IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 22048.0000 47.5360000 0 0 34 102231 1451 1 0.0 0.0 0 0 0 62231 1451 2 1.00000000 60000000.0 0 0 10 20022231 1451 3 300.000000 0.0 1 0 218 72231 1451 4 22-Ti- 48 FEI EVAL-Feb02 K.I.Zolotarev 2231 1451 5 DIST-May12 Rev. 3 May 2012 2231 1451 6 ----IRDF-2002 MATERIAL 2231 2231 1451 7 -----INCIDENT NEUTRON DATA 2231 1451 8 ------ENDF-6 FORMAT 2231 1451 9 ******************************************************************2231 1451 10 IAEA, July 2011 (A. Trkov) 2231 1451 11 Extended cross sections and covariances from 20 to 60 MeV 2231 1451 12 by TENDL-2010, renormalised for continuity. 2231 1451 13 ******************************************************************2231 1451 14 ***************************************************************** 2231 1451 15 ********************** SPECIAL PURPOSE FILE ***************** 2231 1451 16 DOSIMETRY 2231 1451 17 2231 1451 18 For this special purpose library it was decided the reaction 2231 1451 19 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2231 1451 20 This was done after processing through the codes. The 2231 1451 21 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2231 1451 22 ***************************************************************** 2231 1451 23 Author of evaluation: K.I.Zolotarev 2231 1451 24 ***************************************************************** 2231 1451 25 MF= 3 2231 1451 26 MT= 28 -(n,np+pn+d) cross section data 2231 1451 27 -------------------------------------- 2231 1451 28 In this section is given the sum of cross sections of the re- 2231 1451 29 actions Ti-48(n,np)Sc-47 , Ti-48(n,pn)Sc-47 and Ti-48(n,d)Sc-47. 2231 1451 30 Excitation function for the Ti-48(n,x)Sc-47 reaction in the 2231 1451 31 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 32 statistical analysis of experimental cross section data [1-12] 2231 1451 33 and data from STAPRE [13] calculation. 2231 1451 34 Analised experimental data were corrected to the new stan- 2231 1451 35 dards for monitor reactions cross sections and decay data. 2231 1451 36 The final procedure of evaluation Ti-48(n,x)Sc-47 excitation 2231 1451 37 function from threshold to 20 MeV has been carried out within the 2231 1451 38 framework of generalized least squares method. Rational function 2231 1451 39 was used as model function [14]. Calculations was performed by 2231 1451 40 means of Pade-2 code [15]. 2231 1451 41 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 42 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 43 evaluated Ti-48(n,x)Sc-47 excitation function are the following: 2231 1451 44 2231 1451 45 -------------------------------------------- 2231 1451 46 TYPE OF SPECTRUM I , mb (calc.) 2231 1451 47 --------------------------+----------------- 2231 1451 48 U-235 neutron fission I 1.5730E-3 2231 1451 49 CF-252 spontan. fission I 4.3507E-3 2231 1451 50 2231 1451 51 MT=103 -(n,p) cross section 2231 1451 52 ------------------------------------- 2231 1451 53 Excitation function for the Ti-48(n,p)Sc-48 reaction in the 2231 1451 54 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 55 of statistical analysis of experimental cross section data [1-2], 2231 1451 56 [4], [6-11], [18-41] and data from STAPRE [13] calculation. 