Back
94-Pu-239 BRC,CAD,+ EVAL-SEP06 ROMAIN, MORILLON, DOSSANTOS
DIST-JAN09 20090105
----JEFF-311 MATERIAL 9437
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
*************************** JEFF-3.1.1 *************************
** **
** Original data taken from: JEFF-3.1 Updated **
** Modification: New MF1/MT456 & MF2/MT18,102 **
******************************************************************
***************************** JEFF-3.1 *************************
** **
** Original data taken from: JEFF-3.0 (slight changes) **
** **
******************************************************************
2006-09 Bernard, Fort, Santamarina, Noguere, Courcelle (CEA)
Slight modifications in the subthermal-thermal range:
1) MF1: MT452 and MT456 from E. Fort and A. Courcelle
from 1meV to 23eV:
[Proceeding of WONDER'06 Workshop]
[To be published at ND2007 Conference]
to reduce the systematic keff overestimation:
* of MISTRAL-MOx mock-up [JEF/DOC-1143]
* of ICSBEP/PU-SOL-THERM [JEF/DOC-1107]
2) MF2: MT151 from D. Bernard and A. Santamarina:
1 Bound level added to reduce the overestimation of
the Isothermal Temperature Coefficient
in MOx lattices [NSE 144,47-74 (2003)].
This implies slight changes on thermal values,
consistent with differential uncertainties (see
Mughabghab, BNL 2006):
* thermal capture XS change: +2.1b
* thermal fission XS change: -0.8b
* thermal total XS change: +1.2b
2005-06 E. Dupont (CEA) and Ch. Dean (Serco Group)
Unresolved Resonance Parameters (MF2,MT151,LRU=2)
Extension of average parameters up to EH = 30 keV
2005-01 NEA/OECD (Rugama) 8 delayed neutron groups, Jefdoc-976,
Spriggs, Campbel, Piksaikin, Prog Nucl Ener 41,223(2002)
2003-06 CAD (Dupont) Unresolved Resonance Parameters
(MF=2,MT=151,LRU=2) for L=1 and AJ=1.0 resonances:
AMUN changed from 1. to 2. AND GN0 divided by 2.
***************************** JEFF-3.0 ***********************
NEW evaluation
This evaluation is built from contributions of several individuals
in various laboratories.
** BRC : J.P. Delaroche, P. Dossantos-Uzarralde, S. Hilaire,
C. Le luel, M. Lopez-Jimenez, P. Morel, B. Morillon,
P. Romain.
** CAD : E. Dupont, E. Fort, O. Serot, J-Ch Sublet.
** + : H. Derrien, T. Nakagawa.
***************************************************************
MF=1 Descriptive and Nubar Information ************************
***************************************************************
MT=452: Number of neutrons per fission
Total Nubar. Sum of MT=455 and 456.
MT=455: Delayed nubar evaluation (from WPEC/SG 6)
See JEF/DOC-920
Energy dependent delayed neutron spectrum introduced
MT=456: Prompt nubar evaluation (E.Fort and B.Morillon)
The evaluation below 650 eV is based on experimental
data [30].
From 650 eV to 30 MeV, the adopted values are obtained
from the Los Alamos model upgraded by G.Vladuca and
A.Tudora (multiple fission chances included) [31].
The model parameters are slightly different from those
adopted in [31].
MT=458: Energy release due to fission (O.Serot and B.Morillon).
The kinetic energy of the fragments results from a
compilation of recent experimental measurements.
The kinetic energy of the prompt fission neutrons is
consistent with the nup-value and the average energy
deduced from prompt fission spectra. The energy
released by the emission of prompt gamma rays is
obtained from the systematics proposed by Frehaut.
The components of the prompt energy release are
consistent with the nup calculations performed with
a model similar to Madland's. (JEFDOC xxx)
Actually they are all energy dependent but ENDF format
does not allow for such representation.
******************************************************************
MF=2
*******************************************
PU239 RESONANCE DATA 0 keV TO 2.5 keV
Principal evaluators: H.Derrien, T.Nakagawa
*******************************************
Resonance region evaluation by H.Derrien and T. Nakagawa
discussed below. This evaluation extended resonance region to
2.5 keV.
The present file contains the resonance parameters obtained
from a SAMMY fit analysis of high resolution experimental data,
performed at ORNL (Oak Ridge Nationnal Laboratory, USA) by
H.Derrien and G.De Saussure and at JAERI (Tokai-Mura Research
Establishment, Japan) by H.Derrien and T.Nakagawa.
The file contains three independant sections:
1) the first corresponds to the energy range 0 keV to 1 keV.
The corresponding set of resonance parameters contains 398
resonances in the energy range 0 keV to 1 keV, 4 ficticious
negative energy resonances and 3 ficticious resonances above
1 keV;
2) the second corresponds to the energy range 1 keV to 2 keV.
The corresponding set of resonance parameters contains 435
resonances in the energy range 0.980 keV to 2.02 keV, 3
ficticious resonances below 0.9 keV and 3 ficticious resonances
above 2.02 keV;
3) the third corresponds to the energy range 2 keV to 2.5 keV.
The corresponding set of resonance parameters contains 218
resonances in the energy range 1.98 keV to 2.53 keV, 3
ficticious resonances below 1.98 keV and 3 ficticious resonances
above 2.53 keV.
In all sections the ficticious resonance parameters take
into account the contribution of all the external truncated
resonances in such a way that no total, scattering, fission and
capture smooth files are needed in the corresponding energy
ranges for the reproduction of the cross sections within the
experimental errors.
