74-W -184

 74-W -184 JAEA       EVAL-Feb10 N.Iwamoto                        
                      DIST-MAY10                       20100301   
----JENDL-4.0         MATERIAL 7437                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
10-02 The resolved resonance parameters were evaluated by         
      N.Iwamoto.                                                  
      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 parameters for MLBW formula were given in the energy 
      below 3.5 keV. Parameters were evaluated in examining both  
      the experimental data/1,2,3/ and the recommended            
      data of BNL/4/. For unknown radiative width, an average     
      value of 57 milli-eV was assumed. The scattering radius was 
      assumed to be 7.5 fm.                                       
      For JENDL-4.0 the upper limit of the resolved resonance     
      energy was changed due to significant level missing. The    
      negative resonance was placed so as to reproduce the cross  
      sections at thermal energy recommended by Mughabghab /5/.   
                                                                  
    Unresolved resonance region : 3.5 keV - 300 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           9.0699e+00                                 
       Elastic         7.3720e+00                                 
       n,gamma         1.6980e+00           1.6603e+01            
       n,alpha         3.8281e-08                                 
      ----------------------------------------------------------  
         (*) 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= 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= 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/.                               
                                                                  
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= 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= 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: 0,1,2,3,11 (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 W-184                                    
  -------------------                                             
  No.  Ex(MeV)  J  PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.11121   2  +  *                                           
   2  0.36406   4  +  *                                           
   3  0.74831   6  +  *                                           
   4  0.90328   2  +                                              
   5  1.00248   0  +                                              
   6  1.00597   3  +                                              
   7  1.12144   2  +                                              
   8  1.13003   2  -                                              
   9  1.13384   4  +                                              
  10  1.22129   3  -                                              
  11  1.25230   8  +  *                                           
  12  1.28360   2  -                                              
  13  1.28499   5  -                                              
  14  1.29492   5  +                                              
  15  1.32213   0  +                                              
  16  1.34538   4  -                                              
  17  1.36037   4  +                                              
  18  1.38631   2  +                                              
  19  1.42499   3  +                                              
  20  1.43100   2  +                                              
  21  1.44626   6  -                                              
  22  1.47700   6  +                                              
  23  1.49200   5  -                                              
  24  1.50154   7  -                                              
  25  1.52328   3  +                                              
  26  1.53686   4  +                                              
  27  1.57024   2  +                                              
  28  1.58145   6  -                                              
  29  1.61356   1  +                                              
  30  1.61487   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         
  --------------------------------------------------------        
    W-185 22.7200  0.8823  1.2247  0.4842 -0.8101  4.9225         
    W-184 22.1100  1.7693  1.2350  0.5106 -0.1439  6.1796         
    W-183 21.5000  0.8871  1.1150  0.5015 -0.7444  4.9247         
    W-182 21.6000  1.7790  1.2320  0.4968  0.1520  5.7824         
   Ta-184 21.4138  0.0000  1.3866  0.3893 -0.5346  2.1728         
   Ta-183 20.6300  0.8871  1.5183  0.4673 -0.2784  4.1815         
   Ta-182 20.4000  0.0000  1.1768  0.4917 -1.3065  3.5848         
   Ta-181 22.2900  0.8920  1.4278  0.4666 -0.5553  4.5420         
   Hf-183 22.0008  0.8871  1.6503  0.4912 -0.8609  4.9869         
   Hf-182 21.3733  1.7790  1.7280  0.5008  0.0219  5.9188         
   Hf-181 22.0100  0.8920  1.4311  0.4845 -0.7193  4.8126         
   Hf-180 20.9400  1.7889  1.5810  0.5569 -0.5865  6.8900         
   Hf-179 21.3300  0.8969  1.6163  0.4997 -0.8128  4.9790         
   Hf-178 21.4500  1.7989  1.8369  0.5406 -0.5584  6.7847         
  --------------------------------------------------------        
                                                                  
Table 3. Gamma-ray strength function for  W-185                   
  --------------------------------------------------------        
  K0 = 1.140   E0 = 4.500 (MeV)                                   
  * E1: ER = 12.59 (MeV) EG = 2.29 (MeV) SIG = 211.00 (mb)        
        ER = 14.88 (MeV) EG = 5.18 (MeV) SIG = 334.00 (mb)        
        ER =  5.30 (MeV) EG = 1.80 (MeV) SIG =   2.50 (mb)        
  * M1: ER =  7.20 (MeV) EG = 4.00 (MeV) SIG =   1.03 (mb)        
  * E2: ER = 11.06 (MeV) EG = 3.89 (MeV) SIG =   4.53 (mb)        
  --------------------------------------------------------        
                                                                  
References                                                        
 1) Camarda H.S. et al.: Phys. Rev. C8, 1813 (1973).              
 2) Ohkubo M.: JAERI-M 5624 (1974).                               
 3) Macklin R.L. et al.: LA-9200-MS (1982).                       
 4) Mughabghab S.F.:"Neutron Cross Sections", Vol. 1, part B      
    (1984).                                                       
 5) Mughabghab,S.F.: "Atlas of Neutron Resonances, Fifth          
     Edition: Resonance Parameters and Thermal Cross Sections.    
     Z=1-100", Elsevier Science (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).