IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 14028.0000 27.7366000 0 0 41 11425 1451 1 0.0 0.0 0 0 0 61425 1451 2 1.00000000 60000000.0 0 0 10 11425 1451 3 300.000000 0.0 1 0 199 41425 1451 4 14-Si- 28 FEI+HUNDEB EVAL-Apr14 K.I.Zolotarev 1425 1451 5 DIST-May14 1425 1451 6 ----BROND-1 MATERIAL 1425 1425 1451 7 -----INCIDENT NEUTRON DATA 1425 1451 8 ------ENDF-6 FORMAT 1425 1451 9 ******************************************************************1425 1451 10 The evaluation was performed by K. Zolotarev under contract with 1425 1451 11 the IAEA, August 2014. 1425 1451 12 ******************************************************************1425 1451 13 ***************************************************************** 1425 1451 14 Authors of evaluation: K.Zolotarev, P.Zolotarev and J.Csikai 1425 1451 15 ***************************************************************** 1425 1451 16 MF=3 1425 1451 17 MT=103 - (n,p) cross section 1425 1451 18 ------------------------------------- 1425 1451 19 Microscopic experimental data [1-42] were analyzed in the pro-1425 1451 20 cess of preparation of input data base for the evaluation of exci-1425 1451 21 tation function of the Si-28(n,p)Al-28 reaction. 1425 1451 22 During this procedure experimental data were corrected to the 1425 1451 23 new standards. Recommended decay data for Al-28 were taken from 1425 1451 24 Ref. [43]. 1425 1451 25 Special correction was applied to the experimental data [3-4],1425 1451 26 [7-8], [12], [37] and [41]. 1425 1451 27 Experimental data of Marion et al. [3], Mainsbridge et al.[8] 1425 1451 28 and Bass et al. [12] were renormalized to the results of precise 1425 1451 29 measurements of W.Mannhart and D.Schmidt [40] in the overlapping 1425 1451 30 energy ranges. Correction factors for the experimental data [3], 1425 1451 31 [8], [12] were Fc=0.61225, Fc=0.52921, Fc=0.77906, respectively. 1425 1451 32 Experimental data of Furuta et al. [41] were also renormalized to 1425 1451 33 the results of precise measurements by W.Mannhart and D.Schmidt 1425 1451 34 via corrected data of Mainsbridge et al. [8] in the overlapping 1425 1451 35 energy 5.060 - 5.930 MeV, Fc=0.64855. 1425 1451 36 Measured in the energy range 12.33-18.24 MeV widely scattered 1425 1451 37 data of Kern et al. [4] were sorted out for a 5 sets of data with 1425 1451 38 approximately equivalent neutron flux density in the process of 1425 1451 39 irradiation. Each set of the data had been renormalized to corres-1425 1451 40 pondent cross section value determined from representative experi-1425 1451 41 mental data [32], [38], [40]. in the interval 13-14 MeV. 1425 1451 42 Cross section data obtained by Jeronymo et al. [7] at neutron 1425 1451 43 energies 12.55, 13.55, 14.90, 16.50, 19.60 MeV were renormalized 1425 1451 44 to the experimental data of Hongyu Zhou et al. [42] at 14.90 MeV, 1425 1451 45 Fc= 1.37274. 1425 1451 46 Experimental data by Y.Kasugai et al. [37] were renormalized 1425 1451 47 for the integral of cross section calculated from experimental 1425 1451 48 data of Ikeda et al. [32] in the overlapping energy range 13.40 - 1425 1451 49 14.87 MeV, Fc= 1.09426. 1425 1451 50 All experimental data except [32], [37], [39], [40] obtained 1425 1451 51 with using activation method and samples of natural isoptopic com-1425 1451 52 position were corrected above 12 MeV for contribution from reac- 1425 1451 53 tion Si29(n,x)Al28. 1425 1451 54 Cross section value measured by Klochkova et al. [35] was cor- 1425 1451 55 rected also for contribution from Si29(n,p)Al29 and Si30(n,p)Al30 1425 1451 56 reactions. 