ENTRY 31673 20090512 31363167300000001 SUBENT 31673001 20090512 31363167300100001 BIB 14 63 3167300100002 TITLE The (3He,tf) as a surrogate reaction to determine 3167300100003 (n,f) cross sections in the 10 to 20 MeV energy range 3167300100004 AUTHOR (M.S.Basunia,R.M.Clark,B.L.Goldblum,L.A.Bernstein, 3167300100005 L.Phair,J.T.Burke,C.W.Beausang,D.L.Bleuel, 3167300100006 B.Darakchieva,F.S.Dietrich,M.Evtimova,P.Fallon, 3167300100007 J.Gibelin,R.Hatarik,C.C.Jewett,S.R.Lesher,M.A.McMahan, 3167300100008 E.Rodriguez-Vieitez,M.Wiedeking) 3167300100009 INSTITUTE (1USABRK,1USALRL,1USARUT) 3167300100010 (1USAUSA) University of Richmond, Richmond, Virginia 3167300100011 23173, USA 3167300100012 REFERENCE (J,NIM/B,267,1899,2009) 3167300100013 FACILITY (CYCLO,1USABRK) 42 MeV 3He2+ beam from the 88-Inch 3167300100014 Cyclotron at Lawrence Berkeley National Laboratory 3167300100015 SAMPLE A self-supporting 99.99% pure metallic 238U foil of 3167300100016 thickness 4709 + - 235 Angstroms (761 +- 38 mu-g/cm2) 3167300100017 DETECTOR (TELES,SI,SI) one 140 mu-m and one 1000 mu-m Micron 3167300100018 S2 type silicon detectors as a dE+E particle telescope 3167300100019 covering an angular range of 36. to 66. degr. with 3167300100020 respect to the beam axis. Another 140 mu-m silicon 3167300100021 detector was used to detect fission fragments during 3167300100022 the experiment over the backward angular range of 106. 3167300100023 to 131.degr. 3167300100024 All silicon detectors were energy calibrated using 3167300100025 241Am and 226Ra alpha-lines. 3167300100026 METHOD (EDE) 3167300100027 (COINC) Coincidence of LCP and FF. 3167300100028 INC-SOURCE He-3 beam at 42 MeV 3167300100029 PART-DET (P,D,T,HE3,A) p,d,t, 3He, 4He 3167300100030 (FF) 3167300100031 ANALYSIS (SURGT) Neutron induced fission cross sections are 3167300100032 determined by the use of transfer reaction technique 3167300100033 to measure the fission probability of neptunium 3167300100034 formed in the reaction 238U(3He,t)238Np. The calculated3167300100035 compound nucleus formation cross section of a 3167300100036 neutron-induced reaction is combined with the 3167300100037 experimentally measured decay probability of the 3167300100038 compound nucleus, produced via a surrogate reaction, to3167300100039 the desired exit channel to obtain the neutron-induced 3167300100040 cross section indirectly. 3167300100041 The compound nucleus formation cross section for the 3167300100042 237Np+n was calculated for different incident neutron 3167300100043 energies using the Flap2.2 optical-model potential. 3167300100044 (see J.E.Escher and F.S.Dietrich, Tech. Rep. 3167300100045 UCRL-TR-212509, Lawrence Livermore National Laboratory,3167300100046 Livermore, CA (2005). 3167300100047 ERR-ANALYS (EN-ERR) Uncertainty of excitation energy. The 3167300100048 sources of uncertainty for the neutron energy were 3167300100049 the energy straggle of the triton, angular 3167300100050 resolution, beam energy resolution and detector 3167300100051 resolution. 3167300100052 (ERR-S) Statistical uncertainty. The statistical 3167300100053 uncertainty for the fission events in each 120-keV 3167300100054 energy bin was between 1 and 3%. 3167300100055 (ERR-1) The uncertainty of the fission detector 3167300100056 efficiency estimated to be 10%. 3167300100057 (ERR-2) An uncertainty of 3% for the compound 3167300100058 nucleus formation cross section calculation is 3167300100059 estimated for the 10 to 20 MeV neutron energy range. 3167300100060 (MISC2-ERR) Statistical uncertainty fission 3167300100061 probability of 238Np 3167300100062 STATUS (TABLE) Data were received from Dr. M.S.Basunia 3167300100063 (APRVD) Entry approved by Dr. M.S.Basunia 3167300100064 HISTORY (20090512C) SD 3167300100065 ENDBIB 63 0 3167300100066 COMMON 3 3 3167300100067 EN-ERR ERR-1 ERR-2 3167300100068 MEV PER-CENT PER-CENT 3167300100069 0.2 10.0 3.0 3167300100070 ENDCOMMON 3 0 3167300100071 ENDSUBENT 70 0 3167300199999 SUBENT 31673002 20090512 31363167300200001 BIB 2 7 3167300200002 REACTION (93-NP-237(N,F),,SIG,,,DERIV) 3167300200003 Fission c-s deduced via the product of measured 3167300200004 