ENTRY 31711 20110511 31503171100000001 SUBENT 31711001 20110511 31503171100100001 BIB 15 28 3171100100002 TITLE Measurements of neutron fission cross sections of 3171100100003 237Np, 240Pu, 241Pu, 242Pu, 244Pu, and 241Am at 3171100100004 14.7 MeV 3171100100005 AUTHOR (K.Gul, M.Ahmad, M.Anwar, S.M.Saleem) 3171100100006 INSTITUTE (3PAKNIL) Nuclear Physics Division 3171100100007 REFERENCE (J,NSE,94,42,1986) 3171100100008 FACILITY (NGEN,3PAKNIL) 14-MeV Neutron Generator Facility 3171100100009 INC-SOURCE (D-T) 120 keV incident deuteron 3171100100010 INC-SPECT Sample were irradiated at 6 cm from the neutron source.3171100100011 SAMPLE Areal distributions in 40 to 100 micro-gram/cm2 3171100100012 DETECTOR (SI) Fission fragment detection 3171100100013 (GELI) 1369 keV gamma detection 3171100100014 MONITOR (13-AL-27(N,A)11-NA-24,,SIG) 3171100100015 DECAY-MON (11-NA-24,15.HR,DG,1369.) 3171100100016 REL-REF Another work (Sample was 15.1 cm from neutron source) 3171100100017 (N,30548001,N.A.Khan+,J,NIM,173,137,1980) 3171100100018 for Pu-240,241 3171100100019 (N,30548001,N.A.Khan+,J,NIM,173,163,1980) 3171100100020 for Pu-242,244 and Am-241 3171100100021 CORRECTION Np-237, Pu-242, Pu-244: 3171100100022 No alpha pile-up. 2-pi geometry used. No correction 3171100100023 for anisotropy of angular distribution. 3171100100024 Pu-240, Pu-241, Am-241: 3171100100025 Alpha pile-up problem overcome. 3171100100026 Solid angle reduced, and correction was made for 3171100100027 anisotropy of angular distribution. 3171100100028 STATUS (TABLE) Table III of Nucl.Sci.Eng.94(1986)42 3171100100029 HISTORY (20110511C) On 3171100100030 ENDBIB 28 0 3171100100031 COMMON 1 3 3171100100032 MONIT 3171100100033 MB 3171100100034 113. 3171100100035 ENDCOMMON 3 0 3171100100036 ENDSUBENT 35 0 3171100199999 SUBENT 31711002 20110511 31503171100200001 BIB 2 7 3171100200002 REACTION (93-NP-237(N,F),,SIG) 3171100200003 ERR-ANALYS (ERR-T) Total uncertainty (4.9%) 3171100200004 (ERR-S) Statistics (3%) 3171100200005 (ERR-1) Neutron flux (3%) 3171100200006 (ERR-2) 2-pi geometry (1%) 3171100200007 (ERR-3) Fissile material (2%) 3171100200008 (ERR-4) Extrapolation to zero pulse height (1%) 3171100200009 ENDBIB 7 0 3171100200010 COMMON 5 3 3171100200011 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 3171100200012 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT 3171100200013 3. 3. 1. 2. 1. 3171100200014 ENDCOMMON 3 0 3171100200015 DATA 3 1 3171100200016 EN DATA ERR-T 3171100200017 MEV B B 3171100200018 14.7 2.0 0.1 3171100200019 ENDDATA 3 0 3171100200020 ENDSUBENT 19 0 3171100299999 SUBENT 31711003 20110511 31503171100300001 BIB 2 7 3171100300002 REACTION (94-PU-244(N,F),,SIG) 3171100300003 ERR-ANALYS (ERR-T) Total uncertainty (4.9%) 3171100300004 (ERR-S) Statistics (3%) 3171100300005 (ERR-1) Neutron flux (3%) 3171100300006 (ERR-2) 2-pi geometry (1%) 3171100300007 (ERR-3) Fissile material (2%) 3171100300008 (ERR-4) Extrapolation to zero pulse height (1%) 3171100300009 ENDBIB 7 0 3171100300010 COMMON 5 3 3171100300011 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 3171100300012 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT 3171100300013 3. 3. 1. 2. 1. 