B-1604,ISTC, JINER Am-243 9549 0 0 0 9.524300+4 2.409730+2 1 1 0 49549 1451 1 0.000000+0 1.000000+0 0 0 0 69549 1451 2 1.000000+0 2.000000+7 3 0 10 39549 1451 3 0.000000+0 0.000000+0 0 0 408 619549 1451 4 95-Am-243 9549 1451 5 9549 1451 6 Evaluation is accomplished under the Project Agreement 9549 1451 7 B-1604 with the International Science and Technology by9549 1451 8 V.M. Maslov, N.A. Tetereva, A.M. Kolesov, 9549 1451 9 of JINER (Belarus), 9549 1451 10 V.G.Pronyaev, K.I.Zolotarev, of IPPE (Russia) 9549 1451 11 in collaboration with T. Granier, CEA/DAM (France), 9549 1451 12 F.-J. Hambsch, JRC IRMM (Belgium). 9549 1451 13 Support of IAEA within Coordinated Research Project 9549 1451 14 MANREAD is acknowledged. 9549 1451 15 JINER 2010-001 DIST-OCT2010 20100930 9549 1451 16 ----B-1604 MATERIAL 9549 9549 1451 17 -----INCIDENT NEUTRON DATA 9549 1451 18 ------ENDF-6 FORMAT 9549 1451 19 HISTORY 9549 1451 20 Our previous activity /1/ regarding 243-Am neutron data 9549 1451 21 evaluation lead to data file compiled in 1996. Since then 9549 1451 22 new measurements as well as theoretical tools have emerged9549 1451 23 based on major actinides data analysis and evaluation. 9549 1451 24 9549 1451 25 2010-10 Evaluated are unresolved resonance parameters, 9549 1451 26 smooth cross sections, neutron angular distributions, 9549 1451 27 secondary neutron spectra, prompt fission neutron spectra,9549 1451 28 prompt fission neutron multiplicity. 9549 1451 29 9549 1451 30 Resolved neutron parameters of our previous evaluation are9549 1451 31 adopted /1/. 9549 1451 32 MF=1 General information 9549 1451 33 MT=451 Descriptive data and dictionary 9549 1451 34 MT=452 fission neutron multiplicity 9549 1451 35 sum of MT=455 and MT=456 9549 1451 36 MT=455 as delayed fission neutron decay constants 9549 1451 37 are adopted those for 237-Np /2/ 9549 1451 38 the energy dependence of average delayed neutron yield 9549 1451 39 was modeled comparing the values of the group yields 9549 1451 40 for 243-Am and 237-Np at thermal energy /3,4/ and rela- 9549 1451 41 tive energy dependence of the group yields known for 9549 1451 42 237-Np /2/. 9549 1451 43 MT=456 prompt fission neutron multiplicity 9549 1451 44 GMA code /5/ was used to fit measured data by Khokhlov 9549 1451 45 et al. /6/, normalized to nu-bar of 252-Cf(sf) /7/, and 9549 1451 46 Frehaut et al. /8/. 9549 1451 47 The thermal value was taken the same as value predicted 9549 1451 48 by the GMA fit at 0.1 keV, taking into account general 9549 1451 49 trend, observed for most nuclides in the resonance 9549 1451 50 region /9/. 9549 1451 51 At incident neutron energies higher than (n,nf) reaction 9549 1451 52 threshold slight bump in nu-bar was reproduced taking 9549 1451 53 into account partial contributions of (n,F) and (n,nf) 9549 1451 54 reactions as described in /10/. 9549 1451 55 9549 1451 56 MF=2 RESONANCE PARAMETERS 9549 1451 57 MT=151 Resolved resonance parameters (MLBW) 9549 1451 58 Resolved resonance region: 1.0E-5 - 250 EV 9549 1451 59 Parameters for Breit-Wigner formula are adopted from 9549 1451 60 /1/. They are based on total data measurements by 9549 1451 61 Berreth et al. /11/ and Simpson et al. /12/, fission 9549 1451 62 cross section measurement by Knitter et al /13/ and 9549 1451 63 Seeger et al. /14/. 