280 Nuclide V_A hw_A V_B hw_B 9914 1451 281 MeV MeV MeV MeV 9914 1451 282 ---------------------------------------- 9914 1451 283 Es-255 6.120 0.800 4.800 0.520 9914 1451 284 Es-254 6.200 0.650 5.800 0.450 9914 1451 285 Es-253 6.200 0.800 5.400 0.520 9914 1451 286 Es-252 6.200 0.650 5.800 0.450 9914 1451 287 Es-251 6.200 0.800 5.400 0.520 9914 1451 288 ---------------------------------------- 9914 1451 289 9914 1451 290 Table 5. Level density above inner saddle 9914 1451 291 -------------------------------------------------------- 9914 1451 292 Nuclide a* Pair Eshell T E0 Ematch 9914 1451 293 1/MeV MeV MeV MeV MeV MeV 9914 1451 294 -------------------------------------------------------- 9914 1451 295 Es-255 21.8312 0.8767 2.4000 0.3193 -1.5382 2.8767 9914 1451 296 Es-254 21.7575 0.0000 2.4000 0.3199 -2.4149 2.0000 9914 1451 297 Es-253 21.6837 0.8802 2.4000 0.3205 -1.5348 2.8802 9914 1451 298 Es-252 21.6098 0.0000 2.4000 0.3211 -2.4150 2.0000 9914 1451 299 Es-251 21.5360 0.8837 2.4000 0.3217 -1.5314 2.8837 9914 1451 300 -------------------------------------------------------- 9914 1451 301 9914 1451 302 Table 6. Level density above outer saddle 9914 1451 303 -------------------------------------------------------- 9914 1451 304 Nuclide a* Pair Eshell T E0 Ematch 9914 1451 305 1/MeV MeV MeV MeV MeV MeV 9914 1451 306 -------------------------------------------------------- 9914 1451 307 Es-255 21.8312 0.8767 1.1200 0.3449 -0.8359 2.8767 9914 1451 308 Es-254 21.7575 0.0000 1.0800 0.3459 -1.7120 2.0000 9914 1451 309 Es-253 21.6837 0.8802 1.0400 0.3470 -0.8313 2.8802 9914 1451 310 Es-252 21.6098 0.0000 1.0000 0.3481 -1.7109 2.0000 9914 1451 311 Es-251 21.5360 0.8837 0.9600 0.3492 -0.8267 2.8837 9914 1451 312 -------------------------------------------------------- 9914 1451 313 9914 1451 314 Table 7. Gamma-ray strength function for Es-255 9914 1451 315 -------------------------------------------------------- 9914 1451 316 K0 = 1.500 E0 = 4.500 (MeV) 9914 1451 317 * E1: ER = 11.38 (MeV) EG = 2.71 (MeV) SIG = 261.63 (mb) 9914 1451 318 ER = 14.16 (MeV) EG = 4.11 (MeV) SIG = 523.27 (mb) 9914 1451 319 * M1: ER = 6.47 (MeV) EG = 4.00 (MeV) SIG = 1.87 (mb) 9914 1451 320 * E2: ER = 9.93 (MeV) EG = 3.05 (MeV) SIG = 7.50 (mb) 9914 1451 321 -------------------------------------------------------- 9914 1451 322 9914 1451 323 9914 1451 324 References 9914 1451 325 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 9914 1451 326 2) R.J.Tuttld: INDC(NDS)-107/G+Special, p.29 (1979). 9914 1451 327 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 9914 1451 328 4) R.Waldo et al.: Phys. Rev., C23, 1113 (1981). 9914 1451 329 5) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989). 9914 1451 330 6) R.J.Howerton: Nucl. Sci. Eng., 62, 438 (1977). 9914 1451 331 7) G.Audi: Private communication (April 2009). 9914 1451 332 8) J.Katakura et al.: JAERI 1343 (2001). 9914 1451 333 9) T.R.England et al.: LA-11151-MS (1988). 9914 1451 334 10) R.Sher, C.Beck: EPRI NP-1771 (1981). 9914 1451 335 11) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 9914 1451 336 12) H.Diamond et al.: J. Inorg. Nucl. Chem., 30, 2553 (1968). 9914 1451 337 13) K.W.MacMurdo, R.M.Harbour : J. Inorg. Nucl. Chem., 34, 9914 1451 338 449 (1972). 9914 1451 339 14) J.Halperin et al.: Nucl. Sci. Eng., 90, 298 (1985). 9914 1451 340 15) Y.Danon et al.: Nucl. Sci. Eng., 109, 341 (1991). 9914 1451 341 16) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005). 9914 1451 342 17) Y.Danon et al.: 1994 Gatlinburg, Vol.1, p.245 (1994). 9914 1451 343 18) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973). 