IRDFF-v1-05_g.zip separated 31.08.15 by O.Gritzay using div-lib 6000 0 0 0 22049.0000 48.5274000 0 0 34 202234 1451 1 0.0 0.0 0 0 0 62234 1451 2 1.00000000 60000000.0 0 0 10 20022234 1451 3 300.000000 0.0 1 0 100 52234 1451 4 22-Ti- 49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 5 DIST-Feb2004 2234 1451 6 ----IRDF-2002 MATERIAL 2234 2234 1451 7 -----INCIDENT NEUTRON DATA 2234 1451 8 ------ENDF-6 FORMAT 2234 1451 9 ******************************************************************2234 1451 10 IAEA, July 2012 (A. Trkov) 2234 1451 11 Extended cross sections and covariances from 20 to 60 MeV 2234 1451 12 by TENDL-2011, renormalised for continuity. 2234 1451 13 ******************************************************************2234 1451 14 ********************** SPECIAL PURPOSE FILE ***************** 2234 1451 15 DOSIMETRY 2234 1451 16 2234 1451 17 For this special purpose library it was decided the reaction 2234 1451 18 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2234 1451 19 This was done after processing through the codes. The 2234 1451 20 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2234 1451 21 ***************************************************************** 2234 1451 22 22-TI-49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 23 DIST-Feb02 2234 1451 24 ----BROND-2 MATERIAL 2234 2234 1451 25 -----INCIDENT NEUTRON DATA 2234 1451 26 ------ENDF-6 FORMAT 2234 1451 27 ------Russian Reactor Dosimetry File RRDF-2002 2234 1451 28 ***************************************************************** 2234 1451 29 Author of evaluation: K.I.Zolotarev 2234 1451 30 ***************************************************************** 2234 1451 31 MF= 3 2234 1451 32 MT= 28 -(n,np+pn+d) cross section data 2234 1451 33 -------------------------------------- 2234 1451 34 In this section is given the sum of cross section of the reac-2234 1451 35 tions Ti-49(n,np)Sc-48 , Ti-49(n,pn)Sc-48 and Ti-49(n,d)Sc-48. 2234 1451 36 Excitation function for the Ti-49(n,x)Sc-48 reaction in the 2234 1451 37 energy region from threshold to 20 MeV was evaluated by means of 2234 1451 38 statistical analysis of experimental cross section data [1-5] and 2234 1451 39 data from STAPRE [6] calculation. 2234 1451 40 Analysed experimental data were renormalized to the new stan- 2234 1451 41 dards for monitor reactions cross sections and decay data. Data 2234 1451 42 of Pai [2] measured in the energy region 16-19.5 MeV with using 2234 1451 43 Van de Graaff accelerator were renormalized to the results of 2234 1451 44 the theoretical model calculation. 2234 1451 45 The final procedure of evaluation Ti-49(n,x)Sc-48 excitation 2234 1451 46 function from threshold to 20 MeV has been carried out within the 2234 1451 47 framework of generalized least squares method. Rational function 2234 1451 48 was used as model function [7]. Calculations was performed by 2234 1451 49 means of Pade-2 code [8]. 2234 1451 50 U-235 thermal fission [9] and Cf-252 spontaneous fission 2234 1451 51 neutron spectra [10] averaged cross-sections calculated from the 2234 1451 52 evaluated Ti-49(n,x)Sc-48 excitation function are the following: 2234 1451 53 2234 1451 54 -------------------------------------------- 2234 1451 55 TYPE OF SPECTRUM I , mb (calc.) 2234 1451 56 --------------------------I----------------- 2234 1451 57 U-235 neutron fission I 1.0041E-3 2234 1451 58 CF-252 spontan. fission I 2.6070E-3 2234 1451 59 2234 1451 60 MF=33 2234 1451 61 MT= 28 -(n,np+pn+d) cross section cov. matrix 2234 1451 62 --------------------------------------------- 2234 1451 63 Uncertainties in the evaluated excitation function for the 2234 1451 64 reaction Ti-49(n,x)Sc-48 are given in the form of relative cova- 2234 1451 65 riance matrix for the 14-neutron energy groups (LB=5). Covariance 2234 1451 66 matrix of uncertainties was calculated simultaneously with 2234 1451 67 recommended cross section data by means of PADE-2 code. 2234 1451 68 Eigenvalues of the 6-th digits relative covariance matrix 2234 1451 69 given in the 33-file are the following: 2234 1451 70 2234 1451 71 2.34704E-08 3.44912E-08 5.30971E-08 9.47934E-08 2234 1451 72 1.99158E-07 5.01449E-07 1.49592E-06 5.55617E-06 2234 1451 73 1.56887E-04 2.23251E-03 8.77261E-03 1.50447E-02 2234 1451 74 3.12585E-02 8.36148E-02 2234 1451 75 2234 1451 76 References : 2234 1451 77 1. W.G.Cross, H.L.Pai Progress Report, EANDC(CAN)-16, p.1, 2234 1451 78 January 1963 2234 1451 79 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2234 1451 80 3. S.M.Qaim Nucl. Phys., v.A382, n.2, p.255, July 1982 2234 1451 81 4. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2234 1451 82 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2234 1451 83 6. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2234 1451 84 Activation Cross Section and Related Quantities, Report 2234 1451 85 IRK 76-01, Vienna, 1976 2234 1451 86 7. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2234 1451 87 st's Meeting on Evaluation and Processing of Covariance Data, 2234 1451 88 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2234 1451 89 8. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2234 1451 90 9. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2234 1451 91 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2234 1451 92 10. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2234 1451 93 ***************************************************************** 2234 1451 94 File 2 added to the pointwise file containing only the effective 2234 1451 95 scattering radius with no resonance parameters given. 2234 1451 96 Taken from JENDL-3.2 2234 1451 97 ***************************************************************** 2234 1451 98 ***************** Program LINEAR (VERSION 2012-1) ***************2234 1451 99 For All Data Greater than 1.0000D-10 barns in Absolute Value 2234 1451 100 Data Linearized to Within an Accuracy of .100000000 per-cent 2234 1451 101 ***************** Program GROUPIE (VERSION 2012-1) **************2234 1451 102 Unshielded Group Averages Using 640 Groups 2234 1451 103 Weighting Spectrum: Flat (Constant) Spectrum 2234 1451 104 1 451 98 12234 1451 105 2 151 4 12234 1451 106 8 5 2 12234 1451 107 10 5 34 12234 1451 108 40 5 27 12234 1451 109 2234 1 0 110 2234 0 0 111 22049.0000 48.5274000 0 0 1 02234 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