JENDL-5 (Neutron sublibrary (activation cs)) Th-227 0 0 0 0 9.022700+4 2.250770+2 0 0 6 19025 1451 1 0.000000+0 1.000000+0 0 0 0 69025 1451 2 1.000000+0 2.000000+7 0 0 10 59025 1451 3 2.936000+2 0.000000+0 4 0 349 149025 1451 4 90-Th-227 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al. 9025 1451 5 DIST-DEC21 20100303 9025 1451 6 ----JENDL-5 MATERIAL 9025 9025 1451 7 -----INCIDENT NEUTRON DATA 9025 1451 8 ------ENDF-6 FORMAT 9025 1451 9 9025 1451 10 History 9025 1451 11 07-06 Theoretical calculation was performed with CCONE code. 9025 1451 12 07-11 Re-calculation with CCONE code was done. 9025 1451 13 Data were compiled as JENDL/AC-2008/1/. 9025 1451 14 09-08 (MF1,MT458) was evaluated. 9025 1451 15 10-01 Data of prompt gamma rays due to fission were given. 9025 1451 16 10-03 Covariance data were given. 9025 1451 17 9025 1451 18 21-11 revised by O.Iwamoto 9025 1451 19 (MF3/MT19-21,38) deleted 9025 1451 20 (MF8/MT4,16-18,37,102) added 9025 1451 21 9025 1451 22 MF=1 General information 9025 1451 23 MT=452 Number of Neutrons per fission 9025 1451 24 Sum of MT's=455 and 456. 9025 1451 25 9025 1451 26 MT=455 Delayed neutron data 9025 1451 27 (Same as JENDL-3.3) 9025 1451 28 Average values of systematics of Tuttle/2/, Benedetti et 9025 1451 29 al./3/ and Waldo et al./4/ were adopted. 9025 1451 30 Decay constants were calculated by Brady and England/5/. 9025 1451 31 9025 1451 32 MT=456 Number of prompt neutrons per fission 9025 1451 33 (Same as JENDL-3.3) 9025 1451 34 Based on the systematics recommended by Manero and Konshin/6/ 9025 1451 35 and by Howerton/7/. 9025 1451 36 9025 1451 37 MT=458 Components of energy release due to fission 9025 1451 38 Total energy and prompt energy were calculated from mass 9025 1451 39 balance using JENDL-4 fission yields data and mass excess 9025 1451 40 evaluation. Mass excess values were from Audi's 2009 9025 1451 41 evaluation/8/. Delayed energy values were calculated from 9025 1451 42 the energy release for infinite irradiation using JENDL FP 9025 1451 43 Decay Data File 2000 and JENDL-4 yields data. For delayed 9025 1451 44 neutron energy, as the JENDL FP Decay Data File 2000/9/ does 9025 1451 45 not include average neutron energy values, the average values 9025 1451 46 were calculated using the formula shown in the report by 9025 1451 47 T.R. England/10/. The fractions of prompt energy were 9025 1451 48 calculated using the fractions of Sher's evaluation/11/ when 9025 1451 49 they were provided. When the fractions were not given by Sher,9025 1451 50 averaged fractions were used. 9025 1451 51 9025 1451 52 9025 1451 53 MF= 2 Resonance parameters 9025 1451 54 MT=151 9025 1451 55 No resonance parameters are given. 9025 1451 56 9025 1451 57 9025 1451 58 Thermal cross sections and resonance integrals (at 300K) 9025 1451 59 ------------------------------------------------------- 9025 1451 60 0.0253 eV reson. integ.