20-Ca- 44

 20-Ca- 44 JAEA       EVAL-JUN06 K.Shibata                        
                      DIST-MAY10                       20091228   
----JENDL-4.0         MATERIAL 2037                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
06-06 Evaluated by K.Shibata.                                     
09-12 Compiled by K.Shibata                                       
                                                                  
MF=1  General information                                         
  MT=451  Descriptive data and dictionary                         
                                                                  
MF=2  Resonance parameters                                        
  MT=151  Resolved resonance parameters                           
   The resolved resonance region remains unchanged from JENDL-3.3.
   Resolved parameters for MLBW formula were given in the energy  
   region from 1.0e-5 eV to 500 keV.  Parameters were taken from  
   the recommended data of BNL/1/ and the data for a negative     
   resonance were added so as to reproduce the recommended thermal
   cross sections for capture and scatterng/1/.  The scattering   
   radius was assumed to be 3.6 fermi.                            
                                                                  
    Thermal cross sections and resonance integrals at 300 K       
    ----------------------------------------------------------    
                     0.0253 eV           res. integ. (*)          
                      (barns)              (barns)                
    ----------------------------------------------------------    
     Total           4.2473E+00                                   
     Elastic         3.3587E+00                                   
     n,gamma         8.8863E-01           4.2453E-01              
    ----------------------------------------------------------    
       (*) Integrated from 0.5 eV to 10 MeV.                      
                                                                  
MF=3  Neutron cross sections                                      
   Below 500 keV, zero background cross section was given and all 
   the cross-section data are reproduced from the evaluated       
   resolved resonance parameters with MLBW formula.               
                                                                  
   The cross sections were calcualted /2/ by using the TNG code   
   /3/.  The optilcal model parameters of Koning and Delaroche /4/
   were used for neutrons and protons.  The alpha-particle        
   potential parameters were derived from the code developed by   
   Kumar and Kailas./5/                                           
                                                                  
  MT= 1 Total                                                     
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 2 Elastic scattering                                        
   Obtained by subtracting the sum of the partial cross sections  
   from the total cross section.                                  
                                                                  
  MT= 4, 51-89, 91 Inelastic scattering                           
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 16      (n,2n)                                              
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 17      (n,3n)                                              
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 22      (n,na)                                              
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 28      (n,np)                                              
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 102     Capture                                             
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 103     (n,p)                                               
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 107     (n,a)                                               
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 600-649 partial (n,p) cross sections                        
   The cross sections were calculated with the TNG code./3/       
                                                                  
  MT= 800-849 partial (n,a) cross sections                        
   The cross sections were calculated with the TNG code./3/       
                                                                  
MF=4  Angular distributions of secondary neutrons                 
  MT=2                                                            
    Calculated with the TNG code/3/.                              
                                                                  
MF=6  Energy-angle distributions of secondary particles           
  MT= 16 (n,2n) reaction                                          
   Neutron and gamma-ray spectra calculated with TNG/3/.          
                                                                  
  MT= 17 (n,3n) reaction                                          
   Neutron calculated with TNG/3/.  Gamma-ray channel is not open.
                                                                  
  MT= 22 (n,na) reaction                                          
   Neutron, alpha-particle, and gamma-ray spectra calculated with 
   TNG/3/.                                                        
                                                                  
  MT= 28 (n,np) reaction                                          
   Neutron, proton, and gamma-ray spectra calculated with TNG/3/. 
                                                                  
  MT= 51-89 (n,n') reaction                                       
   Neutron angular distributions and discrete gamma-ray spectra   
   calculated with TNG/3/.                                        
                                                                  
  MT= 91 (n,n') reaction                                          
   Neutron spectra, and discrete-continuous gamma-ray spectra     
   calculated with  with TNG/3/.                                  
                                                                  
  MT= 102                                                         
   Calculated with the TNG code /3/.                              
                                                                  
  MT= 600-603 (n,p) reactions leading to discrete levels          
   Proton angular distributions and discrete gamma-ray spectra    
   calculated with TNG/3/.                                        
                                                                  
  MT= 649 (n,p) reaction leading to continuum levels              
   Proton spectra and discrete-continuous gamma-ray spectra       
   calculated with TNG/3/.                                        
                                                                  
  MT= 800-829 (n,a) reactions leading to discrete levels          
   Alpha-particle angular distributions and gamma-ray spectra     
   calculated with TNG/3/.                                        
                                                                  
