47-Ag-109 JAEA EVAL-Dec09 N.Iwamoto,K.Shibata
DIST-MAY10 20100119
----JENDL-4.0 MATERIAL 4731
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
09-12 The resolved resonance parameters were evaluated by
K.Shibata.
The data above the resolved resonance region were evaluated
and compiled by N.Iwamoto.
MF= 1 General information
MT=451 Descriptive data and directory
MF= 2 Resonance parameters
MT=151 Resolved and unresolved resonances
Resolved resonance parameters (below 7.0095 keV)
Resolved resonance parameters (below 7.0095keV) for
JENDL-3.3 were based on the experimental data by Moxon and
Rae /1/, Garg et al./$ga/, Asghar et al./2/, Muradjan
and Adamchuk /3/, de Barros et al./4/, Pattenden and
Jolly /5/, macklin /6/ and Mizumoto et al./7/.
Total spin j and angular momentum l of some resonances were
estimated with a random number method and a method of
Bollinger and Thomas/8/, respectively.
The capture cross section of JENDL-3.3 between 1.3 and 2.6
keV is too low compared with interpolated values from the
lower and higher energy regions. To compensate the lower
capture cross section, p-wave resonances with a capture area
of 0.02 eV were added every 20 eV between 1.25 and 1.59 keV,
and every 40 eV between 1.59 and 2.59 keV. The neutron
width was modified so as to reproduce teh capture area
measured by Macklin /6/.
In JENDL-4, the data for 5.19 eV - 499.8 eV were replaced
with the ones obtained by Lowie et al./9/ A value of 130
meV was used for unknown radiation widths. The neutron
width for 264.7 eV was taken from JENDL-3.3. Part of
unknown j values were determined by considering the work of
Zanini et al./10/ The remaining unknown j values were
estimated by a random number method.
Unresolved resonance region : 7.0095 keV - 130 keV
The unresolved resonance paramters (URP) were determined by
ASREP code /11/ so as to reproduce the evaluated total and
capture cross sections calculated with optical model code
OPTMAN /12/ and CCONE /13/. The unresolved parameters
should be used only for self-shielding calculation.
Thermal cross sections and resonance integrals at 300 K
----------------------------------------------------------
0.0253 eV res. integ. (*)
(barn) (barn)
----------------------------------------------------------
Total 9.2769e+01
Elastic 2.5069e+00
n,gamma 9.0262e+01 1.4647e+03
n,alpha 1.1306e-12
----------------------------------------------------------
(*) Integrated from 0.5 eV to 10 MeV.
MF= 3 Neutron cross sections
MT= 1 Total cross section
Sum of partial cross sections.
MT= 2 Elastic scattering cross section
Obtained by subtracting non-elastic scattering cross sections
from total cross section.
MT= 4 (n,n') cross section
Calculated with CCONE code /13/.
MT= 16 (n,2n) cross section
Calculated with CCONE code /13/.
MT= 17 (n,3n) cross section
Calculated with CCONE code /13/.
MT= 22 (n,na) cross section
Calculated with CCONE code /13/.
MT= 28 (n,np) cross section
Calculated with CCONE code /13/.
MT= 32 (n,nd) cross section
Calculated with CCONE code /13/.
MT= 33 (n,nt) cross section
Calculated with CCONE code /13/.
MT= 41 (n,2np) cross section
Calculated with CCONE code /13/.
MT= 51-91 (n,n') cross section
Calculated with CCONE code /13/.
MT=102 Capture cross section
Calculated with CCONE code /13/.
MT=103 (n,p) cross section
Calculated with CCONE code /13/.
MT=104 (n,d) cross section
Calculated with CCONE code /13/.
MT=105 (n,t) cross section
Calculated with CCONE code /13/.
MT=106 (n,He3) cross section
Calculated with CCONE code /13/.
MT=107 (n,a) cross section
Calculated with CCONE code /13/.
MF= 4 Angular distributions of emitted neutrons
MT= 2 Elastic scattering
Calculated with CCONE code /13/.
MF= 6 Energy-angle distributions of emitted particles
MT= 16 (n,2n) reaction
Calculated with CCONE code /13/.
MT= 17 (n,3n) reaction
Calculated with CCONE code /13/.
MT= 22 (n,na) reaction
Calculated with CCONE code /13/.
MT= 28 (n,np) reaction
Calculated with CCONE code /13/.
MT= 32 (n,nd) reaction
Calculated with CCONE code /13/.
MT= 33 (n,nt) reaction
Calculated with CCONE code /13/.
MT= 41 (n,2np) reaction
Calculated with CCONE code /13/.
MT= 51-91 (n,n') reaction
Calculated with CCONE code /13/.
MT=102 Capture reaction
Calculated with CCONE code /13/.