2231 1451 57 The energy dependence of cross-section from 4.7 MeV to thres- 2231 1451 58 hold was extrapolated with L=0 penetrability function for the 2231 1451 59 the outgoing p + Sc48 channal [42]. 2231 1451 60 Analised microscopic experimental data were renormalized to 2231 1451 61 the new recommended standards for monitor reaction cross sections 2231 1451 62 and decay data. All experiments performed with Ti-natural samples 2231 1451 63 were corrected for the contribution of the Ti-49(n,np+pn)Sc-48 2231 1451 64 and Ti-49(n,d)Sc-48 reactions. Total cross section for these reac-2231 1451 65 tions were taken from ref. [43]. Experimental data from ref. [24] 2231 1451 66 were used partially. It was taken into account only data measured 2231 1451 67 relative Fe-56(n,p)Mn-56 monitor reaction. Data of Firkin [6] 2231 1451 68 obtained in the experiment with neutrons from D(d,n)He3 reaction 2231 1451 69 were renormalized to the results of his measurements with 14.1 2231 1451 70 MeV neutrons from T(d,n)He4 reaction. Cross section data of Smith 2231 1451 71 and Meadows [26] measured in the energy range 5.964 - 9.952 MeV 2231 1451 72 with using D(d,n)He3 neutron source were renormalized to the 2231 1451 73 value 30.7 mb at 10 MeV [38]. The correction factors for the 2231 1451 74 experimental data [6] and [26] were Fc=0.91853 and Fc=1.11111 , 2231 1451 75 respectively 2231 1451 76 Experimental cross section data [44-51] were rejected due to 2231 1451 77 their discrepancy with the main bulk of experimental data [1-2], 2231 1451 78 [4], [6-11], [18-41] and data from theoretical model calculatiod. 2231 1451 79 Statistical analysis of input cross section data was carried 2231 1451 80 out by means of PADE-2 code [15]. Rational function was used as 2231 1451 81 the model function [14]. 2231 1451 82 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 83 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 84 evaluated Ti-48(n,p)Sc-48 excitation function are the following: 2231 1451 85 2231 1451 86 ----------------------+-----------------+----------------------- 2231 1451 87 TYPE OF SPECTRUM | ,mb (calc.) | , mb (measured) 2231 1451 88 ----------------------+-----------------+----------------------- 2231 1451 89 U-235 neutron fission | 0.30156 | 0.305 +- 0.020 [52] 2231 1451 90 | | 0.302 +- 0.010 [53] 2231 1451 91 | | 0.2996+- 0.0054 [54] 2231 1451 92 ----------------------+-----------------+----------------------- 2231 1451 93 Cf-252 spont. fission | 0.42700 | 0.4247+- 0.0080 [55] 2231 1451 94 ----------------------+-----------------+----------------------- 2231 1451 95 2231 1451 96 MF=33 2231 1451 97 MT= 28 -(n,np+pn+d) cross section cov. matrix 2231 1451 98 --------------------------------------------- 2231 1451 99 Uncertainties in the evaluated excitation function for the 2231 1451 100 reaction Ti-48(n,x)Sc-47 are given in the form of relative cova- 2231 1451 101 riance matrix for the 15-neutron energy groups (LB=5). Covariance 2231 1451 102 matrix of uncertainties was calculated simultaneously with 2231 1451 103 recommended cross section data by means of PADE-2 code. 2231 1451 104 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 