ENTRY 23150 20220317 23102315000000001 SUBENT 23150001 20220317 23102315000100001 BIB 17 105 2315000100002 INSTITUTE (2FR SAC) CEA in Saclay. Exp. setup. 2315000100003 (2FR BRC) J.Laurec, A.Adam, T.de Bruyne, E.Bauge, 2315000100004 T.Granier, J.Aupiais, O.Bersillon, G.Le Petit. 2315000100005 (2FR ITL) N.Authier, P.Casoli 2315000100006 REFERENCE (J,NDS,111,2965,2010) 2315000100007 REL-REF (O,21707001,J.Laurec+,J,NDS,111,2965,2010) Data 2315000100008 measured in fission spectrum.2315000100009 (O,21708001,J.Laurec+,J,NDS,111,2965,2010) Data 2315000100010 measured at 14.7 MeV. 2315000100011 AUTHOR (J.Laurec, A.Adam, T.de Bruyne, E.Bauge, T.Granier, 2315000100012 J.Aupiais, O.Bersillon, G.LePetit, N.Authier, P.Casoli)2315000100013 TITLE Fission Product Yields of 233-U, 235-U, 238-U and 2315000100014 239-Pu in Fields of Thermal Neutrons, Fission Neutrons 2315000100015 and 14.7-MeV Neutrons. 2315000100016 FACILITY (REAC,2FR SAC) Thermal column of the EL-3 heavy 2315000100017 water-cooled reactor (decommissioned in 1982) at CEA 2315000100018 in Saclay. 2315000100019 INC-SOURCE (THCOL) 2315000100020 INC-SPECT Flux at the irradiation point was 10**9 neutron/cm2/s.2315000100021 SAMPLE Thin deposits ( made by electroplating) - for fission 2315000100022 chambers; 2315000100023 thicker ones (made by electrospraying) - for fission 2315000100024 products analysis. 2315000100025 METHOD 1(FISCT) Fission counting 2315000100026 2(GSPEC) Gamma-spectrometry of the irradiated targets. 2315000100027 DETECTOR 1(FISCH) Plane fission chamber for the measurements 2315000100028 of the numbers of fissions. 2315000100029 Efficiency of 100 % was confirmed by two measurements: 2315000100030 counting of alpha-activity of the sample; 2315000100031 measurement at thermal flux by gold and cobalt foils. 2315000100032 2(GELI ) Calibrated Ge-Li spectrometers. 2315000100033 Volume 1 cm**3 (planar), 40 cm**3 (coaxial). 2315000100034 Energy resolution from 0.6 at 0.9 keV at 122 keV, 2315000100035 from 1.9 to 1.33 keV at 1.33MeV, depending on detector.2315000100036 Shielded with 10 cm lead. 2315000100037 The efficiency curves of the detectors were determined 2315000100038 by standard sources of 51Cr, 54Mn, 57Co, 60Co, 85Sr, 2315000100039 88Y, 109Cd, 113Sn, 137Cs, 139Ce, 141Ce, 144Ce, 203Hg, 2315000100040 241Am; 166Hom, 169Yb, 182Ta; Eu-152. 2315000100041 Overall uncertainty on the efficiency taking into 2315000100042 account solid angle correction (2 sigma) 2.0% . 2315000100043 ANALYSIS Final value for each fission product is the combination2315000100044 of these different measurements through weighted 2315000100045 averages, the weight being defined as the inverse of 2315000100046 the square of the statistical uncertainty associated 2315000100047 with each individual measurement. 2315000100048 Old yields data were updated as: 2315000100049 Y-new=Y-old(w1*(I1-old/I1-new)+w2*(I2-old/I2-new)), 2315000100050 w1, w2 - normalized weights corresponding to each gamma2315000100051 line: 2315000100052 95Zr 724 keV 50%, 757 keV 50%; 2315000100053 97Zr 658 keV 50%, 743 keV 50%; 2315000100054 140Ba 163 keV 50%, 537 keV 50%; 2315000100055 147Nd 91 keV 50%, 531 keV 50%; 2315000100056 DECAY-DATA (40-ZR-95,64.032D,DG,724.2,0.4427,DG,756.7,0.5438) 2315000100057 44.27+-0.22; 64.032+-0.006; 54.38+-0.22. 2315000100058 (40-ZR-97,0.698D,DG,657.9,0.9823,DG,743.4,0.9309) 2315000100059 98.23+-0.08; 0.698+-3E-4; 93.09+-0.16 . 2315000100060 (42-MO-99,2.747D,DG,140.5,0.8906) 2315000100061 89.06+-0.24 ; 2.747+-4E-4 . 2315000100062 (44-RU-103,39.26D,DG,497.1,0.910) 2315000100063 91.0+-1.2 ; 39.26+-0.02 . 