JENDL-5 (Neutron sublibrary (activation cs)) Bk-246 0 0 0 0 9.724600+4 2.439550+2 0 0 6 19743 1451 1 0.000000+0 1.000000+0 0 0 0 69743 1451 2 1.000000+0 2.000000+7 0 0 10 59743 1451 3 2.936000+2 0.000000+0 4 0 316 149743 1451 4 97-Bk-246 JAEA+ EVAL-FEB10 O.Iwamoto, T.Nakagawa, et al. 9743 1451 5 DIST-DEC21 20100304 9743 1451 6 ----JENDL-5 MATERIAL 9743 9743 1451 7 -----INCIDENT NEUTRON DATA 9743 1451 8 ------ENDF-6 FORMAT 9743 1451 9 9743 1451 10 History 9743 1451 11 07-10 Theoretical calculation was performed with CCONE code. 9743 1451 12 07-11 Data were compiled as JENDL/AC-2008/1/. 9743 1451 13 10-02 Data of prompt gamma rays due to fission were given. 9743 1451 14 10-03 Covariance data were given. 9743 1451 15 9743 1451 16 ********************************************************* 9743 1451 17 Bk-246 2- state is considered as an isomer/2/. 9743 1451 18 However, its level energy is unknown. Therefore, the 2- 9743 1451 19 state was treated as a ground state in this file. 9743 1451 20 ********************************************************* 9743 1451 21 21-11 revised by O.Iwamoto 9743 1451 22 (MF3/MT19-21,38) deleted 9743 1451 23 (MF8/MT4,16-18,37,102) added 9743 1451 24 9743 1451 25 MF=1 General information 9743 1451 26 MT=452 Number of Neutrons per fission 9743 1451 27 Sum of MT's=455 and 456. 9743 1451 28 9743 1451 29 MT=455 Delayed neutron data 9743 1451 30 Estimated from systematics by Tuttle/3/, Benedetti et al./4/ 9743 1451 31 and Waldo et al./5/ 9743 1451 32 9743 1451 33 MT=456 Number of prompt neutrons per fission 9743 1451 34 Estimated from Howerton's systematics/6/. 9743 1451 35 9743 1451 36 9743 1451 37 MF= 2 Resonance parameters 9743 1451 38 MT=151 9743 1451 39 No resonance parameters are given. 9743 1451 40 9743 1451 41 9743 1451 42 Thermal cross sections and resonance integrals (at 300K) 9743 1451 43 ------------------------------------------------------- 9743 1451 44 0.0253 eV reson. integ.(*) 9743 1451 45 (barns) (barns) 9743 1451 46 ------------------------------------------------------- 9743 1451 47 total 2511.9 9743 1451 48 elastic 10.22 9743 1451 49 fission 1801.2 836 9743 1451 50 capture 700.2 312 9743 1451 51 ------------------------------------------------------- 9743 1451 52 (*) In the energy range from 0.5 eV to 10 MeV. 9743 1451 53 9743 1451 54 9743 1451 55 MF= 3 Neutron cross sections 9743 1451 56 Below 0.25 eV: 9743 1451 57 * Elastic scattering cross section is 10.2 b calculated from 9743 1451 58 scattering radius of 9.000 fm/7/. 9743 1451 59 * Fission and capture cross sections are in the 1/v shape. 9743 1451 60 CCONE calculation was extrapolated to 1.0e-5 eV. 9743 1451 61 9743 1451 62 Above 0.25 eV: 9743 1451 63 Cross sections calculated with CCONE code/7/ were adopted. 9743 1451 64 9743 1451 65 MT= 1 Total cross section 9743 1451 66 The cross section was calculated with CC OMP of Soukhovitskii 9743 1451 67 et al./8/ 9743 1451 68 9743 1451 69 MT=18 Fission cross section 9743 1451 70 The simulated (n,f) cross sections of Britt and Wilhelmy/9/ 9743 1451 71 were used to determine the parameters in the CCONE 9743 1451 72 calculation. 