Back
9.424000+4 2.379916+2 1 1 2 0
0.000000+0 0.000000+0 0 0 0 6
1.000000+0 3.000000+7 0 0 10 31
0.000000+0 0.000000+0 0 0 573 1
94-Pu-240 BRC,CAD EVAL-JUL04 Bouland Derrien Morillon Romain
DIST-MAY05 REV1-MAY05 20050504
----JEFF-31 MATERIAL 9440
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
***************************** JEFF-3.1 *************************
** **
** Original data taken from: JEFF-3.0 + New eval. **
** **
******************************************************************
05-01 NEA/OECD (Rugama) 8 delayed neutron groups
Jefdoc-976(Spriggs,Campbel and Piksaikin,Prg Nucl Eng 41,223(2002)
******************************************************************
JEFF-3.1 evaluation above the unresolved resonance region
based on model calculations, from 40 keV to 30 MeV.
B. Morillon and P. Romain
CEA/DAM Bruyeres-le-Chatel
New resonance range evaluations:
by H.Derrien, O.Bouland in the Resolved Range (see below)
by O.Bouland in the Unresolved Range (see below)
CEA/DEN Cadarache
Old modifications (relevant anymore?):
by S.Masetti. A.Ventura.
ENEA (Bologna)
by A.Trkov.
IAEA (Vienna)
******************************************************************
SUMMARY OF REVISED PARTS
******************************************************************
MF=1 General Information
The prompt fission neutron multiplicity and spectra
are calculated using the BRC improved Los Alamos model from
Vladuca and Tudora [1]. The model parameters are slightly
different from those adopted in [1]. The prompt fission
neutron multiplicity is obtained from an energetic balance
ratio. The available energy (the average fission energy
released minus the average fission fragment kinetic energy
minus the average prompt gamma ray energy) is divided by the
energy carry away by the neutron (the average fission
fragment neutron separation energy plus the average
center-of-mass energy of the emitted neutrons). The main
improvement is the dependence of the average total
fission-fragment kinetic energy and the average gamma energy
on neutron incident energy.
MT=452 Total Nubar. Sum of MT=455 and 456
MT=455 Delayed Neutron Yields. BRC modified ENDF/B-VI r7
MT=456 Prompt Neutron Yields.
Vladuca and Tudora BRC improved Madland-Nix model
MT=458 Energy Release. BRC modified JEFF3.0
MF=2 Resonance Parameters
MT=151 RESOLVED RESONANCE PARAMETERS UP TO 5.7 KeV
H.Derrien and O.Bouland (January,1996)
NSE 127,105-129(1997)
JEF/DOC-551 ORNL/TM-13450
Note: The multigroup capture and fission cross sections values
published in NSE 127 (Tables III and V respectively) have been
superseeded by the values published in JEF/DOC-551. These updated
values are also available from ORNL/TM-13450.
----------------------------
Work supported by EDF(France),CEA(France) and OECD
----------------------------
The JEF-2 validation has been recently performed on a
large integral data base including thermal and fast critical
data [2]. It was found that the capture and the fission cross
sections of 240Pu could be significantly too large particularly
in the resolved resonance region. The resonance parameters
proposed in the present file were obtained by a sequential
SAMMY analysis of existing experimental data. The input
parameters of the analysis where those found in the ENDF/B-VI
file in the energy range from thermal to 5700 eV. The reaction
formalism used in SAMMY is the Reich-Moore formalism. The 240Pu
cross sections could be represented by the multilevel Breit-
Wigner formalism in the energy ranges between the class II
states; but the Reich-Moore representation is very useful in
the resonances near the class II states where the fission
widths could be very large. In the next section the main
results of the new evaluation are given. More details can
be found in reference[3].
