IAEA-NDS ENDF Download System, v-2008-07-22 0 0 0 0 2.40500E+4 4.95170E+1 0 0 1 12440 1451 1 0.0 0.0 0 0 0 02440 1451 2 0.0 0.0 0 0 48 22440 1451 3 24-Cr- 50 4CCPFEI EVAL-NOV78 V.M.BYCHKOV,K.I.ZOLOTAREV ET AL.2440 1451 4 YK-3,(42),60 ,8109 DIST-JAN81 820212 2440 1451 5 IDENTIFIER=INDL 2440 REV1 2440 1451 6 FORMAT= ENDF/B-5 2440 1451 7 HISTORY= 8106 RECEIVED AT IAEA IN INTERNAL FORMAT 2440 1451 8 820212 DATA WERE CONVERTED FROM INTERNAL FORMAT TO 2440 1451 9 ENDF/B-5 FORMAT BY NDS IAEA.APPROVED BY AUTHOR. VP. 2440 1451 10 TEXT= 2440 1451 11 24-CR-50(N,2N)24-CR-49 EN.REG. UP TO 20.MEV IN STEPS OF 100.KEV 2440 1451 12 THESE EVALUATED DATA ARE KNOWN AS BOSPOR-80 LIBRARY(THE LIBRARY 2440 1451 13 OF THRESHOLD REACTION CROSS-SECTIONS OF THE NUCLEAR DATA CENTRE, 2440 1451 14 OBNINSK,USSR). 2440 1451 15 THE EVAL.DATA OF BOSPOR-80 LIBRARY WERE OBTAINED AS A RESULT 2440 1451 16 OF CRITICAL ANALYSIS OF ALL THE DATA FROM EXFOR LIBRARY AND 2440 1451 17 CINDA REFERENCES ON THE BASIS OF STATISTICAL AND 2440 1451 18 PRE-EQUILIBRIUM MODELS OF NUCLEAR REACTIONS. 2440 1451 19 FOR A COMPLETE DISCUSSION OF THE EVALUATION METHOD SEE JADERNYE 2440 1451 20 KONSTANTY,32(1),PAGE 105,1979 2440 1451 21 (ENGLISH TRANSLATION INDC(CCP)-147,JUNE 1980). 2440 1451 22 BELOW ARE GIVEN THE VALUES OF EVALUATED CROSS-SECTIONS 2440 1451 23 AVERAGED ON DIFFERENT TYPES OF U-235 NEUTRON FISSION SPECTRA 2440 1451 24 INDUCED BY THERMAL NEUTRONS (IN MBARNS) 2440 1451 25 THESE SPECTRUM AVERAGED VALUES WERE COMPUTED AT FEI AS GIVEN IN 2440 1451 26 YAD. KOSTANTY 3,(42),P.60,(1981) 2440 1451 27 | TYPE OF SPECTRUM (SEE ZIJP W.L.,IN PROC.OF SYMP.2440 1451 28 | ON NUCLEAR DATA IN SC. AND TECHN.,PARIS,1973, 2440 1451 29 REACTION | V.2,P.271,IAEA,VIENNA) 2440 1451 30 | WATT,B.E. |CRANBERG,L. | LEACHMAN,R. 2440 1451 31 -------------|-------------|---------------|------------------ 2440 1451 32 24-CR-50(N,2N)| 1.8E-03 | 1.4E-03 | 2.8E-03 2440 1451 33 ***************** Program LINEAR (VERSION 2002-1) ***************2440 1451 34 For All Data Greater than 1.0000E-10 barns in Absolute Value 2440 1451 35 Data Linearized to Within an Accuracy of .100000000 per-cent 2440 1451 36 ***************** Program FIXUP (Version 2002-1) ****************2440 1451 37 Corrected ZA/AWR in All Sections-----------------------------Yes 2440 1451 38 Corrected Thresholds-----------------------------------------Yes 2440 1451 39 Extended Cross Sections to 20 MeV----------------------------No 2440 1451 40 Allow Cross Section Deletion---------------------------------No 2440 1451 41 Allow Cross Section Reconstruction---------------------------No 2440 1451 42 Make All Cross Sections Non-Negative-------------------------Yes 2440 1451 43 Delete Energies Not in Ascending Order-----------------------Yes 2440 1451 44 Deleted Duplicate Points-------------------------------------Yes 2440 1451 45 Check for Ascending MAT/MF/MT Order--------------------------Yes 2440 1451 46 Check for Legal MF/MT Numbers--------------------------------Yes 2440 1451 47 Allow Creation of Missing Sections---------------------------No 2440 1451 48 Allow Insertion of Energy Points-----------------------------No 2440 1451 49 Create Uniform Energy Grid-----------------------------------No 2440 1451 50 Delete Section if Cross Section =0 at All Energies-----------Yes 2440 1451 51 1 451 53 12440 1451 52 3 16 10 12440 1451 53 2440 1 0 54 2440 0 0 55 2.40500E+4 4.95170E+1 0 0 0 02440 3 16 56 2.93000E+2-1.29400E+7 0 0 1 202440 3 16 57 20 2 2440 3 16 58 13201324.4 0.0 13300000.0 .000500000 13600000.0 .0035000002440 3 16 59 13800000.0 .006500000 13900000.0 .008500000 14000000.0 .0100000002440 3 16 60 14300000.0 .016400000 14600000.0 .022720000 15000000.0 .0310000002440 3 16 61 15200000.0 .035440000 15400000.0 .039660000 15600000.0 .0436600002440 3 16 62 15800000.0 .047440000 16000000.0 .051000000 16600000.0 .0598200002440 3 16 63 17400000.0 .071060000 18000000.0 .079000000 19100000.0 .0925200002440 3 16 64 19600000.0 .098750000 20000000.0 .103300000 2440 3 16 65 2440 3 0 66 2440 0 0 67 0 0 0 68 0.000000+0 0.000000+0 0 0 0 0 -1 0 0 0