9743 1451 73 9743 1451 74 9743 1451 75 MF= 4 Angular distributions of secondary neutrons 9743 1451 76 MT=2 Elastic scattering 9743 1451 77 Calculated with CCONE code/7/. 9743 1451 78 9743 1451 79 MT=18 Fission 9743 1451 80 Isotropic distributions in the laboratory system were assumed.9743 1451 81 9743 1451 82 9743 1451 83 MF= 5 Energy distributions of secondary neutrons 9743 1451 84 MT=18 Prompt neutrons 9743 1451 85 Calculated with CCONE code/7/. 9743 1451 86 9743 1451 87 9743 1451 88 MF= 6 Energy-angle distributions 9743 1451 89 Calculated with CCONE code/7/. 9743 1451 90 Distributions from fission (MT=18) are not included. 9743 1451 91 9743 1451 92 9743 1451 93 MF=12 Photon production multiplicities 9743 1451 94 MT=18 Fission 9743 1451 95 Calculated from the total energy released by the prompt 9743 1451 96 gamma-rays due to fission which was estimated from its 9743 1451 97 systematics, and the average energy of gamma-rays. 9743 1451 98 9743 1451 99 9743 1451 100 MF=14 Photon angular distributions 9743 1451 101 MT=18 Fission 9743 1451 102 Isotoropic distributions were assumed. 9743 1451 103 9743 1451 104 9743 1451 105 MF=15 Continuous photon energy spectra 9743 1451 106 MT=18 Fission 9743 1451 107 Experimental data measured by Verbinski et al./10/ for 9743 1451 108 Pu-239 thermal fission were adopted. 9743 1451 109 9743 1451 110 9743 1451 111 MF=31 Covariances of average number of neutrons per fission 9743 1451 112 MT=452 Number of neutrons per fission 9743 1451 113 Sum of covariances for MT=455 and MT=456. 9743 1451 114 9743 1451 115 MT=455 9743 1451 116 Error of 15% was assumed. 9743 1451 117 9743 1451 118 MT=456 9743 1451 119 Covariance was obtained by fitting a linear function to the 9743 1451 120 at 0.0 and 5.0 MeV with an uncertainty of 10%. 9743 1451 121 9743 1451 122 9743 1451 123 MF=33 Covariances of neutron cross sections 9743 1451 124 Covariances were given to all the cross sections by using 9743 1451 125 KALMAN code/11/ and the covariances of model parameters 9743 1451 126 used in the cross-section calculations. 9743 1451 127 9743 1451 128 Covariances of the fission cross section were determined from 9743 1451 129 experimental data. 9743 1451 130 9743 1451 131 For the following cross sections, standard deviations in the 9743 1451 132 energy region below 0.25 eV were assumed as follows: 9743 1451 133 9743 1451 134 Total 69 % 9743 1451 135 Elastic scattering 90 % 9743 1451 136 Fission 90 % 9743 1451 137 Capture 90 % 9743 1451 138 9743 1451 139 9743 1451 140 MF=34 Covariances for Angular Distributions 9743 1451 141 MT=2 Elastic scattering 9743 1451 142 Covariances were given only to P1 components. 9743 1451 143 9743 1451 144 9743 1451 145 MF=35 Covariances for Energy Distributions 9743 1451 146 MT=18 Fission spectra 9743 1451 147 Estimated with CCONE and KALMAN codes. 