ENTRY M0692 20110119 M056M069200000001 SUBENT M0692001 20110119 M056M069200100001 BIB 15 97 M069200100002 TITLE Photofission cross sections for 237Np in the energy M069200100003 interval from 5.27 to 10.83 MeV. M069200100004 AUTHOR (L.P.GERALDO,R.SEMMLER,O.I.GONCALEZ,J.MESA, M069200100005 J.D.T.ARRUDA-NETO,F.GARCIA,O.RODRIGUES) M069200100006 REFERENCE (J,NSE,136,357,2000) M069200100007 INSTITUTE (3BZLBZL,3BZLIPE,3BZLCTA,3BZLUSP) M069200100008 FACILITY (REAC,3BZLIPE) IPEN IAE-R1 MW pool-type research M069200100009 reactor. M069200100010 INC-SOURCE (MPH) High-resolution (4 - 20 eV) gamma-rays, with M069200100011 discrete energies up to 10.83 MeV, were produced by M069200100012 thermal neutron capture in 30 different target M069200100013 materials, when they were positioned near the M069200100014 reactor core: M069200100015 Capture target Gamma-ray energy, keV M069200100016 ---------------------------------------------------- M069200100017 Yb 5265.70 M069200100018 S 5420.50 M069200100019 Dy 5607.75 M069200100020 Hf 5723.50 M069200100021 In 5891.90 M069200100022 Hg 5966.20 M069200100023 Y 6080.49 M069200100024 Er 6228.23 M069200100025 Na 6395.40 M069200100026 Ca 6419.90 M069200100027 Nd 6501.70 M069200100028 V 6517.26 M069200100029 Gd 6748.70 M069200100030 Ti 6760.13 M069200100031 Be 6809.41 M069200100032 As 7019.45 M069200100033 Sm 7214.20 M069200100034 Mn 7243.79 M069200100035 Pb 7367.70 M069200100036 Cl 7413.80 M069200100037 Se 7418.47 M069200100038 Fe 7631.13 M069200100039 Al 7723.85 M069200100040 K 7770.22 M069200100041 Zn 7863.30 M069200100042 Cu 7914.50 M069200100043 Cd 8484.80 M069200100044 Ni 8998.80 M069200100045 Cr 9720.30 M069200100046 N 10829.18 M069200100047 SAMPLE The sample consisted of 36.10 (0.13 mg) of 237Np M069200100048 deposited in form of NpO-2 on six titanium disks, M069200100049 forming an active diameter of ~ 40 mm. M069200100050 DETECTOR (TRD) Makrofol KG foils (10 micron thickness) in form M069200100051 of a sandwich (2PI geometry). M069200100052 METHOD (EXTB,STTA) M069200100053 ANALYSIS (UNFLD) The method of unfolding of cross section data M069200100054 for definite gamma-lines from those obtained for M069200100055 composition of all gamma-lines used employed M069200100056 least-squares methods and the covariance matrix M069200100057 methodology. M069200100058 REL-REF (N,,O.L.GONCALES+,J,NSE,132,135,1999) The method for M069200100059 unfolding. M069200100060 ERR-ANALYS (DATA-ERR) The overall uncertainties were estimated M069200100061 by summing in quadrature all following components: M069200100062 Error Sources Range (%) M069200100063 --------------------------------------------------- M069200100064 Random errors M069200100065 (0 % uncorrelated errors) M069200100066 Fission track counting 0.43 - 6.83 M069200100067 Gamma-ray counting 0.19 - 4.83 M069200100068 --------------------------------------------------- M069200100069 Systematic errors M069200100070 (100 % completely correlated errors) M069200100071 (ERR-1) 237Np deposit mass 0.35 M069200100072 (ERR-2) Gamma-ray intensity ratio 0.29 M069200100073 (ERR-3) Gamma-ray intensity attenuation 0.34 M069200100074 (ERR-4) 252Cf activity 1.8 M069200100075 (ERR-5) Fission track calibration factor 0.80 M069200100076 (ERR-6,0.77,3.51) Ge(Li) calibration factor 0.77 - M069200100077 3.51 M069200100078 COMMENT Photofission cross sections for 237 Np have been M069200100079 measured as a function of energy, in the interval M069200100080 from 5.27 to 10.83 MeV. The gamma-ray spectra were M069200100081 those produced by thermal neutron capture, in 30 M069200100082 different target materials, as a tangential beam M069200100083 hole of the Instituto de Pesquisas Energeticas e M069200100084 Nucleares IAE-R1 1-MW research reactor. The set of M069200100085 experimental data has been unfolded employing M069200100086 least-squares methods and the covariance matrix M069200100087 methodology. The determined photofission cross M069200100088 sections for 237Np, together with the complete M069200100089 correlation matrix for the involved errors, are M069200100090 presented and are compared with previous M069200100091 measurements reported in the literature. A M069200100092 statistical calculation for the 237Np photofission M069200100093 cross sections was performed, and the results are M069200100094 compared with the experimental data. M069200100095 STATUS (TABLE) Data from the Table III were compiled at the M069200100096 Russia MSU SINP CDFE by V.Varlamov. M069200100097 HISTORY (20060626C) M069200100098 (20110119A) Corrected by V.Varlamov: COMMON. M069200100099 ENDBIB 97 0 M069200100100 COMMON 5 3 M069200100101 ERR-1 ERR-2 ERR-3 ERR-4 ERR-5 M069200100102 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT M069200100103 .35 .29 .34 1.8 .8 M069200100104 ENDCOMMON 3 0 M069200100105 ENDSUBENT 104 0 M069200199999 SUBENT M0692002 20110119 M056M069200200001 BIB 2 5 M069200200002 REACTION (93-NP-237(G,F),,SIG) M069200200003 REL-REF (N,,A.S.SOLDATOV+,R,INDC(CCP)-379,1994) The overall M069200200004 agreement is reasonable between the 237Np(g,f) M069200200005 reaction cross section data obtained and evaluated M069200200006 before. M069200200007 ENDBIB 5 0 M069200200008 NOCOMMON 0 0 M069200200009 DATA 3 30 M069200200010 EN DATA DATA-ERR M069200200011 MEV MB MB M069200200012 5.2657 4.38 .7 M069200200013 5.4205 2.75 1.15 M069200200014 5.60775 8.19 .96 M069200200015 5.7235 14.53 1.28 M069200200016 5.8919 6.72 1.89 M069200200017 5.9662 16.23 .88 M069200200018 6.08049 12.57 .52 M069200200019 6.22823 14.56 3.76 M069200200020 6.3954 22.43 .6 M069200200021 6.4199 18.08 1.7 M069200200022 6.5017 24.5 1.43 M069200200023 6.51726 16.34 4.53 M069200200024 6.7487 37.34 2.16 M069200200025 6.76013 19.23 1.36 M069200200026 6.80941 22.26 1.27 M069200200027 7.01945 20.59 3.85 M069200200028 7.2142 19.73 4.34 M069200200029 7.24379 41.61 4.34 M069200200030 7.3677 34.74 1.08 M069200200031 7.4138 30.99 8.44 M069200200032 7.41847 27.04 12.48 M069200200033 7.63113 30.9 1.94 M069200200034 7.72385 42.22 1.45 M069200200035 7.77022 31.12 3.43 M069200200036 7.8633 49.67 3.26 M069200200037 7.9145 43.45 1.98 M069200200038 8.4848 42.39 32.17 M069200200039 8.9988 53.78 14.89 M069200200040 9.7203 186.02 20.77 M069200200041 10.82918 252.89 16. M069200200042 ENDDATA 32 0 M069200200043 ENDSUBENT 42 0 M069200299999 ENDENTRY 2 0 M069299999999