ENTRY            10226   20010309                             00001022600000001 
SUBENT        10226001   20010309                             00001022600100001 
BIB                 15         30                                 1022600100002 
INSTITUTE  (1USALAS)                                              1022600100003 
REFERENCE  (J,NP/A,191,641,197209)                                1022600100004 
AUTHOR     (I.BERGQVIST,D.M.DRAKE,D.K.MC DANIELS)                 1022600100005 
TITLE      Radiative capture of energetic neutrons by 208Pb       1022600100006 
FACILITY   (VDG) LASL tandem Van de Graaff accelerator            1022600100007 
INC-SOURCE (P-T) 3H(p,n)3He reaction                              1022600100008 
SAMPLE     Lead sample consisted of 98.7% isotopically pure 208Pb 1022600100009 
            in cylindrical form.                                  1022600100010 
METHOD     (BCINT)                                                1022600100011 
           Time of flight technique used to reduce background and 1022600100012 
            to separate gammas due to other sources               1022600100013 
DETECTOR   (NAICR) Gamma rays detected with an anti-Compton       1022600100014 
            spectrometer consisting of a center NaI(Tl) crystal   1022600100015 
            surrounded by a NaI(Tl) annulus.                      1022600100016 
           Neutron flux measured using counter telescope          1022600100017 
            consisting of a NaI(Tl) crystal.                      1022600100018 
ANALYSIS   (4PI1A) Data obtained from angular distribution at 90  1022600100019 
            degrees.                                              1022600100020 
MONITOR    Gamma ray detector efficiency determined as a function 1022600100021 
            of energy by utilizing gammas from 11B(p,gamma)12C,   1022600100022 
            12C(p,p'gamma)12C, and 89Y(p,gamma)90Zr.              1022600100023 
CORRECTION Corrected for background.                              1022600100024 
ERR-ANALYS (DATA-ERR) Total uncertainty in cross section.         1022600100025 
           Uncertainty in neutron flux across sample estimated to 1022600100026 
            be 5.0%.                                              1022600100027 
STATUS     (APRVD) Approved by author                             1022600100028 
           Tabulated data received from D. Drake, February 1972.  1022600100029 
HISTORY    (19730803C)                                            1022600100030 
           (19821015A) Converted to REACTION formalism            1022600100031 
           (20010309A) Updated to new date formats, lower case.   1022600100032 
ENDBIB              30          0                                 1022600100033 
NOCOMMON             0          0                                 1022600100034 
ENDSUBENT           33          0                                 1022600199999 
SUBENT        10226002   20010309                             00001022600200001 
BIB                  1          1                                 1022600200002 
REACTION   (82-PB-208(N,G)82-PB-209,,SIG)                         1022600200003 
ENDBIB               1          0                                 1022600200004 
COMMON               1          3                                 1022600200005 
DATA-ERR                                                          1022600200006 
PER-CENT                                                          1022600200007 
 15.                                                              1022600200008 
ENDCOMMON            3          0                                 1022600200009 
DATA                 2          8                                 1022600200010 
EN         DATA                                                   1022600200011 
MEV        MICRO-B                                                1022600200012 
 7.2        831.0                                                 1022600200013 
 8.2       1243.0                                                 1022600200014 
 9.2       1699.0                                                 1022600200015 
10.2       1786.0                                                 1022600200016 
11.2       1879.0                                                 1022600200017 
12.2       1520.0                                                 1022600200018 
13.2       1201.0                                                 1022600200019 
14.7        943.0                                                 1022600200020 
ENDDATA             10          0                                 1022600200021 
ENDSUBENT           20          0                                 1022600299999 
ENDENTRY             2          0                                 1022699999999