47-Ag-107

 47-Ag-107 JAEA       EVAL-Dec09 N.Iwamoto,K.Shibata              
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 4725                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
09-12 The resolved resonance parameters were evaluated by         
      K.Shibata.                                                  
      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 resonances                       
    Resolved resonance parameters (below 7.0095 keV)              
      Resolved resonance parameters (below 7.0095keV) for         
      JENDL-3.3 were based on the experimental data by Moxon and  
      Rae /1/, Garg et al./2/, Asghar et al./3/, Muradjan         
      and Adamchuk /4/, de Barros et al./5/, Pattenden and        
      Jolly /6/, macklin /7/ and Mizumoto et al./8/.              
      Total spin j and angular momentum l of some resonances were 
      estimated with a random number method and a method of       
      Bollinger and Thomas/9/, respectively.                      
      The capture cross section of JENDL-3.3 between 1.3 and 2.6  
      keV is too low compared with interpolated values from the   
      lower and higher energy regions.  To compensate the lower   
      capture cross section, p-wave resonances with a capture area
      of 0.04 eV were added every 40 eV between 1.28 and 2.04 keV,
      and every 15 eV between 2.04 and 2.64 keV.  The neutron     
      width was modified so as to reproduce teh capture area      
      measured by Macklin /7/.                                    
      In JENDL-4, the data for -11.1 eV - 778.8 eV were replaced  
      with the ones obtained by Lowie et al./10/  A value of 140  
      meV was used for unknown radiation widths.  Part of unknown 
      j values were determined by considering the work of Zanini  
      et al./11/  The remaining unknown j values were estimated   
      by a random number method.                                  
                                                                  
    Unresolved resonance region : 7.0095 keV - 120 keV            
      The unresolved resonance paramters (URP) were determined by 
      ASREP code /12/ so as to reproduce the evaluated total and  
      capture cross sections calculated with optical model code   
      OPTMAN /13/ and CCONE /14/. 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           4.5197e+01                                 
       Elastic         7.5518e+00                                 
       n,gamma         3.7645e+01           1.0552e+02            
       n,alpha         1.4043e-08                                 
      ----------------------------------------------------------  
         (*) 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 /14/.                              
                                                                  
  MT= 16 (n,2n) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 17 (n,3n) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 22 (n,na) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 24 (n,2na) cross section                                    
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 28 (n,np) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 32 (n,nd) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 33 (n,nt) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 41 (n,2np) cross section                                    
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 51-91 (n,n') cross section                                  
    Calculated with CCONE code /14/.                              
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code /14/.                              
                                                                  
  MT=103 (n,p) cross section                                      
    Calculated with CCONE code /14/.                              
                                                                  
  MT=104 (n,d) cross section                                      
    Calculated with CCONE code /14/.                              
                                                                  
  MT=105 (n,t) cross section                                      
    Calculated with CCONE code /14/.                              
                                                                  
  MT=106 (n,He3) cross section                                    
    Calculated with CCONE code /14/.                              
                                                                  
  MT=107 (n,a) cross section                                      
    Calculated with CCONE code /14/.                              
                                                                  
  MT=111 (n,2p) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
  MT=112 (n,pa) cross section                                     
    Calculated with CCONE code /14/.                              
                                                                  
MF= 4 Angular distributions of emitted neutrons                   
  MT=  2 Elastic scattering                                       
    Calculated with CCONE code /14/.                              
                                                                  
MF= 6 Energy-angle distributions of emitted particles             
  MT= 16 (n,2n) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 17 (n,3n) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 22 (n,na) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 24 (n,2na) reaction                                         
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 28 (n,np) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 32 (n,nd) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 33 (n,nt) reaction                                          
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 41 (n,2np) reaction                                         
    Calculated with CCONE code /14/.                              
                                                                  
  MT= 51-91 (n,n') reaction                                       
    Calculated with CCONE code /14/.                              
                                                                  
  MT=102 Capture reaction                                         
    Calculated with CCONE code /14/.                              
                                                                  
                                                                  
                                                                  
***************************************************************** 
       Nuclear Model Calculation with CCONE code /14/             
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE calculation             
  1) Optical model                                                
    * coupled channels calculation                                
      coupled levels: 0,3,4,9,14 (see Table 1)                    
                                                                  
    * optical model potential                                     
      neutron  omp: Kunieda,S. et al./15/ (+)                     
      proton   omp: Koning,A.J. and Delaroche,J.P./16/            
      deuteron omp: Lohr,J.M. and Haeberli,W./17/                 
      triton   omp: Becchetti Jr.,F.D. and Greenlees,G.W./18/     
      He3      omp: Becchetti Jr.,F.D. and Greenlees,G.W./18/     
      alpha    omp: Huizenga,J.R. and Igo,G./19/                  
      (+) omp parameters were modified.                           
                                                                  
