29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK      
 Ch98,Ch99            DIST-NOV 1 REV4-                 20011108   
----ENDF/B-VI         MATERIAL 2931         REVISION 4            
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
Oct 22,2004, Holly Trellue remade mf6mt5 using corrected gnash    
code, to fix an earlier bug. Impact is reduced sec. particle prod.
for new ENDF/B-VII release                                        
 **************************************************************** 
                                                                  
 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy,    
                 P.G. Young (LANL)                                
                                                                  
 The secondary gamma-ray spectrum for radiative capture (MF 12,   
 MT 102) has been updated for new experimental data at incident   
 neutron energies up to 1 keV.                                    
 The MF=12, MT=102 yields above 1 keV were adjusted slightly to   
 force energy conservation.                                       
 The Q-value for radiative capture was also updated in File 3.    
 Details of these changes are described in Frankel et al. [Fr01]. 
                                                                  
 **************************************************************** 
                                                                  
 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN),    
                 M.B. Chadwick, 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 (Release 2) evaluation by D. Hetrick, C.Y. Fu, and D. Larson.  
                                                                  
INCIDENT NEUTRON ENERGIES > 20 MeV                                
                                                                  
  The ENDF/B-VI Release 2 evaluation extends to 20 MeV and        
includes cross sections and energy-angle data for all             
significant reactions.  The present evaluation utilizes a more    
compact composite reaction spectrum representation above 20 MeV   
in order to reduce the length of the file. No essential data for  
applications is lost with this representation.                    
  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, tritons,      
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           
                                                                  
  MF=4 MT=  2  Elastic Angular Distributions                      
                                                                  
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         
               Distributions for Emission Neutrons, Protons,      
               Deuterons, Tritons, and Alphas; and Angle-         
               Integrated Spectra for Gamma Rays and Residual     
                                                                  
               Nuclei That Are Stable Against Particle Emission   
                                                                  
  The evaluation is based on nuclear model calculations that      
have been benchmarked to experimental data, especially for n +    
Cu65 and p + Cu65 reactions (Ch98). We use the GNASH code system  
(Yo92), which utilizes Hauser-Feshbach statistical, preequilib-   
rium and direct-reaction theories. 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. 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 (Ch96a). The recoil   
energy distributions are represented in the laboratory system in  
MT=5, MF=6, and are given as isotropic in the lab system. All     
other data in MT=5,MF=6 are given in the center-of-mass system.   
This method of representation utilizes the LCT=3 option approved  
at the November, 1996, CSEWG meeting.                             
  Preequilibrium corrections were performed in the course of the  
GNASH calculations using the exciton model of Kalbach (Ka77,      
Ka85), validated by comparison with calculations using Feshbach,  
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     
nuclear data sheets were matched to continuum level densities     
using the formulation of Ignatyuk et al. (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).                                                     
                                                                  
SPECIFIC INFORMATION CONCERNING THE 65Cu EVALUATION               
                                                                  
This evaluation is documented in some detail in Ref. (Ko98b).     
                                                                  
The neutron total cross section above 20 MeV was obtained by      
evaluating experimental data, with a particular emphasis on the   
Finlay (Fi93) elemental data. This resulted in an evaluated       
elemental Cu total cross section; to obtain an isotopic 65Cu total
cross section, it was assumed that 63Cu and 65Cu have total cross 
sections in an A**2/3 ratio to one another. The total neutron     
nonelastic cross section was obtained directly from an optical    
model calculation (see below), after verifying that it was in good
agreement with the experimental data (Ko98b).                     
                                                                  
To obtain the neutron optical potential we used total cross       
section data from 1.2 to 4.5 MeV (Gu86) and from 5.3 to 600 MeV   
(Fi93), and elastic scattering angular distribution data from 1.6 
to 96 MeV (Br50, Sa60, Ki74, El82, Gu86). The optical potential   
parameters were obtained using a combination of a grid search code
and the interactive optical model viewer ECISVIEW [Ko97], both    
built around the coupled channels code ECIS96 [Ra94]. The energy  
dependence of the optical model parameters is as described in     
[Ko98]. This optical potential was used for the calculation, with 
ECIS96, of neutron transmission coefficients and DWBA cross       
sections for the entire energy region above 20 MeV.               
                                                                  
