66-Dy-163

 66-Dy-163 JAEA       EVAL-Nov09 N.Iwamoto,S.Chiba                
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6646                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
09-11 The resolved resonance parameters were evaluated by         
      S.Chiba.                                                    
      The data above the resolved resonance region were evaluated 
      and compiled by N.Iwamoto.                                  
                                                                  
MF= 1 General information                                         
  MT=451 Descriptive data and directory                           
                                                                  
MF= 2  Resonance parameters                                       
  MT=151 Resolved and unresolved resonance parameters             
    Resolved resonance region (MLBW formula) : below 1.003 keV    
      The evaluation is based on the work of Mughabghab /1/       
      A scattering radius of 7.5 fm was used.                     
                                                                  
    Unresolved resonance region : 1.003 keV - 160.0 keV           
      The unresolved resonance paramters (URP) were determined by 
      ASREP code /2/ so as to reproduce the evaluated total and   
      capture cross sections calculated with optical model code   
      OPTMAN /3/ and CCONE /4/. The unresolved parameters         
      should be used only for self-shielding calculation.         
                                                                  
      Thermal cross sections and resonance integrals at 300 K     
      ----------------------------------------------------------  
                       0.0253 eV           res. integ. (*)        
                        (barn)               (barn)               
      ----------------------------------------------------------  
       Total           1.2753e+02                                 
       Elastic         3.3418e+00                                 
       n,gamma         1.2419e+02           1.4908e+03            
       n,alpha         4.7968e-06                                 
      ----------------------------------------------------------  
         (*) Integrated from 0.5 eV to 10 MeV.                    
                                                                  
MF= 3 Neutron cross sections                                      
  MT=  1 Total cross section                                      
    Sum of partial cross sections.                                
                                                                  
  MT=  2 Elastic scattering cross section                         
    Obtained by subtracting non-elastic scattering cross sections 
      from total cross section.                                   
                                                                  
  MT=  4 (n,n') cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 16 (n,2n) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 17 (n,3n) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 22 (n,na) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 24 (n,2na) cross section                                    
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 28 (n,np) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 32 (n,nd) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 33 (n,nt) cross section                                     
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 41 (n,2np) cross section                                    
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 51-91 (n,n') cross section                                  
    Calculated with CCONE code /4/.                               
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code /4/.                               
                                                                  
  MT=103 (n,p) cross section                                      
    Calculated with CCONE code /4/.                               
                                                                  
  MT=104 (n,d) cross section                                      
    Calculated with CCONE code /4/.                               
                                                                  
  MT=105 (n,t) cross section                                      
    Calculated with CCONE code /4/.                               
                                                                  
  MT=106 (n,He3) cross section                                    
    Calculated with CCONE code /4/.                               
                                                                  
  MT=107 (n,a) cross section                                      
    Calculated with CCONE code /4/.                               
                                                                  
MF= 4 Angular distributions of emitted neutrons                   
  MT=  2 Elastic scattering                                       
    Calculated with CCONE code /4/.                               
                                                                  
MF= 6 Energy-angle distributions of emitted particles             
  MT= 16 (n,2n) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 17 (n,3n) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 22 (n,na) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 24 (n,2na) reaction                                         
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 28 (n,np) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 32 (n,nd) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 33 (n,nt) reaction                                          
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 41 (n,2np) reaction                                         
    Calculated with CCONE code /4/.                               
                                                                  
  MT= 51-91 (n,n') reaction                                       
    Calculated with CCONE code /4/.                               
                                                                  
  MT=102 Capture reaction                                         
    Calculated with CCONE code /4/.                               
                                                                  
                                                                  
                                                                  
***************************************************************** 
       Nuclear Model Calculation with CCONE code /4/              
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE calculation             
  1) Optical model                                                
    * coupled channels calculation                                
      coupled levels: 0,1,2,4,10,19,30 (see Table 1)              
                                                                  
    * optical model potential                                     
      neutron  omp: Kunieda,S. et al./5/                          
      proton   omp: Koning,A.J. and Delaroche,J.P./6/             
      deuteron omp: Lohr,J.M. and Haeberli,W./7/                  
      triton   omp: Becchetti Jr.,F.D. and Greenlees,G.W./8/      
      He3      omp: Becchetti Jr.,F.D. and Greenlees,G.W./8/      
      alpha    omp: Huizenga,J.R. and Igo,G./9/                   
                                                                  
