ENTRY M0692 20110119 M056M069200000001
SUBENT M0692001 20110119 M056M069200100001
BIB 15 97 M069200100002
TITLE Photofission cross sections for 237Np in the energy M069200100003
interval from 5.27 to 10.83 MeV. M069200100004
AUTHOR (L.P.GERALDO,R.SEMMLER,O.I.GONCALEZ,J.MESA, M069200100005
J.D.T.ARRUDA-NETO,F.GARCIA,O.RODRIGUES) M069200100006
REFERENCE (J,NSE,136,357,2000) M069200100007
INSTITUTE (3BZLBZL,3BZLIPE,3BZLCTA,3BZLUSP) M069200100008
FACILITY (REAC,3BZLIPE) IPEN IAE-R1 MW pool-type research M069200100009
reactor. M069200100010
INC-SOURCE (MPH) High-resolution (4 - 20 eV) gamma-rays, with M069200100011
discrete energies up to 10.83 MeV, were produced by M069200100012
thermal neutron capture in 30 different target M069200100013
materials, when they were positioned near the M069200100014
reactor core: M069200100015
Capture target Gamma-ray energy, keV M069200100016
---------------------------------------------------- M069200100017
Yb 5265.70 M069200100018
S 5420.50 M069200100019
Dy 5607.75 M069200100020
Hf 5723.50 M069200100021
In 5891.90 M069200100022
Hg 5966.20 M069200100023
Y 6080.49 M069200100024
Er 6228.23 M069200100025
Na 6395.40 M069200100026
Ca 6419.90 M069200100027
Nd 6501.70 M069200100028
V 6517.26 M069200100029
Gd 6748.70 M069200100030
Ti 6760.13 M069200100031
Be 6809.41 M069200100032
As 7019.45 M069200100033
Sm 7214.20 M069200100034
Mn 7243.79 M069200100035
Pb 7367.70 M069200100036
Cl 7413.80 M069200100037
Se 7418.47 M069200100038
Fe 7631.13 M069200100039
Al 7723.85 M069200100040
K 7770.22 M069200100041
Zn 7863.30 M069200100042
Cu 7914.50 M069200100043
Cd 8484.80 M069200100044
Ni 8998.80 M069200100045
Cr 9720.30 M069200100046
N 10829.18 M069200100047
SAMPLE The sample consisted of 36.10 (0.13 mg) of 237Np M069200100048
deposited in form of NpO-2 on six titanium disks, M069200100049
forming an active diameter of ~ 40 mm. M069200100050
DETECTOR (TRD) Makrofol KG foils (10 micron thickness) in form M069200100051
of a sandwich (2PI geometry). M069200100052
METHOD (EXTB,STTA) M069200100053
ANALYSIS (UNFLD) The method of unfolding of cross section data M069200100054
for definite gamma-lines from those obtained for M069200100055
composition of all gamma-lines used employed M069200100056
least-squares methods and the covariance matrix M069200100057
methodology. M069200100058
REL-REF (N,,O.L.GONCALES+,J,NSE,132,135,1999) The method for M069200100059
unfolding. M069200100060
ERR-ANALYS (DATA-ERR) The overall uncertainties were estimated M069200100061
by summing in quadrature all following components: M069200100062
Error Sources Range (%) M069200100063
--------------------------------------------------- M069200100064
Random errors M069200100065
(0 % uncorrelated errors) M069200100066
Fission track counting 0.43 - 6.83 M069200100067
Gamma-ray counting 0.19 - 4.83 M069200100068
--------------------------------------------------- M069200100069
Systematic errors M069200100070
(100 % completely correlated errors) M069200100071
(ERR-1) 237Np deposit mass 0.35 M069200100072
(ERR-2) Gamma-ray intensity ratio 0.29 M069200100073
(ERR-3) Gamma-ray intensity attenuation 0.34 M069200100074
(ERR-4) 252Cf activity 1.8 M069200100075