9743 1451 148 9743 1451 149 9743 1451 150 ***************************************************************** 9743 1451 151 Calculation with CCONE code 9743 1451 152 ***************************************************************** 9743 1451 153 9743 1451 154 Models and parameters used in the CCONE/7/ calculation 9743 1451 155 1) Coupled channel optical model 9743 1451 156 Levels in the rotational band were included. Optical model 9743 1451 157 potential and coupled levels are shown in Table 1. 9743 1451 158 9743 1451 159 2) Two-component exciton model/12/ 9743 1451 160 * Global parametrization of Koning-Duijvestijn/13/ 9743 1451 161 was used. 9743 1451 162 * Gamma emission channel/14/ was added to simulate direct 9743 1451 163 and semi-direct capture reaction. 9743 1451 164 9743 1451 165 3) Hauser-Feshbach statistical model 9743 1451 166 * Moldauer width fluctuation correction/15/ was included. 9743 1451 167 * Neutron, gamma and fission decay channel were included. 9743 1451 168 * Transmission coefficients of neutrons were taken from 9743 1451 169 coupled channel calculation in Table 1. 9743 1451 170 * The level scheme of the target is shown in Table 2. 9743 1451 171 * Level density formula of constant temperature and Fermi-gas 9743 1451 172 model were used with shell energy correction and collective 9743 1451 173 enhancement factor. Parameters are shown in Table 3. 9743 1451 174 * Fission channel: 9743 1451 175 Double humped fission barriers were assumed. 9743 1451 176 Fission barrier penetrabilities were calculated with 9743 1451 177 Hill-Wheler formula/16/. Fission barrier parameters were 9743 1451 178 shown in Table 4. Transition state model was used and 9743 1451 179 continuum levels are assumed above the saddles. The level 9743 1451 180 density parameters for inner and outer saddles are shown in 9743 1451 181 Tables 5 and 6, respectively. 9743 1451 182 * Gamma-ray strength function of Kopecky et al/17/,/18/ 9743 1451 183 was used. The prameters are shown in Table 7. 9743 1451 184 9743 1451 185 9743 1451 186 ------------------------------------------------------------------9743 1451 187 Tables 9743 1451 188 ------------------------------------------------------------------9743 1451 189 9743 1451 190 Table 1. Coupled channel calculation 9743 1451 191 -------------------------------------------------- 9743 1451 192 * rigid rotor model was applied 9743 1451 193 * coupled levels = 0,1,2,3 (see Table 2) 9743 1451 194 * optical potential parameters /8/ 9743 1451 195 Volume: 9743 1451 196 V_0 = 49.97 MeV 9743 1451 197 lambda_HF = 0.01004 1/MeV 9743 1451 198 C_viso = 15.9 MeV 9743 1451 199 A_v = 12.04 MeV 9743 1451 200 B_v = 81.36 MeV 9743 1451 201 E_a = 385 MeV 9743 1451 202 r_v = 1.2568 fm 9743 1451 203 a_v = 0.633 fm 9743 1451 204 Surface: 9743 1451 205 W_0 = 17.2 MeV 9743 1451 206 B_s = 11.19 MeV 9743 1451 207 C_s = 0.01361 1/MeV 9743 1451 208 C_wiso = 23.5 MeV 9743 1451 209 r_s = 1.1803 fm 9743 1451 210 a_s = 0.601 fm 9743 1451 211 Spin-orbit: 9743 1451 212 V_so = 5.75 MeV 9743 1451 213 lambda_so = 0.005 1/MeV 9743 1451 214 W_so = -3.1 MeV 9743 1451 215 B_so = 160 MeV 9743 1451 216 r_so = 1.1214 fm 9743 1451 217 a_so = 0.59 fm 9743 1451 218 Coulomb: 9743 1451 219 C_coul = 1.3 9743 1451 