60-Nd-144

 60-Nd-144 JAEA+      EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata    
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6031                               
-----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 12 keV        
      Resonance parameters adopted in JENDL-3.1 were taken from   
      JENDL-2/1/:  resonance energies were taken from Tellier     
      /2/ and Musgrove et al./3/ by adjusting to those of ref.    
      /2/. Nneutron widths were taken from ref./2/, and           
      radiation widths were deduced from the capture areas of     
      Musgrove et al.  For the resonances not measured by Tellier,
      neutron widths were estimated from the capture areas by     
      assuming the average radiation widths of 0.047 eV for s-wave
      resonances and of 0.041 eV for p-wave ones.  For the lowest 
      2 levels, the capture widths of Karzhavina et al./4/ were   
      adopted.  A negative resonance was added at -76 eV so as to 
      reproduce the capture cross section of 3.8+-0.3 barns at    
      0.0253 eV /5/.                                              
      For JENDL-3.2, the capture data measured at ORELA of ORNL   
      were renormalized (factor = 0.967)/6/.  The neutron width   
      and/or the radiation width was revised to reproduce the     
      renormalized capture area for each resonance above 2.6 keV. 
      Effective scattering radius recommended in ref./7/ was      
      adopted and parameters of the nagative level were adjusted  
      to thermal cross sections/7/.                               
      In JENDL-4, the data for 2.8 - 19.9 keV were updated by     
      considering the capture area and g*Gamma_n data obtained by 
      Wisshak et al./8/  The resonance energies and angular       
      momenta (L, J) remain unchanged from JENDL-3.3.  When the   
      derived radiation width became negative, another J value    
      was assumed.  Moreover, the parameters for 374 eV were      
      replaced with the ones obtained by Barry et al./9/  The     
      parameters for the negative resonance were re-adjusted so as
      to reproduce the thermla capture cross section recommended  
      by Mughabghab./10/                                          
                                                                  
    Unresolved resonance region : 12.0 keV - 200.0 keV            
      The unresolved resonance paramters (URP) were determined by 
      ASREP code /11/ so as to reproduce the evaluated total and  
      capture cross sections calculated with optical model code   
      OPTMAN /12/ and CCONE /13/. 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           4.5280e+00                                 
       Elastic         9.0222e-01                                 
       n,gamma         3.6257e+00           6.3033e+00            
       n,alpha         1.5628e-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 /13/.                              
                                                                  
  MT= 16 (n,2n) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 17 (n,3n) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 22 (n,na) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 24 (n,2na) cross section                                    
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 28 (n,np) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 32 (n,nd) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 33 (n,nt) cross section                                     
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 41 (n,2np) cross section                                    
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 51-91 (n,n') cross section                                  
    Calculated with CCONE code /13/.                              
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code /13/.                              
                                                                  
  MT=103 (n,p) cross section                                      
    Calculated with CCONE code /13/.                              
                                                                  
  MT=104 (n,d) cross section                                      
    Calculated with CCONE code /13/.                              
                                                                  
  MT=105 (n,t) cross section                                      
    Calculated with CCONE code /13/.                              
                                                                  
  MT=106 (n,He3) cross section                                    
    Calculated with CCONE code /13/.                              
                                                                  
  MT=107 (n,a) cross section                                      
    Calculated with CCONE code /13/.                              
                                                                  
MF= 4 Angular distributions of emitted neutrons                   
  MT=  2 Elastic scattering                                       
    Calculated with CCONE code /13/.                              
                                                                  
MF= 6 Energy-angle distributions of emitted particles             
  MT= 16 (n,2n) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 17 (n,3n) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 22 (n,na) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 24 (n,2na) reaction                                         
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 28 (n,np) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 32 (n,nd) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 33 (n,nt) reaction                                          
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 41 (n,2np) reaction                                         
    Calculated with CCONE code /13/.                              
                                                                  
  MT= 51-91 (n,n') reaction                                       
    Calculated with CCONE code /13/.                              
                                                                  
  MT=102 Capture reaction                                         
    Calculated with CCONE code /13/.                              
                                                                  
                                                                  
                                                                  
***************************************************************** 
       Nuclear Model Calculation with CCONE code /13/             
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE calculation             
  1) Optical model                                                
    * coupled channels calculation                                
      coupled levels: 0,1,2,3 (see Table 1)                       
                                                                  
    * optical model potential                                     
      neutron  omp: Kunieda,S. et al./14/ (+)                     
      proton   omp: Koning,A.J. and Delaroche,J.P./15/            
      deuteron omp: Lohr,J.M. and Haeberli,W./16/                 
      triton   omp: Becchetti Jr.,F.D. and Greenlees,G.W./17/     
      He3      omp: Becchetti Jr.,F.D. and Greenlees,G.W./17/     
      alpha    omp: McFadden,L. and Satchler,G.R./18/             
      (+) omp parameters were modified.                           
                                                                  
