92-U -235
92-U -235 JAEA+ EVAL-OCT09 O.Iwamoto,N.Otuka,S.Chiba,+
DIST-SEP12 20111206
----JENDL-4.0u1 MATERIAL 9228
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
Update File Distribution
Sep.14,2012 JENDL-4.0u1
History
07-07 Calculation with CCONE code was performed.
07-09 Fission spectra up to 5 MeV were replaced with JENDL-3.3.
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. New CCONE calculation
was adopted.
08-02 Fission and capture cross sections, and nu-p were revised.
CCONE calculation was made with revised parameters.
Data were compiled as JENDL/AC-2008/1/.
09-08 (MF1,MT458) was evaluated.
09-10 nu-p and fission cross section were revised.
10-03 Covarinace data were given.
11-07 Covariance data in RRR were revised.
MF= 1
MT=452 Total number of neutrons per fission
Sum of MT=455 and 456.
MT=455 Delayed neutron data
(same as JENDL-3.3/2/)
Evaluated by using the least-squares method on the basis of
the following experimental data in each energy region.
Thermal region: Keepin/3/, Conant/4/, Synetos/5/,
Reeder/6/, Borzakov/7/
50 keV - 7 MeV: Keepin/3/, Maksyutenko/8/, Masters/9/,
Krick/10/, Evans/11/, Cox/12/,
Besant/13/, Gudkov/14/, Loaiza/15/
14 - 15 MeV : Keepin/16/
Decay constants at the thermal energy were adopted from
Keepin et al./17/
MT=456 Number of prompt neutrons per fission
Below 500 eV, JENDL-3.3 was adopted, which was evaluated on
the basis of experimental data of Gwin et al./18,19/
Above 500 eV, experimental data were anlyzed by the GMA code
/20/ with Chiba-Smith approach/21/ for PPP minimization.
Experimental data are renormalized with nu-p of CF-252
spontaneous fission (3.756+/-0.031) reported by Vorobyev et
al./22/ if standards to derive original data were known.
Experimental data sets are summarized below.
r: re-normalized by nu-p(252Cf spon) of A.S.Vorobyev et al.
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
r 12326.004 2.80E+5 - 1.45E+7 J.C.Hopkins+ /23/
r 12870.004 1.70E+7 - 1.96E+7 R.E.Howe /24/
r 13101.003 5.00E+2 - 9.00E+6 R.Gwin+ /25/
r 20427.002 2.25E+5 - 1.36E+6 F.Kaeppeler+ /26/
21696.004 2.50E+6 - 1.41E+7 I.Johnstone /27/
r 21785.003 1.14E+6 - 1.47E+7 J.Frehaut+ /28/
r 40262.002 8.60E+5 - 5.35E+6 M.V.Savin+ /29/
r 40493.002 1.98E+5 - 9.85E+5 M.V.Savin+ /30/
40785.002 1.43E+7 Ju.A.Vasilev+ /31/
--------------------------------------------------------------
In the energy region from 1 keV to 1 MeV, GMA results were
increased by 0.2%.
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/32/. 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 2000/33/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/34/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/35/ 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: 1.0E-5 - 500 eV)
Adopted are parameters for Reich-Moore formula evaluated by
Leal et al./36/ In the present file, the upper boundary of
resolved resonance region is set to 500 eV.
See Appendix A-1
Unresolved resonance parameters (500 eV - 30 keV)
Unresolved resonance parameters were determined with ASREP
code/37/ so as to reproduce the cross sections.
The parameters are used only for calculation of self-shielding
factors.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ.(*)
(barns) (barns)
-------------------------------------------------------
total 698.90
elastic 15.12
fission 585.08 274
capture 98.71 139
-------------------------------------------------------
(*) In the energy range from 0.5 eV to 10 MeV.
MF= 3 Neutron cross sections
Between 500 eV and 2.25 keV:
Cross sections were calculated with resonance parameters of
JENDL-3.3 which were taken from ENDF/B-VI.5/36/ and
broadened with a resolution function of R(E)=0.03*E.
