94-Pu-240
94-Pu-240 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,+
DIST-MAY10 20100325
----JENDL-4.0 MATERIAL 9440
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-09 Numbers of neutrons per fission were revised.
07-05 Theoretical calculation was made with CCONE code.
07-11 Fission cross section was revised with simultaneous
evaluation.
07-11 Parameters of the -3 eV resonance was modified.
07-12 Fission cross section was revised with a new result of
simultaneous evaluation.
08-01 Fission cross section was revised.
08-02 Fission cross section and nu-p were revised.
CCONE calculation was made with revised parameters.
Data were compiled as JENDL/AC-2008/1/.
09-03 (1,452) and (1,455) were revised.
09-08 (MF1,MT458) was evaluated.
09-10 nu-p and fission cross section were revised.
10-01 Data of prompt gamma rays due to fission were given.
10-02 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
Determined from nu-d of three nuclides and partial fission
cross sections calculated with CCONE code/2/.
Pu-241 = 0.00911 measured by Benedetti et al./3/
Pu-240 = 0.0065 evaluation for Pu-239
Pu-239 = 0.00406 measured by Benedetti et al./3/
Decay constants were taken from evaluation by Brady and
England/4/.
MT=456 Number of prompt neutrons per fission
Experimental data of Frehaut et al./5/, Vorob'jova et al./6/
and Khokhlov et al./7/ were reproduced with a straight line.
The nu-p of Cf-252 spontaneous fission of 3.756/8/ was
applied.
Nu-p below 5 MeV was slightly increased.
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 2.7 keV)
Reich-Moore type resonance parameters given by Bouland et al.
/13/ were adopted. Capture widths of -3 and 1.056-eV levels
were slightly modified so as to reproduce the thermal capture
cross section better.
Bouland et al. analyzed the resonance parameters up to 5.7
keV. However, the upper boundary was set to 2.7 keV, because
the capture cross section was too small above this energy.
Small background to the capture cross section was given in
the energy range from 1 to 2.7 keV to the capture cross
sections by comparing with the data of Weston and Todd/14/.
The fission width of the negative resonance assumed at -3 eV
was modified to reproduce the fission cross section of 0.03 b
at 0.0253 eV/15/.
Unresolved resonance parameters (2.7 - 90 keV)
Parameters were determined with ASREP code/16/ to reproduce
the total and capture cross sections calculated with CCONE
code, and average fission cross section. The parameters are
used only for self-shielding calculations.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 292.01
elastic 2.667
fission 0.036 8.10
capture 289.31 8500
-------------------------------------------------------
(*) 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 (MT=18, 19, 20, 21, 38)
and capture cross sections were calculated with CCONE code/2/.
The model parameters were determined by considering integral
data as well as measured cross-section data. The fission cross
section of JENDL-3.3 was considered for determination of fission
related parameters.
MT= 1 Total cross section
The cross section was calculated with CC OMP of Soukhovitskii
et al./17/
MT= 2 Elastic scattering cross section
Calculated as total - non-elastic scattering cross sections.
MT=18 Fission cross section
Below 100 keV, JENDL-3.3 was adopted. JENDL-3.3 adopted the
data of JENDL-3.2 which were based on the data of Weston and
Todd/18/.
Above, 100 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 its
ratios by the SOK code/19/. Covariance matrix reported in
Iwasaki et al./20/ was also considered in the analysis.
--------------------------------------------------------------
Cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
21821.003 2.47E+6 M.Cance+ /21/
21821.002 2.47E+6 M.Cance+ /21/
30548.002 1.48E+7 N.A.Khan+ /22/
40636.005 1.46E+7 M.I.Kazarinova+ /23/
--------------------------------------------------------------
Ratio to U-235(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
21764.004 1.96E+5 - 9.75E+6 C.Budtz-Jorgensen+/24/
21764.002 1.51E+5 - 2.97E+5 C.Budtz-Jorgensen+/24/
41444.002 5.88E+5 - 2.97E+7 A.V.Fomichev+ /25/
13801.003 5.14E+5 - 2.95E+7 P.Staples+ /26/
22211.002 6.69E+5 - 6.57E+6 T.Iwasaki+ /20/
13576.002 5.20E+4 - 3.28E+5 J.W.Behrens /27/
12714.002 3.35E+5 - 9.60E+6 J.W.Meadows /28/
40509.002 1.27E+5 - 7.40E+6 V.M.Kuprijanov+ /29/
20766.005 5.21E+4 - 7.31E+4 K.Wisshak+ /30/
20766.003 5.09E+4 - 7.25E+4 K.Wisshak+ /30/
20766.002 5.62E+4 - 2.13E+5 K.Wisshak+ /30/
10597.002 3.39E+5 - 2.89E+7 J.W.Behrens+ /31/
--------------------------------------------------------------
Ratio to Pu-239(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
12766.003 5.29E+4 - 2.10E+7 L.W.Weston+ /32/
--------------------------------------------------------------
In the energy region from 500 keV to 10 MeV, obtained cross
section was increased slightly.
