60-Nd-146

 60-Nd-146 JAEA+      EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata    
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6037                               
-----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 10 keV        
      Resonance energies were taken from Tellier/1/ and           
      Musgrove et al./2/  Neutron widths were adopted from        
      Tellier, and radiation widths were deduced from capture     
      areas measured by Musgrove et al.  The average radiation    
      widths were assumed to be 0.051 eV for s-wave resonances and
      0.040 eV for p-wave ones.  A negative resonance was added so
      as to reproduce the capture cross section of 1.4+-0.1 barns 
      at 0.0253 eV/3/.                                            
      In JENDL-4, the parameters at 359.9 eV were replaced with   
      those obtained by Barry et al./4/  The parameters for the   
      negative resonance were re-adjusted.                        
                                                                  
    Unresolved resonance region : 10.0 keV - 200.0 keV            
      The unresolved resonance paramters (URP) were determined by 
      ASREP code /5/ so as to reproduce the evaluated total and   
      capture cross sections calculated with optical model code   
      OPTMAN /6/ and CCONE /7/. 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           1.1281e+01                                 
       Elastic         9.7909e+00                                 
       n,gamma         1.4901e+00           2.6989e+00            
       n,alpha         1.4564e-07                                 
      ----------------------------------------------------------  
         (*) 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 /7/.                               
                                                                  
  MT= 16 (n,2n) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 17 (n,3n) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 22 (n,na) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 24 (n,2na) cross section                                    
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 28 (n,np) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 29 (n,n2a) cross section                                    
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 32 (n,nd) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 33 (n,nt) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 51-91 (n,n') cross section                                  
    Calculated with CCONE code /7/.                               
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code /7/.                               
                                                                  
  MT=103 (n,p) cross section                                      
    Calculated with CCONE code /7/.                               
                                                                  
  MT=104 (n,d) cross section                                      
    Calculated with CCONE code /7/.                               
                                                                  
  MT=105 (n,t) cross section                                      
    Calculated with CCONE code /7/.                               
                                                                  
  MT=106 (n,He3) cross section                                    
    Calculated with CCONE code /7/.                               
                                                                  
  MT=107 (n,a) cross section                                      
    Calculated with CCONE code /7/.                               
                                                                  
MF= 4 Angular distributions of emitted neutrons                   
  MT=  2 Elastic scattering                                       
    Calculated with CCONE code /7/.                               
                                                                  
MF= 6 Energy-angle distributions of emitted particles             
  MT= 16 (n,2n) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 17 (n,3n) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 22 (n,na) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 24 (n,2na) reaction                                         
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 28 (n,np) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 29 (n,n2a) reaction                                         
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 32 (n,nd) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 33 (n,nt) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 51-91 (n,n') reaction                                       
    Calculated with CCONE code /7/.                               
                                                                  
  MT=102 Capture reaction                                         
    Calculated with CCONE code /7/.                               
                                                                  
                                                                  
                                                                  
***************************************************************** 
       Nuclear Model Calculation with CCONE code /7/              
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE calculation             
  1) Optical model                                                
    * coupled channels calculation                                
      coupled levels: 0,1,3,4 (see Table 1)                       
                                                                  
    * optical model potential                                     
      neutron  omp: Kunieda,S. et al./8/ (+)                      
      proton   omp: Koning,A.J. and Delaroche,J.P./9/             
      deuteron omp: Lohr,J.M. and Haeberli,W./10/                 
      triton   omp: Becchetti Jr.,F.D. and Greenlees,G.W./11/     
      He3      omp: Becchetti Jr.,F.D. and Greenlees,G.W./11/     
      alpha    omp: McFadden,L. and Satchler,G.R./12/             
      (+) omp parameters were modified.                           
                                                                  
