92-U -233
92-U -233 JAEA+ EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,et al.
DIST-SEP12 20111206
----JENDL-4.0u1 MATERIAL 9222
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
Update File Distribution
Sep.14,2012 JENDL-4.0u1
History
06-07 Total and fission cross sections were modified.
06-10 Nu-p was modified.
07-06 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 new results 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.
08-03 Interpolation of (5,18) was changed.
Data were compiled as JENDL/AC-2008/1/.
09-04 MF01 was revised.
09-08 (MF1,MT458) was evaluated.
09-10 fission cross section was revised.
10-01 Data of prompt gamma rays due to fission were given.
10-03 Covariance data were given.
11-07 Covariance data in RRR were revised.
MF= 1 General information
MT=452 Number of Neutrons per fission
Sum of MT=455 and 456.
MT=455 Delayed neutron data
At the thermal neutron energy, nu-d of 0.0067 was obtained
from the data of Borzakov et al./2/, Conant et al./3/ and
Keepin et al./4/.
Nu-d above 10 keV was determined from experimental data
measured by Krick and Evans/5,6/, Piksaykin et al./7/,
and Masters et al./8/.
Decay constants were taken from Ref./9/.
MT=456 Number of prompt neutrons per fission
Nu-p was determined from the data of Protopopov et al./10/,
Smirenkin et al./11/, Flerov et al./12/, Hopkins et al./13/,
Colvin et al./14,15/, Mather et al./16/, Walsh et al./17/,
Hockenbury/18/, Nurpeisov et al./19,20/, Sergachev et al./21/,
Nefedov et al./22/, and Gwin et al./23,24/
Nu-p of Cf-252 SF = 3.756+-0.031 /25/ was used.
They were reproduced with two straight lines below and above
about 1.5 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/26/. 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 2000/27/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/28/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/29/ 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: 1.0-5 eV - 600 eV)
Evaluation by Leal et al. for ENDF/B-VII.0 was adopted.
See Appendix A1.
Unresolved resonance parameters (600 eV - 30 keV)
Parameters were determined with ASREP code/30/ so as to
reproduce total, fission and capture cross sections.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 588.77
elastic 12.18
fission 531.34 775
capture 45.26 139
-------------------------------------------------------
(*) 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 total (MT=1), elastic scattering (MT=2) and fission cross
sections (MT=18, 19, 20, 21, 38) were calculated with CCONE
code/31/. The model parameters were determined by considering
integral experimental data as well as measured cross-section
data.
In the CCONE calculation the CC OMP of Soukhovitskii et al./32/
was modified so as to reproduce well the experimental data of
total cross section measured by Poenitz et al./33,34/ and
Guber et al./35/
Other parameters of CCONE calculation were adjusted to the
fission cross section of JENDL-3.3 and the capture cross section
measured by Hopkins and Diven/36/.
The results of CC calculation for the elastic scattering was
increased by 0.2 b above 1.5 MeV to improve integral benchmark
tests.
MT=1 Total cross section
Experimental data measured after 1960 were anlyzed by the GMA
code/37/ with the Chiba and Smith approach/38/ for PPP
minimization.
Experimental data sets are summarized below.
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
10047.095 2.26E+6 - 1.50E+7 D.G.Foster Jr.+ /39/
10225.028 5.05E+5 - 7.96E+6 L.Green+ /40/
10025.028 9.00E+5 - 9.89E+6 L.Green+ /40/
10833.003 4.00E+4 - 2.09E+5 W.P.Poenitz+ /41/
10833.002 5.80E+4 - 4.43E+6 W.P.Poenitz+ /41/
10935.005 4.80E+4 - 4.81E+6 W.P.Poenitz+ /33/
12323.002 3.40E+3 - 1.61E+6 D.C.Stupegia /42/
12333.002 6.01E+2 - 8.81E+3 N.J.Pattenden+ /43/
12853.053 1.82E+6 - 2.03E+7 W.P.Poenitz+ /34/
13891.004 6.09E+2 - 6.85E+5 K.H.Guber+ /35/
--------------------------------------------------------------
MT=2 Elastic scattering cross section
Calculated as total - non-elstic scattering cross sections
MT=18 Fission cross section
Below 10 keV, experimental data measured after 1960 were
anlyzed by the GMA code/37/ with the Chiba and Smith
approach/38/ for PPP minimization. Data were normalized to
absolute cross section by adopting the JENDL-3.3 U-235(n,f)
cross section if the data were given as the ratios to the
U-235(n,f) cross section.
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
ratio by the SOK code/44/. Covariance matrix reported in
Manabe et al./45/ was also considered in the analysis.