A consistent SAMMY fit of Block[4] total cross section,
Harvey transmission[5] and Leonard fission data[6] was performed
in the energy range from 0.02 eV to 1.5 eV in order to obtain
the values of the cross sections at 0.0253 eV and the parameters
of the resonance at 1.056 eV which contributes to more than 90%
to the capture resonance integral. The parameters obtained for
this resonance are very close to those obtained by Spencer[7] and
used in JEF-2 and ENDF/B-VI. The values of the cross sections at
0.0253 eV are the following:
Total 288.66 b
Scattering 2.67 b
Capture 285.93 b
Fission 0.059 b
In the energy range from 10 eV to 5700 eV, the SAMMY fits
were performed on the experimental transmissions of Kolar[8]
two thicknesses in the energy range 20 eV to 700 eV and one
thickness in the energy range 20 eV to 5700 eV and on the
experimental fission cross sections of Weston[9]. Some
preliminary fits were performed on the transmissions of Kolar
in order to check the normalization and background correction
parameters and the experimental resolution parameters. Compared
to the current evaluated data files, much more resonances were
used in the present evaluation, particularly above the energy
of 1500 eV. These added resonances are resonances with small
neutron widths which could be identified above the small
background in the experimental fission data or in the
experimental statistical fluctuations of the transmission
data. This attempt to identify the small resonances in the
high energy region of the data leads to a more realistic
average value of the resonance spacing over the entire energy
range of the analysis and allows to avoid the use of a smooth
background cross section in the high energy range. The
comparison between ENDF/B-VI and the present evaluation is given
in the following Table, for the strength function and the number
of resonances:
Energy Strength Number of
Range eV Function Resonances
Present ENDF/B-VI Present ENDF/B-VI
0- 500 1.089 1.102 42 36
500-1000 1.049 1.027 42 33
1000-1500 1.021 1.008 45 32
1500-2000 1.221 1.167 39 26
2000-2500 0.993 0.911 40 25
2500-3000 1.041 0.948 36 21
3000-3500 0.731 0.628 37 17
3500-4000 0.661 0.539 34 16
4000-4500 1.215 0.952 35 18
4500-5000 1.032 0.896 31 18
5000-5700 1.206 1.047 44 25
The low values of the strength function between 3 keV and
4 keV, whicy not consistent with the values in other energy
ranges (sampling error of about 22%), are not due to the
missing resonances in the corresponding energy ranges. The
same kind of fluctuations of the local values of the strength
function are also observed in 238U data[11] and others nuclei.
The average values of the capture cross section are given
in the following Table:
Energy Range Weston[11] Present B-VI JEF-2 JENDL-3
eV b b b b b
0.02- 1.5 5922 5930 5897 5652
1.5- 50 56.85 56.83 55.34 57.33
50- 100 49.96 48.40 48.57 48.71
100- 200 23.30 24.64 25.57 25.64
200- 300 8.71+/-0.61 7.27 7.41 9.07 9.08
300- 400 10.27+/-0.72 7.93 7.89 9.92 9.94
400- 500 6.60+/-0.46 6.01 5.97 7.02 7.03
500- 600 7.14+/-0.50 6.22 5.91 7.15 7.16
600- 700 5.09+/-0.36 4.44 4.64 4.65 4.65
700- 800 2.63+/-0.18 2.04 1.64 3.31 3.31
800- 900 6.63+/-0.46 5.70 5.25 5.31 5.31
900-1000 5.53+/-0.39 5.75 5.47 6.15 6.15
1000-1500 3.50+/-0.25 3.13 2.89 3.46 3.46
1500-2000 3.03+/-0.21 2.52 2.24 3.05 3.05
2000-3000 2.42+/-0.17 1.90 1.54 2.40 2.40
3000-4000 1.89+/-0.13 1.20 1.29 1.89 1.90
4000-5000 1.67+/-0.12 1.13 1.55 1.76 1.75
5000-5700 0.95 1.54 1.60 1.60
0.02- 200 81.76 82.09 81.99 80.73
200-5000 3.02 2.42 2.37 3.03 3.03
These average cross sections were calculated by NJOY-94.0
at the NEA Data Bank.
In the energy range 200 eV to 5000 eV, the values of
JEF-2 and JENDL-3 were normalized to the experimental values
of Weston. The values of the present evaluation and of ENDF/B-VI
are 25% and 27% lower respectively. One should note that Weston
in his evaluation for ENDF/B-VI[12] did not take into account
his own experimental data.