fission probability distribution of 238Np using 3167300200005 the transfer reaction 238U(3He,t)238Np. 3167300200006 MISC-COL (MISC1) 238Np* excitation energy 3167300200007 (MISC2) fission probabilities of the 238Np* produced 3167300200008 from the 238U(3He,tf) reaction 3167300200009 ENDBIB 7 0 3167300200010 NOCOMMON 0 0 3167300200011 DATA 6 82 3167300200012 EN DATA ERR-S MISC1 MISC2 MISC2-ERR 3167300200013 MEV B B MEV NO-DIM NO-DIM 3167300200014 10.20 2.20 0.07 15.6 0.74 0.023167300200015 10.30 2.13 0.06 15.8 0.72 0.023167300200016 10.40 2.19 0.06 15.9 0.74 0.023167300200017 10.60 2.16 0.06 16.0 0.73 0.023167300200018 10.70 2.18 0.06 16.1 0.73 0.023167300200019 10.80 2.17 0.05 16.2 0.73 0.023167300200020 10.90 2.15 0.05 16.4 0.72 0.023167300200021 11.00 2.29 0.05 16.5 0.77 0.023167300200022 11.20 2.22 0.05 16.6 0.75 0.023167300200023 11.30 2.19 0.05 16.7 0.74 0.023167300200024 11.40 2.08 0.04 16.8 0.70 0.013167300200025 11.50 2.18 0.04 17.0 0.73 0.013167300200026 11.60 2.14 0.04 17.1 0.72 0.013167300200027 11.80 2.14 0.04 17.2 0.72 0.013167300200028 11.90 2.14 0.04 17.3 0.72 0.013167300200029 12.00 2.14 0.04 17.4 0.72 0.013167300200030 12.10 2.05 0.04 17.6 0.69 0.013167300200031 12.20 2.11 0.04 17.7 0.71 0.013167300200032 12.40 2.16 0.04 17.8 0.73 0.013167300200033 12.50 2.12 0.04 17.9 0.72 0.013167300200034 12.60 2.15 0.04 18.0 0.72 0.013167300200035 12.70 2.13 0.03 18.2 0.72 0.013167300200036 12.80 2.19 0.03 18.3 0.74 0.013167300200037 13.00 2.16 0.03 18.4 0.73 0.013167300200038 13.10 2.20 0.03 18.5 0.74 0.013167300200039 13.20 2.17 0.03 18.6 0.73 0.013167300200040 13.30 2.13 0.03 18.8 0.72 0.013167300200041 13.40 2.17 0.03 18.9 0.73 0.013167300200042 13.60 2.15 0.03 19.0 0.72 0.013167300200043 13.70 2.12 0.03 19.1 0.71 0.013167300200044 13.80 2.19 0.03 19.2 0.74 0.013167300200045 13.90 2.19 0.03 19.4 0.74 0.013167300200046 14.00 2.15 0.03 19.5 0.72 0.013167300200047 14.20 2.20 0.03 19.6 0.74 0.013167300200048 14.30 2.20 0.03 19.7 0.74 0.013167300200049 14.40 2.20 0.03 19.8 0.74 0.013167300200050 14.50 2.21 0.03 20.0 0.75 0.013167300200051 14.60 2.24 0.03 20.1 0.76 0.013167300200052 14.80 2.30 0.03 20.2 0.78 0.013167300200053 14.90 2.28 0.03 20.3 0.77 0.013167300200054 15.00 2.26 0.03 20.4 0.76 0.013167300200055 15.10 2.29 0.03 20.6 0.77 0.013167300200056 15.20 2.32 0.03 20.7 0.78 0.013167300200057 15.40 2.30 0.03 20.8 0.77 0.013167300200058 15.50 2.30 0.03 20.9 0.77 0.013167300200059 15.60 2.40 0.03 21.0 0.81 0.013167300200060 15.70 2.36 0.03 21.2 0.80 0.013167300200061 15.90 2.42 0.03 21.3 0.81 0.013167300200062 16.00 2.36 0.03 21.4 0.79 0.013167300200063 16.10 2.36 0.03 21.5 0.79 0.013167300200064 16.20 2.38 0.03 21.6 0.80 0.013167300200065 16.30 2.35 0.03 21.8 0.79 0.013167300200066 16.50 2.43 0.03 21.9 0.82 0.013167300200067 16.60 2.36 0.03 22.0 0.79 0.013167300200068 16.70 2.42 0.03 22.1 0.82 0.013167300200069 16.80 2.37 0.03 22.2 0.80 0.013167300200070 16.90 2.38 0.03 22.4 0.80 0.013167300200071 17.10 2.41 0.03 22.5 0.81 0.013167300200072 17.20 2.30 0.03 22.6 0.78 0.013167300200073 17.30 2.40 0.03 22.7 0.81 0.013167300200074 17.40 2.34 0.03 22.8 0.79 0.013167300200075 17.50 2.36 0.03 23.0 0.80 0.013167300200076 17.70 2.35 0.03 23.1 0.79 0.013167300200077 17.80 2.34 0.03 23.2 0.79 0.013167300200078 17.90 2.35 0.03 23.3 0.79 0.013167300200079 18.00 2.29 0.03 23.4 0.77 0.013167300200080 18.10 2.26 0.03 23.6 0.76 0.013167300200081 18.30 2.31 0.03 23.7 0.78 0.013167300200082 18.40 2.35 0.03 23.8 0.79 0.013167300200083 18.50 2.27 0.03 23.9 0.77 0.013167300200084 18.60 2.28 0.03 24.0 0.77 0.013167300200085 18.70 2.28 0.03 24.2 0.77 0.013167300200086 18.90 2.24 0.03 24.3 0.75 0.013167300200087 19.00 2.26 0.03 24.4 0.76 0.013167300200088 19.10 2.23 0.03 24.5 0.75 0.013167300200089 19.20 2.18 0.03 24.6 0.73 0.013167300200090 19.30 2.16 0.03 24.8 0.73 0.013167300200091 19.50 2.18 0.03 24.9 0.73 0.013167300200092 19.60 2.21 0.03 25.0 0.74 0.013167300200093 19.70 2.15 0.03 25.1 0.73 0.013167300200094 19.80 2.15 0.03 25.2 0.72 0.013167300200095 19.90 2.17 0.03 25.4 0.73 0.013167300200096 ENDDATA 84 0 3167300200097 ENDSUBENT 96 0 3167300299999 ENDENTRY 2 0 3167399999999