92-U -232 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al.
DIST-MAY10 20100323
----JENDL-4.0 MATERIAL 9219
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-01 Fission cross section was modified.
06-10 Nu-p was modified.
07-05 New calculation was performed with CCONE code.
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=455 and 456.
MT=455 Delayed neutron data
Nu-d was determined from Tuttle's systematics/2/. Six group
decay constants were taken from the paper of Brady and
England/3/.
MT=456 Number of prompt neutrons per fission
Nu-p measured by Jaffey and Lerner/4/ at the thermal neutron
energy was adopted. An energy dependent term was determined
from the systematics of Ohsawa /5/.
Nu-p of Cf-252 SF = 3.756+-0.031 /6/ was used.
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/7/. 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 2000/8/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/9/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/10/ 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 - 200 eV)
For JENDL-3.3, recommendation by Mughabghab /11/ was adopted,
and its formula was changed from Reich-Moore to Multilevel
Breit-Wigner type. Background cross section was given to
reproduce measured fission cross sections/12,13/ at valleys
of resonance levels.
For the present file, the capture and fission widths of a
negative resonance were modified so as to reproduce thermal
cross sections.
The thermal cross sections to be reproduced:
Fission = 76.5 +- 4.1 b
Elson et al./14/, Cabell et al./15/, Gryntakis/16/
Capture = 75.4 +- 1.5 b
Halperin et al./17/, Cabell et al./15/
Unresolved resonance parameters (200 eV- 40 keV)
Parameters were determined with ASREP code/18/ so as to
repruduce the following cross sections:
Total = sum of partial cross sections
Elastic = calculated with CCONE code/19/
Fission = results of GMA analysis
Capture = calculated with CCONE code
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 162.72
elastic 10.81
fission 76.52 364
capture 75.39 173
-------------------------------------------------------
(*) 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 total (MT=1), elastic scattering (MT=2) and fission cross
sections (MT=18, 19, 20, 21, 38) were calculated with CCONE
code/19/.
MT=1 Total cross section
From 200 eV to 40 keV: sum of partial cross sections.
Above 40 keV: the cross section was calculated with CCONE
code using CC OMP of Soukhovitskii et al./20/
MT=2 Elastic scattering cross section
From 200 eV to 40 keV: calculated with CCONE code.
Above 40 keV: total - nonelastic scattering cross sections
MT=18 Fission cross section
Below 200 eV, background cross section was given.
Above 200 eV, the following experimental data were analyzed
with the GMA code /21/:
Authors Energy range Data points Reference
Auchampaugh+ 0.15 - 1.93 keV 265 /12/
Farrell 0.15 - 21.3 keV 1173 /13/
Fursov+ 0.135 - 14.7 MeV 77 /22/(*1)
(*1) Relative to Pu-239 fission. Data were converted
to cross sections using JENDL-3.3 data.
In the energy range where experimental data were scarce,
cross-section curve was determined by eye-guiding.
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).
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 neutron spectra
Calculated with CCONE code.
MT=455 Delayed neutron spectra
Results of summation calculation made by Brady and England /3/
were adopted.
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./23/ for
U-235 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 of 15% was assumed below 4 MeV and 4 - 7 MeV, and above
7 MeV, respectively.
MT=456
Covariance was obtained by fitting a linear function to the
data at thermal energy and 5 MeV assuming errors of 2% and
5%, respectively.
MF=32 Covariances of resonance parameters
Format of LCOMP=0 was adopted.
Standard deviations were adopted from the data of Mughabghab
/11/.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/24/ and the covariances of model parameters
used in the theoretical calculations.
Covariances of the total, elastic-scattering and capture cross
sections were determined by considering the experimental data
(see MF=3).
For the following cross sections, covariances were further
modified.
MT=1,2 Total and elastic scattering cross sections
In the resonance region (up to 200 eV), uncertainty of 10 %
was added.
MT=18 Fission cross section
In the resonance region, error of 5% was added to the
contributions from uncertainties of resonance parameters.
Above the resonance region, cross section was evaluated with
GMA code/21/. Standard deviations of 20% were added in the
energy region from 200 eV to 20 keV. Above 8 MeV, they were
assumed to be 10%.
MT=102 Capture cross section
In the resonance region, addtional error of 2 % was given.
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
Estimated with CCONE and KALMAN codes.
*****************************************************************
Calculation with CCONE code
*****************************************************************
Models and parameters used in the CCONE/19/ 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/25/
* Global parametrization of Koning-Duijvestijn/26/
was used.
* Gamma emission channel/27/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/28/ 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/29/. 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/30/,/31/
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 /20/
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= 1.05e-4 S1= 2.49e-4 R'= 9.65 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of U-232
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 0 + *
1 0.04757 2 + *
2 0.15657 4 + *
3 0.32260 6 + *
4 0.54100 8 + *
5 0.56319 1 -
6 0.62897 3 -
7 0.69121 0 +
8 0.73456 2 +
9 0.74690 5 -
10 0.80580 10 +
11 0.83307 4 +
12 0.86679 2 +
13 0.91142 3 +
14 0.91510 7 -
15 0.97068 4 +
16 0.98480 6 +
17 1.01685 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
--------------------------------------------------------
U-233 18.3972 0.7861 2.4694 0.3819 -0.8199 2.9895
U-232 18.3293 1.5757 2.6095 0.3887 -0.1141 3.8805
U-231 18.2614 0.7895 2.6793 0.4123 -1.1756 3.4264
U-230 18.1935 1.5825 2.6739 0.3937 -0.1508 3.9419
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
U-233 5.970 0.800 5.450 0.520
U-232 5.800 1.040 5.000 0.600
U-231 6.000 0.800 5.600 0.520
U-230 3.800 1.040 3.900 0.600
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-233 20.2008 0.9172 2.6000 0.3312 -1.4942 2.9172
U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383
U-231 20.0518 0.9211 2.6000 0.3325 -1.4903 2.9211
U-230 19.9772 1.8463 2.6000 0.3332 -0.5651 3.8463
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-233 20.2008 0.9172 0.0200 0.3576 -0.6166 2.7172
U-232 20.1263 1.8383 -0.0200 0.3745 0.1401 3.8383
U-231 20.0518 0.9211 -0.0600 0.3758 -0.7765 2.9211
U-230 19.9772 1.8463 -0.1000 0.3771 0.1493 3.8463
--------------------------------------------------------
Table 7. Gamma-ray strength function for U-233
--------------------------------------------------------
K0 = 1.501 E0 = 4.500 (MeV)
* E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.00 (mb)
ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)
* M1: ER = 6.66 (MeV) EG = 4.00 (MeV) SIG = 2.69 (mb)
* E2: ER = 10.24 (MeV) EG = 3.31 (MeV) SIG = 6.53 (mb)
--------------------------------------------------------
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