| |
-
endfDoc: click to expand
-
94-Pu-239 LANL EVAL-SEP06 Young,Chadwick,MacFarlane,Derrien
DIST-DEC06 20111222
----ENDF/B-VII MATERIAL 9437
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
*****************************************************************
12/09/2010 S.T. Holloway on behalf of LANL
Reverted MF=1/5, MT=455 Delayed-neutron 6-grp data to
ENDF/B-VI.8 values.
*****************************************************************
UPDATED evaluation in the resolved range
L. Luiz et al. (ORNL)
Features of the evaluation:
1. One single set of resonances parameters covering the energy
range from 0.00001 eV to 2.5 keV.
2. SAMMY (Reich-Moore) analysis of the experimental data led to
a set of resonance parameters that fit well the experimental
data and improves benchmark calculations in the thermal
region.
3. Integral quantities such as K1 and eta (nu*capture/fission)
were the two major indicators on how to fix the problem with
thermal benchmark results.
4. Experimental data used in the previous Pu-239 resonance
were used. The experimental data were well represented
with the new resonance evaluation.
ENDF/B-VII EVALUATION
P.G.Young, M.B.Chadwick, R.E.MacFarlane, W.B.Wilson, D.G.Madland,
P.Talou, T. Kawano (LANL)
and
H. Derrien, G. de Saussure, T. Nakawa (ORNL,JAERI)
and
D.A. Brown, J.Pruet (LLNL)
----- SUMMARY -----
Major features of the ENDF/B-VII evaluation are:
1. A new evaluation of the (n,f) cross section based on ENDF/B-
VII standard cross section analysis is incorporated.
2. A new evaluation of nubar consistent with experimental data
within uncertainties and with fast critical benchmark
measurements is included.
3. A new analysis of the prompt fission neutron spectrum matrix
based on the Los Alamos model is used to calculate neutron
spectra at all incident neutron energies.
4. Improved delayed neutron data are incorporated.
5. New experimental data from a LANSCE GEANIE experiment are
combined with older data and GNASH theoretical calculations to
produce a new evaluation of the 239Pu(n,2n) cross section.
6. Direct reaction cross sections and angular distributions are
extended to an excitation energy of 4 MeV using a DWBA
analysis of 238U fission neutron spectra.
7. Improved fission energy release values are incorporated.
8 A new evaluation of the resonance parameters in the energy
range 0 to 2.5 KeV.
9. A resonance parameter covariance matrix evaluation.
10.A fast-energy covariance matrix evaluation.
----- DETAILED DESCRIPTION -----
>> MF=1 GENERAL INFORMATION
MT=452: TOTAL AVERAGE NEUTRON MULTIPLICITY PER FISSION (TOTAL
NUBAR)
Sum of MT=455 and 456.
MT=455: AVERAGE DELAYED NEUTRON MULTIPLICITY PER FISSION
Based on new delayed neutron 6-group parameters from CINDER'90
summation calculations (Wi05). The CINDER calculations are based
on a new CINDER library in which beta-decay half-lives and beta-
delayed neutron-emission probablities are obtained from the
evaluated experimental data file NuBase2003 (Au03), when
available there. When experimental data are not available, the
data are calculated in a model where allowed Gamow-Teller decays
are treated in a microscopic quasi-particle random-phase
approximation (QRPA) and the first forbidden decays are treated
in the statistical gross theory (Mo03).
MT=456: AVERAGE PROMPT NEUTRON MULTIPLICITY PER FISSION
1.e-5 eV - 1.0 keV: Taken from ENDF/B-VI without change. The
ENDF/B-VI evaluation in this energy range is based on an
evaluation by Fort (Fo88), after a very small (1.000411)
renormalization for consistency with the CSEWG thermal nubar
value from the standards analysis.
1.0 keV - 20.0 MeV: Minor modifications were made to the
ENDF/B-VI evaluation to improve agreement with the results of the
covariance analysis of experimental data that was used for that
evaluation and with integral experimental results. Also, the
ENDF/B-VI data were adjusted above 6-8 MeV for consistency with
the ENDF/B-VII standard 252Cf nubar value. We attempted
to follow the covariance data as well as possible but mainly to
stay roughly within uncertainties in the data and to keep good
agreement with fast critical benchmarks. The most serious
departure from the covariance data occurs below 1.5 MeV, where
the evaluation lies about two standard deviations above the
experimental data. This difference, however, was influenced
strongly by the need to match the integral data results for the
JEZEBEL fast critical experiment.