(*) 9025 1451 61 (barns) (barns) 9025 1451 62 ------------------------------------------------------- 9025 1451 63 total 620.06 9025 1451 64 elastic 12.53 9025 1451 65 fission 202.08 330 9025 1451 66 capture 405.16 661 9025 1451 67 ------------------------------------------------------- 9025 1451 68 (*) In the energy range from 0.5 eV to 10 MeV. 9025 1451 69 9025 1451 70 9025 1451 71 MF= 3 Neutron cross sections 9025 1451 72 Below 0.65 eV: 9025 1451 73 Elastic scattering cross section is 12.5 b calculated from 9025 1451 74 scattering radius/12/. 9025 1451 75 Fission cross section is in the 1/v shape, and 202 b at 9025 1451 76 0.0253 eV/13/. 9025 1451 77 Capture cross section is also in the 1/v shape, and 405 9025 1451 78 b at 0.0253 eV. A capture to fission ratio calculated with 9025 1451 79 CCONE code/12/ at 1 eV was used. 9025 1451 80 9025 1451 81 Above 0.65 eV: 9025 1451 82 All the Cross sections were calculated with CCONE code/12/. 9025 1451 83 9025 1451 84 MT= 1 Total cross section 9025 1451 85 The cross section was calculated with CC OMP of Soukhovitskii 9025 1451 86 et al./14/ 9025 1451 87 9025 1451 88 9025 1451 89 MF= 4 Angular distributions of secondary neutrons 9025 1451 90 MT=2 Elastic scattering 9025 1451 91 Calculated with CCONE code/12/. 9025 1451 92 9025 1451 93 MT=18 Fission 9025 1451 94 Isotropic distributions in the laboratory system were assumed.9025 1451 95 9025 1451 96 9025 1451 97 MF= 5 Energy distributions of secondary neutrons 9025 1451 98 MT=18 Prompt neutrons 9025 1451 99 Calculated with CCONE code/12/. 9025 1451 100 9025 1451 101 MT=455 Delayed neutrons 9025 1451 102 Calculated by Brady and England/5/. 9025 1451 103 9025 1451 104 9025 1451 105 MF= 6 Energy-angle distributions 9025 1451 106 Calculated with CCONE code/12/. 9025 1451 107 Distributions from fission (MT=18) are not included. 9025 1451 108 9025 1451 109 9025 1451 110 MF=12 Photon production multiplicities 9025 1451 111 MT=18 Fission 9025 1451 112 Calculated from the total energy released by the prompt 9025 1451 113 gamma-rays due to fission given in MF=1/MT=458 and the 9025 1451 114 average energy of gamma-rays. 9025 1451 115 9025 1451 116 9025 1451 117 MF=14 Photon angular distributions 9025 1451 118 MT=18 Fission 9025 1451 119 Isotoropic distributions were assumed. 9025 1451 120 9025 1451 121 9025 1451 122 MF=15 Continuous photon energy spectra 9025 1451 123 MT=18 Fission 9025 1451 124 Experimental data measured by Verbinski et al./15/ for 9025 1451 125 U-235 thermal fission were adopted. 9025 1451 126 9025 1451 127 9025 1451 128 MF=31 Covariances of average number of neutrons per fission 9025 1451 129 MT=452 Number of neutrons per fission 9025 1451 130 Sum of covariances for MT=455 and MT=456. 9025 1451 131 9025 1451 132 MT=455 9025 1451 133 Error of 15% was assumed. 9025 1451 134 9025 1451 135 MT=456 9025 1451 136 Covariance was obtained by fitting a linear function to the 9025 1451 137 at 0.0 and 5.0 MeV with an uncertainty of 5%. 9025 1451 138 9025 1451 139 9025 1451 140 MF=33 Covariances of neutron cross sections 9025 1451 141 Covariances were given to all the cross sections by using 9025 1451 142 KALMAN code/16/ and the covariances of model parameters 9025 1451 143 used in the cross-section calculations. 