The following experimental data base has been used in the
SAMMY fits:
- absorption and fission from R. Gwin et al. [1,4];
- fission from R. Gwin et al. [5,7], J. Blons [3], L.W. Weston
et al. [8,15];
- transmission from R.R. Spencer et al. [10], J.A. Harvey et al.
[9].
Prior to the fits the experimental fission and absorption cross
sections were normalised,directly or indirectly to the 0.0253 eV
values obtained by the ENDF/B-VI standard evaluation group [11].
The transmission data were considered as accurate absolute
measurements (R.R.Spencer total cross section at 0.0253 eV is
1025.0 b in excellent agreement with the 1027.3 b standard value)
Details on the analysis are found in [14],[16],[17].
----------------------------------------------------------------
COMMENTS ON THE THERMAL AND LOW ENERGY RANGES
The thermal cross-section values calculated at 293 K by the
resonance parameters of the first section are given in the
following table at 293 K and in barns.
SAMMY RESENDD Proposed standard [11]
------- -------- ----------------------
Fission 747.64 747.90 747.99+-1.87
Capture 271.10 270.73 271.43+-2.14
Scattering 7.97 7.99 7.88+-0.97
------- -------- ----------------------
Total 1026.71 1026.62 1027.30+-5.00
One should note that the 293 K cross sections calculated at
0.0253 eV depend on the way the Doppler broadening calculation
is performed. For instance using a Gaussian broadening function
will give a fission cross section about 2.5 barns larger than the
one obtained from the accurate calculation which conserves the
1/v shape of the thermal cross section. The values given in the
table above were obtained from SAMMY (Leal-Hwang method) [13,18]
and from RESENDD with 0.1% for the interpolation accuracy [20].
The following table shows experimental cross sections
averaged over the energy ranges 0.02 eV to 0.06eV and 0.02 eV
to 0.65 eV, compared to the calculated values:
References Average Cross Sections (barns)
[1-10] 0.02 - 0.06 eV 0.02 - 0.65 eV
---------------- ----------------------- -----------------------
Exp Calc (293K) Exp Calc (293K)
Gwin71 fiss 631.41 843.71
Gwin76 fiss 631.41 838.39
Gwin84 fiss(*) 631.41 631.75(+0.05%) 837.18 838.69(+0.18%)
Deruyter70 fiss 631.41 859.43
Wagemans80 fiss 631.41 862.56
Wagemans88 fiss 631.41 841.80
Gwin71 capture 243.84 243.22(-0.25%) 524.75 518.13(-1.26%)
Gwin76 absorpt(*) 875.90 874.29(-0.18%) 1359.96 1357.14(-0.21%)
Spencer84 tot(*) 883.20 882.86(-0.04%) 1361.69 1367.6 (+0.43%)
----------------- ---------------------- -----------------------
(*)These data had the largest weight in the thermal fit. The
values between the parentheses give the percentage deviation
between the calculated data and the experimental data.
The value of 631.4 barns for all the averaged experimental
fission cross sections in the energy range 0.02 eV to 0.06 eV
corresponds to the renormalisation of the fission experiments to
748.0+-1. barns at 0.0253 eV. ORNL data are consistent within
0.8% over the energy range 0.02 eV to 0.65 eV (i.e. over the 0.3
eV resonance). Deruyter70 and Wagemans80 data are about
2.5% larger and were not included in the SAMMY fit.
When normalised on the standard value at 0.0253 eV, Gwin 76
absorption agrees with the absorption obtained from Spencer total
cross section within 0.7% over the 0.3 eV resonance. The present
evaluation is essentially the result of a consistent SAMMY
analysis of all the available ORNL data with a larger weight on
Gwin 1984 fission, Gwin 1976 absorption and Spencer transmission
data.
After renormalisation of the calculated fission cross section
on the preliminary 1991 Weston and Todd fission data (see next
section) a slight adjustment of the negative resonance parameters
was performed to keep the values calculated at 0.0253 eV in close
agreement with the standard values. The 1988 data of Wagemans et
al.[21] agree within 0.4% with the calculated values over the
energy range from 0.02 eV to 0.65 eV after adjustment of the
energy scale to the ORNL scale (the difference was 0.27 eV at
20 eV between 1988 Wagemans and ORNL SAMMY fit energy scales).
----------------------------------------------------------------
COMMENTS ON THE 0 keV TO 1 keV ENERGY RANGE.
At the end of 1987, an analysis was completed up to 1 keV.
In a preliminary step, a correlated fit of Harvey transmission
data, Weston 84 fission data, and Blons fission data was
performed with possible adjustment of the normalisation
coefficients and of the background corrections. This preliminary
step has shown that this adjustment was not necessary to achieve
consistency between Harvey data and Weston data. The Blons data
needed a large readjustment of the background and normalisation.
Therefore, the final fit was performed only on the Harvey
transmission data, Gwin 84 fission data (below 30 eV), and
Weston 84 fission data, with no background and normalisation
adjustment. Blons data, which have better resolution than Weston
84 data, were used only to obtain more accurate fission widths
of some narrow resonances in the high energy range.
In 1989, preliminary results of the 1988 Weston fission
measurement [15] were included in the SAMMY experimental data
base. One expected from this measurement, which was performed by
using a 86-m flight path with a resolution comparable to that of
Harvey transmission, a confirmation of the excellent quality of
the 1984 measurement. A consistent SAMMY fit of Harvey transmis-
sion, Weston 84 fission and preliminary Weston 88 fission was re-
started from the parameter and covariance files obtained in 1987.
