JENDL-5 (Neutron sublibrary (activation cs)) Np-238 0 0 0 0 9.323800+4 2.360060+2 0 0 6 19349 1451 1 0.000000+0 1.000000+0 0 0 0 69349 1451 2 1.000000+0 2.000000+7 0 0 10 59349 1451 3 2.936000+2 0.000000+0 4 0 410 149349 1451 4 93-Np-238 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,K.Furutaka,+9349 1451 5 DIST-DEC21 20100318 9349 1451 6 ----JENDL-5 MATERIAL 9349 9349 1451 7 -----INCIDENT NEUTRON DATA 9349 1451 8 ------ENDF-6 FORMAT 9349 1451 9 9349 1451 10 History 9349 1451 11 07-07 New theoretical calculation was made with CCONE code. 9349 1451 12 07-10 New theoretical calculation was made with CCONE code. 9349 1451 13 08-01 Resolved resonance parameters were revised. 9349 1451 14 Data were compiled as JENDL/AC-2008/1/. 9349 1451 15 09-02 (1,452), (1,455) and (1,456) were revised. 9349 1451 16 09-08 (MF1,MT458) was evaluated. 9349 1451 17 10-01 Data of prompt gamma rays due to fission were given. 9349 1451 18 10-03 Covariance data were given. 9349 1451 19 9349 1451 20 21-11 revised by O.Iwamoto 9349 1451 21 (MF3/MT19-21,38) deleted 9349 1451 22 (MF8/MT4,16-18,37,102) added 9349 1451 23 9349 1451 24 MF= 1 General information 9349 1451 25 MT=452 Number of Neutrons per fission 9349 1451 26 Sum of MT's=455 and 456. 9349 1451 27 9349 1451 28 MT=455 Delayed neutron data 9349 1451 29 Determined from systematics by Tuttle/2/, Benedetti et al./3/ 9349 1451 30 and Waldo et al./4/, and partial fission cross sections 9349 1451 31 calculated with CCONE code/5/. 9349 1451 32 Decay constants were taken from the evaluation of Brady and 9349 1451 33 England/6/. 9349 1451 34 9349 1451 35 MT=456 Number of prompt neutrons per fission 9349 1451 36 Based on the data of Solonkin et al./7/ (2.3+-0.5 at 0.0253 9349 1451 37 eV) and Ohsawa's systematics/8/. A constant term is an average9349 1451 38 of these two. 9349 1451 39 9349 1451 40 MT=458 Components of energy release due to fission 9349 1451 41 Total energy and prompt energy were calculated from mass 9349 1451 42 balance using JENDL-4 fission yields data and mass excess 9349 1451 43 evaluation. Mass excess values were from Audi's 2009 9349 1451 44 evaluation/9/. Delayed energy values were calculated from 9349 1451 45 the energy release for infinite irradiation using JENDL FP 9349 1451 46 Decay Data File 2000 and JENDL-4 yields data. For delayed 9349 1451 47 neutron energy, as the JENDL FP Decay Data File 2000/10/ does 9349 1451 48 not include average neutron energy values, the average values 9349 1451 49 were calculated using the formula shown in the report by 9349 1451 50 T.R. England/11/. The fractions of prompt energy were 9349 1451 51 calculated using the fractions of Sher's evaluation/12/ when 9349 1451 52 they were provided. When the fractions were not given by Sher,9349 1451 53 averaged fractions were used. 9349 1451 54 9349 1451 55 9349 1451 56 MF= 2 Resonance parameters 9349 1451 57 MT=151 9349 1451 58 Resolved resonance parameters (MLBW: 1.0e-5 - 6.65 eV) 9349 1451 59 Evaluated by Furutaka/13/. 