95-Am-241

 95-AM-241 JAERI+     EVAL-Mar00 T.Nakagawa, O.Iwamoto, et al.    
JAERI-R 2001-059      DIST-Dec03 REV3-Dec03            20031217   
----JENDL-3.3u        MATERIAL 9543                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
82-03 EVALUATION FOR JENDL-2 WAS MADE BY Y.KIKUCHI (JAERI) /1/.   
88-03 RE-EVALUATION FOR JENDL-3 WAS MADE BY T.NAKAGAWA (JAERI)    
      /2/                                                         
01-03 JENDL-3.3.                                                  
      New evaluation made by Maslov et al. /3/ was adopted and    
      modified by by T.Nakagawa (NDC/JAERI)                       
                                                                  
     *****   Modified parts from JENDL-3.2   ******************** 
      All data                                                    
     ***********************************************************  
                                                                  
03-12 (MF4,MT18) added. (MF5,MT18) modified.                      
                                                                  
                                                                  
MF=1  General Information                                         
  MT=451  Descriptive data and directory records                  
                                                                  
  MT=452  Number of neutrons per fission                          
     Sum of MT's= 455 and 456                                     
                                                                  
  MT=455  Delayed neutron data                                    
     Nu-d was based on the experimental data of Saleh et al./4/,  
     and the semi-empirical formula of Tuttle /5/ above 8 MeV .   
     Decay constants were adopted from Saleh et al./4/ and Brady  
     and England /6/.                                             
                                                                  
  MT=456  Number of prompt neutrons per fission                   
     Taken from measurements Khokhlov et al./7/.                  
                                                                  
MF=2 Resonance Parameters                                         
  MT=151 Resolved and unresolved resonance parameters             
  1) Resolved resonance parameters for MLBW formula (below 150 eV)
      The resonance parameters of Maslov et al. which were mainly 
      based on the data of Derrien and Lucas /8/, were slightly   
      modified to reproduce well the Yamamoto et al./9/           
      The parameters of low energy levels were adjusted to the    
      thermal cross sections and resonance integrals.             
                                                                  
  2) Unresolved resonance parameters (150 eV - 40 keV)            
      Average fission cross section to be reproduced was          
      determined from experimental data of Yamamoto et al./9/     
      and Dabbs et al./10/, and the capture cross section from    
      Vanpraet et al./11/ and Gayther et al./12/                  
                                                                  
      The average resonance parameters were determined with ASREP 
      /13/ to reproduce those average cross sections.             
                                                                  
                                                                  
         Thermal cross sections and resonance integrals           
                                                                  
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.                    
    -------------------------------------------------------       
    total          654.44                                         
    elastic         11.83                                         
    fission          3.14             14.8                        
    capture        639.47           1460                          
    -------------------------------------------------------       
                                                                  
                                                                  
MF=3 Neutron Cross Sections                                       
  Except for the following reactions, the evaluated data of Maslov
  et al. /3/ were adopted.                                        
                                                                  
  MT= 1 Total cross section                                       
    Data of JENDL-3.2 were adopted. They were calculated with     
    spherical optical model parameters determined to reproduce    
    the total cross section measured by Phillips and Howe /14/    
                                                                  
           V = 43.4 - 0.107*EN                      (MeV)         
           Ws= 6.95 - 0.339*EN + 0.0531*EN**2       (MeV)         
           Wv= 0             , Vso = 7.0            (MeV)         
           r = rso = 1.282   , rs  = 1.29            (fm)         
           a = aso = 0.60    , b   = 0.5             (fm)         
                                                                  
  MT= 2 Elastic scattering cross section                          
    Calculated as (total - sum of partiacl cross sections)        
                                                                  
  MT=18 Fission cross section                                     
    Based on the experimental data of Hirakawa /15/,              
    Prindre et al. /16/, Aleksandrov et al. /17,18/, Cance        
    et al. /19/, Dabbs et al. /10/, Vorotnikov et al./20/,        
    Wisshak anf Kaeppeler /21/.                                   
                                                                  
  MT=102 Capture cross section                                    
    Based on the evaluated data of Maslov et al. In the MeV       
    region, the cross section of direct and semi-direct process   
    was calculated with DSD code /22/.                            
                                                                  
MF=4 Angular Distributions of Secondary Neutrons                  
  All data were taken from the evaluation by Maslov et al. /3/    
                                                                  
MF=5 Energy Distributions of Secondary Neutrons                   
  All data were taken from the evaluation by Maslov et al. /3/    
                                                                  
MF=8 Radiactive Decay Data                                        
  MT=102                                                          
    Decay data were taken from ENSDF.                             
                                                                  
MF=9 Multiplicities for Production of Radioactive Elements        
  MT=102                                                          
    En = 1e-5 - 0.9 eV:                                           
      Based on the experimental data of Shinohara et al. /23/     
      and Wisshak et al./24/                                      
    En > 0.9 eV:                                                  
      Calculation based on Hauser-Feshbach statistical model and  
      normalized to the experimental data /24/ at 29 keV within   
      the experimental error. For the calculation, optical        
      potential was taken from Maslov et al./3/, level scheme     
      from Wisshak et al./24/                                     
                                                                  
References                                                        
 1) Kikuchi Y.: JAERI-M 82-096 (1982).                            
 2) Nakagawa T.: JAERI-M 88-008 (1989).                           
 3) Malov V.M. et al.: INDC(BLR)-5 (1996).                        
 4) Saleh H.H. et al.: Nucl. Sci. Eng., 125, 51 (1997).           
 5) Tuttle R.J.: INDC(NDS)-107/G+Special, p.29 (1979).            
 6) Brady M.C. and England T.R.: Nucl. Sci. Eng., 103, 129 (1989).
 7) Khokhlov Yu.A. et al.: 1994 Gatlinburg, Vol.1, p.272 (1994).  
 8) Derrien H. and Lucas B.: 1975 Washington, Vol.II, p.637       
    (1975).                                                       
 9) Yamamoto S. et al.: Nucl. Sci. Eng., 126, 201 (1997).         
10) Dabbs J.W.T. et al.: Nucl. Sci. Eng., 83, 22 (1983).          
11) Vanpraet G. et al.: 1985 Santa Fe, Vol.1, p.493 (1985).       
12) Gayther D.B. and Thomas B.W.: 1977 Kiev, Vol. 3, p.3 (1977).  
13) Kikuchi Y. et al.: JAERI-Data/Code 99-025 (1999).             
14) Phillips T.W. and Howe R.E.: Nucl. Sci. Eng., 69, 375(1979).  
15) Hirakawa N.: JNC TJ9400 99-007 (1999). [in Japanese]          
16) Prindre A.L., et al.: Phys. Rev., C20, 1824 (1979)            
17) Aleksandrov B.M. et al.: Sov. At. Energy, 46, 475 (1979).     
18) Aleksandrov B.M. et al.: Yadernye Konstanty, 1/50, 3 (1983).  
19) Cance M., et al.: CEA-N-2194 (1981).                          
20) Vorotnikov P.E. et al.: Sov. J. Nucl. Phys., 44, 912 (1986).  
21) Wisshak K. and Kaeppeler F.: Nucl. Sci. Eng., 76, 148 (1980). 
22) Kawano T.: private communication (1999).                      
23) Shinohara N. et al.: J. Nucl. Sci. Technol., 34, 613 (1997).  
24) Wisshak K. et al.: Nucl. Sci. Eng., 81, 396 (1982).