2315000100064 (45-RH-105-G,1.473D,DG,318.9,0.191) 2315000100065 19.1+-0.6 ; 1.473+-0.002. 2315000100066 (53-I-131,8.025D,DG,364.5,0.815) 2315000100067 81.5+-0.8 ; 8.025+-0.001 . 2315000100068 (52-TE-132,3.204D,DG,228.2,0.88) 2315000100069 88+-3 ; 3.204+-0.013 . 2315000100070 (56-BA-140,12.753D,DG,162.7,0.0622,DG,537.2,0.2439) 2315000100071 6.22+-0.09 ; 12.753+-0.002 ; 24.39+-0.22 . 2315000100072 (58-CE-141,32.508D,DG,145.4,0.4829) 2315000100073 48.29+-0.20 ; 32.508+-0.013 . 2315000100074 (58-CE-143,1.376D,DG,293.3,0.428) 2315000100075 42.8+-0.4 ; 1.376+-0.002 . 2315000100076 (60-ND-147,10.98D,DG,91.0,0.281,DG,531.0,0.1337) 2315000100077 28.1+-0.5 ; 10.98+-0.01 ; 13.37+-1.1 . 2315000100078 ERR-ANALYS (ERR-4,0.3,6.) Statistical uncertainty of photopeak 2315000100079 area. 2315000100080 (ERR-5,1.,2.) Statistical uncertainty of samples mass 2315000100081 ratio. 2315000100082 (ERR-6) Statistical uncertainty of measurement of 2315000100083 number of fissions. 2315000100084 (ERR-7,,1.5) Statistical uncertainty of neutron flux 2315000100085 readjustment. 2315000100086 (ERR-8) Systematical uncertainty of gamma detection 2315000100087 efficiency. 2315000100088 (ERR-9) Systematical uncertainty of samples mass 2315000100089 ratio. 2315000100090 COMMENT Of compiler. In text of J,NDS,111,2965,2010 page 2972 2315000100091 it's mentioned, that U-235 old data were published in 2315000100092 R,CEA-R-5147,1981,Determination of the fission yields 2315000100093 of U-233, U-235, U-238, Pu-239 induced by a spectra of 2315000100094 fission neutrons and neutrons of 14.7 MeV. 2315000100095 (Determination des rendements de fissions induites 2315000100096 par un spectre de neutrons de fission et des neutrons 2315000100097 de 14.7 MeV dans U-233,U-235,U-238,Pu-239). 2315000100098 Old Pu-235 data at thermal spectrum were published in 2315000100099 J.Laurec,Internal CEA DAM report, 1976. 2315000100100 Old data at thermal spectrum were not found in 2315000100101 EXFOR, only data for fission spectrum and 14.7 MeV 2315000100102 incident neutrons: 2315000100103 CORRECTION For contribution from fissile materials. 2315000100104 For background. 2315000100105 HISTORY (20111113C) M.M. 2315000100106 (20220317A) SD: Decay-data of Ba-140 corrected. 2315000100107 ENDBIB 105 0 2315000100108 COMMON 3 3 2315000100109 ERR-6 ERR-8 ERR-9 2315000100110 PER-CENT PER-CENT PER-CENT 2315000100111 0.5 2.0 0.8 2315000100112 ENDCOMMON 3 0 2315000100113 ENDSUBENT 112 0 2315000199999 SUBENT 23150002 20111113 22282315000200001 BIB 5 23 2315000200002 REACTION (92-U-235(N,F)ELEM/MASS,CUM,FY,,MXW) 2315000200003 Fission yields in a thermal neutron spectrum. 2315000200004 SAMPLE .U(3)-O(8) targets on aluminum or titanium foils. 2315000200005 Diameter 14 mm, enrichment 99.9 percent U-235. 2315000200006 Isotopic ratios U-234/U-235 7.0*10**-6, 2315000200007 U-238/U-235 9.0*10**-6. 2315000200008 Mass 5.36+-0.10 microg (measured by alpha-spectrometry2315000200009 5.49+-0.03 (measured by thermo-ionization mass 2315000200010 spectrometry). 2315000200011 ERR-ANALYS (ERR-T) Total uncertainty ( 1 sigma) as combination 2315000200012 in quadrature of: 2315000200013 (ERR-1) Uncertainty (1 sigma) associated with 2315000200014 measurements, i.e. combination of statistical and 2315000200015 systematic uncertainty; 2315000200016 derived from authors' errors - see MISC-COL. 2315000200017 (ERR-2) Uncertainty ( 1 sigma) due to nuclear data. 2315000200018 MISC-COL (MISC) Uncertainty (2 sigma) associated with 2315000200019 measurements, i.e. combination of statistical and 2315000200020 systematic uncertainty, given by authors in article. 