105 given in the 33-file are the following: 2231 1451 106 2231 1451 107 3.25031E-07 3.96552E-07 7.16579E-07 1.55391E-06 2231 1451 108 4.04176E-06 1.39260E-05 8.49003E-05 2.17726E-03 2231 1451 109 8.18710E-03 1.48647E-02 2.37004E-02 3.37652E-02 2231 1451 110 4.90298E-02 9.66009E-02 5.46438E-01 2231 1451 111 2231 1451 112 MT=103 -(n,p) cross section cov. matrix 2231 1451 113 --------------------------------------------- 2231 1451 114 Uncertainties in the evaluated excitation function for the 2231 1451 115 reaction Ti-48(n,p)Sc-48 are given in the form of relative cova- 2231 1451 116 riance matrix for the 29-neutron energy groups (LB=5). Covariance 2231 1451 117 matrix of uncertainties was calculated simultaneously with 2231 1451 118 recommended cross section data by means of PADE-2 code. 2231 1451 119 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 120 given in the 33-file are the following: 2231 1451 121 2231 1451 122 1.82983E-06 1.87119E-06 1.93307E-06 2.03636E-06 2231 1451 123 2.14719E-06 2.31398E-06 2.49898E-06 2.70004E-06 2231 1451 124 3.01245E-06 3.24044E-06 3.62018E-06 4.06991E-06 2231 1451 125 4.35046E-06 4.94940E-06 5.62390E-06 5.93559E-06 2231 1451 126 8.04090E-06 1.16767E-05 1.78927E-05 3.06627E-05 2231 1451 127 6.95321E-05 2.90791E-04 3.53832E-03 6.96897E-03 2231 1451 128 9.39686E-03 1.29685E-02 3.00706E-02 4.84005E-02 2231 1451 129 1.29631E-01 2231 1451 130 2231 1451 131 References: 2231 1451 132 1. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2231 1451 133 2. H.L.Pai, Canadian J. of Physics, v.44, p.2337, 1966 2231 1451 134 3. S.M.Grimes et al., Nucl. Sci. Eng., v.62, p.187, Feb. 1977 2231 1451 135 4. S.M.Qaim, N.I.Molla, Nucl. Phys., v.A283, p.269, June 1977 2231 1451 136 5. M.Viennot et al., Progress Report MOH-5, p.10, 1982 2231 1451 137 6. S.Firkin, Report AERE-M-3350, Harwell, September 1983 2231 1451 138 7. I.Ribansky, S.Gmuca, J. Phys.G, v.9, p.1537, December 1983 2231 1451 139 8. N.I.Molla et al., Report INDC(BAN)-003, September 1986 2231 1451 140 9. Y.Ikeda et al., Report JAERI-1312, March 1988 2231 1451 141 10. W.V.Hecker et al., Nucl. Inst. Meth., Sec.B, v.40/41, p.478, 2231 1451 142 April 1989 2231 1451 143 11. M.Viennot et al., Nucl. Sci. Eng., v.108, p.289, July 1991 2231 1451 144 12. Y.Uno et al., Report JAERI-M-93-046, p.247-256, 1993 2231 1451 145 13. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2231 1451 146 Activation Cross Section and Related Quantities, Report 2231 1451 147 IRK 76-01, Vienna, 1976 2231 1451 148 14. S.Badikov, N.Rabotnov, K.Zolotarev, Proc. of NEANSC Speciali- 2231 1451 149 st's Meeting on Evaluation and Processing of Covariance Data, 2231 1451 150 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2231 1451 151 15. S.A.Badikov et al., Preprint FEI-1686, Obninsk, 1985 2231 1451 152 16. