IRDF-ext 777 0 0 0 3.006400+4 6.338000+1 0 0 41 13025 1451 1 0.000000+0 0.000000+0 0 0 0 63025 1451 2 1.000000+0 2.000000+7 0 0 10 23025 1451 3 0.000000+0 0.000000+0 0 0 248 33025 1451 4 30-Zn- 64 FEI EVAL-Mar06 K.I.Zolotarev 3025 1451 5 DIST-Jan07 BEST-13 33223-15 3025 1451 6 ----BROND-2 MATERIAL 3025 3025 1451 7 -----INCIDENT NEUTRON DATA 3025 1451 8 ------ENDF-6 FORMAT 3025 1451 9 ***************************************************************** 3025 1451 10 * Extension to the International Reactor Dosimetry Library * 3025 1451 11 * supported partially by the International Atomic Energy Agency * 3025 1451 12 * through IAEA research contract 13335. * 3025 1451 13 * Published as a technical report INDC(NDS)-0526 (2008). * 3025 1451 14 * Available online at * 3025 1451 15 * http://www-nds.iaea.org/reports-new/indc-reports/indc-nds/ * 3025 1451 16 * indc-nds-0526.pdf * 3025 1451 17 ***************************************************************** 3025 1451 18 ------Russian Reactor Dosimetry File RRDF-2006 3025 1451 19 ***************************************************************** 3025 1451 20 Author of evaluation: K.I.Zolotarev 3025 1451 21 ***************************************************************** 3025 1451 22 MF=3 3025 1451 23 MT=103 - (n,p) cross section 3025 1451 24 ------------------------------------- 3025 1451 25 Microscopic experimental data [1-46] were analyzed in the 3025 1451 26 process of preparation of input data base for the evaluation of 3025 1451 27 cross sections and their uncertainty for the Zn-64(n,p)Cu-64 3025 1451 28 reaction. During this procedure all experimental data if it was 3025 1451 29 possible were corrected to the new recommended cross section data 3025 1451 30 for monitor reactions used in the measurements and to the new re- 3025 1451 31 commended decay data from Refs. [47] and [48]. 3025 1451 32 Excitation function for the Zn-64(n,p)Cu-64 reaction in the 3025 1451 33 energy region from threshold to 20 MeV was evaluated by means of 3025 1451 34 statistical analysis of experimental cross section data [1-32]. 3025 1451 35 Special correction was done with experimental data [6], [11], 3025 1451 36 [12], [17], [24], [28], [30] and [31]. 3025 1451 37 The results of relative measurements of Paulsen and Liskien 3025 1451 38 for the incident neutron energies 0.99 - 2.21 MeV [6] were norma- 3025 1451 39 lized to the absolute cross section value 13.416 mb at 2.20 MeV 3025 1451 40 determined from Smith and Meadows measurements [12] carried out 3025 1451 41 with Li-7(p,n)Be-7 neutron source. Microscopic cross sections 3025 1451 42 measured by Smith and Meadows for Zn-64(n,p)Cu-64 reaction in the 3025 1451 43 energy interval 1.159-5.576 MeV agree well with experimental data 3025 1451 44 of Ikeda et al. [21] and integral experimental data for Cf-252 3025 1451 45 spontaneous fission neutron spectrum. 3025 1451 46 Experimental data of Santry and Butler [11] obtained in the 3025 1451 47 energy range 1.50 - 5.33 MeV and cross section data of King et al.3025 1451 48 at 2.12 - 4.84 MeV [17] were also renormalized to the results of 3025 1451 49 Smith and Meadows measurements with Li-7(p,n)Be-7 neutron source. 3025 1451 50 Correction factors applied to the experimental data [11] and [17] 3025 1451 51 were equal Fc= 0.91582 and Fc= 1.08431, respectively. Correction 3025 1451 52 factor Fc= 0.91582 was used for all experimental data of Santry 3025 1451 53 and Butler given in the Ref. [11]. 3025 1451 54 Data of Smith and Meadows [12] measured with using neutrons 3025 1451 55 from D(d,n)He3 reaction were renormalized to the cross section of 3025 1451 56 183.6 mb at 5.384 MeV obtained from the results of measurements 3025 1451 57 carried out with Li-7(p,n)Be-7 neutron source. After correction 3025 1451 58 to the new standard cross sections for U238(n,f) monitor reaction 3025 1451 59 [49] and recommended yield for Cu-64 annihilation gammas from Ref.3025 1451 60 [48] the D(d,n)He3 data of Smith and Meadows in the energy range 3025 1451 61 5.384 - 9.939 MeV were increased to the factor Fc= 1.15386 . 