MT=458: ENERGY RELEASE FROM FISSION
Modifications were made to MT=458 based on a new analysis by
Madland (Ma06). The average total fission product kinetic energy
(EFR) and the average total prompt fission gamma-ray energy (EGP)
were taken from the Madland analysis. The average total prompt
fission neutron kinetic energy (ENP) was obtained from NJOY,
using the MF=5,MT=18 fission neutron spectra and prompt nubar
(MT=1,MT=456) from the evaluation. (This value of total neutron
energy is close to Madland's result.) The kinetic energy of
delayed fission neutrons (END), the total energy from delayed
gamma rays (EGD), the total energy released by delayed betas
(EB), and the energy released by neutrinos (ENU) were carried
over from the ENDF/B-VI evaluation. The total energy release per
fission is: ET = EFR+ENP+END+EGP+EGD+EB+ENU, and the quantity ER
(total energy less the neutrino energy, or pseudo-Q value) is
simply ET-ENU. ER is also included as the fission Q-value in
MF=3, MT=18,19,20,21,38.
The uncertainties in EFR, ENP, END, EGP, EGD, EB, ENU, ER, and
ET are carried over from ENDF/B-VI.
There is currently no format for including energy dependence
in the ENDF/B file. These are included in the NJOY processing.
MT=460: BETA-DELAYED FISSION GAMMA DATA
See D.A. Brown (Br06). Evaluated by J. Pruet, based on Pruet
et al. (Pr04).
----- REFERENCES (MF=1) -----
Au03 G. Audi, O. Bersillon, J. Blanchot, and A. H. Wapstra, Nucl.
Phys. A729 (2003) p. 3-129.
Br06 D.A. Brown, Lawrence Livermore National Laboratory report
UCRL-TR-223148 (2006).
Fo88 E. Fort et al., Nuc.Sci.Eng. 99, 375 (1988).
He80 D. M. Hetrick and C. Y. FU, "GLUCS: A Generalized Least-
Squares Program for Updating Cross-Section Evaluations with
Correlated Data Sets," Oak Ridge National Laboratory report
ORNL/TM-7341 (ENDF-303) (1980).
Ma06 D. G. Madland, Nucl. Phys. A772, 113 (2006).
Mo03 P. Moller, B. Pfeiffer, and K.-L. Kratz, Phys. Rev. C. 67
(2003) 055802.
Pr04 J. Pruet, J. Hall, M.-A. Descalle, S.G. Prussin, Nucl. Inst.
Meth. B 222, 403 (2004).
Wi05 W. B. Wilson, personal communication, 2005.
>> MF=2 RESONANCE PARAMETERS
** THE PREVIOUS EVALUATION PRESENT IN ENDF/B-VII.0 (SEE FURTHER
BELOW) IS KEPT FOR NOW, AND THE FOLLOWING NOTES CORRESPOND TO A
PRELIMINARY NEW EVALUATION ONLY. HOWEVER, THE NEW COVARIANCE
MATRIX FOR MF32 (PROCESSED INTO MF33) IS KEPT FOR ENDF/B-VII.1 **
MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS
NEW EVALUATION of the RESONANCE PARAMETERS of Pu-239 in the ENERGY
RANGE 0 to 2.5 keV- H. Derrien, L.C. Leal, and N.M. Larson (2007)
In order to have more flexibility for the calculation of a
complete Resonance Parameter Covariance Matrix (RPCM) the three
uncorrelated energy ranges were merged in a single energy range.
A reevaluation of external parameters was needed to conserve the
values of the cross sections in the thermal energy range. New
experimental data were added to the SAMMY experimental data base,
particularly the last fission measurements of Weston et al.(1993)
and of Wagemans et al.(1993) The problem of the normalization of
the fission cross section in the energy range 0.1 to 1.0 keV was
solved by Weston(1993) and by Wagemans (1993), close to the
results of the standard evaluation(1990). Comparisons of results
are found in the following Table.
