62-Sm-154

 62-Sm-154 JAEA       EVAL-Nov09 N.Iwamoto                        
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6255                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
09-11 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 3.0 KEV      
      RESONANCE PARAMETERS WERE TAKEN FROM JENDL-2 EVALUATED BY   
      KIKUCHI ET AL./1/ AND WERE MODIFIED FOR JENDL-3.            
      FOR JENDL-2, PARAMETERS WERE ADOPTED FROM RAHN ET AL./2/    
      FOR THE LEVELS WHOSE RADIATION WIDTH WAS NOT MEASURED, THE  
      AVERAGE VALUE OF 0.079+-0.013 EV WAS ASSUMED.  A NEGATIVE   
      RESONANCE WAS ADDED AT -35 EV SO AS TO REPRODUCE THE CAPTURE
      CROSS SECTION OF 5.5+-1.1 BARNS AT 0.0253 EV/3/.            
      FOR JENDL-3, THE RADIATION WIDTH OF THE NEGATIVE RESONANCE  
      WAS CHANGED FROM 0.079 EV TO 0.1266 EV AND THE SCATTERING   
      RADIUS FROM 8.34 FM TO 9.67 FM SO AS TO REPRODUCE WELL THE  
      THERMAL CROSS SECTIONS (CAPTURE = 8.4 B, SCATTERING = 11 B) 
      COMPILED BY MUGHABGHAB/4/.                                  
                                                                  
    Unresolved resonance region : 3.0 keV - 250.0 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   
      CCOM /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           1.9427e+01                                 
       Elastic         1.1032e+01                                 
       n,gamma         8.3951e+00           3.6490e+01            
       n,alpha         6.8088e-16                                 
      ----------------------------------------------------------  
         (*) 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= 28 (n,np) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 32 (n,nd) 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= 28 (n,np) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 32 (n,nd) 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: 0,1,2,3,4 (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 Sm-154                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.08198   2  +  *                                           
   2  0.26679   4  +  *                                           
   3  0.54373   6  +  *                                           
   4  0.90264   8  +  *                                           
   5  0.92140   1  -                                              
   6  1.01239   3  -                                              
   7  1.09933   0  +                                              
   8  1.10400   4  -                                              
   9  1.12000   4  -                                              
  10  1.17781   2  +                                              
  11  1.18065   5  -                                              
  12  1.20238   0  +                                              
  13  1.28636   2  +                                              
  14  1.29500   2  +                                              
  15  1.33280  10  +                                              
  16  1.33764   4  +                                              
  17  1.36500   3  -                                              
  18  1.37100   4  +                                              
  19  1.43100   7  -                                              
  20  1.44005   2  +                                              
  21  1.47213   4  +                                              
  22  1.47500   6  +                                              
  23  1.47571   1  -                                              
  24  1.51519   2  -                                              
  25  1.53926   3  +                                              
  26  1.57661   6  +                                              
  27  1.58455   3  -                                              
  28  1.61482   0  +                                              
  29  1.66190   4  -                                              
  30  1.66489   4  +                                              
  31  1.67383   0  +                                              
  32  1.70681   4  +                                              
  33  1.74100   8  +                                              
  34  1.75464   0  +                                              
  35  1.75589   0  +                                              
  36  1.76000   9  -                                              
  37  1.76440   0  +                                              
  38  1.77424   5  -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Sm-155 19.5000  0.9639  2.9414  0.5495 -1.3709  5.8007         
   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         
   Pm-154 18.4033  0.0000  2.5027  0.3188  0.0149  1.0000         
   Pm-153 17.6600  0.9701  3.1546  0.5829 -1.3375  5.8693         
   Pm-152 18.2003  0.0000  3.4439  0.4590 -1.0726  3.0071         
   Pm-151 17.4614  0.9765  3.7662  0.5765 -1.3653  5.8316         
   Nd-153 19.0261  0.9701  2.6911  0.3086  0.9675  1.9701         
   Nd-152 18.3157  1.9467  3.0281  0.5483 -0.0707  6.4042         
   Nd-151 19.8000  0.9765  3.4048  0.5128 -1.0731  5.3158         
   Nd-150 20.0000  1.9596  3.4363  0.5263 -0.3405  6.6204         
   Nd-149 20.9000  0.9831  3.5199  0.4992 -1.1865  5.3955         
   Nd-148 21.1000  1.9728  2.8636  0.4784  0.2048  5.9010         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Sm-155                   
  --------------------------------------------------------        
  K0 = 1.660   E0 = 4.500 (MeV)                                   
  * E1: ER = 12.45 (MeV) EG = 3.21 (MeV) SIG = 129.00 (mb)        
        ER = 16.14 (MeV) EG = 5.27 (MeV) SIG = 257.99 (mb)        
  * M1: ER =  7.63 (MeV) EG = 4.00 (MeV) SIG =   1.10 (mb)        
  * E2: ER = 11.73 (MeV) EG = 4.25 (MeV) SIG =   3.47 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) KIKUCHI,Y. ET AL.: JAERI-M 86-030 (1986).                     
 2) RAHN,F. ET AL.: PHYS. REV., C6, 251 (1972).                   
 3) MUGHABGHAB,S.F. AND GARBER,D.I.: "NEUTRON CROSS SECTIONS,     
    VOL.1, RESONANCE PARAMETERS", BNL 325, 3RD ED., VOL. 1,       
    (1973).                                                       
 4) MUGHABGHAB,S.F.: "NEUTRON CROSS SECTIONS, VOL. I, PART B",    
    ACADEMIC PRESS (1984).                                        
 5) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)              
    [in Japanese].                                                
 6) Iwamoto,O.: JAERI-Data/Code 2003-020 (2003).                  
 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).