Experimental data sets are summarized below.
g: used in GMA analysis, s: used in SOK analysis
--------------------------------------------------------------
Cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
gs 13890.002 6.00E+2 - 7.01E+5 K.H.Guber+ /46/
gs 40927.002 1.94E+6 V.I.Shpakov /47/
gs 12910.002 1.46E+7 K.R.Zasadny+ /48/
s 40911.003 1.47E+7 I.D.Alkhazov+ /49/
s 40610.002 4.40E+4 E.A.Zhagrov+ /50/
s 40587.002 2.45E+4 A.V.Murzin+ /51/
gs 10756.002 1.37E+5 - 8.05E+6 W.P.Poenitz /52/
gs 40547.003 1.48E+7 V.M.Adamov+ /53/
gs 32625.002 5.00E+5 - 1.00E+6 W.G.Yan+ /54/
s 20446.002 5.00E+3 - 3.00E+4 S.Nizamuddin+ /55/
g 20003.005 6.00E+2 - 3.00E+3 M.G.Cao+ /56/
gs 10056.002 6.00E+2 - 9.87E+3 D.W.Bergen /57/
s 30035.003 1.41E+7 R.H.Iyer+ /58/
s 10267.041 7.50E+3 - 8.50E+4 R.Gwin+ /59/
g 10056.002 6.41E+4 - 2.85E+6 D.W.Bergen /57/
g 12360.002 6.00E+2 - 9.78E+5 D.W.Bergen+ /60/
g 21463.002 4.00E+4 - 5.05E+5 P.H.White+ /61/
g 40650.002 2.80E+5 - 2.63E+6 G.N.Smirenkin+ /62/
g 12341.002 6.13E+2 - 9.60E+2 M.S.Moore+ /63/
--------------------------------------------------------------
Ratio to U-235(n,f) cross section
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
gs 41455.002 5.77E+5 - 9.75E+6 O.A.Shcherbakov+ /64/
gs 41455.002 1.01E+7 - 1.94E+7 O.A.Shcherbakov+ /64/
gs 13134.004 1.47E+7 J.W.Meadows+ /65/
g 41432.003 2.00E+4 - 6.38E+6 D.L.Shpak /66/
gs 22014.003 4.90E+5 - 6.97E+6 K.Kanda+ /67/
gs 40607.002 6.42E+5 - 8.25E+5 D.L.Shpak+ /68/
gs 40474.002 2.40E+4 - 7.40E+6 B.I.Fursov+ /69/
g 40474.002 1.27E+5 - 7.00E+6 B.I.Fursov+ /69/
s 40361.003 1.50E+4 - 1.94E+6 D.L.Shpak+ /70/
gs 10236.002 1.42E+5 - 9.37E+6 J.W.Meadows /71/
g 10562.003 8.50E+2 - 1.95E+7 G.W.Carlson+ /72/
gs 20363.002 5.20E+3 - 1.01E+6 E.Pfletschinger+ /73/
g 10084.003 6.60E+2 - 2.40E+4 W.K.Lehto+ /74/
g 40309.003 4.85E+5 - 2.51E+6 V.G.Nesterov+ /75/
g 40027.004 3.30E+5 - 2.58E+6 G.N.Smirenkin+ /76/
s 22282.003 1.35E+7 - 1.49E+7 F.Manabe+ /45/
--------------------------------------------------------------
The obtained cross section in the energy range from 1 to 4
MeV and from 7 to 8 MeV was slightly modified 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).
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 neutron spectrum
Below 5 MeV, data of JENDL-3.3/77/ were adopted.
Comment of JENDL-3.3:
* Distributions were calculated with the modified Madland-Nix
model/78,79/. The compound nucleus formation cross
sections for fission fragments (FF) were calculated using
Becchetti-Greenlees potential/80/. Up to 4th-chance-fission
were considered at high incident neuttron energies.
The Ignatyuk formula/81/ were used to generate the level
density parameters.
Parameters adopted:
Total average FF kinetic energy = 172.311-0.0212*E(MeV)
Average energy release = 188.438 MeV
Average mass number of light FF = 95
Average mass number of heavy FF = 139
Level density of the light FF = 9.999- 10.094
Level density of the heavy FF = 11.89 - 12.20
Note that the parameters vary with the incident energy
within the indicated range.
Above 5.5 MeV, the distributions were calculated with CCONE
code/31/.
MT=455 Delayed neutron spectrum.
Summation calculation made by Brady and England/9/ was
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./82/ for
U-235 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
Nu-d at 0.0253eV was estimated as 0.00067+-0.0002 from
experimental data/2,3,4/. Error of 3% was adopted.