9025 1451 144 9025 1451 145 For the following cross sections, standard deviations in the 9025 1451 146 energy region below 0.65 eV were assumed as follows: 9025 1451 147 9025 1451 148 Total 33 % 9025 1451 149 Elastic scattering 50 % 9025 1451 150 Fission 7.7 % estimated from experimental data 9025 1451 151 Capture 50 % 9025 1451 152 9025 1451 153 9025 1451 154 MF=34 Covariances for Angular Distributions 9025 1451 155 MT=2 Elastic scattering 9025 1451 156 Covariances were given only to P1 components. 9025 1451 157 9025 1451 158 9025 1451 159 MF=35 Covariances for Energy Distributions 9025 1451 160 MT=18 Fission spectra 9025 1451 161 Estimated with CCONE and KALMAN codes. 9025 1451 162 9025 1451 163 9025 1451 164 9025 1451 165 ***************************************************************** 9025 1451 166 Calculation with CCONE code 9025 1451 167 ***************************************************************** 9025 1451 168 9025 1451 169 Models and parameters used in the CCONE/12/ calculation 9025 1451 170 1) Coupled channel optical model 9025 1451 171 Levels in the rotational band were included. Optical model 9025 1451 172 potential and coupled levels are shown in Table 1. 9025 1451 173 9025 1451 174 2) Two-component exciton model/17/ 9025 1451 175 * Global parametrization of Koning-Duijvestijn/18/ 9025 1451 176 was used. 9025 1451 177 * Gamma emission channel/19/ was added to simulate direct 9025 1451 178 and semi-direct capture reaction. 9025 1451 179 9025 1451 180 3) Hauser-Feshbach statistical model 9025 1451 181 * Moldauer width fluctuation correction/20/ was included. 9025 1451 182 * Neutron, gamma and fission decay channel were included. 9025 1451 183 * Transmission coefficients of neutrons were taken from 9025 1451 184 coupled channel calculation in Table 1. 9025 1451 185 * The level scheme of the target is shown in Table 2. 9025 1451 186 * Level density formula of constant temperature and Fermi-gas 9025 1451 187 model were used with shell energy correction and collective 9025 1451 188 enhancement factor. Parameters are shown in Table 3. 9025 1451 189 * Fission channel: 9025 1451 190 Double humped fission barriers were assumed. 9025 1451 191 Fission barrier penetrabilities were calculated with 9025 1451 192 Hill-Wheler formula/21/. Fission barrier parameters were 9025 1451 193 shown in Table 4. Transition state model was used and 9025 1451 194 continuum levels are assumed above the saddles. The level 9025 1451 195 density parameters for inner and outer saddles are shown in 9025 1451 196 Tables 5 and 6, respectively. 9025 1451 197 * Gamma-ray strength function of Kopecky et al/22/,/23/ 9025 1451 198 was used. The prameters are shown in Table 7. 9025 1451 199 9025 1451 200 9025 1451 201 ------------------------------------------------------------------9025 1451 202 Tables 9025 1451 203 ------------------------------------------------------------------9025 1451 204 9025 1451 205 Table 1. Coupled channel calculation 9025 1451 206 -------------------------------------------------- 9025 1451 207 * rigid rotor model was applied 9025 1451 208 * coupled levels = 0,2,1 (see Table 2) 9025 1451 209 * optical potential parameters /14/ 9025 1451 210 Volume: 9025 1451 211 V_0 = 49.97 MeV 9025 1451 212 lambda_HF = 0.01004 1/MeV 9025 1451 213 C_viso = 15.9 MeV 9025 1451 214 A_v = 