9549 1451 64 The latter being normalized to give average fission 9549 1451 65 width =0.228 meV. 9549 1451 66 9549 1451 67 Thermal total based on Berreth et al. /11/. 9549 1451 68 Thermal fission based on data by Hulet et al. /15/, 9549 1451 69 Belanova et al. /16/, Asghar et al. /17/ and Wagemans 9549 1451 70 et al. /18/. 9549 1451 71 9549 1451 72 2200-M/S cross sections and calculated resonance integrals: 9549 1451 73 2200 m/s(b) Res. integ.(b) 9549 1451 74 Total 84.2340 9549 1451 75 ELASTIC 7.46417 - 9549 1451 76 FISSION 0.0638 7.46507 9549 1451 77 CAPTURE 76.7069 1781.13 9549 1451 78 9549 1451 79 9549 1451 80 2) Unresolved resonance parameters (250 eV - 96.8001 eV) 9549 1451 81 The average resonance parameters were determined 9549 1451 82 to reproduce average capture cross sections 9549 1451 83 (Gg=0.033 eV, D_obs=0.621 eV.): 9549 1451 84 9549 1451 85 total: optical model calculations, with coupled 9549 1451 86 channels potential parameters, fitting data by 9549 1451 87 Phillips and Howe /18/ for 241-Am in the energy 9549 1451 88 range 0.5-24 MeV 9549 1451 89 fission: Wisshak et al. /19/, Kanda et al. /20/, Behrens 9549 1451 90 et al. /21/, Knitter et al. /12, Butler et al. 9549 1451 91 /22/. 9549 1451 92 capture: Wisshak et al./19/, Weston and Todd /23/ 9549 1451 93 9549 1451 94 9549 1451 95 MF=3 Neutron cross sections 9549 1451 96 From .250 up to 96.8001 keV evaluated cross sections were9549 1451 97 represented with energy-dependent unresolved resonance 9549 1451 98 parameters (EDURP). EDURP reproduce total, scattering, 9549 1451 99 fission, capture and inelastic scattering cross sections,9549 1451 100 calculated with a Hauser-Feshbach-Moldauer statistical 9549 1451 101 model. The latter is used at higher energy domain as 9549 1451 102 well. 9549 1451 103 9549 1451 104 MT= 1, 2, 4, 51-60,91 - total, elastic and inelastic scattering 9549 1451 105 cross sections. 9549 1451 106 Total, elastic and direct inelastic for rotational ground9549 1451 107 state band levels MT=51,53,56(coupled levels), as well 9549 1451 108 as optical transmission coefficients are obtained 9549 1451 109 with a rigid rotator model coupled channel calculations. 9549 1451 110 Direct inelastic contributions were added incoherently 9549 1451 111 to Hauser-Feshbach calculated values of compound nucleus 9549 1451 112 inelastic scattering cross sections. 9549 1451 113 Total inelastic and continuum inelastic cross sections 9549 1451 114 were calculated with the same model and level density pa-9549 1451 115 rameterization approach, which allowed to reproduce 9549 1451 116 inelastic scattering data for 237-Np target nuclide by 9549 1451 117 Kornilov et al. /24/ for the excitation of specific 9549 1451 118 groups of continuum levels. Similar effect is envisioned 9549 1451 119 in case of Z-odd Am -241 target nuclide. 9549 1451 120 The sharp increase evidenced in the inelastic scattering 9549 1451 121 data /24/ was explained by the excitation of three quasi-9549 1451 122 particle states in residual nuclide Np-237 /25/. 