Due to the lack of proton elastic scattering data in numerical    
form, we used a combination of global optical models for the      
proton channel. The Becchetti-Greenlees potential [Be69]was       
adopted below 47 MeV, and the non-relativistic version of the     
Madland potential [Ma88] above 47 MeV. At this particular energy  
point the two potentials join smoothly.                           
                                                                  
For deuterons, the Lohr-Haeberli global potential [Lo74] was used;
for alpha particles the Moyen potential (MacFadden-Satchler       
[Ma66]) was used; and for tritons the Becchetti-Greenlees         
potential [Be71] was used. The He-3 channel was ignored, due to   
its small importance.                                             
                                                                  
Following Delaroche et al. [De82], we adopted the weak-coupling   
model for direct collective inelastic scattering for Cu-65, using 
Ni-64 as a basis. For the calculation of the cross sections,      
ECIS96 was used in DWBA mode. We used the following direct        
transitions for Cu-65  (ground state 3/2- ) :                     
                                                                  
Jpi Ex(MeV) Deformation length (Delta) or parameter (Beta)        
 0.5- 0.771  Beta(2)=0.0566                                       
 2.5- 1.116  Beta(2)=0.0980                                       
 3.5- 1.481  Beta(2)=0.1132                                       
 1.5- 1.743  Beta(2)=0.0800                                       
 1.5- 3.185  Delta(3)=0.3167                                      
 2.5- 3.435  Delta(3)=0.3879                                      
 3.5- 3.685  Delta(3)=0.4479                                      
 4.5- 3.935  Delta(3)=0.5008                                      
                                                                  
Only one measurement exists for neutron-induced emission spectra  
above 20 MeV for 65Cu: the 25.7 MeV (n,xn) data by Marcinkowski et
al (Ma83). Without adjusting any of the level density or pre-     
equilibrium parameters the GNASH calculation was in good agreement
with these data (Ko98b). Hence we adopted these parameters for the
whole energy region.                                              
                                                                  
****************************************************************  
                                                                  
REFERENCES                                                        
                                                                  
[Ab93]  W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra,   
   Phys.Rev. C47, 1033 (1993)                                     
[Al83]  R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986)   
[Ar80]  E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross  
   Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross   
   Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National
   Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 
   1980] p. 731.                                                  
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182,   
   1190 (1969)                                                    
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization  
   Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W.    
   Haeberli, The University of Wisconsin Press, 1971) p.682.      
[Be92]  O. Bersillon, "SCAT2 - A Spherical Optical Model Code,"   
   in Proc. ICTP Workshop on Computation and Analysis of Nuclear  
   data Relevant to Nuclear Energy and Safety, February-March,    
   1999 Trieste, Italy, to be published in World Scientific Press,
   and Progress Report of the Nuclear Physics Division, Bruyeres- 
   le-Chatel 1977, CEA-N-2037 (1978) p.111                        
[Br50]  S. Bratenahl, S. Fernbach, R.H. Hildebrand et al.,        
   Phys.Rev. 77, 597 (1950)                                       
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 
[Ch96]  M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   
   Driven Technologies:  Calculational Method for Heavy Recoils," 
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996     
[Ch98]  M.B. Chadwick and P.G. Young, "GNASH Calculations of      
   n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 
   REPORT: 1 February - 1 March 1998, internal Los Alamos National
   Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 
[Ch99]  M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 
   National Laboratory report, LA-UR-99-1222 (1999)               
[Co67]  J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J.    
   Phys. 20, 477 (1967)                                           
[De82]  J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and   
   R.L. Walter, Nucl. Phys. A390, 541 (1982).                     
[El82]  S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 
   A390, 509 (1982)                                               
[Fi93]  R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 
   C 47, 237 (1993)                                               
[Fr01]  S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos      
   National Laboratory Report, LA-13812 (2001).                   
[Gu86]  P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl.   
   Phys. A448, 280 (1986)                                         
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    
   Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485     
   (1975)                                                         
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         
   LA-10248-MS (1985)                                             
[Ka88]  C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also          
   C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981)           
[Ki74]  W.E. Kinney, F.G. Perey, report ORNL-4908 (1974)          
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990)         
[Ko97]  A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:
   A Graphical Interface for ECIS95", Proceedings of the NEA      
   Specialists' Meeting on the Nucleon Nucleus Optical Model up to
   200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 
   http://db.nea.fr/html/science/om200/.                          
[Ko98]  A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear   
  Data for Accelerator-Driven Systems: Nuclear models, Experiments
   and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 
[Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI    
   neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 
   Los Alamos National Laboratory report LAUR- (1998); ECN lab and
   JEFF report (1998).                                            
[Lo74]  J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974)    
[Ma66]  Macfadden and Satchler, Nuc.Phys. 84, 177 (1966)          
[Ma83]  A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al.,  
   Nucl.Phys. A402, 220 (1983)                                    
[Ma88]   D.G. Madland, "Recent Results in the Development of a    
   Global Medium-Energy Nucleon-Nucleus Optical-Model Potential,  
   "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 
   Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 
   (1988).                                                        
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      
[Ra94]  J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772  
   (1994)                                                         
[Sa60]  G.L. Salmon, Nucl.Phys. 21, 15 (1960)                     
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, report        
   LA-12343-MS (1992)                                             
                                                                  