220 r_c = 1.2452 fm 9743 1451 221 a_c = 0.545 fm 9743 1451 222 Deformation: 9743 1451 223 beta_2 = 0.213 9743 1451 224 beta_4 = 0.066 9743 1451 225 beta_6 = 0.0015 9743 1451 226 9743 1451 227 * Calculated strength function 9743 1451 228 S0= 0.97e-4 S1= 3.73e-4 R'= 9.00 fm (En=1 keV) 9743 1451 229 -------------------------------------------------- 9743 1451 230 9743 1451 231 Table 2. Level Scheme of Bk-246 9743 1451 232 ------------------- 9743 1451 233 No. Ex(MeV) J PI 9743 1451 234 ------------------- 9743 1451 235 0 0.00000 2 - * 9743 1451 236 1 0.04800 3 - * 9743 1451 237 2 0.11200 4 - * 9743 1451 238 ------------------- 9743 1451 239 *) Coupled levels in CC calculation 9743 1451 240 9743 1451 241 Table 3. Level density parameters 9743 1451 242 -------------------------------------------------------- 9743 1451 243 Nuclide a* Pair Eshell T E0 Ematch 9743 1451 244 1/MeV MeV MeV MeV MeV MeV 9743 1451 245 -------------------------------------------------------- 9743 1451 246 Bk-247 18.9645 0.7635 1.3193 0.3836 -0.7996 2.9113 9743 1451 247 Bk-246 18.8984 0.0000 0.9401 0.2886 -0.6877 1.0000 9743 1451 248 Bk-245 18.8322 0.7667 1.2089 0.4136 -1.0946 3.2897 9743 1451 249 Bk-244 18.7661 0.0000 1.0000 0.2893 -0.6864 1.0000 9743 1451 250 Bk-243 18.6999 0.7698 1.1486 0.4243 -1.1853 3.4128 9743 1451 251 -------------------------------------------------------- 9743 1451 252 9743 1451 253 Table 4. Fission barrier parameters 9743 1451 254 ---------------------------------------- 9743 1451 255 Nuclide V_A hw_A V_B hw_B 9743 1451 256 MeV MeV MeV MeV 9743 1451 257 ---------------------------------------- 9743 1451 258 Bk-247 6.300 0.800 5.970 0.520 9743 1451 259 Bk-246 6.200 0.650 6.600 0.500 9743 1451 260 Bk-245 6.300 0.800 5.970 0.520 9743 1451 261 Bk-244 6.200 0.650 5.550 0.450 9743 1451 262 Bk-243 6.200 0.800 4.800 0.520 9743 1451 263 ---------------------------------------- 9743 1451 264 9743 1451 265 Table 5. Level density above inner saddle 9743 1451 266 -------------------------------------------------------- 9743 1451 267 Nuclide a* Pair Eshell T E0 Ematch 9743 1451 268 1/MeV MeV MeV MeV MeV MeV 9743 1451 269 -------------------------------------------------------- 9743 1451 270 Bk-247 21.2402 0.8908 2.6000 0.2836 -0.9848 2.3908 9743 1451 271 Bk-246 21.1662 0.0000 2.6000 0.3230 -2.4113 2.0000 9743 1451 272 Bk-245 21.0921 0.8944 2.6000 0.2846 -0.9813 2.3944 9743 1451 273 Bk-244 21.0180 0.0000 2.6000 0.3242 -2.4113 2.0000 9743 1451 274 Bk-243 20.9439 0.8981 2.6000 0.3248 -1.5132 2.8981 9743 1451 275 -------------------------------------------------------- 9743 1451 276 9743 1451 277 Table 6. Level density above outer saddle 9743 1451 278 -------------------------------------------------------- 9743 1451 279 Nuclide a* Pair Eshell T E0 Ematch 9743 1451 280 1/MeV MeV MeV MeV MeV MeV 9743 1451 281 -------------------------------------------------------- 9743 1451 282 Bk-247 24.6538 0.8908 0.8800 0.2869 -0.3705 2.3908 9743 1451 283 Bk-246 21.3551 0.0000 0.8400 0.3594 -1.7906 2.1000 9743 1451 284 Bk-245 24.4819 0.8944 0.8000 