1425 1451 57 Experimental data presented in works [1-2], [5], [9], [10-11],1425 1451 58 [13-14], [17], [19], [22-23], [33], [34] were rejected completely.1425 1451 59 Results obtained by Cohen and White [2] above 14 MeV contradi-1425 1451 60 cts to the trend of Si-28(n,p)Al-28 reaction excitation function 1425 1451 61 determined in the experimental investigations [4], [26], [39]. 1425 1451 62 Data of Birk et al. [9] measured in the interval of neutron 1425 1451 63 energies 5.35-6.40 MeV gives a significantly overestimated cross- 1425 1451 64 sections in comparison with the renormalized experimental data 1425 1451 65 [3], [8], [12]. 1425 1451 66 Experimental data from works [1], [5], [10-11], [13-14], [17],1425 1451 67 [19], [22-23] and [33-34]] gives information only in one energy 1425 1451 68 point and were rejected due to uderestimation or overestimation 1425 1451 69 of cross section in the energy interval 14-15 MeV. 1425 1451 70 Excitation function for the Si-28(n,p)Al-28 reaction in the 1425 1451 71 energy region from threshold to 21.0 MeV was evaluated by means 1425 1451 72 of statistical analysis of experimental cross section data [3-4], 1425 1451 73 [6-7], [12], [15-16], [18], [20-21], [24-32], [35-42]. 1425 1451 74 Statistical analysis of input cross section data was carried 1425 1451 75 out by means of PADE-2 code [44]. Rational function was used as 1425 1451 76 the model function [45]. 1425 1451 77 Evaluated Si-28(n,p)Al-28 reaction excitation function was 1425 1451 78 stested with using integral experimental data [46-49] for U-235 1425 1451 79 thermal fission neutron spectrum [50] and Cf-252 spontaneous fis- 1425 1451 80 sion neutron spectrum [51]. Integral experimental data were cor- 1425 1451 81 rected to the new recommended cross sections [52], [53] for the 1425 1451 82 monitor reactions. Calculated and measured average cross-section 1425 1451 83 values for are given below. 1425 1451 84 1425 1451 85 ----------------------+-----------------+----------------------- 1425 1451 86 TYPE OF SPECTRUM | ,mb (calc.) | , mb (measured) 1425 1451 87 ----------------------+-----------------+----------------------- 1425 1451 88 U-235 neutron fission | 5.4836 | 5.470 +- 0.168 [ *] 1425 1451 89 ----------------------+-----------------+----------------------- 1425 1451 90 CF-252 spont. fission | 7.1173 | 6.900 +- 0.437 [49] 1425 1451 91 ----------------------+-----------------+----------------------- 1425 1451 92 * - averaged value obtained from experimental data [46-48] 1425 1451 93 1425 1451 94 MF=33 1425 1451 95 MT=103 - (n,p) cross section cov. matrix 1425 1451 96 ---------------------------------------- 1425 1451 97 Uncertainties in the evaluated excitation function for the 1425 1451 98 reaction Si-28(n,p)Al-28 are given in the form of relative covari-1425 1451 99 ance matrix for the 47-neutron energy groups (LB=5). Covariance 1425 1451 100 matrix of uncertainties was calculated simultaneously with 1425 1451 101 recommended cross section data by means of PADE-2 code [44]. 1425 1451 102 Eigenvalues of the 6-th digits relative covariance matrix 1425 1451 103 given in the 33-file are the following: 1425 1451 104 1425 1451 105 3.51564E-06 3.52823E-06 3.54907E-06 3.57789E-06 1425 1451 106 3.61472E-06 3.66117E-06 3.71115E-06 3.77705E-06 1425 1451 107 3.83668E-06 3.92041E-06 3.99476E-06 4.07835E-06 1425 1451 108 4.19493E-06 4.30483E-06 4.41557E-06 4.55429E-06 1425 1451 109 4.93802E-06 5.62141E-06 6.59375E-06 8.10454E-06 1425 1451 110 1.03073E-05 1.34009E-05 1.76791E-05 2.35391E-05 1425 1451 111 3.11619E-05 3.32529E-05 4.37642E-05 1.54908E-04 1425 1451 112 2.25270E-04 6.60505E-04 7.08582E-04 8.01326E-04 1425 1451 113 8.79098E-04 9.45802E-04 1.02200E-03 1.18158E-03 1425 1451 114 1.42206E-03 1.51558E-03 