95-Am-243
95-Am-243 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa,+
DIST-MAY10 20100318
----JENDL-4.0 MATERIAL 9549
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
07-05 Theoretical calculation was made with CCONE code.
07-08 Theoretical calculation was made with CCONE code.
08-03 Interpolation of (5,18) was changed.
Data were compiled as JENDL/AC-2008/1/
09-03 (1,452) and (1,455) were revised.
09-08 (MF1,MT458) was evaluated.
09-10 Fission cross section was revised.
10-01 Data of prompt gamma rays due to fission were given.
10-03 Covariance data were given.
MF= 1
MT=452 Total neutron per fission
Sum of MT=455 and 456.
MT=455 Delayed neutrons
Determined from nu-d of the following three nuclides and
partial fission cross sections calculated with CCONE code/2/.
Am-244 = 0.0085 Saleh et al./3/, Charlton et al./4/
Am-243 = 0.006659 *1)
Am-242 = 0.0049 Saleh et al./3/
*1) an average of systematics by Tuttle/5/,
Benedetti et al./6/ and Waldo et al./7/
Decay constants were taken from Saleh et al. and Brady and
England.
MT=456 Prompt neutrons per fission
The data measured by Khokhlov et al./8/ and Drapchinsky et
al./9/ were fitted by a linear function/10/.
MT=458 Components of energy release due to fission
Total energy and prompt energy were calculated from mass
balance using JENDL-4 fission yields data and mass excess
evaluation. Mass excess values were from Audi's 2009
evaluation/11/. Delayed energy values were calculated from
the energy release for infinite irradiation using JENDL FP
Decay Data File 2000 and JENDL-4 yields data. For delayed
neutron energy, as the JENDL FP Decay Data File 2000/12/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/13/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/14/ when
they were provided. When the fractions were not given by Sher,
averaged fractions were used.
MF= 2 Resonance parameters
MT=151
Resolved resonance parameters (below 250eV)
JENDL-3.3 parameters of resonances below 1.744 eV were
modified:
to reproduce an effective capture cross section measured
by Ohta et al./15/,
to delete background cross sections of fission given in
JENDL-3.3.
Unresolved resonance parameters (250eV - 40keV)
Parameters were determined with ASREP code/16/ so as to
reproduce the cross sections in the energy range from 250 eV
to 40 keV. They are used only for self-shielding calculations.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 85.831
elastic 6.490
fission 0.0816 6.31
capture 79.259 2040
-------------------------------------------------------
(*) In the energy range from 0.5 eV to 10 MeV.
MF= 3 Neutron cross sections
Cross sections above the resolved resonance region except for
the elastic scattering (MT=2) and fission cross sections (MT=18,
19, 20, 21, 38) were calculated with CCONE code/2/.
MT= 1 Total cross section
The cross section was calculated with CC OMP of Soukhovitskii
et al./17/ The OMP was adjusted to the Am-241(n,tot) cross
section/18/.
MT= 2 Elastic scattering cross section
Calculated as total - non-elastic scattering cross sections.
MT=18 Fission cross section (Above 250eV)
The following experimental data were analyzed with the GMA
code /19/:
Wisshak+/20/, Fomushkin+/21/, Fursov+/22/, Kanda+/23/,
Knitter+/24/, Golovnya+/25/, Kobayashi+/26/, Laptev+/27/,
Aiche+/28/, Baba+/29/
MT=19, 20, 21, 38 Multi-chance fission cross sections
Calculated with CCONE code, and renormalized to the total
fission cross section (MT=18).
MT=102 Capture cross section
The experimental data of Weston and Todd/30/ and Wisshak and
Kaeppeler/31/ were used to determine the parameters in the
CCONE calculation.
MF= 4 Angular distributions of secondary neutrons
MT=2 Elastic scattering
Calculated with CCONE code.
MT=18 Fission
Isotropic distributions in the laboratory system were assumed.