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2231 1451 153 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2231 1451 154 17. W.Mannhart Report, INDC(NDS)-220/L,p.158, IAEA, Vienna, 1989 2231 1451 155 18. A.Poularikas, R.W.Fink, Physical Review, v.115, p.989, 1959 2231 1451 156 19. F.Gabbard, B.D.Kern, Physical Review, v.128, p.1276, 1962 2231 1451 157 20. M.Hillman, Nucl. Phys., v.37, p.78, 1962 2231 1451 158 21. M.Bormann et al., Nucl. Phys., v.63, p.438, March 1965 2231 1451 159 22. V.N.Levkovskij et al., Yadernaja Fizika (Sov.), v.10, n.1, 2231 1451 160 p.44, July 1969 2231 1451 161 23. S.K.Ghorai et al., J. Nucl. Energy, v.25, p.319, August 1971 2231 1451 162 24. Y.Lukic, E.E.Carroll, Nucl. Sci. Eng., v.43, p.233, 1971 2231 1451 163 25. J.Prasad, D.C.Sarcar, Nuovo Cimento, v.A3, p.467, June 1971 2231 1451 164 26. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2231 1451 165 27. W.Mannhart, H.Vonach, Zeitsch. fur Physik, Sect.A, v.272, 2231 1451 166 p.279, March 1975 2231 1451 167 28. K.Kayashima e.a., Prog. Report NEANDC(J)-61U, p.94, Sep. 1979 2231 1451 168 29. M.Viennot e.a. Proc. of Int. Conf.on Nuclear Data for Science 2231 1451 169 and Technology, Antwerpen, Belgium, 6-10 Sept. 1982, p.406 2231 1451 170 30. L.R.Greenwood, Report DOE-ER-0046-21, p.15, May 1985 2231 1451 171 31. R.Pepelnik et al. Progress Report NEANDC(E)-262U, (5), p.32, 2231 1451 172 June 1985 2231 1451 173 32. J.P.Gupta et al., Pramana, v.24, p.637, 1985 2231 1451 174 33. Hoang Dac Luc et al., Progress Report INDC(VN)-5, Sep. 1986 2231 1451 175 34. K.Kobayashi, I.Kimura, Proc. of an Intern. Conf. on Nuclear 2231 1451 176 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2231 1451 177 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 2231 1451 178 35. Lu Han-Lin et al., Report INDC(CPR)-16, August 1989 2231 1451 179 36. Y.Ikeda et al., Report JAERI-M-91-032, p.281, March 1991 2231 1451 180 37. M.Viennot et al., Nucl. Sci. Eng., v.108, p.289, July 1991 2231 1451 181 38. S.M.Qaim et al., Proc. of Int. Conference on Nuclear Data for 2231 1451 182 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 2231 1451 183 Verlag, Berlin - Heidelberg, 1992, p.297-300 2231 1451 184 39. I.Garlea et al., J. Rev. Roum. Phys., v.37, n.1, p.19, 1992 2231 1451 185 40. Yuan Junqian et al., High Energy Physics and Nuclear Physics 2231 1451 186 (China), v.16, n.1, p.57, January 1992 2231 1451 187 41. A.Grallert et al., Progress Report,INDC(NDS)-286, p.131, 1993 2231 1451 188 42. S.A.Badikov, A.B.Pashchenko, Voprosy Atomnoy Nauki i Tekhniki 2231 1451 189 Ser. Jadernye Konstanty, 2(53), p.70, 1987 2231 1451 190 43. K.Zolotarev, RRDF-2002, MAT=2243, eval. February 2002 2231 1451 191 44. E.B.Paul, R.L.Clarke, Canadian J. Phys., v.31, p.267, 1953 2231 1451 192 45. D.L.Allan, Nucl. Phys., v.24, p.274, April 1961 2231 1451 193 46. J.E.Strain, W.J.Ross, Report ORNL-3672, January 1965 2231 1451 194 47. H.K.Vonach et al., Proc. of 2-nd Conference on Nuclear Cross- 2231 1451 195 Sections and Technology, Washington D.C., 4-7 March 1968, 2231 1451 196 v.2, p.885 2231 1451 197 48. D.Crumpton, J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969 2231 1451 198 49. V.K.Tikku et al., Proc. of Nucl. Phys. and Solid State Phys. 2231 1451 199 Symp., Chandigarh, v.2, p.115, December 1972 2231 1451 200 50. R.Spangler et al., J. Trans. Amer. Nucl. Soc., v.22, p.818, 2231 1451 201 November 1975 2231 1451 202 51. K.T.Osman, F.I.Habbani, Report, INDC(SUD)-001, October 1996 2231 1451 203 52. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2231 1451 204 Washington D.C., 25-28 April 1989, v.2, p.923 2231 1451 205 53. W.Mannhart, Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2231 1451 206 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2231 1451 207 54. W.Mannhart, Status of the Evaluation of the Neutron Spectrum 2231 1451 208 of 235U + n-th., Report INDC(NDS)-0540, IAEA, Vienna, 2008 2231 1451 209 55. W.Mannhart, Status of the Evaluation of the Neutron Spectrum 2231 1451 210 of 252Cf(sf), Report INDC(NDS)-0540, IAEA, Vienna, 2008 2231 1451 211 ***************************************************************** 2231 1451 212 File 2 added to the pointwise file containing only the effective 2231 1451 213 scattering radius with no resonance parameters given. 2231 1451 214 Taken from ENDF/B-VI 2231 1451 215 ***************************************************************** 2231 1451 216 ***************** Program LINEAR (VERSION 2012-1) ***************2231 1451 217 For All Data Greater than 1.0000D-10 barns in Absolute Value 2231 1451 218 Data Linearized to Within an Accuracy of .100000000 per-cent 2231 1451 219 ***************** Program GROUPIE (VERSION 2012-1) **************2231 1451 220 Unshielded Group Averages Using 640 Groups 2231 1451 221 Weighting Spectrum: Flat (Constant) Spectrum 2231 1451 222 1 451 223 12231 1451 223 2 151 4 12231 1451 224 3 103 66 12231 1451 225 8 5 2 12231 1451 226 10 5 45 12231 1451 227 33 103 93 12231 1451 228 40 5 30 12231 1451 229 2231 1 0 230 2231 0 0 231 22048.0000 47.5360000 0 0 1 02231 2151 232 2.204800+4 1.000000+0 0 0 1 02231 2151 233 1.000000-5 1.000000-5 0 0 0 02231 2151 234 0.000000+0 5.733000-1 0 0 0 02231 2151 235 2231 2 0 236 2231 0 0 237 22048.0000 47.5360000 0 0 0 02231 3103 238 -3211800.00-3211800.00 0 0 1 1692231 3103 239 169 1 2231 3103 240 3200000.00 4.44705E-9 3300000.00 1.46148E-7 3400000.00 3.52717E-72231 3103 241 3500000.00 5.59287E-7 3600000.00 7.65856E-7 3700000.00 9.72426E-72231 3103 242 3800000.00 1.17900E-6 3900000.00 1.75487E-6 4000000.00 2.60258E-62231 3103 243 4100000.00 3.40719E-6 4200000.00 4.52734E-6 4300000.00 6.44311E-62231 3103 244 4400000.00 9.78106E-6 4500000.00 1.53460E-5 4600000.00 2.41598E-52231 3103 245 4700000.00 3.75024E-5 4800000.00 5.69529E-5 4900000.00 8.44353E-52231 3103 246 5000000.00 1.22266E-4 5100000.00 1.73193E-4 5200000.00 2.40420E-42231 3103 247 5300000.00 3.27615E-4 5400000.00 4.38894E-4 5500000.00 5.78755E-42231 3103 248 5600000.00 7.51970E-4 5700000.00 9.63413E-4 5800000.00 .0012178252231 3103 249 5900000.00 .001519525 6000000.00 .001872085 6100000.00 .0022779952231 3103 250 6200000.00 .002738340 6300000.00 .003252590 6400000.00 .0038184952231 3103 251 6500000.00 .004432120 6600000.00 .005088075 6700000.00 .0057798702231 3103 252 6800000.00 .006500350 6900000.00 .007242180 7000000.00 .0079983202231 3103 253 7100000.00 .008762405 7200000.00 .009529020 7300000.00 .0102938752231 3103 254 7400000.00 .011053750 7500000.00 .011806650 7600000.00 .0125515502231 3103 255 7700000.00 .013288200 7800000.00 .014017100 7900000.00 .0147393002231 3103 256 8000000.00 .015456250 8100000.00 .016169550 8200000.00 .0168810002231 3103 257 8300000.00 .017592500 8400000.00 .018306000 8500000.00 .0190233502231 3103 258 8600000.00 .019746300 8700000.00 .020476550 8800000.00 .0212156002231 3103 259 8900000.00 .021964950 9000000.00 .022725900 9100000.00 .0234996002231 3103 260 9200000.00 .024287100 9300000.00 .025089300 9400000.00 .0259069502231 3103 261 9500000.00 .026740650 9600000.00 .027590950 9700000.00 .0284582002231 3103 262 9800000.00 .029342550 9900000.00 .030244000 10000000.0 .0311625002231 3103 263 10100000.0 .032097800 10200000.0 .033049450 10300000.0 .0340169002231 3103 264 10400000.0 .034999400 10500000.0 .035996000 10600000.0 .0370056002231 3103 265 10700000.0 .038026900 10800000.0 .039058450 10900000.0 .0400986502231 3103 266 11000000.0 .041145600 11100000.0 .042197200 11200000.0 .0432513502231 3103 267 11300000.0 .044305650 11400000.0 .045357500 11500000.0 .0464043002231 3103 268 11600000.0 .047443150 11700000.0 .048471000 11800000.0 .0494848502231 3103 269 11900000.0 .050481550 12000000.0 .051457850 12100000.0 .0524105002231 3103 270 12200000.0 .053336200 12300000.0 .054231750 12400000.0 .0550940002231 3103 271 12500000.0 .055919800 12600000.0 .056706200 12700000.0 .0574504502231 3103 272 12800000.0 .058149900 12900000.0 .058802050 13000000.0 .0594048002231 3103 273 13100000.0 .059956250 13200000.0 .060454700 13300000.0 .0608989002231 3103 274 13400000.0 .061287850 13500000.0 .061620800 13600000.0 .0618974002231 3103 275 13700000.0 .062117600 13800000.0 .062281650 13900000.0 .0623901502231 3103 276 14000000.0 .062443900 14100000.0 .062444050 14200000.0 .0623919502231 3103 277 14300000.0 .062289200 14400000.0 .062137650 14500000.0 .0619393002231 3103 278 14600000.0 .061696250 14700000.0 .061410700 14800000.0 .0610850502231 3103 279 14900000.0 .060721700 15000000.0 .060323200 15100000.0 .0598920002231 3103 280 15200000.0 .059430600 15300000.0 .058941500 15400000.0 .0584271502231 3103 281 15500000.0 .057890000 15600000.0 .057332300 15700000.0 .0567564002231 3103 282 15800000.0 .056164500 15900000.0 .055558600 16000000.0 .0549407502231 3103 283 16100000.0 .054312900 16200000.0 .053676800 16300000.0 .0530341002231 3103 284 16400000.0 .052386400 16500000.0 .051735250 16600000.0 .0510819502231 3103 285 16700000.0 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2231 8 0 302 2231 0 0 303 22048.0000 47.5360000 0 0 1 0223110 5 304 -9219999.00-9219999.00 21047 0 1 107223110 5 305 107 1 223110 5 306 9400000.00 2.09726E-7 9500000.00 7.70813E-7 9600000.00 1.33742E-6223110 5 307 9700000.00 1.90403E-6 9800000.00 2.47063E-6 9900000.00 3.03724E-6223110 5 308 10000000.0 3.60385E-6 10100000.0 4.17046E-6 10200000.0 5.32343E-6223110 5 309 10300000.0 6.96840E-6 10400000.0 8.47552E-6 10500000.0 9.97182E-6223110 5 310 10600000.0 1.16052E-5 10700000.0 1.35478E-5 10800000.0 1.59992E-5223110 5 311 10900000.0 1.91905E-5 11000000.0 2.33890E-5 11100000.0 2.89037E-5223110 5 312 11200000.0 3.60910E-5 11300000.0 4.53630E-5 11400000.0 5.71957E-5223110 5 313 11500000.0 7.21354E-5 11600000.0 9.08205E-5 11700000.0 1.13983E-4223110 5 314 11800000.0 1.42472E-4 11900000.0 1.77270E-4 12000000.0 2.19511E-4223110 5 315 12100000.0 2.70512E-4 12200000.0 3.31792E-4 12300000.0 4.05106E-4223110 5 316 12400000.0 4.92478E-4 12500000.0 5.96228E-4 12600000.0 7.19007E-4223110 5 317 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