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6-C - 0 LANL,ORNL EVAL-JUN96 M.B.CHADWICK, P.G.YOUNG, C.Y. FU
Ch96a,Ch96b,Fu90,Ch99DIST-JAN09 20090105
----JEFF-311 MATERIAL 600 REVISION 2
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
*************************** JEFF-3.1.1 *************************
** **
** Original data taken from: JEFF-3.1 **
** **
******************************************************************
***************************** JEFF-3.1 *************************
** **
** Original data taken from: ENDF/B-VI.8 **
** **
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ENDF/B-VI MOD 3 Evaluation, June 1996, M.B. Chadwick and
P.G. Young (LANL)
Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code
in cooperation with ECN Petten.
This evaluation provides a complete representation of the
nuclear data needed for transport, damage, heating, radioactivity
and shielding applications over the incident neutron energy
range from 1.0E-11 to 150 MeV. The discussion here is divided
into the region below and above 20 MeV.
INCIDENT NEUTRON ENERGIES < 20 MeV
Below 20 MeV the evaluation is based completely on the ENDF/B-
VI.1 (Release 1) evaluation by Fu [Fu90]. The following minor
modifications were made to the ENDF/B-VI.1 evaluation:
1. The energy range from En = 20 MeV to 32 MeV was replaced by
the LANL evaluation (described below);
2. The elastic, nonelastic, and total cross sections from 19 to
20 MeV were varied to join smoothly with the higher energy
values at 20 MeV.
3. The new flag, LTT=3, is used in MF=4,MT=2 to indicate that
Legendre polynomials are used below 20 MeV and probability
tabulations at higher energies. A small discontinuity exists for
MF=4,MT=2 at 20 MeV where the two different representations join.
The higher energy evaluation utilizes a tabulation in order to
overcome the inaccuracies caused by the ENDF-6 limitation of 20
for the maximum number of Legendre coefficients.
INCIDENT NEUTRON ENERGIES > 20 MeV (12C analysis)
The evaluation above 20 MeV utilizes MF=6, MT=5 to represent
all reaction data. Production cross sections and emission
spectra are given for neutrons, protons, deuterons, alpha
particles, gamma rays, and all residual nuclides produced (A>5)
in the reaction chains. To summarize, the ENDF sections with
non-zero data above En = 20 MeV are:
MF=3 MT= 1 Total Cross Section
MT= 2 Elastic Scattering Cross Section
MT= 3 Nonelastic Cross Section
MT= 5 Sum of Binary (n,n') and (n,x) Reactions
MT=102 Radiative Capture Cross Section (Estimate Only)
MF=4 MT= 2 Elastic Angular Distributions
MF=6 MT= 5 Production Cross Sections and Energy-Angle
Distributions for Emission Neutrons, Protons,
Deuterons, and Alphas; and Angle-Integrated
Spectra for Gamma Rays and Residual Nuclei That
Are Stable Against Particle Emission
MF=33 MT= 1 Covariance file for total cross section
MT= 2 Covariance file for elastic cross section
MT= 3 Covariance file for nonelastic cross section
MT= 5 Covariance file for composite reaction cross sect.
MT=102 Covariance file for capture cross section
The evaluation is based on nuclear model calculations that have
been benchmarked to experimental data, especially for n + C12 and
p + C12 reactions [Ch96a]. We use the GNASH code system [Yo92],
which utilizes Hauser-Feshbach statistical, preequilibrium and
direct-reaction theories. Coupled-channel and spherical optical
model calculations are used to obtain particle transmission
coefficients for the Hauser-Feshbach calculations, as well as for
the elastic neutron angular distributions.
Cross sections and spectra for producing individual residual
nuclei are included for reactions that exceed a cross section of
approximately 1 nb at any energy. The energy-angle-correlations
for all outgoing particles are based on Kalbach systematics
[Ka88].
A model was developed to calculate the energy distributions of
all recoil nuclei in the GNASH calculations (Ch96b). The recoil
energy distributions are represented in the laboratory system in
MT=5, MF=6, and are given as isotropic in the lab system. Note
that all other data in MT=5,MF=6 are given in the center-of-mass
system. This method of representation requires a modification of
the original ENDF-6 format.