*****************************************************************
Nuclear Model Calculation with CCONE code /13/
*****************************************************************
Models and parameters used in the CCONE calculation
1) Optical model
* coupled channels calculation
coupled levels: 0,3,4,18 (see Table 1)
* optical model potential
neutron omp: Kunieda,S. et al./14/ (+)
proton omp: Koning,A.J. and Delaroche,J.P./15/
deuteron omp: Lohr,J.M. and Haeberli,W./16/
triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./17/
He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./17/
alpha omp: Huizenga,J.R. and Igo,G./18/
(+) omp parameters were modified.
* resonance for pseudo levels was calculated on the basis of
DWBA.
ER= 2.500 (MeV) WIDTH= 0.400 (MeV) L= 3 BETA= 0.155
2) Two-component exciton model/19/
* Global parametrization of Koning-Duijvestijn/20/
was used.
* Gamma emission channel/21/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Width fluctuation correction/22/ was applied.
* Neutron, proton, deuteron, triton, He3, alpha and gamma
decay channel were taken into account.
* Transmission coefficients of neutrons were taken from
optical model calculation.
* The level scheme of the target is shown in Table 1.
* Level density formula of constant temperature and Fermi-gas
model were used with shell energy correction/23/.
Parameters are shown in Table 2.
* Gamma-ray strength function of generalized Lorentzian form
/24/,/25/ was used for E1 transition.
For M1 and E2 transitions the standard Lorentzian form was
adopted. The prameters are shown in Table 3.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Level Scheme of Ag-109
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 1/2 - *
1 0.08803 7/2 +
2 0.13274 9/2 +
3 0.31138 3/2 - *
4 0.41521 5/2 - *
5 0.42000 7/2 +
6 0.70191 3/2 -
7 0.70600 1/2 +
8 0.70700 3/2 -
9 0.72435 3/2 +
10 0.73529 5/2 +
11 0.77350 11/2 +
12 0.78900 7/2 -
13 0.86276 5/2 -
14 0.86947 5/2 +
15 0.87000 5/2 -
16 0.89000 9/2 +
17 0.91100 7/2 +
18 0.91210 7/2 - *
19 0.93080 13/2 +
-------------------
*) Coupled levels in CC calculation
Table 2. Level density parameters
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Ag-110 15.3000 0.0000 3.3762 0.5836 -1.7755 4.1405
Ag-109 15.6000 1.1494 2.9604 0.6222 -1.0590 5.9420
Ag-108 14.6000 0.0000 2.5216 0.6708 -2.2742 5.0724
Ag-107 15.6000 1.1601 1.9728 0.5979 -0.4714 5.2925
Pd-109 16.0000 1.1494 3.8267 0.6463 -1.8210 6.7457
Pd-108 14.3000 2.3094 3.1785 0.6359 0.3436 6.8413
Pd-107 15.0000 1.1601 3.1932 0.6723 -1.5375 6.6188
Pd-106 14.4000 2.3311 2.3412 0.6736 0.1590 7.3089
Rh-108 15.6000 0.0000 4.1884 0.5114 -1.3068 3.3132
Rh-107 13.0075 1.1601 4.0295 0.7116 -1.4170 6.5036
Rh-106 14.2000 0.0000 3.7991 0.5945 -1.6674 4.0000
Rh-105 15.8000 1.1711 3.4219 0.6193 -1.2405 6.1130
Rh-104 14.1000 0.0000 2.9724 0.6799 -2.3482 5.1092
--------------------------------------------------------
Table 3. Gamma-ray strength function for Ag-110
--------------------------------------------------------
* E1: ER = 15.90 (MeV) EG = 6.71 (MeV) SIG = 150.00 (mb)
ER = 6.40 (MeV) EG = 1.80 (MeV) SIG = 1.50 (mb)
* M1: ER = 8.56 (MeV) EG = 4.00 (MeV) SIG = 1.21 (mb)
* E2: ER = 13.15 (MeV) EG = 4.79 (MeV) SIG = 2.50 (mb)
--------------------------------------------------------
References
1) Moxon, M.C., Rae, E.R., "Proc. EANDC Conf. on Time-of-Flight
Methods, Saclay, 1961",439.
2) Asghar, M., et al., "Proc. Int. Conf. on the Study of Nuclear
Structure with Neutrons, Antwerp 1965",(65).
3) Muradjan, G.V., Adamchuk, Ju. V., Jaderno-Fizicheskie
Issledovanija, 6, 64 (1968).
4) de Barros, S., et al., Nucl. Phys., A131, 305(1969).
5) Pattenden, N.J., Jolly, J.E., AERE-PR/NP-16(1969).
6) Macklin, R.L., Nucl. Sci. Eng., 82, 400(1982).
7) Mizumoto, M., et al., J. Nucl. Sci. Technol., 20, 883(1983).
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