=============================================================
Energy Weston Weston Standard ENDF/B-6 Present
Range keV 1984 1988 1990 1993 2007
=============================================================
0.10-0.20 18.50 19.00 18.67 18.69 18.41 -1.5%
0.20-0.30 17.82 17.83 17.88 17.80 17.67 -0.7%
0.30-0.40 8.31 8.203 8.430 8.301 8.268 -0.4%
0.40-0.50 9.54 9.403 9.570 9.580 9.447 -1.4%
0.50-0.60 15.51 15.31 15.56 15.40 15.36 -0.3%
0.60-0.70 4.28 4.189 4.459 4.358 4.278 -1.9%
0.70-0.80 5.50 5.502 5.630 5.509 5.525 +0.3%
0.80-0.90 4.87 4.873 4.979 4.842 4.877 +0.7%
0.90-1.00 8.36 8.437 8.297 8.436 8.424 -0.1%
-------------------------------------------------------------
0.10-1.00 10.30 10.31 10.37 10.32 10.25 -0.7%
-------------------------------------------------------------
1.00-1.10 5.657 5.564 5.551 5.488 -1.1%
1.10-1.20 6.026 6.118 5.975 5.960 -0.3%
1.20-1.30 4.466 4.684 4.592 4.496 -2.1%
1.30-1.40 7.331 7.299 6.980 7.253 +3.9%
1.40-1.50 3.943 4.130 4.033 3.984 -1.2%
1.50-1.60 2.422 2.583 2.555 2.435 -4.9%
1.60-1.70 3.830 4.075 3.944 3.890 -1.4%
1.70-1.80 3.204 3.513 3.387 3.340 -1.4%
1.80-1.90 5.315 5.448 5.165 5.307 +2.7%
1.90-2.00 1.996 2.139 2.155 1.960 -9.9%
-------------------------------------------------------------
1.00-2.00 4.419 4.555 4.466 4.432 4.419 -0.3%
-------------------------------------------------------------
2.00-2.10 1.983 2.014 2.033 1.947 -4.4%
2.10-2.20 3.028 3.029 2.942 2.988 +1.6%
2.20-2.30 2.399 2.342 2.351 2.242 -4.9%
2.30-2.40 3.534 3.715 3.637 3.636 -0.0%
2.40-2.50 4.118 4.050 3.961 4.034 +1.8%
-------------------------------------------------------------
2.00-2.50 3.012 3.030 2.985 2.968 -0.6%
=============================================================
Weston 1984 and 1988 experimental data, ENDF/B-6, and the
results of the present analysis are displayed in the table. Weston
table. Weston 1984 data and Weston 1988 data were normalized to
the integral standard proposed by Wagemans in 1993.The last column
1993.The last column shows the percentage deviation between
ENDF/B-6 and the present evaluation. The cross sections are given
in barns. The average cross sections were calculated with the
code SAMMY.
The cross sections at 0.0253 eV are given in the following
Table:
===============================================================
Standard B-VI Mughabghab Present
1992 1993 2005 2007
===============================================================
Total (b) (1027.30) 1026.30 1025.3+/-2.9 1027.30
Fission (b) 747.99+/-0.25 747.66 748.1+/-2.0 747.09
Capture (b) 271.43+/-0.79 270.65 269.3+/-2.9 271.40
Scattering (b) 7.88+/-2.30 7.99 7.94+/-0.4 8.81
===============================================================
The capture resonance integrals are compared to the previous
evaluation in several energy range, in the following Table.
Table.
==============================================================
Energy Range ENDF/B-VI Present Percentage
(eV) ( barn) (barn) difference
==============================================================
0.01-0.10 626.7 627.7 +0.2%
0.10-0.30 777.9 767.5 -1.4%
0.30-0.50 371.1 376.1 +1.3%
--------------------------------------------------------------
0.01-0.50 1775.7 1771.3 -0.2%
--------------------------------------------------------------
0.50-0.70 17.58 17.67 +0.5%
0.70-7.00 11.82 11.32 -4.4%
7.00-100. 118.26 117.91 -0.3%
100.-1000. 26.09 25.26 -3.3%
1000.-2000. 2.62 2.51 -4.4%
2000.-2500. 0.77 0.71 -8.5%
--------------------------------------------------------------
0.5-2500. 177.09 175.29 -1.0%
==============================================================
Resonance parameter covariance was generated in the resolved
energy region (1.0-5 eV to 2.5 keV) with the computer code SAMMY
at ORNL. Experimental data and uncertainties were used to generate
covariance data. SAMMY covariance information was converted in
the ENDF FILE32 representation. However, because of computer
storage concern the FILE32 was converted into the FILE33 (MT=1,
2, 18, and 102) representation with reduced computer storage but
approximating the features of the FILE32 representation. The full
FILE32 matrices can be obtained directly from ORNL. The
conversion was done with the PUFF-IV code. Above the resolved
resonance region (> 2.5 keV) covariance data were generated at
LANL. Uncertainty on nubar was also generated at LANL.
The details of the reevaluation are in progress of publication in
an ORNL/TM report.
References:
We 93, Weston, NSE, 115, 164, (1993)
Wa 93, Wagemans, NSE, 115, 173 (1993)
We 84, Weston, NSE, 88, 567 (1984)
We 92, Weston, NSE, 14, 415 (1992)
--------------------- PREVIOUS EVALUATION --------------------
----- RESOLVED RESONANCE PARAMETERS (1.e-5 - 2500 eV) -----
The resolved resonance parameters are the same as those in
ENDF/B-VI Release 8. They were first installed in MOD 2 of
ENDF/B-VI in January, 1993, by H. Derrien (ORNL) and T. Nakagawa
(JAERI). This revision extended the resonance region to
2.5 keV. This description is taken from the ENDF/B-VI
evaluation.
PU239 RESONANCE DATA 0 keV TO 2.5 keV
(H.Derrien, T.Nakagawa)
The present file contains the resonance parameters obtained
from a SAMMY fit analysis of high resolution experimental data,
performed at ORNL (Oak Ridge National 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 independent sections:
1) the first corresponds to the energy range 0 keV to 1 keV.