The capture cross section was multiplied by ratios of average
capture cross sections of JENDL-3.2 and JENDL-3.3 to lower the
cross sections to those of JENDL-3.2.
Above 2.25 keV:
Cross sections except for the total (M=1), elastic scattering
(MT=2), fission (MT=18, 19, 20, 21, 38) and capture cross
sections were calculated with CCONE code/38/.
The model parameters were determined by considering integral
experimental data as well as measured cross-section data.
MT= 1 Total cross section
In the energy range from 2.25 to 500 keV, cross section was
calculated with CCONE code/38/.
Above 500 keV, the cross section was determined by spline
fitting to experimental data of Schwartz et al./39/, Poenitz
et al./40,41/, Harvey et al./42/, Cabe and Cance/43/, and
Uttley et al./44/ These experimental data were used also
for adjustment of the OMP of Soukhovitskii et al./45/
MT=2 Elastic scattering cross section
Calculated as total - non-elstic scattering cross section
MT=16 (n,2n) cross section
Calculated with CCONE code. The experimental data of Frehaut
et al./46/ were considered to determine the model parameters
of CCONE calculation.
MT=18 Fission cross sections
From 2.25 keV to 10 keV, JENDL-3.3/2/ was adopted. The data
of JENDL-3.3 were based on the experimental data of Weston
and Todd/47/.
Above 10 keV, experimental data measured after 1980 were
analyzed by simultaneous fitting of U-233, U-235, U-238,
Pu-239, Pu-240 and Pu-241 fission cross section and its ratio
by the SOK code /48/.
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
41112.002 1.88E+6 - 2.37E+6 V.A.Kalinin+ /49/
22304.006 2.60E+6 - 1.47E+7 K.Merla+ /50/
22304.002 4.45E+6 - 1.88E+7 K.Merla+ /50/
12924.002 1.07E+6 - 5.99E+6 R.G.Johnson+ /51/
40969.011 6.24E+5 - 7.85E+5 N.N.Buleeva+ /52/
22091.002 1.35E+7 - 1.49E+7 T.Iwasaki+ /53/
30721.002 1.42E+7 J.W.Li+ /54/
12877.004 5.05E+3 - 2.05E+5 L.W.Weston+ /47/
10987.002 3.10E+5 - 2.82E+6 A.D.Carlson+ /55/
30634.002 1.47E+7 J.W.Li+ /56/
12826.002 1.46E+7 M.Mahdavi+ /57/
10971.002 1.41E+7 O.A.Wasson+ /58/
21620.002 2.50E+6 - 4.45E+6 M.Cance+ /59/
21777.002 5.40E+3 - 8.25E+4 F.Corvi+ /60/
10950.002 2.45E+5 - 1.20E+6 O.A.Wasson+ /61/
40601.002 6.00E+3 - 4.50E+4 A.A.Bergman+ /62/
-----.--- 3.03E+6 - 2.96E+7 A.D.Carlson+ /63/
--------------------------------------------------------------
The obtained cross section from 1 to 4 MeV and from 7 to 8
MeV was slightly modified for JENDL-4.
The fission cross section of JENDL-3.3 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
From 2.25 keV to 1 MeV, experimental data measured after 1970
were analyzed by the GMA code/20/ with the Chiba and Smith
approach/21/ for PPP minimization/64/. All experimental
data are given in the form of alpha-value (=ratios to the
U-235(n,f) cross section), which were normalized to absolute
cross section by the JENDL-3.3 U-235(n,f) cross section. Data
points of Kononov et al./65/ below 20 keV were not
considered because their systematic error was very large in
this energy region.
Experimental data sets are summarized below.
