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4-Be- 9 IRK-VIENNA EVAL-JAN97 VIENNA, OBNINSK
EFF-DOC DIST-JAN09 20090105
----JEFF-311 MATERIAL 425
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
*************************** JEFF-3.1.1 *************************
** **
** Original data taken from: JEFF-3.1 **
** **
******************************************************************
***************************** JEFF-3.1 *************************
The Original data is taken from: EFF3 MOD6 however, due to
unresolved processing/handling problems related to the 8xx MT's
and other considerations, the following actions have been taken
- correction of MF-2
- addition of MF-12-14 MT-102 (ENDF/B-VI r8)
- removal of MF-3, MF-6, MF-33 MT16 (n,2n)
The informations being redundant to those of MT875-890
The partials data have been proved to be of better quality
J-Ch Sublet and S Tagesen
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EUROPEAN FUSION FILE - VERSION 3.0 - NMOD=6 BE-9
******************************************************************
NMOD=6
Addition of MF3 MT102 and corresponding MF33 MT102 based on
the BE-9 N,G KOPECKY-2000 evaluation. Added at the NEA Databank.
Sections revised October 2003 by S. Tagesen and H. Vonach, Vienna.
DATA from EAF-2000, renormalized by a factor of 1.166 according
to a new evaluation of thermal capture cross sections by
S.F. Mughabghab, INDC(NDS)-440, Feb. 2003
NMOD=2
VERSION WITH LAW=7 REPRESENTATION IN MF6, MT16
NMOD=3
added single reaction channels of MT16 in MT875-890
******************************************************************
MF6 MT16 data renormalised
ZERO angular distributions in MF6 MT16 replaced by isotropic ones
MF6 MT16 data converted from LAW=1 to LAW=7
Modifications by A. Hogenbirk, ECN PETTEN, AUG97
******************************************************************
The evaluation was performed using the code GLUCS based on the
Bayesian approach. Evaluated cross sections and their
covariances were derived for the cross sections (n,total),
(n,2n), (n,alpha0), (n,p), (n,d) and (n,t) of 9-Be, which form
a complete set of basic non-redundant cross sections, for the
whole neutron energy range 10-5 to 20 MeV. In addition to the
experimental data base on the mentioned basic cross sections,
experimental data on sigma-el, sigma-nonel and sigma-He-prod,
which can be expressed as linear functions of the basic cross
sections were also included in the evaluation.
Generally the output of the computer code GLUCS was used directly
to assemble MF3 and MF33.
Special treatment was necessary for the total neutron cross
section MT1:
In the incident neutron energy range up to 4.5 MeV the cross
section exhibits considerable structure, which is well established
in the measurements of (Bilpuch 61) and (Schwartz 71). For the
least-squares adjustment procedure it is, however, very
impractical to treat the full original data sets with several
thousand data points each. The adjustment was therefore done
with group average values spanning a 500 keV neutron energy
range each. Next, the adjustment factors calculated by GLUCS
were used to scale the experimental points to the evaluation
results. Finally a thinning and smoothing procedure combining
at least 5 data points was applied, to get a good
representation of the existing structures without reflecting
large statistical fluctuations.
Thus the complete excitation function was assembled in the
following way:
incident neutron energy range data source
10-5 eV - 10 keV ENDF/B-VI
24 keV experimental point (Aizawa 83, Block 75)
55 keV - 490 keV Bilpuch 61, thinned to 5 keV steps
500 keV - 1.4 MeV Schwartz 71, thinned to 5 keV steps
1.4 MeV - 3.0 MeV Schwartz 71, thinned to 10 keV steps
3.0 MeV - 4.5 MeV Schwartz 71, thinned to 25 keV steps
4.5 MeV - 20 MeV GLUCS results, .5 MeV group averages
In addition to the cross section evaluation an evaluation of
the energy and angular distribution of the secondary neutrons
was performed. For this purpose the energy and angular
distributions of all partial reaction channels contributing to
the secondary neutron production (neutron inelastic scattering
followed by further neutron decay of 9-Be levels, (n,alpha)
reactions followed by two neutron breakup of 6-He and various
other three-body breakup reactions) were investigated and their
energy and angular neutron distributions calculated in the
laboratory system. Using this information, the total secondary
neutron energy and angular distribution was expressed as sum of
the distributions for all reaction channels weighted according
to their cross sections, which were used as fit parameters to
adjust the calculated distributions to the experimental data
existing at four energies (5.9, 10.1, 14.1 and 18.0 MeV). For this
purpose the mentioned code GLUCS, after some modification,
could also be used. As a result of this process it was possible
to reproduce the experimental data within their uncertainties
by our model calculations and to derive a set of partial (n,2n)
cross sections and their covariances at the mentioned energies.
By suitable inter- and extrapolation procedures (guided by
theory) subsequently such partial reaction cross sections were
derived for the whole energy range from the (n,2n) threshold to
20 MeV. Using these cross sections the energy and angular
distribution of the secondary neutrons was calculated for the
whole energy range of the evaluation.
Description of double differential spectra of individual reaction
components
Authors: V. Pronyaev, S. Tagesen and H. Vonach
I.R.K. (1998)
Representation of reaction channels for the reaction
9Be + n -> n + n + a + a
16 channels are given for neutron emission,
17 channels given for alpha emission, (n,a0) 6He -> beta only
channels are characterized by the following MT's:
MT 107: 9Be (n, a0) 6He (g.s. => one alpha particle only)
MT 875: inel.scatt. through level at 1.684 MeV
MT 876: inel.scatt. through level at 2.429 MeV
MT 877: inel.scatt. through level at 2.78 MeV
MT 878: inel.scatt. through level at 3.049 MeV
MT 879: inel.scatt. through level at 4.704 MeV
MT 880: inel.scatt. through level at 5.59 MeV
MT 881: inel.scatt. through level at 6.38 MeV
MT 882: inel.scatt. through level at 6.76 MeV
MT 883: inel.scatt. through level at 7.94 MeV
MT 884: inel.scatt. through level at 11.283 MeV
MT 885: inel.scatt. through level at 11.81 MeV
MT 886: 9Be (n, a1) 6He* level at 2.4 MeV
MT 887: 9Be (n, a2) 6He* level at 4.0 MeV
MT 888: 9Be (n, 5He#) 5He# # =: unstable to particle decay
MT 889: 9Be (n, n + n + 8Be#)
MT 890: 9Be (n, n + a + 5He#)
To permit correct calculation of the energy balance, the available
energy QM = -1.574 MeV for all channels MT875 to 890,
corresponding to total disintegration of the system 9Be + n.
QI gives the Q-value of the first step in the respective chain
and thus determines the reaction threshold in that channel.
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