The corresponding set of resonance parametres contains 398
resonances in the energy range 0 keV to 1 keV, 4 fictitious
negative energy resonances and 3 fictitious 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
fictitious resonances below 0.9 keV and 3 fictitious
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
fictitious resonances below 1.98 keV and 3 fictitious
resonances above 2.53 keV.
In all sections the fictitious 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 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). Deruyter 1970 and Wagemans 1980 data are about
2.5% larger and were not included in the SAMMY fit.
When normalized 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 renormalization 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 normalization
coefficients and of the background corrections. This preliminary
step has shown that this adjustment was not necessary to have
consistency between Harvey data and Weston data. The Blons data
needed a large readjustment of the background and normalization.
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 normalization 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
transmission, Weston 84 fission and preliminary Weston 88 fission
was restarted from the parameter and covariance files obtained in
1987. It appeared that large background and normalisation
corrections were needed on the new Weston fission to obtain
consistency with Harvey transmission data. These corrections were
comparable to those found on Blons data and were not understood
by the authors of the experiment. The last SAMMY runs were
performed by not allowing background and normalization 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
normalization of the data in the thermal energy range showed that
the 1984 data should be renormalized by about +3%. To take into
account this renormalization, 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
corresponding resonances;
2) adjustment of the neutron width of the 0+ resonances by a
refit of the transmission data and of the renormalized 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
calculated 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 normalized 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 fictitious 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 renormalization 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 Westin 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] normalized to ENDF/B-VI standard in the energy range
from 1.0 keV to 2.0 keV.
(d) Gwin 1971 experimental data normalized 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 normalized to Gwin 1976 average values; there are large
differences between the calculated data and the experimental data
averaged over 0.1 keV 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 normalization 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 normalized 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]
normalized 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 PARAMETERS (2.5 - 30 keV) -----
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
normalized 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
statistical 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. In 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 file 3.
The cross sections obtained at 0RNL 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 0RNL 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 (MF=2) -----
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).
>> MF=3 NEUTRON CROSS SECTIONS
----- GENERAL INFORMATION -----
The maximum energy of the evaluation remains at 20 MeV, the
same as ENDF/B-VI.
Three major changes were made to MF=3 in going from the
ENDF/B-VI 239Pu evaluation to the ENDF/B-VII evaluation: (1) the
(n,f) cross section was revised based on the ENDF/B-VII standard
cross section analysis; (2) the (n,2n) cross section was revised
based on new experimental data; and, (3) direct reaction cross
sections and angular distributions, inferred from neutron
spectrum measurements on 238U, are included for groups of states
in the MT=72-90 data.
The evaluation of other reactions above 10 keV is based on
ENDF/B-VI which, in turn, is based on a detailed theoretical
analysis utilizing the available experimental data. In the
ENDF/B-VI analysis, coupled-channels optical model calculations
with the ECIS code (Ra70) were used to provide the total,
elastic, and inelastic cross sections to the first 7 members of
the ground state rotational band, as well as neutron elastic and
inelastic angular distributions to the rotational levels. The
deformed optical model potential used for the calculations is
potential #7 in the International Atomic Energy Agency's
Reference Input Parameter Library (RIPL-2) optical model
parameter library (see Yo94). The ECIS code was also used to
calculate neutron transmission coefficients. Hauser-Feshbach
statistical theory calculations were carried out with the GNASH
(Ar88,Yo98) and COMNUC (Du71) code systems, including
preequilibrium and fission channels. DWBA calculations were
performed with the DWUCK code (Ku70) for several vibrational
levels, using B(El) values inferred from (d,d') data on Pu238 and
Pu240, as well as Coulomb excitation measurements. A weak
coupling model (Pe69) was used to apply the Pu238 and Pu240
results to states in Pu239.
----- DETAILS OF LANL REVISION -----
MT=1: TOTAL CROSS SECTION
0.03 to 20 MeV: Based on coupled-channel optical calculations
and the exp. data of (Po81,Sh78,Po83,Sc74,Fo71,Sm73,Na73,Pe60,
Ca73,Li90). The experimental and theoretical results were
combined through a covariance analysis with the GLUCS code system
(He80). The covariance analysis results, which agreed well with
Derrien's (De89) unresolved resonance analysis at 30 keV, were
smoothly joined to those results between 30 and 50 keV.
MT=2: ELASTIC SCATTERING CROSS SECTION
0.030 to 20 MeV: Based on subtraction of MT=4,16,17,18,37,
and 102 from MT=1.
MT=3: NONELASTIC CROSS SECTION
The (redundant) nonelastic cross section is the sum of the
following reactions: MT=4,16,17,18,37,102.
MT=4: INELASTIC CROSS SECTION
Sum of MT=51-91.
