ENTRY 10115 20000202 00001011500000001 SUBENT 10115001 20000202 00001011500100001 BIB 11 24 1011500100002 INSTITUTE (1USALAS) 1011500100003 REFERENCE (J,NSE,43,54,197101) 1011500100004 AUTHOR (C.J.ORTH) 1011500100005 TITLE The average number of neutrons emitted in the 1011500100006 spontaneous fission of some even-even heavy nuclides 1011500100007 METHOD Determined by counting neutrons and fissions separately1011500100008 DETECTOR Neutrons counted in eight B10-lined tubes embedded on 1011500100009 an 8 in. diameter in a 16 in. diameter by 15 in. long 1011500100010 cylinder of high density polyethylene. Tubes connected1011500100011 in parallel. With discriminator set above 60Co gammas,1011500100012 Efficiency for 252Cf spontaneous neutrons - 3.84%. 1011500100013 (PROPC) Fissions counted with gas-flow proportional 1011500100014 counters in 2 pi geometry, biased above alphas. 1011500100015 MONITOR (98-CF-252(0,F),,NU) 1011500100016 CORRECTION Not corrected for delayed neutrons (considered 1011500100017 negligible). 1011500100018 Corrected for neutrons from (alpha,n) reactions. 1011500100019 ERR-ANALYS No information. 1011500100020 STATUS (APRVD) Approved by author. 1011500100021 Values from table in reference. 1011500100022 HISTORY (19710618C) 1011500100023 (19820512A) BIB corrections. 1011500100024 (19820512A) Converted to REACTION formalism 1011500100025 (20000202A) Converted to new date format, lower case. 1011500100026 ENDBIB 24 0 1011500100027 COMMON 1 3 1011500100028 MONIT 1011500100029 NO-DIM 1011500100030 3.77 1011500100031 ENDCOMMON 3 0 1011500100032 ENDSUBENT 31 0 1011500199999 SUBENT 10115002 20000202 00001011500200001 BIB 3 9 1011500200002 REACTION (94-PU-244(0,F),,NU) 1011500200003 SAMPLE 13.6-mg PuO2 containing 99.06 atom-percent 244Pu 1011500200004 enclosed in pyrex tube for neutron counting. 1011500200005 ERR-ANALYS (DATA-ERR) Error includes uncertainties in 1011500200006 . spontaneous fission half-life, 1011500200007 . neutron counter efficiency, 1011500200008 . sample weight, 1011500200009 . small difference in delayed neutron contribution 1011500200010 compared to 252Cf. 1011500200011 ENDBIB 9 0 1011500200012 NOCOMMON 0 0 1011500200013 DATA 2 1 1011500200014 DATA DATA-ERR 1011500200015 PART/FIS PART/FIS 1011500200016 2.30 0.19 1011500200017 ENDDATA 3 0 1011500200018 ENDSUBENT 17 0 1011500299999 SUBENT 10115003 20000202 00001011500300001 BIB 3 6 1011500300002 REACTION (96-CM-248(0,F),,NU) 1011500300003 SAMPLE 2 micrograms of 248Cm. Diluted to 1 ml for neutron 1011500300004 counting. 1011500300005 ANALYSIS 7.5% of alphas from sample are due to 246Cm, therefore,1011500300006 calculated that more than 99.5% of spontaneous 1011500300007 fissions due to 248Cm. 1011500300008 ENDBIB 6 0 1011500300009 NOCOMMON 0 0 1011500300010 DATA 2 1 1011500300011 DATA DATA-ERR 1011500300012 PART/FIS PART/FIS 1011500300013 3.11 0.09 1011500300014 ENDDATA 3 0 1011500300015 ENDSUBENT 14 0 1011500399999 SUBENT 10115004 20000202 00001011500400001 BIB 3 6 1011500400002 REACTION (96-CM-250(0,F),,NU) 1011500400003 SAMPLE Curium fraction from a portion of the radioactive 1011500400004 debris from a heavy-element-producing nuclear 1011500400005 explosion (Hutch). 1011500400006 ANALYSIS Alpha analysis indicated 98.5% of spontaneous fissions 1011500400007 due to 250Cm. 1011500400008 ENDBIB 6 0 1011500400009 NOCOMMON 0 0 1011500400010 DATA 2 1 1011500400011 DATA DATA-ERR 1011500400012 PART/FIS PART/FIS 1011500400013 3.31 0.08 1011500400014 ENDDATA 3 0 1011500400015 ENDSUBENT 14 0 1011500499999 SUBENT 10115005 20000202 00001011500500001 BIB 3 4 1011500500002 REACTION (98-CF-250(0,F),,NU) 1011500500003 SAMPLE Three 250Cf samples prepared by irradiation of freshly 1011500500004 purified 249Bk in reactors. 1011500500005 ANALYSIS 250Cf spontaneous fission contributions > 90%. 1011500500006 ENDBIB 4 0 1011500500007 NOCOMMON 0 0 1011500500008 DATA 2 1 1011500500009 DATA DATA-ERR 1011500500010 PART/FIS PART/FIS 1011500500011 3.53 0.09 1011500500012 ENDDATA 3 0 1011500500013 ENDSUBENT 12 0 1011500599999 SUBENT 10115006 20000202 00001011500600001 BIB 3 5 1011500600002 REACTION (98-CF-254(0,F),,NU) 1011500600003 SAMPLE CF samples radiochemically separated from debris of 1011500600004 heavy element production shot. 1011500600005 ANALYSIS 98% of spontaneous fissions in sources due to 254Cf, 1011500600006 2% to 252Cf. 1011500600007 ENDBIB 5 0 1011500600008 NOCOMMON 0 0 1011500600009 DATA 2 1 1011500600010 DATA DATA-ERR 1011500600011 PART/FIS PART/FIS 1011500600012 3.93 0.05 1011500600013 ENDDATA 3 0 1011500600014 ENDSUBENT 13 0 1011500699999 ENDENTRY 6 0 1011599999999