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2.805800+4 5.743800+1 1 0 2 0
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 0 338 1
28-Ni- 58 IRK-IJS EVAL-AUG99 EUROPEAN JOINT COLLABORATION
DIST-MAY05 REV1-MAY05 20050504
----JEFF-31 MATERIAL 2825
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
***************************** JEFF-3.1 *************************
** **
** Original data taken from: JEFF-3.0 **
** **
******************************************************************
***************************** JEFF-3.0 ***********************
MF-6 91 Law 22 changed to 2
DATA TAKEN FROM :- EFF-3.1 (DIST-AUG99 REV1-SEP00)
******************************************************************
Authors and Responsibilities:
S.Tagesen, H.Vonach and A.Wallner, I.R.K.:
- Complete evaluation of the cross sections including covariance
matrices by generalized least squares cross section update
code GLUCS (9, 10). The total cross section is evaluated in
broad energy bins.
A.Trkov, I.J.S.:
- Final assembly of the file.
- Consistency corrections (interp.law in MF6 starter file).
- Implementation of the resonance fluctuations on the smooth
newly evaluated cross sections.
- Preliminary data verification and benchmarking.
Evaluation Details:
The Oak Ridge Ni-58 ENDF/B-VI Revision 1 evaluation (MAT 2825)
by Larson et.al. was the chosen starter file. All neutron
cross sections above 810 keV were re-evaluated and include
covariance data. The following data sets were selected as
"priors":
MT1 EFF-2
MT16 IRDF-90 (evaluation Pavlik)(11)
MT22 ENDF-B/VI
MT28 EFF-2
MT51 EFF-2
MT52 EFF-2
MT53 EFF-2
MT54-58 EFF-2
MT91 ENDF/B-VI
MT102 EFF-2
MT103 evaluation Badikov (12)
MT104 ENDF-B/VI
MT105 JENDL-3
MT106 EFF-2
MT107 EFF-2
MT112 EFF-2
The Evaluation was performed in three steps:
step 1: Individual excitation functions updated with new experi-
mental data points
MT1 28 data sets, 240 datapoints
EXFOR #: 10047(Foster), 10225 023(Green), 10225 024
(Green), 10342 002(Perey), 10342 003(Perey)
10416(Schwartz), 10593(Guenther),10823(Smith)
11056(Coon), 11057(Goodman), 11108(Peterson)
11155(Brathenahl), 12882(larson), 20012(Cier-
jacks), 20480 015(Cabe), 21122(McCallum)
30037(Mazari), 30141(Angeli), 31486(Poli-
croniades), 40559(Tutubalin), 68023(Tsukada)
10037(Boschung), 12752(Budtz), 113523(Smith)
40614(Fedorov), 40813(Dukarevich), 60949
(Thibault), Larson
Note: The high resolution data (EXFOR entry 22314001,
Brusegan et.al., 1997) are excluded from the
evaluation process due to a systematic
discrepancy in the data at higher energies.
MT51 8 data sets, 57 datapoints
EXFOR #: 10037(Boschung), 11276(Rodgers), 12752(Budtz)
12930(Guss), 12997(Pedroni), 13523(Smith)
40065(Pasechnik), 40531(Korzh)
MT52 4 data sets, 26 datapoints
EXFOR #: 10037(Boschung), 10852(Traiforos)
12752(Budtz), 13523(Smith)
MT53 3 data sets, 23 datapoints
EXFOR #: 10852(Traiforos), 12752(Budtz), 13523(Smith)
MT54-58 2 data sets, 13 datapoints
EXFOR #: 10852(Traiforos), 133523(Smith)
MT104 3 data sets, 3 datapoints
EXFOR #: 10827(Grimes), 12999(Graham), 30407(Glover)
MT105 2 data sets, 5 datapoints
EXFOR #: 22156(Katoh), 30473(Sudar)
MT107 6 data sets, 48 datapoints
EXFOR #: 31446(Tang), 31481(Majeddin), Sanami
41239(Ketlerov), Fessler-dd, Fessler-dt
step 2: combined evaluation of alpha and deuteron production xsec.
MT22+107 12 data sets, 80 datapoints
EXFOR #: 10827(Grimes), Baba, 13598(Haight96)
21658(Paulsen), 21873(Wattecamps), Sanami
Haight98, 10827a(Grimes), 10933(Kneff)
12999(Graham), 13598a(Haight), Tsabaris
MT28+104 7 data sets, 71 datapoints
EXFOR #: 21965 004(Pavlik), 21965 005(Pavlik)
22089(Ikeda), 30604(Raics), 30979(Viennot)
31444(Lu Hanlin), 41240(Filatenkov)
step 3: combined evaluation of all reactions with redundant data
MT2 9 data sets, 81 datapoints
EXFOR #: 10037(Boschung), 10113(Kinney), 12752(Budtz)
12930(Guss), 12997(Pedroni), 13523(Smith)
20019(Holmquist), 22048(Olson), 1 eval.pt.