MT=16: (n,2n) CROSS SECTIONS
Based on new experimental data from a LANSCE-GEANIE
experiment, combined with older data and GNASH theoretical
calculations (Be02). The new results are deduced from a
combination of measured partial gamma-ray cross sections and
enhanced Hauser-Feshbach reaction modeling. Older measurements by
Mather (Ma72), Frehaut (Fr80), and Lo(02) are also included in
the analysis. The evaluation is discussed in Be02.
MT=17 and 37: (n,3n) and (n,4n) CROSS SECTIONS
Taken from ENDF/B-VI. Based on the GNASH Hauser-Feshbach
statistical/preequilibrium calculations described above.
MT=18: FISSION NEUTRON CROSS SECTION
The Pu239(n,f) cross section that resulted from the
simultaneous standards analysis for ENDF/B-VII (Pr05) was used
with minimal smoothing at all incident neutron energies above the
resonance region. The original standard energy grid is included
as a subset of a denser grid. The expansion to the denser grid
was accomplished using a spline fit to a log-log file of the
standard data.
The Q-value was changed from 199.92 MeV to 198.8438 MeV to
maintain consistency with the MF=1,MT=458 data.
MT=19,20,21,38: MULTI-CHANCE FISSION CROSS SECTIONS
The ratios of the multi-chance fission cross sections to the
total fission cross section were obtained from GNASH
calculations. The evaluated multi-chance fission cross sections
were then obtained by multiplying the MT=18 total fission cross
section by the ratios.
The Q-value was changed from 199.92 MeV to 198.8438 MeV to
maintain consistency with the MF=1,MT=458 data.
MT=51-55,57: DISCRETE INELASTIC LEVEL CROSS SECTIONS (GROUND-
STATE ROTATIONAL BAND)
Thres. to 20 MeV: Coupled-channels optical model calculations
were made to the 3/2+ to 13/2+ members of the K=1/2 ground state
rotational band using the ECIS code. Compound nucleus
contributions, obtained from COMNUC calculations, are also
included. These were taken from ENDF/B-VI, with one-to-one
correspondance of MT numbers.
MT=56,58-63,65,67,68: DISCRETE INELASTIC LEVEL CROSS SECTIONS
(COMPOUND NUCLEUS REACTIONS TO COMBINED
LEVELS)
Threshold to 6.0 MeV, Compound nucleus reaction theory
calculation with width fluctuations, using the COMNUC code. These
were taken from ENDF/B-VI, combining levels as follows:
ENDF/B-VII ENDF/B-VI
MT=58 MT58 + MT59
MT=59 MT60 + MT61
MT=60 MT62 + MT63
MT=61 MT64 + MT65
MT=62 MT66 + MT67
MT=63 MT68 + MT69
MT=65 MT71 + MT72
MT=67 MT74 + MT75
MT=68 MT76 + MT77
MT=64,66,69-71: DISCRETE INELASTIC LEVEL CROSS SECTIONS (DIRECT
REACTIONS TO INDIVIDUAL STATES)
Threshold to 20 MeV, distorted wave Born approximation
calculations with the DWUCK code (Ku70) for l=2 and l=3
vibrational states. The l=2 states are MT=69,71 and the l=3
states are MT=64,66, and 70. Compound nucleus contributions were
included in the data for MT = 64 and 66. The cross sections were
taken from the ENDF/B-VI evaluation, and the correspondance of MT
numbers is as follows:
ENDF/B-VII ENDF/B-VI
MT=64 MT=70
MT=66 MT=73
MT=69 MT=78
MT=70 MT=79
MT=71 MT=80
MT=72-90: DISCRETE INELASTIC LEVEL CROSS SECTIONS (DIRECT
REACTIONS TO GROUPS OF STATES)
The cross sections for MT=72-90 are based upon Baba's neutron
emission spectra measurements (Ba89) at 14 MeV for n + 238U
reactions. To accomplish this, we adopted in the present
evaluation the MT=72-90 data from the ENDF/B-VII 238U evaluation
(MAT=9237). To free up the necessary MT slots in the present
evaluation, it was necessary to combine data for some of the
weaker levels, as tabulated above.
The 238U results are based on DWBA calculations run with the
ECIS94 code (Ra94), assuming a set of 2+ or 3- (mainly)
vibrational states. Deformation parameters were determined by
matching the 14-MeV Baba data. The tabulated cross sections are
actually sums of contributions to several states. The DWBA
calculations were used to extrapolate the 14-MeV cross sections
to lower and higher energies, and to obtain the MF=4 angular
distributions for each assumed state. The spins, parities, and
deformation parameters used in the calculations are given in the
table below. These results affect the evaluation in the
excitation energy range Ex=1.17-3.91 MeV.
