IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 30064.0000 63.3800000 0 0 41 13025 1451 1 0.0 0.0 0 0 0 63025 1451 2 1.00000000 60000000.0 0 0 10 23025 1451 3 300.000000 0.0 1 0 258 43025 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 IAEA, June 2012 (A. Trkov) 3025 1451 11 Extended cross sections and covariances from 20 to 60 MeV 3025 1451 12 by TENDL-2011, renormalised for continuity. 3025 1451 13 ******************************************************************3025 1451 14 *** MF2 added *************************************************** 3025 1451 15 * Extension to the International Reactor Dosimetry Library * 3025 1451 16 * supported partially by the International Atomic Energy Agency * 3025 1451 17 * through IAEA research contract 13335. * 3025 1451 18 * Published as a technical report INDC(NDS)-0526 (2008). * 3025 1451 19 * Available online at * 3025 1451 20 * http://www-nds.iaea.org/reports-new/indc-reports/indc-nds/ * 3025 1451 21 * indc-nds-0526.pdf * 3025 1451 22 ***************************************************************** 3025 1451 23 ------Russian Reactor Dosimetry File RRDF-2006 3025 1451 24 ***************************************************************** 3025 1451 25 Author of evaluation: K.I.Zolotarev 3025 1451 26 ***************************************************************** 3025 1451 27 MF=3 3025 1451 28 MT=103 - (n,p) cross section 3025 1451 29 ------------------------------------- 3025 1451 30 Microscopic experimental data [1-46] were analyzed in the 3025 1451 31 process of preparation of input data base for the evaluation of 3025 1451 32 cross sections and their uncertainty for the Zn-64(n,p)Cu-64 3025 1451 33 reaction. During this procedure all experimental data if it was 3025 1451 34 possible were corrected to the new recommended cross section data 3025 1451 35 for monitor reactions used in the measurements and to the new re- 3025 1451 36 commended decay data from Refs. [47] and [48]. 3025 1451 37 Excitation function for the Zn-64(n,p)Cu-64 reaction in the 3025 1451 38 energy region from threshold to 20 MeV was evaluated by means of 3025 1451 39 statistical analysis of experimental cross section data [1-32]. 3025 1451 40 Special correction was done with experimental data [6], [11], 3025 1451 41 [12], [17], [24], [28], [30] and [31]. 3025 1451 42 The results of relative measurements of Paulsen and Liskien 3025 1451 43 for the incident neutron energies 0.99 - 2.21 MeV [6] were norma- 3025 1451 44 lized to the absolute cross section value 13.416 mb at 2.20 MeV 3025 1451 45 determined from Smith and Meadows measurements [12] carried out 3025 1451 46 with Li-7(p,n)Be-7 neutron source. Microscopic cross sections 3025 1451 47 measured by Smith and Meadows for Zn-64(n,p)Cu-64 reaction in the 3025 1451 48 energy interval 1.159-5.576 MeV agree well with experimental data 3025 1451 49 of Ikeda et al. [21] and integral experimental data for Cf-252 3025 1451 50 spontaneous fission neutron spectrum. 3025 1451 51 Experimental data of Santry and Butler [11] obtained in the 3025 1451 52 energy range 1.50 - 5.33 MeV and cross section data of King et al.3025 1451 53 at 2.12 - 4.84 MeV [17] were also renormalized to the results of 3025 1451 54 Smith and Meadows measurements with Li-7(p,n)Be-7 neutron source. 