9349 1451 60 Parameters were obtained, starting from the parameters 9349 1451 61 evaluated by Morogovskij/14/, with SAMMY code /15/ to 9349 1451 62 reproduce the fission cross section measured by Danon et 9349 1451 63 al./16/ Their data were normalized to 2130 b at 0.0253 eV. 9349 1451 64 The capture width was fixed to 50 meV. A negative resonance 9349 1451 65 was assumed to reproduce the thermal cross sections: 9349 1451 66 efective capture = 479+-24 /17/ 9349 1451 67 fission = 2201+-34 /18,16,19/ 9349 1451 68 Doppler as well as resolution broadenings were taken into 9349 1451 69 account in the analysis: temperature was assumed to be 300 K. 9349 1451 70 For resolution broadening, parameters of SAMMY's original 9349 1451 71 resolution-broadening function were chosen to approximately 9349 1451 72 reproduce the experimental resolution function described by 9349 1451 73 equation (11) of ref./16/. 9349 1451 74 9349 1451 75 Un-resolved resonance parameters (6.65 eV - 10 keV) 9349 1451 76 Parameters (URP) were determined with ASREP code /20/ so as to9349 1451 77 reproduce the cross sections in this energy region. URP are 9349 1451 78 used only for self-shielding calculations. 9349 1451 79 9349 1451 80 9349 1451 81 Thermal cross sections and resonance integrals (at 300K) 9349 1451 82 ------------------------------------------------------- 9349 1451 83 0.0253 eV reson. integ.(*) 9349 1451 84 (barns) (barns) 9349 1451 85 ------------------------------------------------------- 9349 1451 86 total 2693.3 9349 1451 87 elastic 12.26 9349 1451 88 fission 2201.6 1100 9349 1451 89 capture 479.5 201 9349 1451 90 ------------------------------------------------------- 9349 1451 91 (*) In the energy range from 0.5 eV to 10 MeV. 9349 1451 92 9349 1451 93 9349 1451 94 MF= 3 Neutron cross sections 9349 1451 95 All the cross-section data above 6.65 eV were calculated with 9349 1451 96 CCONE code/5/. 9349 1451 97 9349 1451 98 MT= 1 Total cross section 9349 1451 99 The cross section was calculated with CC OMP of Soukhovitskii 9349 1451 100 et al./21/ 9349 1451 101 9349 1451 102 MT=18 Fission cross section 9349 1451 103 Calculated with CCONE code. The simulated (n,f) cross section 9349 1451 104 of Britt and Wilhelmy/22/, and the experimental data of Danon 9349 1451 105 et al./16/ were used to determine the parameters in the CCONE 9349 1451 106 calculation. 9349 1451 107 9349 1451 108 9349 1451 109 MF= 4 Angular distributions of secondary neutrons 9349 1451 110 MT=2 Elastic scattering 9349 1451 111 Calculated with CCONE code. 9349 1451 112 9349 1451 113 MT=18 Fission 9349 1451 114 Isotropic distributions in the laboratory system were assumed.9349 1451 115 9349 1451 116 9349 1451 117 MF= 5 Energy distributions of secondary neutrons 9349 1451 118 MT=18 Prompt neutrons 9349 1451 119 Calculated with CCONE code. 9349 1451 120 9349 1451 121 MT=455 Delayed neutrons 9349 1451 122 Calculated by Brady and England /6/. 9349 1451 123 9349 1451 124 9349 1451 125 MF= 6 Energy-angle distributions 9349 1451 126 Calculated with CCONE code. 9349 1451 127 Distributions from fission (MT=18) are not included. 9349 1451 128 9349 1451 129 9349 1451 130 MF=12 Photon production multiplicities 9349 1451 131 MT=18 Fission 9349 1451 132 Calculated from the total energy released by the prompt 9349 1451 133 gamma-rays due to fission given in MF=1/MT=458 and the 9349 1451 134 average energy of gamma-rays. 9349 1451 135 9349 1451 136 9349 1451 137 MF=14 Photon angular distributions 9349 1451 138 MT=18 Fission 9349 1451 139 Isotoropic distributions were assumed. 