1.59922E-03 1.95391E-03 1425 1451 115 3.05175E-03 4.12368E-03 7.39707E-03 1.53135E-02 1425 1451 116 1.86107E-02 5.42632E-02 6.77829E-02 1425 1451 117 1425 1451 118 References : 1425 1451 119 1. E.B.Paul, R.L.Clarke, Can. J. of Physics 31 (1953) 267 1425 1451 120 2. A.V.Cohen, P.H.White, Nucl. Phys. 1 (1956) 73 1425 1451 121 3. J.B.Marion et al., Phys. Rev. 101 (1956) 247 1425 1451 122 4. B.D.Kern et al. Nucl. Phys. 10 (1959) 226 1425 1451 123 5. C.S.Khurana, H.S.Hans Proc. of 4th Nuclear Physics and Solid 1425 1451 124 Stata Physics Symp., 24-26 Feb. 1960, Waltair, India, p.297 1425 1451 125 6. D.L.Allan, Nucl. Phys. 24 (1961) 274 1425 1451 126 7. J.M.F.Jeronymo et al., Nucl. Phys. 47 (1963) 157 1425 1451 127 8. B.Mainsbridge et al., Nucl. Phys. 48 (1963) 83 1425 1451 128 9. M.Birk et al., Nucl. Instrum. Methods 21 (1963) 197 1425 1451 129 10. J.E.Strain, W.J.Ross, Report ORNL-3672, January 1965 1425 1451 130 11. C.S.Khurana, I.M.Govil, Nucl. Phys. 69 (1965) 153 1425 1451 131 12. R.Bass et al., Report EANDC(E)-66, p.64, February 1966 1425 1451 132 13. B.Mitra, A.M.Ghose, Nucl. Phys. 83 (1966) 157 1425 1451 133 14. A.Pasquarelli, Nucl. Phys. A93 (1967) 218 1425 1451 134 15. 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P.N.Ngoc et al., Report ZFK-410, p.192, January 1980 1425 1451 150 28. J.Janczyszyn Proc. of Int. Conf. on Nucl. Data for Science 1425 1451 151 and Techn., 6-10 September 1982, Antwerp, Holland, D.Reidel 1425 1451 152 Publishing Company, p.869 1425 1451 153 29. R.Pepelnik et al., Progress Report NEANDC(E)-262U,(5), p. 32, 1425 1451 154 June 1985 1425 1451 155 30. D.A.Bradley et al., Proc. of Int. Symp. on Fast Neutrons in 1425 1451 156 Science and Technology, Chiang Mai, 4-8 February 1985, p.19 1425 1451 157 31. J.Csikai, T.Chimoye et al., Zeitschrift fuer Physik A325 1425 1451 158 (1986) 69 1425 1451 159 32. Y.Ikeda et al., Report JAERI-1312, March 1988 1425 1451 160 33. A.Ercan et al., Proc. of an Intern. Conf. on Nuclear Data for 1425 1451 161 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-1425 1451 162 Verlag, 1992, pp.376-377 1425 1451 163 34. M.Belgaid et al., J. of Radioanalytical and Nuclear Chemistry,1425 1451 164 Letters 166 (1992) 493 1425 1451 165 35. 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S.Badikov, N.Rabotnov, K.Zolotarev, Proc. of NEANSC Speciali- 1425 1451 178 st's Meeting on Evaluation and Processing of Covariance Data, 1425 1451 179 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 1425 1451 180 46. I.Kimura et al., Nucl. Sci. Technology 8 (1971) 59 1425 1451 181 47. E.I.Grigor'ev, V.P.Jaryna, Atomaja Energija 43 (1977) 14 (in 1425 1451 182 Russian). 1425 1451 183 48. M.A.Arribere et al., J. Radioanalytical and Nuclear Chemistry 1425 1451 184 244 (2000) 417 1425 1451 185 49. Z.Dezso, J.Csikai, Proc. of Int. Conf. on Nuclear Data for 1425 1451 186 Science and Technology, 6-10 September 1982, Antwerp, Holland,1425 1451 187 D.Reidel Publishing Company, p.418 1425 1451 188 50. P.G.Young et al., Evaluated Neutron Data for Uranium-235, 1425 1451 189 ENDF/B-VII.1 Library, MAT=9228, MF=5, MT=18, eval. Sept. 2006 1425 1451 190 51. W.Mannhart, Report, INDC(NDS)-220/L,p.158, IAEA, Vienna, 1989 1425 1451 191 52. 