92-U -238
92-U -238 JAEA+ EVAL-NOV09 O.Iwamoto,N.Otuka,S.Chiba,+
DIST-MAY10 20100311
----JENDL-4.0 MATERIAL 9237
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-07 (n,2n) cross section was revised.
06-10 nu-p was revised.
07-04 Calculation with CCONE code was performed.
07-06 Fission spectra above 5.5 MeV were revised.
07-11 Fission cross section was revised with simultaneous
evaluation.
07-12 Fission cross section was revised with results of new
simultaneous evaluation. Unresolved resonance parameters
were revised.
08-01 Fission cross section was revised.
08-03 Fission and capture cross sections, and nu-p were revised.
CCONE calculation was made with revised parameters.
Interpolation of (5,18) was changed.
Data were compiled as JENDL/AC-2008/1/.
09-08 (MF1,MT458) was evaluated.
09-10 fission cross section was modified slightly.
09-11 Results of new CCONE calculation were adopted.
10-03 Covariance data were given.
MF= 1
MT=452 Total number of neutrons per fission
Sum of MT=455 and 456.
MT=455 Delayed neutron data
(same as JENDL-3.3/2/)
Experimental data of Krick and Evans /3/ were renormalized
to those of Meadows /4/, and the least-squares fitting was
carried out with the SOK code /5/.
Decay constants were adopted from Keepin et al. /6/
MT=456 Number of prompt neutrons per fission
Experimental data reported after 1960 were considered:
Butler et al./7/, Conde et al./8/, Asplund-Nilsson/9/,
Mather et al./10/, Vorob'jova et al./11/, Bao/12/,
Nurpeisov et al./13/, Frehaut et al./14,15/, Malinovskij et
al./16/, Boykov et al./17/, Smirenkin et al./18/.
Cf-252 nu-p of 3.756 /19/ was used. These experimental data
were fitted with two straight lines below and above 14 MeV.
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/20/. 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 2000/21/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/22/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/23/ 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 (RM: 1.0E-5 - 20 keV)
The data of ENDF/B-VII/24/ were adopted. The parameters were
analyzed by Derrien et al./25/ with SAMMY code /26/ and
experimental data up to 20 keV/27,28,29,30,31,etc./
The thermal capture cross section was adjusted to 2.683 b
evaluated by Trkov et al./32/
Unresolved resonance parameters (20 keV - 150 keV)
The parameters are used only for calculation of self-
shielding factors.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 11.983
elastic 9.300
fission 1.68E-5 1.24
capture 2.683 276
-------------------------------------------------------
(*) 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), fission cross (MT=18, 19, 20,
21, 38) and capture cross sections were calculated with CCONE
code/33/.
Model parameters for CCONE code were determined by considering
experimental data of total and (n,2n) cross sections, and
fission and capture cross sections of JENDL-3.3. OMPs were
based on those of Kunieda et al./34/, and adjusted to the total
cross section. The model parameters were further adjusted by
considering integral data.
MT= 1 Total cross section
Below about 12 MeV, cross section was calculated with CCONE
code. Above 12 MeV, a smooth cross-section curve was obtained
by spline fitting to the experimental data of Abfalterer et
al./35/
MT=2 Elastic scattering cross section
Calculated as total - non-elastic scattering cross sections
except for the 1.3-4.0MeV region where the differences among
adopted fission and calculated fission were reflected in the
inelastic scattering cross sections.
MT=16 (n,2n) cross section
Calculated with CCONE code. Following experimental data were
considered for the determination of model parameters:
Frehaut et al./36,37/, Kornilov et al./38/, Karius
et al./39/, Raics et al./40/, Golovnya et al./41/,
Konno et al./42/, Filatenkov et al./43/, Veeser et al./44/,
Ryves et al./45/, Pepelnik et al./46/.
The data of Frehaut et al. were multiplied by a factor of 1.1.
The data of activation measurements were re-normalized by
adopting a intensity of 21.2% to the 208-keV gamma-rays from
Np-237.