0.2888 -0.3662 2.3944 9743 1451 285 Bk-244 21.0180 0.0000 0.7600 0.3563 -1.7065 2.0000 9743 1451 286 Bk-243 20.9439 0.8981 0.7200 0.3575 -0.8078 2.8981 9743 1451 287 -------------------------------------------------------- 9743 1451 288 9743 1451 289 Table 7. Gamma-ray strength function for Bk-247 9743 1451 290 -------------------------------------------------------- 9743 1451 291 K0 = 1.500 E0 = 4.500 (MeV) 9743 1451 292 * E1: ER = 11.41 (MeV) EG = 2.72 (MeV) SIG = 250.94 (mb) 9743 1451 293 ER = 14.31 (MeV) EG = 4.19 (MeV) SIG = 501.88 (mb) 9743 1451 294 * M1: ER = 6.53 (MeV) EG = 4.00 (MeV) SIG = 1.78 (mb) 9743 1451 295 * E2: ER = 10.04 (MeV) EG = 3.15 (MeV) SIG = 7.21 (mb) 9743 1451 296 -------------------------------------------------------- 9743 1451 297 9743 1451 298 9743 1451 299 References 9743 1451 300 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 9743 1451 301 2) ENSDF, Evaluated Nuclear Structure Data File (2007). 9743 1451 302 3) R.J.Tuttld: INDC(NDS)-107/G+Special, p.29 (1979). 9743 1451 303 4) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 9743 1451 304 5) R.Waldo et al.: Phys. Rev., C23, 1113 (1981). 9743 1451 305 6) R.J.Howerton: Nucl. Sci. Eng., 62, 438 (1977). 9743 1451 306 7) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 9743 1451 307 8) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005). 9743 1451 308 9) H.C.Britt, J.B.Wilhelmy: Nucl. Sci. Eng., 72, 222 (1979). 9743 1451 309 10) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973). 9743 1451 310 11) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in 9743 1451 311 Japanese. 9743 1451 312 12) C.Kalbach: Phys. Rev. C33, 818 (1986). 9743 1451 313 13) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004). 9743 1451 314 14) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985). 9743 1451 315 15) P.A.Moldauer: Nucl. Phys. A344, 185 (1980). 9743 1451 316 16) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953). 9743 1451 317 17) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990). 9743 1451 318 18) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). 9743 1451 319 9743 1451 320 1 451 334 19743 1451 321 2 151 4 19743 1451 322 3 4 26 19743 1451 323 3 16 13 19743 1451 324 3 17 8 19743 1451 325 3 18 113 19743 1451 326 3 37 4 19743 1451 327 3 102 146 19743 1451 328 8 4 2 19743 1451 329 8 16 2 19743 1451 330 8 17 2 19743 1451 331 8 18 2 19743 1451 332 8 37 2 19743 1451 333 8 102 2 19743 1451 334 9743 1 099999 9743 0 0 0 9.724600+4 2.439550+2 0 0 1 09743 2151 1 9.724600+4 1.000000+0 0 0 1 09743 2151 2 1.000000-5 2.500000-1 0 0 0 09743 2151 3 2.000000+0 9.000000-1 0 0 0 09743 2151 4 9743 2 099999 9743 0 0 0 9.724600+4 2.439550+2 0 0 0 09743 3 4 1 0.000000+0-4.800000+4 0 0 1 689743 3 4 2 68 2 9743 3 4 3 4.819680+4 0.000000+0 5.000000+4 1.566550-2 7.000000+4 8.629570-29743 3 4 4 1.000000+5 1.917400-1 1.071020+5 2.125695-1 1.124590+5 2.282831-19743 3 4 5 1.200000+5 2.545387-1 1.400000+5 3.283022-1 1.700000+5 4.269908-19743 3 4 6 2.000000+5 5.114073-1 2.500000+5 7.558250-1 3.000000+5 8.009430-19743 3 4 7 