9349 1451 140 9349 1451 141 9349 1451 142 MF=15 Continuous photon energy spectra 9349 1451 143 MT=18 Fission 9349 1451 144 Experimental data measured by Verbinski et al./23/ for 9349 1451 145 Pu-239 thermal fission were adopted. 9349 1451 146 9349 1451 147 9349 1451 148 MF=31 Covariances of average number of neutrons per fission 9349 1451 149 MT=452 Number of neutrons per fission 9349 1451 150 Sum of covariances for MT=455 and MT=456. 9349 1451 151 9349 1451 152 MT=455 9349 1451 153 Error of 15% was assumed. 9349 1451 154 9349 1451 155 MT=456 9349 1451 156 Covariance was obtained by fitting a linear function to the 9349 1451 157 data at 0.0 and 5.0 MeV with an uncertainty of 22% which was 9349 1451 158 estimated from the experimental data of Solonkin et al./7/ 9349 1451 159 9349 1451 160 9349 1451 161 MF=32 Covariances of resonance parameters 9349 1451 162 MT=151 Resolved resonance parameterss 9349 1451 163 Format of LCOMP=1 was adopted. 9349 1451 164 9349 1451 165 Covariances of parameters were taken from the results of SAMMY9349 1451 166 analysis/13/. The uncertainty of capture width was assumed 9349 1451 167 to be 30%. 9349 1451 168 9349 1451 169 9349 1451 170 MF=33 Covariances of neutron cross sections 9349 1451 171 Covariances were given to all the cross sections by using 9349 1451 172 KALMAN code/24/ and the covariances of model parameters 9349 1451 173 used in the cross-section calculations. 9349 1451 174 9349 1451 175 Covariances of the fission cross section were determined by 9349 1451 176 considering the experimental data (see MF=3). 9349 1451 177 9349 1451 178 In the resolved resonance region, the following standard 9349 1451 179 deviations were added to the contributions from resonance 9349 1451 180 parameters: 9349 1451 181 Total 0 - 10 % 9349 1451 182 Elastic scattering 20 % 9349 1451 183 Fission 0 - 10 % 9349 1451 184 Capture 0 - 10 % 9349 1451 185 9349 1451 186 9349 1451 187 MF=34 Covariances for Angular Distributions 9349 1451 188 MT=2 Elastic scattering 9349 1451 189 Covariances were given only to P1 components. 9349 1451 190 9349 1451 191 9349 1451 192 MF=35 Covariances for Energy Distributions 9349 1451 193 MT=18 Fission spectra 9349 1451 194 Estimated with CCONE and KALMAN codes. 9349 1451 195 9349 1451 196 9349 1451 197 ***************************************************************** 9349 1451 198 Calculation with CCONE code 9349 1451 199 ***************************************************************** 9349 1451 200 9349 1451 201 Models and parameters used in the CCONE/5/ calculation 9349 1451 202 1) Coupled channel optical model 9349 1451 203 Levels in the rotational band were included. Optical model 9349 1451 204 potential and coupled levels are shown in Table 1. 9349 1451 205 9349 1451 206 2) Two-component exciton model/25/ 9349 1451 207 * Global parametrization of Koning-Duijvestijn/26/ 9349 1451 208 was used. 9349 1451 209 * Gamma emission channel/27/ was added to simulate direct 9349 1451 210 and semi-direct capture reaction. 9349 1451 211 9349 1451 212 3) Hauser-Feshbach statistical model 9349 1451 213 * Moldauer width fluctuation correction/28/ was included. 