It appeared that large background and normalisation corrections
were needed on the new Weston fission data to obtain consistency
with Harvey transmission data. These corrections were comparable
to those found in Blons data and were not understood by the
authors of the experiment. The last SAMMY runs were performed by
not allowing background and normalisation variations on Harvey
transmission and Weston 84 fission (very small error bars were
assigned to the corresponding parameters in the covariance
matrix) and by allowing these variations on Weston 88 data. A
new set of resonance parameters was obtained, which was improved
compared to the previous set due to the very high resolution of
the new Weston fission measurement.
The calculated average fission cross section in the energy
range from 0.1 keV to 1.0 keV was 3.7% smaller than the values
obtained by the ENDF/B-VI standard evaluation group due to the
fact that Weston 84 data were 3.1% lower than the average
standard value. A new measurement was performed by Weston and
Todd in 1991 [22] in order to check their 1984 data. A careful
normalisation of the data in the thermal energy range showed
that the 1984 data should be renormalised by about +3%. To take
into account this renormalisation, the 1989 resonance parameters
were modified at JAERI [17] in the following way:
1) increase of the fission width by 3% and decrease of the
capture width by a quantity equal to the variation of the
fission width in the narrow resonances(mainly 1+ resonances);
that does not modify the total cross section in the correspond-
ing resonances;
2) adjustment of the neutron width of the 0+ resonances by a
refit of the transmission data and of the renormalised Weston
and Todd 1984 data in energy ranges where the contribution of
the 0+ resonances is dominant, and increase of the other(small)
0+ neutron widths by 3%. No severe inconsistency was observed
between the transmission data and the new fission data over the
dominant 0+ resonances; the differences between the 1989 fits of
the transmission and the new fits were consistent within the
experimental error bars.
The following table shows the fission cross sections calculat-
ed from the resonance parameters, the experimental values and
the results of the ENDF/B-VI standard evaluation group averaged
in the same energy intervals. Weston 1991 data are preliminary.
Weston 1984 data are normalised on preliminary Weston 1991.
Energy Cross Sections (barn)
(eV) Calc. Weston 1991 Weston 1984 Standard
---------- ------ ----------- ----------- --------
0.010-10. 80.12 79.98
9-20 94.74 94.91
20-40 17.52 17.76 17.97
40-60 50.64 50.90 50.87
60-100 54.42 54.38 54.33
100-200 18.63 18.59 18.56 18.66
200-300 17.85 17.89 17.88
300-400 8.31 8.34 8.43
400-500 9.59 9.58 9.57
---------- ------ ----------- ----------- --------
200-500 11.92 11.93 11.93 11.96
---------- ------ ----------- ----------- --------
500-600 15.39 15.57 15.86
600-700 4.37 4.30 4.46
700-800 5.51 5.53 5.63
800-900 4.84 4.89 4.98
900-1000 8.33 8.38 8.30
---------- ------ ----------- ----------- --------
500-1000 7.69 7.73 7.73 7.79
---------- ------ ----------- ----------- --------
20-1000 13.09 13.11 13.11
--------------------------------------------------------
Gwin 1971 and 1976 absorption data were not included in the
SAMMY fit in the energy range above 1 eV. Accurate absorption
cross sections should be calculated from the parameters obtained
from the analysis of the transmission and fission data. The
following table shows the calculated average values of the
capture, absorption and alpha compared to Gwin 1971 and Gwin
1976 data. The calculations were performed with RESENDD, 1%
accuracy.
Energy (eV) Cross Sections (barn)
calc. values (293 K) Gwin data
------------ --------------------- ------------------
CAPT ABSORP ALPHA ABSORP ALPHA
7.3- 16.0 76.61 196.04 0.64 208.00 0.74(*)
16.0- 37.5 20.51 44.55 0.85 46.50 0.89(*)
37.5- 50.0 48.72 70.00 2.29 83.15 2.96(*)
50.0-100.0 33.60 92.13 0.57 92.84 0.63
100.0-200.0 15.58 34.29 0.83 33.66 0.87
200.0-300.0 15.85 33.68 0.89 34.69 0.94
300.0-400.0 9.69 18.01 1.16 18.31 1.16
400.0-500.0 3.96 13.56 0.41 13.56 0.44
500.0-600.0 10.87 26.30 0.70 26.54 0.72
600.0-700.0 6.53 10.90 1.49 11.57 1.54
700.0-800.0 4.95 10.47 0.90 10.52 0.97
800.0-900.0 3.65 8.50 0.75 9.30 0.82
900.0-999.9 5.06 13.51 0.60 13.23 0.70
------------------------------------------------------
(*) Gwin 1971 data
If one excepts the energy range 37.5-50 eV, the calculated
absorption values agree well with Gwin experimental data; they
are on average 1.2% lower in the energy range from 50 eV to
1000 eV.
----------------------------------------------------------------
COMMENTS ON THE 1 keV TO 2 keV ENERGY RANGE
Preliminary resonance parameters were obtained in 1989 from
the analysis of the Harvey thick sample transmission data and of
the preliminary results of Weston 88 fission measurement. Due to
lack of time, the medium and thin sample transmission data were
not included in the SAMMY data base, and the contribution of the
truncated external resonances was not carefully investigated.
