95-Am-243

 95-Am-243 JAEA+      EVAL-JAN10 O.Iwamoto,T.Nakagawa,T.Ohsawa,+  
                      DIST-MAY10                       20100318   
----JENDL-4.0         MATERIAL 9549                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
07-05 Theoretical calculation was made with CCONE code.           
07-08 Theoretical calculation was made with CCONE code.           
08-03 Interpolation of (5,18) was changed.                        
      Data were compiled as JENDL/AC-2008/1/                      
09-03 (1,452) and (1,455) were revised.                           
09-08 (MF1,MT458) was evaluated.                                  
09-10 Fission cross section was revised.                          
10-01 Data of prompt gamma rays due to fission were given.        
10-03 Covariance data were given.                                 
                                                                  
                                                                  
MF= 1                                                             
  MT=452 Total neutron per fission                                
    Sum of MT=455 and 456.                                        
                                                                  
  MT=455 Delayed neutrons                                         
    Determined from nu-d of the following three nuclides and      
    partial fission cross sections calculated with CCONE code/2/. 
                                                                  
      Am-244 = 0.0085   Saleh et al./3/, Charlton et al./4/       
      Am-243 = 0.006659 *1)                                       
      Am-242 = 0.0049   Saleh et al./3/                           
                *1) an average of systematics by Tuttle/5/,       
                    Benedetti et al./6/ and Waldo et al./7/       
                                                                  
    Decay constants were taken from Saleh et al. and Brady and    
    England.                                                      
                                                                  
  MT=456 Prompt neutrons per fission                              
    The data measured by Khokhlov et al./8/ and Drapchinsky et    
    al./9/ were fitted by a linear function/10/.                  
                                                                  
  MT=458 Components of energy release due to fission              
    Total energy and prompt energy were calculated from mass      
    balance using JENDL-4 fission yields data and mass excess     
    evaluation. Mass excess values were from Audi's 2009          
    evaluation/11/. Delayed energy values were calculated from    
    the energy release for infinite irradiation using JENDL FP    
    Decay Data File 2000 and JENDL-4 yields data. For delayed     
    neutron energy, as the JENDL FP Decay Data File 2000/12/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/13/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/14/ when  
    they were provided. When the fractions were not given by Sher,
    averaged fractions were used.                                 
                                                                  
                                                                  
MF= 2 Resonance parameters                                        
  MT=151                                                          
  Resolved resonance parameters (below 250eV)                     
    JENDL-3.3 parameters of resonances below 1.744 eV were        
    modified:                                                     
      to reproduce an effective capture cross section measured    
      by Ohta et al./15/,                                         
      to delete background cross sections of fission given in     
      JENDL-3.3.                                                  
                                                                  
  Unresolved resonance parameters (250eV - 40keV)                 
    Parameters were determined with ASREP code/16/ so as to       
    reproduce the cross sections in the energy range from 250 eV  
    to 40 keV. They are used only for self-shielding calculations.
                                                                  
                                                                  
     Thermal cross sections and resonance integrals (at 300K)     
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.(*)                 
                     (barns)       (barns)                        
    -------------------------------------------------------       
    total           85.831                                        
    elastic          6.490                                        
    fission          0.0816            6.31                       
    capture         79.259          2040                          
   -------------------------------------------------------        
         (*) In the energy range from 0.5 eV to 10 MeV.           
                                                                  
                                                                  
MF= 3 Neutron cross sections                                      
  Cross sections above the resolved resonance region except for   
  the elastic scattering (MT=2) and fission cross sections (MT=18,
  19, 20, 21, 38) were calculated with CCONE code/2/.             
                                                                  
  MT= 1 Total cross section                                       
    The cross section was calculated with CC OMP of Soukhovitskii 
    et al./17/ The OMP was adjusted to the Am-241(n,tot) cross    
    section/18/.                                                  
                                                                  
  MT= 2 Elastic scattering cross section                          
    Calculated as total - non-elastic scattering cross sections.  
                                                                  
  MT=18 Fission cross section (Above 250eV)                       
    The following experimental data were analyzed with the GMA    
    code /19/:                                                    
      Wisshak+/20/, Fomushkin+/21/, Fursov+/22/, Kanda+/23/,      
      Knitter+/24/, Golovnya+/25/, Kobayashi+/26/, Laptev+/27/,   
      Aiche+/28/, Baba+/29/                                       
                                                                  
  MT=19, 20, 21, 38 Multi-chance fission cross sections           
    Calculated with CCONE code, and renormalized to the total     
    fission cross section (MT=18).                                
                                                                  
  MT=102 Capture cross section                                    
    The experimental data of Weston and Todd/30/ and Wisshak and  
    Kaeppeler/31/ were used to determine the parameters in the    
    CCONE calculation.                                            
                                                                  
                                                                  
MF= 4 Angular distributions of secondary neutrons                 
  MT=2 Elastic scattering                                         
    Calculated with CCONE code.                                   
                                                                  
