ENTRY 21390 20161001 22532139000000001 SUBENT 21390001 20161001 22532139000100001 BIB 12 21 2139000100002 INSTITUTE (2BLGGHT) 2139000100003 REFERENCE (J,RCA,10,156,1968) 2139000100004 AUTHOR (P.De Regge,R.Dams,J.Hoste) 2139000100005 TITLE The Sc-45(n,p)Ca-44 and Sc-45(n,alpha)K-42 mean 2139000100006 fission cross sections 2139000100007 FACILITY (REAC,2BLGGHT) BR-1 reactor. 2139000100008 INC-SOURCE (REAC) Equivalent fission spectrum. 2139000100009 INC-SPECT Fission neutron spectrum from U-235. 2139000100010 METHOD (ACTIV) Activation, followed by chemical separation 2139000100011 of reaction products. 2139000100012 DETECTOR (SCIN) Liquid scintillator and 4 Pi beta counting 2139000100013 for calcium activity. 2139000100014 (GEMUC) Calibrated Geiger-Muller tube and gamma 2139000100015 spectroscopy for potassium activity. 2139000100016 MONITOR ((MONIT1)16-S-32(N,P)15-P-32,,SIG,,FIS) Fission 2139000100017 spectrum average taken as 60 mb. 2139000100018 ERR-ANALYS (DATA-ERR) Standard deviation for average of several 2139000100019 independent irradiations. 2139000100020 HISTORY (19800229C) 2139000100021 (20161001A) SD:Updated to new date formats,lower case. 2139000100022 BIB update. 2139000100023 ENDBIB 21 0 2139000100024 COMMON 2 3 2139000100025 EN-DUMMY MONIT1 2139000100026 MEV MB 2139000100027 1.5 60. 2139000100028 ENDCOMMON 3 0 2139000100029 ENDSUBENT 28 0 2139000199999 SUBENT 21390002 20161001 22532139000200001 BIB 6 10 2139000200002 REACTION (21-SC-45(N,P)20-CA-45,,SIG,,FIS) Fission spectrum 2139000200003 average cross section. 2139000200004 INC-SPECT Thermal neutron flux of 4.x 10**11 n/sec/cm2. 2139000200005 10 daily irradiations of 2 hr 2139000200006 SAMPLE 100 mg high purity SC2O3 and 100 mg (NH4)2SO4 were 2139000200007 weighed and sealed in separate silica tubes. Both 2139000200008 samples were packed in a standard aluminum container 2139000200009 METHOD (CHSEP) 2139000200010 DECAY-DATA (20-CA-45,165.D,B-) 2139000200011 STATUS (TABLE) Table 3 from RCA,10,156,1968 2139000200012 ENDBIB 10 0 2139000200013 NOCOMMON 0 0 2139000200014 DATA 2 1 2139000200015 DATA DATA-ERR 2139000200016 MB MB 2139000200017 34.6 0.5 2139000200018 ENDDATA 3 0 2139000200019 ENDSUBENT 18 0 2139000299999 SUBENT 21390003 20161001 22532139000300001 BIB 10 30 2139000300002 REACTION (21-SC-45(N,A)19-K-42,,SIG,,FIS) Fission spectrum 2139000300003 average cross section. 2139000300004 MONITOR ((MONIT2)19-K-41(N,G)19-K-42,,SIG) 2139000300005 ((MONIT3)15-P-31(N,G)15-P-32,,SIG) 2139000300006 MONIT-REF ((MONIT2),D.J.Hughes+,R,BNL-325,1960) Suppl. 2139000300007 ((MONIT3),D.J.Hughes+,R,BNL-325,1958) 2139000300008 DECAY-DATA (19-K-42,12.36HR,DG) 2139000300009 INC-SPECT The thermal neutron flux was determined by irradiating 2139000300010 NH4H2PO4 with and without cadmium cover to correct for 2139000300011 the activity produced by epi-cadmium neutrons. 2139000300012 SAMPLE Two samples of 20 mg Sc2O3 and 4 mg K2CO3, one of each 2139000300013 with cadmium cover, were irradiated for 2 h together 2139000300014 with 100 mg (NH4)2SO4 and 40 mg NH4H2PO4 as fast and 2139000300015 thermal flux monitors respectively. 2139000300016 ANALYSIS Cross section was determined using two methods: 2139000300017 1. Determination of the mean fission neutron cross- 2139000300018 section on the basis of fluence ratios and cross- 2139000300019 section ratios for fission and thermal neutrons 2139000300020 2. Determination of the mean fission neutron cross- 2139000300021 section on the basis of absolute fission neutron 2139000300022 fluence measurements and absolute activity 2139000300023 determinations 2139000300024 The agreement of the results obtained by both methods 2139000300025 is not merely a check but illustrates also the relative2139000300026 accuracy of the reference cross sections used. 2139000300027 CORRECTION .Corrected for thermal neutrons by comparing 2139000300028 monitor irradiations with and without cadmium 2139000300029 cover. 2139000300030 STATUS (TABLE) Table 4 from RCA,10,156,1968 2139000300031 HISTORY (20161001A) SD: MONITOR was added. BIB update. 2139000300032 ENDBIB 30 0 2139000300033 NOCOMMON 0 0 2139000300034 DATA 4 1 2139000300035 DATA DATA-ERR MONIT2 MONIT3 2139000300036 MB MB B B 2139000300037 0.158 0.004 1.30 0.190 2139000300038 ENDDATA 3 0 2139000300039 ENDSUBENT 38 0 2139000399999 ENDENTRY 3 0 2139099999999