MF= 5 Energy distributions of secondary neutrons
MT=18 Prompt neutrons
Below 6 MeV, calculated by Ohsawa/32/ with modified
Madland-Nix formula considering multi-mode fission processes
(standard-1, standard-2, superlong).
Above 7 MeV, calculated with CCONE code/2/.
MT=455 Delayed neutrons
Taken from Brady and England /33/. Normalized yields of
6 groups are those measured by Saleh et al./3/
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF=12 Photon production multiplicities
MT=18 Fission
Calculated from the total energy released by the prompt
gamma-rays due to fission given in MF=1/MT=458 and the
average energy of gamma-rays.
MF=14 Photon angular distributions
MT=18 Fission
Isotoropic distributions were assumed.
MF=15 Continuous photon energy spectra
MT=18 Fission
Experimental data measured by Verbinski et al./34/ for
Pu-239 thermal fission were adopted.
MF=31 Covariances of average number of neutrons per fission
MT=452 Number of neutrons per fission
Combination of covariances for MT=455 and MT=456.
MT=455
Error was assumed as follows:
E < 5 MeV 5 % /3/
above 15 MeV 15 %
MT=456
Covariances were obtained by fitting a linear function to the
experimental data of Khokhlov et al./8/ and Drapchinsky et
al./9/ Obtained standard deviation was multiplied by a
factor of 3 so that the minimum error was about 1%.
MF=32 Covariances of resonance parameters
Format of LCOMP=0 was adopted.
Standard deviations of resonance parameters were taken from
Mughabghab /35/ If no uncertainties were given by
Mughabghab, 0.1% and 10% were assumed for resonance energies
and other parameters, respectively.
Additional errors of 5 % were given to the fission cross
section, and the following values for the capture cross
section:
energy range additional errors
1.0e-5 - 13 eV 5%
13 - 50 eV 10%
50 - 250 eV 15%
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/36/ and the covariances of model parameters
used in the theoretical calculations.
For the following cross sections, covariances were determined
by different methods.
MT=1, 2 Total and elastic scattering cross sections
In the resolved resonance region, uncertainty of 10% was
added to the contributions from resonance parameter
uncertainties.
Above 250 eV, estimated with CCONE and KALMAN codes.
MT=18 Fission cross section
Above 250 eV, evaluated with GMA code/19/.
Standard deviations were multiplied by a factor of 2.0.
MT=102 Capture cross section
Above 250 eV, covariance matrix was obtained with CCONE and
KALMAN codes/36/.
MF=34 Covariances for Angular Distributions
MT=2 Elastic scattering
Covariances were given only to P1 components.
MF=35 Covariances for Energy Distributions
MT=18 Fission spectra
Below 6 MeV, covarinaces of Pu239 fission spectra given in
JENDL-3.3 were adopted after multiplying a factor of 9.
Above 6 MeV, estimated with CCONE and KALMAN codes.
*****************************************************************
Calculation with CCONE code
*****************************************************************
Models and parameters used in the CCONE/2/ calculation
1) Coupled channel optical model
Levels in the rotational band were included. Optical model
potential and coupled levels are shown in Table 1.
2) Two-component exciton model/37/
* Global parametrization of Koning-Duijvestijn/38/
was used.
* Gamma emission channel/39/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/40/ was included.
* Neutron, gamma and fission decay channel were included.
* Transmission coefficients of neutrons were taken from
coupled channel calculation in Table 1.
* The level scheme of the target is shown in Table 2.
* Level density formula of constant temperature and Fermi-gas
model were used with shell energy correction and collective
enhancement factor. Parameters are shown in Table 3.
* Fission channel:
Double humped fission barriers were assumed.
Fission barrier penetrabilities were calculated with
Hill-Wheler formula/41/. Fission barrier parameters were
shown in Table 4. Transition state model was used and
continuum levels are assumed above the saddles. The level
density parameters for inner and outer saddles are shown in
Tables 5 and 6, respectively.