MT=18 Fission cross section
Below 400 keV, JENDL-3.3/2/ was adopted.
Above 400 keV, experimental data measured after 1960 were
analyzed by simultaneous fitting of U-233, U-235, U-238,
Pu-239, Pu-240 and Pu-241 fission cross sections and their
ratio by the SOK code/5/. Covariance matrix reported in
Manabe et al./47/ was also considered in the analysis.
--------------------------------------------------------------
Cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
13586.011 1.00E+7 J.W.Meadows+ /48/
22304.003 4.80E+6 - 1.88E+7 K.Merla+ /49/
30669.002 4.00E+6 - 5.50E+6 J.X.Wu+ /50/
20779.003 1.39E+7 - 1.46E+7 M.Cance+ /51/
40547.007 1.48E+7 V.M.Adamov+ /52/
40483.002 1.60E+5 - 1.55E+6 P.E.Vorotnikov+ /53/
40081.002 2.50E+6 I.M.Kuks+ /54/
21209.002 1.27E+7 - 1.94E+7 B.Adams+ /55/
22565.002 1.45E+7 G,Winkler+ /56/
--------------------------------------------------------------
Ratio to U-233(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
10422.006 1.00E+6 - 2.85E+7 J.W.Behrens+ /57/
--------------------------------------------------------------
Ratio to U-235(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
10635.002 1.51E+5 - 2.39E+7 F.C.Difilippo+ /58/
41455.003 5.77E+5 - 2.94E+7 O.A.Shcherbakov+ /59/
30722.002 1.47E+7 J.W.Li+ /60/
13134.007 1.47E+7 J.W.Meadows /61/
40831.004 1.38E+7 - 1.48E+7 A.A.Goverdovskij+ /62/
40831.003 5.44E+6 - 1.04E+7 A.A.Goverdovskij+ /62/
30588.002 1.35E+7 - 1.48E+7 M.Varnagy+ /63/
40506.002 9.81E+5 - 7.00E+6 B.I.Fursov+ /64/
10653.004 1.44E+5 - 2.92E+7 J.W.Behrens+ /65/
20870.002 2.65E+6 - 7.01E+6 M.Cance+ /66/
20869.002 4.67E+6 - 8.85E+6 C.Nordborg+ /67/
20409.002 1.37E+6 - 2.96E+7 S.Cierjacks+ /68/
10506.002 5.33E+6 - 1.04E+7 J.W.Meadows /69/
10504.002 1.09E+6 - 3.03E+6 J.W.Meadows /70/
10237.003 8.99E+5 - 5.15E+6 J.W.Meadows /71/
10232.006 2.00E+6 - 3.00E+6 W.P.Poenitz+ /72/
22282.006 1.35E+7 - 1.49E+7 F.Manabe+ /47/
-----.--- 8.33E+5 - 2.96E+7 P.W.Lisowski+ /73/
--------------------------------------------------------------
The cross section was slightly modified in the energy region
from 1 to 4 MeV and from 7 to 8 MeV.
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
Based on JENDL-3.3/2/ data which were evaluated as follows:
Below 300 keV, evaluation was mainly based on the data
measured by Kazakov et al./74/. Above 300 keV, data were
taken from JENDL-2 which were determined mainly from the
measurements by Poenitz/75/, Panitkin and Sherman/76/,
Moxon/77/, Fricke et al./78/ and Menlove and Poenitz/79/.
Above 1 MeV, statistical model calculation was made, and
direct and semi-direct capture cross section was calculated
with DSD code/80/.
They were slightly modified by considering integral data.
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 fission neutrons
Below 5 MeV, data of JENDL-3.3/2/ were adopted.
Comment of JENDL-3.3:
* Distributions were calculated with the modified Madland-Nix
model/81,82/. The compound nucleus formation cross
sections for fission fragments (FF) were calculated using
Bechetti-Greenlees potential/83/. Up to 3rd-chance-fission
were considered at high incident neutron energies.
