94-Pu-239
94-Pu-239 JAEA+ EVAL-DEC09 O.Iwamoto,N.Otuka,S.Chiba,+
DIST-MAY10 20100325
----JENDL-4.0 MATERIAL 9437
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
07-03 Data were calculated with CCONE code.
07-06 Fission spectrum was revised.
07-07 Theoretical calculation was made with CCONE code.
07-08 Theoretical calculation was made with CCONE code.
07-11 Fission cross section was revised with simultaneous
evaluation.
07-12 Fission cross section was revised with a new result of
simultaneous evaluation.
08-01 Fission cross section and resonance parameters were
revised. New CCONE calculation was adopted.
08-02 Re-calculated with CCONE code. Fission cross section and
nu-p were revised.
08-03 nu-p was revised.
Interpolation of (5,18) was changed.
Data were compiled as JENDL/AC-2008/1/.
09-08 (MF1,MT458) was evaluated.
09-10 nu-p and fission cross section were revised.
09-12 nu-p was revised.
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
(same as JENDL-3.3/2/)
Evaluated data by Tuttle/3/ were adopted.
Decay constants were adopted from Keepin et al./4/
MT=456 Number of prompt neutrons per fission
Below 20 eV:
Determined from experimental data of Gwin et al./5/
reducing by 0.1%. Standard Cf-252 sf nu-p was taken to be
3.756.
40 - 500 eV:
Determined from experimental data of Frehaut et al./6/,
Gwin et al./7, 5/
500 eV - 20 keV:
Determined from experimental data of Gwin et al./7, 8/
Above 20 keV:
Experimental data were anlyzed by the GMA code/9/ with
the Chiba and Smith approach/10/ for PPP minimization.
Experimental data were renormalized with nu-p of Cf-252
spontaneous fission (3.756+/-0.031) reported by Vorobyev et
al./11/ if standards to derive original data were known.
Experimental data sets are summarized below.
r: re-normalized by nu-p(Cf-252 spon) of A.S.Vorobyev et al.
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
30600.002 1.86E+5 - 1.44E+6 H.Q.Zhang+ /12/
r 21135.007 9.90E+5 - 4.02E+6 D.S.Mather+ /13/
r 12326.006 2.50E+5 - 1.45E+7 J.C.Hopkins+ /14/
21696.006 1.41E+7 I.Johnstone /15/
21685.004 2.28E+7 - 2.83E+7 J.Frehaut /16/
13101.004 5.50E+2 - 9.50E+6 R.Gwin+ /8/
10759.004 5.60E+2 - 6.80E+6 R.Gwin+ /7/
20453.003 5.50E+5 - 8.50E+5 D.S.Mather+ /17/
20453.002 7.75E+4 - 1.15E+6 D.S.Mather+ /17/
--------------------------------------------------------------
Slight modification was made for JENDL-4.0 in the energy
range from 1 keV to 4 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/18/. 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 2000/19/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/20/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/21/ 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: below 2.5 keV)
Parameters were evaluated by Derrien/22/.
Unresolved resonance parameters (2.5 keV - 30 keV)
The energy dependent S0, S1 and fission width were determined
with ASREP code/23/ so as to reproduce the evaluated total,
capture and fission cross sections. 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 1027.40
elastic 8.82
fission 742.76 301
capture 275.82 180
-------------------------------------------------------
(*) 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/24/.
The model parameters were determined by considering integral
data as well as measured total, (n,2n) and capture cross
sections, and fission cross section of JENDL-3.3.
MT= 1 Total cross section
The cross section was calculated with CC OMP of Soukhovitskii
et al./25/ The OMPs were modified based on the experimental
data of Harvey et al./26/ and Poenitz et al./27,28/
MT=2 Elastic scattering cross section
Calculated as total - non-elstic scattering cross sections
MT=16 (n,2n) cross section
Calculated with CCONE code. The experimental data of Becker
et al./29/ and Lougheed et al./30/ were used to
determine the model parameters in the CCONE calculation.
MT=18 Fission cross section
Below 10 keV, JENDL-3.3/2/ was adopted, which was based on
measurements of ref./31/ and ref./32/.
Above, 10 keV, experimental data measured after 1960 were
analyzed in the present work by simultaneous fitting of U-
233, U-235, U-238, Pu-239, Pu-240 and Pu-241 fission cross
sections and their ratios by the SOK code /33/.