  MT= 849 (n,a) reaction leading to continuum levels              
   Alpha-particle spectra and discrete-continuous gamma-ray       
   spectra calculated with TNG/3/.                                
                                                                  
< Appendix >                                                      
******************************************************************
*            Nuclear Model Calcualtions with TNG Code /3/        *
******************************************************************
The description of the model calculations is given in Ref.2.      
                                                                  
< Optical model parameters >                                      
Neutron and protons:                                              
  Koning and Delaroche /4/                                        
Alphas:                                                           
  The potential parameters were obtained using the code developed 
  by Kumar and Kailas./5/                                         
                                                                  
< Level scheme of Ca- 44 >                                        
  -------------------------                                       
   No.   Ex(MeV)     J  PI                                        
  -------------------------                                       
    0    0.00000     0   +                                        
    1    1.15710     2   +                                        
    2    1.88350     0   +                                        
    3    2.28310     4   +                                        
    4    2.65650     2   +                                        
    5    3.04430     4   +                                        
    6    3.28500     6   +                                        
    7    3.30130     2   +                                        
    8    3.30790     3   -                                        
    9    3.35720     3   +                                        
   10    3.58040     3   +                                        
   11    3.66150     1   -                                        
   12    3.67610     2   -                                        
   13    3.71180     4   -                                        
   14    3.77620     2   +                                        
   15    3.86500     5   -                                        
   16    3.91350     5   -                                        
   17    3.92260     3   +                                        
   18    4.01140     3   +                                        
   19    4.09200     2   +                                        
   20    4.16900     1   +                                        
   21    4.19570     2   +                                        
   22    4.26020     3   +                                        
   23    4.31520     3   +                                        
   24    4.35840     3   -                                        
   25    4.39940     3   -                                        
   26    4.40920     1   -                                        
   27    4.43700     2   +                                        
   28    4.47980     2   +                                        
   29    4.55260     1   -                                        
   30    4.56490     5   -                                        
   31    4.57300     3   -                                        
   32    4.58410     3   +                                        
   33    4.60400     1   +                                        
   34    4.65100     2   +                                        
   35    4.69020     2   +                                        
   36    4.80370     1   -                                        
   37    4.82400     2   -                                        
   38    4.86610     2   +                                        
   39    4.88400     2   -                                        
                                                                  
The direct-reaction process was taken into account for the 1st,   
3rd, 4th, 8th, 16th, and 25th levels by DWBA.                     
                                                                  
< Level density parameters >                                      
Energy dependent parameters of Mengoni-Nakajima /6/ were used.    
  ----------------------------------------------------------      
  Nuclei    a*    Pair    Esh     T     E0    Ematch Econt        
          1/MeV   MeV     MeV    MeV    MeV    MeV    MeV         
  ----------------------------------------------------------      
  Ca- 45   7.212  1.789  0.882  1.223 -0.908  9.109  3.556        
  Ca- 44   6.485  3.618  1.143  1.463 -0.604 13.607  4.905        
  Ca- 43   6.962  1.830  0.922  1.322 -1.548 10.316  3.419        
  Ca- 42   6.243  3.703  0.470  1.550 -0.451 13.961  5.357        
  K - 44   6.494  0.000  1.303  1.184 -1.660  5.909  0.812        
  K - 43   6.062  1.830  2.010  1.410 -1.729 10.733  3.393        
  Ar- 41   8.342  1.874  2.006  0.721  1.512  4.471  3.968        
  Ar- 40   5.998  3.795  1.822  1.464 -0.092 13.254  5.378        
  ----------------------------------------------------------      
                                                                  
References                                                        
 1) Mughaghab S.F. et al.:"Neutron Cross Sections", Vol. 1, Part  
    A (1981).                                                     
 2) Shibata, K: J. Nucl. Sci. Technol., 44, 10 (2007).            
 3) Fu, C.Y.: ORNL/TM-7042 (1980); Shibata, K., Fu, C.Y.: ORNL/TM-
    10093.                                                        
 4) Koning, A.J., Delaroche, J.P.: Nucl. Phys., A713, 231 (2003). 
 5) Kumar, A., Kailas, S: a computer code contained in RIPL-2,    
    private communication (2002).                                 
 6) Mengoni, A., Nakajima, Y. Nucl. Sci. Technol., 31, 151 (1994).