96-Cm-248
96-Cm-248 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa+
DIST-MAY10 20100319
----JENDL-4.0 MATERIAL 9649
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-01 Fission cross section was evaluated with GMA code.
06-05 Resolved resonance parameters were modified.
07-03 Fission spectra were evaluated.
07-05 New calculation was made with CCONE code.
08-03 Interpolation of (5,18) was changed.
Recalculation with CCONE code was performed.
Data were compiled as JENDL/AC-2008/1/.
09-08 (MF1,MT458) was evaluated.
10-01 Data of prompt gamma rays due to fission were given.
10-03 Covariance data were given.
MF= 1 General information
MT=452 Number of Neutrons per fission
Sum of MT's=455 and 456.
MT=455 Delayed neutron data
(same as JENDL-3.3)
Semi-empirical formula by Tuttle/2/.
MT=456 Number of prompt neutrons per fission
(same as JENDL-3.3)
At the 0 eV, the experimental data of Zhuravlev et al./3/
was adopted. An energy-dependent term was based on the semi-
empirical formula by Howerton/4/.
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/5/. 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 2000/6/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/7/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/8/ 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: 1.0E-5 - 1500 eV)
Resonance parameters for JENDL-3.3 were evaluated on the basis
of the data of Benjamin et al./9/, Moore and Keyworth/10/
and Maguire et al./11/. Comment for JENDL-3.3 is as follows:
* RESONANCE ENERGIES, NEUTRON AND RADIATIVE WIDTHS WERE
TAKEN FROM THE EXPERIMENTAL DATA OF BENJAMIN ET AL./9/.
FOR RESONANCES WHOSE RADIATIVE WIDTH WAS UNKNOWN, THE
AVERAGE VALUE OF 0.026 EV/9/ WAS ADOPTED. THE FISSION
WIDTHS were adopted from Moore and Keyworth /10/ and
Maguire et al./11/ THE AVERAGE FISSION WIDTH OF 0.0013
EV/10/ WAS USED FOR ALL RESONANCES OF WHICH FISSION
WIDTH HAD NOT BEEN MEASURED. Then the fission widths were
roughly adjusted to the fission cross section measured by
Maguire et al. R=9.1 FM WAS ASSUMED TO REPRODUCE THE
POTENTIAL SCATTERING CROSS SECTION OF 10.4 BARNS ASSUMED
BY BENJAMIN ET AL./9/. THE NEUTRON WIDTH OF THE FIRST
RESONANCE WAS SLIGHTLY ADJUSTED TO REPRODUCE THE CAPTURE
CROSS SECTION OF 2.57 BARNS AT 0.0253 EV. BACKGROUND
CROSS SECTIONS WERE GIVEN ONLY FOR THE FISSION AND TOTAL
CROSS SECTIONS BY ASSUMING THE FORM OF 1/V. THE THERMAL
CROSS SECTIONS TO BE REPRODUCED WERE ESTIMATED FROM
AVAILABLE EXPERIMENTAL DATA.
For the present file, a negative resonance was added at -30 eV
to reproduce the thermal cross sections, and background cross
sections were removed.
The thermal cross sections to be reproduced:
Fission = 0.337 +- 0.032 b
Benjamin et al./12/, Zhuravlev et al./13/,
Serot et al./14/
Capture = 2.87 +- 0.26 b
Druschel et al./15/, Gavrilov et al./16/
Unresolved resonance parameters (1.5 - 200 keV)
Parameters were determined with ASREP code/17/ so as to
reproduce the cross sections. They are used only for self-
shielding calculations.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 10.198
elastic 6.989
fission 0.337 7.84
capture 2.872 267
-------------------------------------------------------
(*) 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/18/.
MT= 1 Total cross section
The cross section was calculated with CC OMP of Soukhovitskii
et al./19/
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 in the energy
range from 1.5 keV to 7 MeV with the GMA code/20/:
Authors Energy range Data points Reference
Moore+ 1.4 keV - 2.8 MeV 413 /10/
Fomushkin+ 0.3 - 5.5 MeV 20 /21/
Maguire+ 1.4 keV - 80 keV 44 /11/
Fomushkin+ 14.1MeV 1 /22/
Fursov+ 510 keV - 6.8 MeV 37 /23/*1)
*1) Ratio to Pu-239 fission, converted to cross section
by using JENDL-3.3 data.
The results of GMA were used to determine the parameters in
the CCONE calculation.
Above 8 MeV, JENDL-3.3 was adopted.
MT=19, 20, 21, 38 Multi-chance fission cross sections
Calculated with CCONE code, and renormalized to the total
fission cross section (MT=18).
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 Fission neutron spectra
Below 6 MeV, calculated by Ohsawa /24/ with modified
Madland-Nix formula considering multi-mode fission processes
(standard-1, standard-2, superlong).