9549 1451 123 The deformed optical potential adopted was that obtained 9549 1451 124 for 237-Np, /25/, then the beta_2 and beta_4 deformation 9549 1451 125 parameters were slightly varied to describe neutron 9549 1451 126 strength functions. Obtained partitioning of the total 9549 1451 127 cross section into reaction and scattering cross sections9549 1451 128 allows consistent description of fission capture cross 9549 1451 129 sections in 0.001-5 MeV energy 9549 1451 117 9549 1451 130 range within a statistical model. 9549 1451 131 Strength function So = 0.9009x10-4(EV)-1/2 at 0.25 keV. 9549 1451 132 9549 1451 133 VR=(45.722-0.334xE) MeV; RR =1.2600 fm; AR =.6300 fm; 9549 1451 134 WD=(3.690+0.400xE) MeV; E< 10 MeV RD =1.24 fm; 9549 1451 135 WD= 7.690 MeV; E>=10 MeV AD =.5200 fm; 9549 1451 136 VSO= 6.2 MeV; RS0=1.12 fm; ASO=.47 fm; 9549 1451 137 beta_2= .180; beta_4=.080; 9549 1451 138 9549 1451 139 9549 1451 140 Fission, capture and compound inelastic scattering cross 9549 1451 141 sections were calculated with Hauser-Feshbach-Moldauer /26/ 9549 1451 142 approach, at incident neutron energies higher than 0.266 MeV 9549 1451 143 Tepel et al. /27/ theory was employed. 9549 1451 144 9549 1451 145 243-Am level schema as taken from Nuclear Data Sheets: 9549 1451 146 9549 1451 147 NO. ENERGY(MEV) SPIN-PARITY 9549 1451 148 GS 0.0 5/2- 9549 1451 149 1 0.0422 7/2- 9549 1451 150 2 0.0840 5/2+ 9549 1451 151 3 0.0964 9/2- 9549 1451 152 4 0.1092 7/2+ 9549 1451 153 5 0.1435 9/2+ 9549 1451 154 6 0.1623 11/2- 9549 1451 155 7 0.1893 11/2+ 9549 1451 156 8 0.238 13/2- 9549 1451 157 9 0.244 13/2+ 9549 1451 158 10 0.266 15/2- 9549 1451 159 9549 1451 160 9549 1451 161 9549 1451 162 Continuum levels were assumed above 0.267 eV 9549 1451 163 9549 1451 164 MT=16,17,37. (n,2n), (n,3n) and (n,4n) cross sections from 9549 1451 165 statistical model calculations with account of 9549 1451 166 pre-equilibrium neutron emission (modified STAPRE code /28/ 9549 1451 167 was used). Pre-equilibrium neutron emission contribution was 9549 1451 168 fixed according to consistent description of (n,F) and (n,xn)9549 1451 169 reaction data for 238-U and 232-Th target nuclides /10/. 9549 1451 170 In case of 243-Am(n,2n) reaction experimental datum at 14 MeV9549 1451 171 by Gangarz, cited in/29/ on the absolute yield of short-lived9549 1451 172 ground state in the reaction Am(n,2n)242g-Am was reproduced. 9549 1451 173 Isomer ratio of the yields of 242m-Am and 243g-Am was model- 9549 1451 174 led from the (n,2n) threshold up to 20 MeV. The energy levels9549 1451 175 of 242-Am were adopted from /30/, based on residual 9549 1451 176 interaction of quasiparticle states in odd-odd nuclides. 9549 1451 177 9549 1451 178 MT=18, 19, 20, 21,38. Fission cross section is calculated 9549 1451 179 within statistical model /26,27,29,31/. 9549 1451 180 Measured fission data /13,19-22, 32-40/ analysis was accomp- 9549 1451 181 lished witin GMA /3/ approach. Statistical model calculations9549 1451 182 at energies of 1 keV-20 MeV are maintained which deviate from9549 1451 183 the GMA evaluation within GMA-estimated uncertainties 9549 1451 184 The contribution of emissive (n,nf) and (n,2nf) fission to 9549 1451 185 the total fission cross section was calculated using the 9549 1451 186 model /10/, which consistently describes Am(n,2n)242g-Am 9549 1451 187 data by Gangarz /29/. 9549 1451 188 9549 1451 189 MT=102 Capture cross section 9549 1451 190 Capture cross section is calculated within a statistical mo- 9549 1451 191 del. Above neutron energy 5.5 MeV capture cross section is 9549 1451 192 assumed to be 0.001 barn. Competition of (n,gf) and (n,gn') 9549 1451 193 reactions is taken into account. 