Parameters adopted:
Total average FF kinetic energy = 167.41 - 172.65 MeV
Average energy release = 186.115 - 186.364 MeV
Average mass number of light FF = 99 - 111
Average mass number of heavy FF = 128 - 140
Level density of the light FF = 10.106 - 10.963
Level density of the heavy FF = 11.441 - 7.811
Ratio of nuclear temperature
for light to heavy FF = 1.0
Note that the parameters vary with the incident energy
within the indicated range.
Above 5.5 MeV, the spectra calculated with CCONE code /33/
were adopted.
MT=455 Delayed neutrons
(same as JENDL-3.3)
Taken from Brady and England/84/. Group abundace parameters
were adjusted so as to reproduce total delayed neutron
emission rate measured by Keepin/6/, and East et al./85/.
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF=12 Photon Production Multiplicities (option 1)
MT=18 Fission
(same as JENDL-3.3)
The thermal neutron-induced fission gamma spectrum of U-235
measured by Verbinski et al./86/ was adopted for the whole
energy region. The intensity of photon below 0.14 MeV, where
no data were given, was assumed to be the same as that
between 0.14 and 0.3 MeV.
Data were extended up to 20 MeV for JENDL-4.0.
MF=14 Angular Distributions of Photons
Isotropic distributions were assumed for all sections.
MF=15 Continuous Photon Energy Spectra
MT=18 Fission
(same as JENDL-3.3)
U-235 spectra measured by Verbinski et al./86/.
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
Same as JENDL-3.3/2/.
MT=456
Covariance was obtained by fitting to the experimental
data (see MF1,MT456).
MF=32 Covariances of resonance parameters
Format of LCOMP=2 was adopted.
The covariance matrix was given to the resonance parameters
up to 2 keV, and the energy range was set to 1.0e-5 eV to
1.5 keV. The covariance matrix of resonance parameters was
taken from ORNL evaluation /87,88/.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/89/ 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 energy range from 1.5 to 20 keV, uncertainty of 5% was
assumed. Above 20 keV, covariance of the CCONE calculation
was adopted.
MT=18 Fission cross section
In the energy range below 400 keV, uncertainty of 80% was
assumed.
Above 400 keV, covariance matrix was obtained by simultaneous
evaluation among the fission cross sections of U-233, U-235,
U-238, Pu-239, Pu-240 and Pu-241(See MF=3, MT=18, and /1/).
Since the variances are very small, they were adopted by
multiplying a factor of 2.
MT=102 Capture cross section
In the energy range from 1.5 to 20 keV, uncertainty of 10%
was assumed. Above 20 keV, the covariance matrix was taken
from JENDL-3.3.
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 5 MeV, based on the covarinaces given in JENDL-3.3.
Above 5 MeV, estimated with CCONE and KALMAN codes.
*****************************************************************
Calculation with CCONE code
*****************************************************************
Models and parameters used in the CCONE/33/ 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/90/
* Global parametrization of Koning-Duijvestijn/91/
was used.