Nevertheless, the results were used in the ENDF/B-VI file, along
with a smooth file in order to agree with the average values of
a previous ENDF/B-VI evaluation (this preliminary set of
parameters was considered as more useful than the statistical
parameters in the energy range 1 keV to 2 keV for the
calculation of the self-shielding factors).
The analysis was restarted in April 1991 at JAERI with an
updated version of SAMMY adapted by T. Nakagawa to the FACOM 780.
The preliminary set of parameters obtained at Oak Ridge in 1989
was used as prior information to start the SAMMY calculations.
Also prior to the analysis, the contribution of the external
resonances was calculated by using the set of the 0 keV to 1 keV
known resonances, shifted in the energy ranges -1 keV to 0 keV,
2 keV to 3 keV, and 3 keV to 4 keV; equivalent contribution was
obtained by using 3 ficticious resonances below 1 keV and 3
ficticious resonances above 2 keV [17]. The analysis was
performed on the thick and medium sample transmissions of Harvey
(the thin sample data was not useful in the high energy range)
and on the 1988 fission data released by Weston at the beginning
of 1991 [15]. The definitive SAMMY fits were performed in April
1992 after renormalisation of the 1988 data of Weston to the
ENDF/B-VI standard values between 1 keV and 2 keV, in agreement
with the 1991 new measurements of Weston and Todd.
The average cross sections calculated from the resonance
parameters are compared to the experimental values in the
following table.
Energy Cross Sections (barn)
(keV) Total Fission Capture
-------- ---------------- ---------------- ---------------
CALC(a) EXP(b) CALC(a) EXP(c) CALC(a) EXP(d)
1.0-1.1 24.47 24.95 5.549 5.581 4.728 5.04
1.1-1.2 22.82 23.10 5.985 6.017 3.757 2.95
1.2-1.3 22.29 22.90 4.601 4.501 4.287 4.00
1.3-1.4 22.63 22.85 6.997 6.997 3.012 2.52
1.4-1.5 20.42 20.95 4.041 4.059 3.450 3.57
1.5-1.6 18.30 18.95 2.564 2.613 3.521 3.89
1.6-1.7 21.82 21.90 3.952 3.955 3.833 4.36
1.7-1.8 21.26 21.35 3.400 3.425 4.091 4.37
1.8-1.9 23.76 23.30 5.178 5.187 3.639 3.14
1.9-2.0 18.48 18.90 2.152 2.180 3.205 4.06
-------- ---------------- ---------------- ---------------
1.0-2.0 21.63 21.92 4.442 4.446 3.752 3.79
-------------------------------------------------------------
(a) total,fission and capture cross sections calculated by
RESEND from the resonance parameters.
(b) experimental total cross sections from Derrien [23].
(c) Weston and Todd 1988 high resolution fission cross sections
[15] normalised to ENDF/B-VI standard in the energy range
from 1.0 keV to 2.0 keV.
(d) Gwin 1971 experimental data normalised to Gwin 1976 data.
The difference of 1.3% between the average calculated total
cross section and the average experimental cross section in the
energy range from 1.0 keV and 2.0 keV is mainly due to the method
of evaluating the total cross section from the effective cross
section of Derrien [23]. The accuracy of the Sammy fit of the
experimental transmission data is better than 0.5% on the cross
section. The calculated fission cross sections are in very good
agreement with the experimental data. The capture data [1] are
average values obtained from the data available in the EXFOR
file and normalised to Gwin 1976 average values; there are large
differences between the calculated data and the experimental
data averaged over 0.1keV intervals; but on the interval from
1.0 keV to 2.0 keV the average values are consistent within 1.0%.
----------------------------------------------------------------
COMMENTS ON THE 2.0 keV TO 2.5 keV REGION
This energy range was also analysed at JAERI [17]. No
preliminary set of resonance parameters was available prior to
the analysis. More than 90% of the resonances, compared to the
low energy range, could still be identified in the transmission
data between 2 keV and 2.5 keV. Therefore, the correlated SAMMY
analysis of Harvey transmissions and Weston fission was still
feasible in this energy range. The resonance parameters obtained
are consistent and have nearly the same statistical properties
as those of the resonances in the 0 to 2 keV energy range. A
quite good fit of the transmission and fission data was obtained
without background and normalisation adjustment. However, the
calculated fission cross sections are, on average, 1.4% lower
than the experimental values. This difference, which however is
not larger than the systematic errors on the experimental data,
could be due to the difficulties of identifying the wide j=0+
resonances in the experimental data, because the effects of the
increasing resolution and Doppler widths. Prior to the SAMMY
fits, the fission data of Weston and Todd (1988 high resolution
data) were normalised to the ENDF/B-VI standard in the energy
range from 1 keV to 2 keV.
The cross sections, calculated from the resonance parameters
and averaged over 0.1 keV intervals, are given in the following
table.
Energy Cross Sections (barn)
(keV) TOTAL FISSION CAPTURE
--------- ---------------- ---------------- -------
CALC(a) EXP(b) CALC(a) EXP(c) CALC(a)
2.0-2.1 17.34 17.30 2.034 2.062 3.223
2.1-2.2 20.27 19.80 2.949 2.999 4.051
2.2-2.3 19.34 19.10 2.357 2.393 3.324
2.3-2.4 21.28 21.20 3.646 3.679 3.640
2.4-2.5 20.03 20.60 3.956 4.024 3.128
--------- ---------------- ---------------- -----------
2.0-2.5 19.65 19.60 2.989 3.031 3.473
-----------------------------------------------------------
(a) total,fission and capture cross sections calculated by
RESENDD, 1% accuracy at 300 K, from the resonance
parameters.