Above 100 eV error of 8% was assumed.
MT=456
Covariance of obtained by fitting a stlight line to
experimental data (See MF1,MT456).
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/84/ 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
Below 600 eV, covariance matrices was calculated from those
of resonance parameters/83/.
Above 600 keV, Covariance matrix was obtained with CCONE and
KALMAN codes/84/.
MT=18 Fission cross section
Below 600 eV, covariance matrices was calculated from those
of resonance parameters/83/.
600 eV - 9 keV, covariances were obtained by GMA code.
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 600 eV, covariance matrices was calculated from those
of resonance parameters/83/.
Above 600 keV, Covariance matrix was obtained with CCONE and
KALMAN codes/84/.
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/31/ 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/85/
* Global parametrization of Koning-Duijvestijn/86/
was used.
* Gamma emission channel/87/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/88/ 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/89/. 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/90/,/91/
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 (see Table 2)
* optical potential parameters /32/
Volume:
V_0 = 50.1895 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 = 1.25 fm
a_v = 0.63 fm
Surface:
W_0 = 16.2027 MeV
B_s = 11.19 MeV
C_s = 0.01361 1/MeV
C_wiso = 23.5 MeV
r_s = 1.1803 fm
a_s = 0.65 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 = 0.205125
beta_4 = 0.076
beta_6 = 0.0015
* Calculated strength function
S0= 0.92e-4 S1= 2.11e-4 R'= 9.64 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of U-233
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 5/2 + *
1 0.04035 7/2 + *
2 0.09216 9/2 + *
3 0.15523 11/2 + *
4 0.19700 7/2 +
5 0.22947 13/2 +
6 0.29881 5/2 -
7 0.30194 13/2 -
8 0.31190 3/2 +
9 0.31460 15/2 +
10 0.32083 7/2 -
11 0.33042 7/2 +
12 0.34048 5/2 +
13 0.35379 9/2 -
14 0.38043 7/2 +
15 0.39756 11/2 -
16 0.39850 1/2 +
17 0.41117 17/2 +
18 0.41576 3/2 +
19 0.42500 17/2 +
20 0.43200 9/2 +
21 0.45611 5/2 +
22 0.49700 11/2 +
23 0.50362 7/2 -
24 0.51755 19/2 +
25 0.52200 15/2 -
26 0.54654 5/2 +
27 0.56160 9/2 -
28 0.56700 5/2 -
29 0.57200 1/2 -
30 0.57500 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
--------------------------------------------------------
U-234 18.4623 1.5689 2.5578 0.3903 -0.1513 3.9089
U-233 18.3945 0.7861 2.4694 0.3820 -0.8201 2.9898
U-232 18.3266 1.5757 2.6095 0.3888 -0.1142 3.8806
U-231 18.2588 0.7895 2.6793 0.3781 -0.7806 2.9485
U-230 18.1909 1.5825 2.6739 0.3937 -0.1509 3.9421
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
U-234 6.000 1.040 5.400 0.600
U-233 5.600 0.800 5.400 0.520
U-232 5.500 1.040 5.000 0.600
U-231 6.000 0.800 5.600 0.520
U-230 5.800 1.040 5.100 0.600
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-234 21.7235 1.8304 2.6000 0.3321 -0.7507 4.0304
U-233 20.2008 0.9172 2.6000 0.3667 -2.0299 3.4172
U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383
U-231 20.0518 0.9211 2.6000 0.3325 -1.4903 2.9211
U-230 19.9772 1.8463 2.6000 0.3332 -0.5651 3.8463
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-234 21.7235 1.8304 0.0600 0.3643 0.0626 3.9304
U-233 19.2990 0.9172 0.0200 0.4202 -1.2096 3.4172
U-232 18.3293 1.8383 -0.0200 0.4264 -0.2156 4.2383
U-231 20.0518 0.9211 -0.0600 0.3758 -0.7765 2.9211
U-230 19.9772 1.8463 -0.1000 0.3771 0.1493 3.8463
--------------------------------------------------------
Table 7. Gamma-ray strength function for U-234
--------------------------------------------------------
K0 = 1.600 E0 = 4.500 (MeV)
* E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 299.61 (mb)
ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)
* M1: ER = 6.65 (MeV) EG = 4.00 (MeV) SIG = 2.81 (mb)
* E2: ER = 10.22 (MeV) EG = 3.30 (MeV) SIG = 6.52 (mb)
--------------------------------------------------------
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Appendix A1: Resolved resonance parameters (ENDF/B-VII.0)
=========================================================
MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETER EVALUATIONS
L. Leal, H. Derrien, Guber, N. Larson, R. Wright,
and R.Spencer (ORNL)
January, 2003
The resonance parameter evaluation was done by Leal,
Derrien, Guber, Larson, Wright and Spencer [LE01] using the
multilevel R-matrix analysis code SAMMY [LA96]. The resolved
resonance evaluation were performed in the energy range from 0 to
600 eV. The unresolved resonance region evaluation covers the
energy range from 600 eV to 40 keV.
