ENTRY 10921 20091002 13571092100000001 SUBENT 10921001 20091002 13571092100100001 BIB 15 73 1092100100002 INSTITUTE (1USAANL) 1092100100003 REFERENCE (J,NSE,76,30,198010) 1092100100004 AUTHOR (G.Winkler,D.L.Smith,J.W.Meadows) 1092100100005 TITLE Measurement of cross sections for the 1092100100006 63Cu(n,alpha)Co60 reaction from threshold to 10-MeV 1092100100007 FACILITY (DYNAM,1USAANL) Argonne fast neutron generator. 1092100100008 SAMPLE Natural copper, 2.54-cm diameter, 0.635-cm thick. 1092100100009 Cobalt impurity less than one part per million. 1092100100010 INC-SPECT Energy resolution affected by source target thickness, 1092100100011 source characteristics, and kinematic effects 1092100100012 DETECTOR (NAICR) Pair of NaI(Tl) detectors, one on either side 1092100100013 of sample, dimensions 7.62 by 7.62-cm, surrounded 1092100100014 by 5-cm lead shield in addition to normal shielding 1092100100015 of low-level counting room. Third detector used to 1092100100016 monitor background. 1092100100017 (LONGC) Long counter used to measure relative neutron 1092100100018 flux. 1092100100019 (IOCH) Methane-filled low-mass ionization chamber with 1092100100020 deposits of 238U used to determine neutron fluence by 1092100100021 detection of fission fragments. 1092100100022 (GELI) Germanium-lithium detector used for 60Co 1092100100023 measurements in counting efficiency calibration and 1092100100024 purity check of sample and gamma spectra. 1092100100025 MONITOR Source reactions used to calibrate accelerator energy 1092100100026 scale. 1092100100027 Standard 60Co sources used in counting efficiency 1092100100028 calibration. 1092100100029 2(92-U-238(N,F),,SIG) Taken from ENDF-B/V. 1092100100030 Monitor energy resolution is +-1/2 FWHM. 1092100100031 DECAY-DATA (27-CO-60-G,5.271YR,B) 1092100100032 METHOD (ACTIV) Activation analysis. Sample attached to outer 1092100100033 wall of ionization chamber. Checks made to assure that1092100100034 no activity was transferred to sample from ionization 1092100100035 chamber wall. Counting efficiency determined in 1092100100036 separate experiment using cobalt samples and standard 1092100100037 60Co sources. 1092100100038 CORRECTION Corrections made for: 1092100100039 .neutrons produced in non-primary source reactions, 1092100100040 .neutron scattering in Ta cup,p-Li7 source,or 1092100100041 deuteron gas cell components, 1092100100042 .background from deuterium gas cell, 1092100100043 .contributions from minority isotopes of uranium, 1092100100044 .extrapolation-to-zero correction (for fissions), 1092100100045 .fission products not escaping from deposit, 1092100100046 .elastic and inelastic scattering of neutrons in 1092100100047 sample and from ionization chamber, 1092100100048 .neutron absorption, 1092100100049 .geometry, 1092100100050 .source characteristics, 1092100100051 .effects of inhomogeneous activation of sample on the 1092100100052 gamma counting efficiency. 1092100100053 ERR-ANALYS (ERR-T) total error 1092100100054 Partial errors and their sources: 1092100100055 (ERR-S) statistical error 1092100100056 (ERR-1,,,1.0) Gamma-ray detection efficiency. 1092100100057 (ERR-2,,,0.0) Irradiation geometry. 1092100100058 (ERR-3,,,1.0) Uranium deposit,mass,and isotope content.1092100100059 (ERR-4,,,1.0) Extrapolation correction for fissions and1092100100060 correction for finite thickness of deposit. 