MT Ex (MeV) J Pi Beta
0.00000000 0.0 +1 0.0000E+00
72 1.17000000 3.0 -1 3.8087E-02
73 1.25000000 2.0 +1 3.0175E-02
74 1.44000000 3.0 -1 5.6001E-02
75 1.59000000 3.0 -1 3.8111E-02
76 1.75000000 3.0 -1 3.9460E-02
77 1.85000000 3.0 -1 3.5265E-02
78 1.95000000 3.0 -1 4.0750E-02
79 2.15000000 3.0 -1 4.7400E-02
80 2.30000000 3.0 -1 5.3002E-02
81 2.39000000 4.0 +1 8.8154E-03
82 2.49000000 2.0 +1 2.5122E-02
83 2.94000000 2.0 +1 2.7150E-02
84 3.18900000 2.0 +1 2.5287E-02
85 3.38800000 2.0 +1 2.5070E-02
86 3.53800000 2.0 +1 1.5390E-02
87 3.63700000 2.0 +1 1.6125E-02
88 3.73700000 2.0 +1 1.6472E-02
89 3.83700000 2.0 +1 1.4293E-02
90 3.90900000 2.0 +1 1.5091E-02
MT=91: INELASTIC CONTINUUM NEUTRON CROSS SECTION
Based on the GNASH Hauser-Feshbach statistical/preequilibrium
calculations, described above. Note that MT=91 thresholds at 0.63
MeV. Therefore, discrete states with excitation energies above
0.63 MeV (MT=69-90) lie in the MT=91 continuum region.
MT=102: NEUTRON RADIATIVE CAPTURE CROSS SECTION
The radiative capture cross section above the resonance region
is taken from the ENDF/B-VI evaluation.
----- REFERENCES (MF=3) -----
Ar88 E.D. Arthur, LA-UR-88-382 (1988).
Ba89 M. Baba, H. Wakabayashi, N. Itoh, K. Maeda, and N. Hirakawa,
"Measurements of Prompt Fission Neutron Spectra and Double-
Differential Neutron Inelastic Scattering Cross Sections for
238-U and 232-Th," IAEA Int. Nucl. Data report INDC(JPN)-129
(1989).
Be02 L. A. Bernstein, J. A. Becker, P. E. Garrett, W. Younes, D.
P. McNabb, D. E. Archer, C. A. McGrath, H. Chen, W. E.
Ormand, M. A. Stoyer, R. O. Nelson, M. B. Chadwick, G. D.
Johns, W. W. Wilburn, M. Devlin, D. M. Drake, and P. G.
Young, Phys. Rev. C 65, 021601(R).
Ca73 J. Cabe et al., CEA-R-4524 (1973).
De89 H. Derrien and G. de Saussure, Oak Ridge National Laboratory
report ORNL-10986 (1989).
Du71 C.L.Dunford, "Compound Nucleus Reaction Analysis Programs
COMNUC and CASCADE," Atomics International North American
Rockwell report AI-AEC-12931 (document no. TI-707-130-013)
(February, 1971).
Fo71 D. Foster and D. Glasgow, Phys.Rev. C3, 576 (1971).
Fr80 J. Frehaut et al., Nucl. Sci. Eng. 74, 29 (1980).
He80 D. M. Hetrick and C. Y. FU, "GLUCS: A Generalized Least-
Squares Program for Updating Cross-Section Evaluations with
Correlated Data Sets," Oak Ridge National Laboratory report
ORNL/TM-7341 (ENDF-303) (1980).
Ku70 P.D. Kunz, DWUCK: A Distorted-Wave Born Approximation
Program, unpublished report (1970).
Li90 P.W. Lisowski, personal communication of a measurement
done at WNR in 1985 (1990).
Lo02 R. W. Lougheed et al., personal communication, 2002.
Ma72 D. S. Mather et al., EANDC(UK) 142-AL, 1972.
Na73 K. Nadolny et al., USNDC-9 (1973)p.170
Pe60 J.Peterson et al., Phys.Rev.120, 521(1960).
Pe69 R.J.Peterson, Ann.Phys. 53, 40 (1069).
Po81 W.Poenitz et al., Nuc.Sci.Eng.78, 333(1981).
Po83 W.Poenitz et al., ANL-NDM-80, 1983.
Pr05 V.G. Pronyaev, S.A.Badikov, A.D. Carlson, Z. Chen, E.V. Gai,
G.M. Hale, F.-J. Hambsch, H.M. Hofmann, T. Kawano, N.M.
Larson, S.-Y. Oh, D.L. Smith, S. Tagesen, and H. Vonach,
personal communication (2005); see also: "An International
Evaluation of the Neutron Cross Section Standards," to be
published as an IAEA Technical report (2006).
Ra70 J.Raynal,IAEA SMR-9/8 (1970).
Ra94 J. Raynal, "Notes on ECIS94," Centre d'Etudes Nucleaires
(Saclay) report CEA-N-2772 (1994).
Sc74 R.Schwartz et al., Nuc.Sci.Engr.54,322(1974).
Sh78 R. Shamu et al., private communication, 1978.
Sm73 A. Smith et al., J.Nuc.En. 27, 317 (1973).