(b) average total cross sections obtained from the average
experimental effective cross sections of Derrien [23].
(c) 1988 high resolution data of Weston and Todd [15]
normalised to ENDF/B-VI standard in the energy range
from 1 keV to 2 keV.
----------------------------------------------------------------
FISSION AND CAPTURE RESONANCE INTEGRALS
The fission and capture resonance integrals are compared to
JENDL3 data in the following table:
Energy range (eV) Fission(barn) Capture(barn)
----------------- ----------------- -----------------
JENDL3 present JENDL3 present
0.5 - 5.0 85.725 84.879 28.651 28.723
5.0 - 10.0 25.081 25.147 19.059 18.950
10.0 - 50.0 96.856 99.715 77.181 74.686
50.0 - 100.0 40.479 41.552 25.930 25.376
100.0 - 301.0 19.677 20.252 17.952 17.729
301.0 -1000.0 10.047 10.317 8.348 8.418
1000.0 -2000.0 3.484 3.206 2.840 2.634
2000.0 -2.E+07 17.783 (17.783) 5.224 (5.224)
----------------- ----------------- -----------------
Total 299.132 302.851 185.185 181.739
----------------------------------------------------------
The JENDL3 resonance parameters are those obtained in 1987 in
the energy range 0 keV to 1 keV. They are sligthly different from
those published in 1989. Which explains the small differences
observed between JENDL3 and the present results in this energy
range. In the energy range 1 keV to 2 keV, JENDL3 is unresolved
range. The fission and capture resonance integrals calculated
from ENDF/B-V and those found in BNL-325 are the following:
ENDF/B-V Fission: 302.13 b Capture: 194.10 b
BNL-325 Fission: 310+-10 b Capture: 200+-20 b
The consequence of changing from the old sets of resonance
parameters(ENDF/B-V and previous sets) to the new set is that
the capture resonance integral will decrease by 6.7% compared
with the ENDF/B-V value.
----------------------------------------------------------------
UNRESOLVED RESONANCE REGION
The average resonance prameters are given in the energy range
2.5 keV to 30 keV for 70 energy points. They were obtained by
using the Cadarache statistical code FISINGA to fit the gross
structure of the Saclay experimental total cross sections [26]
below 4 keV and of selected experimental fission cross sections
normalised to ENDF/B-VI standard evaluation [11]. Above 4 keV no
high resolution total cross section data are available; average
total cross sections were calculated to be consistent with the
stastistical paramaters obtained in the resolved resonance
region [14] and with the Optical Model parameters of Lagrange
and Madland [24] obtained by fitting the experimental data in
the high energy range. A value of 9.46 fm was used for the
effective radius. The values obtained for alpha are consistent
with the experimental data.
The competitive width is not used for the inelastic scattering
cross section. For each energy point of the unresolved region the
neutron width corresponds only to the elastic scattering cross
section. The inelastic scattering cross section should be found
in file 3.
The cross sections obtained at ORNL by processing the
evaluated file using NJOY-87.1 are given in the following table,
'FISS' for the fission values and 'CAPT' for the capture values.
Energy Cross sections Energy Cross sections
(keV) (barn) (keV) (barn)
------ ----------------- ------ ----------------
FISS CAPT FISS CAPT
2.500 4.280 2.456 13.750 1.715 0.942
2.550 2.725 2.754 14.250 1.492 0.948
2.650 3.103 3.425 14.750 1.797 0.854
2.750 4.169 2.010 15.250 1.883 0.797
2.850 4.126 2.077 15.750 1.697 0.843
2.950 3.362 3.710 16.250 1.801 0.782
3.050 3.017 1.998 16.750 1.628 0.824
3.150 4.896 1.934 17.250 1.498 0.819
3.250 3.954 2.277 17.750 1.862 0.701
3.350 1.710 2.166 18.250 1.711 0.736
3.450 2.198 2.572 18.750 1.632 0.748
3.550 2.214 1.885 19.250 1.738 0.694
3.650 2.394 2.948 19.750 1.743 0.677
3.750 3.067 1.624 20.500 1.672 0.679
3.850 3.556 2.122 21.500 1.646 0.661
3.950 2.931 2.397 22.500 1.472 0.697
4.125 2.114 2.270 23.500 1.632 0.619
4.375 2.509 2.129 24.500 1.636 0.597
4.625 2.772 1.715 25.500 1.547 0.607
4.875 1.980 2.186 26.500 1.628 0.562
5.125 2.406 1.916 27.500 1.544 0.572
5.375 2.153 1.953 28.500 1.568 0.549
5.625 2.294 1.807 29.500 1.609 0.521
Average values of the fission cross sections compared to the
ENDF/B-VI standard evaluation [11] and alpha values compared to
some experimental data are given in the following table.
