94-Pu-241
94-Pu-241 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,+
DIST-MAY10 20100325
----JENDL-4.0 MATERIAL 9443
-----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.
07-12 Fission cross section was revised with a result of new
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.
09-09 fission and total sigs and URP were revised.
09-10 fission cross section was revised.
09-12 New theoretical calculation was made with CCONE code.
Nu-d was revised.
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's = 455 and 456.
MT=455 Delayed neutron data
Determined from nu-d of the following three nuclides and
partial fission cross sections calculated with CCONE code/2/.
Pu-242 = 0.0160 measured by Benedetti et al./3/
Pu-241 = 0.0064 data measured by Benedetti et al./3/
reduced by a factor of 0.7
Pu-240 = 0.0046 data for Pu-239, reduced by a factor
of 0.7
Decay constants were adopted from Ref./4/.
MT=456 Number of prompt neutrons per fission
Experimental data of Conde et al./5/, Frehaut et al./6/ and
D'yachenko et al./7/ were reproduced with two straight lines.
A thermal value of 2.929 evaluated by Holden and Zucker/8/
was also considered. They were re-normalized to the nu-p of
Cf-252 spontaneous fission of 3.756/9/.
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/10/. 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 2000/11/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/12/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/13/ 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 300 eV)
Revised Derrien's evaluation/14/ below 20 eV was adopted.
Numerical data were taken from ENDF/B-VII.0/15/.
See Apendix A-1.
Unresolved resonance parameters (300 eV - 30 keV)
Parameters were determined with ASREP code/16/ to reproduce
capture cross section measured by Weston and Todd/17/,
fission cross section measured by Weston and Todd/17/,
Migneco et al./18/ and Gerasimov et al./19/, and total cross
section calculated with CCONE code.
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 1386.67
elastic 11.26
fission 1012.34 567
capture 363.06 180
-------------------------------------------------------
(*) 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), 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 capture cross sections data 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./20/.
MT=2 Elastic scattering cross section
From 300 eV to 30 keV, CCONE calculation was adopted.
Above 30 keV, it was calculated as (total)-(non-elastic
scattering cross sections).
MT=18 Fission cross section
From 300 eV to 10 keV, the data of Weston and Todd/17/,
Migneco et al./18/ and Gerasimov et al. /19/ were taken
into consideration.
Above 10 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
ratios by the SOK code /21/.
--------------------------------------------------------------
Cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
30548.003 1.48E+7 N.A.Khan+ /22/
10636.002 5.00E+3 - 6.50E+4 G.W.Carlson+ /23/
20570.004 1.18E+6 - 2.63E+6 I.Szabo+ /24/
20567.004 3.50E+4 - 9.70E+5 I.Szabo+ /24/
20484.002 5.00E+3 - 2.98E+4 J.Blons+ /25/
40636.007 1.46E+7 M.I.Kazarinova+ /26/
--------------------------------------------------------------
Ratio to U-235(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
40474.003 2.40E+4 - 7.40E+6 B.I.Fursov+ /27/
40474.003 1.27E+5 - 7.00E+6 B.I.Fursov+ /27/
10563.002 5.00E+3 - 2.96E+7 J.W.Behrens+ /28/
20364.002 1.37E+4 - 1.13E+6 F.Kaeppeler+ /29/
--------------------------------------------------------------
Cross section was slightly modified in the energy region from
1 to 4 MeV, and from 7 to 8 MeV for JENDL-4.
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
From 300 eV to 30 keV, the cross section was calculated from
the unresolved resonance parameters mentioned above. CCONE
calculation was adopted in the energy range above 50 keV.
The experimental data of Weston and Todd/17/ 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 were assumed in the laboratory system.
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/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./30/ 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 in the energy region
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.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/31/ 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 5% was added
to the contributions from resonance parameter uncertainties.
Above 300 eV, covariances for CCONE calculation were adopted.