 **************************************************************** 
                                                                  
 ENDF/B-VI MOD 3 Revision, March 1991, ORNL                       
                                                                  
 MOD 3 changes                                                    
   1) Corrections to MF=6, MT=63 at 17.5 MeV to prevent negative  
      values in the angular distribution.                         
                                                                  
 **************************************************************** 
                                                                  
 * Note there was no MOD 2 released.                              
                                                                  
 **************************************************************** 
                                                                  
 ENDF/B-VI MOD 1 Evaluation, November 1989, D. Hetrick, F.Y. Fu,  
           D. Larson (ORNL)                                       
                                                                  
    This work employed several nuclear model codes including the  
 optical-model code GENOA [1], the Distorted Wave Born            
 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach  
 code TNG [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 WERE TAKEN FROM MUGHABGHAB [6]. POINT
        WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER  
        FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 
        99.5 KEV.  REICH-MOORE PARAMETERS ARE GIVEN.  AGREEMENT   
        WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 
        FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN       
        AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND     
        COMPARED WITH THE BINNED DATA.  THIS IS PROBABLY DUE TO   
        TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES    
        SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION.             
        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 TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS  
        FROM 1.E-5 EV TO 99.5 KEV. THE THERMAL CROSS SECTION      
        (16.3B) IS REPRODUCED.  A SMALL CONTRIBUTION FROM 3/102 IS
        REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151   
        REPRODUCES THE CAPTURE DATA.  FROM 99.5 KEV TO 1.12 MEV   
        CU65 DATA FROM [7] IS USED, AFTER APPROPRIATE AVERAGING.  
        ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE.  FROM 1.12 TO  
        4.0 MEV, NAT CU DATA OF PEREY [8] USED IN V5 IS RETAINED. 
        FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL [9] IS    
        AVERAGED AND USED.  COMPARISONS FROM 1.2 TO 4.5 MEV       
        WITH AVERAGED ARGONNE DATA FOR NAT CU [10] SHOW 1%        
        AGREEMENT.  LARSON'S DATA ABOVE 10 MEV WERE ADJUSTED      
        SLIGHTLY (<2 %) TO FIT 65CU DATA OF DYUMIN ET AL. [11].   
  3-2   ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY        
        SUBTRACTING THE NONELASTIC FROM THE TOTAL. THE THERMAL    
        VALUE OF 14.1B IS REPRODUCED.                             
  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, AND 3-107.       
  3-4   TOTAL INELASTIC CROSS SECTION; SUM OF 3-51, 3-52, ..      
        .., 3-63, AND 3-91                                        
  3-16  (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS [12]      
        CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL-     
        TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS          
        SECTIONS [13].                                            
  3-22  (N,NA) CROSS SECTIONS WERE CALCULATED BY THE TNG          
        CODE [3,4,5] WHICH AGREES WELL WITH AVAILABLE DATA [5].   
  3-28  (N,NP) CROSS SECTIONS WERE CALCULATED BY THE TNG          
        CODE [3,4,5].  DATA ARE SCARCE FOR THIS REACTION [5].     
  3-51 TO 3-63 INELASTIC SCATTERING EXCITING LEVELS; RESULTS ARE  
        FROM TNG [3,4,5] WHICH INCLUDES DIRECT INTERACTION        
        CROSS SECTIONS FROM DWUCK [2] CALCULATIONS. EXTENSIVE     
        COMPARISONS WITH EXPERIMENTAL DATA USED IN THE EVALUATION 
        PROCESS ARE SHOWN IN [5].                                 
  3-91  INELASTIC SCATTERING EXCITING THE CONTINUUM (TNG          
        CALCULATED).  GUIDANCE FROM (N,XN) EMISSION DATA [5].     
  3-102 (N,G) CAPTURE CROSS SECTION IS OBTAINED FROM RESONANCE    
        PARAMETERS FROM 1.E-5 EV TO 99.5 KEV.  A SMALL BACKGROUND 
        IS GIVEN HERE, WHICH WHEN ADDED TO THE RESONANCE          