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, and multiplied by a factor to
reproduce well experimental data in the energy range from
170 keV to 4 MeV.
The experimental data of Weston and Todd/33/ and Wisshak
and Kaeppeler/34,35/ 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/2/.
MT=18 Fission
Isotropic distributions were assumed in the laboratory system.
MF= 5 Energy distributions of secondary neutrons
MT=18 Fission spectra
Calculated with CCONE code/2/.
MT=455 Delayed neutron spectra
(Same as JENDL-3.3)
Results of summation calculation made by Brady and England/4/
were adopted.
MF= 6 Energy-angle distributions
Calculated with CCONE code/2/.
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./36/ 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 Number of delayed neutrons per fission
Uncertainty of 5% /3/ was assumed below 5 MeV and 15%
above 5 MeV.
MT=456 Number of prompt neutrons per fission
Covariance matrix was obtained by fitting to the experimental
data of nu-p (See MF1/MT456).
MF=32 Covariances of resonance parameters
Only standard deviations of resonance parameters were given
on the basis of SAMMY fitting results /13/.
No correlation matrix was given.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/37/ 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
In the resolved resonance region, uncertainty of 3% was added
to the contributions from resonance parameter uncertainties.
Above 2.7 keV, covariances for CCONE calculation were
adopted.
MT=18 Fission cross section
In the resolved resonance region, uncertainty of 2% was added
to the contributions from resonance parameter uncertainties.
Between 2.7 to 90 keV, covariance matrix was estimated from
Ref./18/
Above 90 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
Obtained with CCONE and KALMAN codes/37/.
Uncertainty of 5% is given in the resolved resonance region.
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/38/
* Global parametrization of Koning-Duijvestijn/39/
was used.
* Gamma emission channel/40/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/41/ 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/42/. 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/43/,/44/
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 /17/
Volume:
V_0 = 49.8075 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 = 17.171 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 = 0.233263
beta_4 = 0.06237
beta_6 = -0.02167
* Calculated strength function
S0= 1.03e-4 S1= 2.66e-4 R'= 9.42 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Pu-240
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 0 + *
1 0.04282 2 + *
2 0.14169 4 + *
3 0.29432 6 + *
4 0.49752 8 + *
5 0.59734 1 -
6 0.64885 3 -
7 0.74233 5 -
8 0.74780 10 +
9 0.86071 0 +
10 0.90032 2 +
11 0.93806 1 -
12 0.95885 2 -
13 0.99220 4 +
14 1.00193 3 -
15 1.03053 3 +
16 1.03752 4 -
17 1.04180 12 +
18 1.05570 9 -
19 1.07622 4 +
20 1.08945 0 +
21 1.11553 5 -
22 1.13095 2 +
23 1.13697 2 +
24 1.16153 6 -
25 1.17750 3 +
26 1.18050 2 +
27 1.19900 3 +
28 1.22300 2 +
29 1.23246 4 +
30 1.24080 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
--------------------------------------------------------
Pu-241 18.5557 0.7730 2.1853 0.3475 -0.4724 2.5178
Pu-240 18.4895 1.5492 2.1440 0.3873 -0.0924 3.7908
Pu-239 18.4232 0.7762 1.8503 0.3562 -0.5010 2.5665
Pu-238 18.3569 1.5557 1.9652 0.3806 0.0280 3.6616
Pu-237 18.2906 0.7795 1.8799 0.3588 -0.5099 2.5875
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Pu-241 5.949 0.580 5.478 0.520
Pu-240 6.150 1.040 4.900 0.600
Pu-239 6.050 0.700 5.700 0.600
Pu-238 6.000 1.040 4.800 0.600
Pu-237 5.800 0.800 5.800 0.520
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Pu-241 20.4242 0.9018 2.6000 0.3647 -2.0579 3.4018
Pu-240 20.3513 1.8074 2.6000 0.3517 -0.9375 4.1074
Pu-239 20.2784 0.9056 2.6000 0.3523 -1.8394 3.2056
Pu-238 20.2054 1.8150 2.6000 0.3313 -0.6081 3.8150
Pu-237 20.1324 0.9094 2.6000 0.3320 -1.5137 2.9094
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
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
Pu-238 20.2054 1.8150 0.3400 0.3694 0.1086 3.8150
Pu-237 20.1324 0.9094 0.3000 0.3706 -0.7963 2.9094
--------------------------------------------------------
Table 7. Gamma-ray strength function for Pu-241
--------------------------------------------------------
K0 = 2.000 E0 = 4.500 (MeV)
* E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.96 (mb)
ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)
* M1: ER = 6.59 (MeV) EG = 4.00 (MeV) SIG = 3.29 (mb)
* E2: ER = 10.12 (MeV) EG = 3.22 (MeV) SIG = 6.78 (mb)
--------------------------------------------------------
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