Preequilibrium corrections were performed in the course of the
GNASH calculations using either Feshbach, Kerman, Koonin (FKK)
theory [Ch93] or the exciton model of Kalbach [Ka77, Ka85].
Discrete level data from nuclear data sheets were matched to
continuum level densities using the formulation of Ignatyuk
[Ig75] and pairing and shell parameters from the Cook [Co67]
analysis. Neutron and charged-particle transmission coefficients
were obtained from the optical potentials, as discussed below.
Gamma-ray transmission coefficients were calculated using the
Kopecky-Uhl model [Ko90].
****************************************************************
ENDF/B-VI MOD 2 Revision, July 1991
The only changes for MOD 2 are completion of some references,
and addition of total elastic scattering uncertainties as
recommended by the Standards Subcommittee.
"Following concerns expressed about the seemingly small
standards uncertainties, the standards subcommitte has provided
expanded uncertainties. These uncertainties are estimates such
that if a modern day experiment were performed today on a given
standard using the best techniques, those results should fall
within these expanded uncertainties (2/3 of the time). They take
into account data inconsistencies and concerns about R-Matrix
parameters. Note that it is not assumed that the uncertainties
are totally correlated within the energy ranges given. It is
recommended that these expanded uncertainties be put in file 1
and in the documentation for the standards."
Quote from Standards Subcommittee CSEWG minutes for the May 1990
meeting.
C(n,n) Total elastic scattering cross section
Energy (keV) Uncertainty (%)
1 - 500 0.46
500 - 1500 0.53
1500 - 1800 0.60
****************************************************************
ENDF/B-VI MOD 1 Evaluation, C.Y. Fu, E.J. Axton, and F.G. Perey
ORNL
NEW EVALUATION FOR VERSION VI:
1. Total and elastic scattering from 0.1 to 0.25 MeV and from 1.6
to 1.9 MeV due to inclusion of C13 resonances [Fu88].
2. Elastic angular distribution: 0.1 to 2.0 MeV due to 13C
effects [Fu88].
3. New evaluation for all File 3 data 5-32 MeV by Axton [Ax88],
including new data for (n,n'3a) [An86, Br84, Me84, Ol87], for
(n,n') [Ba85, Gu81, Sa81], and for kerma factors [Be81, Br83,
Bu85, De84, De85, De86, Di82, Ha84, Mc86, We79].
4. Extention of energy and angular distributions to 32 MeV were
mostly based on extrapolation. File 33 was extended to 32 MeV
with the addition of the new LB=8 subsubsection.
Retained from ENDF/B-V (with negligible changes such as
Q-values): all data between 2.0 and 5.0 MeV [Fu78], all energy
and angular data for neutron and gamma-ray productions, and
uncertainty files (LB=8 subsubsection added).
Data and evaluation techniques used in the new evaluation and
ENDF/B-V evaluation, as adopted here, are summerized below.
MF-MT
3-1 TOTAL
1.E-5 eV TO 4.81 MeV -- sum of 3-2 and 3-102.
4.81 MeV TO 20 MeV -- [Sc67, Ci69, Pe72].
3-2 ELASTIC SCATTERING
1.E-5 eV to 4.81 MeV -- R-Matrix analysis with data [Sc67,
Ci69, Pe72, Ah70, Bl75, Di68, Fr70, He75, La57, La61, La69,
Me70, Ki76, Ho72, Ho75, St70, Me54, Wi58, Kn73, Pu64, El62,
We65, Ke65, Go65, Ga72, Pk72]. Bayes theorem (or nonlinear
least square) used for energies less then 2 MeV, resulting
weights were then used in the R-Matrix analysis. A thermal
total cross section of 4.746 +- 0.25 (evaluated by Lubitz
[Lu76]) was also used in the R-Matrix fit.
4.81 MeV to 8 MeV -- [Ga72, Ve73, Pk72].
8 MeV to 14 MeV -- [Ha75, Ve73, Pu76].
14 MeV to 32 MeV -- [Bo68, Me84].
R-Matrix analysis for C13 below 2 MeV with data of [La81,
He75, Co61, Au79].