3025 1451 55 Correction factors applied to the experimental data [11] and [17] 3025 1451 56 were equal Fc= 0.91582 and Fc= 1.08431, respectively. Correction 3025 1451 57 factor Fc= 0.91582 was used for all experimental data of Santry 3025 1451 58 and Butler given in the Ref. [11]. 3025 1451 59 Data of Smith and Meadows [12] measured with using neutrons 3025 1451 60 from D(d,n)He3 reaction were renormalized to the cross section of 3025 1451 61 183.6 mb at 5.384 MeV obtained from the results of measurements 3025 1451 62 carried out with Li-7(p,n)Be-7 neutron source. After correction 3025 1451 63 to the new standard cross sections for U238(n,f) monitor reaction 3025 1451 64 [49] and recommended yield for Cu-64 annihilation gammas from Ref.3025 1451 65 [48] the D(d,n)He3 data of Smith and Meadows in the energy range 3025 1451 66 5.384 - 9.939 MeV were increased to the factor Fc= 1.15386 . 3025 1451 67 In the neutron energy range between 8.4-14.3 MeV more repre- 3025 1451 68 sentative are the results of new precise measurements carried out 3025 1451 69 by Mannhart and Schmidt [32]. Experimental data [2], [16], [25], 3025 1451 70 [31] and renormalized to the factor Fc= 1.15386 data of Santry 3025 1451 71 and Butler [11] are agree well with new measurements of Mannhart 3025 1451 72 and Schmidt [32]. 3025 1451 73 Cross section data of Viennot et al. [24], Molla et al. [29], 3025 1451 74 Kielan and Marcinkowski [30] were renormalized to the integral of 3025 1451 75 cross sections calculated from of experimental data of Mannhart 3025 1451 76 and Schmidt [32] in the overlapping energy ranges. Experimental 3025 1451 77 data of Ghorai et al. [28] were renormalized to preliminary evalu-3025 1451 78 ated integral of cross sections in the energy interval from 14.2 3025 1451 79 to 16.2 MeV. After corrections to the new standards experimental 3025 1451 80 data [24], [28], [29] and [30] were multiplied to the factors: 3025 1451 81 Fc= 0.81195, Fc= 0.84351, Fc= 0.75385, Fc= 0.93987, respectively. 3025 1451 82 Data of Huang Xiaolong et al. [31] measured in the energy 3025 1451 83 range 14.65 - 19.02 MeV by means of T(d,n)He4 neutron source were 3025 1451 84 renormalized to the cross section value of 152.9 mb (+-2.4%) at 3025 1451 85 14.65 MeV, evaluated from experimental data [4], [7], [8], [14] 3025 1451 86 and [27]. 3025 1451 87 Experimental data [34], [36], [38] and [46] were rejected due 3025 1451 88 to systematical and significant underestimation of cross sections 3025 1451 89 above 2.8 MeV. The results of Bormann and Lammers [9] measure- 3025 1451 90 ments obtained in the energy range 14.10-18.19 MeV were not taken 3025 1451 91 into account from 15.46 MeV to 18.19 MeV due to systematical over-3025 1451 92 estimation of cross sections. For the same reason data of Ghorai 3025 1451 93 et al. [28] at 17.2 MeV and Kielan and Marcinkowski [30] at 15.9 3025 1451 94 and 16.6 MeV were rejected also. 3025 1451 95 Cross section data given in the Refs. [33-46] were rejected 3025 1451 96 completely due to their big discrepancy with the main bulk of ex- 3025 1451 97 perimental data. In the rejected experimental data [33], [35], 3025 1451 98 [39], [41-45] cross section values were measured only at a one 3025 1451 99 energy point in the interval 14 - 15 MeV. 3025 1451 100 Statistical analysis of input cross section data was carried 3025 1451 101 out by means of PADE-2 code [50]. Rational function was used as 3025 1451 102 the model function [51]. 