3025 1451 62 In the neutron energy range between 8.4-14.3 MeV more repre- 3025 1451 63 sentative are the results of new precise measurements carried out 3025 1451 64 by Mannhart and Schmidt [32]. Experimental data [2], [16], [25], 3025 1451 65 [31] and renormalized to the factor Fc= 1.15386 data of Santry 3025 1451 66 and Butler [11] are agree well with new measurements of Mannhart 3025 1451 67 and Schmidt [32]. 3025 1451 68 Cross section data of Viennot et al. [24], Molla et al. [29], 3025 1451 69 Kielan and Marcinkowski [30] were renormalized to the integral of 3025 1451 70 cross sections calculated from of experimental data of Mannhart 3025 1451 71 and Schmidt [32] in the overlapping energy ranges. Experimental 3025 1451 72 data of Ghorai et al. [28] were renormalized to preliminary evalu-3025 1451 73 ated integral of cross sections in the energy interval from 14.2 3025 1451 74 to 16.2 MeV. After corrections to the new standards experimental 3025 1451 75 data [24], [28], [29] and [30] were multiplied to the factors: 3025 1451 76 Fc= 0.81195, Fc= 0.84351, Fc= 0.75385, Fc= 0.93987, respectively. 3025 1451 77 Data of Huang Xiaolong et al. [31] measured in the energy 3025 1451 78 range 14.65 - 19.02 MeV by means of T(d,n)He4 neutron source were 3025 1451 79 renormalized to the cross section value of 152.9 mb (+-2.4%) at 3025 1451 80 14.65 MeV, evaluated from experimental data [4], [7], [8], [14] 3025 1451 81 and [27]. 3025 1451 82 Experimental data [34], [36], [38] and [46] were rejected due 3025 1451 83 to systematical and significant underestimation of cross sections 3025 1451 84 above 2.8 MeV. The results of Bormann and Lammers [9] measure- 3025 1451 85 ments obtained in the energy range 14.10-18.19 MeV were not taken 3025 1451 86 into account from 15.46 MeV to 18.19 MeV due to systematical over-3025 1451 87 estimation of cross sections. For the same reason data of Ghorai 3025 1451 88 et al. [28] at 17.2 MeV and Kielan and Marcinkowski [30] at 15.9 3025 1451 89 and 16.6 MeV were rejected also. 3025 1451 90 Cross section data given in the Refs. [33-46] were rejected 3025 1451 91 completely due to their big discrepancy with the main bulk of ex- 3025 1451 92 perimental data. In the rejected experimental data [33], [35], 3025 1451 93 [39], [41-45] cross section values were measured only at a one 3025 1451 94 energy point in the interval 14 - 15 MeV. 3025 1451 95 Statistical analysis of input cross section data was carried 3025 1451 96 out by means of PADE-2 code [50]. Rational function was used as 3025 1451 97 the model function [51]. 3025 1451 98 Evaluated excitation function for the reaction Zn64(n,p)Cu64 3025 1451 99 was tested with using integral experimental data [52-53] for 3025 1451 100 U-235 thermal fission neutron spectrum and integral experimental 3025 1451 101 data [54-57] for Cf-252 spontaneous fission neutron spectrum. 3025 1451 102 Calculated and measured average cross section values for U-235 3025 1451 103 thermal fission neutron spectrum [58] and Cf-252 spontaneous 3025 1451 104 fission neutron spectrum [59] are given in the table 1. 3025 1451 105 Table 1 3025 1451 106 ================================================================= 3025 1451 107 TYPE OF SPECTRUM ,mb (calc.) , mb (measured) 3025 1451 108 ----------------------------------------------------------------- 3025 1451 109 U-235 neutron fission 38.914 38.89 +- 2.82 [ *] 3025 1451 110 35.39 +- 1.07 [57] 3025 1451 111 ----------------------------------------------------------------- 3025 1451 112 CF-252 spont. fission 42.718 42.34 +- 0.94 [**] 3025 1451 113 40.59 +- 0.67 [57] 3025 1451 114 ================================================================= 3025 1451 115 * - averaged value obtained from experimental data [52-53] 3025 1451 116 ** - averaged value obtained from experimental data [54-56] 3025 1451 117 3025 1451 118 MF=33 3025 1451 119 MT=103 - (n,p) cross section cov. matrix 3025 1451 120 ---------------------------------------- 3025 1451 121 Uncertainties in the evaluated excitation function for the 3025 1451 122 reaction Zn-64(n,p)Cu-64 are given in the form of relative covari-3025 1451 123 ance matrix for the 49-neutron energy groups (LB=5). Covariance 3025 1451 124 matrix of uncertainties was calculated simultaneously with 3025 1451 125 recommended cross section data by means of PADE-2 code [50]. 