        CONTRIBUTION REPRODUCES EXPERIMENTAL DATA, INCLUDING THE  
        THERMAL VALUE OF 2.17B.  FROM 99.5 KEV TO 20 MEV, DATA    
        FROM THE CSISRS LIBRARY [14] WERE USED TO CONSTRUCT AN    
        AVERAGE CURVE THROUGH THE DATA; RESULTS ARE SIMILAR TO EYE
        GUIDE IN REF [15].                                        
  3-103 (N,P) CROSS SECTIONS WERE CALCULATED BY THE TNG           
        CODE [3,4,5] WHICH AGREES WELL WITH MEASURED DATA [5].    
  3-104 (N,D) CROSS SECTION - SHAPE OF (N,P) CROSS SECTION USED   
        NORMALIZED TO AVAILABLE DATA [5].                         
  3-105 (N,T) CROSS SECTION - SHAPE OF (N,P) CROSS SECTION        
        USED NORMALIZED TO SYSTEMATICS OF QAIM AND STOCKLIN [16]  
        AT 14.6 MEV.                                              
  3-106 (N,3HE) CROSS SECTION - SHAPE OF (N,A) CROSS SECTION      
        USED NORMALIZED TO AVAILABLE DATA [17].                   
  3-107 (N,A) CROSS SECTIONS WERE CALCULATED BY THE               
        TNG CODE [3,4,5].  DATA ARE SCARCE AND DISCREPANT FOR     
        THIS REACTION [5].                                        
  4-2   ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS GIVEN FOR     
        ELASTIC SCATTERING ARE RESULTS OF FITTING DATA WITH GENOA.
        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 64CU 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,XN) AND (N,XG) 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 61CO 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,XA) AND (N,XG) D-D EMISSION
        DATA HEAVILY USED TO BENCHMARK THE TNG CALCULATIONS [5].  
  6-28  (N,NP) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR THE  
        NEUTRON AND 64NI 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) AND (N,XG) D-D EMISSION    
        DATA USED TO BENCHMARK THE TNG CALCULATIONS [5].          
  6-51 THROUGH 6-63 INELASTIC SCATTERING EXCITING LEVELS; EACH    
        SECTION INCLUDES SIMPLE CONSTANT YIELDS FOR THE NEUTRON   
        AND 65CU RESIDUAL; ANGULAR DISTRIBUTIONS ARE GIVEN FOR    
        THE OUTGOING NEUTRON (LEGENDRE COEFFICIENTS COMPUTED      
        BY DWUCK [2] AND TNG [3,4,5]).  COMPARISONS WITH ANGULAR  
        DISTRIBUTION DATA ARE GIVEN IN [5].                       
  6-91  INELASTIC SCATTERING EXCITING THE CONTINUUM; INCLUDES     
        SIMPLE CONSTANT YIELDS FOR THE NEUTRON AND 65CU           
        RESIDUAL AND ENERGY DEPENDENT YIELD BASED ON TNG          
        CALCULATED GAMMA-RAY SPECTRA FOR THE GAMMA RAY;           
        CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN             
        FOR EACH (ANGULAR DISTRIBUTIONS ARE GIVEN ONLY FOR THE    
        OUTGOING NEUTRON).  (N,XN) AND (N,XG) D-D EMISSION DATA   
        RELIED UPON TO BENCHMARK THE TNG CALCULATIONS [5].        
  6-103 (N,P) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR P     
        AND 65NI RESIDUAL, AND ENERGY DEPENDENT YIELD BASED       
        ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY;            
        CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH    
        PRODUCT.  (N,XP) AND (N,XG) D-D EMISSION DATA USED TO     
        BENCHMARK THE TNG CALCULATIONS [5].                       
  6-107 (N,A) REACTION; INCLUDES SIMPLE CONSTANT YIELDS FOR A     
        AND 62CO RESIDUAL, AND ENERGY DEPENDENT YIELD BASED       
        ON CALCULATED GAMMA-RAY SPECTRA FOR GAMMA RAY;            
        CALCULATED NORMALIZED DISTRIBUTIONS ARE GIVEN FOR EACH    
        PRODUCT.  (N,XA) AND (N,XG) D-D EMISSION DATA USED TO     
        BENCHMARK THE TNG CALCULATIONS [5].                       
 12-51 THROUGH 12-63 BRANCHING RATIOS FOR THE LEVELS, COMPILED    
        BY HETRICK ET AL. [5] ARE GIVEN.                          
 12-102 (N,G) CAPTURE; THERMAL VALUES TAKEN FROM DELFINI ET AL.   
        [18].  HIGHER-ENERGY VALUES BASED ON TNG CALCULATIONS     
        USING PRIMARY BRANCHING RATIOS OF DELFINI ET AL. FOR      
        S-WAVES                                                   
 14-51 THROUGH 14-63 GAMMA RAY ANGULAR DISTRIBUTIONS ASSUMED TO   
        BE ISOTROPIC.                                             
 14-102 (N,G) CAPTURE; ISOTROPIC DISTRIBUTIONS ASSUMED            
 15-102 (N,G) CAPTURE; AS IN 12-102                               
--------------------------------------------------------------    
                UNCERTAINTY FILES                                 
        ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS       
        REQUIRED BY ENDF/B-VI FORMATS                             
                                                                  