3-3 NONELASTIC
1.E-5 eV to 4.81 MeV -- same as 3-102.
4.81 MeV to 32 MeV -- 3-1 minus 3-2.
3-4 INELASTIC -- sum of 3-51 through 3-91
3-16 (N,2N) -- [An81, We81]
3-24 (N,2NA) -- empirical estimate
3-28 (N,NP) -- empirical estimate
3-32 (N,ND) -- empirical estimate
3-33 (N,NT) -- include (N,NPT) and (N,N2P)
3-41 (N,2NP) -- include (N,2NHE-3)
3-51 INELASTIC SCATTERING TO 4.439-MeV LEVEL
4.81 MeV to 6.32 MeV -- 3-3 minus 3-102.
6.32 MeV to 8.796 MeV -- 3-3 minus 3-102 minus 3-107.
8.796 MeV TO 32 MeV -- same refs. as in 3-2 and gamma-ray data
of [Mo72]. New data considered for Version-VI: [Ba85, Gu81,
Sa81].
3-52 to 3-91 (N,N PRIME 3A)
MT=52 to 57: real levels
MT=58 to 73: pseudo levels with 1-MeV width
MT=91: an evaporation component with T=0.3 to reproduce
correct threshold effect and the decay of the 2.43-MeV
level of 9Be.
New data considered for ENDF/B-VI: [An86, Br84].
3-102 CAPTURE
1.E-5 eV to 1 MeV -- 1/v with 3.36 mb at thermal.
1 MeV to 32 MeV -- derived from (G,N) cross section of [Co57].
3-103 (N,P) -- [Ri68].
3-104 (N,D) -- derived from (D,N) of [Am57].
3-105 (N,T) -- estimated shape normalized to [Qa78]
3-107 (N,A) -- [Da63, Ve68, Re60, Gr55, Ob72, Va70].
3-111 (N,2P) -- estimate
3-112 (N,PA) -- includes (N,LI-6), (N,DA), (N,TA), (N,DT),
(N,HE6), (N,HE-3 A), (N,PTA)
3-115 (N,PD) -- estimate
3-116 (N,PT) -- estimate
3-203 PROTON PRODUCTION --
(3-28)+(3-41)+(3-103)+2*(3-111)+(3-112)+(3-115)+(3-116)
3-204 DEUTERON PRODUCTION -- (3-32)+(3-104)+(3-115)
3-205 TRITON PRODUCTION -- (3-33)+(3-105)+(3-116)
3-207 ALPHA PRODUCTION -- (3-24)+3*(3-52 to 3-91)+(3-107)+(3-112)
4-2 ANGULAR DISTRIBUTION OF ELASTICALLY SCATTERED NEUTRONS
Same data and analysis as in 3-2. Legendre coefficients in
center-of-mass.
4-51 INELASTIC SCATTERING TO 4.439-MeV LEVEL
Same data sources as in 3-51.
4-52 INELASTIC SCATTERING TO 7.653-MeV LEVEL -- [Gr69].
4-53 INELASTIC SCATTERING TO 9.638-MeV LEVEL -- [Gr69].
4-54 to 4-91 -- isotropic in center-of-mass.
5-28 evaporation spectrum with T=0.3 to 0.5 MeV.
5-91 evaporation spectrum with T=0.3 MeV. This is a small
component of (N,N PRIME 3A) and is used mainly for the decay
of the 2.43-MeV level of 9Be [An75] and for reproducing the
correct threshold effect [Fr55].
12-51 MULTIMLICITY OF THE SINGLE GAMMA-RAY -- from 3-51
12-102 MULTIMLICITY OF (N,G) GAMMA-RAYS -- [Aj70].
14-51 ANGULAR DISTRIBUTION OF 4.439-MeV GAMMA-RAYS -- [Mo72,
Dr69, Ha59, Pr66, Ne64, An58, Ko65, En64].
14-102 ANGULAR DISTRIBUTION OF CAPTURE GAMMA-RAYS -- isotropic
in center-of-mass.
33-1 to 33-107 UNCERTAINTY FILES FOR FILE 3 DATA.
*****************************************************************
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