* Gamma emission channel/92/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/93/ 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/94/. 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/95/,/96/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,2,3,4,7 (see Table 2)
* optical potential parameters /97/
Real:
VRO = -42.0786 MeV
VR1 = 0.027
VR2 = 0.00012 1/MeV
VR3 = 3.5e-07 1/MeV^2
VRLA = 96.2445 MeV
ALAVR = 0.00386032 1/MeV
Imaginary-surface:
WD0 = 0 MeV
WD1 = 0
WDA1 = 0
WDBW = 14.5489 MeV
WDWID = 12.0143 MeV
ALAWD = 0.013353 1/MeV
Imaginary-volume:
WC0 = 0 MeV
WC1 = 0
WCA1 = 0
WCBW = 17 MeV
WCWID = 105 MeV
BNDC = 0 MeV
Spin-orbit:
VS = 6.634 MeV
ALASO = 0.005 1/MeV
WSO = 0 MeV
WS1 = 0
WSBW = -3.1 MeV
WSWID = 160 MeV
Radius and diffuseness:
RR = 1.23906 fm
RRBWC = 0 fm
RRWID = 0 MeV
PDIS = 2
AR0 = 0.650409 fm
AR1 = 0 fm
RD = 1.21933 fm
AD0 = 0.66086 fm
AD1 = 0 fm/MeV
RC = 1.21 fm
AC0 = 0.685 fm
AC1 = 0 fm/MeV
RW = 0 fm
AW0 = 0 fm
AW1 = 0 fm/MeV
RS = 1.0751 fm
AS0 = 0.59 fm
AS1 = 0 fm/MeV
RZ = 1.264 fm
RZBWC = 0 fm
RRWID = 0 MeV
AZ = 0.341 fm
CCOUL = 0.9 MeV
ALF = 0
Coulomb correction:
CISO = 24.3 MeV
WCISO = 18 MeV
Deformation:
beta_2 = 0.223632
beta_4 = 0.09
beta_6 = -0.0031
* Calculated strength function
S0= 1.10e-4 S1= 1.81e-4 R'= 9.48 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of U-238
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 0 + *
1 0.04492 2 + *
2 0.14838 4 + *
3 0.30718 6 + *
4 0.51810 8 + *
5 0.68011 1 -
6 0.73193 3 -
7 0.77590 10 + *
8 0.82664 5 -
9 0.92721 0 +
10 0.93055 1 -
11 0.95012 2 -
12 0.96613 2 +
13 0.96631 7 -
14 0.99723 0 +
15 0.99758 3 -
16 1.02800 4 -
17 1.03725 2 +
18 1.05638 4 +
19 1.05773 3 +
20 1.05966 3 +
21 1.06027 2 +
22 1.07670 12 +
23 1.10571 3 +
24 1.12884 2 -
25 1.13075 4 +
26 1.13570 7 -
27 1.15070 9 -
-------------------
*) 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-239 17.1586 0.7762 3.0776 0.3946 -0.8805 3.0192
U-238 19.7315 1.5557 3.0121 0.3609 -0.0539 3.7504
U-237 18.2087 0.7795 2.7455 0.3666 -0.6759 2.7868
U-236 20.1615 1.5623 2.7551 0.3852 -0.3630 4.1437
U-235 18.5328 0.7828 2.6265 0.3874 -0.9246 3.1046
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
U-239 6.452 0.702 5.574 0.474
U-238 6.331 0.950 4.952 0.600
U-237 6.036 0.650 5.641 0.500
U-236 6.220 1.040 5.034 0.550
U-235 5.790 0.400 5.470 0.300
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-239 20.6471 0.9056 2.6000 0.3418 -1.7200 3.1056
U-238 20.5728 1.8150 2.6000 0.3324 -0.6607 3.8750
U-237 20.4984 0.9094 2.6000 0.3286 -1.5020 2.9094
U-236 20.4241 1.8226 2.6000 0.3293 -0.5887 3.8226
U-235 20.3497 0.9133 2.6000 0.3299 -1.4981 2.9133
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-239 22.3136 0.9056 0.2600 0.3564 -0.8646 3.0056
U-238 20.2054 1.8150 0.2200 0.4201 -0.4806 4.5150
U-237 21.5966 0.9094 0.1800 0.3569 -0.7807 2.9094
U-236 20.4241 1.8226 0.1400 0.3693 0.1220 3.8226
U-235 20.3497 0.9133 0.1000 0.3706 -0.7868 2.9133
--------------------------------------------------------
Table 7. Gamma-ray strength function for U-239
--------------------------------------------------------
K0 = 2.946 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 = 6.61 (MeV) EG = 4.00 (MeV) SIG = 4.47 (mb)
* E2: ER = 10.15 (MeV) EG = 3.24 (MeV) SIG = 6.50 (mb)
--------------------------------------------------------
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