10002111111001 (col. 11 = 1 = allow MT reconstruction) FIXUP.IN FIXUP.OUT 0.0 0.0 0 0 0 1 (total) 0.0 0.0 0 0 0.0 0.0 0 0 0 4 (total inelastic) 0.0 0.0 0 0 0.0 0.0 0 0 0 16 (total n,2n) 0.0 0.0 0 0 0.0 0.0 0 0 0103 (total n,p) 0.0 0.0 0 0 0.0 0.0 0 0 0104 (total n,d) 0.0 0.0 0 0 0.0 0.0 0 0 0105 (total n,t) 0.0 0.0 0 0 0.0 0.0 0 0 0106 (total n,He-3) 0.0 0.0 0 0 0.0 0.0 0 0 0107 (total n,alpha) 0.0 0.0 0 0 === This and the below lines are NOT read as input, so you can store === === anything you want below. Here I have defined how the input data === === is interpreted = so you know what each input field means. === ------------------------------------------------------------------------------ Interpretation of Input Test/Correction Options ------------------------------------------------------------------------------ Correct ZA/AWR in All Sections---------- Yes Correct Thresholds---------------------- No Extend Cross Section to 20 MeV---------- No Allow Cross Section Deletion------------ No Allow Cross Section Reconstruction------ Yes (Use Built-in Table) Make All Cross Sections Non-Negative---- Yes Delete Energies Not in Ascending Order-- Yes Delete Duplicate Points----------------- Yes Check for Ascending MAT/MF/MT Order----- Yes Check for Legal MF/MT Numbers----------- Yes Allow Creation of Missing Sections------ Yes Allow Insertion of Energy Points-------- No Uniform Energy Grid for ALL MT---------- No Delete Section if Cross Section =0------ Yes ------------------------------------------------------------------------------ ENDF/B Input and Output Data FilenameS FIXUP.IN FIXUP.OUT ------------------------------------------------------------------------------ Built-in Summation/Deletion/Threshold Exclusion Rules ------------------------------------------------------------------------------ MT = MT Ranges and Messages ------------------------------------------------------------------------------ ------------------------------------------------------------------------------ Neutrons - ENDF/B-V and Earlier Versions of ENDF/B ------------------------------------------------------------------------------ 4 =+( 51, 91) 103 =+(700,718) 104 =+(720,738) 105 =+(740,758) 106 =+(760,778) 107 =+(780,798) 16 =+(875,891) 101 =+(102,114) 18 =+( 19, 19)+( 20, 21)+( 38, 38) 27 =+( 18, 18)+(101,101) 3 =+( 4, 4)+( 6, 9)+( 16, 17)+( 22, 37) 19 =+( 18, 18)-( 20, 21)-( 38, 38) 1 =+( 2, 3) ------------------------------------------------------------------------------ Neutrons - ENDF/B-VI ------------------------------------------------------------------------------ 4 =+( 50, 91) 103 =+(600,649) 104 =+(650,699) 105 =+(700,749) 106 =+(750,799) 107 =+(800,849) 16 =+(875,891) 101 =+(102,117) 18 =+( 19, 19)+( 20, 21)+( 38, 38) 27 =+( 18, 18)+(101,101) 3 =+( 4, 5)+( 11, 17)+( 22, 37)+( 41, 45) 19 =+( 18, 18)-( 20, 21)-( 38, 38) 1 =+( 2, 3) ------------------------------------------------------------------------------ Photons - ENDF/B-VI ------------------------------------------------------------------------------ 516 =+(515,515)+(517,517) 522 =+(534,572) 501 =+(502,502)+(504,504)+(516,516)+(522,522) ------------------------------------------------------------------------------ If Not Present the Following Sections will be Created ------------------------------------------------------------------------------ C1 C2 L1 L2 MAT MT ------------------------------------------------------------------------------ 0.0 0.0 0 0 0 1 0.0 0.0 0 0 0 1 0.0 0.0 0 0 0 4 0.0 0.0 0 0 0 4 0.0 0.0 0 0 0 16 0.0 0.0 0 0 0 16 0.0 0.0 0 0 0 103 0.0 0.0 0 0 0 103 0.0 0.0 0 0 0 104 0.0 0.0 0 0 0 104 0.0 0.0 0 0 0 105 0.0 0.0 0 0 0 105 0.0 0.0 0 0 0 106 0.0 0.0 0 0 0 106 0.0 0.0 0 0 0 107 0.0 0.0 0 0 0 107 ------------------------------------------------------------------------------