3025 1451 126 Eigenvalues of the 6-th digits relative covariance matrix 3025 1451 127 given in the 33-file are the following: 3025 1451 128 3025 1451 129 5.50784E-08 5.71733E-08 5.94423E-08 6.15434E-08 3025 1451 130 6.41589E-08 6.67452E-08 6.87802E-08 7.09362E-08 3025 1451 131 7.33156E-08 7.62346E-08 7.97587E-08 8.39517E-08 3025 1451 132 8.83311E-08 9.24351E-08 9.75569E-08 1.04183E-07 3025 1451 133 1.11626E-07 1.18053E-07 1.24390E-07 1.34278E-07 3025 1451 134 1.46587E-07 1.56514E-07 1.69670E-07 1.96551E-07 3025 1451 135 2.10571E-07 2.58348E-07 2.86453E-07 3.37288E-07 3025 1451 136 4.18775E-07 5.04883E-07 5.48409E-07 6.51679E-07 3025 1451 137 7.95195E-07 9.69913E-07 1.51323E-06 6.87735E-06 3025 1451 138 4.70011E-05 4.54712E-04 8.44040E-04 1.00266E-03 3025 1451 139 1.39839E-03 1.60043E-03 2.35419E-03 2.47891E-03 3025 1451 140 5.19930E-03 6.20593E-03 6.78721E-03 2.04942E-02 3025 1451 141 3.28605E-02 3025 1451 142 3025 1451 143 References : 3025 1451 144 1. 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W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 3025 1451 250 ***************************************************************** 3025 1451 251 3025 1451 252 1 451 255 13025 1451 253 3 103 68 13025 1451 254 33 103 224 13025 1451 255 0.000000+0 0.000000+0 0 0 0 03025 1 099999 0.000000+0 0.000000+0 0 0 0 03025 0 0 0 3.006400+4 6.338000+1 0 0 0 03025 3103 1 2.035000+5 2.035000+5 0 0 1 1943025 3103 2 194 2 3025 3103 3 1.000000-5 0.000000+0 5.000000+5 0.000000+0 9.000000+5 1.474650-53025 3103 4 1.000000+6 6.765810-5 1.100000+6 1.301490-4 1.200000+6 2.182080-43025 3103 5 1.300000+6 3.542330-4 1.400000+6 5.690330-4 1.500000+6 9.042170-43025 3103 6 1.600000+6 1.415020-3 1.700000+6 2.173390-3 1.800000+6 3.270910-33025 3103 7 1.900000+6 4.821060-3 2.000000+6 6.959280-3 2.100000+6 9.839400-33025 3103 8 2.200000+6 1.362410-2 2.300000+6 1.846740-2 2.400000+6 2.448800-23025 3103 9 2.500000+6 3.173720-2 2.600000+6 4.016670-2 2.700000+6 4.960850-23025 3103 10 2.800000+6 5.977880-2 2.900000+6 7.031060-2 3.000000+6 8.080820-23025 3103 11 3.100000+6 9.090900-2 3.200000+6 1.003320-1 3.300000+6 1.089020-13025 3103 12 3.400000+6 1.165460-1 3.500000+6 1.232770-1 3.600000+6 1.291650-13025 3103 13 3.700000+6 1.343150-1 3.800000+6 1.388390-1 3.900000+6 1.428530-13025 3103 14 4.000000+6 1.464580-1 4.100000+6 1.497450-1 4.200000+6 1.527880-13025 3103 15 4.300000+6 1.556500-1 4.400000+6 1.583790-1 4.500000+6 1.610150-13025 3103 16 4.600000+6 1.635850-1 4.700000+6 1.661140-1 4.800000+6 1.686170-13025 3103 17 4.900000+6 1.711050-1 5.000000+6 1.735880-1 5.100000+6 1.760690-13025 3103 18 5.200000+6 1.785510-1 5.300000+6 1.810340-1 5.400000+6 1.835170-13025 3103 19 5.500000+6 1.859980-1 5.600000+6 1.884740-1 5.700000+6 1.909400-13025 3103 20 5.800000+6 1.933920-1 5.900000+6 1.958250-1 6.000000+6 1.982340-13025 3103 21 6.100000+6 2.006130-1 6.200000+6 2.029580-1 6.300000+6 2.052620-13025 3103 22 6.400000+6 2.075220-1 6.500000+6 2.097310-1 6.600000+6 2.118860-13025 3103 23 6.700000+6 2.139820-1 6.800000+6 2.160160-1 6.900000+6 2.179840-13025 3103 24 7.000000+6 2.198840-1 7.100000+6 2.217140-1 7.200000+6 2.234710-13025 3103 25 7.300000+6 2.251550-1 7.400000+6 2.267660-1 7.500000+6 2.283040-13025 3103 26 7.600000+6 2.297700-1 7.700000+6 2.311660-1 7.799990+6 2.324940-13025 3103 27 7.899990+6 2.337570-1 8.000000+6 2.349580-1 8.099990+6 2.361020-13025 3103 28 8.200000+6 2.371930-1 8.300000+6 2.382350-1 8.400000+6 2.392350-13025 3103 29 8.500000+6 2.401970-1 8.600000+6 2.411270-1 8.700000+6 2.420300-13025 3103 30 8.800000+6 2.429120-1 8.900000+6 2.437800-1 9.000000+6 2.446370-13025 3103 31 9.100000+6 2.454880-1 9.200000+6 2.463390-1 9.300000+6 2.471920-13025 3103 32 9.400000+6 2.480500-1 9.500000+6 2.489150-1 9.600000+6 2.497870-13025 3103 33 9.700000+6 2.506670-1 9.800000+6 2.515510-1 9.900000+6 2.524370-13025 3103 34 1.000000+7 2.533200-1 1.010000+7 2.541920-1 1.020000+7 2.550460-13025 3103 35 1.030000+7 2.558710-1 1.040000+7 2.566540-1 1.050000+7 2.573820-13025 3103 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