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W.Mannhart, Status of the Evaluation of the Neutron Spectrum 1425 1451 194 of 252Cf(sf), Report INDC(NDS)-0540, IAEA, Vienna, 2008 1425 1451 195 1425 1451 196 ***************************************************************** 1425 1451 197 ***************** Program LINEAR (VERSION 2012-1) ***************1425 1451 198 For All Data Greater than 1.0000D-10 barns in Absolute Value 1425 1451 199 Data Linearized to Within an Accuracy of .100000000 per-cent 1425 1451 200 ***************** Program GROUPIE (VERSION 2012-1) **************1425 1451 201 Unshielded Group Averages Using 640 Groups 1425 1451 202 Weighting Spectrum: Flat (Constant) Spectrum 1425 1451 203 1 451 197 11425 1451 204 2 151 4 11425 1451 205 3 103 227 11425 1451 206 33 103 214 11425 1451 207 1425 1 0 208 1425 0 0 209 14028.0000 27.7366000 0 0 1 01425 2151 210 1.402800+4 1.000000+0 0 0 1 01425 2151 211 1.000000-5 1.750000+6 0 0 0 01425 2151 212 0.000000+0 4.136400-1 0 0 0 01425 2151 213 1425 2 0 214 1425 0 0 215 14028.0000 27.7366000 0 0 0 01425 3103 216 -3859920.00-3859920.00 0 0 1 1621425 3103 217 162 1 1425 3103 218 3900000.00 2.78370E-9 4000000.00 3.65791E-5 4100000.00 1.08472E-41425 3103 219 4200000.00 1.80365E-4 4300000.00 3.35979E-4 4400000.00 .0010341591425 3103 220 4500000.00 .002013580 4600000.00 .003231060 4700000.00 .0047395001425 3103 221 4800000.00 .006576175 4900000.00 .008752860 5000000.00 .0112196251425 3103 222 5100000.00 .013747950 5200000.00 .015225270 5300000.00 .0146615601425 3103 223 5400000.00 .021811960 5500000.00 .026449050 5600000.00 .0392183201425 3103 224 5700000.00 .040219740 5800000.00 .043491290 5900000.00 .0560553901425 3103 225 6000000.00 .069110928 6100000.00 .076905265 6200000.00 .0855128491425 3103 226 6300000.00 .107752431 6400000.00 .170419125 6500000.00 .1384060251425 3103 227 6600000.00 .145877375 6700000.00 .185359650 6800000.00 .1946155501425 3103 228 6900000.00 .169575250 7000000.00 .189214850 7100000.00 .1693275001425 3103 229 7200000.00 .228039425 7300000.00 .220888825 7400000.00 .2216625501425 3103 230 7500000.00 .177083050 7600000.00 .238379850 7700000.00 .2601657131425 3103 231 7800000.00 .275268050 7900000.00 .249066100 8000000.00 .1969320361425 3103 232 8100000.00 .181705467 8200000.00 .196774659 8300000.00 .2002936911425 3103 233 8400000.00 .233463418 8500000.00 .230983806 8600000.00 .2752309571425 3103 234 8700000.00 .214813238 8800000.00 .222815782 8900000.00 .2339267401425 3103 235 9000000.00 .235855535 9100000.00 .223215000 9200000.00 .2265885001425 3103 236 9300000.00 .228353500 9400000.00 .229123000 9500000.00 .2299550001425 3103 237 9600000.00 .232369000 9700000.00 .238288000 9800000.00 .2477055001425 3103 238 9900000.00 .256809000 10000000.0 .262093000 10100000.0 .2639815001425 3103 239 10200000.0 .264268500 10300000.0 .263193000 10400000.0 .2610595001425 3103 240 10500000.0 .258785000 10600000.0 .256583000 10700000.0 .2545175001425 3103 241 10800000.0 .252568000 10900000.0 .250704000 11000000.0 .2493460001425 3103 242 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14900000.0 .2456365001425 3103 255 15000000.0 .242005500 15100000.0 .238306000 15200000.0 .2345525001425 3103 256 15300000.0 .230757000 15400000.0 .226930500 15500000.0 .2230835001425 3103 257 15600000.0 .219225000 15700000.0 .215362500 15800000.0 .2115025001425 3103 258 15900000.0 .207804500 16000000.0 .204309500 16100000.0 .2008360001425 3103 259 16200000.0 .197320500 16300000.0 .193770500 16400000.0 .1901945001425 3103 260 16500000.0 .186600500 16600000.0 .182995500 16700000.0 .1793870001425 3103 261 16800000.0 .175782000 16900000.0 .172186500 17000000.0 .1686075001425 3103 262 17100000.0 .165051000 17200000.0 .161523000 17300000.0 .1580285001425 3103 263 17400000.0 .154572500 17500000.0 .151160000 17600000.0 .1477945001425 3103 264 17700000.0 .144480500 17800000.0 .141222000 17900000.0 .1380210001425 3103 265 18000000.0 .134880500 18100000.0 .131804000 18200000.0 .1287930001425 3103 266 18300000.0 .125849000 18400000.0 .122973500 18500000.0 .1201675001425 3103 267 18600000.0 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