--------------------------------------------------------------
EXFOR Energy range (eV) Authors Reference
--------------------------------------------------------------
12424.003 3.47E+3 - 2.56E+4 J.B.Czirr+ /66/
20158.002 8.50E+3 - 5.50E+4 R.E.Bandl+ /67/
20880.002 1.06E+4 - 1.96E+5 H.Beer+ /68/
20880.003 1.04E+5 - 3.07E+5 H.Beer+ /68/
20880.004 1.70E+4 - 6.40E+4 H.Beer+ /68/
20880.005 3.79E+5 - 4.81E+5 H.Beer+ /68/
21777.004 2.50E+3 - 8.25E+4 F.Corvi+ /60/
40412.002 2.04E+4 - 1.10E+6 V.N.Kononov+ /65/
40502.002 3.49E+3 - 8.89E+3 Ju.V.Ryabov /69/
40581.002 2.50E+3 - 4.50E+4 G.V.Muradyan+ /70/
40609.004 2.45E+4 V.P.Vertebnyy+ /71/
12409.003 2.00E+5 - 6.00E+5 G.de Saussure+ /72/
12407.002 1.23E+4 - 6.90E+5 L.W.Weston+ /73/
12331.005 3.00E+4 - 1.00E+6 J.C.Hopkins+ /74/
12416.002 1.78E+5 - 1.01E+6 B.C.Diven+ /75/
--------------------------------------------------------------
The results of the GMA were used to determine the parameters
in the CCONE calculation.
Above 1.3 MeV, results of the CCONE calculation were adopted.
The present results were slightly modified by considering
integral data.
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 spectra
Below 5 MeV, spectra given in JENDL-3.3/2/ were adopted.
Relevant part of JENDL-3.3 comments:
*DISTRIBUTIONS WERE CALCULATED WITH A MODIFIED MADLAND-NIX
MODEL WITH CONSIDERATION FOR MULTIMODAL NATURE OF THE FISSION
PROCESS/76,77/. THE COMPOUND NUCLEUS FORMATION CROSS SEC-
TIONS FOR FISSION FRAGMENTS WERE CALCULATED USING BECCHETTI-
GREENLEES POTENTIAL/78/. THE IGNATYUK FORMULA/79/ WERE
USED TO GENERATE THE LEVEL DENSITY PARAMETERS. UP TO
3rd-CHANCE-FISSION WERE CONSIDERED AT HIGH INCIDENT NEUTRON
ENERGIES.
PARAMETERS ADOPTED FOR THERMAL-NEUTRON FISSION:
(S1: STANDARD-1, S2: STANDARD-2, SL: SUPERLONG MODES)
TOTAL AVERAGE FRAGMENT KINETIC ENERGY
= 187 MEV FOR S1
= 167 MEV FOR S2
= 157 MEV FOR SL
AVERAGE ENERGY RELEASE
= 194.49 MEV FOR S1
= 184.86 MEV FOR S2
= 190.95 MEV FOR SL
AVERAGE MASS NUMBER OF LIGHT FF = 96
AVERAGE MASS NUMBER OF HEAVY FF = 140
LEVEL DENSITY OF THE LIGHT FF = 10.31(S2), 11.43(S1)
LEVEL DENSITY OF THE HEAVY FF = 8.89(S1), 13.25(S2)
MODE BRANCHING RATIO = 0.18342(S1), 0.81589(S2),
0.00069(SL)
NOTE THAT THE PARAMETERS VARY WITH THE INCIDENT ENERGY
WITHIN THE INDICATED RANGE.
Above 5.5 MeV, calculated with CCONE code/38/.
MT=455 Delayed neutron spectra
(same as JENDL-3.3)
Taken from Brady and England/80/. Group abundace parameters
were adjusted so as to reproduce total delayed neutron
emission rate measured by Keepin/17/, Piksaikin/81/ and
East/82/.
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF=12 Photon Production Multiplicities (option 1)
MT=18 Fission
(same as JENDL-3.3)
The thermal neutron-induced fission gamma spectrum measured
by Verbinski et al./83/ was adopted.