Pavlik
MT3 4 data sets, 7 datapoints
EXFOR #: 11216(Beyster), 11217(Taylor), 11220(Beyster)
1 eval.pt. (Pavlik)
MT4 4 data sets, 35 datapoints
EXFOR #: 10852(Traiforos), 11218(Day), Larson
1 eval.pt. (Pavlik)
MT91 1 data set, 1 data point
derived by spectrum integration 5 - 11 MeV
overall Chisq./deg.of freedom 0.8023
A detailed description including literature references will be
given in an IAEA(NDS) - Report, published by the IAEA.
Resonance Fluctuations in the Cross Sections:
The cross sections below 810 keV are not affected by the re-
evaluation process. Above this energy, broad bin average total
cross section from the starter file was calculated. The bins
correspond to those used in the re-evaluation process. A smooth
cross section curve was generated, conserving the bin average
values. Fluctuations modulating function was defined as the
ratio of the original and the smoothed cross section. This
modulating function is applied on the re-evaluated smooth total
cross section and the inelastic cross sections.
Other Data:
- Above 14 MeV the cross section for inelastic scattering into
continuum was forced from the ENDF/B-VI starter file.
- The interpolation laws in the MF6 starter file has inconsistent
incident energy grid for the corresponding point interpolation,
therefore the flag was changed to unit base interpolation.
- Neutron spectrum in MF6 MT91 was reduced slightly above
7.5 MeV for incident neutrons at 13 and 14.5 MeV. This
adjustment is justified by comparison with the neutron emission
spectrum at 14 MeV evaluated by Vonach et.al. and by the
evidence from integral benchmark results.
The remaining comments are taken over from the starter file,
except for the sections referring to the data, which have been
superseeded. Modified sections are identified by the use of
lower case characters.
******************************************************************
CAPTURE WIDTHS CORRECTED FOR 58.7 AND 439.52 KEV RESONANCES.
THE ELASTIC TRANSFORMATION MATRIX WAS REMOVED.
FIXED TYPO IN MINIMUM ENERGY IN MF=6, MT=51
******************************************************************
THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE
DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1)
AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES
ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE
COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL
REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE
CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF
CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).
----- DESCRIPTION OF FILES
(MF-MT)
1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS.
2-151 RESONANCE PARAMETERS ARE USED EXCLUSIVELY TO PROVIDE THE
TOTAL, SCATTERING AND CAPTURE CROSS SECTIONS FROM 1.E-5 EV
TO 812 KEV, EXCEPT FOR CAPTURE FROM 450-812 KEV. THE
RESONANCE PARAMETERS ARE TAKEN FROM AN
EXTENSIVE RESONANCE PARAMETER ANALYSIS OF TRANSMISSION,
SCATTERING AND CAPTURE DATA (6). NEGATIVE ENERGY
RESONANCES HAVE BEEN USED TO GIVE THE CORRECT THERMAL
VALUES. THE SCATTERING CROSS SECTION IS GIVEN COMPLETELY
BY THE RESONANCE PARAMETERS FROM 1.E-5 TO 812 KEV, BUT FOR
THE CAPTURE CROSS SECTION, AVERAGE RESONANCE PARAMETERS
USED FROM 450 TO 812 KEV. THUS, THERE IS A BACKGROUND
CONTRIBUTION IN 3/102 FROM 450 TO 812 KEV TO ACCOUNT FOR
THE SMALL DIFFERENCE IN CAPTURE CROSS SECTION FROM THE
AVERAGE RESONANCE PARAMETERS AND THE AVERAGED EXPERIMENTAL
DATA (6). THE REICH-MOORE CODE SAMMY (7) WAS USED FOR THE
RESONANCE PARAMETER ANALYSIS.
THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE
RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 29.4 B,
ELASTIC SCATTERING 24.8 B, AND CAPTURE 4.62 B.
NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION
OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE
PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON
A FINER ENERGY GRID THAN GIVEN IN 4/2.
3-1 The broad bin total cross section from .812 to 20 MeV was
re-evaluated, but the datailed shape corresponds to the
same high resolution measurement that was used for the
resonance parameter analysis in 2/151.
3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY
SUBTRACTING THE NONELASTIC FROM THE TOTAL
3-3 NONELASTIC CROSS SECTION; SUM OF 3-4, 3-16, 3-22, 3-28,
3-102, 3-103, 3-104, 3-105, 3-106, 3-107 and 3-112.
