95-Am-241
95-Am-241 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa,+
DIST-JUL13 20130626
----JENDL-4.0u1 MATERIAL 9543
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
07-05 New theoretical calculation was made with CCONE code.
07-09 Isomeric ratio was revised.
07-11 Isomeric ratio was revised.
08-03 Interpolation of (5,18) was changed.
Data were compiled as JENDL/AC-2008/1/.
09-03 (MF2,MT151) was revised.
09-08 (MF1,MT458) was evaluated.
09-12 (1,455) and (2,151) were revised.
10-01 Data of prompt gamma rays due to fission were given.
10-03 Covariance data were given.
13-06 QI for an isomer in (MF9,MT102) was corrected.
(MF8,MT102) MATP was removed.
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-242 = 0.0049 Saleh et al./3/
Am-241 = 0.0016 *1)
Am-240 = 0.0011 *1)
*) A half of systematics by Tuttle/4/,
Benedetti et al./5/ and Waldo et al./6/
Decay constants were adopted from Saleh et al./3/ and Brady
and England/7/.
MT=456 Prompt neutrons per fission
The data measured by Jaffey et al./8/, Khokhlov et al./9/
and Drapchinsky et al./10/ were fitted by a linear
function/11/.
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/12/. 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 2000/13/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/14/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/15/ 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 (MLBW, below 150 eV)
Resonance parameters adopted in JENDL-3.3 were revised mainly
in a low energy region. The positive resonances below 12 eV
were modified on the basis of new parameters reported by
Jandel el al./16/. The neutron widths were multiplied by
a factor of 1.029. The parameters below 1.27-eV resonance
were further adjusted to the fission cross section of Dabbs
et al./17/ Parameters of negative resonances were adjusted
so as to reproduce the following thermal cross sections:
Capture = 684 +- 15 b
Kalebin et al./18/, Shinohara et al./19/, Fioni et
al./20/, Bringer et al./21/, Jandel et al./16/
Capture to ground state = 620.1 +- 7.8 b
Harbour et al./22/, Gavrilov et al./23/, Shinohara
et al./19/, Maidana et al./24/, Fioni et al./20/,
Nakamura et al./25/
Capture to meta stable st= 64.8 +- 3.6 b
Shinohara et al./19/, Fioni et al./20/ and others
Fission = 3.12 +- 0.06 b
Zhuravlev et al./26/, Dabbs et al./17/, Yamamoto
et al./27/, and others.
The present resolved resonance parameters give the total
cross section larger than experimental data measured by
Adamchuk et al./28/ and Kalebin et al./18/
Unresolved resonance parameters
Parameters were determined with ASREP code/29/ so as to
reproduce the cross sections in the energy range from 150 eV
to 40 keV. They are used only for self-shielding calculations.
Thermal cross sections and resonance integrals (300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
-------------------------------------------------------
total 699.22
elastic 11.82
fission 3.122 13.3
capture 684.28 1590
-------------------------------------------------------
(*) 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./30/ The parameters were adjusted using the experimental
data of Phillips and Howe/31/.
MT= 2 Elastic scattering cross section
Calculated as total - non-elastic scattering cross sections.
MT=18 Fission cross section
The following experimental data were analyzed with the GMA
code /32/:
Cance+/33/, Hage+/34/, Aleksandrov+/35/, Dabbs+/17/,
Vorotnikov+/36/, Yamamoto+/27/, Golovnya/37/, Baba+/38/.
The results of GMA were used to determine the parameters in
the CCONE calculation.
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
Calculated with CCONE code. The experimental data of Wisshak
and Kaeppeler/39/, Vanpraet et al./40/, Gayther and Thomas/41/
and Ivanova et al./42/ 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 with modified Madland-Nix formula
considering multi-mode fission processes (standard-1,
standard-2, superlong) by Ohsawa/43/.
Above 7 MeV, calculated with CCONE code/2/.
MT=455 Delayed neutrons
Calculated by Brady and England/7/. Fractions of 6
temporal groups were adopted from Saleh et al./3/.
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF= 8 Radioactive nuclide production
MT=102 Capture cross section
(same as JENDL-3.3)
Decay data were taken from ENSDF.
MF= 9 Multiplicities for production of radioactive elements
MT=102 Capture cross section
Isomeric ratio (IR) was calculated with CCONE code above 100
eV. IR to ground state (IR-g) was normalized to 0.84 at 300
keV/44/.
IR-g below 0.1 eV was based on experimental data of Harbour
et al./22/, Gavlilov et al./23/, Wisshak et al./45/, Shinohara
et al./19/, Fioni et al./20/, Bringer et al./46/. Average
IR-g = 0.896 +-0.002.