4.000000+5 1.134812+0 5.000000+5 1.482877+0 6.000000+5 1.725840+09743 3 4 8 7.000000+5 1.935811+0 8.000000+5 2.075439+0 9.000000+5 2.192886+09743 3 4 9 1.000000+6 2.302037+0 1.100000+6 2.382106+0 1.200000+6 2.447618+09743 3 4 10 1.300000+6 2.496425+0 1.400000+6 2.534003+0 1.500000+6 2.551714+09743 3 4 11 1.600000+6 2.559260+0 1.700000+6 2.548057+0 1.800000+6 2.529044+09743 3 4 12 1.900000+6 2.496084+0 2.000000+6 2.458721+0 2.100000+6 2.413571+09743 3 4 13 2.300000+6 2.316375+0 2.500000+6 2.215296+0 2.700000+6 2.116700+09743 3 4 14 3.000000+6 1.981195+0 3.500000+6 1.786249+0 4.000000+6 1.630359+09743 3 4 15 4.500000+6 1.506597+0 5.000000+6 1.409549+0 5.500000+6 1.337158+09743 3 4 16 6.000000+6 1.282752+0 6.500000+6 1.153275+0 7.000000+6 9.525980-19743 3 4 17 7.500000+6 7.897590-1 8.000000+6 6.756369-1 8.500000+6 5.991347-19743 3 4 18 9.000000+6 5.473849-1 9.500000+6 5.107033-1 1.000000+7 4.828992-19743 3 4 19 1.050000+7 4.601760-1 1.100000+7 4.429805-1 1.150000+7 4.240991-19743 3 4 20 1.200000+7 4.090759-1 1.250000+7 3.944026-1 1.300000+7 3.799787-19743 3 4 21 1.350000+7 3.687725-1 1.400000+7 3.547080-1 1.450000+7 3.438589-19743 3 4 22 1.500000+7 3.303668-1 1.550000+7 3.202186-1 1.600000+7 3.101136-19743 3 4 23 1.650000+7 3.007419-1 1.700000+7 2.883092-1 1.750000+7 2.796232-19743 3 4 24 1.800000+7 2.709652-1 1.850000+7 2.659229-1 1.900000+7 2.580117-19743 3 4 25 1.950000+7 2.506846-1 2.000000+7 2.434106-1 9743 3 4 26 9743 3 099999 9.724600+4 2.439550+2 0 0 0 09743 3 16 1 -5.918330+6-5.918330+6 0 0 1 299743 3 16 2 29 2 9743 3 16 3 5.942590+6 0.000000+0 6.500000+6 8.662480-2 7.000000+6 2.280900-19743 3 16 4 7.500000+6 3.053060-1 8.000000+6 3.402720-1 8.500000+6 3.513310-19743 3 16 5 9.000000+6 3.492260-1 9.500000+6 3.413430-1 1.000000+7 3.324990-19743 3 16 6 1.050000+7 3.256820-1 1.100000+7 3.187220-1 1.150000+7 3.200860-19743 3 16 7 1.200000+7 3.200080-1 1.250000+7 3.213910-1 1.300000+7 3.201130-19743 3 16 8 1.350000+7 3.106040-1 1.400000+7 3.022030-1 1.450000+7 2.875830-19743 3 16 9 1.500000+7 2.769900-1 1.550000+7 2.630510-1 1.600000+7 2.500750-19743 3 16 10 1.650000+7 2.390230-1 1.700000+7 2.337670-1 1.750000+7 2.254690-19743 3 16 11 1.800000+7 2.180640-1 1.850000+7 2.072400-1 1.900000+7 2.014390-19743 3 16 12 1.950000+7 1.962000-1 2.000000+7 1.913100-1 9743 3 16 13 9743 3 099999 9.724600+4 2.439550+2 0 0 0 09743 3 17 1 -1.288970+7-1.288970+7 0 0 1 159743 3 17 2 15 2 9743 3 17 3 1.294250+7 0.000000+0 1.350000+7 2.887950-5 1.400000+7 4.888520-49743 3 17 4 1.450000+7 2.071500-3 1.500000+7 5.086720-3 1.550000+7 9.344550-39743 3 17 5 1.600000+7 1.434800-2 1.650000+7 1.797830-2 1.700000+7 2.139620-29743 3 17 6 1.750000+7 2.462470-2 1.800000+7 2.756610-2 1.850000+7 3.024360-29743 3 17 7 1.900000+7 3.267830-2 1.950000+7 3.496030-2 2.000000+7 3.702020-29743 3 17 8 9743 3 099999 9.724600+4 2.439550+2 0 0 0 09743 3 18 1 2.102000+8 2.102000+8 0 0 1 3289743 3 18 2 328 2 9743 3 18 3 1.000000-5 9.154102+4 1.103037-5 8.715959+4 1.216691-5 8.298776+49743 3 18 4 1.342055-5 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