MF=14 Photon Angular Distributions
MT=18
(same as JENDL-3.3)
Isotropic distributions were assumed.
MF=15 Continuous Photon Energy Spectra
MT=18
(same as JENDL-3.3)
Experimental data by Verbinski et al./83/ 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
(Same as JENDL-3.3/2/)
MT=456
Below 500 eV, the covariance of JENDL-3.3 was adopted.
Above 500 eV, it was obtained by fitting to the experimental
data described above. The error of nu-p was multiplied by
a factor of 2.0.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/85/ 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 500 eV, covariance matrices was calculated from those
of resonance parameters/84/.
In the energy range from 200 to 500 eV, uncertainty of 2%
was assumed for the total cross section, and 4% for the
elastic scattering. From 0.5 to 2.25 keV, uncertainty of 5%
was assumed. Above 2.25 keV, covariance of the CCONE
calculation was adopted.
MT=18 Fission cross section
Below 500 eV, covariance matrices was calculated from those
of resonance parameters/84/.
In the energy range from 200 to 500 eV, uncertainty of 2%
was assumed. From 500 eV to 9 keV, uncertainty was assumed
to be 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 500 eV, covariance matrices was calculated from those
of resonance parameters/84/.
In the energy range from 200 to 500 eV, uncertainty of 3%
was assumed.
From 500 eV to 2.25 keV, uncertainty was assumed to be 10%.
In the energy region from 2.25 keV to 1 MeV, capture cross
section was obtained with GMA code/64/. Its covariance matrix
was obtained simultaneously.
Above 1 MeV, covariance of the CCONE calculation was adopted.
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/38/ 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/86/
* Global parametrization of Koning-Duijvestijn/87/
was used.
* Gamma emission channel/88/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/89/ 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/90/. 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/91/,/92/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,3,6,9 (see Table 2)
* optical potential parameters /45/
Volume:
V_0 = 49.8613 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 = 0.61 fm
Surface:
W_0 = 17.1117 MeV
B_s = 11.19 MeV
C_s = 0.01361 1/MeV
C_wiso = 23.5 MeV
r_s = 1.1803 fm
a_s = 0.61 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 = 0.201044
beta_4 = 0.11
beta_6 = 0.0015
* Calculated strength function
S0= 0.93e-4 S1= 2.11e-4 R'= 9.50 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of U-235
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 7/2 - *
1 0.00008 1/2 +
2 0.01304 3/2 +
3 0.04621 9/2 - *
4 0.05171 5/2 +
5 0.08174 7/2 +
6 0.10304 11/2 - *
7 0.12930 5/2 +
8 0.15047 9/2 +
9 0.17071 13/2 - *
10 0.17139 7/2 +
11 0.19712 11/2 +
12 0.22542 9/2 +
13 0.24913 15/2 -
14 0.25900 7/2 +
15 0.29114 11/2 +
16 0.29467 13/2 +
17 0.33285 5/2 +
18 0.33852 17/2 -
19 0.35730 15/2 +
20 0.36707 7/2 +
21 0.36900 13/2 +
22 0.39322 3/2 +
23 0.41478 9/2 +
24 0.42675 5/2 +
25 0.43860 19/2 -
26 0.44572 7/2 +
27 0.45450 15/2 +
28 0.47130 11/2 +
29 0.47430 7/2 +
30 0.48500 17/2 +
31 0.50992 9/2 +
32 0.53240 13/2 +
33 0.53323 9/2 +
34 0.55040 21/2 -
35 0.56800 19/2 +
36 0.58780 11/2 +
37 0.60808 11/2 +
38 0.61640 15/2 +
39 0.63317 5/2 -
40 0.63781 3/2 -
-------------------
*) 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-236 18.6157 1.5623 2.7551 0.3856 -0.1537 3.8909
U-235 18.4419 0.7828 2.6265 0.3721 -0.7434 2.8828
U-234 18.4800 1.5689 2.5578 0.3899 -0.1502 3.9076
U-233 18.4122 0.7861 2.4694 0.3817 -0.8188 2.9881
