90-Th-228

 90-Th-228 JAEA+      EVAL-FEB10 O.Iwamoto, T.Nakagawa, et al.    
                      DIST-MAY10                       20100319   
----JENDL-4.0         MATERIAL 9028                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
07-06 Theoretical calculation was performed with CCONE code.      
07-07 Data were compiled as JENDL/AC-2008/1/.                     
10-02 Data of prompt gamma rays due to fission were given.        
      Nu-total, nu-p and nu-d were revised.                       
10-03 Covariance data were given.                                 
                                                                  
                                                                  
MF=1 General information                                          
  MT=452 Number of Neutrons per fission                           
    Sum of MT's=455 and 456.                                      
                                                                  
  MT=455 Delayed neutron data                                     
    Average values of systematics of Tuttle/2/, Benedetti et      
    al./3/ and Waldo et al./4/ were adopted.                      
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Based on the semi-empirical formula of Ohsawa/5/.             
                                                                  
                                                                  
MF= 2 Resonance parameters                                        
  MT=151                                                          
  Resolved resonance parameters (MLBW: 10e-5 - 13 eV)             
    Parameters of JENDL-3.3 which were based on the data obtained 
    by Simpson et al./6/ were adopted. A negative resonance       
    recommended by Mughabghab/7/ was added so as to reproduce     
    the thermal cross sections.                                   
    Fission widths were ignored. The fission cross section is     
    given as background cross sections.                           
                                                                  
  Unresolved resonance parameters (13 eV - 70 keV)                
    Parameters (URP) were determined with ASREP code /8/ so as to 
    reproduce the cross sections in this energy region. URP are   
    used only for self-shielding calculations.                    
                                                                  
     Thermal cross sections and resonance integrals (at 300K)     
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.(*)                 
                     (barns)       (barns)                        
    -------------------------------------------------------       
    total            154.35                                       
    elastic           31.30                                       
    fission            0.15            1.38                       
    capture          122.90         1120                          
    -------------------------------------------------------       
      (*) In the energy range from 0.5 eV to 10 MeV.              
                                                                  
                                                                  
MF= 3 Neutron cross sections                                      
  Cross sections except for the elastic scattering (MT=2) and     
  fission cross sections (MT=18,19) below 100 keV were calculated 
  with CCONE code/9/.                                             
                                                                  
  MT= 1 Total cross section                                       
    The cross section was calculated with CC OMP of Soukhovitskii 
    et al./10/.                                                   
                                                                  
  MT= 2 Elastic scattering cross section                          
    Calculated as total - non-elastic scattering cross sections.  
                                                                  
  MT=18,19  Fission cross section, (n,f) cross section            
    Below 4.5 keV, 1/v shape was assumed. The cross section at    
    0.0253 eV was assumed to be 0.15 b which was determined from  
    "< 0.3 b" recommended by Mughabghab/7/.                       
    From 4.5 to 100 keV, assumed to be 0.34 mb.                   
    From 100keV to 20 MeV, calculated with CCONE code/9/.         
                                                                  
    The fission cross section measured by James et al./11/ were   
    used to adjust parameters of CCONE code.                      
                                                                  
                                                                  
MF= 4 Angular distributions of secondary neutrons                 
  MT=2  Elastic scattering                                        
    Calculated with CCONE code/9/.                                
                                                                  
  MT=18 Fission                                                   
    Isotropic distributions in the laboratory system were assumed.
                                                                  
                                                                  
MF= 5 Energy distributions of secondary neutrons                  
  MT=18 Prompt neutrons                                           
    Calculated with CCONE code/9/.                                
                                                                  
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code/9/.                                
    Distributions from fission (MT=18) are not included.          
                                                                  
                                                                  
MF=12 Photon production multiplicities                            
  MT=18 Fission                                                   
    Calculated from the total energy released by the prompt       
    gamma-rays due to fission which was estimated from its        
    systematics, and the average energy of gamma-rays.            
                                                                  
                                                                  
MF=14 Photon angular distributions                                
  MT=18 Fission                                                   
    Isotoropic distributions were assumed.                        
                                                                  
                                                                  
MF=15 Continuous photon energy spectra                            
  MT=18 Fission                                                   
    Experimental data measured by Verbinski et al./12/ for        
    U-235 thermal fission were adopted.                           
                                                                  
