ENTRY            13913   20240314                             15071391300000001 
SUBENT        13913001   20240314                             15071391300100001 
BIB                 10         21                                 1391300100002 
AUTHOR     (J.A.Grundl)                                           1391300100003 
TITLE      Measurement of the Average Fission Cross-Section       1391300100004 
           Ratio, sigmaF(235U)/sigmaF(238U) for 235U and 239Pu    1391300100005 
           Fission Neutrons                                       1391300100006 
REFERENCE  (C,71VIENNA,,107,1971)                                 1391300100007 
REL-REF    (N,10304001,J.A.Grundl+,J,ANS,15,945,1972)             1391300100008 
           Measurements with 252Cf neutron source.                1391300100009 
INSTITUTE  (1USANBS)                                              1391300100010 
FACILITY   (REAC,1USANBS)                                         1391300100011 
SAMPLE     (92-U-235,ENR=0.997) Two foils: Enriched and natural   1391300100012 
           uranium, 200 microg/cm2 thick.                         1391300100013 
DETECTOR   (IOCH)                                                 1391300100014 
MONITOR    (92-U-235(N,F),,SIG)   Run to run thermal neutron flux 1391300100015 
           levels in the cavity were monitored with a commercial  1391300100016 
           235U fission chamber exposed to a beam taken of the    1391300100017 
           front face of the thermal column. A triple scaler      1391300100018 
           arrangement was also used to record pulses from this   1391300100019 
           monitor chamber.                                       1391300100020 
HISTORY    (20140804C) BP                                         1391300100021 
           (20230604A) OS. Uncertainties modified in subs.2,3     1391300100022 
           (20240314A) On. Major alterations in 002.              1391300100023 
ENDBIB              21          0                                 1391300100024 
NOCOMMON             0          0                                 1391300100025 
ENDSUBENT           24          0                                 1391300199999 
SUBENT        13913002   20240314                             15071391300200001 
BIB                  7         20                                 1391300200002 
REACTION   ((92-U-235(N,F),,SIG,,FIS)/(92-U-238(N,F),,SIG,,FIS))  1391300200003 
STATUS     (TABLE,,J.A.Grundl,C,71VIENNA,,107,1971) page 108.     1391300200004 
METHOD     (PHD)                                                  1391300200005 
INC-SOURCE Fission rates were observed with 235U fission          1391300200006 
           neutrons source.                                       1391300200007 
INC-SPECT  Energy of 235U fission neutrons is not listed in the   1391300200008 
           article. We can assume average fission neutrons        1391300200009 
           energy of ~2 MeV.                                      1391300200010 
ERR-ANALYS (ERR-T) Uncertainties are due to:                      1391300200011 
           (ERR-S) counting statistics - 0.3%.                    1391300200012 
           (ERR-1) thermal column monitor stability - 0.6%.       1391300200013 
           (ERR-2) source-detector position - 2%.                 1391300200014 
           (ERR-3) cavity position - 1.2%.                        1391300200015 
           (ERR-4) reactor background - 2%.                       1391300200016 
           (ERR-5) cavity return background - 2.5%.               1391300200017 
           (ERR-6) relative detector efficiency - 0.3%.           1391300200018 
           (ERR-7) intrinsic source-detector scattering - 1.5%.   1391300200019 
           (ERR-8) foil weight ratio - 2%.                        1391300200020 
HISTORY    (20230604A) ERR-T in DATA section deleted              1391300200021 
           (20240314A) On. EN-DUMMY=2.0 MeV -> KT-DUMMY=1.32 MeV  1391300200022 
ENDBIB              20          0                                 1391300200023 
COMMON              10          6                                 1391300200024 
ERR-T      ERR-S      ERR-1      ERR-2      ERR-3      ERR-4      1391300200025 
ERR-5      ERR-6      ERR-7      ERR-8                            1391300200026 
PER-CENT   PER-CENT   PER-CENT   PER-CENT   PER-CENT   PER-CENT   1391300200027 
PER-CENT   PER-CENT   PER-CENT   PER-CENT                         1391300200028 
        4.7        0.3        0.6        2.0        1.2        2.01391300200029 
        2.5        0.3        1.5        2.0                      1391300200030 
ENDCOMMON            6          0                                 1391300200031 
DATA                 2          1                                 1391300200032 
KT-DUMMY   DATA                                                   1391300200033 
MEV        NO-DIM                                                 1391300200034 
       1.32       3.71                                            1391300200035 
ENDDATA              3          0                                 1391300200036 
ENDSUBENT           35          0                                 1391300299999 
SUBENT        13913003   20230707                             15071391300300001 
BIB                  8         23                                 1391300300002 
REACTION   (((92-U-235(N,F),,SIG,,FIS)/                           1391300300003 
           (92-U-238(N,F),,SIG,,FIS))//                           1391300300004 
           ((92-U-235(N,F),,SIG,,FIS)/(92-U-238(N,F),,SIG,,FIS))) 1391300300005 
STATUS     (TABLE,,J.A.Grundl,C,71VIENNA,,107,1971) page 108.     1391300300006 
METHOD     (PHD)                                                  1391300300007 
INC-SOURCE Fission rates were observed with 235U (denominator     1391300300008 
           ratio) and 239Pu (nominator ratio) fission neutron     1391300300009 
           sources.                                               1391300300010 
INC-SPECT  Energies of 235U and 239Pu fission neutrons are not    1391300300011 
           listed in the article. We can assume that 235U and     1391300300012 
           239Pu average fission neutrons energies are ~2 and     1391300300013 
           ~2.1 MeV, respectively.                                1391300300014 
COMMENT    From the knowledge of 235U fission source ratio of     1391300300015 
           3.71 and 239Pu/235U sources ratio of 9.70, one may     1391300300016 
           reconstruct a fission cross section ratio for a pure   1391300300017 
           239Pu source as 3.5987 +- 0.1704.                      1391300300018 
ERR-ANALYS (ERR-T) Uncertainties are due to:                      1391300300019 
           (ERR-S) counting statistics - 0.4%.                    1391300300020 
           (ERR-1) thermal column monitor stability - 0.6%.       1391300300021 
           (ERR-2) source-detector position - 0.7%.               1391300300022 
           (ERR-3) cavity position - 0.5%.                        1391300300023 
           (ERR-4) reactor background - 0.6%.                     1391300300024 
HISTORY    (20230604A) ERR-T in DATA section deleted              1391300300025 
ENDBIB              23          0                                 1391300300026 
COMMON               6          3                                 1391300300027 
ERR-T      ERR-S      ERR-1      ERR-2      ERR-3      ERR-4      1391300300028 
PER-CENT   PER-CENT   PER-CENT   PER-CENT   PER-CENT   PER-CENT   1391300300029 
        1.3        0.4        0.6        0.7        0.5        0.61391300300030 
ENDCOMMON            3          0                                 1391300300031 
DATA                 3          1                                 1391300300032 
EN-DUM-NM  EN-DUM-DN  DATA                                        1391300300033 
MEV        MEV        NO-DIM                                      1391300300034 
        2.1        2.0      0.970                                 1391300300035 
ENDDATA              3          0                                 1391300300036 
ENDSUBENT           35          0                                 1391300399999 
ENDENTRY             3          0                                 1391399999999