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3.007000+3 6.955732+0 0 0 2 1
0.000000+0 0.000000+0 0 0 0 6
1.000000+0 2.000000+7 0 0 10 31
0.000000+0 0.000000+0 0 0 399 1
3-Li- 7 ECN EVAL-AUG90 BIRMINGHAM, PETTEN, GEEL, LASL
DIST-MAY05 REV1-MAY05 20050504
----JEFF-31 MATERIAL 328
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
***************************** JEFF-3.1 *************************
** **
** Original data taken from: JEFF-3.0 **
** **
******************************************************************
***************************** JEFF-3.0 *************************
DATA TAKEN FROM :- EFF-2.4 (DIST-NOV94 REV-NOV94)
******************************************************************
EUROPEAN FUSION FILE - VERSION 2 Li-7
********************************* User's version with
USER'S VERSION lumped reaction
********************************* mechanism (LIP=0)
PROCESSING NOTES:
This is the users' version with lumped reaction mechanisms
(LIP=0). Processing of this file is possible with NJOY91.91
A basic version, where all reaction mechanisms are
explicitly specified by the LIP parameter, is also
available. Although the format used is strictly ENDF-6,
not all present processing codes may work properly, because
of the fact that the energy-angle distributions are quite
complex for light targets.
Loss of information may occur if processing codes translate
the laboratory frame energy-angle distributions into
Legendre coefficients (for PN group cross-sections).
In order to take advantage of the possibility of MCNP4A to
use MF6 data from a tabulated distribution in LAW=7 format,
MF6 LAW=1 data were converted to LAW=7, for all reaction
products, except for gamma's (ZAP=0.). The latter data
remain in the LAW=1 format, as processing in the LAW=7
format is not allowed currently with NJOY.
Processing with the most recent version of NJOY91 is easily
possible.
******************************************************************
The EFF-2 data file is made in the framework of the European Fu-
sion Programme of the European Community. The data file is main-
tained at the Netherlands Energy Research Foundation ECN, P.O. Box
1, 1755 ZG Petten, The Netherlands (contact: H. Gruppelaar, Nuc-
lear Analysis, Business Unit Nuclear Energy).
The evaluation for Li-7 has been sponsored by a NET contract (Next
European Torus programme).
*****************************************************************-
Authors:
T. D. Beynon, G.M. Field
School of Physics and Space Research, University of Birmingham,
Edgbaston, Birmingham, B15 2TT, U.K.
H. Gruppelaar, D. Nierop
Netherlands Energy research Foundation ECN, P.O. Box 1, 1755 ZG
Petten N.H., The Netherlands
H. Liskien,
Central Bureau of Nuclear Measurements, Geel, Belgium
P. G. Young,
Los Alamos National Laboratory, Los Alamos, New Mexico, USA
SUMMARY OF THE DATA FILE EFF-2:
The primary reference for this evaluation is BE90, based upon
earlier work BE70, BE79, BE88.
The EFF-2 evaluation of Li-7 neutron induced cross sections can be
viewed as an update of the ENDF/B-VI evaluation of Young (YO88),
mainly with respect to a revision of angle-energy coupled neutron
distributions. In fact, there are only small differences in the
excitation functions (MF=3) and no differences in the elastic
scattering angular distributions. However, the description of the
neutron angle-energy distributions is entirely different in EFF-2,
because of the adoption of a new method, storing the data in the
new MF=6 format of ENDF-VI, rather than adopting "pseudo" levels
in MF=3 and MF=4. The neutron distributions of all reactions are
explicitly given; but in the present user's version different
reaction mechanisms have been lumped together (LIP=0). In MF=6
the distributions are specified in the laboratory system in a
tabular form as a function of angle and outgoing energy. In this
version no distributions for other particles than neutrons are
given explicitly. A more detailed description is given below.
Note that there are no changes in the photon-production
cross-sections (MF=12,14). The covariance data (MF=33) have not
been included, although ENDF/B-VI data may still be relevant for
those cross-sections that were not modified.
EXCITATION FUNCTIONS (MF=3)
The angle-energy integrated total cross-section (MT1), elastic
scattering cross-section (MT2), total inelastic scattering
cross-section (MT4), (n,2n) cross-section (MT16), (n,2n)d
cross-section (MT24), (n,3n)p cross-section (MT25) and (n,gamma)
cross-section (MT102) are equal to those of ENDF/B-VI. Likewise,
the total tritium production cross-section (MT52+53+91) and the
inelastic scattering cross section for excitation of the first-ex-
cited state (MT51) are the same.