MT=18 Fission cross section
Error of 2 % was assumed in the resolved resonance region up
to 300 eV.
In the energy range from 300 eV to 10 keV, the fission cross
section was determined from the experimetal data of Gerassimov
et al./19/ Error of 5% was assumed in the region up to
9 keV.
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
Error of 10 % was assumed in the resolved resonance region.
In the energy region from 300 eV to 2 keV, error of 15% was
assumed, and from 2 to 30 keV, error of 10%, by considering
dispersion of average cross sections of Weston and Todd/17/.
Above 30 keV, Covariance matrix was obtained with CCONE and
KALMAN codes/31/.
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/32/
* Global parametrization of Koning-Duijvestijn/33/
was used.
* Gamma emission channel/34/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/35/ 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/36/. 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/37/,/38/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,2,3,9 (see Table 2)
* optical potential parameters /20/
Volume:
V_0 = 49.682 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 = 17.45 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 = 0.24731
beta_4 = 0.05852
beta_6 = -0.02486
* Calculated strength function
S0= 1.19e-4 S1= 2.62e-4 R'= 9.47 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Pu-241
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 5/2 + *
1 0.04197 7/2 + *
2 0.09578 9/2 + *
3 0.16131 11/2 + *
4 0.16168 1/2 +
5 0.17094 3/2 +
6 0.17505 7/2 +
7 0.22299 5/2 +
8 0.23194 9/2 +
9 0.23500 13/2 + *
10 0.24489 7/2 +
11 0.30117 11/2 +
12 0.33700 1/2 +
13 0.33714 9/2 +
14 0.37300 11/2 +
15 0.37600 1/2 -
16 0.38500 13/2 +
17 0.40445 9/2 -
18 0.40890 7/2 -
19 0.44600 11/2 -
20 0.47300 1/2 -
21 0.49500 7/2 +
22 0.50300 13/2 +
23 0.51881 5/2 -
24 0.53420 7/2 +
25 0.56142 7/2 -
26 0.57000 15/2 -
27 0.61484 9/2 -
28 0.64500 13/2 -
29 0.68100 1/2 -
30 0.75517 1/2 +
31 0.76927 1/2 -
32 0.77915 3/2 -
33 0.78415 3/2 +
34 0.80044 3/2 +
35 0.80048 5/2 +
36 0.81095 5/2 -
37 0.83159 5/2 +
38 0.83330 7/2 -
39 0.83484 3/2 +
40 0.84196 1/2 -
-------------------
*) 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-242 18.5806 1.5428 2.4520 0.3710 0.0263 3.6230
Pu-241 19.1301 0.7730 2.1853 0.3372 -0.4308 2.4675
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
Pu-238 18.3685 1.5557 1.9652 0.3804 0.0287 3.6608
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Pu-242 6.172 0.998 4.674 0.600
Pu-241 5.795 0.580 5.487 0.520
Pu-240 6.250 1.040 4.920 0.600
Pu-239 6.050 0.700 5.700 0.600
Pu-238 6.000 1.040 4.800 0.600
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Pu-242 20.4971 1.7999 2.6000 0.3433 -0.8376 3.9999
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
Pu-238 20.2054 1.8150 2.6000 0.3313 -0.6081 3.8150
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
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
Pu-238 20.2054 1.8150 0.3400 0.3694 0.1086 3.8150
--------------------------------------------------------
Table 7. Gamma-ray strength function for Pu-242
--------------------------------------------------------
K0 = 2.481 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 = 6.58 (MeV) EG = 4.00 (MeV) SIG = 3.87 (mb)
* E2: ER = 10.11 (MeV) EG = 3.21 (MeV) SIG = 6.78 (mb)
--------------------------------------------------------
References
1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009).
2) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007).
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4) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989).
5) H.Conde et al.: J. Nucl. Energy, 22, 53 (1968).
6) J.Frehaut et al.: CEA-R-4626 (1974).