 33-1   TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV  
        EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8.       
 33-2   EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV
 33-3   DERIVED FROM 1.E-5 TO 99.5 KEV, EXPLICIT USING LB=1,8 FROM
        99.5 KEV TO 20 MEV.                                       
 33-4   DERIVED FROM THRESHOLD TO 20 MEV.                         
 33-16  (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS [12]         
        CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL-     
        TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS          
        SECTIONS [13].                                            
 33-22  (N,NA) UNCERTAINTIES ESTIMATED FROM TNG AND DATA.         
 33-28  (N,NP) COVARIANCES ESTIMATED FROM TNG AND DATA.           
 33-51-91  INELASTIC SCATTERING UNCERTAINTIES ESTIMATED FROM DATA 
        AND CALCULATIONS.                                         
 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 
        ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 
        DATA [5,14,15] FROM 99.5 KEV TO 20 MEV.                   
 33-103 (N,P) COVARIANCES - ESTIMATED USING CSISRS DATA AS GUIDE. 
 33-104 (N,D) COVARIANCES - ESTIMATED, BASED ON DATA.             
 33-105 (N,T) COVARIANCES - ESTIMATED.                            
 33-106 (N,3HE) COVARIANCES - ESTIMATED, BASED ON DATA.           
 33-107 (N,A) COVARIANCES WERE ESTIMATED BASED ON TNG AND DATA.   
                                                                  
 **************************************************************** 
                                                                  
 REFERENCES:                                                      
                                                                  