Energy Cross sections (barn) Alpha
(keV) (1) (2) (3) (4) (5) (6) (7) (8)
------ ------------------------- --------------------------
3- 4 2.992 3.000 2.213 2.20 0.740 0.720 0.895 0.820
4- 5 2.394 2.383 2.073 2.07 0.866 0.870 0.821 0.837
5- 6 2.266 2.301 1.863 1.91 0.822 0.820 0.867 0.834
6- 7 2.006 2.008 1.677 1.63 0.836 0.790 0.816 0.793
7- 8 2.134 2.054 1.409 1.34 0.660 0.640 0.630 0.605
8- 9 2.207 2.216 1.245 1.23 0.564 0.540 0.575 0.530
9-10 1.867 1.864 1.136 1.05 0.608 0.550 0.617 0.569
1-10 2.628 2.622 2.014 2.06 0.767 0.752 0.806 0.768
10-20 1.762 1.764 0.876 0.85 0.497 0.480 0.466 0.498
20-30 1.597 1.595 0.606 0.58 0.379 0.350 0.373 0.388
-------------------------------------------------------------
(1) Fission cross section, present evaluation (0K)
(2) Fission cross section, ENDF/B-VI standard [11]
(3) Capture cross section, present evaluation (293 K)
(4) Capture cross section, Gwin et al. 1976 [4]
(5) Alpha value, present evaluation (293 K)
(6) Alpha value from Gwin et al. 1976 [4]
(7) Alpha value from Sowerby-Konshin evaluation 1971 [25]
(8) Average alpha value from experimental data
The fission and capture resonance integrals obtained at ORNL
are compared to ENDF/B-5 data in the following table.
Energy range Fission (barn) Capture (barn)
(eV) ENDF/B-5 present ENDF/B-5 present
--------------- ----------------- -----------------
0.5 - 5.0 86.02 85.71 32.31 28.65
5.0 - 10.0 26.03 25.08 20.14 19.06
10.0 - 50.0 100.25 96.87 78.66 77.19
50.0 - 100.0 40.32 40.47 27.23 25.93
100.0 - 301.0 19.98 19.68 19.52 17.95
301.0 -1000.0 10.15 10.05 8.54 8.35
--------------- ----------------- -----------------
0.5 -1000.0 282.76 277.85 186.30 177.13
--------------------------------------------------------
The fission and capture resonance integrals are obtained by
adding the ENDF/B-V value above 1 keV to the present evaluation.
These and the corresponding values from ENDF/B-V evaluation are:
Present - Fission: 297.22 b Capture: 184.93 b
ENDF/B-V - Fission: 302.13 b Capture: 194.10 b
----------------------------------------------------------------
REFERENCES
1. R. Gwin et al., Nucl.Sci.Eng. 45, 25 (1971).
2. A.J. Deruyter et al., J.Nucl.En. 26, 293 (1972).
3. J. Blons, Nucl.Sci.Eng. 51, 130 (1973).
4. R. Gwin et al., Nucl.Sci.Eng. 59, 79 (1976).
5. R. Gwin et al., Nucl.Sci.Eng. 61, 116 (1976).
6. W. Wagemans, Ann.Nucl.En. 7 #9, 495 (1980).
7. R. Gwin et al., Nucl.Sci.Eng. 88, 37 (1984).
8. L.W. Weston et al., Nucl.Sci.Eng. 88, 567 (1984).
9. J.A. Harvey et al., Nuclear Data for Sci. and Technol., Proc.
Int. Conf., May 30 - June 3, 1988, Mito, Japan (Saikon
Publishing Co., 1988) p.115.
10. R.R. Spencer et al., Nucl.Sci.Eng. 96, 318 (1987).
11. A. Carlson et al., preliminary results of the ENDF/B-6
standard evaluation (Sep.8, 1987); see W. P. Poenitz et al.,
Argonne National Laboratory report ANL/NDM-139 [ENDF-358]
(1997)
12. A.J. Deruyter, J.Nucl.En. 26, 293 (1972).
13. N.M. Larson et al., Oak Ridge National Laboratory reports
ORNL/TM-7485, ORNL/TM-9179, and ORNL/TM-9719/R1
14. H. Derrien and G. DeSaussure, Oak Ridge National Laboratory
report ORNL-TM-10986 (1988).
15. L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992).
16. H. Derrien et al., Nucl.Sci.Eng. 106, 434 (1990).
17. H. Derrien and T. Nakagawa, to be published.
18. L. Leal and R.N. Hwang, Trans.Am.Nucl.Soc. 55, 340 (1987).
19. H. Derrien et aL., Nucl.Sci.Eng. 106, 434 (1990).
20. T. Nakagawa, RESENDD a JAERI version of RESEND
21. C. Wagemans et al., Nuclear Data for Sci. and Technol., Proc.
Int. Conf., May 30 - June 3, 1988, Mito, Japan (Saikon
Publishing Co., 1988) p.91.
22. L.W. Weston et al., Nucl.Sci.Eng. 115, 164 (1993).
23. H. Derrien, to be published in J.Nucl.Sci.Technol.
24. Ch. Lagrange and D.G. Madland, Phys.Rev. C 33, 1616 (1986).
25. M.G. Sowerby et al., At.En.Rev. 10, 453 (1972)
26. H. Derrien, thesis, Univ. Paris - Sud, Orsay Serie A No. 1172
(1973).
27. A. Lendl et al.,Atomnaya Energiya Vol61,N3,pp215-216,(1986)
28. E. Fort et al., paper to SGC/WPEC, (2002)
29. F.J. Hambsch et al, Jour. of Nuc. Sci. and Tech., ND-2001
procedings, to be published, (2002)
30. E. Fort et al., NSE99,pp375-389, (1988)
31. G.Vladuca, A.Tudora., Ann.Nuc.Energy. 28, 689 (2001).
******************************************************************
ENERGY REGION 0.03 TO 30 MeV ***********************************
Principal evaluators : J.P. Delaroche, S.Hilaire, B. Morillon
P. Romain.
******************************************************************
The evaluation above 30 keV is based on a detailed theoretical
model analysis utilizing the available experimental data and
microscopic level densities as guides to phenomenological models.