(ERR-5) Fission track calibration factor 0.80 M069200100076
(ERR-6,0.77,3.51) Ge(Li) calibration factor 0.77 - M069200100077
3.51 M069200100078
COMMENT Photofission cross sections for 237 Np have been M069200100079
measured as a function of energy, in the interval M069200100080
from 5.27 to 10.83 MeV. The gamma-ray spectra were M069200100081
those produced by thermal neutron capture, in 30 M069200100082
different target materials, as a tangential beam M069200100083
hole of the Instituto de Pesquisas Energeticas e M069200100084
Nucleares IAE-R1 1-MW research reactor. The set of M069200100085
experimental data has been unfolded employing M069200100086
least-squares methods and the covariance matrix M069200100087
methodology. The determined photofission cross M069200100088
sections for 237Np, together with the complete M069200100089
correlation matrix for the involved errors, are M069200100090
presented and are compared with previous M069200100091
measurements reported in the literature. A M069200100092
statistical calculation for the 237Np photofission M069200100093
cross sections was performed, and the results are M069200100094
compared with the experimental data. M069200100095
STATUS (TABLE) Data from the Table III were compiled at the M069200100096
Russia MSU SINP CDFE by V.Varlamov. M069200100097
HISTORY (20060626C) M069200100098
(20110119A) Corrected by V.Varlamov: COMMON. M069200100099
ENDBIB 97 0 M069200100100
COMMON 5 3 M069200100101
ERR-1 ERR-2 ERR-3 ERR-4 ERR-5 M069200100102
PER-CENT PER-CENT PER-CENT PER-CENT PER-CENT M069200100103
.35 .29 .34 1.8 .8 M069200100104
ENDCOMMON 3 0 M069200100105
ENDSUBENT 104 0 M069200199999
SUBENT M0692002 20110119 M056M069200200001
BIB 2 5 M069200200002
REACTION (93-NP-237(G,F),,SIG) M069200200003
REL-REF (N,,A.S.SOLDATOV+,R,INDC(CCP)-379,1994) The overall M069200200004
agreement is reasonable between the 237Np(g,f) M069200200005
reaction cross section data obtained and evaluated M069200200006
before. M069200200007
ENDBIB 5 0 M069200200008
NOCOMMON 0 0 M069200200009
DATA 3 30 M069200200010
EN DATA DATA-ERR M069200200011
MEV MB MB M069200200012
5.2657 4.38 .7 M069200200013
5.4205 2.75 1.15 M069200200014
5.60775 8.19 .96 M069200200015
5.7235 14.53 1.28 M069200200016
5.8919 6.72 1.89 M069200200017
5.9662 16.23 .88 M069200200018
6.08049 12.57 .52 M069200200019
6.22823 14.56 3.76 M069200200020
6.3954 22.43 .6 M069200200021
6.4199 18.08 1.7 M069200200022
6.5017 24.5 1.43 M069200200023
6.51726 16.34 4.53 M069200200024
6.7487 37.34 2.16 M069200200025
6.76013 19.23 1.36 M069200200026
6.80941 22.26 1.27 M069200200027
7.01945 20.59 3.85 M069200200028
7.2142 19.73 4.34 M069200200029
7.24379 41.61 4.34 M069200200030
7.3677 34.74 1.08 M069200200031
7.4138 30.99 8.44 M069200200032
7.41847 27.04 12.48 M069200200033
7.63113 30.9 1.94 M069200200034
7.72385 42.22 1.45 M069200200035
7.77022 31.12 3.43 M069200200036
7.8633 49.67 3.26 M069200200037
7.9145 43.45 1.98 M069200200038
8.4848 42.39 32.17 M069200200039
8.9988 53.78 14.89 M069200200040
9.7203 186.02 20.77 M069200200041
10.82918 252.89 16. M069200200042
ENDDATA 32 0 M069200200043
ENDSUBENT 42 0 M069200299999
ENDENTRY 2 0 M069299999999