3171100300014 ENDCOMMON 3 0 3171100300015 DATA 3 1 3171100300016 EN DATA ERR-T 3171100300017 MEV B B 3171100300018 14.7 1.7 0.1 3171100300019 ENDDATA 3 0 3171100300020 ENDSUBENT 19 0 3171100399999 SUBENT 31711004 20110511 31503171100400001 BIB 2 7 3171100400002 REACTION (94-PU-242(N,F),,SIG) 3171100400003 ERR-ANALYS (ERR-T) Total uncertainty (4.9%) 3171100400004 (ERR-S) Statistics (3%) 3171100400005 (ERR-1) Neutron flux (3%) 3171100400006 (ERR-2) 2-pi geometry (1%) 3171100400007 (ERR-3) Fissile material (2%) 3171100400008 (ERR-4) Extrapolation to zero pulse height (1%) 3171100400009 ENDBIB 7 0 3171100400010 COMMON 5 3 3171100400011 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 3171100400012 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT 3171100400013 3. 3. 1. 2. 1. 3171100400014 ENDCOMMON 3 0 3171100400015 DATA 3 1 3171100400016 EN DATA ERR-T 3171100400017 MEV B B 3171100400018 14.7 2.1 0.1 3171100400019 ENDDATA 3 0 3171100400020 ENDSUBENT 19 0 3171100499999 SUBENT 31711005 20110511 31503171100500001 BIB 4 9 3171100500002 REACTION (94-PU-241(N,F),,SIG) 3171100500003 ASSUMED (ASSUM,94-PU-241(N,F),,DA,FF,RSD) 3171100500004 REL-REF (N,,R.B.Leachman+,J,PR,B137,814,1965) Anisotropy 3171100500005 ERR-ANALYS (ERR-T) Total uncertainty (7.7%) 3171100500006 (ERR-S) Statistics (6%) 3171100500007 (ERR-1) Neutron flux (3%) 3171100500008 (ERR-2) Anisotropy (2%) 3171100500009 (ERR-3) Fissile material (3%) 3171100500010 (ERR-4) Extrapolation to zero pulse height (1%) 3171100500011 ENDBIB 9 0 3171100500012 COMMON 6 3 3171100500013 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 ASSUM 3171100500014 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT NO-DIM 3171100500015 6. 3. 2. 3. 1. 0.2 3171100500016 ENDCOMMON 3 0 3171100500017 DATA 3 1 3171100500018 EN DATA ERR-T 3171100500019 MEV B B 3171100500020 14.7 2.0 0.2 3171100500021 ENDDATA 3 0 3171100500022 ENDSUBENT 21 0 3171100599999 SUBENT 31711006 20110511 31503171100600001 BIB 4 9 3171100600002 REACTION (94-PU-240(N,F),,SIG) 3171100600003 ASSUMED (ASSUM,94-PU-240(N,F),,DA,FF,RSD) 3171100600004 REL-REF (N,,J.E.Simons+,J,PR,B137,809,1965) Anisotropy 3171100600005 ERR-ANALYS (ERR-T) Total uncertainty (7.7%) 3171100600006 (ERR-S) Statistics (6%) 3171100600007 (ERR-1) Neutron flux (3%) 3171100600008 (ERR-2) Anisotropy (2%) 3171100600009 (ERR-3) Fissile material (3%) 3171100600010 (ERR-4) Extrapolation to zero pulse height (1%) 3171100600011 ENDBIB 9 0 3171100600012 COMMON 6 3 3171100600013 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 ASSUM 3171100600014 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT NO-DIM 3171100600015 6. 3. 2. 3. 1. 0.3 3171100600016 ENDCOMMON 3 0 3171100600017 DATA 3 1 3171100600018 EN DATA ERR-T 3171100600019 MEV B B 3171100600020 14.7 2.5 0.2 3171100600021 ENDDATA 3 0 3171100600022 ENDSUBENT 21 0 3171100699999 SUBENT 31711007 20110511 31503171100700001 BIB 3 8 3171100700002 REACTION (95-AM-241(N,F),,SIG) 3171100700003 ASSUMED (ASSUM,95-AM-241(N,F),,DA,FF,RSD) Pu-241 value adopted 3171100700004 ERR-ANALYS (ERR-T) Total uncertainty (7.7%) 3171100700005 (ERR-S) Statistics (6%) 3171100700006 (ERR-1) Neutron flux (3%) 3171100700007 (ERR-2) Anisotropy (2%) 3171100700008 (ERR-3) Fissile material (3%) 3171100700009 (ERR-4) Extrapolation to zero pulse height (1%) 3171100700010 ENDBIB 8 0 3171100700011 COMMON 6 3 3171100700012 ERR-S ERR-1 ERR-2 ERR-3 ERR-4 ASSUM 3171100700013 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT NO-DIM 3171100700014 6. 3. 2. 3. 1. 0.2 3171100700015 ENDCOMMON 3 0 3171100700016 DATA 3 1 3171100700017 EN DATA ERR-T 3171100700018 MEV B B 3171100700019 14.7 3.0 0.2 3171100700020 ENDDATA 3 0 3171100700021 ENDSUBENT 20 0 3171100799999 ENDENTRY 7 0 3171199999999