IAEA-NDS ENDF Download System, v-2008-07-22 0 0 0 0 2.60560E+4 5.54544E+1 0 0 1 22620 1451 1 0.0 0.0 0 0 0 02620 1451 2 0.0 0.0 0 0 101 42620 1451 3 26-Fe- 56 4CCPFEI EVAL-NOV78 V.M.BYCHKOV,K.I.ZOLOTAREV ET AL.2620 1451 4 YK-3,(42),60 ,8109 DIST-JAN81 820212 2620 1451 5 IDENTIFIER=INDL 2620 REV2 2620 1451 6 FORMAT= ENDF/B-5 2620 1451 7 =============================================================== 2620 1451 8 ======================== For MF = 16 ======================== 2620 1451 9 2620 1451 10 HISTORY= 8106 RECEIVED AT IAEA IN INTERNAL FORMAT 2620 1451 11 820212 DATA WERE CONVERTED FROM INTERNAL FORMAT TO 2620 1451 12 ENDF/B-5 FORMAT BY NDS IAEA.APPROVED BY AUTHOR. VP. 2620 1451 13 (N,P) DATA TAKEN OVER FROM BOSPOR-78=INDL-2620,REV1 2620 1451 14 (N,2N) DATA ADDED FOR BOSPOR-80. VP. 2620 1451 15 TEXT= 2620 1451 16 26-FE-56(N,P )26-FE-55 EN.REGION UP TO 20.MEV IN STEPS OF 100.KEV 2620 1451 17 26-FE-56(N,2N)26-FE-55 EN.REGION UP TO 20.MEV IN STEPS OF 100.KEV 2620 1451 18 THESE EVALUATED DATA ARE KNOWN AS BOSPOR-80 LIBRARY(THE LIBRARY 2620 1451 19 OF THRESHOLD REACTION CROSS-SECTIONS OF THE NUCLEAR DATA CENTRE, 2620 1451 20 OBNINSK,USSR). 2620 1451 21 THE EVAL.DATA OF BOSPOR-80 LIBRARY WERE OBTAINED AS A RESULT 2620 1451 22 OF CRITICAL ANALYSIS OF ALL THE DATA FROM EXFOR LIBRARY AND 2620 1451 23 CINDA REFERENCES ON THE BASIS OF STATISTICAL AND 2620 1451 24 PRE-EQUILIBRIUM MODELS OF NUCLEAR REACTIONS. 2620 1451 25 FOR A COMPLETE DISCUSSION OF THE EVALUATION METHOD SEE JADERNYE 2620 1451 26 KONSTANTY,32(1),PAGE 105,1979 2620 1451 27 (ENGLISH TRANSLATION INDC(CCP)-147,JUNE 1980). 2620 1451 28 BELOW ARE GIVEN THE VALUES OF EVALUATED CROSS-SECTIONS 2620 1451 29 AVERAGED ON DIFFERENT TYPES OF U-235 NEUTRON FISSION SPECTRA 2620 1451 30 INDUCED BY THERMAL NEUTRONS (IN MBARNS) 2620 1451 31 THESE SPECTRUM AVERAGED VALUES WERE COMPUTED AT FEI AS GIVEN IN 2620 1451 32 YAD. KOSTANTY 3,(42),P.60,(1981) 2620 1451 33 | TYPE OF SPECTRUM (SEE ZIJP W.L.,IN PROC.OF SYMP.2620 1451 34 | ON NUCLEAR DATA IN SC. AND TECHN.,PARIS,1973, 2620 1451 35 REACTION | V.2,P.271,IAEA,VIENNA) 2620 1451 36 | WATT,B.E. |CRANBERG,L. | LEACHMAN,R. 2620 1451 37 -------------|-------------|---------------|------------------ 2620 1451 38 26-FE-56(N,2N)| 0.0754 | 0.0617 | 0.105 2620 1451 39 26-FE-56(N,P) | 1.08 | 1.00 | 1.08 2620 1451 40 =============================================================== 2620 1451 41 ======================== For MF = 103 ======================== 2620 1451 42 2.60560+ 4 5.54540+ 1 0 0 0 12620 1451 43 0.0 + 0 0.0 + 0 0 0 0 02620 1451 44 0.0 + 0 0.0 + 0 0 0 39 32620 1451 45 26-Fe- 56 ORNL EVAL-JUL78 C.Y.FU 2620 1451 46 DIST-AUG79 790612 2620 1451 47 EXPERIMENTAL DATA AVAILABLE SINCE THE EVALUATION OF FE-56 2620 1451 48 (N,P) CROSS SECTIONS BY DUDEY AND KENNERLEY(1) WERE REVIEWED. 2620 1451 49 SMITH AND MEADOWS(2) REPORTED NEW DATA FROM 4 TO 10 MEV, AN ENERGY2620 1451 50 RANGE MOST IMPORTANT FOR APPLICATION OF THIS REACTION TO FISSION 2620 1451 51 REACTOR DOSIMETRY. VONACH(3) HAS RECENTLY CALLED FOR ATTENTION 2620 1451 52 TO HIS DATA(4) FROM 13 TO 15 MEV. FURTHER CHECKS(3) RESULTED 2620 1451 53 IN A REDUCTION OF THE DATA ERRORS FROM 4% TO 2.5%. THE DATA HAD 2620 1451 54 NOT BEEN USED IN THE DUDEY AND KENNERLEY EVALUATION BUT ARE 2620 1451 55 2 TO 3% LOWER THAN THE EVALUATED VALUES. THREE SINGLE-ENERGY 2620 1451 56 MEASUREMENTS(5-7) NEAR 14.6 MEV HAVE BEEN REPORTED. THE ONE 2620 1451 57 BY DYER(5) IS 1% HIGHER THAN THE ENDF/B-IV VALUE, ROBERTSON 7% 2620 1451 58 HIGHER, AND QAIM 4% LOWER. 