U-232 18.3442 1.5757 2.6095 0.3885 -0.1133 3.8795
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
U-236 6.201 1.040 5.417 0.550
U-235 5.700 0.400 5.600 0.300
U-234 6.050 1.040 5.400 0.600
U-233 5.970 0.800 5.450 0.520
U-232 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-236 20.9712 1.8226 2.6000 0.3244 -0.5720 3.8226
U-235 20.3497 0.9133 2.6000 0.3515 -1.8195 3.2133
U-234 20.2753 1.8304 2.6000 0.3522 -0.9023 4.1304
U-233 20.2008 0.9172 2.6000 0.3312 -1.4942 2.9172
U-232 20.1263 1.8383 2.6000 0.3319 -0.5731 3.8383
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
U-236 21.5182 1.8226 0.1400 0.3654 0.0498 3.9226
U-235 20.3497 0.9133 0.1000 0.4065 -1.2054 3.4133
U-234 20.2753 1.8304 0.0600 0.3939 -0.1191 4.1304
U-233 20.2008 0.9172 0.0200 0.3732 -0.7817 2.9172
U-232 20.1263 1.8383 -0.0200 0.3745 0.1401 3.8383
--------------------------------------------------------
Table 7. Gamma-ray strength function for U-236
--------------------------------------------------------
K0 = 1.500 E0 = 4.500 (MeV)
* E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 303.18 (mb)
ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)
* M1: ER = 6.63 (MeV) EG = 4.00 (MeV) SIG = 2.69 (mb)
* E2: ER = 10.19 (MeV) EG = 3.28 (MeV) SIG = 6.51 (mb)
--------------------------------------------------------
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*****************************************************************
Appendix A-1 from ENDF/B-VI.8
*****************************************************************
File 2
MT=151 Resonance parameters, from a new analysis by Leal
et al. [LE97], using the multilevel R-matrix analysis code
SAMMY [LA96]. Energy range for U235 is 0 to 2.25 keV.
For the first time, integral data were fitted during the
analysis process: Thermal cross sections (fission,
capture, and elastic), Westcott g-factors (fission and
absorption) are from the ENDF/B-6 standards [CA93], and
the K1 value is from Hardy [HA79].
Thermal parameters obtained in the present evaluation,
first using the microscopic experimental data only, and
second including the integral data as well, are compared
to the SAMMY input in the following Tabld:
Parameter SAMMY input Fit to Fit to diff.
value diff data & integ.
alone data
----------- ----------------- -------- ------------
Fission 584.25 +/- 1.11 582.28 584.88
Capture 98.96 +/- 0.74 99.18 98.66
Scattering 15.46 +/- 1.06 15.44 15.12
Westcott gf 0.9771 +/- 0.0008 0.9743 0.9764
Westcott ga 0.9790 +/- 0.0008 0.9774 0.9785
Westcott gg 0.9956 0.9910
K1(barn) 722.70 +/- 3.90 717.48 722.43
The final adjustment of nu by SAMMY to the recommended K1
value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission
and absorption cross sections calculated from the final
resonance parameters.
In the following Tables, the fission and capture cross
sections calculated in this evaluation with the code
SAMMY are compared with experimental data.
Experimental and calculated total cross sections.
Energy Range Calculated Schrack Weston Weston
(eV) (b.eV) (b.eV) (b.eV) (b.eV)
--------------- ---------- ------- ------ ------
0.5 - 20.0 910.4 929.9
20.0 - 60.0 1867.8 1882.8 1869.9
60.0 - 100.0 954.0 968.0 954.2
100.0 - 200.0 2032.7 2092.7 2089.5 2073.9
200.0 - 300.0 2062.2 2007.0 2060.0 2054.6
300.0 - 400.0 1280.8 1321.6 1297.1 1292.9
400.0 - 500.0 1333.2 1391.5 1351.8 1347.9
500.0 - 600.0 1489.2 1467.9 1499.2 1494.3
600.0 - 700.0 1126.6 1156.4 1134.1 1132.6
700.0 - 800.0 1088.7 1085.8 1093.3 1075.7
800.0 - 900.0 797.6 784.0 813.0 804.9
900.0 - 1000.0 724.4 723.9 738.2 721.4
1000.0 - 2000.0 7036.1 7054.2
Experimental and calculated capture cross sections.