The main difference is the subdivision of the inelastic scattering
cross section into cross-sections for three discrete levels at
0.478, 4.63 and 6.68 MeV (MT51,52,53, respectively) and a conti-
nuum (MT91), rather than into a mixture of cross-sections to dis-
crete and "pseudo" levels (MT51-82) as adopted in ENDF/B-VI. In
fact, the (n,n')gamma cross-section (MT51) is the same in the two
evaluations, as well the cross-sections for excitation of the
excited state at 4.63 MeV (MT52 and MT56, respectively). The
cross-section for excitation of the 6.68 MeV level (MT53) was ta-
ken from the evaluation of Liskien LI86. It is noted that the
threshold for the continuum (MT91) is below the energy of the
last discrete level (MT53).
Another (small) difference is that the (n,d) cross-section of
ENDF/B-VI has been splitted into an (n,d) part (MT104) and an
(n,np) part (MT28), each of 50% of the ENDF/B-VI values.
ANGULAR DISTRIBUTIONS (MF=4)
Although it is possible to store angular distributions for all
reactions in the new MF=6 format, it was decided to use the new
MF=6 format only for distributions of particles emitted from the
continuum or a broadened level. The neutron distributions of elas-
tic and inelastic scattering cross-sections via "sharp" discrete
states are well described by excitation functions in MF=3 and cen-
ter-of-mass angular distributions in MF=4. In all cases Legendre
coefficients are given.
In the EFF-2 evaluation the elastic (MT=2) angular distribution is
equal to that of ENDF/B-VI. The same applies to the neutron angu-
lar distribution of inelastic scattering to the second excited
state at 4.63 MeV (MT=52, equal to MT56 of ENDF/B-VI). For the
angular distribution of neutrons emitted from the first-excited
state at 0.478 MeV (MT=51) EFF-1 data were used, evaluated by H.
Liskien (LI86). All other angular distributions are included in
the combined angle-energy distributions stored in file MF=6. No
seperate energy distribution file (MF=4) is given in EFF-2. It is
noted that the angular distributions of neutrons emitted from the
level at 6.68 MeV (MT53) are not given here, because of the fact
that this level has a finite natural width.
ENERGY ANGLE DISTRIBUTIONS (MF=6)
The major change with respect to ENDF/B-VI is the re-evaluation of
angle-energy distribution functions for neutrons emitted in the
continuum from the following reactions: (n,2n), (n,2n)d,
(n,3n)t-alpha, (n,np), and (n,n')alpha-t (MT=16,24,25,28,91, res-
pectively). In addition, the angle-energy distributions of
neutrons coming from the excitation of the 6.68 MeV state via
the (n,n')alpha-t reaction (MT53) are specified here, rather than
in MF=4. The reason is that the 6.68 MeV state has a significant
natural width, and it appears as a broadened peak in the emission
spectrum.
The format of MF=6 was chosen to allow a physical description of
angle-energy coupled neutron distributions. The method of using
"pseudo" levels (used in ENDF/B-VI) can do the same job, but its
mayor drawback is that the levels are not real and therefore it is
not an optimal physical description. In the present evaluation
the following subdivision has been made:
__________________________________________________________________
MT= 16 (N,2N) CROSS SECTION (THREE-BODY BREAK-UP)
MT= 24 (N,2N)D-ALPHA CROSS SECTION (THREE-BODY BREAK-UP
FOLLOWED BY DECAY OF FIRST-EXCITED STATE IN LI-6)
NOTE: DOUBLE SEQUENTIAL BREAK-UP TROUGH HE-6 AND
HE-5 HAS BEEN NEGLECTED.
MT= 25 (N,3N)ALPHA-P CROSS SECTION (MULTI-BODY BREAK-UP)
MT= 28 (N,NP) CROSS SECTION (THREE-BODY BREAK-UP)
MT= 53 (N,N')T-ALPHA CROSS SECTION (INELASTIC
SCATTERING FOLLOWED BY DECAY OF EXCITED STATE
WITH NATURAL WIDTH)
MT= 91 (N,N')T-ALPHA CROSS SECTION
_________________________________________________________________
Throughout subfile MF6 the EFF-2 data are stored using LAW=7
(Continuum angle-energy distributions), adopting tabular
representations of neutron yields and tabulated laboratory energy
distributions as well as outcoming energy distributions for each
laboratory angle (21 equidistant values of the cosine of the
scattering angle).