1092100100061 (ERR-5,,,1.0) Correction for neutron absorption in 1092100100062 Cu sample. 1092100100063 (ERR-6,,,1.0) Error in correction for neutron 1092100100064 scattering by the sample and fission chamber 1092100100065 components. 1092100100066 (ERR-7,,,0.5) Neutron source characteristics. 1092100100067 (ERR-8,,,0.0) Correction of fissions for neutron 1092100100068 background and neutron scattering due to tantalum 1092100100069 cup. 1092100100070 STATUS (APRVD) Approved by G. Winkler, 1982/10/13. 1092100100071 HISTORY (19801118C) 1092100100072 (19820514A) BIB, COMMON corrections. 1092100100073 (19850708A) Updated error analysis, BIB corrections. 1092100100074 (20081031A) BIB section updated. Error formats updated.1092100100075 ENDBIB 73 0 1092100100076 COMMON 1 3 1092100100077 ERR-1 1092100100078 PER-CENT 1092100100079 1.5 1092100100080 ENDCOMMON 3 0 1092100100081 ENDSUBENT 80 0 1092100199999 SUBENT 10921002 20091002 13571092100200001 BIB 4 7 1092100200002 REACTION 1((29-CU-63(N,A)27-CO-60,,SIG)/(92-U-238(N,F),,SIG)) 1092100200003 2(29-CU-63(N,A)27-CO-60,,SIG) 1092100200004 INC-SOURCE (P-LI7) 1092100200005 ERR-ANALYS Uncertainty in neutron energy about 10 keV. 1092100200006 2(MONIT-ERR,,,0.5) Error in 238U standard fission cross 1092100200007 section. 1092100200008 HISTORY (20081031A) Errors updated. 1092100200009 ENDBIB 7 0 1092100200010 COMMON 3 3 1092100200011 ERR-6 ERR-7 MONIT-ERR 2 1092100200012 PER-CENT PER-CENT PER-CENT 1092100200013 2.2 1.5 4. 1092100200014 ENDCOMMON 3 0 1092100200015 DATA 15 8 1092100200016 EN EN-RSL-HW DATA 1ERR-T 1ERR-S ERR-2 1092100200017 ERR-3 ERR-4 ERR-5 ERR-8 DATA 2ERR-T 21092100200018 EN-NRM 2EN-NRM-RSL2MONIT 2 1092100200019 MEV MEV NO-DIM PER-CENT PER-CENT PER-CENT 1092100200020 PER-CENT PER-CENT PER-CENT PER-CENT MB MB 1092100200021 MEV MEV B 1092100200022 3.560 0.044 7.27 -05 50. 50. 3.0 1092100200023 1.0 0.5 1.8 0.039 0.020 1092100200024 3.570 0.043 0.5365 1092100200025 3.800 0.081 2.42 -04 13. 12. 3.3 1092100200026 1.0 0.2 1.8 0.132 0.018 1092100200027 3.818 0.062 0.5446 1092100200028 3.800 0.082 1.84 -04 22. 21. 2.6 1092100200029 1.0 0.2 1.8 0.100 0.022 1092100200030 3.818 0.063 0.5446 1092100200031 4.065 0.041 5.142 -04 6.5 4.7 2.0 1092100200032 1.0 1.5 1.8 0.281 0.021 1092100200033 4.072 0.040 0.5465 1092100200034 4.361 0.041 9.769 -04 5.7 3.8 2.0 1092100200035 1.0 0.7 1.8 0.1 0.536 0.037 1092100200036 4.368 0.038 0.5486 1092100200037 4.656 0.042 1.475 -03 5.6 3.6 2.0 1092100200038 1.0 1.0 1.8 0.1 0.802 0.055 1092100200039 4.663 0.038 0.5440 1092100200040 4.954 0.043 2.409 -03 5.0 2.4 2.0 1092100200041 1.0 1.2 1.9 0.1 1.288 0.083 1092100200042 4.960 0.038 0.5347 1092100200043 5.268 0.045 3.912 -03 8.7 7.5 1.7 1092100200044 1.3 1.0 1.9 0.3 2.12 0.20 1092100200045 5.277 0.041 0.5412 1092100200046 ENDDATA 30 0 1092100200047 ENDSUBENT 46 0 1092100299999 SUBENT 10921003 20091002 13571092100300001 BIB 4 10 1092100300002 REACTION 1((29-CU-63(N,A)27-CO-60,,SIG)/(92-U-238(N,F),,SIG)) 1092100300003 2(29-CU-63(N,A)27-CO-60,,SIG) 1092100300004 INC-SOURCE (D-D) 1092100300005 ERR-ANALYS Uncertainty in neutron energy: 1092100300006 .about 22 keV from 5.1-5.6 MeV, 1092100300007 .about 20 keV from 5.6-6.6 MeV, 1092100300008 .about 15 keV from 6.9-10 MeV. 1092100300009 (ERR-9,,,0.5) Correction for activity induced in sample1092100300010 by neutron background (empty gas cell). 1092100300011 HISTORY (20081031A) Errors updated. 