* Gamma-ray strength function of Kopecky et al/42/,/43/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,3,6,8 (see Table 2)
* optical potential parameters /17/
Volume:
V_0 = 48 MeV
lambda_HF = 0.004 1/MeV
C_viso = 15.9 MeV
A_v = 12.04 MeV
B_v = 81.36 MeV
E_a = 385 MeV
r_v = 1.255 fm
a_v = 0.58 fm
Surface:
W_0 = 17.2 MeV
B_s = 11.19 MeV
C_s = 0.01361 1/MeV
C_wiso = 23.5 MeV
r_s = 1.15 fm
a_s = 0.601 fm
Spin-orbit:
V_so = 5.75 MeV
lambda_so = 0.005 1/MeV
W_so = -3.1 MeV
B_so = 160 MeV
r_so = 1.1214 fm
a_so = 0.59 fm
Coulomb:
C_coul = 1.3
r_c = 1.2452 fm
a_c = 0.545 fm
Deformation:
beta_2 = 0.213
beta_4 = 0.08
beta_6 = 0.0015
* Calculated strength function
S0= 0.91e-4 S1= 2.65e-4 R'= 9.45 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Am-243
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 5/2 - *
1 0.04220 7/2 - *
2 0.08400 5/2 +
3 0.09640 9/2 - *
4 0.10920 7/2 +
5 0.14350 9/2 +
6 0.16230 11/2 - *
7 0.18930 11/2 +
8 0.23800 13/2 - *
9 0.24400 13/2 +
10 0.26600 3/2 -
-------------------
*) Coupled levels in CC calculation
Table 3. Level density parameters
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Am-244 18.7661 0.0000 1.9765 0.2797 -0.6834 1.0000
Am-243 18.6999 0.7698 2.0985 0.3873 -0.8965 3.0385
Am-242 18.6337 0.0000 1.6845 0.2795 -0.6541 0.9592
Am-241 18.1961 0.7730 1.7328 0.3819 -0.7226 2.8365
Am-240 18.5012 0.0000 1.3474 0.2883 -0.6831 1.0000
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Am-244 6.450 0.650 6.000 0.530
Am-243 6.200 0.800 5.400 0.520
Am-242 6.510 0.600 6.050 0.550
Am-241 6.100 0.800 5.500 0.520
Am-240 6.100 0.650 6.000 0.450
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Am-244 21.5810 0.0000 2.6000 0.3194 -2.3947 2.0000
Am-243 21.5049 0.8981 2.6000 0.3201 -1.4966 2.8981
Am-242 20.8697 0.0000 2.6000 0.3254 -2.4113 2.0000
Am-241 21.3526 0.9018 2.6000 0.3213 -1.4929 2.9018
Am-240 21.2764 0.0000 2.6000 0.3219 -2.3947 2.0000
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Am-244 21.5810 0.0000 0.6400 0.3664 -1.8698 2.2000
Am-243 21.5049 0.8981 0.6000 0.3532 -0.8019 2.8981
Am-242 21.4288 0.0000 0.5600 0.3688 -1.8681 2.2000
Am-241 21.3526 0.9018 0.5200 0.3556 -0.7969 2.9018
Am-240 21.2764 0.0000 0.4800 0.3567 -1.6981 2.0000
--------------------------------------------------------
Table 7. Gamma-ray strength function for Am-244
--------------------------------------------------------
* E1: ER = 11.51 (MeV) EG = 2.76 (MeV) SIG = 245.81 (mb)
ER = 14.29 (MeV) EG = 4.18 (MeV) SIG = 491.62 (mb)
* M1: ER = 6.56 (MeV) EG = 4.00 (MeV) SIG = 1.29 (mb)
* E2: ER = 10.08 (MeV) EG = 3.18 (MeV) SIG = 6.92 (mb)
--------------------------------------------------------
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