9549 1451 194 Trend of measured data by Wisshak et al. /19/ and Weston and 9549 1451 195 Todd /23/ are reproduced. 9549 1451 196 9549 1451 197 9549 1451 198 MF=4 Angular distributions of secondary neutrons 9549 1451 199 for MT=2,51,53,56 - from coupled channel calculations 9549 1451 200 (rigid rotator model), with added isotropic compound contri- 9549 1451 201 bution. 9549 1451 202 9549 1451 203 MT=16, 17, 18-21, 38, 52, 54, 55, 57-60 and 91 are isotropic in 9549 1451 204 the lab. system 9549 1451 205 9549 1451 206 MF=5 Energy distributions of secondary neutrons 9549 1451 207 9549 1451 208 Energy distributions for MT=16,17,91 were calculated with 9549 1451 209 a Hauser-Feshbach statistical model of cascade neutron 9549 1451 210 emission, taking into account exclusive pre-fission (n,xnf) 9549 1451 211 and (n,xng) neutron spectra, with the allowance of pre-equi- 9549 1451 212 librium emission of the first neutron. The calculated spectra9549 1451 213 are strictly correlated with (n,F) and (n,xn) reaction cross 9549 1451 214 sections /10/. 9549 1451 215 9549 1451 216 9549 1451 217 MT=18,19,20,21,38 9549 1451 218 Prompt fission neutron spectra (PFNS) were calculated with 9549 1451 219 phenomenological model /10/, exclusive pre-fission 9549 1451 220 neutron spectra of(n,xnf) reactions, either equilibrium and 9549 1451 221 pre-equilibrium spectra of pre-fission (n,xnf) neutrons are 9549 1451 222 calculated with Hauser-Feshbach statistical model. PFNS from 9549 1451 223 fission fragments are calculated as a superposition of two 9549 1451 224 Watt distributions for heavy and light fission fragments 9549 1451 225 (FF), the partial contributions being equal, while the 9549 1451 226 temperatures different. FF kinetic energy is the superimposed9549 1451 227 phenomenological parameter, generally lower, than TKE. That 9549 1451 228 peculiarity roughly reflects its dependence on the moment of 9549 1451 229 prompt fission neutron emission /41/. 9549 1451 230 The model reproduces the shape of available PFNS data for 9549 1451 231 235-U, 238-U, 232-Th, 237-Np measured from thermal up to 18 9549 1451 232 MeV incident neutron energies /42/. 9549 1451 233 Average energies of PFNS predict distinct lowering in 9549 1451 234 the vicinity of (n,nf) and (n,2nf) reaction thresholds 9549 1451 235 in general compliance with fission chances contributions to 9549 1451 236 the observed fission cross section. 9549 1451 237 9549 1451 238 MF= 8 Radioactive nuclide production 9549 1451 239 MT=16 (n,2n) reaction cross section 9549 1451 240 Decay data were taken from ENSDF. 9549 1451 241 9549 1451 242 MT=102 Capture cross section 9549 1451 243 Decay data were taken from ENSDF. 