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8) N.E.Holden, M.S.Zucker: Nucl. Sci. Eng., 98, 174 (1988).
9) A.S.Vorobyev et al.: 2004 Santa Fe, Vol.1, p.613 (2004).
10) G.Audi: Private communication (April 2009).
11) J.Katakura et al.: JAERI 1343 (2001).
12) T.R.England et al.: LA-11151-MS (1988).
13) R.Sher, C.Beck: EPRI NP-1771 (1981).
14) H.Derrien et al.: Nucl. Sci. Eng., 150, 109 (2005).
15) M.B.Chadwick et al.: Nucl. Data Sheets, 107, 2931 (2006).
16) Y.Kikuchi et al.: JAERI-Data/Code 99-025 (1999) in Japanese.
17) L.W.Weston, J.H.Todd: Nucl. Sci. Eng., 65, 454 (1978).
18) E.Migneco et al.: 1970 Helsinki, p.437 (1970).
19) V.F.Gerasimov et al.: JINR-E3-97-213, p.348 (1997).
20) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).
21) T.Kawano et al.: JAERI-Research 2000-004 (2000).
22) N.A.Khan et al.: Nucl. Instrum. Methods, 173, 137 (1980).
23) G.W.Carlson et al.: Nucl. Sci. Eng., 63, 149 (1977).
24) I.Szabo et al.: 1976 ANL, p.208 (1976).
25) J.Blons et al.: 1971 Knoxville, Vol.2, p.836 (1971).
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27) B.I.Fursov et al.: At. Energy, 44, 236 (1978).
28) J.W.Behrens et al.: Nucl. Sci. Eng., 68, 128 (1978).
29) F.Kaeppeler et al.: Nucl. Sci. Eng., 51, 124 (1973).
30) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).
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Japanese.
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33) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).
34) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).
35) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).
36) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).
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38) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).
Appendix A-1: Resonance parameters
----------------------------------------------------------------
Reevaluation of rhe resonance parameters in the energy range up
to 20 eV - Colaboration ORNL(USA)-CEN/CAD(FRANCE)
H. Derrien and L.C Leal, ORNL
A. Courcelle and A. Santamarina, CEN/CAD
(submitted to NSE, 2003)
Cross section integrals in the energy range 0.0021 eV to 20.0 eV
-------------------------------------------------------------
Energy Range Capture b-eV Fission b-eV
eV 1993 Present 1993 Present
-------------------------------------------------------------
0.0021-0.020 12.25 12.51 2.1% 31.06 31.43 1.2%
0.0200-0.030 3.67 3.72 1.4% 10.24 10.36 1.2%
0.0300-0.100 15.28 15.59 2.0% 49.02 49.14 0.2%
0.1000-0.500 110.58 117.40 6.2% 262.76 263.84 0.4%
0.5000-1.000 5.90 6.03 2.2% 17.93 17.89 -0.2%
1.0000-3.000 7.30 7.16 -2.0% 54.88 52.89 -3.8%
3.0000-20.00 1213. 1234. 1.7% 3039. 3026. -0.4%
-------------------------------------------------------------
The cross sections at 0.0253 eV are the following:
Total 1386.5 b
Fission 1012.2 b
Capture 363.0 b
very close to the current standard.
The fit of the experimental data (Young total, Wagemans fission,
Weston capture and Weston fission ) were performed by assuming
that young total and Wagemans fission were OK with the standard
and that Weston original data(from the EXFOR file)needed a renor-
zation to agree with the standard. A consistent normalization
was obtained by a SAMMY fit leading to a decrease of 4.5%
in the original Weston fission and a decrease of 3.0% in the
original Weston capture.The differences between the 1988, 1993
and the present results are consistent with the experimental
errors given by Weston for the measured fission (2-3%) and
capture (10-15%). The default of normalization of the Weston
fission,due to severe experimental errors in the thermal range,
was responsible for the large uncertainty in the capture results.