63-Eu-155

 63-Eu-155 JAEA       EVAL-Nov09 N.Iwamoto                        
                      DIST-MAY10                       20100119   
----JENDL-4.0         MATERIAL 6337                               
-----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 29.7 EV      
      RESONANCE PARAMETERS WERE BASED ON JENDL-2 EVALUATION BY    
      KUKICHI ET AL./1/ WHICH WERE MADE ON THE BASIS OF THE DATA  
      MEASURED BY ANUFRIEV ET AL./2/  A NEGATIVE RESONANCE WAS    
      ADDED SO AS TO REPRODUCE THE THERMAL CAPTURE CROSS SECTION  
      GIVEN BY MUGHABGHAB/3/.                                     
      FOR JENDL-3, TOTAL SPIN J WAS TENTATIVELY ESTIMATED WITH A  
      RANDOM NUMBER METHOD. PARAMETERS OF THE NEGATIVE LEVEL WERE 
      ADJUSTED TO THE THERMAL CAPTURE CROSS SECTION AND RESONANCE 
      INTEGRAL MEASURED BY SEKINE ET AL./4/.                      
                                                                  
    Unresolved resonance region : 29.7 eV - 100 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           3.7671e+03                                 
       Elastic         6.6014e+00                                 
       n,gamma         3.7605e+03           1.5509e+04            
       n,alpha         1.6577e-10                                 
      ----------------------------------------------------------  
         (*) 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= 33 (n,nt) cross section                                     
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 41 (n,2np) 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= 33 (n,nt) reaction                                          
    Calculated with CCONE code /7/.                               
                                                                  
  MT= 41 (n,2np) 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: 0,1,4,7,11 (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 Eu-155                                   
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000  5/2 +  *                                           
   1  0.07864  7/2 +  *                                           
   2  0.10433  5/2 -                                              
   3  0.16901  7/2 -                                              
   4  0.17916  9/2 +  *                                           
   5  0.24578  3/2 +                                              
   6  0.25466  9/2 -                                              
   7  0.30069 11/2 +  *                                           
   8  0.30738  5/2 +                                              
   9  0.35717 11/2 -                                              
  10  0.39148  7/2 +                                              
  11  0.44303 13/2 +  *                                           
  12  0.48709 13/2 -                                              
  13  0.50101  9/2 +                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 2. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Eu-156 18.0000  0.0000  2.8275  0.5361 -1.7176  4.0906         
   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         
   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         
   Pm-150 17.9970  0.0000  4.0234  0.4210 -0.7878  2.5000         
   Pm-149 17.2625  0.9831  3.6138  0.5926 -1.4731  6.0264         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for Eu-156                   
  --------------------------------------------------------        
  K0 = 2.300   E0 = 4.500 (MeV)                                   
  * E1: ER = 12.43 (MeV) EG = 3.20 (MeV) SIG = 130.61 (mb)        
        ER = 16.12 (MeV) EG = 5.26 (MeV) SIG = 261.22 (mb)        
  * M1: ER =  7.62 (MeV) EG = 4.00 (MeV) SIG =   1.45 (mb)        
  * E2: ER = 11.70 (MeV) EG = 4.24 (MeV) SIG =   3.57 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) KIKUCHI, Y. ET AL.: JAERI-M 86-030 (1986).                    
 2) ANUFRIEV, V.A., ET AL.: SOV. AT. ENERGY, 46, 182 (1979).      
 3) MUGHABGHAB, S.F.: "NEUTRON CROSS SECTIONS, VOL. I, PART B",   
    ACADEMIC PRESS (1984).                                        
 4) SEKINE, T. ET AL.: APPL. RADIAT. ISOT., 38, 513 (1987).       
 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).