**64-Gd-156 Review**

By V.G. Pronyaev, NDS, IAEA

21 February 2003

**General Comment:**

Taking into account that the accuracy of nuclear data for fission
products is most important in neutron transport and nuclear
transmutation calculations, the following priority to the quality of the
data can be given when we compare different evaluations:

1.  Most important energy region is a region of spectrum with maximum
    neutron fluxes in thermal and fast fission reactors, namely around
    0.0253 eV and below 5 MeV.

2.  Most important cross sections in order of reducing of their
    importance are integral cross sections: capture, total, elastic,
    inelastic with excitation of levels and other threshold cross
    sections; differential cross sections: secondary neutron spectrum
    from inelastic scattering and elastic scattering differential cross
    sections, gamma production cross section and spectra.

##### **Files**

ENDF/B-VI -- 1977 evaluation. The resolved resonance region with 30 MLBW
s- resonances taken from Mughabghab73 is given for En 10^-5^ to 983.4
eV. No unresolved resonance energy region is given. Results of optical
and statistical model calculations with total cross section as for
natural Gd are given for En above 983.4 eV.

JEF-2 is 1982 evaluation with 36 MLBW s- resonances in the RRR below
1.58 keV and URR with s-, p- and d- waves and energy-independent
parameters. Results of optical and statistical model calculations
without account of direct and pre-equilibrium processes are given for
higher neutron energy.

JENDL-3.2 (and JENDL-3.3) -- 1990 revision of 1984 JENDL-2 evaluation.
90 MLBW s- resonances in the RRR below 2.214 keV are taken from
Karzhavina69 and Coceva73. S-, p- and d- wave average resonance
parameters in the unresolved resonance region between 2.214 keV and 100
keV were chosen according Mughabghab84 recommendations and fitting to
the experimental cross sections. Results of spherical optical,
statistical and pre-equilibrium model calculations without account of
direct processes in the inelastic scattering with excitation of discrete
levels are given in the fast neutron energy.

BROND-2 -- 1989 evaluation with 105 RM s- resonaces in the energy range
up to 2.265 keV (RRR is from 10^-5^ to 983.4 eV). No URR is given.
Results of optical, statistical, pre-equilibrium decay and direct
reaction models are given for En above 983.4 eV. Secondary neutron
spectra are given in the temperature approximation.

CENDL-3 -- 2000 evaluation with RRR and URR parameters generally taken
from JENDL-3.2 library. Gamma-width of first resonance was modified to
give 1.635 b capture cross section at 0.0253 eV. Optical, statistical
and pre-equilibrium decay model calculations are used in evaluation of
cross section above 100 keV. No direct processes with excitation of
levels in the inelastic neutron scattering are accounted. Spline
function fit of experimental data was used for (n,2n) cross section and
for (n,) cross section in the energy range 100 keV -- 1 MeV.

###### Thermal and resonance region

Table below shows the cross section at 0.0253 eV (barns), effective
scattering radius R' (fm), RI (barns) and fast rector spectrum averaged
capture cross section (capture fast) for all libraries:

  --------- ----------- --------- ----------- --------- ---------- ----------------
            ENDF/B-VI   JEF-2     JENDL-3.2   BROND-2   CENDL-3    Mughabghab2002
                                  (and 3.3)                        

  Total     7.348       7.109     7.008       6.307     6.455      

  Elastic   5.638       5.573     4.820       4.251     4.820      

  Capture   1.710       1.536     2.188       2.056     1.635      1.8±0.7

  R' (RRR)  8.1         8.1       8.1         8.1       8.1        

  R' (URR)  \-          8.1       7.363       \-        7.363      

  RI        105.1       100\.     121\.       114.3     99.9       104.±15.

  Capture   0.63576     0.69938   0.60532     0.71018   0.68717    
  fast                                                             
  --------- ----------- --------- ----------- --------- ---------- ----------------

Thermal capture cross section was increased in Mughabghab2002 to 1.8 b
from 1.5 b in Mughabghab84 but uncertainty is still large. There are no
data on total cross section in thermal energy region. JENDL-3.2 and
CENDL-3 have more wide energy region for resolved resonances than other
libraries. In the unresolved resonance region, Wisshak95 measurements
for Gd isotopes show that the total cross section is underestimated at
10-15% in all libraries for exclusion of JEF-2. The reason is that
scattering radius R' is adjusted on Gd-natural cross section and is
taken too low in JENDL-3.2 and CENDL-3 for low-A isotopes. The spherical
optical model used in other libraries is not well justified for keV
energies of incident neutrons with the model parameters adjusted on
total cross section for natural Gd without account of (N-Z) isotopic
term. This underestimation should be seen also for elastic scattering.
ENDF/B-VI, BROND-2 and CENDL-3 give better description of the
experimental capture cross section in the URR.

###### Fast region

Fast region covers the energy above 100 keV. The energy dependence of
the optical model parameters used in JENDL-3.2 library overestimate the
total cross section in high energies. Spline used in CENDL-3 for capture
cross section below 1 MeV gives best fit of experimentally observed
cross sections but smoothes the structures (cusps) caused by competition
between capture and inelastic scattering channels. ENDF/B-VI model
calculation gives good fit of capture cross section in all energy range
and CENDL-3 gives too low values for En above 3 MeV, probably because
has too high level density for continuum inelastic scattering channel or
too low level density or gamma-width for the capture channel. ENDF/B-VI
and BROND-2 uses the emission cross section of first neutron as
inelastic scattering cross section MT=4. Because of required consistency
between total and partials, this leads to wrong elastic cross section
for En above (n,2n) reaction threshold. The inelastic scattering cross
section with excitation of discrete levels is too low in all libraries
for exclusion of BROND-2 because of direct processes with excitation of
low-laying collective levels were not accounted. Optical model
parameters in CENDL-3 library give about 10% higher values of
non-elastic (and reaction) cross section comparing with experimental
data for neighboring nuclei. Temperature presentation of the secondary
neutron spectra in ENDF/B-VI and BROND-2 is good only below few MeV of
incident neutron energy. MF=4 presentation of secondary angular
distributions for continuum of levels in CENDL-3 library is rather
useless, because it does not account the existing energy-angular
correlations. MF=6 presentation should be used. Spline fit on
experimental data is used in CENDL-3 for (n,2n) reaction. Other
threshold reactions have rather low cross sections and they are usually
normalized on the experimental data at 14 MeV energy point or on
systematics at this point.

###### Recommendation and conclusion

The CENDL-3 can be recommended as the best in the whole energy region
10^-5^ eV -- 20 MeV.

###### Further possible improvements

R' and other average parameters in the URR should be taken consistent
with the average for those from RRR and adjusted to the isotopic elastic
scattering and total cross section. Total and elastic cross section in
the URR should be revised. Deformed potential should be used in the
optical model calculations with account of low-laying collective states.
The reasons of reaction (inelastic scattering, non-elastic) cross
section should be reduced. The contribution of direct processes in the
inelastic scattering with excitation of states of collective nature
should be accounted. Secondary neutron energy-angular distributions
should be presented in the file MF=6. The calculations of capture cross
section for En above 2 MeV should be done with re-analysis of level
densities used in the inelastic scattering and capture channels.
Evaluation of gamma-production cross sections and spectra should be
done.