Above 0.1 eV, the data were connected straighly to 0.859 at
1 eV, and the CCONE calculations above 100 eV in the log-
linear scale.
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./47/ 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
Standard deviation was roughly estimated as 5% below 5 MeV,
15% above 5 MeV.
MT=456
Covariance was obtained by fitting to the experimental
data (see MF1,MT456).
MF=32 Covariances of resonance parameters
Format of LCOMP=0 was adopted.
Standard diviations of resonance parameters were taken from
Ref./16/ below 12 eV, and from the JENDL-3.3 covariance
file /11/ above 12 eV. They were based on Derrien and
Lucas /48/ and Mughabghab /49/. For the levels whose
information are not given in those references, assumed were
standard deviations of 0.1 % for resonance energies, and 10
% for neutron, capture and fission widths.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/50/ 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 15% was
added to the contributions from resonance parameter
uncertainties.
Above 150 eV, estimated with CCONE and KALMAN codes.
MT=18 Fission cross section
In the resolved resonance region, uncertainty of 5% was added
to the contributions from resonance parameter uncertainties.
Above 150 eV, the fission cross section was evaluated with
GMA code/32/. The following errors were added to the GMA
results:
150 - 5000 eV 10 %
5 - 50 keV 5 %
50 - 500 keV 3 %
0.5 - 5 MeV 1 %
MT=102 Capture cross section
In the resolved resonance region, uncertainty of 5% was added
to the contributions from resonance parameter uncertainties.
In the energy range below 150 eV, the following addtional
errors were added:
0.1 - 5 eV 10 %
5 - 150 eV 15 %
Above 150 eV, covariance matrix was obtained with CCONE and
KALMAN codes/50/.
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/51/
* Global parametrization of Koning-Duijvestijn/52/
was used.
* Gamma emission channel/53/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/54/ 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/55/. 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/56/,/57/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,2,3,5,11 (see Table 2)
* optical potential parameters /30/
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= 0.99e-4 S1= 2.66e-4 R'= 9.49 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Am-241
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 5/2 - *
1 0.04118 7/2 - *
2 0.09370 9/2 - *
3 0.15750 11/2 - *
4 0.20588 5/2 +
5 0.23368 13/2 - *
6 0.23520 7/2 +
7 0.23900 7/2 +
8 0.27300 13/2 -
9 0.27320 9/2 +
10 0.31980 11/2 +
11 0.31982 15/2 - *
12 0.38110 13/2 +
13 0.41818 17/2 -
14 0.45310 15/2 +
15 0.45900 11/2 -
16 0.47181 3/2 -
17 0.49500 21/2 +
18 0.50445 5/2 -
19 0.52567 19/2 -
20 0.53090 17/2 +
-------------------
*) 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-242 18.6337 0.0000 1.6845 0.2832 -0.6794 0.9948
Am-241 18.1961 0.7730 1.7328 0.3819 -0.7226 2.8365
Am-240 17.7611 0.0000 1.3474 0.2960 -0.6883 1.0000
Am-239 18.4349 0.7762 1.5592 0.3648 -0.5528 2.6354
Am-238 18.3685 0.0000 1.3698 0.2894 -0.6819 1.0000
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Am-242 6.510 0.600 6.050 0.550
Am-241 6.100 0.800 5.500 0.520
Am-240 6.000 0.650 5.600 0.450
Am-239 6.000 0.800 5.400 0.520
Am-238 5.700 0.650 2.100 0.450
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
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
Am-239 21.2001 0.9056 2.6000 0.3225 -1.4891 2.9056
Am-238 21.1238 0.0000 2.6000 0.3231 -2.3946 2.0000
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Am-242 21.4288 0.0000 0.5600 0.3688 -1.8681 2.2000
Am-241 21.3526 0.9018 0.5200 0.3405 -0.6297 2.7018
Am-240 21.2764 0.0000 0.4800 0.3416 -1.5310 1.8000
Am-239 21.2001 0.9056 0.4400 0.3428 -0.6250 2.7056
Am-238 21.1238 0.0000 0.4000 0.3440 -1.5301 1.8000
--------------------------------------------------------
Table 7. Gamma-ray strength function for Am-242
--------------------------------------------------------
* E1: ER = 11.53 (MeV) EG = 2.77 (MeV) SIG = 243.63 (mb)
ER = 14.32 (MeV) EG = 4.20 (MeV) SIG = 487.26 (mb)
* M1: ER = 6.58 (MeV) EG = 4.00 (MeV) SIG = 1.27 (mb)
* E2: ER = 10.11 (MeV) EG = 3.21 (MeV) SIG = 6.93 (mb)
--------------------------------------------------------
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