3-4 TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, ..
.., 3-58, AND 3-91
3-16 (n,2n) cross sections were re-evaluated.
3-22 (n,na) + (n,an) cross sections were re-evaluated.
3-28 (n,np) + (n,pn) cross sections were re-evaluated.
3-51 to 3-58 and 91 INELASTIC SCATTERING EXCITING LEVELS; the
cross sections were re-evaluated.
3-102 (N,G) CAPTURE CROSS SECTION IS PROVIDED ONLY BY RESONANCE
PARAMETERS FROM 1.E-5 EV TO 450 KEV. FROM 450 TO 812 KEV
A SMALL BACKGROUND FILE IS GIVEN IN 3/102 TO COMBINE WITH
THE 2/151 RESULTS. Above 812 keV the cross sections were
re-evaluated.
3-103 (n,p) cross sections were re-evaluated.
3-104 (n,d) cross sections were re-evaluated.
3-105 (n,t) cross sections were re-evaluated.
3-106 (n,3He) cross sections from EFF-2, re-evaluated.
3-107 (n,a) cross sections were re-evaluated.
3-112 (n,pa) cross sections from EFF-2, re-evaluated.
4-2 ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS FOR ELASTIC
SCATTERING WERE REVIEWED AND ADOPTED FROM ENDF/B-V (8).
IF DESIRED, ANGULAR DISTRIBUTIONS CAN BE CALCULATED BY
THE USER ON A FINER ENERGY GRID FROM THE R-M RESONANCE
PARAMETERS IN 2/151.
6-16 (N,2N) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE
NEUTRON AND 57NI RESIDUAL, AND ENERGY DEPENDENT YIELDS
BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA
RAY; TNG CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR
EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR
THE OUTGOING NEUTRON). (N,XN) D-D EMISSION DATA HEAVILY
USED TO BENCHMARK THE TNG CALCULATIONS (5).
6-22 (N,NA) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE
NEUTRON, ALPHA, AND 54FE RESIDUAL, AND ENERGY DEPENDENT
YIELD BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE
GAMMA RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN
FOR EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY
FOR THE OUTGOING NEUTRON; ISOTROPY IS ASSUMED FOR THE
ALPHA AND RESIDUAL). (N,XA) D-D EMISSION DATA HEAVILY USED
TO BENCHMARK THE TNG CALCULATIONS (5).
6-28 (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE
NEUTRON, P, AND 57CO RESIDUAL, AND ENERGY DEPENDENT YIELD
BASED ON TNG CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA
RAY; CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR
EACH PRODUCT (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR
THE OUTGOING NEUTRON). (N,XP) D-D EMISSION DATA HEAVILY
USED TO BENCHMARK THE TNG CALCULATIONS (5).
6-51 THROUGH 6-58 INELASTIC SCATTERING EXCITING LEVELS; EACH
SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON
AND 58NI RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR
THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED
BY DWUCK (1) AND TNG (2,3,4,5)). EXTENSIVE COMPARISONS
WITH ANGULAR DISTRIBUTION DATA ARE GIVEN IN (5).
6-91 INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES
SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 58NI
RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG
CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY; TNG
CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN
FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE
OUTGOING NEUTRON). (N,XN) D-D EMISSION DATA HEAVILY USED
TO BENCHMARK THE TNG CALCULATIONS (5).
6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P
AND 58CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED
ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY;
CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH
PRODUCT. (N,XP) D-D EMISSION DATA HEAVILY USED TO BENCH-
MARK THE TNG CALCULATIONS (5).
6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A
AND 55FE RESIDUAL, AND ENERGY DEPENDENT YIELD BASED
ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY;
CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH
PRODUCT. (N,XA) D-D EMISSION DATA HEAVILY USED TO BENCH-
MARK THE TNG CALCULATIONS (5).
12-51 THROUGH 12-58 BRANCHING RATIOS FOR THE LEVELS, COMPILED
BY HETRICK ET AL. (5), ARE GIVEN.
12-102 (N,G) CAPTURE; MULTIPLICITIES WERE TNG CALCULATED.
14-51 THROUGH 14-58 AND 14-102 GAMMA RAY ANGULAR DISTRIBUTIONS
ASSUMED TO BE ISOTROPIC.
15-102 (N,G) CAPTURE; TNG CALCULATED.
--------------------------------------------------------------
UNCERTAINTY FILES
ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS
REQUIRED BY ENDF/B-VI FORMATS
For all evaluated reactions full covariance matrices are
given as calculated by the bayesian evaluation update code
GLUCS. This includes full inter-reaction covariances.