                                                                  
MF=31 Covariances of average number of neutrons per fission       
  MT=452 Number of neutrons per fission                           
    Sum of covariances for MT=455 and MT=456.                     
                                                                  
  MT=455                                                          
    Error of 15% was assumed.                                     
                                                                  
  MT=456                                                          
    Covariance was obtained by fitting a linear function to the   
    at 0.0 and 5.0 MeV with an uncertainty of 5%.                 
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
  MT=151                                                          
    Format of LCOMP=0 was adopted.                                
    Uncertainties of parameters were taken from Mughabghab /7/    
    and Simpson et al./6/ For the parameters without any          
    information on uncertainty, unceratainties of 0.1% and 10%    
    were assumed to resonance energies and resonance widths,      
    respectively.                                                 
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/13/ and the covariances of model parameters         
  used in the cross-section calculations.                         
                                                                  
  In the resolved resonance region (up to 13 eV), the following   
  standard deviations were added to the contributions from        
  resonance parameters:                                           
                                                                  
     Total               12.6 %                                   
     Elastic scattering  50 %                                     
     Fission             90 %                                     
     Capture             10 %                                     
                                                                  
                                                                  
MF=34 Covariances for Angular Distributions                       
  MT=2 Elastic scattering                                         
    Covariances were given only to P1 components.                 
                                                                  
                                                                  
MF=35 Covariances for Energy Distributions                        
  MT=18 Fission spectra                                           
    Estimated with CCONE and KALMAN codes.                        
                                                                  
                                                                  
***************************************************************** 
  Calculation with CCONE code                                     
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE/9/ calculation          
  1) Coupled channel optical model                                
     Levels in the rotational band were included. Optical model   
     potential and coupled levels are shown in Table 1.           
                                                                  
  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                            
    * Moldauer width fluctuation correction/17/ was included.     
    * Neutron, gamma and fission decay channel were included.     
    * Transmission coefficients of neutrons were taken from       
      coupled channel calculation in Table 1.                     
    * The level scheme of the target is shown in Table 2.         
    * Level density formula of constant temperature and Fermi-gas 
      model were used with shell energy correction and collective 
      enhancement factor. Parameters are shown in Table 3.        
    * Fission channel:                                            
      Double humped fission barriers were assumed.                
      Fission barrier penetrabilities were calculated with        
      Hill-Wheler formula/18/. Fission barrier parameters were    
      shown in Table 4. Transition state model was used and       
      continuum levels are assumed above the saddles. The level   
      density parameters for inner and outer saddles are shown in 
      Tables 5 and 6, respectively.                               
    * Gamma-ray strength function of Kopecky et al/19/,/20/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,2,4,7 (see Table 2)                      
  * optical potential parameters /10/                             
    Volume:                                                       
      V_0       = 49.97    MeV                                    
      lambda_HF = 0.01004  1/MeV                                  
      C_viso    = 15.9     MeV                                    
      A_v       = 12.04    MeV                                    
      B_v       = 81.36    MeV                                    
      E_a       = 385      MeV                                    
      r_v       = 1.2568   fm                                     
      a_v       = 0.633    fm                                     
    Surface:                                                      
      W_0       = 17.2     MeV                                    
      B_s       = 11.19    MeV                                    
      C_s       = 0.01361  1/MeV                                  
      C_wiso    = 23.5     MeV                                    
      r_s       = 1.1803   fm                                     
      a_s       = 0.601    fm                                     
    Spin-orbit:                                                   
      V_so      = 5.75     MeV                                    
      lambda_so = 0.005    1/MeV                                  
      W_so      = -3.1     MeV                                    
      B_so      = 160      MeV                                    
      r_so      = 1.1214   fm                                     
      a_so      = 0.59     fm                                     
    Coulomb:                                                      
      C_coul    = 1.3                                             
      r_c       = 1.2452   fm                                     
      a_c       = 0.545    fm                                     
    Deformation:                                                  
      beta_2    = 0.213                                           
      beta_4    = 0.066                                           
      beta_6    = 0.0015                                          
                                                                  
