IAEA-NDS ENDF Download System, v-2008-07-22 0 0 0 0 3.40740E+4 7.32874E+1 0 0 1 13430 1451 1 0.0 0.0 0 0 0 03430 1451 2 0.0 0.0 0 0 48 23430 1451 3 34-Se- 74 4CCPFEI EVAL-NOV78 V.M.BYCHKOV,K.I.ZOLOTAREV ET AL.3430 1451 4 YK-3,(42),60 ,8109 DIST-JAN81 820212 3430 1451 5 IDENTIFIER=INDL 3430 REV1 3430 1451 6 FORMAT= ENDF/B-5 3430 1451 7 HISTORY= 8106 RECEIVED AT IAEA IN INTERNAL FORMAT 3430 1451 8 820212 DATA WERE CONVERTED FROM INTERNAL FORMAT TO 3430 1451 9 ENDF/B-5 FORMAT BY NDS IAEA.APPROVED BY AUTHOR. VP. 3430 1451 10 TEXT= 3430 1451 11 34-SE-74(N,P)33-AS-74 EN.REG. UP TO 20.MEV IN STEPS OF 100.KEV 3430 1451 12 THESE EVALUATED DATA ARE KNOWN AS BOSPOR-80 LIBRARY(THE LIBRARY 3430 1451 13 OF THRESHOLD REACTION CROSS-SECTIONS OF THE NUCLEAR DATA CENTRE, 3430 1451 14 OBNINSK,USSR). 3430 1451 15 THE EVAL.DATA OF BOSPOR-80 LIBRARY WERE OBTAINED AS A RESULT 3430 1451 16 OF CRITICAL ANALYSIS OF ALL THE DATA FROM EXFOR LIBRARY AND 3430 1451 17 CINDA REFERENCES ON THE BASIS OF STATISTICAL AND 3430 1451 18 PRE-EQUILIBRIUM MODELS OF NUCLEAR REACTIONS. 3430 1451 19 FOR A COMPLETE DISCUSSION OF THE EVALUATION METHOD SEE JADERNYE 3430 1451 20 KONSTANTY,32(1),PAGE 105,1979 3430 1451 21 (ENGLISH TRANSLATION INDC(CCP)-147,JUNE 1980). 3430 1451 22 BELOW ARE GIVEN THE VALUES OF EVALUATED CROSS-SECTIONS 3430 1451 23 AVERAGED ON DIFFERENT TYPES OF U-235 NEUTRON FISSION SPECTRA 3430 1451 24 INDUCED BY THERMAL NEUTRONS (IN MBARNS) 3430 1451 25 THESE SPECTRUM AVERAGED VALUES WERE COMPUTED AT FEI AS GIVEN IN 3430 1451 26 YAD. KOSTANTY 3,(42),P.60,(1981) 3430 1451 27 | TYPE OF SPECTRUM (SEE ZIJP W.L.,IN PROC.OF SYMP.3430 1451 28 | ON NUCLEAR DATA IN SC. AND TECHN.,PARIS,1973, 3430 1451 29 REACTION | V.2,P.271,IAEA,VIENNA) 3430 1451 30 | WATT,B.E. |CRANBERG,L. | LEACHMAN,R. 3430 1451 31 -------------|-------------|---------------|------------------ 3430 1451 32 34-SE-74(N,2N) | 6.6 | | 3430 1451 33 ***************** Program LINEAR (VERSION 2002-1) ***************3430 1451 34 For All Data Greater than 1.0000E-10 barns in Absolute Value 3430 1451 35 Data Linearized to Within an Accuracy of .100000000 per-cent 3430 1451 36 ***************** Program FIXUP (Version 2002-1) ****************3430 1451 37 Corrected ZA/AWR in All Sections-----------------------------Yes 3430 1451 38 Corrected Thresholds-----------------------------------------Yes 3430 1451 39 Extended Cross Sections to 20 MeV----------------------------No 3430 1451 40 Allow Cross Section Deletion---------------------------------No 3430 1451 41 Allow Cross Section Reconstruction---------------------------No 3430 1451 42 Make All Cross Sections Non-Negative-------------------------Yes 3430 1451 43 Delete Energies Not in Ascending Order-----------------------Yes 3430 1451 44 Deleted Duplicate Points-------------------------------------Yes 3430 1451 45 Check for Ascending MAT/MF/MT Order--------------------------Yes 3430 1451 46 Check for Legal MF/MT Numbers--------------------------------Yes 3430 1451 47 Allow Creation of Missing Sections---------------------------No 3430 1451 48 Allow Insertion of Energy Points-----------------------------No 3430 1451 49 Create Uniform Energy Grid-----------------------------------No 3430 1451 50 Delete Section if Cross Section =0 at All Energies-----------Yes 3430 1451 51 1 451 53 13430 1451 52 3 103 9 13430 1451 53 3430 1 0 54 3430 0 0 55 3.40740E+4 7.32874E+1 0 0 0 03430 3103 56 2.93000E+2-5.72000E+5 0 0 1 163430 3103 57 16 2 3430 3103 58 579800.000 0.0 4700000.00 0.0 5500000.00 .0040000003430 3103 59 6000000.00 .007000000 7000000.00 .014000000 8000000.00 .0270000003430 3103 60 9000000.00 .044000000 10000000.0 .065000000 12000000.0 .1110000003430 3103 61 13000000.0 .133000000 14000000.0 .145000000 15000000.0 .1480000003430 3103 62 16000000.0 .145000000 17000000.0 .141000000 19000000.0 .1290000003430 3103 63 20000000.0 .120000000 3430 3103 64 3430 3 0 65 3430 0 0 66 0 0 0 67 0.000000+0 0.000000+0 0 0 0 0 -1 0 0 0