IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 27059.0000 58.4269000 0 0 41 12725 1451 1 0.0 0.0 0 0 0 62725 1451 2 1.00000000 60000000.0 0 0 10 22725 1451 3 300.000000 0.0 1 0 506 142725 1451 4 27-Co- 59 FEI EVAL-Jan10 K.I.Zolotarev 2725 1451 5 DIST-Mar10 Rev.2 January 2010 2725 1451 6 ----BROND-2 MATERIAL 2725 2725 1451 7 -----INCIDENT NEUTRON DATA 2725 1451 8 ------ENDF-6 FORMAT 2725 1451 9 ******************************************************************2725 1451 10 IAEA, July 2011 (A. Trkov) 2725 1451 11 Extended cross sections and covariances from 20 to 60 MeV 2725 1451 12 by TENDL-2010, renormalised for continuity. 2725 1451 13 ******************************************************************2725 1451 14 ***************************************************************** 2725 1451 15 ------Russian Reactor Dosimetry File 2725 1451 16 ***************************************************************** 2725 1451 17 Author of evaluation: K.I.Zolotarev 2725 1451 18 ***************************************************************** 2725 1451 19 MF=3 2725 1451 20 MT= 16 -(n,2n) cross section 2725 1451 21 ------------------------------------- 2725 1451 22 Microscopic experimental data [1-45] were analyzed in the 2725 1451 23 process of preparation of input data base for the evaluation of 2725 1451 24 cross sections and their uncertainty for the Co-59(n,2n)Co-58 2725 1451 25 reaction. During this procedure all experimental data if it was 2725 1451 26 possible were corrected to the new standards. 2725 1451 27 Excitation function for the Co-59(n,2n)Co-58 reaction in the 2725 1451 28 energy region from threshold to 60 MeV was evaluated by means of 2725 1451 29 statistical analysis of experimental cross section data [1-33] 2725 1451 30 and data obtained from theoretical model calculation carried out 2725 1451 31 by means of GNASH code [46]. 2725 1451 32 Experimental cross section data [34-45] were rejected due to 2725 1451 33 inconsistency with the main bulk of experimental data. 2725 1451 34 Statistical analysis of input cross section data was carried 2725 1451 35 out by means of PADE-2 code [47]. Rational function was used as 2725 1451 36 the model function [48]. 2725 1451 37 Evaluated excitation function of the Co-59(n,2n)Co-58 reaction 2725 1451 38 was tested by comparison with integral experimental data [49] for 2725 1451 39 U-235 thermal fission and Cf-252 spontaneous fission neutron spe- 2725 1451 40 ctra. Data for spectra were taken from Refs. [50] and [51], res- 2725 1451 41 pectively. The results of test are given in the Table 1. 2725 1451 42 Table 1 2725 1451 43 ----------------------+-----------------+----------------------- 2725 1451 44 Type of spectrum | ,mb (calc.) | , mb (measured) 2725 1451 45 ----------------------+-----------------+----------------------- 2725 1451 46 U-235 neutron fission | 0.19067 | 0.2028 +- 0.0800 [49] 2725 1451 47 ----------------------+-----------------+----------------------- 2725 1451 48 Cf-252 spont. fission | 0.40799 | 0.4051 +- 0.0102 [49] 2725 1451 49 ----------------------+-----------------+----------------------- 2725 1451 50 2725 1451 51 MT= 17 -(n,3n) cross section 2725 1451 52 ------------------------------------- 2725 1451 53 Microscopic cross sections for the Co-59(n,3n)Co-57 reaction 2725 1451 54 are given only in a three works [1-3]. Experimental data of Uno 2725 1451 55 et al. [2] obtained in measurements with Li-7(p,n)Be-7 neutron 2725 1451 56 source were renormalized to factor Fc= 1.07378, which was deter- 2725 1451 57 mined from analysis of Co-59(n,3n)Co-57 data of these authors. 2725 1451 58 The database used to evaluate the Co-59(n,3n)Co-57 reaction 2725 1451 59 cross section from threshold to 85 MeV was assembled from micro- 2725 1451 60 scopic experimental data [1-3] and data from theoretical modeling 2725 1451 61 calculation carried out by means of EMPIRE-2.19 code [4]. 