  MT=18 Fission                                                   
    Isotropic distributions in the laboratory system were assumed.
                                                                  
                                                                  
MF= 5 Energy distributions of secondary neutrons                  
  MT=18 Prompt neutrons                                           
    Below 6 MeV, calculated by Ohsawa/32/ with modified           
    Madland-Nix formula considering multi-mode fission processes  
    (standard-1, standard-2, superlong).                          
    Above 7 MeV, calculated with CCONE code/2/.                   
                                                                  
  MT=455 Delayed neutrons                                         
    Taken from Brady and England /33/. Normalized yields of       
    6 groups are those measured by Saleh et al./3/                
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code.                                   
    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 given in MF=1/MT=458 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./34/ for        
    Pu-239 thermal fission were adopted.                          
                                                                  
                                                                  
MF=31 Covariances of average number of neutrons per fission       
  MT=452 Number of neutrons per fission                           
    Combination of covariances for MT=455 and MT=456.             
                                                                  
  MT=455                                                          
    Error was assumed as follows:                                 
          E < 5 MeV      5 % /3/                                  
       above 15 MeV     15 %                                      
                                                                  
  MT=456                                                          
    Covariances were obtained by fitting a linear function to the 
    experimental data of Khokhlov et al./8/ and Drapchinsky et    
    al./9/ Obtained standard deviation was multiplied by a        
    factor of 3 so that the minimum error was about 1%.           
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Format of LCOMP=0 was adopted.                                
                                                                  
    Standard deviations of resonance parameters were taken from   
    Mughabghab /35/ If no uncertainties were given by             
    Mughabghab, 0.1% and 10% were assumed for resonance energies  
    and other parameters, respectively.                           
                                                                  
    Additional errors of 5 % were given to the fission cross      
    section, and the following values for the capture cross       
    section:                                                      
         energy range    additional errors                        
         1.0e-5 - 13 eV     5%                                    
         13 - 50 eV        10%                                    
         50 - 250 eV       15%                                    
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/36/ and the covariances of model parameters         
  used in the theoretical calculations.                           
                                                                  
  For the following cross sections, covariances were determined   
  by different methods.                                           
                                                                  
  MT=1, 2 Total and elastic scattering cross sections             
    In the resolved resonance region, uncertainty of 10% was      
    added to the contributions from resonance parameter           
    uncertainties.                                                
                                                                  
    Above 250 eV, estimated with CCONE and KALMAN codes.          
                                                                  
  MT=18 Fission cross section                                     
    Above 250 eV, evaluated with GMA code/19/.                    
    Standard deviations were multiplied by a factor of 2.0.       
                                                                  
  MT=102 Capture cross section                                    
    Above 250 eV, covariance matrix was obtained with CCONE and   
    KALMAN codes/36/.                                             
                                                                  
                                                                  
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                                           
    Below 6 MeV, covarinaces of Pu239 fission spectra given in    
    JENDL-3.3 were adopted after multiplying a factor of 9.       
    Above 6 MeV, estimated with CCONE and KALMAN codes.           
                                                                  
                                                                  
***************************************************************** 
  Calculation with CCONE code                                     
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE/2/ 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/37/                              
    * Global parametrization of Koning-Duijvestijn/38/            
      was used.                                                   
    * Gamma emission channel/39/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/40/ 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/41/. 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/42/,/43/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,3,6,8 (see Table 2)                      
  * optical potential parameters /17/                             
    Volume:                                                       
      V_0       = 48       MeV                                    
      lambda_HF = 0.004    1/MeV                                  
      C_viso    = 15.9     MeV                                    
      A_v       = 12.04    MeV                                    
      B_v       = 81.36    MeV                                    
      E_a       = 385      MeV                                    
      r_v       = 1.255    fm                                     
      a_v       = 0.58     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       = 1.15     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    = 0.08                                            
      beta_6    = 0.0015                                          
                                                                  