9349 1451 214 * Neutron, gamma and fission decay channel were included. 9349 1451 215 * Transmission coefficients of neutrons were taken from 9349 1451 216 coupled channel calculation in Table 1. 9349 1451 217 * The level scheme of the target is shown in Table 2. 9349 1451 218 * Level density formula of constant temperature and Fermi-gas 9349 1451 219 model were used with shell energy correction and collective 9349 1451 220 enhancement factor. Parameters are shown in Table 3. 9349 1451 221 * Fission channel: 9349 1451 222 Double humped fission barriers were assumed. 9349 1451 223 Fission barrier penetrabilities were calculated with 9349 1451 224 Hill-Wheler formula/29/. Fission barrier parameters were 9349 1451 225 shown in Table 4. Transition state model was used and 9349 1451 226 continuum levels are assumed above the saddles. The level 9349 1451 227 density parameters for inner and outer saddles are shown in 9349 1451 228 Tables 5 and 6, respectively. 9349 1451 229 * Gamma-ray strength function of Kopecky et al/30/,/31/ 9349 1451 230 was used. The prameters are shown in Table 7. 9349 1451 231 9349 1451 232 9349 1451 233 ------------------------------------------------------------------9349 1451 234 Tables 9349 1451 235 ------------------------------------------------------------------9349 1451 236 9349 1451 237 Table 1. Coupled channel calculation 9349 1451 238 -------------------------------------------------- 9349 1451 239 * rigid rotor model was applied 9349 1451 240 * coupled levels = 0,1,2,4,7 (see Table 2) 9349 1451 241 * optical potential parameters /21/ 9349 1451 242 Volume: 9349 1451 243 V_0 = 49.97 MeV 9349 1451 244 lambda_HF = 0.01004 1/MeV 9349 1451 245 C_viso = 15.9 MeV 9349 1451 246 A_v = 12.04 MeV 9349 1451 247 B_v = 81.36 MeV 9349 1451 248 E_a = 385 MeV 9349 1451 249 r_v = 1.2568 fm 9349 1451 250 a_v = 0.633 fm 9349 1451 251 Surface: 9349 1451 252 W_0 = 17.2 MeV 9349 1451 253 B_s = 11.19 MeV 9349 1451 254 C_s = 0.01361 1/MeV 9349 1451 255 C_wiso = 23.5 MeV 9349 1451 256 r_s = 1.1803 fm 9349 1451 257 a_s = 0.601 fm 9349 1451 258 Spin-orbit: 9349 1451 259 V_so = 5.75 MeV 9349 1451 260 lambda_so = 0.005 1/MeV 9349 1451 261 W_so = -3.1 MeV 9349 1451 262 B_so = 160 MeV 9349 1451 263 r_so = 1.1214 fm 9349 1451 264 a_so = 0.59 fm 9349 1451 265 Coulomb: 9349 1451 266 C_coul = 1.3 9349 1451 267 r_c = 1.2452 fm 9349 1451 268 a_c = 0.545 fm 9349 1451 269 Deformation: 9349 1451 270 beta_2 = 0.213 9349 1451 271 beta_4 = 0.066 9349 1451 272 beta_6 = 0.0015 9349 1451 273 9349 1451 274 * Calculated strength function 9349 1451 275 S0= 0.87e-4 S1= 3.05e-4 R'= 9.37 fm (En=1 keV) 9349 1451 276 -------------------------------------------------- 9349 1451 277 9349 1451 278 Table 2. Level Scheme of Np-238 9349 1451 279 ------------------- 9349 1451 280 No. Ex(MeV) J PI 9349 1451 281 ------------------- 9349 1451 282 0 0.00000 2 + * 9349 1451 283 1 0.02643 3 + * 9349 1451 284 2 0.06233 4 + * 9349 1451 285 3 0.08667 3 + 9349 1451 286 4 0.10615 5 + * 9349 1451 287 5 0.12165 4 + 9349 1451 288 6 0.13604 3 - 9349 1451 289 7 0.16168 6 + * 9349 1451 290 8 0.16553 5 + 9349 1451 291 9 0.17915 4 - 9349 1451 292 10 0.18288 2 - 9349 1451 293 11 0.21552 3 - 9349 1451 294 12 0.21795 0 - 9349 1451 295 13 0.21870 6 + 9349 1451 