12.04 MeV 9025 1451 215 B_v = 81.36 MeV 9025 1451 216 E_a = 385 MeV 9025 1451 217 r_v = 1.2568 fm 9025 1451 218 a_v = 0.633 fm 9025 1451 219 Surface: 9025 1451 220 W_0 = 17.2 MeV 9025 1451 221 B_s = 11.19 MeV 9025 1451 222 C_s = 0.01361 1/MeV 9025 1451 223 C_wiso = 23.5 MeV 9025 1451 224 r_s = 1.1803 fm 9025 1451 225 a_s = 0.601 fm 9025 1451 226 Spin-orbit: 9025 1451 227 V_so = 5.75 MeV 9025 1451 228 lambda_so = 0.005 1/MeV 9025 1451 229 W_so = -3.1 MeV 9025 1451 230 B_so = 160 MeV 9025 1451 231 r_so = 1.1214 fm 9025 1451 232 a_so = 0.59 fm 9025 1451 233 Coulomb: 9025 1451 234 C_coul = 1.3 9025 1451 235 r_c = 1.2452 fm 9025 1451 236 a_c = 0.545 fm 9025 1451 237 Deformation: 9025 1451 238 beta_2 = 0.213 9025 1451 239 beta_4 = 0.066 9025 1451 240 beta_6 = 0.0015 9025 1451 241 9025 1451 242 * Calculated strength function 9025 1451 243 S0= 0.98e-4 S1= 1.27e-4 R'= 9.91 fm (En=1 keV) 9025 1451 244 -------------------------------------------------- 9025 1451 245 9025 1451 246 Table 2. Level Scheme of Th-227 9025 1451 247 ------------------- 9025 1451 248 No. Ex(MeV) J PI 9025 1451 249 ------------------- 9025 1451 250 0 0.00000 1/2 + * 9025 1451 251 1 0.00929 5/2 + * 9025 1451 252 2 0.02438 3/2 + * 9025 1451 253 3 0.03786 3/2 - 9025 1451 254 4 0.07367 3/2 - 9025 1451 255 5 0.07623 3/2 + 9025 1451 256 6 0.07762 3/2 + 9025 1451 257 7 0.08678 9/2 + 9025 1451 258 8 0.09919 1/2 + 9025 1451 259 9 0.12730 3/2 + 9025 1451 260 10 0.18371 1/2 - 9025 1451 261 11 0.20002 5/2 - 9025 1451 262 12 0.20710 3/2 - 9025 1451 263 13 0.22848 3/2 - 9025 1451 264 14 0.23146 7/2 - 9025 1451 265 ------------------- 9025 1451 266 *) Coupled levels in CC calculation 9025 1451 267 9025 1451 268 Table 3. Level density parameters 9025 1451 269 -------------------------------------------------------- 9025 1451 270 Nuclide a* Pair Eshell T E0 Ematch 9025 1451 271 1/MeV MeV MeV MeV MeV MeV 9025 1451 272 -------------------------------------------------------- 9025 1451 273 Th-228 17.7035 1.5894 3.0590 0.3964 -0.1539 3.9563 9025 1451 274 Th-227 17.6369 0.7965 3.1200 0.4219 -1.2457 3.5201 9025 1451 275 Th-226 17.5702 1.5965 2.8637 0.4068 -0.2170 4.0544 9025 1451 276 Th-225 17.5034 0.8000 2.7304 0.3997 -0.9049 3.1303 9025 1451 277 Th-224 17.4367 1.6036 2.4990 0.4108 -0.1826 4.0385 9025 1451 278 -------------------------------------------------------- 9025 1451 279 9025 1451 280 Table 4. Fission barrier parameters 9025 1451 281 ---------------------------------------- 9025 1451 282 Nuclide V_A hw_A V_B hw_B 9025 1451 283 MeV MeV MeV MeV 9025 1451 284 ---------------------------------------- 9025 1451 285 Th-228 3.900 1.040 6.400 0.600 9025 1451 286 Th-227 4.100 0.800 6.400 0.520 9025 1451 287 Th-226 3.900 1.040 8.200 0.600 9025 1451 288 Th-225 4.200 0.800 8.000 0.520 9025 1451 289 Th-224 4.100 1.040 7.600 0.600 9025 1451 290 ---------------------------------------- 9025 1451 291 9025 1451 292 Table 5. Level density above inner saddle 9025 1451 293 -------------------------------------------------------- 9025 1451 294 Nuclide a* Pair Eshell T E0 Ematch 9025 1451 295 1/MeV MeV MeV MeV MeV MeV 9025 1451 296 -------------------------------------------------------- 