The average values of the fission cross section are given
in the following Table:
Energy Range Present B-VI JEF-2 JENDL-3
eV mb mb mb mb
0.02- 1.5 1649 1170 1140 1048
1.5- 50 91 94 381 94
50- 100 74 76 346 76
100- 200 46 50 337 50
200- 300 52 53 222 53
300- 400 15 18 228 18
400- 500 47 49 188 49
500- 600 20 23 185 21
600- 700 54 54 208 66
700- 800 879 905 1020 938
800- 900 698 615 693 613
900-1000 86 80 155 75
1000-1500 206 199 257 147
1500-2000 316 297 422 312
2000-3000 210 181 332 242
3000-4000 75 74 116 6
4000-5000 60 50 88 67
5000-5700 150 145 91 124
1.5 -5700 159.5 149 228 158
These average cross sections were calculated by NJOY-94.0
at the NEA Data Bank.
In the energy range 1.5 eV to 700 eV, the average fis-
sions of J are much larger than the other values. The present
results are in general consistent with B-VI and JENDL-3. Note
an inconsistent value in JENDL-3 in the energy range 3000 eV to
4000 eV.
The capture and fission resonance integrals are given
in the following Table:
Present B-VI JEF-2 JENDL-3
b b b b
Capture 8481 8494 8445 8102
Fission 3.16 2.46 3.52 2.29
For the capture, the difference between JENDL-3 and the
others is mainly due to a smaller value of the neutron width of
the resonance at 1.056 eV.
The comparison of the present results with JEF-2 shows
a significant decrease of the capture cross section and of the
fission cross section in the resolved energy range,in agreement
with the tendancy observed in the validation of the JEF-2 general
purpose file[2].
MF=2
MT=151 Unresolved Resonance energy Range between 5.7-40 KeV
O.Bouland (April,2002)
(see JEF/DOC-917)
----------------------------
Work supported by CEA(France), EDF(France)
----------------------------
In the unresolved region, the choice was made to tabulate the
entire dilute pointwise cross sections in file mf=3 because the
use of the parameters with ENDF processing does not lead to cross
sections consistent with mf=3 as the codes use a more primitive
version of the formalism. The unresolved parameters given in mf=2
mt=151 are only to be used for self-shielding calculations (flag
LSSF set to 1).
AVERAGE TOTAL CROSS SECTION ADJUSTEMENT:
From the experimental data selected in the unresolved range[13,
14,15] and a prior estimate of the average parameters, the cal-
culated average total cross section was fitted with the Bayesian
code FITACS[16] which employs Hauser-Feshbach theory and Moldauer
prescription for overlapping resonances. The posterior average
resonance parameter values obtained are presented in the Table
below. The uncertainties which are given reflect only the sta-
tistical uncertainties on the experimental data and the quality
of this adjustment.
_______________________________________________________________
Orbital | Strength | Distant-level | Mean level |Effective|
Angular | function | parameter | spacing |Radius |
Momentum | | (R_c^infinity | | |
(hbar) | (1/10000) | | (eV) | (fm) |
---------|--------------|---------------|------------|---------|
0 | 1.102+-0.052 | 0.034+-0.011 | 13.43 | 9.10 |
1 | 1.842+-0.083 | 0.284+-0.028 | | |
2 | 1.030+-0.121 | 0.046+-0.027 | | |
3 | 2.022+-0.135 | 0.126+-0.092 | | |
_______________________________________________________________
AVERAGE CAPTURE CROSS SECTION ADJUSTEMENT:
By reference to the very large amount of work on the validation
of the JEF2.2 general purpose file[17], it appears that among the
set of experimental capture data available (Weston and Todd[11],
Wisshak and Kappeler[18] and Hockenbury et al.[19]) none of them
were acceptable in magnitude; even the most satisfactory one
(Weston and Todd) being too high of about (7+-8) percent on ave-
rage in the energy range (5.7-1000 keV). Since the conclusion of
the recent 1995 re-evaluation[3] of the resolved range was also
to decrease the average capture cross section (20% too high
in the energy range 200-5000 eV), a significant decrease of the
average capture cross section in the present work has been una-
voidable. Keeping the most adequate capture data set (Weston and
Todd) but renormalised, a fit of the capture width of the s-, p-
and d- waves was performed starting from the previously fitted
neutron channel average parameters and from the fission channel
parameters determined in parallel (see next section). In order
to keep reasonable the fitted value of the s- wave radiative cap-
ture width, a renormalisation factor of only -12% was applied to
the Weston and Todd capture measurement. Table below presents
the chosen prior and the fitted posterior values for the various
average capture widths involved in this work.