A Cartesian linear-linear interpolation scheme is prescribed for
outgoing energy and cosine of scattering angle.
For a detailed discussion of the modelling the user is referred to
BE90. The reaction kinetics determines the distributions, once
the excitation functions for the various processes and the
center-of-mass angular distributions are known. These parameters
are largely determined by experimental (or evaluated) data. If
not, they were varied within their physical boundaries to obtain
an optimum fit with experimental angle-energy distributions. With
respect to ENDF/B-VI only very few additional data (DE87, SH89,
CH85, TA87) were used in this process: the main innovation comes
from the modelling itself.
PHOTON PRODUCTION DATA
The photon production data stored in MF12 (photon multiplicities)
and MF14 (photon angular distributions) are identical to those of
ENDF/B-VI. Photons are only given for inelastic scattering via
the first-excited state (MT51) and for radiative capture (MT102).
Photons from the (n,2n)d-alpha reaction (MT24) are not explicitly
given.
*****************************************************************
EFF-2 DETAILED FILE INFORMATION
UPDATED FROM THE ENDF/B-VI EVALUATION
(V-C ANALYSIS MEANS VARIANCE-COVARIANCE ANALYSIS, SEE
YO-81,82,88)
*****************************************************************
*********** MF=2 RESONANCE PARAMETERS ***************************
MT=151 SCATTERING RADIUS ONLY.
*********** MF=3 SMOOTH NEUTRON CROSS SECTIONS ******************
MT= 1 TOTAL CROSS SECTION. ENDF/B-V.1 ADOPTED BELOW 0.1 MEV
EXCEPT THE THERMAL DATA OF MU81 WAS USED AND THE ELAST.
DATA OF AL82 WAS MATCHED. V-C ANALYSIS ABOVE 0.1
MEV BASED ON ME70, GO71, HA78, FO71, KA57, BR58, PE60,
CO52, LA79, AND HI68. ERRORS DOUBLED NEAR 260-KEV
RESONANCE FOR HI68, ME70 DATA.
MT= 2 ELASTIC CROSS SECTION. V-C ANALYSIS USED DATA OF TH56,
WI56, KN68, BA63A, HO68, LI80, CO67, WO62, RE66, AR64,
HY68, LA61, KN81, KN79, HO79, LA64, CH85, SH84, AND
DR86. THE ERRORS IN THE V-C ANALYSIS WERE TRIPLED NEAR
260-KEV RESONANCE FOR LA61, LA64 DATA. OPTICAL MODEL
ANALYSIS OF HO79 DATA USED TO CALCULATE X/S ABOVE
14 MEV FOR V-C ANALYSIS. THERMAL CROSS SECTION OF MU81
USED, AND EVALUATION MATCHED TO AL82 DATA BELOW 0.1 MEV.
MT= 4 (N,NPRIME)GAMMA + (N,NPRIME)ALPHA-T. SUM OF MT=51,52,
53,91 (EQUAL TO ENDF/B-VI)
MT= 16 (N,2N) CROSS SECTION. V-C ANALYSIS OF MT16+MT24 USED
DATA OF CH85, AS58, MC61, MA69. SEPARATION OF MT16 AND
MT24 FOLLOWS THE RATIO OF THE ENDF/B-V.1 DATA.
MT= 24 (N,2N)ALPHA-D CROSS SECTION. SEE COMMENT FOR MT16.
MT= 25 (N,3N)ALPHA-P CROSS SECTION. SMOOTH CURVE DRAWN ABOVE 14
MEV SO AS TO APPROXIMATELY AGREE WITH MA69, KO71 DATA.
MT= 28 (N,NP) CROSS SECTION. 50 PCT. OF MT104 OF ENDF/B-VI
SUM OF (N,D) + (N,NP) AGREES APPROXIMATELY WITH EXPS OF
BA53, BA63B, LI73 AND DISAGREE WITH MI61.
MT= 51 (N,NPRIME)GAMMA CROSS SECTION. SEPARATE V-C ANALYSIS
USED DATA OF PR72, OL80, SM76, DI74, MO78, FR55, KN81,
BE60, BA53, CH61, KN68, GL63, HO68, BA63A. ERRORS
DOUBLED ON FR55, BE60, AND TRIPLED ON CH61.