9025 1451 297 Th-228 19.8280 1.8543 2.6000 0.3346 -0.5570 3.8543 9025 1451 298 Th-227 19.7533 0.9292 2.6000 0.3352 -1.4822 2.9292 9025 1451 299 Th-226 19.6786 1.8625 2.6000 0.3359 -0.5488 3.8625 9025 1451 300 Th-225 19.6038 0.9333 2.6000 0.3366 -1.4780 2.9333 9025 1451 301 Th-224 19.5291 1.8708 2.6000 0.3373 -0.5405 3.8708 9025 1451 302 -------------------------------------------------------- 9025 1451 303 9025 1451 304 Table 6. Level density above outer saddle 9025 1451 305 -------------------------------------------------------- 9025 1451 306 Nuclide a* Pair Eshell T E0 Ematch 9025 1451 307 1/MeV MeV MeV MeV MeV MeV 9025 1451 308 -------------------------------------------------------- 9025 1451 309 Th-228 19.8280 1.8543 -0.3000 0.3812 0.1590 3.8543 9025 1451 310 Th-227 19.7533 0.9292 -0.3400 0.3826 -0.7655 2.9292 9025 1451 311 Th-226 19.6786 1.8625 -0.3800 0.3840 0.1685 3.8625 9025 1451 312 Th-225 19.6038 0.9333 -0.4200 0.3854 -0.7601 2.9333 9025 1451 313 Th-224 19.5291 1.8708 -0.4600 0.3868 0.1781 3.8708 9025 1451 314 -------------------------------------------------------- 9025 1451 315 9025 1451 316 Table 7. Gamma-ray strength function for Th-228 9025 1451 317 -------------------------------------------------------- 9025 1451 318 K0 = 1.503 E0 = 4.500 (MeV) 9025 1451 319 * E1: ER = 11.03 (MeV) EG = 2.71 (MeV) SIG = 302.00 (mb) 9025 1451 320 ER = 13.87 (MeV) EG = 4.77 (MeV) SIG = 449.00 (mb) 9025 1451 321 * M1: ER = 6.71 (MeV) EG = 4.00 (MeV) SIG = 2.82 (mb) 9025 1451 322 * E2: ER = 10.31 (MeV) EG = 3.37 (MeV) SIG = 6.27 (mb) 9025 1451 323 -------------------------------------------------------- 9025 1451 324 9025 1451 325 9025 1451 326 References 9025 1451 327 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 9025 1451 328 2) R.J.Tutle: INDC(NDS)-107/G+SPECIAL, P.29 (1979), 9025 1451 329 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 9025 1451 330 4) R.Waldo et al.: Phys. Rev., C23, 1113 (1981). 9025 1451 331 5) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989). 9025 1451 332 6) F.Manero, V.A.Konshin: At. Energy Rev.,10, 637 (1972). 9025 1451 333 7) R.J.Howerton: Nucl. Sci. Eng. 62, 438 (1977). 9025 1451 334 8) G.Audi: Private communication (April 2009). 9025 1451 335 9) J.Katakura et al.: JAERI 1343 (2001). 9025 1451 336 10) T.R.England et al.: LA-11151-MS (1988). 9025 1451 337 11) R.Sher, C.Beck: EPRI NP-1771 (1981). 9025 1451 338 12) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 9025 1451 339 13) S.F.Mughabghab: "Atlas of Neutron Resonances," Elsevier 9025 1451 340 (2006). 9025 1451 341 14) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005). 9025 1451 342 15) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973). 9025 1451 343 16) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in 9025 1451 344 Japanese. 9025 1451 345 17) C.Kalbach: Phys. Rev. C33, 818 (1986). 9025 1451 346 18) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 9025 1451 347 19) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 9025 1451 348 20) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 9025 1451 349 21) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 9025 1451 350 22) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 9025 1451 351 23) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). 