9914 1451 344 19) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in 9914 1451 345 Japanese. 9914 1451 346 20) C.Kalbach: Phys. Rev. C33, 818 (1986). 9914 1451 347 21) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 9914 1451 348 22) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 9914 1451 349 23) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 9914 1451 350 24) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 9914 1451 351 25) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 9914 1451 352 26) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). 9914 1451 353 9914 1451 354 1 451 368 19914 1451 355 2 151 4 19914 1451 356 3 4 27 19914 1451 357 3 16 14 19914 1451 358 3 17 9 19914 1451 359 3 18 118 19914 1451 360 3 37 6 19914 1451 361 3 102 173 19914 1451 362 8 4 2 19914 1451 363 8 16 2 19914 1451 364 8 17 2 19914 1451 365 8 18 2 19914 1451 366 8 37 2 19914 1451 367 8 102 2 19914 1451 368 9914 1 099999 9914 0 0 0 9.925400+4 2.519050+2 0 0 1 09914 2151 1 9.925400+4 1.000000+0 0 0 1 09914 2151 2 1.000000-5 5.000000-1 0 0 0 09914 2151 3 7.000000+0 9.000000-1 0 0 0 09914 2151 4 9914 2 099999 9914 0 0 0 9.925400+4 2.519050+2 0 0 0 09914 3 4 1 0.000000+0-8.010000+4 0 0 1 729914 3 4 2 72 2 9914 3 4 3 8.041800+4 0.000000+0 8.453420+4 2.838010-2 8.453430+4 2.838038-29914 3 4 4 1.000000+5 7.142337-2 1.081800+5 8.967715-2 1.146530+5 1.041247-19914 3 4 5 1.200000+5 1.150159-1 1.400000+5 1.519149-1 1.548120+5 1.753423-19914 3 4 6 1.700000+5 1.976187-1 1.717790+5 2.001176-1 2.000000+5 2.533050-19914 3 4 7 2.050110+5 2.604309-1 2.155520+5 2.781084-1 2.500000+5 3.423071-19914 3 4 8 3.000000+5 4.018360-1 4.000000+5 5.477124-1 5.000000+5 7.941188-19914 3 4 9 6.000000+5 1.005160+0 7.000000+5 1.197180+0 8.000000+5 1.352475+09914 3 4 10 9.000000+5 1.494085+0 1.000000+6 1.633605+0 1.100000+6 1.755716+09914 3 4 11 1.200000+6 1.867783+0 1.300000+6 1.965893+0 1.400000+6 2.055086+09914 3 4 12 1.500000+6 2.123203+0 1.600000+6 2.181245+0 1.700000+6 2.216949+09914 3 4 13 1.800000+6 2.243334+0 1.900000+6 2.251970+0 2.000000+6 2.252973+09914 3 4 14 2.100000+6 2.242867+0 2.300000+6 2.205852+0 2.500000+6 2.154598+09914 3 4 15 2.700000+6 2.096537+0 3.000000+6 2.003674+0 3.500000+6 1.887928+09914 3 4 16 4.000000+6 1.789051+0 4.500000+6 1.743832+0 5.000000+6 1.716010+09914 3 4 17 5.500000+6 1.658272+0 6.000000+6 1.352713+0 6.500000+6 1.063918+09914 3 4 18 7.000000+6 8.148541-1 7.500000+6 6.413846-1 8.000000+6 5.346980-19914 3 4 19 8.500000+6 4.649435-1 9.000000+6 4.203612-1 9.500000+6 3.883168-19914 3 4 20 1.000000+7 3.647671-1 1.050000+7 3.443080-1 1.100000+7 3.290558-19914 3 4 21 1.150000+7 3.149735-1 1.200000+7 3.012604-1 1.250000+7 2.884117-19914 3 4 22 1.300000+7 2.751742-1 1.350000+7 2.657394-1 1.400000+7 2.559115-19914 3 4 23 1.450000+7 2.443536-1 1.500000+7 2.354112-1 1.550000+7 2.272096-19914 3 4 24 1.600000+7 2.190506-1 1.650000+7 2.114972-1 1.700000+7 2.039685-19914 3 4 25 1.750000+7 1.970324-1 1.800000+7 1.901426-1 1.850000+7 1.837624-19914 3 4 26 1.900000+7 1.797795-1 1.950000+7 1.737562-1 2.000000+7 1.679375-19914 3 4 27 9914 3 099999 9.925400+4 2.519050+2 0 0 0 09914 3 16 1 -5.093030+6-5.093030+6 0 0 1 319914 3 16 2 31 2 9914 3 16 3 5.113250+6 0.000000+0 5.500000+6 5.348450-2 6.000000+6 3.540770-19914 3 16 4 6.500000+6 5.499520-1 7.000000+6 6.787230-1 7.500000+6 7.350070-19914 3 16 5 8.000000+6 7.562020-1 8.500000+6 7.520660-1 9.000000+6 7.463010-19914 3 16 6 9.500000+6 7.383540-1 1.000000+7 7.358440-1 1.050000+7 7.354440-19914 3 16 7 1.100000+7 7.333180-1 1.150000+7 7.215840-1 1.200000+7 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