2315000200021 Were re-calculated (/2.) as ERR-1 to obtain 1 sigma 2315000200022 error; authors' errors are given as MISC according to 2315000200023 the comment of N.Otsuka (NDS,IAEA) . 2315000200024 STATUS (TABLE) Table VI of J,NDS,111,2965,2010. 2315000200025 ENDBIB 23 0 2315000200026 COMMON 1 3 2315000200027 EN-DUMMY 2315000200028 EV 2315000200029 0.0253 2315000200030 ENDCOMMON 3 0 2315000200031 DATA 7 11 2315000200032 ELEMENT MASS DATA ERR-1 ERR-2 ERR-T 2315000200033 MISC 2315000200034 NO-DIM NO-DIM PC/FIS PER-CENT PER-CENT PER-CENT 2315000200035 PER-CENT 2315000200036 40. 95. 6.42 2.05 0.3 2.1 2315000200037 4.1 2315000200038 40. 97. 5.84 2.2 0.1 2.2 2315000200039 4.4 2315000200040 42. 99. 6.18 2.0 0.3 2.0 2315000200041 4.0 2315000200042 44. 103. 2.97 2.1 1.3 2.5 2315000200043 4.2 2315000200044 45. 105. 0.92 2.75 3.2 4.2 2315000200045 5.5 2315000200046 52. 132. 4.23 2.0 3.4 4.0 2315000200047 4.0 2315000200048 53. 131. 2.86 2.15 1.0 2.4 2315000200049 4.3 2315000200050 56. 140. 6.15 2.1 0.9 2.3 2315000200051 4.2 2315000200052 58. 141. 5.92 2.0 0.4 2.0 2315000200053 4.0 2315000200054 58. 143. 5.81 2.1 0.9 2.3 2315000200055 4.2 2315000200056 60. 147. 2.11 2.1 4.2 4.7 2315000200057 4.2 2315000200058 ENDDATA 26 0 2315000200059 ENDSUBENT 58 0 2315000299999 SUBENT 23150003 20111113 22282315000300001 BIB 5 24 2315000300002 REACTION (94-PU-239(N,F)ELEM/MASS,CUM,FY,,MXW) 2315000300003 Fission yields in a thermal neutron spectrum. 2315000300004 SAMPLE .PU-O(2) Targets on aluminum or titanium foils. 2315000300005 Target diameter 14 mm, enrichment in Pu-239 better 2315000300006 than 98. percent. 2315000300007 Isotopic ratios Pu-240/Pu-239 1.71*10**-2, 2315000300008 Pu-241/Pu-239 3.7*10**-4. 2315000300009 Mass 2.23+-0.06 microg (measured by alpha-spectrometry2315000300010 2.26+-0.02 (measured by thermo-ionization mass 2315000300011 spectrometry). 2315000300012 ERR-ANALYS (ERR-T) Total uncertainty ( 1 sigma) as combination 2315000300013 in quadrature of: 2315000300014 (ERR-1) Uncertainty (1 sigma) associated with 2315000300015 measurements, i.e. combination of statistical and 2315000300016 systematic uncertainty; 2315000300017 derived from authors' errors - see MISC-COL. 2315000300018 (ERR-2) Uncertainty ( 1 sigma) due to nuclear data. 2315000300019 MISC-COL (MISC) Uncertainty (2 sigma) associated with 2315000300020 measurements, i.e. combination of statistical and 2315000300021 systematic uncertainty, given by authors in article. 2315000300022 Were re-calculated (/2.) as ERR-1 to obtain 1 sigma 2315000300023 error; authors' errors are given as MISC according to 2315000300024 the comment of N.Otsuka (NDS,IAEA) . 2315000300025 STATUS (TABLE) Table VII of J,NDS,111,2965,2010. 2315000300026 ENDBIB 24 0 2315000300027 COMMON 1 3 2315000300028 EN-DUMMY 2315000300029 EV 2315000300030 0.0253 2315000300031 ENDCOMMON 3 0 2315000300032 DATA 7 10 2315000300033 ELEMENT MASS DATA ERR-1 ERR-2 ERR-T 2315000300034 MISC 2315000300035 NO-DIM NO-DIM PC/FIS PER-CENT PER-CENT PER-CENT 2315000300036 PER-CENT 2315000300037 40. 95. 4.81 2.3 0.3 2.3 2315000300038 4.6 2315000300039 40. 97. 5.31 1.85 0.1 1.8 2315000300040 3.7 2315000300041 42. 99. 6.54 1.75 0.3 1.8 2315000300042 3.5 2315000300043 44. 103. 7.08 2.0 1.3 2.4 2315000300044 4.0 2315000300045 52. 132. 5.25 1.9 3.4 3.9 2315000300046 3.8 2315000300047 53. 131. 4.06 2.0 1.0 2.2 2315000300048 4.0 2315000300049 56. 140. 5.43 2.25 0.9 2.4 2315000300050 4.5 2315000300051 58. 141. 5.54 1.85 0.4 1.9 2315000300052 3.7 2315000300053 58. 143. 4.39 2.0 0.9 2.2 2315000300054 4.0 2315000300055 60. 147. 2.02 2.0 4.2 4.7 2315000300056 4.0 2315000300057 ENDDATA 24 0 2315000300058 ENDSUBENT 57 0 2315000399999 ENDENTRY 3 0 2315099999999