2725 1451 62 Statistical analysis of input database was performed by means 2725 1451 63 of PADE-2 code [5]. Rational function was used as the model func- 2725 1451 64 tion [6]. 2725 1451 65 Integral experimental data of Qaim et al [7] obtained for the 2725 1451 66 neutron spectrum from Be(d,n) reaction can't be used for testing 2725 1451 67 the evaluated Co-59(n,3n)Co-57 reaction excitation function due 2725 1451 68 to absence of well determined data for neutron spectrum. 2725 1451 69 2725 1451 70 MT=102 - (n,g) cross section 2725 1451 71 ---------------------------- 2725 1451 72 The cross section is taken over from IRDF2002, which in turn 2725 1451 73 adopts the data from ENDF/B-VI Release 2 by A. Smith, D. Smith, 2725 1451 74 P. Guenther, J. Meadows, R. Lawson(ANL), R. Howerton(LLNL), 2725 1451 75 M. Sugimoto(JAERI). 2725 1451 76 The Resonance Parameters Were Evaluated By G.Desaussure, 2725 1451 77 N.M.Larson, J.A.Harvey, N.W.Hill, Ann. Nucl. Energy 19, 393 (1992)2725 1451 78 2725 1451 79 MT=103 -(n,p) cross section 2725 1451 80 ------------------------------------- 2725 1451 81 Microscopic experimental data [1-34] were analyzed in the 2725 1451 82 process of preparation of input data base for the evaluation of 2725 1451 83 cross sections and their uncertainty for the Co-59(n,p)Fe-59 2725 1451 84 reaction. During this procedure all experimental data if it was 2725 1451 85 possible were corrected to the new standards. 2725 1451 86 Excitation function for the Co-59(n,p)Fe-59 reaction in the 2725 1451 87 energy region from threshold to 60 MeV was evaluated by means of 2725 1451 88 statistical analysis of experimental cross section data [1-24] 2725 1451 89 and data obtained from theoretical model calculation carried out 2725 1451 90 by means of GNASH code [35]. 2725 1451 91 Experimental cross section data [24-34] were rejected due to 2725 1451 92 inconsistency with the main bulk of experimental data. 2725 1451 93 Statistical analysis of input cross section data was carried 2725 1451 94 out by means of PADE-2 code [36]. Rational function was used as 2725 1451 95 the model function [37]. 2725 1451 96 Evaluated excitation function of the Co-59(n,p)Fe-59 reaction 2725 1451 97 was tested by comparison with integral experimental data [38] for 2725 1451 98 U-235 thermal fission and Cf-252 spontaneous fission neutron spe- 2725 1451 99 ctra. Data for spectra were taken from Refs. [39] and [40], res- 2725 1451 100 pectively. The results of test are given in the Table 2. 2725 1451 101 Table 2 2725 1451 102 ----------------------+-----------------+----------------------- 2725 1451 103 Type of spectrum | ,mb (calc.) | , mb (measured) 2725 1451 104 ----------------------+-----------------+----------------------- 2725 1451 105 U-235 neutron fission | 1.4163 | 1.396 +- 0.033 [38] 2725 1451 106 ----------------------+-----------------+----------------------- 2725 1451 107 Cf-252 spont. fission | 1.7155 | 1.690 +- 0.042 [38] 2725 1451 108 ----------------------+-----------------+----------------------- 2725 1451 109 2725 1451 110 MT=107 -(n,a) cross section 2725 1451 111 --------------------------- 2725 1451 112 Excitation function for the Co-59(n,a)Mn-56 reaction is taken 2725 1451 113 from IRDF-2002. In the energy region from 2.5 MeV to 21 MeV it 2725 1451 114 was evaluated by means ofstatistical analysis of experimental 2725 1451 115 cross section data [1-28],and data from GNASH [29] calculation. 2725 1451 116 All experimental data were renormalized to the new standards 2725 1451 117 for monitor reactions cross sections and decay data. Uncertainty 2725 1451 118 in the monitor reaction cross section was added to the total 2725 1451 119 uncertainty for Santry and Butler data [5]. Cross section data 2725 1451 120 measured by Huang Jianzhou et al. [18] were renormalized to the 2725 1451 121 preliminary evaluated integral of excitation function in the 2725 1451 122 energy range 13.0 - 15.0 MeV. Correction factor for this data 2725 1451 123 was equal Fc= 1.0507 . 2725 1451 124 Experimental data from ref.