296 14 0.23283 5 - 9349 1451 297 15 0.24396 1 + 9349 1451 298 16 0.24640 1 + 9349 1451 299 17 0.25033 1 + 9349 1451 300 18 0.25039 2 - 9349 1451 301 19 0.25885 4 - 9349 1451 302 20 0.27552 5 + 9349 1451 303 21 0.27764 2 + 9349 1451 304 22 0.28580 1 - 9349 1451 305 23 0.29703 6 - 9349 1451 306 24 0.29837 3 + 9349 1451 307 25 0.29923 3 + 9349 1451 308 26 0.29979 1 - 9349 1451 309 27 0.30068 1 - 9349 1451 310 28 0.30074 6 - 9349 1451 311 29 0.30540 1 - 9349 1451 312 30 0.31270 5 - 9349 1451 313 31 0.31506 4 + 9349 1451 314 32 0.32431 4 - 9349 1451 315 33 0.32521 1 - 9349 1451 316 34 0.32860 6 + 9349 1451 317 35 0.33400 1 - 9349 1451 318 ------------------- 9349 1451 319 *) Coupled levels in CC calculation 9349 1451 320 9349 1451 321 Table 3. Level density parameters 9349 1451 322 -------------------------------------------------------- 9349 1451 323 Nuclide a* Pair Eshell T E0 Ematch 9349 1451 324 1/MeV MeV MeV MeV MeV MeV 9349 1451 325 -------------------------------------------------------- 9349 1451 326 Np-239 18.4349 0.7762 2.6850 0.3834 -0.8836 3.0329 9349 1451 327 Np-238 18.3685 0.0000 2.2742 0.3205 -0.9882 1.4160 9349 1451 328 Np-237 18.3022 0.7795 2.4371 0.3963 -0.9739 3.1569 9349 1451 329 Np-236 18.2358 0.0000 2.1332 0.2999 -0.7994 1.1664 9349 1451 330 Np-235 18.1694 0.7828 2.2924 0.3973 -0.9417 3.1303 9349 1451 331 -------------------------------------------------------- 9349 1451 332 9349 1451 333 Table 4. Fission barrier parameters 9349 1451 334 ---------------------------------------- 9349 1451 335 Nuclide V_A hw_A V_B hw_B 9349 1451 336 MeV MeV MeV MeV 9349 1451 337 ---------------------------------------- 9349 1451 338 Np-239 6.250 0.800 5.250 0.600 9349 1451 339 Np-238 6.200 0.460 5.850 0.370 9349 1451 340 Np-237 6.000 0.950 5.570 0.600 9349 1451 341 Np-236 6.100 0.600 6.080 0.600 9349 1451 342 Np-235 6.250 0.950 5.630 0.600 9349 1451 343 ---------------------------------------- 9349 1451 344 9349 1451 345 Table 5. Level density above inner saddle 9349 1451 346 -------------------------------------------------------- 9349 1451 347 Nuclide a* Pair Eshell T E0 Ematch 9349 1451 348 1/MeV MeV MeV MeV MeV MeV 9349 1451 349 -------------------------------------------------------- 9349 1451 350 Np-239 20.6471 0.9056 2.6000 0.3273 -1.5058 2.9056 9349 1451 351 Np-238 20.5728 0.0000 2.6000 0.3280 -2.4114 2.0000 9349 1451 352 Np-237 21.9626 0.9094 2.6000 0.3162 -1.4591 2.9094 9349 1451 353 Np-236 22.2477 0.0000 2.6000 0.3139 -2.3586 2.0000 9349 1451 354 Np-235 20.3497 0.9133 2.6000 0.3299 -1.4981 2.9133 9349 1451 355 -------------------------------------------------------- 9349 1451 356 9349 1451 357 Table 6. Level density above outer saddle 9349 1451 358 -------------------------------------------------------- 9349 1451 359 Nuclide a* Pair Eshell T E0 Ematch 9349 1451 360 1/MeV MeV MeV MeV MeV MeV 9349 1451 361 -------------------------------------------------------- 9349 1451 362 Np-239 22.1219 0.9056 0.3200 0.3278 -0.5333 2.6056 9349 1451 363 Np-238 20.5728 0.0000 0.2800 0.3661 -1.7021 2.0000 9349 1451 364 Np-237 22.3287 0.9094 0.2400 0.3268 -0.5253 2.6094 9349 1451 365 Np-236 22.2477 0.0000 0.2000 0.3977 -2.2734 2.7000 9349 1451 366 Np-235 22.1666 0.9133 0.1600 0.3517 -0.7691 