Coupled channel optical model calculations were used to provide
the total and direct reaction components of elastic and inelastic
scattering cross sections and angular distributions for
collective levels. These are the (1/2)+, (3/2)+, ..., (11/2)+
members of the ground state band, and the (1/2)-, (3/2)- and
(5/2)- members of the experimentally identified octupole band.
The plain rotation-vibration model is adopted. The coupling
strength for interband transitions is closed to that deduced from
coupled channel analyses of inelastic scattering data for the
Kpi=(0)- vibrational band of U238.
Coupled channel calculations are performed using the ECIS
code [Ra70] which also provides coumpound nucleus cross sections
and transmission coefficients used in pre-equilibrium/evaporation
emission treated in the Exciton and HAUSER-FESHBACH models
implemented in the GNASH code [Yo96].
This reaction code has been modified to include width
fluctuation factors, relativistic kinematics, and a more
realistic treatment of the fission process. Briefly, the
simple double-humped fission barrier model is improved by
treating explicitly the coupling between class I and class
II states and damping of class II states.
Emission of light hadrons up to He4 is explicitly treated in the
model calculations. Fission decay of associated residual nuclei
is also treated. But none of these emission and fission cross
sections are explicitely provided in the files.
Above 16.5 MeV the sigma (n,3n), (n,4n) and (n,5n) include
components from Light Charged Particles (LPCs). For instance
sigma (n,3n) given in MT=17 is the sum of the actual (n,3n)
cross section and cross sections associated with LPCs :
Effective sigma(n,3n)= True sigma(n,3n) +
sigma (n,LPC) * sigma(n,3n) / [sigma(n,3n)+sigma(n,4n)+
sigma(n,5n)]
Below is provided a table of such relationships between true
and effective sigma(n,xn) cross sections.
*****************************************************************
*Neutron* Sigma* Sigma* Sigma* Sigma* Sigma* Sigma* Sigma*
* Energy* (n,3n)* (n,3n)* (n,4n)* (n,4n)* (n,5n)* (n,5n)*(n,LCP)*
* * | * (+LCP)* | * (+LCP)* | * (+LCP)* | *
* (MeV) * (b) * (b) * (b) * (b) * (b) * (b) * (b) *
*****************************************************************
*.1650+2*.1095+0*.1190+0*.0000+0*.0000+0*.0000+0*.0000+0*.9476-2*
*.1700+2*.1430+0*.1538+0*.0000+0*.0000+0*.0000+0*.0000+0*.1077-1*
*.1750+2*.1809+0*.1931+0*.0000+0*.0000+0*.0000+0*.0000+0*.1215-1*
*.1800+2*.2219+0*.2355+0*.0000+0*.0000+0*.0000+0*.0000+0*.1358-1*
*.1850+2*.2633+0*.2784+0*.0000+0*.0000+0*.0000+0*.0000+0*.1512-1*
*.1900+2*.2995+0*.3162+0*.0000+0*.0000+0*.0000+0*.0000+0*.1669-1*
*.1950+2*.3222+0*.3405+0*.1969-4*.2081-4*.0000+0*.0000+0*.1829-1*
*.2000+2*.3363+0*.3562+0*.2584-3*.2737-3*.0000+0*.0000+0*.1996-1*
*.2050+2*.3368+0*.3585+0*.1271-2*.1353-2*.0000+0*.0000+0*.2175-1*
*.2100+2*.3191+0*.3423+0*.3907-2*.4191-2*.0000+0*.0000+0*.2345-1*
*.2150+2*.2989+0*.3235+0*.8409-2*.9100-2*.0000+0*.0000+0*.2526-1*
*.2200+2*.2791+0*.3047+0*.1515-1*.1654-1*.0000+0*.0000+0*.2695-1*
*.2250+2*.2629+0*.2894+0*.2350-1*.2586-1*.0000+0*.0000+0*.2882-1*
*.2300+2*.2291+0*.2557+0*.3475-1*.3878-1*.0000+0*.0000+0*.3058-1*
*.2350+2*.2019+0*.2280+0*.4841-1*.5467-1*.0000+0*.0000+0*.3234-1*
*.2400+2*.1763+0*.2013+0*.6387-1*.7291-1*.0000+0*.0000+0*.3400-1*
*.2450+2*.1560+0*.1798+0*.7793-1*.8984-1*.0000+0*.0000+0*.3575-1*
*.2500+2*.1367+0*.1586+0*.9692-1*.1124+0*.0000+0*.0000+0*.3735-1*
*.2550+2*.1241+0*.1443+0*.1158+0*.1346+0*.0000+0*.0000+0*.3896-1*
*.2600+2*.1098+0*.1280+0*.1342+0*.1564+0*.0000+0*.0000+0*.4034-1*
*.2650+2*.1004+0*.1170+0*.1518+0*.1769+0*.0000+0*.0000+0*.4170-1*
*.2700+2*.9288-1*.1084+0*.1643+0*.1917+0*.4294-7*.5011-7*.4293-1*
*.2750+2*.8309-1*.9739-1*.1733+0*.2031+0*.1862-5*.2182-5*.4412-1*
*.2800+2*.7845-1*.9216-1*.1818+0*.2136+0*.1668-4*.1959-4*.4548-1*
*.2850+2*.7155-1*.8510-1*.1802+0*.2143+0*.8049-4*.9574-4*.4771-1*
*.2900+2*.6966-1*.8362-1*.1784+0*.2142+0*.2391-3*.2870-3*.4976-1*
*.2950+2*.6734-1*.8218-1*.1678+0*.2048+0*.6368-3*.7772-3*.5197-1*
*.3000+2*.6565-1*.8130-1*.1596+0*.1976+0*.1279-2*.1584-2*.5400-1*
*****************************************************************
The (n,5n) cross section is provided in the above table, but not
inserted in the file.
