64-Gd-152

 64-Gd-152 JAEA+      EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata    
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6425                               
-----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 2.66 keV     
      Resonance parameters below 10 eV were evaluated on the      
      basis of Mughabghab/1/.  Above 12 eV, parameters were       
      adopted from Macklin/2/.  For the resonances only whose     
      capture area was measured, neutron widths were determined   
      from the capture area and an average radiation width of     
      0.0586 eV/2/.  The total spin J and orbital angular         
      momentum L were assigned by considering the magnitude       
      of the capture area of each resonance.  A negative resonance
      was added so as to reproduce the thermal capture cross      
      section of 735+-20 barns and the capture resonance integral 
      of 2020+-160 barns/1/.  Scattering radius of 8.2 fm was     
      estimated from an optical model calculation shown in fig. 2 
      of ref./1/                                                  
      In JENDL-4, the data for 12.46 - 184 eV were replaced with  
      the ones obtained by Anufriev et al./3/  The 207.7-eV       
      resonance was removed, since it was not observed by         
      Leinweber et al./4/  The energy of the negative resonance   
      was adjusted so as to reproduce the thermal capture cross   
      section recommended by Mughabghab./5/                       
                                                                  
    Unresolved resonance region : 2.66 keV - 300.0 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           7.4555e+02                                 
       Elastic         1.0452e+01                                 
       n,gamma         7.3509e+02           9.3512e+02            
       n,alpha         7.0393e-03                                 
      ----------------------------------------------------------  
         (*) 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= 25 (n,3na) 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= 33 (n,nt) 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/.                               
                                                                  
  MT=108 (n,2a) 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= 25 (n,3na) 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= 33 (n,nt) 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: 0,1,3,8,20 (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-152                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.34428   2  +  *                                           
   2  0.61540   0  +                                              
   3  0.75540   4  +  *                                           
   4  0.93054   2  +                                              
   5  1.04785   0  +                                              
   6  1.10917   2  +                                              
   7  1.12319   3  -                                              
   8  1.22738   6  +  *                                           
   9  1.28226   4  +                                              
  10  1.31465   1  -                                              
  11  1.31842   2  +                                              
  12  1.43402   3  +                                              
  13  1.46053   1  +                                              
  14  1.47048   5  -                                              
  15  1.55021   4  +                                              
  16  1.60560   2  +                                              
  17  1.64341   2  -                                              
  18  1.66807   6  +                                              
  19  1.69241   3  +                                              
  20  1.74678   8  +  *                                           
  21  1.75598   1  -                                              
  22  1.77157   3  -                                              
  23  1.80766   5  -                                              
  24  1.83962   2  +                                              
  25  1.86158   5  +                                              
  26  1.86205   2  +                                              
  27  1.88030   7  -                                              
  28  1.91542   3  -                                              
  29  1.94116   2  +                                              
  30  1.96200   0  -                                              
  31  1.97567   2  +                                              
  32  1.99789   6  +                                              
  33  2.01163   3  +                                              
  -------------------                                             
  *) 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-153 20.9000  0.9701  3.9793  0.5231 -1.6694  5.9506         
   Gd-152 18.3157  1.9467  3.2774  0.5203  0.2124  5.9623         
   Gd-151 18.8247  0.9765  2.9209  0.5214 -0.8124  5.0822         
   Gd-150 18.1096  1.9596  2.0439  0.5067  0.7184  5.4062         
   Eu-152 19.7700  0.0000  4.2144  0.5244 -2.4180  4.7265         
   Eu-151 21.0000  0.9765  3.8814  0.4854 -1.1094  5.2279         
   Eu-150 20.0000  0.0000  3.1727  0.4165 -0.9266  2.6887         
   Eu-149 17.2625  0.9831  2.5338  0.5772 -0.9591  5.4887         
   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         
   Sm-148 18.4000  1.9728  2.0339  0.5337  0.3686  5.9610         
   Sm-147 18.4207  0.9897  1.4097  0.5385 -0.5090  4.9131         
   Sm-146 17.6964  1.9863  0.5792  0.5450  0.8159  5.5739         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Gd-153                   
  --------------------------------------------------------        
  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)        
  * M1: ER =  7.67 (MeV) EG = 4.00 (MeV) SIG =   1.86 (mb)        
  * E2: ER = 11.78 (MeV) EG = 4.27 (MeV) SIG =   3.73 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Mughabghab, S.F.: "Neutron Cross Sections, Vol. I, Part B",   
    Academic Press (1984).                                        
 2) Macklin, R.L.: Nucl. Sci. Eng. 95, 304 (1987).                
 3) Anufriev, V.A. et al.: 87 Kiev, 2, 225 (1987).                
 4) Leinweber, G et al.: Nucl. Sci. Eng., 154, 261 (2006).        
 5) Mughabghab, S.F.: "Atlas of Neutron Resonances", Elsevier     
    (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).