1092100300012 ENDBIB 10 0 1092100300013 COMMON 1 3 1092100300014 MONIT-ERR 2 1092100300015 PER-CENT 1092100300016 4. 1092100300017 ENDCOMMON 3 0 1092100300018 DATA 18 21 1092100300019 EN EN-RSL-HW DATA 1ERR-T 1ERR-S ERR-2 1092100300020 ERR-3 ERR-4 ERR-5 ERR-6 ERR-7 ERR-8 1092100300021 ERR-9 DATA 2ERR-T 2EN-NRM 2EN-NRM-RSL2MONIT 21092100300022 MEV MEV NO-DIM PER-CENT PER-CENT PER-CENT 1092100300023 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT 1092100300024 PER-CENT MB MB MEV MEV B 1092100300025 5.120 0.120 3.737 -03 7.9 6.5 1.7 1092100300026 1.3 1.0 1.9 2.5 0.5 0.8 1092100300027 1.0 2.01 0.18 5.128 0.115 0.5370 1092100300028 5.185 0.122 3.888 -03 7.9 6.7 2.0 1092100300029 1.0 0.2 1.9 2.2 0.5 0.4 1092100300030 1.0 2.10 0.19 5.195 0.112 0.5389 1092100300031 5.455 0.104 5.520 -03 8.1 6.8 1.7 1092100300032 1.3 1.0 1.9 2.5 0.5 0.8 1092100300033 1.0 3.02 0.27 5.463 0.102 0.5464 1092100300034 5.677 0.104 7.040 -03 5.0 2.5 2.0 1092100300035 1.0 0.2 1.9 2.5 0.5 0.3 1092100300036 1.0 4.03 0.26 5.719 0.102 0.5719 1092100300037 5.860 0.097 8.230 -03 4.5 1.3 2.0 1092100300038 1.0 0.2 1.9 2.5 0.5 0.3 1092100300039 1.0 4.90 0.30 5.872 0.096 0.5959 1092100300040 5.865 0.096 8.364 -03 6.1 4.1 1.7 1092100300041 1.3 1.0 1.9 2.5 0.5 1.0 1092100300042 1.0 4.99 0.36 5.876 0.096 0.5964 1092100300043 6.074 0.095 9.499 -03 4.5 1.2 2.0 1092100300044 1.0 0.2 1.9 2.5 0.5 0.4 1092100300045 1.0 6.13 0.37 6.087 0.094 0.6449 1092100300046 6.349 0.090 1.046 -02 4.5 1.2 2.0 1092100300047 1.0 0.2 1.9 2.5 0.5 0.5 1092100300048 1.0 7.92 0.48 6.362 0.089 0.7572 1092100300049 6.396 0.089 1.130 -02 5.3 2.8 1.7 1092100300050 1.3 1.0 1.9 2.5 0.5 1.0 1092100300051 1.0 8.79 0.58 6.407 0.089 0.7773 1092100300052 6.600 0.088 1.107 -02 4.5 1.0 2.0 1092100300053 1.0 0.3 1.9 2.5 0.5 0.5 1092100300054 1.0 9.34 0.56 6.615 0.088 0.8437 1092100300055 6.844 0.087 1.194 -02 4.5 1.2 2.0 1092100300056 1.0 0.3 1.9 2.5 0.5 0.4 1092100300057 1.0 10.77 0.65 6.859 0.087 0.9018 1092100300058 6.913 0.086 1.314 -02 6.6 4.5 1.7 1092100300059 1.3 1.0 1.9 2.5 0.5 2.0 1092100300060 1.0 11.98 0.92 6.926 0.086 0.9113 1092100300061 7.093 0.086 1.299 -02 4.6 1.2 2.0 1092100300062 1.0 0.3 1.9 2.5 0.5 0.9 1092100300063 1.0 12.16 0.74 7.109 0.079 0.9361 1092100300064 7.420 0.086 1.447 -02 5.3 2.4 1.7 1092100300065 1.3 1.0 1.8 2.5 0.5 2.0 1092100300066 1.0 14.16 0.95 7.434 0.076 0.9785 1092100300067 7.920 0.089 1.838 -02 5.5 1.7 1.7 1092100300068 1.3 1.0 1.7 2.5 0.5 2.4 1092100300069 2.0 18.20 1.24 7.935 0.080 0.9905 1092100300070 8.413 0.097 2.130 -02 5.4 1.5 1.7 1092100300071 1.3 1.0 1.5 2.5 0.5 2.4 1092100300072 2.0 21.2 1.4 8.429 0.087 0.9942 1092100300073 8.428 0.098 2.084 -02 4.6 0.7 2.0 1092100300074 1.0 0.3 1.2 2.5 0.7 1.2 1092100300075 2.0 20.7 1.3 8.447 0.088 0.9943 1092100300076 8.902 0.104 2.510 -02 5.3 1.3 1.7 1092100300077 1.3 1.0 1.2 2.5 0.9 2.4 1092100300078 2.0 25.0 1.7 8.919 0.093 0.9978 1092100300079 9.390 0.112 2.789 -02 6.2 2.5 1.7 1092100300080 1.3 1.0 1.2 2.5 2.5 2.4 1092100300081 2.0 27.7 2.0 9.408 0.099 0.9917 1092100300082 9.417 0.111 2.751 -02 5.8 1.3 1.7 1092100300083 1.3 1.0 1.2 2.5 2.5 2.4 1092100300084 2.0 27.3 1.9 9.434 0.099 0.9912 1092100300085 9.873 0.117 2.977 -02 6.0 1.3 1.7 1092100300086 1.3 1.0 1.2 2.5 3.0 2.4 1092100300087 2.0 29.3 2.1 9.892 0.107 0.9837 1092100300088 ENDDATA 69 0 1092100300089 ENDSUBENT 88 0 1092100399999 ENDENTRY 3 0 1092199999999