64-Gd-155

 64-Gd-155 JAEA+      EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata    
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6434                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
09-12 The resolved resonance parameters were evaluated by         
      A.Zukeran,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 resonance parameters             
    Resolved resonance region (MLBW formula) : below 0.1818 keV   
      For JENDL-2, parameters of 3 levels below 2.6 ev were taken 
      from the data of Moller et al./1/  The data above 3.6 eV    
      were based on the measured data by Friesenhahn et al./2/    
      and by Ribon/3/.  The average radiation width of 0.12865    
      eV was assumed.  Scattering radius of 6.7 fm was adopted    
      from bnl 325(3rd.)/4/.                                      
      For JENDL-3, total spin J of J-unknown levels was estimated 
      with a random number method.                                
      In JENDL-4, the data for 0.02515 - 180.3 eV were replaced   
      with the ones obtained by Leinweber et al./5/  The energy   
      of the 1st level was slightly adjusted.                     
                                                                  
    Unresolved resonance region : 181.8 eV - 100.0 keV            
      The unresolved resonance paramters (URP) were determined by 
      ASREP code /6/ so as to reproduce the evaluated total and   
      capture cross sections calculated with optical model code   
      CCOM /7/ and CCONE /8/. 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           6.0795e+04                                 
       Elastic         5.9487e+01                                 
       n,gamma         6.0735e+04           1.5608e+03            
       n,alpha         8.1543e-05                                 
      ----------------------------------------------------------  
         (*) 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 /8/.                               
                                                                  
  MT= 16 (n,2n) cross section                                     
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 17 (n,3n) cross section                                     
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 22 (n,na) cross section                                     
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 24 (n,2na) cross section                                    
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 28 (n,np) cross section                                     
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 32 (n,nd) cross section                                     
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 41 (n,2np) cross section                                    
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 51-91 (n,n') cross section                                  
    Calculated with CCONE code /8/.                               
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code /8/.                               
                                                                  
  MT=103 (n,p) cross section                                      
    Calculated with CCONE code /8/.                               
                                                                  
  MT=104 (n,d) cross section                                      
    Calculated with CCONE code /8/.                               
                                                                  
  MT=105 (n,t) cross section                                      
    Calculated with CCONE code /8/.                               
                                                                  
  MT=106 (n,He3) cross section                                    
    Calculated with CCONE code /8/.                               
                                                                  
  MT=107 (n,a) cross section                                      
    Calculated with CCONE code /8/.                               
                                                                  
MF= 4 Angular distributions of emitted neutrons                   
  MT=  2 Elastic scattering                                       
    Calculated with CCONE code /8/.                               
                                                                  
MF= 6 Energy-angle distributions of emitted particles             
  MT= 16 (n,2n) reaction                                          
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 17 (n,3n) reaction                                          
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 22 (n,na) reaction                                          
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 24 (n,2na) reaction                                         
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 28 (n,np) reaction                                          
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 32 (n,nd) reaction                                          
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 41 (n,2np) reaction                                         
    Calculated with CCONE code /8/.                               
                                                                  
  MT= 51-91 (n,n') reaction                                       
    Calculated with CCONE code /8/.                               
                                                                  
  MT=102 Capture reaction                                         
    Calculated with CCONE code /8/.                               
                                                                  
                                                                  
                                                                  
***************************************************************** 
       Nuclear Model Calculation with CCONE code /8/              
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE calculation             
  1) Optical model                                                
    * coupled channels calculation                                
      coupled levels: 0,1,7,10,19,32 (see Table 1)                
                                                                  
    * optical model potential                                     
      neutron  omp: Kunieda,S. et al./9/ (+)                      
      proton   omp: Koning,A.J. and Delaroche,J.P./10/ (+)        
      deuteron omp: Lohr,J.M. and Haeberli,W./11/                 
      triton   omp: Becchetti Jr.,F.D. and Greenlees,G.W./12/     
      He3      omp: Becchetti Jr.,F.D. and Greenlees,G.W./12/     
      alpha    omp: Huizenga,J.R. and Igo,G./13/ (+)              
      (+) omp parameters were modified.                           
                                                                  