--------------------------------------------------------------
Cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
22304.009 1.47E+7 K.Merla+ /34/
22304.005 4.90E+6 - 1.88E+7 K.Merla+ /34/
12877.005 5.05E+3 - 2.05E+5 L.W.Weston+ /35/
30670.002 1.00E+6 - 5.60E+6 X.J.Zhou+ /36/
30634.003 1.47E+7 J.W.Li+ /37/
12826.003 1.46E+7 M.Mahdavi+ /38/
40487.003 5.50E+3 - 9.50E+4 Ju.V.Rjabov /39/
20779.005 1.39E+7 - 1.46E+7 M.Cance+ /40/
10314.003 1.40E+5 - 9.64E+5 M.C.Davis+ /41/
20618.003 2.35E+6 - 5.53E+6 I.Szabo+ /42/
20570.003 8.05E+5 - 2.61E+6 I.Szabo+ /42/
20567.003 3.50E+4 - 9.72E+5 I.Szabo+ /42/
20001.002 5.05E+3 - 3.05E+4 J.Blons /43/
20002.002 5.05E+3 - 2.95E+4 B.H.Patrick+ /44/
10267.002 5.50E+3 - 1.50E+5 R.Gwin+ /45/
40927.006 1.92E+6 I.D.Alkhazov+ /46/
--------------------------------------------------------------
Ratio to U-235(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
41455.005 5.77E+5 - 2.94E+7 O.A.Shcherbakov+ /47/
13801.002 8.50E+5 - 2.95E+7 P.Staples+ /48/
13134.009 1.47E+7 J.W.Meadows /49/
30588.005 1.35E+7 - 1.48E+7 M.Varnagy+ /50/
10562.002 5.00E+3 - 2.96E+7 G.W.Carlson+ /51/
40824.003 2.40E+4 - 7.40E+6 B.I.Fursov+ /52/
40824.002 3.20E+5 - 7.00E+6 B.I.Fursov+ /52/
20569.004 1.15E+4 - 1.99E+5 I.Szabo+ /42/
20409.003 3.92E+5 - 2.09E+7 S.Cierjacks+ /53/
20428.004 5.50E+3 - 9.50E+5 D.B.Gayther /54/
10253.002 3.00E+4 - 5.49E+6 W.P.Poenitz /55/
20363.003 5.20E+3 - 1.01E+6 E.Pfletschinger+ /56/
10086.004 1.50E+5 - 1.40E+6 W.P.Poenitz /57/
-----.--- 5.00E+5 - 2.96E+7 P.W.Lisowski+ /58/
--------------------------------------------------------------
Thus obtained fission cross section was further modified by
multiplying a factor of 1.003 in the energy range from 2.5 keV
to 5 MeV. The data in the 7-8 MeV region were also modified.
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
Schomberg et al./59/, Chelnokov et al./60/, Gwin et al./45/,
Kononov et al./61/ and Hopkins and Diven/62/ were used to
determine the model 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 Prompt fission neutrons
Below 5 MeV, data of JENDL-3.3/2/ were adopted.
Comment of JENDL-3.3:
* Distributions were calculated with a modified Madland-Nix
model with consideration for multimodal nature of the
fission process/63,64/. The compound nucleus formation
cross sections for fission fragments were calculated using
Becchetti-Greenlees potential/65/. Up to 3rd-chance-fission
were considered at high incident neutron energies.
Parameters adopted for thermal-neutron fission/64/:
(S1: standard-1, S2: standard-2, S3: standard-3 modes)
total average fragment kinetic energy
= 190.4 MeV for S1
= 174.2 MeV for S2
= 164.2 MeV for S3
average energy release = 205.400 MeV for S1
= 196.279 MeV for S2
= 182.123 MeV for S3
average mass number of light FF = 105 for S1
= 99 for S2
= 83 for s3
average mass number of heavy FF = 135 for S1
= 141 for S2
= 157 for s3
level density of the light FF = 11.236(S1),
10.764(S2), 6.669(S3)
level density of the heavy FF = 9.577(S1),
13.104(S1),16.284(S3)
mode branching ratio = 0.248(S1), 0.742(S2), 0.01(S3)
These data are essentially based on Schillebeeckx et
al./66/.
Note that the parameters vary with the incident energy.
Energy-dependent mode branching ratio data of Brosa et
al./67/ was used.
Above 5.5 MeV, spectra were calculated with CCONE code/24/.
MT=455
(same as JENDL-3.3)
Taken from Brady and England /68/. Group abundance parameters
were adjusted so as to reproduce total delayed neutron
emission rate measured by Keepin et al./4/, Besant et al.
/69/ and Maksyutenko/70/
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF=12 Photon production multiplicities and transition
probability arrays
MT=18 Fission
(same as JENDL-3.3)
Stored under option-1 (multiplicities). The thermal neutron
induced fission gamma spectrum measured by Verbinski et
al./71/ was adopted and used up to 20 MeV neutron. Since
no data were given for the photons below 0.14 MeV, it was
assumed to be the same as that of the photons between 0.14
and 0.3 MeV.
MF=14 Photon angular distributions
MT=18
Isotropic distributions were assumed.