___________________________________________________________
| | gGamma^0_gamma| gGamma^1_gamma | gGamma^2_gamma |
| | (meV) | (meV) | (meV) |
|---------|---------------|----------------|----------------|
| Prior | 31.92+-1.6 | 31.92+-10. | 31.92+-10. |
|---------|---------------|----------------|----------------|
|Posterior| 30.7+-2.5 | 22.53+-5. | 30.7* |
|_________|_______________|________________|________________|
* Due to the conception of the FITACS code, the average capture
width of the d- wave resonances is not a fitting parameter and is
driven by the s- waves average capture width.
AVERAGE FISSION CROSS SECTION ADJUSTEMENT:
Unfortunately the shape of the 240Pu fission cross section is in-
compatible with the single-humped fission barrier model available
in the FITACS code and since the partial cross sections are inter
-connected through the total transmission coefficient, the quest
of a specific program for treating the sub-threshold fission had
been required. In a FIRST REPRESENTATIVE APPROACH the calculation
of the fission cross section has been achieved with the AVXSF pro
-gram of LYNN[20] including a double-humped fission barrier with
moderately weak coupling between the class-II states and the nor-
mal compound nuclear (class-I) resonances. Due to the large num-
ber of parameters involved in the calculation of the sub-thres-
hold fission cross section, no fitting method was actually possi-
ble and thus a trial-error procedure was adopted. Since the pro-
gram AVXSF uses some approximations in the calculation of both
neutron and photon channel transmission coefficients, an itera-
tive procedure which involves the two codes FITACS and AVXSF, was
set up.
The calculation of the average fission cross section using the
AVXSF code has been finally performed in the energy range [5.7keV
- 200 keV]. But, although the AVXSF calculation includes a double
-humped fission barrier and a representative coupling between the
class-II and class-I states, this current modelisation of the
class-II states gives only an average effect on the calculated
fission cross section. From the many sets of experimental average
fission cross section data available in the literature, one sees
very well that the experimental data, even with a poor resolu-
tion, show a gross structure which can not be reproduced by the
formalism proposed by AVXSF. In some other fission cross sec-
tion measurements in the so-called 'unresolved energy range' abo-
ve 5.7 keV such as in the Weston and Todd data[9], a very fine
structure due to partially resolved class-I states appear in the
envelope of the intrinsic class-II states. At higher energy the
class-I states are no more resolved and the class-II states beco-
me badly resolved and thus only the gross structure shows up in
the fission cross section.
Therefore for JEFF3.0, it was decided to follow a PRAGMATIC AP-
PROACH for the fission cross section from only the JEF2.2 vali-
dation trends[21]. So the fictitious fission cross section, now
simulated in the energy range [5.7keV - 40keV], is satisfactory
only in a neutronic sense. Such an approach was made possible
because the 240Pu fission cross section remains very small below
100 keV and subsequently it has no real effect on the adjustment
of the other partial cross sections. Moreover the high quality of
a nuclear model, reproducing the resonances observed in the fis-
sion cross section, is somehow distorted by the transcription of
these evaluated data in ENDF-6 format. The fission cross section,
resulting of this PRAGMATIC approach, exhibits 3 step-like func-
tions covering the (5.7-40) keV energy range with the 9.12 keV
and 24.8 keV boundaries belonging to the ERALIB energy group
structure[21].
COMMENTS ON THE TOTAL CROSS SECTION VALUE:
The Table below highlights the wrong values of the total cross
sections predicted by the jef2.2 (hereby called Ref) and the
jendl3.2 data files in the unresolved range. This work (namely
jeff3.0) decreases significantly the value of the total cross
section and is in agreement with the ENDF/B-VI.5 prediction.