MF=3 Smooth Cross Sections -------------------------------------
MT=1 Neutron Total Cross Section. 0.03 to 30 MeV, analysis
based on coupled-channel optical calculations and the
exp. data of [Po81,Sh78,Po83,Sc74,Fo71,Sm73,Na73,Pe60,
Ca73,Li90]. Calculated as the sum of MT=1 and MT=3.
MT=2 0.030 to 30 MeV, based on coupled channel and
statistical model calculations.
MT=3 0.030 to 30 MeV,
The sum of partial cross sections is calculated using
GNASH, in which the neutron transmission coefficients
we use are from ECIS calculations. Compound elastic
component is not included in the above sum.
MT=4 0.030 to 30 MeV, based on sum of MT=51-91.
MT=16 (n,2n) cross section 0.030 to 30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
MT=17 (n,3n) cross section 0.030 to 30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
For more information see comments and table above.
MT=18 fission cross section 0.030 to 30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
This file includes components stemming from fission
of residuals associated with charged particle emission.
MT=37 (n,4n) cross section 0.030 to 30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
For more information see comments and table above.
MT=51-55 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=56-63 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=64 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=65 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=66-67 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=68-77 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=91 Thres. to 30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
MT=102 0.030-30 MeV,
GNASH Hauser-Feshbach statistical/preequilibrium calc.
MF=4 Neutron Angular Distributions -----------------------------
Tabulated sigma(theta) values
MT=2 Elastic scattering angular distribution based on ECIS
coupled-channel calculations and GNASH calculations.
MT=16,17,37 Isotropic distributions.
MT=18 Isotropic distribution.
MT=51-55 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=56-63 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=64 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=65 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=66-67 Thres. to 30 MeV, coupled-channel optical model
calculations [(3/2)+ to (11/2)+] members of the
Kpi=(1/2)+ ground state rotational band, and (1/2)-,
(3/2)- and (5/2)- members of the octupole band)
using the ECIS code. Compound nucleus contributions,
obtained from GNASH calculations, are also included.
MT=68-77 Thres. to 30 MeV, Compound nucleus reaction theory
calculations using the GNASH code.
MOLDAUER width fluctuation factors are turned off beyond
4 MeV incident energy.
MT=91 Isotropic distribution.
MF=5 Neutron Energy Distributions ------------------------------
MT=16 GNASH Hauser-Feshbach statistical/preequilibrium calc.
Updated Kalbach-Mann systematics used for specifying
neutron distributions [Ka87]. Only neutrons given.
MT=17 GNASH Hauser-Feshbach statistical/preequilibrium calc.
Updated Kalbach-Mann systematics used for specifying
neutron distributions [Ka87]. Only neutrons given.
MT=18 Neutron energy distributions from fission based on the
Los Alamos model, with multiple chances (first, second,
third, fourth and fifth chance), and upgraded by
G.Vladuca and A.Tudora [Vl01].
A linear relation between the average prompt gamma ray
energy and the average prompt neutron multiplicity and a
dependence of the average fission fragments kinetic
energy on the incident neutron energy are used.
The model parameters are slightly different from those
adopted in [Vl01].
MT=37 GNASH Hauser-Feshbach statistical/preequilibrium calc.
Updated Kalbach-Mann systematics used for specifying
neutron distributions [Ka87]. Only neutrons given.
MT=91 GNASH Hauser-Feshbach statistical/preequilibrium calc.
Updated Kalbach-Mann systematics used for specifying
neutron distributions [Ka87]. Only neutrons given.
MT=455 Tal England [En89].
MF=12,13,14,15 Photon-Production Data -----N.Y.I.---------------
----------------------------------------------------------------
REFERENCES
[Ar84] E. Arthur et al., Nuc.Sci.Eng. 88, 56 (1984).
[Ca73] J. Cabe et al., report CEA-R-4524 (1973).
[En89] T.R. England et al, Los Alamos reports LA 11151-MS
(1988) and LA-11534-T (1989); M.C. Brady and T.R. England,
Nucl.Sci.Eng. 103, 129 (1989).
[Fo71] D. Foster and D. Glasgow, Phys.Rev. C3, 576 (1971).
[Ka87] C. Kalbach, Phys.Rev. C 37, 2350 (1988).
[Li90] P. Lisowski, private comm. of WNR data taken in 1985.
[Na73] K. Nadolny et al., USNDC-9 (1973)p.170
[Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960).
[Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981).
[Po83] W. Poenitz et al., Argonne National Laboratory report
ANL-NDM-80 (1983).
[Ra70] J. Raynal,IAEA SMR-9/8 (1970).
[Sc74] R. Schwartz et al., Nucl.Sci.Eng. 54, 322 (1974).
[Sh78] R. Shamu et al., private communication, 1978.
[Sm73] A. Smith et al., J.Nuc.En. 27, 317 (1973).
[Vl01] G.Vladuca, A.Tudora., Ann.Nuc.Energy. 28, 689 (2001).
[Yo96] P.G. Young, E.D. Arthur and M. B. Chadwick,
in Workshop on Nuclear Reaction Data
and Nuclear Reactors, Trieste, Italy (1996).
Back
|