[30-41] were rejected due to their 2725 1451 125 discrepancy with the main bulk of experimental data [1-28] and 2725 1451 126 data from theoretical model calculation. 2725 1451 127 The final procedure of evaluation Co-59(n,a)Mn-56 excitation 2725 1451 128 function from threshold to 21 MeV has been carried out within the 2725 1451 129 framework of generalized least squares method. Rational function 2725 1451 130 was used as model function [42]. Calculations was performed by 2725 1451 131 means of Pade-2 code [43]. 2725 1451 132 The evaluated Co-59(n,a)Mn-56 excitation function averaged 2725 1451 133 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 2725 1451 134 fission neutron spectrum [45] gives the next values : 2725 1451 135 ---------------------------------------------------------------- 2725 1451 136 TYPE OF SPECTRUM I ,mb (calc.) I , mb (measured) 2725 1451 137 ---------------------------------------------------------------- 2725 1451 138 I I 0.1500 +- 0.0080 [46] 2725 1451 139 U-235 neutron fission I 0.15823 I 0.1700 +- 0.0120 [47] 2725 1451 140 I I 0.1568 +- 0.0035 [48] 2725 1451 141 ---------------------------------------------------------------- 2725 1451 142 I I 0.2208 +- 0.0014 [49] 2725 1451 143 Cf-252 spont. fission I 0.22095 I 0.2221 +- 0.0023 [50] 2725 1451 144 I I 0.2218 +- 0.0041 [51] 2725 1451 145 ---------------------------------------------------------------- 2725 1451 146 2725 1451 147 2725 1451 148 MT=33 2725 1451 149 MT= 16 -(n,2n) cross section cov. matrix 2725 1451 150 ------------------------------------------- 2725 1451 151 Uncertainties in the evaluated excitation function for the 2725 1451 152 reaction Co-59(n,2n)Co-58 are given in the form of relative cova- 2725 1451 153 riance matrix for the 48-neutron energy groups (LB=5). Covariance 2725 1451 154 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 155 mended cross section data by means of PADE-2 code. 2725 1451 156 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 157 given in the 33-file are the following: 2725 1451 158 2725 1451 159 1.17519E-05 1.18901E-05 1.20954E-05 1.23231E-05 2725 1451 160 1.25599E-05 1.28152E-05 1.31251E-05 1.35366E-05 2725 1451 161 1.41638E-05 1.53005E-05 1.70968E-05 1.92698E-05 2725 1451 162 2.37801E-05 2.92459E-05 3.75758E-05 4.54525E-05 2725 1451 163 5.84343E-05 7.15999E-05 8.22985E-05 9.41809E-05 2725 1451 164 1.08914E-04 1.24863E-04 1.41512E-04 1.58718E-04 2725 1451 165 1.76290E-04 1.93113E-04 2.04246E-04 2.17762E-04 2725 1451 166 2.36636E-04 2.56933E-04 2.77951E-04 2.99381E-04 2725 1451 167 3.31092E-04 3.69081E-04 4.06286E-04 4.54022E-04 2725 1451 168 5.01935E-04 5.11280E-04 5.73614E-04 6.49448E-04 2725 1451 169 7.36019E-04 8.31088E-04 1.42913E-03 2.23904E-03 2725 1451 170 7.86541E-03 2.03901E-02 9.83103E-02 9.97381E-01 2725 1451 171 2725 1451 172 MT= 17 -(n,3n) cross section cov. matrix 2725 1451 173 ------------------------------------------- 2725 1451 174 Uncertainties in the evaluated excitation function for the 2725 1451 175 reaction Co-59(n,3n)Co-57 are given in the form of relative cova- 2725 1451 176 riance matrix for the 39-neutron energy groups (LB=5). Covariance 2725 1451 177 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 178 mended cross section data by means of PADE-2 code. 2725 1451 179 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 180 given in the 33-file are the following: 2725 1451 181 2725 1451 182 8.33452E-06 8.58035E-06 9.03318E-06 9.59768E-06 2725 1451 183 1.04456E-05 1.14678E-05 1.25643E-05 1.41888E-05 2725 1451 184 1.59404E-05 1.76322E-05 2.01039E-05 2.31221E-05 2725 1451 185 2.72359E-05 3.26333E-05 4.29541E-05 5.07083E-05 2725 1451 186 6.20570E-05 7.99376E-05 9.68474E-05 1.08456E-04 2725 1451 187 1.31791E-04 1.60861E-04 1.93303E-04 2.25159E-04 2725 1451 188 2.43294E-04 2.75947E-04 3.19283E-04 3.67215E-04 2725 1451 189 4.19165E-04 4.75056E-04 5.34932E-04 5.98807E-04 2725 1451 190 6.66521E-04 7.14806E-04 3.72924E-03 1.99128E-02 2725 1451 191 2.88573E-02 1.09197E-01 