_________________________________________________________________
Sigma_t |(BVI.5-Ref) |(jeff3.0-Ref) |(jendl3.2-Ref)|(ERALIB-Ref)|
| --------- | ----------- | ------------ | ---------- |
Energy | Ref | Ref | Ref | Ref |
range | | | | |
(keV) | (%) | (%) | (%) | (%) |
_________|____________|______________|______________|____________|
5.7-9.12 | -9.1 | -9.6 | 1.3 | No |
_________|____________|______________|______________| ERALIB |
9.12-24.8| -12.4 | -12.0 | 1.9 | trends |
_________|____________|______________|______________| |
24.8-40. | -12.2 | -11.5 | -0.6 | |
_________|____________|______________|______________|____________|
COMMENTS ON THE CAPTURE CROSS SECTION VALUE:
The Table below well exhibits the wrong value of the capture
cross sections calculated from any of the current evaluated data
file. All of them are based on a too large value of the s-waves
average capture width. The decrease of the s-waves average cap-
ture width in this work has made possible an agreement with the
ERALIB trends but the target of -20 percent suggested by the 1995
study[3] from a lower energy range (200 eV - 5 keV) was impossi-
ble to reach since it would have requested a s-waves average cap
ture width much smaller than 30.7 meV; value recommended by the
co-ordinated research project[22].
_________________________________________________________________
Sigma_gam|(BVI.5-Ref) |(jeff3.0-Ref) |(jendl3.2-Ref)|(ERALIB-Ref)|
| --------- | ----------- | ------------ | ---------- |
Energy | Ref | Ref | Ref | Ref |
range | | | | |
(keV) | (%) | (%) | (%) | (%) |
_________|____________|______________|______________|____________|
5.7-9.12 | -2.3 | -6.5 | 0.86 | -8.9+-7.8 |
_________|____________|______________|______________|____________|
9.12-24.8| 0.09 | -7.7 | 6.1 | -7.6+-7.5 |
_________|____________|______________|______________|____________|
24.8-40. | -0.06 | -8.6 | 15.7 | -5.6+-7.2 |
_________|____________|______________|______________|____________|
COMMENTS ON THE FISSION CROSS SECTION VALUE:
Concerning the fission cross section, this work is in an agreement
with the ERALIB trends as expected. The choice of the average sub
-threshold fission cross section calculated with AVXSF would
have given some similar results to those obtained from ENDF/B-
VI.5. The deviation of +120 percent observed for ENDF/B-VI.5 in
the lowest energy group (5.7-9.12) keV (see Table below) is due
to a strong dip in the measured fission cross section which can
not be represented by an average calculation which follows ap-
proximately a 1/v slope.
_________________________________________________________________
Sigma_fis|(BVI.5-Ref) |(jeff3.0-Ref) |(jendl3.2-Ref)|(ERALIB-Ref)|
| --------- | ----------- | ------------ | ---------- |
Energy | Ref | Ref | Ref | Ref |
range | | | | |
(keV) | (%) | (%) | (%) | (%) |
_________|____________|______________|______________|____________|
5.7-9.12| 120. | -10.7 | 109. | -11.8+-18.|
_________|____________|______________|______________|____________|
9.12-24.8| 13.1 | -3.7 | 15.5 | -7.1+-18. |
_________|____________|______________|______________|____________|
24.8-40.| 5.1 | 0.6 | 13.5 | 0.9+-17. |
_________|____________|______________|______________|____________|
MF=3 Reaction Cross-sections
From the energy of 1 keV up to 200 MeV, six states Coupled
Channel Calculations are performed using the ECIS95 [23] code
which also provides compound nucleus cross sections and trans-
mission coefficients used in pre-equilibrium/evaporation
emission treated in the exciton and Hauser-Feshbach models
implemented in the Bruyeres-le-Chatel modified version of the
GNASH code[24]. This reaction code has been modified to
include width fluctuation factors, relativistic kinematics,
and a more realistic treatment of the fission process.
A fission penetrability model taking into account Double
Humped Fission Barrier has been used, explicitly coupling
class I and II states while damping those of class II.
Emission of light hadrons up to He4 are explicitly treated in
the model calculations. Fission decay of associated residual
nuclei is also treated. However, none of these emissions and
fission cross-sections, up to the (n,4nf), are yet explicitly
provided in this file.
The Resolved Resonance Range, ending now at 40 keV, the
model calculations data are implemented from this energy.
MT=1 calculation from BRC deformed optical potential
over the whole energy range 1 keV-200 MeV.
the results have been validated with existing
experimental neutron reaction cross section data.