9549 1451 244 9549 1451 245 MF=9 Multiplicities for Production of Radioactive Nuclides 9549 1451 246 MT=16 (n,2n) reaction cross section 9549 1451 247 9549 1451 248 The of yields of short-lived ground state(1-) and long-lived 9549 1451 249 (5-) excited (55 keV) states of 242-Am is modelled in the 9549 1451 250 same manner as it was done in case of 237-Np(n,2n)236-Np 9549 1451 251 reaction /25/. Levels of 242-Am are modelled using predicted 9549 1451 252 Gallaher-Moshkowski doublets by Sood /30/. Modelled energy- 9549 1451 253 depended ratio of the yields of short-lived ground state(1-) 9549 1451 254 and long-lived (5-) excited (55 keV) is consistent with 9549 1451 255 fission cross section data and data by Gangarz, cited in 9549 1451 256 /29/ on the absolute yield of short-lived ground-state in 9549 1451 257 reaction Am(n,2n)242g-Am at 14 MeV. 9549 1451 258 9549 1451 259 MT=102 Capture cross section 9549 1451 260 9549 1451 261 The of yields of short-lived excited (72.9 keV)state(1-) 9549 1451 262 244m-Am and long-lived (6-) ground state 244g-Am of 244-Am 9549 1451 263 is modelled in the same manner as it was done in case of 9549 1451 264 243-Am(n,2n)242-Am reaction, solving kinetic equation, 9549 1451 265 implemented in /28/. Levels of 244-Am are modelled using 9549 1451 266 predicted Gallaher-Moshkowski doublets by Sood /30/. Modelled9549 1451 267 energy dependend ratio of the yields of long-lived ground 9549 1451 268 state(6-) and short-lived (5-) excited (72.9 keV) is 9549 1451 269 consistent with the measurement by Marie et al. /43/ of the 9549 1451 270 ratio of the yields of 244g-Am and 244(m+g)Am at thermal 9549 1451 271 point. The 243-Am(n,g)244(m+g)Am reaction thermal cross- 9549 1451 272 section was measured in /43/ via the alpha-decay of 244-Cm, 9549 1451 273 distinguishing alpha-rays at 5804.8 keV and 5762.7 keV. 9549 1451 274 The partial 243-Am(n,g)244g-Am thermal cross-section was 9549 1451 275 determined by Marie et al. (2006) /43/ counting the 9549 1451 276 743.97 keV and 897.85 keV gamma-rays emitted after the decay 9549 1451 277 of 244g-Am(beta-)244-Cm. 9549 1451 278 9549 1451 279 References 9549 1451 280 1. Maslov V.M. et al., INDC(BLR)-006, 1996, IAEA, Vienna. 9549 1451 281 2. 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A 556, pp. 547-5559549 1451 410 (2006). 9549 1451 411 9549 1451 412 1 451 473 09549 1451 413 1 452 14 09549 1451 414 1 455 17 09549 1451 415 1 456 14 09549 1451 416 2 151 1054 09549 1451 417 3 1 29 09549 1451 418 3 2 26 09549 1451 419 3 4 28 09549 1451 420 3 16 10 09549 1451 421 3 17 7 09549 1451 422 3 18 26 09549 1451 423 3 19 26 09549 1451 424 3 20 10 09549 1451 425 3 21 8 09549 1451 426 3 37 4 09549 1451 427 3 38 6 09549 1451 428 3 51 28 09549 1451 429 3 52 20 09549 1451 430 3 53 25 09549 1451 431 3 54 19 09549 1451 432 3 55 17 09549 1451 433 3 56 22 09549 1451 434 3 57 15 09549 1451 435 3 58 14 09549 1451 436 3 59 14 09549 1451 437 3 60 13 09549 1451 438 3 91 19 09549 1451 439 3 102 20 09549 1451 440 4 2 121 09549 1451 441 4 16 2 09549 1451 442 4 17 2 09549 1451 443 4 18 2 09549 1451 444 4 19 2 09549 1451 445 4 20 2 09549 1451 446 4 21 2 09549 1451 447 4 37 2 09549 1451 448 4 38 2 09549 1451 449 4 51 121 09549 1451 450 4 52 2 09549 1451 451 4 53 118 09549 1451 452 4 54 2 09549 1451 453 4 55 2 09549 1451 454 4 56 115 09549 1451 455 4 57 2 09549 1451 456 4 58 2 09549 1451 457 4 59 2 09549 1451 458 4 60 2 09549 1451 459 4 91 2 09549 1451 460 5 16 1202 09549 1451 461 5 17 703 09549 1451 462 5 18 3447 09549 1451 463 5 19 3447 09549 1451 464 5 20 2118 09549 1451 465 5 21 1230 09549 1451 466 5 37 97 09549 1451 467 5 38 564 09549 1451 468 5 91 1193 09549 1451 469 8 16 3 09549 1451 470 8 102 11 09549 1451 471 9 16 18 09549 1451 472 9 102 39 09549 1451 473 9549 1 099999