  2) Two-component exciton model/19/                              
    * Global parametrization of Koning-Duijvestijn/20/            
      was used.                                                   
    * Gamma emission channel/21/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Width fluctuation correction/22/ 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/23/.           
      Parameters are shown in Table 2.                            
    * Gamma-ray strength function of generalized Lorentzian form  
      /24/,/25/ 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-144                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.69656   2  +  *                                           
   2  1.31467   4  +  *                                           
   3  1.51087   3  -  *                                           
   4  1.56092   2  +                                              
   5  1.79146   6  +                                              
   6  2.07291   2  +                                              
   7  2.08468   0  +                                              
   8  2.09328   5  -                                              
   9  2.10979   4  +                                              
  10  2.17897   3  +                                              
  11  2.18575   1  -                                              
  12  2.20480   4  -                                              
  13  2.21831   6  +                                              
  14  2.29541   4  +                                              
  15  2.32190   2  -                                              
  16  2.32818   0  +                                              
  17  2.34700   2  +                                              
  18  2.36882   2  +                                              
  19  2.39950   4  -                                              
  20  2.42021   5  +                                              
  21  2.45171   4  +                                              
  22  2.46400   1  -                                              
  23  2.49000   2  +                                              
  24  2.50842   3  -                                              
  25  2.52779   2  +                                              
  26  2.56451   3  +                                              
  27  2.58232   3  +                                              
  28  2.59000   1  -                                              
  29  2.59253   2  +                                              
  30  2.59900   3  -                                              
  31  2.60173   4  +                                              
  32  2.60300   3  +                                              
  33  2.60593   3  -                                              
  34  2.61307   7  -                                              
  35  2.61400   2  -                                              
  36  2.65510   3  +                                              
  37  2.65554   1  +                                              
  38  2.65600   4  +                                              
  39  2.67561   0  +                                              
  40  2.68167   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         
  --------------------------------------------------------        
   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         
   Nd-143 17.7000  1.0035 -0.4179  0.5516  0.0353  4.4179         
   Nd-142 15.0000  2.0140 -1.2557  0.6895  0.7987  6.4278         
   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         
   Pr-142 16.4000  0.0000 -0.4377  0.7390 -2.6336  6.4135         
   Pr-141 16.4637  1.0106 -1.2280  0.6590 -0.3966  5.5793         
   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         
   Ce-139 15.5000  1.0178 -1.1255  0.5922  0.4151  4.0889         
   Ce-138 16.8661  2.0430 -0.4123  0.5781  1.0263  5.6162         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Nd-145                   
  --------------------------------------------------------        
  * E1: ER = 14.95 (MeV) EG = 6.31 (MeV) SIG = 296.00 (mb)        
  * M1: ER =  7.80 (MeV) EG = 4.00 (MeV) SIG =   1.05 (mb)        
  * E2: ER = 11.99 (MeV) EG = 4.37 (MeV) SIG =   3.38 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Kikuchi, Y. et al.: JAERI-M 86-030 (1986).                    
 2) Tellier, H.: CEA-N-1459 (1971).                               
 3) Musgrove, A.R. de L., et al.: AEEC/E401 (1977).               
 4) Karzhavina, E.N., et al.: Sov. J. Nucl. Phys., 8, 371 (1969). 
 5) Fedorova, A.F., et al.: "Proc. 3rd All-union Conf. on Neutron 
    Physics, Kiev 1975", Vol. 1, 169.                             
 6) Allen, B.J. et al.: Nucl. Sci. Eng., 82, 230 (1982).          
 7) Mughabghab, S.F. et al.: "Neutron Cross Sections, Vol. I,     
    Part A", Academic Press (1981).                               
 8) Wisshak, K. et al.: Phys. Rev., C57, 3452 (1998).             
 9) Barry, D.P., et al.: Nucl. Sci. Eng., 153, 8 (2006).          
10) Mughabghab, S.F: "Atlas of Neutron Resonances", Elsevier      
    (2006).                                                       
11) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
12) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).   
13) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).           
14) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).     
15) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)  
    [Global potential].                                           
16) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).      
17) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.             
    J.H.Williams Lab., Univ. Minnesota (1969).                    
18) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966).    
19) Kalbach,C.: Phys. Rev. C33, 818 (1986).                       
20) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).   
21) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).   
22) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).                  
23) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151   
    (1994).                                                       
24) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).              
25) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).