ENTRY 13776 20010927 00001377600000001 SUBENT 13776001 20010927 00001377600100001 BIB 14 48 1377600100002 INSTITUTE (1USAORL) 1377600100003 REFERENCE (J,PR,91,655,195308) 1377600100004 AUTHOR (R.W.Lamphere) 1377600100005 TITLE Fission Cross Section of Uranium-234 1377600100006 FACILITY (VDG) Oak Ridge 5-Mev Van de Graaff. 1377600100007 INC-SOURCE (P-T). Tritium gas target. 1377600100008 SAMPLE 2 foils of 234U plated over a one-inch circle on 2-mm 1377600100009 thick platinum backing, one thin foil containing 1377600100010 about 1 mg of uranium, and one thick foil containing 1377600100011 about 4 mg. The quantity of uranium in the thin foil 1377600100012 known accurately. 1377600100013 2 monitor foils of 235U plated in the same way as 1377600100014 sample foils, thin and thick foils of same weights as 1377600100015 234U foils. 1377600100016 234U and 235U foils placed back to back. 1377600100017 MONITOR (92-U-235(N,F),,SIG) 1377600100018 DETECTOR (FISCH) A double ionization chamber. 1377600100019 A mixture of 97% argon plus 3% carbon dioxide at a 1377600100020 pressure of two atmospheres supplied with a flow of 1377600100021 0.1 cubic foot per hour. The chamber was surrounded 1377600100022 by a 1/32-inch thick shield of cadmium to absorb 1377600100023 slow neutrons. 1377600100024 METHOD For calibration the thin and thick foils of the same 1377600100025 isotopic content were placed in the fission chamber 1377600100026 back to back, and exposed to neutrons from a Po-Be 1377600100027 source for several hours. The fission yields ratios 1377600100028 were found to 0.2% for the pair of 234U foils and 1377600100029 for the pair of 235U foils. 1377600100030 Then the thick foils used to find the 234U to 325U 1377600100031 cross section ratio. 1377600100032 CORRECTION Corrected for: 1377600100033 . background, and counts from pileup of alpha pulses 1377600100034 in the 234U (<0.01%). 1377600100035 . different absorption losses of the fission 1377600100036 fragments for foils facing toward and away neutron 1377600100037 source (~1% at 4 MeV). 1377600100038 Not corrected for effect of room scattered neutrons 1377600100039 on 235U fission cross section (correction factor 1377600100040 determined to be 1.01+-0.01). 1377600100041 ERR-ANALYS Contributions to uncertainties: 1377600100042 . quantity of uranium deposited on foils (14%); 1377600100043 . effect of room scattered neutrons (1%); 1377600100044 . beam momentum (0.15%); 1377600100045 . amplifier gain changes (0.12%); 1377600100046 . statistics (1.5-3%). 1377600100047 STATUS (CURVE) Data read from curve in report LA-1714 (taken 1377600100048 from Barshall Curves). 1377600100049 HISTORY (20010921C) IS 1377600100050 ENDBIB 48 0 1377600100051 NOCOMMON 0 0 1377600100052 ENDSUBENT 51 0 1377600199999 SUBENT 13776002 20010927 00001377600200001 BIB 1 1 1377600200002 REACTION (92-U-234(N,F),,SIG) 1377600200003 ENDBIB 1 0 1377600200004 NOCOMMON 0 0 1377600200005 DATA 2 68 1377600200006 EN DATA 1377600200007 MEV B 1377600200008 0.38 0.130 1377600200009 0.40 0.150 1377600200010 0.43 0.190 1377600200011 0.47 0.300 1377600200012 0.49 0.340 1377600200013 0.53 0.460 1377600200014 0.55 0.560 1377600200015 0.57 0.590 1377600200016 0.60 0.685 1377600200017 0.64 0.740 1377600200018 0.67 0.845 1377600200019 0.70 0.970 1377600200020 0.72 1.020 1377600200021 0.77 1.160 1377600200022 0.81 1.260 1377600200023 0.85 1.315 1377600200024 0.89 1.285 1377600200025 0.94 1.230 1377600200026 0.98 1.210 1377600200027 1.02 1.150 1377600200028 1.07 1.120 1377600200029 1.12 1.150 1377600200030 1.14 1.190 1377600200031 1.16 1.270 1377600200032 1.20 1.280 1377600200033 1.25 1.210 1377600200034 1.29 1.255 1377600200035 1.34 1.245 1377600200036 1.38 1.285 1377600200037 1.43 1.305 1377600200038 1.49 1.310 1377600200039 1.52 1.400 1377600200040 1.62 1.475 1377600200041 1.71 1.490 1377600200042 1.76 1.470 1377600200043 1.81 1.460 1377600200044 1.85 1.580 1377600200045 1.90 1.530 1377600200046 1.95 1.570 1377600200047 2.00 1.585 1377600200048 2.10 1.500 1377600200049 2.20 1.525 1377600200050 2.25 1.490 1377600200051 2.30 1.450 1377600200052 2.35 1.440 1377600200053 2.40 1.470 1377600200054 2.51 1.440 1377600200055 2.61 1.505 1377600200056 2.72 1.500 1377600200057 2.82 1.505 1377600200058 2.88 1.475 1377600200059 2.93 1.560 1377600200060 2.98 1.590 1377600200061 3.04 1.540 1377600200062 3.09 1.540 1377600200063 3.15 1.505 1377600200064 3.20 1.550 1377600200065 3.26 1.545 1377600200066 3.37 1.550 1377600200067 3.48 1.550 1377600200068 3.59 1.590 1377600200069 3.70 1.545 1377600200070 3.76 1.500 1377600200071 3.82 1.525 1377600200072 3.88 1.565 1377600200073 3.94 1.535 1377600200074 4.00 1.525 1377600200075 4.05 1.545 1377600200076 ENDDATA 70 0 1377600200077 ENDSUBENT 76 0 1377600299999 ENDENTRY 2 0 1377699999999