MT=2 calculation from BRC deformed optical potential
MT=4 calculation from BRC deformed optical potential
sum of mt=51-91.
MT=16 (n,2n) cross section
MT=17 (n,3n) cross section
MT=18 (n,f) calculation with BRC modified GNASH code, with
a double humped fission barrier penetration model
MT=37 (n,4n) cross-section
MT=51-74(n,n') cross-section for 1st-24th excited states
MT=91 (n,n') continuum cross-section
MT=102 (n,g) cross-section
MF=4 Angular Distributions of Secondary Particles
MT=2 elastic angular distribution, given up to 30 MeV
MT=18 fission given up to 30 MeV (assumed isotropic)
MT=51-74 inelastic levels, 1st-24th excited states
With a uniform number of angular points (91), equal values
of the tabulated probability distributions may occur.
MF=5 Energy Distributions of Secondary Particles
MT-16 Taken from JEFF3.0 and extended from 20 up to 30 MeV
MT=17 " " "
MT=18 " " "
MT=37 " " "
MT=91 " " "
MT=455 " " "
MF=12 Photon Production Multiplicities
MT=4 From ENDF/B-VI.7 and extended from 20 up to 30 MeV
MT=18 " " "
MT=102 " " "
MF=13 Photon Production Cross-section
MT=3 From ENDF/B-VI.7 and extended from 20 up to 30 MeV
MF=14 Photon Angular Distribution
MT=3 From ENDF/B-VI.7 and extended from 20 up to 30 MeV
MT=4 " " "
MT=18 " " "
MT=102 " " "
MF=15 Continuous Photon Energy Spectra
MT=3 From ENDF/B-VI.7 and extended from 20 up to 30 MeV
MT=18 " " "
MT=102 " " "
----------------------------------------------------------------
REFERENCES
[1] G. Vladuca and A. Tudora, Ann. Nuc. Energy. 28, 689 (2001).
[2] E.Fort et al., Gatlinburg Conference,Tennessee,May 9-16,1994
[3] 0.Bouland, H.Derrien et al.,NSE 127, 2, 105-129 (Oct 1997)
[4] R.C.Block et al.,NSE 8,112(1960)
[5] J.H.Harvey,Private communication at ORNL (1995)
[6] B.R.Leonard et al.,Hanford Report Series 67219,4 (1960)
[7] R.Spencer et al.,NSE 96,318(1987)
[8] W.Kolar et al., J.Nucl.Energy 22,299(1968)
[9] L.W.Weston et al.,NSE 88,567(1984)
[10] D.K.Olsen et al.,NSE 94,102(1986)
[11] L.W.Weston et al.,NSE 63,143(1977)
[12] L.W.Weston et al.,ORNL-TM-10386(1988)
[13] R. GWIN, private communication to L.W. Weston (1985)
[14] W.P. POENITZ and J.F. WHALEN, ANL/NDM-80 (1985)
[15] A.B. SMITH et al., Nucl. Sci. Eng., 47, pp. 19-28 (1972)
[16] F.H. FROEHNER, Nucl.Sci.Eng., 103, pp. 119-128 (1989)
[17] E. FORT, ' ERALIB results', Private communication at CEA/
Cadarache (1998-10-08) and 'The JEF2.2 Nuclear Data Library',
JEFF Report 17, part III, NEA (2000).
[18] K. WISSHAK and F. KAPPELER, Nucl.Sci.Eng., 66, p. 363 (1978)
[19] R.W. HOCKENBURY et al., Nucl.Sci.Eng., 49, pp. 153-161 (1972)
[20] J.E. LYNN, Harwell Report AERE-R 7468 (1974).
[21] E. FORT, ' ERALIB results', Private communication at CEA/
Cadarache (1998-10-08) and 'The JEF2.2 Nuclear Data Library',
JEFF Report {/bf17}, part III, NEA (2000).
[22] Handbook for Calculations of Nuclear Data, RIPL, IAEA-TECDOC
(1998)
[23] J. Raynal, "Code ECIS95" CEA report N-2772, (1994).
[24] P.G. Young, E.D. Arthur and M. B. Chadwick, Workshop on
Nuclear Reaction Data and Nuclear Reactors, Trieste,
Italy (1996).
************************ C O N T E N T S ***********************
1 451 578
Back
|