MF=15 Continuous photon energy spectra
MT=18
(same as JENDL-3.3)
Experimental data by Verbinski et al./71/ 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
(same as JENDL-3.3/2/)
MT=456 Number of prompt neutrons per fission
Below 500 eV, the same covariance as JENDL-3.3 was adopted.
Above 500 eV, covariance was obtained by GMA fitting to the
experimental data of nu-p (see MF=1,MT=456).
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/72/ 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 energy region up to 2.0 keV, covariances were
calculated from the covariances of resonance parameters/73/.
From 2 to 2.5 keV, uncertaintes were assumed to be 3% for
the total, and 4% for elastic scattering cross sections.
Above 2.5 keV, the covariances for CCONE calculation were
adopted.
MT=18 Fission cross section
Below 2.0 keV, covariances were calculated from the
covariances of resonance parameters/73/. From 2 to 9 keV,
uncertaintes were assumed to be 2.5%.
Above 9 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
Below 2.0 keV, covariances were calculated from the
covariances of resonance parameters/73/. From 2 to 2.5 keV,
uncertaintes were assumed to be 5 %.
Above 2.5 keV, covariances were obtained with CCONE and
KALMAN codes/72/.
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 JENDL-3.3 data.
Above 5 MeV, covariances were estimated with CCONE and
KALMAN codes.
*****************************************************************
Calculation with CCONE code
*****************************************************************
Models and parameters used in the CCONE/24/ 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/74/
* Global parametrization of Koning-Duijvestijn/75/
was used.
* Gamma emission channel/76/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/77/ 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/78/. 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/79/,/80/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,2,3,4,5,7,9 (see Table 2)
* optical potential parameters /25/
Volume:
V_0 = 50.054 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 = 17.1463 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 = 0.227635
beta_4 = 0.06501
beta_6 = -0.01837
* Calculated strength function
S0= 1.09e-4 S1= 2.54e-4 R'= 9.31 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Pu-239
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 1/2 + *
1 0.00786 3/2 + *
2 0.05727 5/2 + *
3 0.07570 7/2 + *
4 0.16376 9/2 + *
5 0.19280 11/2 + *
6 0.28546 5/2 +
7 0.31850 13/2 + *
8 0.33012 7/2 +
9 0.35810 15/2 + *
10 0.38742 9/2 +
11 0.39158 7/2 -
12 0.43400 9/2 -
13 0.46200 11/2 +
14 0.46980 1/2 -
15 0.48700 11/2 -
16 0.49210 3/2 -
17 0.50560 5/2 -
18 0.51184 7/2 +
19 0.51930 17/2 +
20 0.53800 7/2 +
21 0.55620 7/2 -
22 0.56500 9/2 +
23 0.57060 19/2 +
24 0.58300 9/2 -
25 0.62000 15/2 -
26 0.63400 11/2 +
27 0.66110 11/2 -
28 0.71600 13/2 -
29 0.75250 1/2 +
30 0.75600 11/2 -
-------------------
*) 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-240 18.5472 1.5492 2.1440 0.3862 -0.0888 3.7864
Pu-239 18.4808 0.7762 1.8503 0.3551 -0.4969 2.5616
Pu-238 18.4143 1.5557 1.9652 0.3796 0.0313 3.6576
Pu-237 18.3478 0.7795 1.8799 0.3578 -0.5058 2.5825
Pu-236 18.2812 1.5623 1.9752 0.3731 0.1229 3.5604
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Pu-240 6.250 1.040 5.000 0.600
Pu-239 5.750 0.700 5.550 0.600
Pu-238 5.400 0.700 5.100 0.600
Pu-237 5.800 0.800 5.800 0.520
Pu-236 6.000 1.040 5.000 0.600
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
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
Pu-236 20.0594 1.8226 2.6000 0.3326 -0.6004 3.8226
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Pu-240 20.9063 1.8074 0.4200 0.3827 -0.1468 4.1074
Pu-239 19.7253 0.9056 0.3800 0.3818 -0.8914 3.0056
Pu-238 19.2870 1.8150 0.3400 0.3797 0.0977 3.8150
Pu-237 20.1324 0.9094 0.3000 0.3706 -0.7963 2.9094
Pu-236 20.0594 1.8226 0.2600 0.3719 0.1175 3.8226
--------------------------------------------------------
Table 7. Gamma-ray strength function for Pu-240
--------------------------------------------------------
K0 = 2.000 E0 = 4.500 (MeV)
* E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 303.06 (mb)
ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)
* M1: ER = 6.60 (MeV) EG = 4.00 (MeV) SIG = 3.31 (mb)
* E2: ER = 10.14 (MeV) EG = 3.23 (MeV) SIG = 6.79 (mb)
--------------------------------------------------------
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