94-Pu-242
94-Pu-242 JAEA+ EVAL-JAN10 O.Iwamoto, T.Nakagawa, Murata, +
DIST-MAY10 20100323
----JENDL-4.0 MATERIAL 9446
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-08 Nu-p was revised.
06-09 Resonance parameters were revised.
06-12 Fission cross section was revised.
07-05 Data were calculated with CCONE code.
Data were compiled as JENDL/AC-2008/1/.
09-03 (1,452), (1,455) and (1,456) were revised.
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 neutrons
Determined from nu-d of the following three nuclides and
partial fission cross sections calculated with CCONE code/2/.
Pu-243 = 0.0153 an average of experimental data of Krick
and Evans /3, 4/
Pu-242 = 0.011 0.00160 /5/ was multiplied by 0.7.
Pu-241 = 0.0064 0.00911 /5/ was multiplied by 0.7.
Values for Pu-242 and 241 were multiplied by 0.7 to
reproduce the nu-d of 0.0113+-0.0009 at 14.7MeV/6/
Decay constants were evaluated by Brady and England/7/.
MT=456 Number of prompt neutrons per fission
Least-squares fitting of a straight line to the experimental
data of Khokhlov et al./8/ Since their data were total
numbers of neutrons per fission, numbers of delayed neutrons
(MT=455) were subtracted.
nu-p = 2.8779 + 0.13755*E(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/9/. 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 2000/10/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/11/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/12/ 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 1 keV)
Resonance parameters of JENDL-3.3 were modified:
* Upper boundary was decreased from 1.9 keV to 1 keV.
* Capture width of 2.67-eV resonances was changed from 22
meV to 26.8 meV.
* Fission width of 53.46-eV resonance was increased from
1.825 micro-eV to 36 micro-eV.
Thermal capture cross section of 19.98+-0.66 b to be
reproduced was determined from Butler et al./13/, Durham
and Molson/14/ and Marie et al./15/.
Unresolved resonance parameters (1 keV - 100 keV)
Parameters were estimated with ASREP code/16/ so as to
total, fission and capture cross sections in this energy
region. They are used only for self-shielding calculations.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 28.213
elastic 8.326
fission 0.00244 4.36
capture 19.885 1130
-------------------------------------------------------
(*) 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
Calculated with CCONE code and modified below 500 keV by
multiplying an energy-dependent factor so as to reproduce
average total cross sections obtained from the data of Young
et al. /17/ The calculation was made with CC OMP of
Soukhovitskii et al./18/
MT= 2 Elastic scattering cross section
Calculated as total cross section - sum of partial cross
sections.
MT=18 Fission cross section
The following experimental data were analyzed in the energy
range from 1 keV to 20 MeV with the GMA code/19/:
Authors Energy range Data points Reference
Bulter 141 keV - 1.66 MeV 65 /20/
Fomushkin+ 14.5 MeV 1 /21/
Bergen+ 0.1 - 2.96 MeV 141 /22/(*1)
Auchampaugh+ 0.95 keV - 3.99 MeV 3102 /23/
Meadows 0.397 - 9.92 MeV 49 /24/(*2)
Behrens 97.2 keV - 20.0 MeV 133 /25/(*2)
Kuprijanov+ 0.127 - 7.4 MeV 71 /26/(*2)
Cance+ 2.47 MeV 2 /27/
Alkhazov+ 14.7 MeV 1 /28/
Weigmann+ 0.3 - 9.7 MeV 222 /29/
Arlt+ 14.7 MeV 1 /30/
Meadows 14.7 MeV 1 /31/(*2)
Iwasaki+ 0.597 - 6.76 MeV 17 /32/(*2)
Staples 0.514 - 19.5 MeV 124 /33/(*2)
*1) only the data above 100 keV were used.
*2) ratio to U-235 fission cross section
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 /34,35/ and Hockenbury et al./36/ 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 Fission spectra
Calculated with CCONE code.
MT=455 Delayed neutron spectra
(Same as JENDL-3.3)
Results of summation calculation made by Brady and England/7/
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./37/ 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 of 10% was assumed below 5 MeV and above 5 MeV,
respectively by comparing with experimental data/4, 6/
MT=456
Covariance was obtained by fitting a linear function to the
experimental data of Khokhlov et al./8/(see MF1,MT456).