MT= 52 (N,NPRIME)ALPHA-T CROSS SECTION. SEPARATE V-C ANALYSIS
PERFORMED USING DATA OF CH85, DR87, SC87, DE87, RE66,
AR64, HY68, WO62, CO67, BI77, HO79, BA79, BA63, LI80,
HO68, RO62 AND YO65. (MT52 IS EQUAL TO MT56 OF ENDF/B-VI)
MT= 53 BASED UPON EVALUATION OF LISKIEN (SEE GRAPH IN LI86)
MT= 91 DIFFERENCE BETWEEN MT4 AND SUM OF MT51,52,53
MT=102 (N,GAMMA) CROSS SECTION. SHAPE FROM ENDF/B-V.1 EXCEPT
DATA RAISED ABOVE 10 EV TO AGREE WITH EXP.DATA (IM59).
THERMAL CROSS SECTION OF MU81 USED.
MT=104 (N,D) CROSS SECTION. 50 PCT. OF MT104 FROM ENDF/B-VI
SUM OF (N,D) + (N,NP) AGREES APPROXIMATELY WITH EXPS OF
BA53, BA63B, LI73 AND DISAGREE WITH MI61.
*********** MF=4 NEUTRON ANGULAR DISTRIBUTIONS ******************
MT= 2 LEGENDRE COEFFICIENTS OBTAINED BY DRAWING SMOOTH CURVE
THROUGH FITTED COEFFICIENTS FROM MEASUREMENTS LISTED
UNDER MF3/MT2. DATA OF LA61, KN68, KN79, KN81, HO79,
EMPHASIZED. OPTICAL MODEL CALCULATIONS USED ABOVE 14 MEV.
MT= 51 LEGENDRE COEFFICIENTS OBTAINED FROM ANALYSIS OF LISKIEN
(LI86), JOINED SMOOTHLY TO DWBA CALCULATION
ABOVE 8 MEV USING DWUCK CODE AND OPTICAL PARAMETERS FROM
MT=2 ANALYSIS. SOME USE ALSO MADE OF LI7(P,PPRIME) DATA.
NOTE: THE ANGULAR DISTRIBUTION IS DIFFERENT FROM THAT OF
ENDF/B-VI
MT= 52 COEFFICIENTS OBTAINED BY DRAWING SMOOTH CURVES THROUGH
FITTED VALUES FROM EXPERIMENTS LISTED ABOVE UNDER
MF3/MT52, ESPECIALLY HO68 AND HO79. (EQUAL TO MT56 OF
ENDF/B-VI).
*********** MF=6 NEUTRON EMISSION DATA **************************
NO OTHER EMITTED PARTICLES THAN NEUTRONS ARE DESCRIBED
BASIC REFERENCE: BE90
MT= 16 (N,2N) CROSS SECTION (THREE-BOBY BREAK-UP)
MT= 24 (N,2N)D-ALPHA CROSS SECTION (THREE-BODY BREAK-UP
FOLLOWED BY DECAY OF FIRST-EXCITED STATE IN LI-7)
NOTE: DOUBLE SEQUENTIAL BREAK-UP TROUGH HE-6 AND
HE-5 HAS BEEN NEGLECTED.
MT= 25 (N,3N)ALPHA-P CROSS SECTION (MULTI-BODY BREAK-UP)
MT= 28 (N,NP) CROSS SECTION (THREE-BOBY BREAK-UP)
MT= 53 (N,N')T-ALPHA CROSS SECTION (INELASTIC
SCATTERING FOLLOWED BY DECAY OF EXCITED STATE
WITH NATURAL WIDTH)
MT= 91 (N,N')T-ALPHA CROSS SECTION
*********** MF=12 PHOTON MULTIPLICITIES *************************
MT= 51 MULTIPLICITY IS 1.0 EVERYWHERE SINCE FIRST LEVEL IN
LI7 IS ONLY KNOWN PHOTON EMITTER.
MT=102 ADOPTED DIRECTLY FROM ENDF/B-V.1. DUE TO THE ABSENCE
OF DATA, THE TRANSITIONS ARE SIMPLY REASONABLE GUESSES.
*********** MF=14 PHOTON ANGULAR DISTRIBUTIONS ******************
MT= 51 ISOTROPY ASSUMED AT ALL ENERGIES.
MT=102 ISOTROPY ASSUMED FOR ALL GAMMAS AT ALL ENERGIES.
*********** REFERENCES ******************************************
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