  2) Two-component exciton model/10/                              
    * Global parametrization of Koning-Duijvestijn/11/            
      was used.                                                   
    * Gamma emission channel/12/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Width fluctuation correction/13/ was applied.               
    * Neutron, proton, deuteron, triton, He3, alpha and gamma     
      decay channel were taken into account.                      
    * Transmission coefficients of neutrons were taken from       
      optical model calculation.                                  
    * The level scheme of the target is shown in Table 1.         
    * Level density formula of constant temperature and Fermi-gas 
      model were used with shell energy correction/14/.           
      Parameters are shown in Table 2.                            
    * Gamma-ray strength function of generalized Lorentzian form  
      /15/,/16/ was used for E1 transition.                       
      For M1 and E2 transitions the standard Lorentzian form was  
      adopted. The prameters are shown in Table 3.                
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Level Scheme of Dy-163                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000  5/2 -  *                                           
   1  0.07344  7/2 -  *                                           
   2  0.16734  9/2 -  *                                           
   3  0.25089  5/2 +                                              
   4  0.28157 11/2 -  *                                           
   5  0.28559  7/2 +                                              
   6  0.33654  9/2 +                                              
   7  0.35115  1/2 -                                              
   8  0.38975  3/2 -                                              
   9  0.41238 11/2 +                                              
  10  0.41524 13/2 -  *                                           
  11  0.42184  3/2 -                                              
  12  0.42768  5/2 -                                              
  13  0.45000  7/2 -                                              
  14  0.47539  5/2 -                                              
  15  0.49720 13/2 +                                              
  16  0.51455  7/2 -                                              
  17  0.55302  7/2 -                                              
  18  0.56600  7/2 -                                              
  19  0.56871 15/2 -  *                                           
  20  0.58793  9/2 -                                              
  21  0.61200  1/2 -                                              
  22  0.64625  9/2 -                                              
  23  0.66000  3/2 +                                              
  24  0.70500  1/2 -                                              
  25  0.71147  5/2 -                                              
  26  0.71200  5/2 +                                              
  27  0.71827 11/2 -                                              
  28  0.72760  9/2 -                                              
  29  0.73766  1/2 +                                              
  30  0.74000 17/2 -  *                                           
  31  0.76621  3/2 +                                              
  32  0.78110  5/2 +                                              
  33  0.79339  1/2 -                                              
  34  0.80131  7/2 -                                              
  35  0.82080  3/2 -                                              
  36  0.82680  9/2 +                                              
  37  0.85112  7/2 +                                              
  38  0.85929  3/2 +                                              
  39  0.88301  5/2 -                                              
  40  0.88429  1/2 +                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Dy-164 19.6000  1.8741  2.0177  0.5619 -0.3641  6.7556         
   Dy-163 19.9000  0.9399  2.1664  0.5502 -1.2684  5.7265         
   Dy-162 19.5000  1.8856  2.4639  0.5582 -0.4126  6.7742         
   Dy-161 20.0000  0.9457  2.7684  0.5468 -1.4246  5.8432         
   Tb-163 18.6492  0.9399  2.1299  0.5457 -0.8637  5.2359         
   Tb-162 18.6000  0.0000  1.8944  0.4625 -0.8591  2.8270         
   Tb-161 18.4519  0.9457  2.4950  0.5607 -1.0970  5.5427         
   Tb-160 18.9000  0.0000  2.5855  0.5446 -1.9885  4.4661         
   Gd-162 19.3418  1.8856  2.3374  0.5664 -0.4417  6.8501         
   Gd-161 19.6000  0.9457  2.2447  0.5241 -0.8691  5.1638         
   Gd-160 19.1372  1.8974  2.4922  0.5804 -0.6190  7.1078         
   Gd-159 19.9000  0.9517  2.6302  0.5300 -1.1252  5.4620         
   Gd-158 19.3000  1.9093  2.8152  0.5596 -0.4648  6.8458         
   Gd-157 20.0000  0.9577  3.0516  0.5315 -1.2892  5.6268         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Dy-164                   
  --------------------------------------------------------        
  * E1: ER = 12.14 (MeV) EG = 3.06 (MeV) SIG = 140.23 (mb)        
        ER = 16.06 (MeV) EG = 5.22 (MeV) SIG = 280.47 (mb)        
        ER =  6.20 (MeV) EG = 4.50 (MeV) SIG =   1.10 (mb)        
        ER =  3.10 (MeV) EG = 1.50 (MeV) SIG =   0.60 (mb)        
  * M1: ER =  7.49 (MeV) EG = 4.00 (MeV) SIG =   0.97 (mb)        
  * E2: ER = 11.51 (MeV) EG = 4.14 (MeV) SIG =   3.82 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Mughabghab, S.F.: "Neutron Cross Sections, Vol. 1, Neutron    
   Resonance Parameters and Thermal Cross Sections, Part B",      
   Academic Press (1984).                                         
 2) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
 3) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).   
 4) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).           
 5) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).     
 6) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)  
    [Global potential].                                           
 7) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).      
 8) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.             
    J.H.Williams Lab., Univ. Minnesota (1969).                    
 9) Huizenga,J.R. and Igo,G.: Nucl. Phys. 29, 462 (1962).         
10) Kalbach,C.: Phys. Rev. C33, 818 (1986).                       
11) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).   
12) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).   
13) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).                  
14) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151   
    (1994).                                                       
15) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).              
16) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).