Yo94 P. G. Young, "Experience at Los Alamos with Use of the
Optical Model for Applied Nuclear Data Calculations," Los
Alamos National Laboratory report LA-UR-94-3104 (1994).
Yo98 P. G. Young, E. D. Arthur, and M. B. Chadwick,
"Comprehensive Nuclear Model Calculations: Theory and Use of
the GNASH Code," Proc. Workshop on Nuclear Reaction Data and
Nuclear Reactors, ICTP, Trieste, Italy, 15 April - 17 May
1996 [Ed: A. Gandini and G. Reffo, World Scientific Publ.
Co., Singapore (1998)] p. 227-404.
>> MF=4 ANGULAR DISTRIBUTIONS OF SECONDARY PARTICLES
MT=2: NEUTRON ELASTIC SCATTERING ANGULAR DISTRIBUTIONS
1.e-5 eV - 20.0 MeV: Elastic scattering angular distribution
based on ECIS70 (Ra70, Yo94) coupled-channels calculations, with
a compound elastic component from COMNUC included below 6 MeV
(Du71). These data are taken from ENDF/B-VI.
MT=51-55,57: DISCRETE INELASTIC LEVEL ANGULAR DISTRIBUTIONS
(GROUND-STATE ROTATIONAL BAND)
Thres. to 20 MeV: Coupled-channel optical model calculations
plus compound-nucleus contributions. Compound elastic component
from COMNUC included below 6 MeV (Du71). These data are taken
from ENDF/B-VI.
MT=56,58-63,65,67,68: DISCRETE INELASTIC LEVEL ANGULAR
DISTRIBUTIONS (COMPOUND NUCLEUS REACTIONS
TO COMBINED LEVELS)
Threshold to 6.0 MeV, Compound nucleus reaction theory
calculation with width fluctuations using the COMNUC code. These
were taken from ENDF/B-VI. For the combined levels (see MF=3,
MT=56,58-63,65,67,68 above), the angular distributions for the
first MT number were used in the level combinations.
MT=64,66,69-71: DISCRETE INELASTIC LEVEL ANGULAR DISTRIBUTIONS
(DIRECT REACTIONS TO INDIVIDUAL STATES)
Threshold to 20 MeV, distorted wave Born approximation
calculations with the DWUCK code (Ku70) for l=2 and l=3
vibrational states. Compound-nucleus contributions are included
for MT=64 and 66. These data are taken from ENDF/B-VI.
MT=72-90: DISCRETE INELASTIC LEVEL ANGULAR DISTRIBUTIONS (DIRECT
REACTIONS TO GROUPS OF STATES)
The angular distributions are obtained from DWBA calculations
run with the ECIS94 code (Ra94), assuming a set of vibrational
levels as described under M=3, MT=72-90 above. Deformation
parameters were determined by matching the 14-MeV Baba data
(Ba89), and the angular distributions are generally consistent
with those of the Baba data.
----- REFERENCES (MF=4) -----
Ba89 M. Baba, H. Wakabayashi, N. Itoh, K. Maeda, and N. Hirakawa,
"Measurements of Prompt Fission Neutron Spectra and Double-
Differential Neutron Inelastic Scattering Cross Sections for
238-U and 232-Th," IAEA Int. Nucl. Data report INDC(JPN)-129
(1989).
Du71 C.L.Dunford, "Compound Nucleus Reaction Analysis Programs
COMNUC and CASCADE," Atomics International North American
Rockwell report AI-AEC-12931 (document no. TI-707-130-013)
(February, 1971).
Ku70 P.D. Kunz, DWUCK: A Distorted-Wave Born Approximation
Program, unpublished report.
Ra70 J.Raynal,IAEA SMR-9/8 (1970).
Ra94 J. Raynal, "Notes on ECIS94," Centre d'Etudes Nucleaires
(Saclay) report CEA-N-2772 (1994).
Yo94 P. G. Young, "Experience at Los Alamos with Use of the
Optical Model for Applied Nuclear Data Calculations," Los
Alamos National Laboratory report LA-UR-94-3104 (1994).
>> MF=5 ENERGY DISTRIBUTIONS OF SECONDARY PARTICLES
MT=18 PROMPT FISSION NEUTRON SPECTRUM MATRIX
Prompt fission neutron spectrum matrix for the n + 239Pu
system was calculated (Ma01) using the Los Alamos (Madland-Nix)
model (MN82) in its exact formulation with energy-dependent
compound nucleus formation cross sections for the inverse
processes. The matrix includes first-, second-, and third-chance
fission components and also includes the neutrons emitted prior
to fission in second- and third-chance fission. The tabulated
distribution law (LF=1) is used.
The matrix contains vectors for 19 incident neutron energies
and 1 vector for a finite incident neutron energy range. These
are:
0.0, 0.5, 1.0, 1.5, 2.0, 2.5, 3.0, 4.0, 5.0, 6.0, 7.0, 8.0,
9.0, 10.0, 11.0, 12.0, 13.0, 14.0, 15.0, and 16.0--20.0.