2620 1451 59 INCLUSION OF THE NEW DATA IN A SUBJECTIVE ANALYSIS RESULTS 2620 1451 60 IN UP TO 10% REDUCTION FROM THE ENDF/B-IV VALUES FROM THRESHOLD TO2620 1451 61 6 MEV, 2-3% REDUCTION FORM 6 TO 10 MEV, AND 1-2% REDUCTION FROM 2620 1451 62 13-15 MEV. THE DATA OF SMITH AND MEADOWS ARE RELATIVE TO ENDF/B-42620 1451 63 U-238(N,F) CROSS SECTIONS AND HAD BEEN RENORMALIZED TO ENDF/B-V 2620 1451 64 VALUES BEFORE BEING USED IN THE PRESENT EVALUATION. THE EFFECT 2620 1451 65 ON THE CALCULATED FISSION SPECTRUM AVERAGE IS ESTIMATED TO BE 2620 1451 66 A REDUCTION OF ABOUT 3% FROM ENDF/B-IV VALUE. 2620 1451 67 REFERENCES: 2620 1451 68 1. N. D. DUDEY AND R. KENNERLEY, "EVALUATION OF FE-56(N,P)MN-56 2620 1451 69 CROSS SECTIONS FOR ENDF/B-III," IN ENDF/B-IV DOSIMETRY FILE, 2620 1451 70 EDITED BY B. A. MAGURNO, BNL-NCS-50446(1975). 2620 1451 71 2. D. L. SMITH AND J. W. MEADOWS, NUCL. SCI. ENG. 58,314(1975). 2620 1451 72 3. H. VONACH,"STATUS OF SOME ACTIVATION CROSS SECTIONS FOR 2620 1451 73 REACTOR NEUTRON DOSIMETRY IN THE RANGE 13-15 MEV," P. 361, 2620 1451 74 IAEA-208(1975). 2620 1451 75 4. H. VONACH ET AL., CONF. NUCL. CROSS SECTIONS AND TECH., 2620 1451 76 WASHINGTON, NBS-SP-299, VOL. 1, P. 885(1968). 2620 1451 77 5. N. C. DYER AND J. H. HAMILTON, J. INORG. NUCL. CHEM. 34, 1119 2620 1451 78 (1972). 2620 1451 79 6. J. C. ROBERTSON ET AL., J. NUCL. EN. 27, 139(1972). 2620 1451 80 7. S. M. QAIM ET AL., CONF. ON CHEM. NUCL. DATA, CANTERBURY, 2620 1451 81 P. 121(1971). 2620 1451 82 ***************** PROGRAM LINEAR (VERSION 82-1) *****************2620 1451 83 DATA LINEARIZED USING A MAXIMUM ERROR OF 0.100 PER-CENT 2620 1451 84 2620 1451 85 =============================================================== 2620 1451 86 ***************** Program LINEAR (VERSION 2002-1) ***************2620 1451 87 For All Data Greater than 1.0000E-10 barns in Absolute Value 2620 1451 88 Data Linearized to Within an Accuracy of .100000000 per-cent 2620 1451 89 ***************** Program FIXUP (Version 2002-1) ****************2620 1451 90 Corrected ZA/AWR in All Sections-----------------------------Yes 2620 1451 91 Corrected Thresholds-----------------------------------------Yes 2620 1451 92 Extended Cross Sections to 20 MeV----------------------------No 2620 1451 93 Allow Cross Section Deletion---------------------------------No 2620 1451 94 Allow Cross Section Reconstruction---------------------------No 2620 1451 95 Make All Cross Sections Non-Negative-------------------------Yes 2620 1451 96 Delete Energies Not in Ascending Order-----------------------Yes 2620 1451 97 Deleted Duplicate Points-------------------------------------Yes 2620 1451 98 Check for Ascending MAT/MF/MT Order--------------------------Yes 2620 1451 99 Check for Legal MF/MT Numbers--------------------------------Yes 2620 1451 100 Allow Creation of Missing Sections---------------------------No 2620 1451 101 Allow Insertion of Energy Points-----------------------------No 2620 1451 102 Create Uniform Energy Grid-----------------------------------No 2620 1451 103 Delete Section if Cross Section =0 at All Energies-----------Yes 2620 1451 104 1 451 108 12620 1451 105 3 16 14 22620 1451 106 3 103 77 12620 1451 107 33 103 32 12620 1451 108 2620 1 0 109 2620 0 0 110 2.60560E+4 5.54544E+1 0 0 0 02620 3 16 111 2.93000E+2-1.12000E+7 0 0 1 332620 3 16 112 33 2 2620 3 16 113 11401967.7 0.0 11700000.0 .020000000 11900000.0 .0440000002620 3 16 114 12000000.0 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