  * Calculated strength function                                  
    S0= 1.07e-4 S1= 2.55e-4 R'=  9.88 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of Th-228                                   
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.05776   2  +  *                                           
   2  0.18682   4  +  *                                           
   3  0.32800   1  -                                              
   4  0.37818   6  +  *                                           
   5  0.39608   3  -                                              
   6  0.51919   5  -                                              
   7  0.62250   8  +  *                                           
   8  0.69560   7  -                                              
   9  0.83182   0  +                                              
  10  0.87447   2  +                                              
  11  0.91180  10  +                                              
  12  0.92080   9  -                                              
  13  0.93858   0  +                                              
  14  0.94420   1  +                                              
  15  0.96837   3  -                                              
  16  0.96897   2  +                                              
  17  0.97950   2  +                                              
  18  1.01641   2  +                                              
  19  1.02253   3  +                                              
  20  1.05993   4  -                                              
  21  1.09102   4  +                                              
  22  1.12295   2  -                                              
  23  1.15347   2  +                                              
  24  1.16000  10  -                                              
  25  1.16837   3  -                                              
  26  1.17451   5  +                                              
  27  1.17539   2  +                                              
  28  1.18980  11  -                                              
  29  1.20054   3  +                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 3. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Th-229 17.7702  0.7930  3.2566  0.4327 -1.4313  3.7239         
   Th-228 17.7035  1.5894  3.0590  0.3964 -0.1539  3.9563         
   Th-227 17.6369  0.7965  3.1200  0.4219 -1.2457  3.5201         
   Th-226 17.5702  1.5965  2.8637  0.4068 -0.2170  4.0544         
   Th-225 17.5034  0.8000  2.7304  0.3997 -0.9049  3.1303         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
   Th-229   5.500   0.800   6.000   0.520                         
   Th-228   3.900   1.040   6.400   0.600                         
   Th-227   4.100   0.800   6.400   0.520                         
   Th-226   3.900   1.040   8.200   0.600                         
   Th-225   4.200   0.800   8.000   0.520                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Th-229 19.9026  0.9251  2.6000  0.3339 -1.4862  2.9251         
   Th-228 19.8280  1.8543  2.6000  0.3346 -0.5570  3.8543         
   Th-227 19.7533  0.9292  2.6000  0.3352 -1.4822  2.9292         
   Th-226 19.6786  1.8625  2.6000  0.3359 -0.5488  3.8625         
   Th-225 19.6038  0.9333  2.6000  0.3366 -1.4780  2.9333         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Th-229 19.9026  0.9251 -0.2600  0.3799 -0.7708  2.9251         
   Th-228 19.8280  1.8543 -0.3000  0.3812  0.1590  3.8543         
   Th-227 19.7533  0.9292 -0.3400  0.3826 -0.7655  2.9292         
   Th-226 19.6786  1.8625 -0.3800  0.3840  0.1685  3.8625         
   Th-225 19.6038  0.9333 -0.4200  0.3854 -0.7601  2.9333         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for Th-229                   
  --------------------------------------------------------        
  K0 = 1.502   E0 = 4.500 (MeV)                                   
  * E1: ER = 11.03 (MeV) EG = 2.71 (MeV) SIG = 302.00 (mb)        
        ER = 13.87 (MeV) EG = 4.77 (MeV) SIG = 449.00 (mb)        
  * M1: ER =  6.70 (MeV) EG = 4.00 (MeV) SIG =   2.81 (mb)        
  * E2: ER = 10.30 (MeV) EG = 3.36 (MeV) SIG =   6.26 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
References                                                        
 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009).     
 2) R.J.Tutle: INDC(NDS)-107/G+SPECIAL, P.29 (1979),              
 3) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982).          
 4) R.Waldo et al.: Phys. Rev., C23, 1113 (1981).                 
 5) T.Ohsawa: J. Nucl. Radiochem. Sci., 9, 19 (2008).             
 6) O.D.Simpson et al.: Nucl. Sci. Eng., 29, 423 (1967).          
 7) S.F.Mughabghab: "Atlas of Neutron Resonances," Elsevier       
   (2006).                                                        
 8) Y.Kikuchi, et al.: JAERI-Data/Code 99-025 (1999) in Japanese. 
 9) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007).            
10) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).     
11) G.D.James et al.:Nucl. Phys. A419, 497 (1984)                 
12) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).            
13) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in         
    Japanese.                                                     
14) C.Kalbach: Phys. Rev. C33, 818 (1986).                        
15) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).     
16) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).     
17) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).                   
18) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).            
19) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990).                
20) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).