  2) Two-component exciton model/14/                              
    * Global parametrization of Koning-Duijvestijn/15/            
      was used.                                                   
    * Gamma emission channel/16/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Width fluctuation correction/17/ 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/18/.           
      Parameters are shown in Table 2.                            
    * Gamma-ray strength function of enhanced generalized         
      Lorentzian form/19/,/20/ 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 Gd-155                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000  3/2 -  *                                           
   1  0.06001  5/2 -  *                                           
   2  0.08655  5/2 +                                              
   3  0.10531  3/2 +                                              
   4  0.10758  9/2 +                                              
   5  0.11800  7/2 +                                              
   6  0.12105 11/2 -                                              
   7  0.14607  7/2 -  *                                           
   8  0.21435 13/2 +                                              
   9  0.23013 11/2 +                                              
  10  0.25171  9/2 -  *                                           
  11  0.26665  5/2 +                                              
  12  0.26862  3/2 +                                              
  13  0.28257 13/2 -                                              
  14  0.28700  3/2 -                                              
  15  0.32138  5/2 -                                              
  16  0.32609  5/2 +                                              
  17  0.35043  7/2 +                                              
  18  0.36763  1/2 +                                              
  19  0.39232 11/2 -  *                                           
  20  0.39353  7/2 -                                              
  21  0.42341  7/2 +                                              
  22  0.42381 17/2 +                                              
  23  0.42724  3/2 +                                              
  24  0.45056  3/2 -                                              
  25  0.45137  1/2 -                                              
  26  0.45367 15/2 +                                              
  27  0.45448  5/2 -                                              
  28  0.46383 15/2 -                                              
  29  0.48000  5/2 -                                              
  30  0.48597  9/2 -                                              
  31  0.48872  5/2 +                                              
  32  0.53430 13/2 -  *                                           
  33  0.55337  7/2 -                                              
  34  0.55937  1/2 -                                              
  35  0.58146  5/2 -                                              
  36  0.59214  3/2 -                                              
  37  0.61084  7/2 +                                              
  38  0.61486  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         
  --------------------------------------------------------        
   Gd-156 19.0000  1.9215  3.2702  0.5513 -0.3880  6.7098         
   Gd-155 20.5000  0.9639  3.7045  0.5229 -1.4800  5.7609         
   Gd-154 18.5215  1.9340  3.6018  0.5706 -0.6048  7.0075         
   Gd-153 20.9000  0.9701  3.9793  0.5231 -1.6694  5.9506         
   Eu-155 17.9000  0.9639  3.3259  0.5676 -1.2578  5.7030         
   Eu-154 19.2000  0.0000  3.6717  0.5485 -2.4486  4.8922         
   Eu-153 17.3400  0.9701  3.8805  0.5963 -1.6297  6.1695         
   Eu-152 19.7700  0.0000  4.2144  0.5244 -2.4180  4.7265         
   Sm-154 18.5215  1.9340  3.2136  0.5576 -0.3117  6.6726         
   Sm-153 20.0000  0.9701  3.6781  0.5579 -1.8633  6.3072         
   Sm-152 19.7000  1.9467  3.6242  0.5066 -0.0488  6.1904         
   Sm-151 20.8000  0.9765  3.9732  0.5224 -1.6295  5.9141         
   Sm-150 19.2000  1.9596  3.2458  0.5078  0.1619  6.0033         
   Sm-149 19.2000  0.9831  2.9030  0.5042 -0.6887  4.8887         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Gd-156                   
  --------------------------------------------------------        
  K0 = 2.000   E0 = 4.500 (MeV)                                   
  * E1: ER = 11.20 (MeV) EG = 2.60 (MeV) SIG = 180.00 (mb)        
        ER = 15.20 (MeV) EG = 3.60 (MeV) SIG = 242.00 (mb)        
        ER =  3.00 (MeV) EG = 1.00 (MeV) SIG =   0.40 (mb)        
        ER =  6.00 (MeV) EG = 2.00 (MeV) SIG =   2.00 (mb)        
  * M1: ER =  7.62 (MeV) EG = 4.00 (MeV) SIG =   2.03 (mb)        
  * E2: ER = 11.70 (MeV) EG = 4.24 (MeV) SIG =   3.69 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Moller, H.B., et al.: Nucl. Sci. Eng., 8, 183 (1960).         
 2) Friesenhahn, S.J., et al.: Nucl. Phys., A146, 337 (1970).     
 3) Ribon, P.: CEA-N-1149 (1969).                                 
 4) Mughabghab, S.F. and Garber, D.I.: "Neutron Cross Sections,   
    Vol. I, Resonance Parameters", BNL 325, 3rd ed., Vol. 1,      
    (1973).                                                       
 5) Leinweber, G et al.: Nucl. Sci. Eng., 154, 261 (2006).        
 6) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
 7) Iwamoto,O.: JAERI-Data/Code 2003-020 (2003).                  
 8) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).           
 9) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).     
10) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)  
    [Global potential].                                           
11) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).      
12) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.             
    J.H.Williams Lab., Univ. Minnesota (1969).                    
13) Huizenga,J.R. and Igo,G.: Nucl. Phys. 29, 462 (1962).         
14) Kalbach,C.: Phys. Rev. C33, 818 (1986).                       
15) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).   
16) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).   
17) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).                  
18) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151   
    (1994).                                                       
19) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).              
20) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).