Variances were multiplied by a factor of 2.
MF=32 Covariances of resonance parameters
Format of LCOMP=0 was adopted.
Standard deviations of resonance energy, neutron and capture
widths were taken from Mughabghab /38/ Those of fission
width were based on the data of fission area reported by
Weigmann et al./29/ and Auchampaugh et al./23/.
If no information was available, uncertainties were assumed.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/39/ 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 resonance region (below 1 keV), uncertainty of 8 %
was added.
Above 1 keV, covariance matrix was obtained with CCONE and
KALMAN codes/39/.
MT=18 Fission cross section
In the resonance region from 10 to 1000 eV, addtional error
of 50% was given.
Above the resonance region, cross section was evaluated with
GMA code/19/. Standard deviation obatianed was multiplied
by a factor of 2.0.
MT=102 Capture cross section
In the resonance region from 10 to 1000 eV, addtional error
of 10% was given.
Above 1 keV, covariance matrix was obtained with CCONE and
KALMAN codes/39/.
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/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/40/
* Global parametrization of Koning-Duijvestijn/41/
was used.
* Gamma emission channel/42/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/43/ 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/44/. 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/45/,/46/
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 /18/
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 = 0.23892
beta_4 = 0.04807
beta_6 = -0.02376
* Calculated strength function
S0= 0.98e-4 S1= 2.93e-4 R'= 9.32 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Pu-242
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 0 + *
1 0.04454 2 + *
2 0.14730 4 + *
3 0.30640 6 + *
4 0.51810 8 + *
5 0.77860 10 +
6 0.78045 1 -
7 0.83230 3 -
8 0.86500 3 +
9 0.92700 5 -
10 0.95600 0 +
11 0.99250 2 +
12 1.01950 3 -
13 1.03920 1 +
14 1.06400 4 -
15 1.08440 12 +
16 1.09210 6 +
17 1.10200 2 +
18 1.12200 5 -
19 1.15100 2 -
20 1.15450 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
--------------------------------------------------------
Pu-243 18.6999 0.7698 2.4578 0.3280 -0.3352 2.3315
Pu-242 18.6337 1.5428 2.4520 0.3701 0.0291 3.6198
Pu-241 18.5675 0.7730 2.1853 0.3473 -0.4715 2.5167
Pu-240 18.5012 1.5492 2.1440 0.3871 -0.0917 3.7899
Pu-239 18.4349 0.7762 1.8503 0.3560 -0.5001 2.5655
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Pu-243 5.750 0.680 5.520 0.520
Pu-242 6.100 1.000 4.850 0.600
Pu-241 5.950 0.580 5.480 0.520
Pu-240 6.250 1.040 4.920 0.600
Pu-239 6.050 0.700 5.700 0.600
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Pu-243 20.5699 0.8981 2.6000 0.3633 -2.0616 3.3981
Pu-242 20.4971 1.7999 2.6000 0.3503 -0.9450 4.0999
Pu-241 20.4242 0.9018 2.6000 0.3647 -2.0579 3.4018
Pu-240 20.3513 1.8074 2.6000 0.3300 -0.6156 3.8074
Pu-239 20.2784 0.9056 2.6000 0.3523 -1.8394 3.2056
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Pu-243 20.9439 0.8981 0.5400 0.3740 -0.9758 3.0981
Pu-242 20.4971 1.7999 0.5000 0.3933 -0.2466 4.1999
Pu-241 20.4242 0.9018 0.4600 0.3804 -0.9744 3.1018
Pu-240 20.5363 1.8074 0.4200 0.3796 -0.0661 4.0074
Pu-239 20.2784 0.9056 0.3800 0.3901 -1.0534 3.2056
--------------------------------------------------------
Table 7. Gamma-ray strength function for Pu-243
--------------------------------------------------------
K0 = 2.100 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 = 6.57 (MeV) EG = 4.00 (MeV) SIG = 3.39 (mb)
* E2: ER = 10.10 (MeV) EG = 3.19 (MeV) SIG = 6.78 (mb)
--------------------------------------------------------
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