  2) Two-component exciton model/13/                              
    * Global parametrization of Koning-Duijvestijn/14/            
      was used.                                                   
    * Gamma emission channel/15/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Width fluctuation correction/16/ 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/17/.           
      Parameters are shown in Table 2.                            
    * Gamma-ray strength function of enhanced generalized         
      Lorentzian form/18/,/19/ 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 Nd-146                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.45377   2  +  *                                           
   2  0.91540   0  +                                              
   3  1.04222   4  +  *                                           
   4  1.18962   3  -  *                                           
   5  1.30310   2  +                                              
   6  1.37681   1  -                                              
   7  1.47059   2  +                                              
   8  1.51752   5  -                                              
   9  1.57200   0  +                                              
  10  1.60257   0  +                                              
  11  1.68800   2  -                                              
  12  1.69676   0  +                                              
  13  1.74496   4  +                                              
  14  1.76940   2  -                                              
  15  1.77716   3  +                                              
  16  1.78002   6  +                                              
  17  1.78730   2  +                                              
  18  1.81200   1  -                                              
  19  1.83400   2  +                                              
  20  1.88470   3  -                                              
  21  1.90534   2  +                                              
  22  1.91880   4  +                                              
  23  1.97845   2  +                                              
  24  1.98892   4  +                                              
  25  2.02700   1  -                                              
  26  2.02942   7  -                                              
  27  2.04553   4  -                                              
  28  2.06900   5  -                                              
  29  2.07259   3  -                                              
  30  2.08352   6  +                                              
  31  2.09000   0  +                                              
  32  2.09579   4  +                                              
  33  2.11945   2  +                                              
  34  2.14386   2  +                                              
  35  2.14895   0  +                                              
  36  2.16600   3  -                                              
  37  2.19728   2  +                                              
  38  2.20836   2  +                                              
  39  2.21975   3  +                                              
  40  2.22500   1  -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Nd-147 19.7000  0.9897  2.4886  0.4934 -0.5694  4.7470         
   Nd-146 18.1900  1.9863  1.6792  0.5692  0.1138  6.4542         
   Nd-145 18.5400  0.9965  1.1101  0.5235 -0.2928  4.6189         
   Nd-144 17.5000  2.0000  0.3419  0.6111  0.2496  6.6190         
   Pr-146 17.5893  0.0000  2.4188  0.5462 -1.6453  4.0472         
   Pr-145 16.8637  0.9965  1.7883  0.6002 -0.8883  5.5766         
   Pr-144 15.5000  0.0000  0.9153  0.6715 -1.9662  5.0412         
   Pr-143 16.6639  1.0035  0.4682  0.6161 -0.5920  5.4208         
   Ce-145 18.2180  0.9965  1.7406  0.5686 -0.8969  5.4793         
   Ce-144 17.4894  2.0000  1.0129  0.5822  0.3675  6.2813         
   Ce-143 19.6000  1.0035  0.4100  0.4774  0.1189  3.9645         
   Ce-142 18.9500  2.0140 -0.3155  0.5558  0.6875  5.9346         
   Ce-141 17.9000  1.0106 -1.0773  0.4985  0.5829  3.4550         
   Ce-140 17.0742  2.0284 -1.9470  0.5674  1.4861  4.9920         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Nd-147                   
  --------------------------------------------------------        
  K0 = 1.800   E0 = 4.500 (MeV)                                   
  * E1: ER = 13.27 (MeV) EG = 3.63 (MeV) SIG = 120.27 (mb)        
        ER = 15.83 (MeV) EG = 5.08 (MeV) SIG = 240.54 (mb)        
  * M1: ER =  7.77 (MeV) EG = 4.00 (MeV) SIG =   1.10 (mb)        
  * E2: ER = 11.94 (MeV) EG = 4.35 (MeV) SIG =   3.35 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Tellier, H.: CEA-N-1459 (1971).                               
 2) Musgrove, A.R. de L., et al.: AEEC/E401 (1977).               
 3) Fedorova, A.F., et al.: "Proc. 3rd All-union Conf. on Neutron 
    Physics, Kiev 1975", Vol. 1, 169.                             
 4) Barry, D.P. et al.: Nucl. Sci. Eng., 153, 8 (2006).           
 5) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
 6) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).   
 7) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).           
 8) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).     
 9) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)  
    [Global potential].                                           
10) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).      
11) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.             
    J.H.Williams Lab., Univ. Minnesota (1969).                    
12) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966).    
13) Kalbach,C.: Phys. Rev. C33, 818 (1986).                       
14) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).   
15) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).   
16) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).                  
17) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151   
    (1994).                                                       
18) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).              
19) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).