3025 1451 103 Evaluated excitation function for the reaction Zn64(n,p)Cu64 3025 1451 104 was tested with using integral experimental data [52-53] for 3025 1451 105 U-235 thermal fission neutron spectrum and integral experimental 3025 1451 106 data [54-57] for Cf-252 spontaneous fission neutron spectrum. 3025 1451 107 Calculated and measured average cross section values for U-235 3025 1451 108 thermal fission neutron spectrum [58] and Cf-252 spontaneous 3025 1451 109 fission neutron spectrum [59] are given in the table 1. 3025 1451 110 Table 1 3025 1451 111 ================================================================= 3025 1451 112 TYPE OF SPECTRUM ,mb (calc.) , mb (measured) 3025 1451 113 ----------------------------------------------------------------- 3025 1451 114 U-235 neutron fission 38.914 38.89 +- 2.82 [ *] 3025 1451 115 35.39 +- 1.07 [57] 3025 1451 116 ----------------------------------------------------------------- 3025 1451 117 CF-252 spont. fission 42.718 42.34 +- 0.94 [**] 3025 1451 118 40.59 +- 0.67 [57] 3025 1451 119 ================================================================= 3025 1451 120 * - averaged value obtained from experimental data [52-53] 3025 1451 121 ** - averaged value obtained from experimental data [54-56] 3025 1451 122 3025 1451 123 MF=33 3025 1451 124 MT=103 - (n,p) cross section cov. matrix 3025 1451 125 ---------------------------------------- 3025 1451 126 Uncertainties in the evaluated excitation function for the 3025 1451 127 reaction Zn-64(n,p)Cu-64 are given in the form of relative covari-3025 1451 128 ance matrix for the 49-neutron energy groups (LB=5). Covariance 3025 1451 129 matrix of uncertainties was calculated simultaneously with 3025 1451 130 recommended cross section data by means of PADE-2 code [50]. 3025 1451 131 Eigenvalues of the 6-th digits relative covariance matrix 3025 1451 132 given in the 33-file are the following: 3025 1451 133 3025 1451 134 5.50784E-08 5.71733E-08 5.94423E-08 6.15434E-08 3025 1451 135 6.41589E-08 6.67452E-08 6.87802E-08 7.09362E-08 3025 1451 136 7.33156E-08 7.62346E-08 7.97587E-08 8.39517E-08 3025 1451 137 8.83311E-08 9.24351E-08 9.75569E-08 1.04183E-07 3025 1451 138 1.11626E-07 1.18053E-07 1.24390E-07 1.34278E-07 3025 1451 139 1.46587E-07 1.56514E-07 1.69670E-07 1.96551E-07 3025 1451 140 2.10571E-07 2.58348E-07 2.86453E-07 3.37288E-07 3025 1451 141 4.18775E-07 5.04883E-07 5.48409E-07 6.51679E-07 3025 1451 142 7.95195E-07 9.69913E-07 1.51323E-06 6.87735E-06 3025 1451 143 4.70011E-05 4.54712E-04 8.44040E-04 1.00266E-03 3025 1451 144 1.39839E-03 1.60043E-03 2.35419E-03 2.47891E-03 3025 1451 145 5.19930E-03 6.20593E-03 6.78721E-03 2.04942E-02 3025 1451 146 3.28605E-02 3025 1451 147 3025 1451 148 References : 3025 1451 149 1. 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W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 3025 1451 255 ***************************************************************** 3025 1451 256 ***************** Program LINEAR (VERSION 2012-1) ***************3025 1451 257 For All Data Greater than 1.0000D-10 barns in Absolute Value 3025 1451 258 Data Linearized to Within an Accuracy of .100000000 per-cent 3025 1451 259 ***************** Program GROUPIE (VERSION 2012-1) **************3025 1451 260 Unshielded Group Averages Using 640 Groups 3025 1451 261 Weighting Spectrum: Flat (Constant) Spectrum 3025 1451 262 1 451 255 13025 1451 263 2 151 4 13025 1451 264 3 103 77 