 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967)    
 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of           
     Colorado, unpublished (1972)                                 
 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu,           
     Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,
     May 1980, Brookhaven National Lab. report BNL-NCS-51245      
     (1980) p.675                                                 
 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986)          
 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report     
     ORNL/TM-9083 [ENDF-337] (1984)                               
 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron    
     Cross Sections, Vol. 1, Neutron Resonance Parameters and     
     Thermal Cross Sections, Part A, Z=1-60," (Academic Press,    
     1981)                                                        
 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600  
     (1977), and private communication.                           
 [8] F.G. Perey, private communication (1977)                     
 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50   
     MeV, Upton, NY, May 1980, Brookhaven National Lab. report    
     BNL-NCS-51245 (1980) p.277                                   
[10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl.   
     Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and    
     W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80      
     (1983) [CSISRS data set 12853]                               
[11] A.I. Dyumin, D.M. Kaminker, G.N. Popova, and V.A. Smolin,    
     Izv.Akad.Nauk SSSR, Ser.Fiz. 36, 852 (1972)                  
[12] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341      
     [ENDF-303] (1980)                                            
[13] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp.    
     on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26,   
     1982 (U.S. National Bureau of Standards) p.877               
[14] CSISRS Library, National Nuclear Data Center, Brookhaven     
     National Laboratory, Upton, N.Y. 11973.                      
[15] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross        
     Sections, Vol. 2, Neutron Cross Section Curves" (Academic    
     Press, 1988)                                                 
[16] S.M. Qaim and G. Stoecklin, Nucl.Phys. A257, 233 (1976)      
[17] S.M. Qaim, Radiochimica Acta, 25, 13 (1978)                  
[18] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys.     
     A404, 250 (1983)                                             
                                                                  
 ************************ C O N T E N T S *********************** 
                                1        451        596          5
                                2        151        184          1
                                3          1       1018          4
                                3          2       1017          4
                                3          3         82          4
                                3          4         19          1
                                3          5         55          4
                                3         16         11          1
                                3         22          7          1
                                3         28          8          1
                                3         51         15          1
                                3         52         14          1
                                3         53         13          1
                                3         54         13          1
                                3         55         13          1
                                3         56         13          1
                                3         57         13          1
                                3         58         13          1
                                3         59         12          1
                                3         60         12          1
                                3         61         12          1
                                3         62         12          1
                                3         63         12          1
                                3         91         12          1
                                3        102         12          5
                                3        103         13          1
                                3        104          8          1
                                3        105          7          1
                                3        106          7          1
                                3        107         11          1
                                4          2        891          4
                                6          5      13168          4
                                6         16        415          1
                                6         22        435          1
                                6         28        534          1
                                6         51         47          1
                                6         52         44          1
                                6         53         44          1
                                6         54         44          1
                                6         55          7          1
                                6         56          7          1
                                6         57          7          1
                                6         58          7          1
                                6         59          7          1
                                6         60          7          1
                                6         61          7          1
                                6         62          7          1
                                6         63         41          3
                                6         91       1547          1
                                6        103        934          1
                                6        107        782          1
                               12         51          3          1
                               12         52          3          1
                               12         53          3          1
                               12         54          3          1
                               12         55          3          1
                               12         56          3          1
                               12         57          4          1
                               12         58          3          1
                               12         59          3          1
                               12         60          3          1
                               12         61          3          1
                               12         62          3          1
                               12         63          3          1
                               12        102       1709          5
                               14         51          1          1
                               14         52          1          1
                               14         53          1          1
                               14         54          1          1
                               14         55          1          1
                               14         56          1          1
                               14         57          1          1
                               14         58          1          1
                               14         59          1          1
                               14         60          1          1
                               14         61          1          1
                               14         62          1          1
                               14         63          1          1
                               14        102          1          5
                               15        102         88          1
                               33          1         20          1
                               33          2         14          1
                               33          3         18          1
                               33          4         10          1
                               33         16         63          1
                               33         22         14          1
                               33         28         14          1
                               33         51         12          1
                               33         52         12          1
                               33         53         12          1
                               33         54         12          1
                               33         55         12          1
                               33         56         12          1
                               33         57         12          1
                               33         58         12          1
                               33         59         12          1
                               33         60         12          1
                               33         61         12          1
                               33         62         12          1
                               33         63         12          1
                               33         91         12          1
                               33        102         17          1
                               33        103         14          1
                               33        104         12          1
                               33        105         10          1
                               33        106         10          1
                               33        107         18          1