ENTRY 23183 20120224 22302318300000001 SUBENT 23183001 20120224 22302318300100001 BIB 17 71 2318300100002 AUTHOR (I.A.Reyhancan) 2318300100003 TITLE Measurements and model calculations of activation cross2318300100004 sections for (n,p) reactions on Sm-152 isotope between 2318300100005 13.57 and 14.83 MeV neutrons. 2318300100006 REFERENCE (J,ARI,70,765,2012) Main reference. 2318300100007 #doi:10.1016/j.apradiso.2011.12.037 2318300100008 INSTITUTE (2TUKSTU) 2318300100009 PART-DET (DG) decay gamma-rays. 2318300100010 REL-REF (R,,A.Pavlik,J,JP/G,8,1283,1982) 2318300100011 90Zr(n,2n)89Zr c-s data used in ZR/Nb method. 2318300100012 (R,,D.R.Nethaway,J,JIN,40,1285,1978) 2318300100013 93Nb(n,2n)92mNb c-s data used in ZR/Nb method. 2318300100014 (M,,V.E.Lewis,J,NIM,174,141,1980) 2318300100015 Zr/Nb ratio method . 2318300100016 (M,,Subasi+,J,NSE,122,(1),423,1996) 2318300100017 Details of method. 2318300100018 (M,,Subasi+,J,NSE,130,254,1998) 2318300100019 Details of method. 2318300100020 (M,,Subasi+,J,NSE,135,260,2000) 2318300100021 Details of method. 2318300100022 FACILITY (NGEN,2TUKSTU) SAMES T-400 neutron generator. 2318300100023 Beam current is kept at about 300 microA with 2318300100024 accelerator voltage of 250 kV. 2318300100025 INC-SOURCE (D-T) Solid tritium target. 2318300100026 The angles of the irradiation positions to the deuteron2318300100027 beam were 0, 30, 60, 90, 120, 130, which correspond to 2318300100028 the neutron energies 14.83, 14.75, 14.52, 14.15, 13.75,2318300100029 13.57 MeV. 2318300100030 INC-SPECT The effective energy of incident neutron at each 2318300100031 irradiation position were determined by the ratio of 2318300100032 90Zr(n,2n)89Zr (3.27 d) (Pavlik+,1982) 2318300100033 and 93Nb(n,2n)92mNb (10.15 d) (Nethaway,1978) reaction 2318300100034 cross sections (the Zr/Nb ratio method) (Lewis and 2318300100035 Zieba, 1980) and are given in detail in Subasi+,1996. 2318300100036 Neutron fluxes are measured in each irradiation place 2318300100037 (70mm from the neutron source )from 9.E+7 to 1.1E+8 2318300100038 n/cm2*s, depending on the neutron target condition. 2318300100039 MONITOR (13-AL-27(N,P)12-MG-27,,SIG) 2318300100040 MONIT-REF (,,3,ENDF/B-VII.0,,2006) 2318300100041 DECAY-MON (12-MG-27,9.458MIN,DG,843.8,0.73) 2318300100042 SAMPLE Sm-152 isotopic ambulance (26.75+-0.16) %. 2318300100043 Sample pellets(13mm diameter,8.5mm thickness, about 2g 2318300100044 weight) were pressed from the natural,high-purity 2318300100045 (99.9%) Sm2O3 and Al2O3 powders. Water contents in the2318300100046 samples were removed by heating the oxide materials up 2318300100047 to 105 deg C, then 0.5h cooled down to the room 2318300100048 temperature, were filled in dry stainless steel 2318300100049 capsules, pressed to form a dense pellet, were tightly 2318300100050 packed into thin polyethylene(PE) boxes and sealed-of. 2318300100051 METHOD (ACTIV) Cyclic irradiation. 2318300100052 DETECTOR (SCIN) During irradiation, the neutron flux was 2318300100053 monitored with plastic scintillation detector NE102A, 2318300100054 placed under the neutron-producing target. The monitor 2318300100055 was calibrated with the standard reaction irradiating 2318300100056 Al2O3 sample placed in the same irradiation position. 2318300100057 (HPGE) Gamma activities induced in the Al sample was 2318300100058 measured by high-resolution gamma-ray spectrometer. 2318300100059 In all gamma-ray measurements an acrylic plate of 10mm 2318300100060 thickness was placed in front of the detector to block 2318300100061 the beta particles from the samples. 2318300100062 CORRECTION For dead time and the self-absorption of the gamma ray 2318300100063 in the pellet(4%). 2318300100064 Neutron flux was corrected for the neutron flux 2318300100065 variations. 2318300100066 For low energy neutron contribution to the investigated2318300100067 reactions - up to 10%. 2318300100068 The effect of low energy neutron in sample pellets 2318300100069 packed in thin polyethylene boxes was performed by the 2318300100070 MCNP5 Monte Carlo code (X-5MonteCarloTeam,2003) and 2318300100071 found less than 1% that can be neglected. 