    * resonance for pseudo levels was calculated on the basis of  
      DWBA.                                                       
      ER= 2.500 (MeV)  WIDTH= 0.400 (MeV)  L=  3  BETA= 0.155     
                                                                  
  2) Two-component exciton model/20/                              
    * Global parametrization of Koning-Duijvestijn/21/            
      was used.                                                   
    * Gamma emission channel/22/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Width fluctuation correction/23/ 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/24/.           
      Parameters are shown in Table 2.                            
    * Gamma-ray strength function of generalized Lorentzian form  
      /25/,/26/ 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 Ag-107                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000  1/2 -  *                                           
   1  0.09312  7/2 +                                              
   2  0.12559  9/2 +                                              
   3  0.32481  3/2 -  *                                           
   4  0.42315  5/2 -  *                                           
   5  0.77331 11/2 +                                              
   6  0.78659  3/2 -                                              
   7  0.92206  5/2 +                                              
   8  0.94970  5/2 -                                              
   9  0.97330  7/2 -  *                                           
  10  0.99100 13/2 +                                              
  11  1.06120  7/2 +                                              
  12  1.14200  1/2 +                                              
  13  1.14306  5/2 -                                              
  14  1.14690  9/2 -  *                                           
  15  1.22200 11/2 -                                              
  16  1.22301  5/2 +                                              
  17  1.25889  3/2 +                                              
  18  1.32580  3/2 +                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Ag-108 14.6000  0.0000  2.5216  0.6708 -2.2742  5.0724         
   Ag-107 15.6000  1.1601  1.9728  0.5979 -0.4714  5.2925         
   Ag-106 13.4291  0.0000  1.2475  0.7902 -2.6899  6.1687         
   Ag-105 12.8006  1.1711  0.7422  0.8304 -1.4372  7.4941         
   Pd-107 15.0000  1.1601  3.1932  0.6723 -1.5375  6.6188         
   Pd-106 14.4000  2.3311  2.3412  0.6736  0.1590  7.3089         
   Pd-105 14.9000  1.1711  2.0672  0.7067 -1.5220  6.8969         
   Pd-104 13.5000  2.3534  1.1560  0.7879 -0.3130  8.4969         
   Rh-106 14.2000  0.0000  3.7991  0.5945 -1.6674  4.0000         
   Rh-105 15.8000  1.1711  3.4219  0.6193 -1.2405  6.1130         
   Rh-104 14.1000  0.0000  2.9724  0.6799 -2.3482  5.1092         
   Rh-103 15.8000  1.1824  2.3988  0.6206 -0.9205  5.8890         
   Rh-102 15.0000  0.0000  1.6557  0.6874 -2.3483  5.3149         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Ag-108                   
  --------------------------------------------------------        
  * E1: ER = 15.90 (MeV) EG = 6.71 (MeV) SIG = 150.00 (mb)        
        ER =  6.40 (MeV) EG = 1.80 (MeV) SIG =   1.50 (mb)        
  * M1: ER =  8.61 (MeV) EG = 4.00 (MeV) SIG =   1.27 (mb)        
  * E2: ER = 13.23 (MeV) EG = 4.81 (MeV) SIG =   2.53 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Moxon, M.C., Rae, E.R., "Proc. EANDC Conf. on Time-of-Flight  
    Methods, Saclay, 1961",439.                                   
 2) Garg, J.B., et al., Phys. Rev., B137, 547(1965).              
 3) Asghar, M., et al., "Proc. Int. Conf. on the Study of Nuclear 
    Structure with Neutrons, Antwerp 1965",(65).                  
 4) Muradjan, G.V., Adamchuk, Ju. V., Jaderno-Fizicheskie         
    Issledovanija, 6, 64 (1968).                                  
 5) de Barros, S., et al., Nucl. Phys., A131, 305(1969).          
 6) Pattenden, N.J., Jolly, J.E., AERE-PR/NP-16(1969).            
 7) Macklin, R.L., Nucl. Sci. Eng., 82, 400(1982).                
 8) Mizumoto, M., et al., J. Nucl. Sci. Technol., 20, 883(1983).  
 9) Bollinger, L.M., Thomas, G.E.: Phys. Rev., 171,1293(1968).    
10) Lowie, L.Y., Phys. Rev., C59, 1119 (1999); Phys. Rev., C56,   
    90 (1997).                                                    
11) Zanini, L, et al., JINR-E3-97-213, p.221 (1997).              
12) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
13) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).   
14) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).           
15) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).     
16) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)  
    [Global potential].                                           
17) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).      
18) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.             
    J.H.Williams Lab., Univ. Minnesota (1969).                    
19) Huizenga,J.R. and Igo,G.: Nucl. Phys. 29, 462 (1962).         
20) Kalbach,C.: Phys. Rev. C33, 818 (1986).                       
21) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).   
22) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).   
23) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).                  
24) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151   
    (1994).                                                       
25) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).              
26) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).