9025 1451 352 9025 1451 353 1 451 367 19025 1451 354 2 151 4 19025 1451 355 3 4 31 19025 1451 356 3 16 13 19025 1451 357 3 17 9 19025 1451 358 3 18 133 19025 1451 359 3 37 5 19025 1451 360 3 102 152 19025 1451 361 8 4 2 19025 1451 362 8 16 2 19025 1451 363 8 17 2 19025 1451 364 8 18 2 19025 1451 365 8 37 2 19025 1451 366 8 102 2 19025 1451 367 9025 1 099999 9025 0 0 0 9.022700+4 2.250770+2 0 0 1 09025 2151 1 9.022700+4 1.000000+0 0 0 1 09025 2151 2 1.000000-5 6.500000-1 0 0 0 09025 2151 3 5.000000-1 9.960000-1 0 0 0 09025 2151 4 9025 2 099999 9025 0 0 0 9.022700+4 2.250770+2 0 0 0 09025 3 4 1 0.000000+0-9.290000+3 0 0 1 829025 3 4 2 82 2 9025 3 4 3 9.331270+3 0.000000+0 1.000000+4 1.168480-4 2.000000+4 4.271980-39025 3 4 4 2.448830+4 9.088890-3 3.000000+4 1.259763-1 3.803220+4 1.672268-19025 3 4 5 5.000000+4 2.799239-1 7.000000+4 3.816291-1 7.399730+4 3.973361-19025 3 4 6 7.656870+4 4.310535-1 7.796490+4 4.630293-1 8.716560+4 5.892416-19025 3 4 7 9.963070+4 6.845523-1 1.000000+5 6.990610-1 1.109800+5 7.635182-19025 3 4 8 1.200000+5 8.164754-1 1.278660+5 8.458092-1 1.400000+5 9.063795-19025 3 4 9 1.700000+5 1.023500+0 1.845260+5 1.081971+0 2.000000+5 1.162263+09025 3 4 10 2.009090+5 1.163420+0 2.080200+5 1.177645+0 2.294950+5 1.228394+09025 3 4 11 2.324880+5 1.239501+0 2.500000+5 1.283700+0 3.000000+5 1.415168+09025 3 4 12 4.000000+5 1.468276+0 5.000000+5 1.637023+0 6.000000+5 1.681743+09025 3 4 13 7.000000+5 1.780339+0 8.000000+5 1.827206+0 9.000000+5 1.896668+09025 3 4 14 1.000000+6 1.985364+0 1.100000+6 2.060780+0 1.200000+6 2.138829+09025 3 4 15 1.300000+6 2.218922+0 1.400000+6 2.299019+0 1.500000+6 2.379434+09025 3 4 16 1.600000+6 2.457355+0 1.700000+6 2.533488+0 1.800000+6 2.606230+09025 3 4 17 1.900000+6 2.675176+0 2.000000+6 2.740512+0 2.300000+6 2.908807+09025 3 4 18 2.500000+6 2.996449+0 2.700000+6 3.066300+0 3.000000+6 3.143206+09025 3 4 19 3.500000+6 3.169928+0 4.000000+6 3.196283+0 4.500000+6 3.162341+09025 3 4 20 5.000000+6 3.148117+0 5.500000+6 3.094472+0 6.000000+6 2.914124+09025 3 4 21 6.500000+6 2.402432+0 7.000000+6 1.942149+0 7.500000+6 1.581514+09025 3 4 22 8.000000+6 1.315506+0 8.500000+6 1.113720+0 9.000000+6 9.621713-19025 3 4 23 9.500000+6 8.453024-1 1.000000+7 7.540315-1 1.050000+7 6.830070-19025 3 4 24 1.100000+7 6.251917-1 1.150000+7 5.812041-1 1.200000+7 5.440872-19025 3 4 25 1.250000+7 5.119200-1 1.300000+7 4.858609-1 1.350000+7 4.648747-19025 3 4 26 1.400000+7 4.447701-1 1.450000+7 4.263986-1 1.500000+7 4.110957-19025 3 4 27 1.550000+7 3.978079-1 1.600000+7 3.827536-1 1.650000+7 3.702634-19025 3 4 28 1.700000+7 3.584706-1 1.750000+7 3.476726-1 1.800000+7 3.370138-19025 3 4 29 1.850000+7 3.269563-1 1.900000+7 3.171102-1 1.950000+7 3.077158-19025 3 4 30 2.000000+7 2.991617-1 9025 3 4 31 9025 3 099999 9.022700+4 2.250770+2 0 0 0 09025 3 16 1 -5.462200+6-5.462200+6 0 0 1 309025 3 16 2 30 2 9025 3 16 3 5.486470+6 0.000000+0 6.000000+6 1.511460-1 6.500000+6 6.133530-19025 3 16 4 7.000000+6 1.009530+0 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