2318300100072 HISTORY (20120224C) M.M. 2318300100073 ENDBIB 71 0 2318300100074 NOCOMMON 0 0 2318300100075 ENDSUBENT 74 0 2318300199999 SUBENT 23183002 20120224 22302318300200001 BIB 8 51 2318300200002 REACTION 1(62-SM-152(N,P)61-PM-152-G,,SIG) 2318300200003 2(62-SM-152(N,P)61-PM-152-M1,,SIG) 2318300200004 REL-REF (A,31439002,A.Kirov+,J,ZP/A,345,(3),285,1993) 2318300200005 (D,12033029,R.G.Wille+,J,PR,118,242,1960) 2318300200006 (R,,A.Artna-Cohen,J,NDS,79,1,1996) 2318300200007 Reference for decay data, also NUDAT2.5, 2011 . 2318300200008 DECAY-DATA1(61-PM-152-G,4.12MIN,DG,121.8,0.16) HL=4.12+-0.08 . 2318300200009 2(61-PM-152-M1,7.52MIN,DG,340.4,0.313) HL=7.52+-0.08 , 2318300200010 Gamma-ray intensity (31.3+-2.0) % . 2318300200011 CORRECTION For self-absorption (73% for 121.8 keV, 18% for 340.4 2318300200012 keV), attenuation of the gamma rays in the acrylic 2318300200013 beta absorber and dead time. 2318300200014 The contribution from the Sm-154(n,t)Pm-152m,g and 2318300200015 Sm-154(n,3He)Nd-152 (Pm-152m,g via beta- decay) were 2318300200016 ignored, because cross sections of these reactions 2318300200017 for 14.5 MeV neutron are less than 10**-6 barn and 2318300200018 10**-18 barn (JEFF-3.1, 2009), respectively. 2318300200019 ANALYSIS Full energy peak (FEP) areas, mass of sample, 2318300200020 the concentration of the element, Avogadro number,the 2318300200021 atomic weight of the element, isotopic abundance, the 2318300200022 relative gamma-ray intensity, effective flux measured 2318300200023 by calibrated Al foil monitor, counting solid angle 2318300200024 with self-absorption correction, FEP efficiency of the 2318300200025 gamma-ray detector, decay constant, the number of 2318300200026 cycles were used to cross-section calculation by the 2318300200027 cyclic activation formula. 2318300200028 For 152gPm radionuclide, gamma energy line 121.8keV 2318300200029 is contributed from Pm-152m (intensity 45%),this 2318300200030 contribution was subtracted from total photo-peak area 2318300200031 counts and cross sections were calculated for 2318300200032 Sm-152(n,p)Pm-152g reaction. 2318300200033 ERR-ANALYS (ERR-1) Error of sample weight. 2318300200034 (ERR-2) Error of concentration of the element. 2318300200035 (ERR-3) Error of isotopic abundance. 2318300200036 (ERR-4) Error of relative gamma-ray intensity. 2318300200037 (ERR-5) Error of counting solid angle and intrinsic 2318300200038 efficiency of the gamma detector. 2318300200039 (ERR-6,10.,25.) Error of full energy photo-peak area. 2318300200040 (ERR-7) Error of neutron flux fluctuation. 2318300200041 (ERR-8) Error of effective neutron flux (reference 2318300200042 cross-section). 2318300200043 (ERR-9) Error of irradiation, cooling and measuring 2318300200044 times. 2318300200045 Uncertainties were assumed to be uncorrelated. 2318300200046 (ERR-T) Total uncertainties in the cross section values2318300200047 could be determined by adding the experimental errors 2318300200048 and the uncertainties of nuclear data in the quadratic 2318300200049 form. 2318300200050 ADD-RES (COMP) Compared with statistical model calculations - 2318300200051 code STAPRE, spin cut-off parameter 1.0 . 2318300200052 STATUS (TABLE) Table 4 of J,ARI,70,765,2012. 2318300200053 ENDBIB 51 0 2318300200054 COMMON 8 6 2318300200055 ERR-1 ERR-2 ERR-3 ERR-4 ERR-5 ERR-7 2318300200056 ERR-8 ERR-9 2318300200057 PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT 2318300200058 PER-CENT PER-CENT 2318300200059 0.1 0.1 0.6 6.4 3.0 1. 2318300200060 3.0 0.1 2318300200061 ENDCOMMON 6 0 2318300200062 DATA 7 6 2318300200063 EN EN-ERR DATA 1ERR-T 1DATA 2ERR-T 22318300200064 MONIT 2318300200065 MEV MEV MB MB MB MB 2318300200066 MB 2318300200067 13.57 0.04 2.85 0.77 0.82 0.21 2318300200068 76.53 2318300200069 13.75 0.05 2.66 0.72 0.97 0.24 2318300200070 75.09 2318300200071 14.15 0.07 2.81 0.76 1.03 0.26 2318300200072 71.89 2318300200073 14.52 0.09 3.75 0.96 1.43 0.36 2318300200074 68.96 2318300200075 14.75 0.09 4.83 1.21 1.85 0.46 2318300200076 67.15 2318300200077 14.83 0.10 5.82 1.45 2.36 0.59 2318300200078 66.51 2318300200079 ENDDATA 16 0 2318300200080 ENDSUBENT 79 0 2318300299999 ENDENTRY 2 0 2318399999999