Above 7 MeV, calculated with CCONE code.
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./25/ for
Pu-239 thermal fission were adopted.
MF=31 Covariances of average number of neutrons per fission
MT=452 Number of neutrons per fission
Sum of covariances for MT=455 and MT=456.
MT=455
Error of 15% was assumed.
MT=456
Covariance was obtained by fitting a linear function to the
data at 0.0 and 5.0 MeV with an uncertainty of 4% and 5%,
respevtively. The uncertainty at 0 eV was estimated from the
experimental data of Zhuravlev et al./3/
MF=32 Covariances of resonance parameters
MT=151 Resolved resonance parameterss
Format of LCOMP=0 was adopted.
Uncertainties of parameters were taken from Mughabghab /26/.
For the parameters without any information on uncertainty,
the following uncertainties were assumed:
Resonance energy 0.1 %
Neutron width 10 %
Capture width 20 %
Fission width 10 %
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/27/ and the covariances of model parameters
used in the cross-section calculations.
For the fission cross section, covariances obtained with the
GMA analysis were adopted. Standard deviations (SD) were
multiplied by a factor of 2. SD of 10% were assumed in the
energy region above 8 MeV.
In the resolved resonance region, the following standard
deviations were added to the contributions from resonance
parameters:
Total 10 - 20 %
Elastic scattering 10 - 20 %
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/18/ 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/28/
* Global parametrization of Koning-Duijvestijn/29/
was used.
* Gamma emission channel/30/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/31/ 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/32/. 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/33/,/34/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,2,3,4 (see Table 2)
* optical potential parameters /19/
Volume:
V_0 = 49.97 MeV
lambda_HF = 0.01004 1/MeV
C_viso = 15.9 MeV
A_v = 12.04 MeV
B_v = 81.36 MeV
E_a = 385 MeV
r_v = 1.2568 fm
a_v = 0.633 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 = 1.1803 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 = 0.066
beta_6 = 0.0015
* Calculated strength function
S0= 1.21e-4 S1= 3.23e-4 R'= 9.08 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Cm-248
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 0 + *
1 0.04340 2 + *
2 0.14360 4 + *
3 0.29810 6 + *
4 0.50500 8 + *
5 0.76070 10 +
6 1.04900 2 +
7 1.04900 1 -
8 1.06130 12 +
9 1.08400 0 +
10 1.09400 3 -
11 1.12600 2 +
12 1.14300 4 +
13 1.17200 5 -
14 1.22200 4 +
15 1.23500 3 -
-------------------
*) Coupled levels in CC calculation
Table 3. Level density parameters
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Cm-249 19.0966 0.7605 2.3053 0.3737 -0.8387 2.9359
Cm-248 18.5357 1.5240 2.0504 0.3482 0.2818 3.2555
Cm-247 18.3955 0.7635 1.7794 0.3734 -0.6804 2.7533
Cm-246 18.8984 1.5302 1.7310 0.3608 0.1621 3.4286
Cm-245 18.8322 0.7667 1.4601 0.3623 -0.5771 2.6382
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Cm-249 5.500 0.750 5.100 0.430
Cm-248 6.100 1.040 4.950 0.600
Cm-247 5.400 0.800 5.650 0.650
Cm-246 6.300 1.040 5.100 0.600
Cm-245 6.050 0.500 5.700 0.420
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Cm-249 21.0062 0.8872 2.6000 0.3388 -1.7502 3.0872
Cm-248 20.9336 1.5000 2.6000 0.3323 -1.0300 3.6000
Cm-247 20.8609 0.8908 2.6000 0.3256 -1.5320 2.8908
Cm-246 20.7882 1.6500 2.6000 0.3263 -0.7728 3.6500
Cm-245 20.7155 0.8944 2.6000 0.3342 -1.6357 2.9944
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Cm-249 19.0966 0.8872 0.9000 0.3432 -0.5123 2.4872
Cm-248 20.9336 1.7780 0.8600 0.3706 -0.1043 3.9780
Cm-247 20.8609 0.8908 0.8200 0.3573 -0.8210 2.8908
Cm-246 20.7882 1.7852 0.7800 0.3658 -0.0107 3.8852
Cm-245 20.7155 0.8944 0.7400 0.3596 -0.8163 2.8944
--------------------------------------------------------
Table 7. Gamma-ray strength function for Cm-249
--------------------------------------------------------
* E1: ER = 11.38 (MeV) EG = 2.71 (MeV) SIG = 332.90 (mb)
ER = 14.28 (MeV) EG = 4.18 (MeV) SIG = 431.66 (mb)
* M1: ER = 6.52 (MeV) EG = 4.00 (MeV) SIG = 1.49 (mb)
* E2: ER = 10.01 (MeV) EG = 3.12 (MeV) SIG = 7.06 (mb)
--------------------------------------------------------
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