33-1 TOTAL UNCERTAINTIES estimated FROM 1E-5 TO 810 keV, full
covariances from GLUCS from 810 keV to 20 MeV.
33-2 DERIVED FROM 1E-5 TO 20 MEV
33-3 DERIVED FROM 1E-5 to 20 MeV.
33-4 DERIVED FROM THRESHOLD TO 20 MEV.
33-16 (N,2N) covariances from GLUCS.
33-22 (N,NA) covariances from GLUCS.
33-28 (N,NP) covariances from GLUCS.
33-51-53 inelastic scattering covariances from GLUCS
33-54-58 inelastic scattering covariances lumped into MT854
33-91 inelastic scattering covariances from GLUCS
33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM 1E-5 to 810 keV,
covariances from GLUCS from 810 keV TO 20 MEV.
33-103 (N,P) covariances from GLUCS.
33-104 (N,D) covariances from GLUCS.
33-105 (n,t) covariances from GLUCS.
33-106 (n,3He) covariances from GLUCS.
33-107 (N,A) covariances from GLUCS.
33-112 (n,pa) covariances from GLUCS.
34- 2 LEGENDRE COEFFICIENTS A1-A3 COVARIANCES FROM EFF-2.4
REFERENCES:
(1) P.D. KUNZ, "DISTORTED WAVE CODE DWUCK72," UNIV. OF
COLORADO, UNPUBLISHED (1972).
(2) C.Y. FU, "A CONSISTENT NUCLEAR MODEL FOR COMPOUND AND
PRECOMPOUND REACTIONS WITH CONSERVATION OF ANGULAR
MOMENTUM," ORNL/TM-7042 (1980).
(3) C.Y FU, "DEVELOPMENT AND APPLICATION OF MULTI-STEP
HAUSER-FESHBACH/PRE-EQUILIBRIUM MODEL THEORY," SYMP.
NEUTRON CROSS SECTIONS FROM 10 TO 50 MEV, UPTON, N.Y.,
MAY 12-14,1980, BNL-NCS-51425, P 675, BROOKHAVEN
NATIONAL LAB.
(4) K. SHIBATA AND C.Y. FU, "RECENT IMPROVEMENTS OF THE TNG
STATISTICAL MODEL CODE", ORNL/TM-10093 (AUGUST, 1986).
(5) D.M. HETRICK, C.Y. FU, AND D.C. LARSON, "CALCULATED NEUTRON
-INDUCED CROSS SECTIONS FOR 58,60NI FROM 1 TO 20 MEV AND
COMPARISONS WITH EXPERIMENT," ORNL/TM-10219,ENDF-344 (1987).
(6) C.M. PEREY, F.G. PEREY, J.A. HARVEY, N.W. HILL, N.M. LARSON,
AND R.L. MACKLIN,"58NI+N TRANSMISSION, DIFFERENTIAL ELASTIC
SCATTERING AND CAPTURE MEASUREMENTS AND ANALYSIS FROM 5 TO
813 KEV", REPORT ORNL/TM-10841, IN PREPARATION
(7) N.M. LARSON AND F.G. PEREY,"USERS GUIDE FOR SAMMY:A COMPUTER
MODEL FOR MULTILEVEL R-MATRIX FITS TO NEUTRON DATA USING
BAYES' EQUATIONS", ORNL/TM-7485, OAK RIDGE NATIONAL LAB, 1980
"UPDATED USERS'GUIDE FOR SAMMY", ORNL/TM-9179, 1984,
ORNL/TM-9179R1, 1985, AND ORNL/TM-9179/R2, 1988.
(8) M. DIVADEENAM,"NI ELEMENTAL NEUTRON INDUCED REACTION CROSS-
SECTION EVALUATION", BNL-NCS-51346, ENDF-294, MARCH 1979.
(9) D.M. HETRICK AND C.Y. FU, "GLUCS: A GENERALIZED LEAST SQUARES
PROGRAM FOR UPDATING CROSS SECTION EVALUATIONS WITH
CORRELATED DATA SETS," ORNL/TM-7341, ENDF-303 (OCT,1980).
(10) S. TAGESEN and D.M. HETRICK, "Enhancements to the Generalized
Least-Squares Cross-Section Evaluation Code GLUCS", Proc.Int.
Conf. on Nuclear Data for Science and Technology, Gatlinburg
Tennessee, May 9 - 13, 1994, p 589.
(11) M. Wagner et al., Physics Data 13-5, Fachinformationszentrum
Karlsruhe, 1990
(12) S. Badikov, S. Tagesen and H. Vonach, Physics Data 13-9,
Fachinformationszentrum Karlsruhe, 1996
1 451 343
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