The single vector for the upper range of incident neutron
energy is an approximation due to the lack of pertinent
experimental measurements.
The multiple-chance fission average prompt neutron
multiplicity (nubar prompt) was calculated simultaneously and, in
reproducing experiment, was crucial in determining the fission
spectrum matrix.
Modified the original outgoing energy grid to include more points
above 10 MeV. The energy-steps are now 200 keV wide compared to
1 MeV earlier. (P.Talou, LANL, 09/08/2010)
MT=455 DELAYED NEUTRON EMISSION SPECTRA FROM FISSION
The abundances are based on the new delayed-neutron
calculations (Wi05) described under MF=1, MT=455. However, the
earlier six-group spectra from ENDF/B-VI are carried over to
ENDF/B-VII because new multigroup spectra have not yet been
calculated with the new library.
----- REFERENCES (MF=5) -----
Ma01 D. G. Madland, private communication, 30 September 2001.
MN82 D. G. Madland and J. R. Nix, Nucl. Sci. Eng. 81 (1982)
213-271.
Wi05 W. B. Wilson, personal communication, 2005.
>> MF=6 CORRELATED ENERGY-ANGLE DISTRIBUTIONS
MT=16,17,37: (n,xn) CONTINUUM DISTRIBUTIONS
Same as ENDF/B-VI. Based on the GNASH Hauser-Feshbach
statistical/preequilibrium analysis described above. Updated
Kalbach-Mann systematics are used for specifying neutron
distributions (Ka88). Only neutron distributions are given.
MT=91: (n,n') CONTINUUM DISTRIBUTIONS
Same as ENDF/B-VI. Based on the GNASH Hauser-Feshbach
statistical/preequilibrium analysis described above. Updated
Kalbach-Mann systematics are used for specifying neutron
distributions (Ka88). Only neutron distributions are given.
----- REFERENCES (MF=6) -----
Ka88 C. Kalbach, "Systematics of Continuum Angular
Distributions: Extensions to Higher Energies," Phys.Rev.C
37, 2350 (1988); see also C. Kalbach and F. M. Mann,
"Phenomenology of Continuum Angular Distributions. I.
Systematics and Parameterization," Phys.Rev.C 23, 112
(1981).
>>MF=12,13,14,15 PHOTON-PRODUCTION DATA
All photon-production data were carried over from ENDF/B-V.2,
MAT=1399. Data are given for MF=12, MT=4,18,102; MF=13, MT=3;
MF=14, MT=3,4,18,102; and, MF=15, MT=3,4,18,102. The ENDF/B-V.2
data were taken from Hunter and Stewart (Hu72, Hu73). The cross
sections are based upon the neutron files (MF=2,3) and calculated
multiplicities. Data above 1.09 MeV are based on measurements by
Drake (Dr71) and Nellis (Ne71).
One modification was made to the ENDF/B-VI data. The yield for
MF=12,MT=18 is changed from 8.095 to 7.7833 in order to maintain
consistency with the MF=1,MT=458 data.
----- REFERENCES (MF=12-15) -----
Dr71 D. Drake, Los Alamos National Lab. personal communication
(1971).
Hu72 R. E. Hunter and L. Stewart, LA-4901 (1972).
Hu73 R. E. Hunter et al., LA-5172 (1973).
Ne71 D. O. Nellis, P. S. Buchanan, G. H. Williams, and J. A.
Stout, Texas Nuclear Corp. report ORO-2791-32-7102, 1971.
>> MF=33 CROSS-SECTION COVARIANCE MATRICES
>>> MF=33 RESONANCE PARAMETER COVARIANCE (ORNL)
See explanations above.
>>> MF=33 FAST-ENERGY COVARIANCE (LANL)
Cross-section covariance matrices in the fast energy region were
obtained through a combination of nuclear reaction sensitivity
calculations using the GNASH code, and available information on
experimental uncertainties for some of the cross sections. The
sensitivity of GNASH results on the choice of model parameters
was assessed, and the KALMAN code (Ka97) was used to merge
sensitivity calculations with experimental uncertainties.
More details on this work can be found in (Ta08).
--------- REFERENCES (MF=33) ------------
Ka97 T.Kawano and K.Shibata, "Covariance Evaluation System"
JAERI-Data/Code 97-037, 1997 (in Japanese).
Ta08 P.Talou, T.Kawano and P.G.Young, "Covariance matrices for
ENDF/B-VII 235,238U and 239Pu evaluated files in the fast
energy range", Proc. of the ND2007 Conference, April 22-27,
2007, Nice, France. [Los Alamos report LA-UR-07-2501].
************************ C O N T E N T S ***********************
**************** Program DICTIN (VERSION 2007-1) ****************
|