1.94886E-01 2725 1451 192 2725 1451 193 MT=102 -(n,g) cross section cov. matrix 2725 1451 194 ------------------------------------------- 2725 1451 195 The covariance matrix is taken over from IRDF90/V2 2725 1451 196 2725 1451 197 MT=103 -(n,p) cross section cov. matrix 2725 1451 198 ------------------------------------------- 2725 1451 199 Uncertainties in the evaluated excitation function for the 2725 1451 200 reaction Co-59(n,p)Fe-59 are given in the form of relative cova- 2725 1451 201 riance matrix for the 49-neutron energy groups (LB=5). Covariance 2725 1451 202 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 203 mended cross section data by means of PADE-2 code. 2725 1451 204 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 205 given in the 33-file are the following: 2725 1451 206 2725 1451 207 4.13473E-07 4.27626E-07 4.51208E-07 4.82385E-07 2725 1451 208 5.24529E-07 5.68750E-07 6.26389E-07 6.81740E-07 2725 1451 209 7.43528E-07 8.13479E-07 8.73417E-07 9.37226E-07 2725 1451 210 1.00863E-06 1.07570E-06 1.13454E-06 1.19653E-06 2725 1451 211 1.26664E-06 1.33965E-06 1.46617E-06 1.65554E-06 2725 1451 212 1.94736E-06 2.31852E-06 2.67098E-06 3.22483E-06 2725 1451 213 3.77462E-06 4.41698E-06 5.55287E-06 6.09540E-06 2725 1451 214 7.44285E-06 9.25147E-06 1.07387E-05 1.17943E-05 2725 1451 215 1.41203E-05 1.69054E-05 2.06437E-05 2.63415E-05 2725 1451 216 2.88626E-05 1.17262E-04 9.10324E-04 1.40176E-03 2725 1451 217 2.23894E-03 3.18553E-03 5.21607E-03 9.11440E-03 2725 1451 218 1.43199E-02 1.78563E-02 3.16668E-02 1.63229E-01 2725 1451 219 8.13509E-01 2725 1451 220 2725 1451 221 MT=107 -(n,a) cross section cov. matrix 2725 1451 222 ------------------------------------------- 2725 1451 223 Uncertainties in the evaluated excitation function for the 2725 1451 224 reaction Co-59(n,a)Mn-56 are given in the form of relative cova- 2725 1451 225 riance matrix for the 32-neutron energy groups (LB=5). Covariance 2725 1451 226 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 227 mended cross section data by means of PADE-2 code. 2725 1451 228 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 229 given in the 33-file are the following: 2725 1451 230 2725 1451 231 1.11275E-08 1.13832E-08 1.19424E-08 1.27401E-08 2725 1451 232 1.37409E-08 1.52188E-08 1.67720E-08 1.86812E-08 2725 1451 233 2.09005E-08 2.38172E-08 2.69210E-08 3.07595E-08 2725 1451 234 3.46853E-08 4.07806E-08 4.51537E-08 5.55085E-08 2725 1451 235 6.19014E-08 7.83507E-08 9.39864E-08 1.01329E-07 2725 1451 236 2.19793E-07 8.71735E-07 1.71048E-05 1.93185E-04 2725 1451 237 1.05065E-03 1.19481E-03 2.01786E-03 2.53733E-03 2725 1451 238 3.28491E-03 1.06037E-02 2.32281E-02 9.30559E-01 2725 1451 239 2725 1451 240 2725 1451 241 References for the Co-59(n,2n)Co-58 reaction 2725 1451 242 1. B.D.Kern et al., Progress Report A-KTY-59/60, p. 3, Feb. 1960 2725 1451 243 2. E.Weigold, Australian J. 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W.Mannhart Validation of Differential Cross Sections with 2725 1451 500 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2725 1451 501 September 2002 2725 1451 502 2725 1451 503 ***************************************************************** 2725 1451 504 ***************** Program LINEAR (VERSION 2012-1) ***************2725 1451 505 For All Data Greater than 1.0000D-10 barns in Absolute Value 2725 1451 506 Data Linearized to Within an Accuracy of .100000000 per-cent 2725 1451 507 ***************** Program GROUPIE (VERSION 2012-1) **************2725 1451 508 Unshielded Group Averages Using 640 Groups 2725 1451 509 Weighting Spectrum: Flat (Constant) Spectrum 2725 1451 510 1 451 518 12725 1451 511 2 151 4 12725 1451 512 3 1 10378 12725 1451 513 3 2 10257 12725 1451 514 3 16 106 12725 1451 515 3 17 99 12725 1451 516 3 102 7386 12725 1451 517 3 103 130 12725 1451 518 3 107 37 12725 1451 519 33 16 216 12725 1451 520 33 17 147 12725 1451 521 33 102 18 12725 1451 522 33 103 224 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