13025 1451 265 33 103 231 13025 1451 266 3025 1 0 267 3025 0 0 268 30064.0000 63.3800000 0 0 1 03025 2151 269 3.006400+4 1.000000+0 0 0 1 03025 2151 270 1.000000+0 2.000000+7 0 0 0 03025 2151 271 0.000000+0 6.700000-1 0 0 0 03025 2151 272 3025 2 0 273 3025 0 0 274 30064.0000 63.3800000 0 0 0 03025 3103 275 203500.000 203500.000 0 0 1 2063025 3103 276 206 1 3025 3103 277 500000.000 4.60828E-7 525000.000 1.38248E-6 550000.000 2.30414E-63025 3103 278 575000.000 3.22580E-6 600000.000 4.23962E-6 630000.000 5.34560E-63025 3103 279 660000.000 6.45159E-6 690000.000 7.55758E-6 720000.000 8.84790E-63025 3103 280 760000.000 1.03225E-5 800000.000 1.17972E-5 840000.000 1.32718E-53025 3103 281 880000.000 1.72077E-5 920000.000 3.59111E-5 960000.000 5.70758E-53025 3103 282 1000000.00 9.89036E-5 1100000.00 1.74179E-4 1200000.00 2.86221E-43025 3103 283 1300000.00 4.61633E-4 1400000.00 7.36625E-4 1500000.00 .0011596193025 3103 284 1600000.00 .001794205 1700000.00 .002722150 1800000.00 .0040459853025 3103 285 1900000.00 .005890170 2000000.00 .008399340 2100000.00 .0117317503025 3103 286 2200000.00 .016045750 2300000.00 .021477700 2400000.00 .0281126003025 3103 287 2500000.00 .035951950 2600000.00 .044887600 2700000.00 .0546936503025 3103 288 2800000.00 .065044700 2900000.00 .075559400 3000000.00 .0858585953025 3103 289 3100000.00 .095620495 3200000.00 .104617000 3300000.00 .1127240003025 3103 290 3400000.00 .119911500 3500000.00 .126221000 3600000.00 .1317400003025 3103 291 3700000.00 .136577000 3800000.00 .140846000 3900000.00 .1446555003025 3103 292 4000000.00 .148101500 4100000.00 .151266500 4200000.00 .1542190003025 3103 293 4300000.00 .157014486 4400000.00 .159696987 4500000.00 .1623000003025 3103 294 4600000.00 .164849500 4700000.00 .167365487 4800000.00 .1698609753025 3103 295 4900000.00 .172346488 5000000.00 .174828500 5100000.00 .1773100003025 3103 296 5200000.00 .179792488 5300000.00 .182275475 5400000.00 .1847574883025 3103 297 5500000.00 .187236000 5600000.00 .189707000 5700000.00 .1921659883025 3103 298 5800000.00 .194608476 5900000.00 .197029488 6000000.00 .1994235003025 3103 299 6100000.00 .201785488 6200000.00 .204109977 6300000.00 .2063919773025 3103 300 6400000.00 .208626489 6500000.00 .210808500 6600000.00 .2129339903025 3103 301 6700000.00 .214998980 6800000.00 .216999980 6900000.00 .2189339913025 3103 302 7000000.00 .220798991 7100000.00 .222592482 7200000.00 .2243129833025 3103 303 7300000.00 .225960484 7400000.00 .227534992 7500000.00 .2290369933025 3103 304 7600000.00 .230467986 7700000.00 .231830053 7800000.00 .2331256143025 3103 305 7900000.00 .234357548 8000000.00 .235530046 8100000.00 .2366475443025 3103 306 8200000.00 .237713990 8300000.00 .238734990 8400000.00 .2397159903025 3103 307 8500000.00 .240661995 8600000.00 .241578495 8700000.00 .2424709913025 3103 308 8800000.00 .243345991 8900000.00 .244208491 9000000.00 .2450624963025 3103 309 9100000.00 .245913496 9200000.00 .246765491 9300000.00 .2476209913025 3103 310 9400000.00 .248482491 9500000.00 .249350991 9600000.00 .2502269913025 3103 311 9700000.00 .251108991 9800000.00 .251993991 9900000.00 .2528784963025 3103 312 10000000.0 .253756000 10100000.0 .254619000 10200000.0 .2554585003025 3103 313 10300000.0 .256262500 10400000.0 .257018000 10500000.0 .2577110003025 3103 314 10600000.0 .258325500 10700000.0 .258844000 10800000.0 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