2.9133 9349 1451 367 -------------------------------------------------------- 9349 1451 368 9349 1451 369 Table 7. Gamma-ray strength function for Np-239 9349 1451 370 -------------------------------------------------------- 9349 1451 371 K0 = 1.300 E0 = 4.500 (MeV) 9349 1451 372 * E1: ER = 10.98 (MeV) EG = 2.17 (MeV) SIG = 311.00 (mb) 9349 1451 373 ER = 14.08 (MeV) EG = 4.66 (MeV) SIG = 540.00 (mb) 9349 1451 374 * M1: ER = 6.61 (MeV) EG = 4.00 (MeV) SIG = 2.08 (mb) 9349 1451 375 * E2: ER = 10.15 (MeV) EG = 3.24 (MeV) SIG = 6.65 (mb) 9349 1451 376 -------------------------------------------------------- 9349 1451 377 9349 1451 378 9349 1451 379 References 9349 1451 380 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009). 9349 1451 381 2) R.J.Tuttle: INDC(NDS)-107/G+Special, p.29 (1979). 9349 1451 382 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982). 9349 1451 383 4) R.Waldo et al.: Phys. Rev., C23, 1113 (1981). 9349 1451 384 5) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007). 9349 1451 385 6) M.C.Brady, T.R.England: Nucl. Sci. 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Rev. C41, 1941 (1990). 9349 1451 412 31) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990). 9349 1451 413 9349 1451 414 1 451 428 19349 1451 415 2 151 4 19349 1451 416 3 4 38 19349 1451 417 3 16 13 19349 1451 418 3 17 9 19349 1451 419 3 18 291 19349 1451 420 3 37 5 19349 1451 421 3 102 348 19349 1451 422 8 4 2 19349 1451 423 8 16 2 19349 1451 424 8 17 2 19349 1451 425 8 18 2 19349 1451 426 8 37 2 19349 1451 427 8 102 2 19349 1451 428 9349 1 099999 9349 0 0 0 9.323800+4 2.360060+2 0 0 1 09349 2151 1 9.323800+4 1.000000+0 0 0 1 09349 2151 2 1.000000-5 6.650000+0 0 0 0 09349 2151 3 2.000000+0 9.400000-1 0 0 0 09349 2151 4 9349 2 099999 9349 0 0 0 9.323800+4 2.360060+2 0 0 0 09349 3 4 1 0.000000+0-2.642700+4 0 0 1 1039349 3 4 2 103 2 9349 3 4 3 2.653900+4 0.000000+0 3.000000+4 1.961150-2 5.000000+4 6.016990-29349 3 4 4 6.259410+4 8.474460-2 7.000000+4 1.024010-1 8.704120+4 1.459637-19349 3 4 5 8.704130+4 1.459639-1 1.000000+5 2.006485-1 1.041800+5 2.145943-19349 3 4 6 1.066050+5 2.226855-1 1.200000+5 2.667020-1 1.221600+5 2.735491-19349 3 4 7 1.366210+5 3.216206-1 1.400000+5 3.430392-1 1.623700+5 4.322990-19349 3 4 8 1.662330+5 4.458957-1 1.700000+5 4.592076-1 1.799130+5 4.930818-19349 3 4 9 1.836520+5 5.086141-1 2.000000+5 5.886427-1 2.164350+5 6.332715-19349 3 4 10 2.188720+5 6.445546-1 2.196270+5 6.483421-1 2.338140+5 6.932827-19349 3 4 11 2.338150+5 6.932853-1 2.449930+5 7.226293-1 2.474440+5 7.311113-19349 3 4 12 2.500000+5 7.410916-1 2.513910+5 7.458851-1 2.514510+5 7.464015-19349 3 4 13 2.599500+5 7.902179-1 2.766860+5 8.628655-1 2.788170+5 8.716067-19349 3 4 14 2.870110+5 9.099128-1 2.982890+5 9.644339-1 2.996320+5 9.703063-19349 3 4 15 3.000000+5 9.724651-1 3.004980+5 9.743058-1 3.010580+5 9.769798-19349 3 4 16 3.019540+5 9.824118-1 3.020170+5 9.832212-1 3.066940+5 1.006607+09349 3 4 17 3.140290+5 1.041146+0 3.163980+5 1.050843+0 3.256880+5 1.086855+09349 3 4 18 3.265880+5 1.090719+0 3.299920+5 1.106070+0 3.354150+5 1.126572+09349 3 4 19 4.000000+5 1.307327+0 5.000000+5 1.520494+0 6.000000+5 1.643436+09349 3 4 20 7.000000+5 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