  * Calculated strength function                                  
    S0= 0.91e-4 S1= 2.65e-4 R'=  9.45 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of Am-243                                   
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000  5/2 -  *                                           
   1  0.04220  7/2 -  *                                           
   2  0.08400  5/2 +                                              
   3  0.09640  9/2 -  *                                           
   4  0.10920  7/2 +                                              
   5  0.14350  9/2 +                                              
   6  0.16230 11/2 -  *                                           
   7  0.18930 11/2 +                                              
   8  0.23800 13/2 -  *                                           
   9  0.24400 13/2 +                                              
  10  0.26600  3/2 -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 3. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Am-244 18.7661  0.0000  1.9765  0.2797 -0.6834  1.0000         
   Am-243 18.6999  0.7698  2.0985  0.3873 -0.8965  3.0385         
   Am-242 18.6337  0.0000  1.6845  0.2795 -0.6541  0.9592         
   Am-241 18.1961  0.7730  1.7328  0.3819 -0.7226  2.8365         
   Am-240 18.5012  0.0000  1.3474  0.2883 -0.6831  1.0000         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
   Am-244   6.450   0.650   6.000   0.530                         
   Am-243   6.200   0.800   5.400   0.520                         
   Am-242   6.510   0.600   6.050   0.550                         
   Am-241   6.100   0.800   5.500   0.520                         
   Am-240   6.100   0.650   6.000   0.450                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Am-244 21.5810  0.0000  2.6000  0.3194 -2.3947  2.0000         
   Am-243 21.5049  0.8981  2.6000  0.3201 -1.4966  2.8981         
   Am-242 20.8697  0.0000  2.6000  0.3254 -2.4113  2.0000         
   Am-241 21.3526  0.9018  2.6000  0.3213 -1.4929  2.9018         
   Am-240 21.2764  0.0000  2.6000  0.3219 -2.3947  2.0000         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Am-244 21.5810  0.0000  0.6400  0.3664 -1.8698  2.2000         
   Am-243 21.5049  0.8981  0.6000  0.3532 -0.8019  2.8981         
   Am-242 21.4288  0.0000  0.5600  0.3688 -1.8681  2.2000         
   Am-241 21.3526  0.9018  0.5200  0.3556 -0.7969  2.9018         
   Am-240 21.2764  0.0000  0.4800  0.3567 -1.6981  2.0000         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for Am-244                   
  --------------------------------------------------------        
  * E1: ER = 11.51 (MeV) EG = 2.76 (MeV) SIG = 245.81 (mb)        
        ER = 14.29 (MeV) EG = 4.18 (MeV) SIG = 491.62 (mb)        
  * M1: ER =  6.56 (MeV) EG = 4.00 (MeV) SIG =   1.29 (mb)        
  * E2: ER = 10.08 (MeV) EG = 3.18 (MeV) SIG =   6.92 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
References                                                        
 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009).     
 2) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007).            
 3) H.H.Saleh et al.: Nucl. Sci. Eng., 125, 51 (1997).            
 4) W,Charlton et al.; 1997 Trieste, Part I, p.491 (1997).        
 5) R.J.Tuttle: INDC(NDS)-107/G+Special, p.29 (1979).             
 6) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982).          
 7) R.Waldo et al.: Phys. Rev., C23, 1113 (1981).                 
 8) Yu.A.Khokhlov et al.: 1994 Gatlinburg, Vol.1, p.272 (1994).   
 9) L.V.Drapchinsky et al.: ISTC 1828-01 (2004).                  
10) T.Nakagawa: J. Nucl. Sci. Eng., 42, 984 (2005).               
11) G.Audi: Private communication (April 2009).                   
12) J.Katakura et al.: JAERI 1343 (2001).                         
13) T.R.England et al.: LA-11151-MS (1988).                       
14) R.Sher, C.Beck: EPRI NP-1771 (1981).                          
15) M.Ohta et al.: J. Nucl. Sci. Technol., 43, 1441 (2006).       
16) Y.Kikuchi, et al.: JAERI-Data/Code 99-025 (1999) in Japanese. 
17) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).     
18) T.W.Phillips, R.E.Howe: Nucl. Sci. Eng., 69, 375 (1979).      
19) W.P.Poenitz: BNL-NCS-51363, Vol.I, p.249 (1981).              
    S.Chiba, D.L.Smith: ANL/NDM-121 (1991).                       
20) K.Wisshak, F.Kaeppeler: Nucl. Sci. Eng.,85, 251 (1983).       
21) Eh.Fomushkin, et al.: Yadernye Konstanty, 3/57, 17 (1984)     
  = INDC(CCP)-281/L, p.23 (1989).                                 
22) B.I.Fursov, et al.: Sov. At. Enery, 58, 899 (1986).           
23) K.Kanda, et al.: J. Nucl. Sci. Technol., 24, 423 (1987).      
24) H.-H.Knitter, C,Budtz-Jorgensen: Nucl. Sci. Eng., 99, 1       
   (1988).                                                        
25) V.Ya.Golovnya et al.: JINR-E3-419, p.293 (1999).              
26) K.Kobayashi, et al.: J. Nucl. Sci. Technol., 36, 20 (1999).   
27) A.B.Laptev et al.: 2007Sanibal Island, p.462 (2007).          
28) M.Aiche et al.: 2007 Nice (ND2007), p.483 (2007).             
29) M.Baba et al.: Private communication (2007).                  
30) L.W.Weston, J.H.Todd: Nucl. Sci. Eng., 91, 444 (1985).        
31) K.Wisshak, F.Kaeppeler: Nucl. Sci. Eng., 85, 251 (1983).      
32) T.Ohsawa: Private communication (2007).                       
33) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989).     
34) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).            
35) S.F.Mughabghab: "Atlas of Neutron Resonances," Elsevier       
    (2006).                                                       
36) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in         
    Japanese.                                                     
37) C.Kalbach: Phys. Rev. C33, 818 (1986).                        
38) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).     
39) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).     
40) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).                   
41) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).            
42) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990).                
43) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).