76-Os-190

 76-Os-190 JAEA       EVAL-Jan10 N.Iwamoto                        
                      DIST-MAY10                       20100121   
----JENDL-4.0         MATERIAL 7643                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
10-01 The resolved resonance parameters were evaluated by         
      N.Iwamoto.                                                  
      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: below 800 eV                       
      Resolved resonance parameters were taken from Mughabghab    
      /1/. The unknown reduced neutron width and angular          
      momentum were assumed to be 1 meV and L=0, respectively,    
      for levels with energy below 721.5 eV.  The negative        
      resonance was placed so as to reproduce the cross sections  
      at thermal energy recommended by Mughabghab /1/.            
                                                                  
    Unresolved resonance region : 800 eV - 200 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   
      CCOM /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           2.9456e+01                                 
       Elastic         1.6348e+01                                 
       n,gamma         1.3108e+01           2.5017e+01            
       n,alpha         1.9876e-09                                 
      ----------------------------------------------------------  
         (*) 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= 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= 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: 0,1,2,7,26 (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/                   
      (+) omp parameters were modified.                           
                                                                  
  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 enhanced 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 Os-190                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.18672   2  +  *                                           
   2  0.54785   4  +  *                                           
   3  0.55798   2  +                                              
   4  0.75602   3  +                                              
   5  0.91178   0  +                                              
   6  0.95537   4  +                                              
   7  1.05038   6  +  *                                           
   8  1.11472   2  +                                              
   9  1.11550   1  -                                              
  10  1.16319   4  +                                              
  11  1.20386   5  +                                              
  12  1.32690   1  +                                              
  13  1.38240   2  +                                              
  14  1.38700   3  -                                              
  15  1.43582   2  +                                              
  16  1.44612   5  +                                              
  17  1.47420   6  +                                              
  18  1.48200   1  -                                              
  19  1.51410   5  +                                              
  20  1.54530   0  +                                              
  21  1.54720   1  +                                              
  22  1.56904   3  +                                              
  23  1.57030   2  +                                              
  24  1.58388   4  -                                              
  25  1.61597   2  +                                              
  26  1.66647   8  +  *                                           
  27  1.67568   2  +                                              
  28  1.67950   3  +                                              
  29  1.68060   1  +                                              
  30  1.68169   5  -                                              
  31  1.68906   2  +                                              
  32  1.70540  10  -                                              
  33  1.70830   2  +                                              
  34  1.72480   1  -                                              
  35  1.73290   0  +                                              
  36  1.77700   0  +                                              
  37  1.80270   0  +                                              
  38  1.81350   0  +                                              
  39  1.82360   0  +                                              
  40  1.83632   6  +                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Os-191 24.1000  0.8683  0.1619  0.5157 -1.2338  5.6493         
   Os-190 23.1000  1.7411  0.3025  0.5036 -0.0238  6.0343         
   Os-189 23.6000  0.8729  0.7207  0.5092 -1.1969  5.5241         
   Os-188 23.1000  1.7504  0.7049  0.4799  0.1757  5.6719         
   Re-190 22.0088  0.0000  0.6386  0.3818 -0.4260  2.0158         
   Re-189 21.1918  0.8729  0.9313  0.5028 -0.6533  4.8062         
   Re-188 22.0000  0.0000  0.8388  0.4802 -1.4013  3.6931         
   Re-187 20.9976  0.8775  1.0287  0.5196 -0.8390  5.1039         
    W-189 22.5864  0.8729  1.4546  0.4721 -0.6930  4.7000         
    W-188 21.9779  1.7504  1.2256  0.5151 -0.1841  6.2107         
    W-187 23.5700  0.8775  1.1868  0.4614 -0.6754  4.6629         
    W-186 23.1400  1.7598  1.2307  0.4935 -0.1424  6.0929         
    W-185 22.7200  0.8823  1.2247  0.4842 -0.8101  4.9225         
    W-184 22.1100  1.7693  1.2350  0.5106 -0.1439  6.1796         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Os-191                   
  --------------------------------------------------------        
  K0 = 1.700   E0 = 4.500 (MeV)                                   
  * E1: ER = 12.68 (MeV) EG = 2.60 (MeV) SIG = 206.00 (mb)        
        ER = 14.40 (MeV) EG = 4.16 (MeV) SIG = 401.00 (mb)        
        ER =  5.50 (MeV) EG = 1.10 (MeV) SIG =   2.50 (mb)        
        ER =  2.10 (MeV) EG = 1.90 (MeV) SIG =   0.06 (mb)        
  * M1: ER =  7.12 (MeV) EG = 4.00 (MeV) SIG =   1.32 (mb)        
  * E2: ER = 10.94 (MeV) EG = 3.82 (MeV) SIG =   4.72 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Mughabghab,S.F.: "Atlas of Neutron Resonances, Fifth          
    Edition: Resonance Parameters and Thermal Cross Sections.     
    Z=1-100", Elsevier Science (2006).                            
 2) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
 3) Iwamoto,O.: JAERI-Data/Code 2003-020 (2003).                  
 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).