ENTRY 20593 20211219 23062059300000001 SUBENT 20593001 20211219 23062059300100001 BIB 18 43 2059300100002 TITLE Measurement of average cross sections for some 2059300100003 threshold reactions by means of a small fission 2059300100004 foil in large thermal neutron field 2059300100005 AUTHOR (I.Kimura,K.Kobayashi,T.Shibata) 2059300100006 INSTITUTE (2JPNKTO) 2059300100007 REFERENCE (J,NST,10,574,1973) Numerical values and comparisons. 2059300100008 FACILITY (REAC,2JPNKTO) Kyoto university reactor (KUR). 2059300100009 INC-SOURCE Fission neutrons induced in U-235 foil inside 2059300100010 thermal column of the KUR reactor. 2059300100011 Fission foil 0.5 mm thickness,1 cm diameter 2059300100012 compossed of a 90% enriched U-Al alloy. 2059300100013 INC-SPECT U-235 fission spectrum. 2059300100014 SAMPLE Threshold foils surrounding fission foil, no 2059300100015 details given. 2059300100016 METHOD (ACTIV) Activation, with simultaneous measurement 2059300100017 of fission neutron number by detection of the 2059300100018 1.596 MeV gamma-ray from La-140 produced by decay 2059300100019 of the fission product Ba-140. 2059300100020 DETECTOR (GELI) Ge-Li detector whose detection 2059300100021 efficiency had previously been calibrated. 2059300100022 ANALYSIS Shape factor for neutron absorption in threshold 2059300100023 foil calculated by two methods, analytical formula, 2059300100024 formulae (4)-(8) of Nucl.Sci.Techn.10,574(1973), 2059300100025 and Monte Carlo method. 2059300100026 MONITOR Absolute. 2059300100027 COMMENT Results lower than those proposed by R.McElroy,et al. 2059300100028 Nucl.Sci.Eng.48,51(1972). 2059300100029 CORRECTION Background 2059300100030 FLAG (1.) Shape factor calculated analyticaly. 2059300100031 (2.) Shape factor calculated by Monte Carlo. 2059300100032 ERR-ANALYS (ERR-T) Total error 2059300100033 (ERR-S,1.,5.) Statistical errors 2059300100034 (ERR-1) Uncertainty in detector efficiency 2059300100035 (ERR-2,1.,2.) Uncertainty in position of irradiation 2059300100036 (ERR-3,,1.) Uncertainty in scattering and absorption 2059300100037 of neutrons in a sample foil 2059300100038 (ERR-4,,0.4) Uncertainty in scattering and absorption 2059300100039 of neutrons and second fission in a U-Al foil 2059300100040 STATUS (TABLE) Table 2 of Nucl.Sci.Techn.10(1973)574 2059300100041 HISTORY (19760802C) 2059300100042 (19840106U) A.P.T. Japan code changed to JPN 2059300100043 (20211219A) SD:Updated to new date formats,lower case. 2059300100044 ERR-ANALYS updated. Corrections in all Subents. 2059300100045 ENDBIB 43 0 2059300100046 COMMON 2 3 2059300100047 EN-DUMMY ERR-1 2059300100048 MEV PER-CENT 2059300100049 1.5 1.5 2059300100050 ENDCOMMON 3 0 2059300100051 ENDSUBENT 50 0 2059300199999 SUBENT 20593002 20211219 23062059300200001 BIB 2 2 2059300200002 REACTION (13-AL-27(N,A)11-NA-24,,SIG,,FIS) 2059300200003 HISTORY (20211219U) SD: DATA-ERR -> ERR-T. Cosmetic corrections2059300200004 ENDBIB 2 0 2059300200005 NOCOMMON 0 0 2059300200006 DATA 3 2 2059300200007 DATA ERR-T FLAG 2059300200008 MB MB NO-DIM 2059300200009 0.635 0.036 1. 2059300200010 0.653 0.036 2. 2059300200011 ENDDATA 4 0 2059300200012 ENDSUBENT 11 0 2059300299999 SUBENT 20593003 20211219 23062059300300001 BIB 2 2 2059300300002 REACTION (28-NI-58(N,P)27-CO-58,,SIG,,FIS) 2059300300003 HISTORY (20211219U) SD: DATA-ERR -> ERR-T. Cosmetic corrections2059300300004 ENDBIB 2 0 2059300300005 NOCOMMON 0 0 2059300300006 DATA 3 2 2059300300007 DATA ERR-T FLAG 2059300300008 MB MB NO-DIM 2059300300009 102.6 5.2 1. 2059300300010 101.6 5.6 2. 2059300300011 ENDDATA 4 0 2059300300012 ENDSUBENT 11 0 2059300399999 SUBENT 20593004 20211219 23062059300400001 BIB 2 4 2059300400002 REACTION (49-IN-115(N,INL)49-IN-115-M,,SIG,,FIS) 2059300400003 HISTORY (20211219A) SD: SF1=In-115m1 -> In-115 in REACTION code2059300400004 DATA-ERR -> ERR-T. HALF-LIFE deleted (not given in the 2059300400005 article). 2059300400006 ENDBIB 4 0 2059300400007 NOCOMMON 0 0 2059300400008 DATA 3 2 2059300400009 DATA ERR-T FLAG 2059300400010 MB MB NO-DIM 2059300400011 176.5 8.8 1. 2059300400012 177.4 9.8 2. 2059300400013 ENDDATA 4 0 2059300400014 ENDSUBENT 13 0 2059300499999 ENDENTRY 4 0 2059399999999