The evaluation included high resolution transmission data
(GU00), fission cross section data [GU98] measured at the Oak
Ridge Electron Linear Accelerator (ORELA), in addition to other
experimental data. Integral data were also included in the
evaluation. Integral quantities and thermal values calculated
with the U-233 resonance parameter are shown in the Table below.
Also shown are the results of calculations using U-233 ENDF/B-VI
evaluation and Axton standard values:
Quantity ENDF/B-VI Axton Present Eval
-------- --------- ----- ------------
Fission 531.14 +/- 1.33 530.70 +/- 1.34 530.70
Capture 45.51 +/- 0.68 45.52 +/- 0.70 45.22
Scattering 12.13 +/- 0.66 12.19 +/- 0.67 12.18
Westcott ga 0.9996 +/- 0.0011 0.9995 +/- 0.0011 1.00325
Westcott gf 0.9955 +/- 0.0014 0.9955 +/- 0.0014 1.00045
K1 (barn) 742.60 +/- 2.40 742.25 +/- 0.0040 746.0
The following Table shows the average values of the fission
and capture cross sections of the present evaluation compared
to the previous ENDF-B6 evaluation in the energy range thermal
to 600 eV.
Energy Range Fission Capture
(eV) Present ENDF/B-VI.5 Present ENDF/B-VI.5
-------------- -------- ----------- ------ -----------
0.001 -0.020 980.19 971.62 82.03 83.27
0.020 -0.050 462.03 460.02 39.84 40.21
0.050 -0.400 201.88 202.62 20.86 20.36
0.400 - 1.00 127.29 126.91 11.93 9.95
1.0 - 2.10 389.14 378.56 66.35 67.37
2.10 - 2.75 206.76 198.02 111.45 112.00
2.75 - 3.00 49.84 50.46 7.91 7.48
3.00 - 15.0 104.25 101.26 17.66 17.65
15.0 - 30.0 94.72 91.80 13.51 13.27
30.0 - 50.0 40.72 38.85 5.66 5.46
50.0 - 75.0 41.24 41.21 5.69 4.61
75.0 - 100 36.92 33.72 8.94 4.35
100 - 125 38.24 29.94 6.10 3.88
125 - 150 21.11 22.10 3.72 3.54
150 - 200 20.99 21.34 3.06 3.18
200 - 300 23.10 19.87 3.51 2.72
300 - 400 18.28 16.66 2.45 2.33
400 - 500 11.06 13.17 1.39 2.08
500 - 600 13.52 13.40 2.00 1.90
-----------------------------------------------------------------
The fission and capture resonance integral calculated from
the present evaluation are 776.64 b and 139.66 b, respectively,
which compare to 760 +/ 17 b and 137 +/- 6 b reported by
Mughabghab.[MU85]
----- REFERENCES (MF=2) -----
GU98 K. H. Guber et al., Nuc. Sci. Eng. 135, 1(2000).
GU00 K. H. Guber et al., to be published in the Nuc. Sci. Eng.
KA85 K. Kanda et al, Measurement of Fast Neutron Induced Fission
Cross Sections of 232-Th, 233-U, and 234-U Relative to
235-U, Nuclear Data for Basic and Applied Science, Vol. 1,
Santa Fe, New Mexico (May 1985).
LA98 N. M. Lasrson, Updated User Guide for SAMMY: Multilevel R-
Matrix Fits to Neutron Data Using Bayes' Equations,
ORNL/TM-9179/R4 (December 1998). See also ORNL/TM-9170/R5.
LE01 L. C. Leal, H. Derrien, J. A. Harvey, K. H. Guber, N. M.
Larson and R. R. Spencer, R-Matrix Resonance Analysis and
Statistical Properties of the Resonance Parameters of U-233
in the Neutron Energy Range from Thermal to 600 eV,
ORNL/TM-2000/372, March 2001.
LE96 L. C. Leal and R. Q. Wright, Assessment of the Available
233-U Cross Section Evaluations in the Calculation of
Critical Benchmark Experiments, ORNL/TM-13313R (Sept 1996).
MU85 S. F. Mughabghab, Neutron Cross Sections, Vol. I. Part B
(1985).