Energy Range Calculated De Saussure Perez
(eV) (b.eV) (b.eV) (b.eV)
--------------- ---------- ----------- ------
0.5 - 20.0 653.5 647
20.0 - 60.0 1066.1 1084 1057
60.0 - 100.0 490.2 477 504
100.0 - 200.0 1158.8 1148 1138
200.0 - 300.0 907.8 904 940
300.0 - 400.0 660.2 658 642
400.0 - 500.0 495.9 506 478
500.0 - 600.0 533.3 506 562
600.0 - 700.0 494.8 481 449
700.0 - 800.0 490.1 513 475
800.0 - 900.0 439.8 444 397
900.0 - 1000.0 504.2 542 482
1000.0 - 1100.0 509.6 522 463
1100.0 - 1200.0 413.7 395 332
1200.0 - 1300.0 340.4 372 267
1300.0 - 1400.0 304.1 304 225
1400.0 - 1500.0 355.7 301 254
--------------- ---------- ----------- ------
20.0 - 1500.0 9164.7 9046 8665
The fission and capture resonance integral calculated from
the present evaluation are 276.04 b and 140.49 b, respectively,
giving a capture-to-fission ratio (alpha value) of 0.509 in
excellent agreement with the value obtained from integral
measurements.
The following energy-differential data were included in the
analysis:
(1) Transmission data of Harvey et al. [HA86] on the ORELA
18-meter flight path, with sample thickness of 0.03269
atoms/barn, cooled to 77 K (0.4 to 68 eV).
(2) Transmission data of Harvey et al. [HA86] on the ORELA
80-meter flight path, with sample thickness of 0.00233
atoms/barn, cooled to 77 K (4 to 2250 eV).
(3) Transmission data of Harvey et al. [HA86] on the ORELA
80-meter flight path, with sample thickness of 0.03269
atoms/barn, cooled to 77 K (4 to 2250 eV).
(4) Fission data of Schrack [SC88] on the RPI Linac at 8.4
meter flight path (0.02 to 20 eV).
(5,6) Fission and capture data of de Saussure et al. [DE67]
on the ORELA 25.2-meter flight path (0.01 to 2250 eV).
(7,8) Fission and capture data of Perez et al. [PE73] on the
ORELA 39-meter flight path (0.01 to 100 eV).
(9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter
flight path (0.01 to 20 eV).
(10) Transmission data of Spencer et al. [SP84] on the ORELA
ORELA 18-meter flight path, sample thickness of 0.001468
atom/barn (0.01 to 1.0 eV).
(11) Fission data of Wagemans et al. [WA88] on the Geel 18-
meter flight path (0.001 to 1.0 eV)
(12,13) Absorption and fission data of Gwin [GW96] at ORELA
(0.01 to 4.0 eV).
(14) Fission data of Weston and Todd [WE84] on the ORELA
18.9-meter flight path (14 to 2250 eV).
(15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to
1.0 eV).
(16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to
0.15 eV).
(17) Fission data of Weston and Todd [WE92] on the ORELA
86.5-meter flight path (100 to 2000 eV).
(18) Fission yield data of Moxon et al. [MO92] at ORELA
(0.01 to 50.0 eV).
----------------------------------------------------------------
REFERENCES FOR RESOLVED RESONANCE REGION
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Neutron Cross Section Measurements Standards," National
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(1993)
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"Simultaneous Measurements of the Neutron Fission and Capture
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