95-Am-242

 95-Am-242G MINSK+    EVAL-FEB97 V.M.Maslov et al.                
INDC(BLR)-008/G       DIST-MAR02 REV2-APR00            20000420   
----JENDL-3.3         MATERIAL 9546                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
HISTORY                                                           
80-03 New evaluation was made by T.Nakagawa and S.Igarasi /Na80/  
88-03 Since no recent experimental data were available, the data  
      of JENDL-2 were adopted for JENDL-3.                        
00-04 JENDL-3.3 was compiled by T.Nakagawa                        
      Evaluated data of Maslov et al. /Ma97/ were extensively     
      adopted.                                                    
                                                                  
     ***** Modified parts from JENDL-3.2 *******************      
     All data                                                     
     *******************************************************      
                                                                  
                                                                  
================================================================= 
Description on modified parts from Maslov's evaluation.           
================================================================= 
                                                                  
MF=2 Resonance Rapameters                                         
  MT=151                                                          
    Unresolved resonance parameters                               
       The level spacing was modified to keep consistency with    
       that of L=0 and J=0.5. The neutron width was also modified 
       with the same factor of the level spacing. Interpolation   
       of cross sections was changed to 5 (log-log).              
                                                                  
    Calculated Thermal cross sections and Resonance integral      
                                                                  
                 at .0253 eV (b)   Res. Integ. (b)                
       Total        2319.8            -                           
       Elastic         7.66           -                           
       Fission      2093.3           997                          
       Capture       218.8           187                          
                                                                  
MF=3 Neutron Cross Sections                                       
  MT=19, 20, 21 Fission                                           
    Maslov's elavualtion was not adopted.                         
                                                                  
  MT=102 Capture                                                  
    Direct and semi-direct capture cross section was calculated   
    with DSD code /Ka99/, and added to Maslov's calculation.      
                                                                  
MF=5 Energy Distributions of Secondary Neutrons                   
  MT=16, 17, 91                                                   
    At the threshold energies, the same shape of distributions at 
    the second incident energy was assumed.  Interpolation was    
    replaced with 22.                                             
                                                                  
 Other parts are the same as Maslov's evaluation.                 
================================================================= 
                                                                  
References                                                        
Ka99) Kawano T.: private communication (1999).                    
Ma97) Maslov V.M. et al.: INDC(BLR)-8 (1997).                     
Na80) Nakagawa T. and Igarasi S.: JAERI-M 8903 (1980). in Japanese
                                                                  
========== Description given in Maslov's data =================== 
 95-Am-242G MINSK BYEL    EVAL-FEB97                              
                          DIST-MAR97                              
                                                                  
                      V.M. MASLOV, E.Sh. SUKHOVITSKIJ,            
                      Yu.V. PORODZINSKIJ, G.B. MOROGOVSKIJ        
                                                                  
 STATUS                                                           
 EVALUATION WAS MADE UNDER THE PROJECT AGREEMENT CIS-03-95        
 WITH INTERNATIONAL SCIENCE AND TECHNOLOGY CENTER (MOSCOW).       
 FINANCING PARTY OF THE ISTC FOR THE PROJECT IS JAPAN.            
 EVALUATION WAS REQUESTED BY Y.KIKUCHI (JAERI, TOKAI)             
 DOCUMENTED IN INDC(BLR)-008, 1997.                               
                                                                  
 MF=1   GENERAL  INFORMATION                                      
                                                                  
   MT=451  COMMENTS AND DICTIONARY                                
   MT=452  TOTAL NUMBER OF NEUTRONS PER FISSION                   
           SUM OF MT=455 AND MT=456.                              
   MT=455  DELAYED NEUTRON DATA                                   
           NUMBER OF DELAYED NEUTRONS AND                         
           DECAY CONSTANTS FROM BRADY ET AL./1/                   
   MT=456  NUMBER OF PROMPT NEUTRONS PER FISSION ADOPTED TO BE    
           IDENTICAL TO THAT FOR AM-242M, WHICH WAS OBTAINED      
           WITH MADLAND-NIX MODEL CALCULATIONS /2/ FITTED TO      
           THE MEASURED DATA OF HOWE ET AL./3/                    
           ABOVE  EMISSIVE FISSION THRESHOLD                      
           SUPERPOSITION OF NEUTRON EMISSION                      
           IN (N,XNF) REACTIONS /4/ AND PROMPT FISSION            
           NEUTRONS IS EMPLOYED.                                  
                                                                  
 MF=2   RESONANCE PARAMETERS                                      
                                                                  
   MT=151  RESONANCE  PARAMETERS  (MLBW)                          
           RESOLVED RESONANCE REGION :     1.0E-5 - 100 EV        
           GENERATED FROM AVERAGE RESONANCE RAMETERS  BASED ON    
           OPTICAL MODEL CALCULATIONS AND FISSION LEVEL DENSITY,  
           BARRIER PARAMETERS AND TRANSITION STATES STRUCTURE,    
           FIXED FOR  AM-242M.                                    
           CALCULATED 2200 M/S CROSS SECTIONS AND RESONANCE       
           INTEGRALS ARE:                                         
                            2200 M/SEC       RES.INTEG.           
              TOTAL         2319.73 b            -                
              ELASTIC          7.66 b            -                
              FISSION       2093.23 b          997.81             
              CAPTURE        218.83 b          187.01             
                                                                  
           UNRESOLVED RESONANCE REGION :                          
           ENERGY INDEPENDENT PARAMETERS:                         
              R=9.42  FM  FROM OPTICAL MODEL CALCULATIONS         
              S1=2.046*10-4  FROM OPTICAL MODEL CALCULATIONS      
              S2=1.924*10-4  FROM OPTICAL MODEL CALCULATIONS      
           ENERGY   DEPENDENT PARAMETERS:                         
           S0-DECREASES FROM 1.363-4(0.10KEV) TO 1.267-4(44.2838  
           keV)                                                   
           D - SPIN DEPENDENT, NORMALIZED TO  =0.877 EV     
           CORRESPONDENT LEVEL DENSITY PARAMETER WAS FITTED TO    
            =0.271 EV FOR AM-242M, THEN GROUND STATE        
           DIFFERENCES ARE TAKEN INTO ACCOUNT.                    
           WF -SPIN DEPENDENT AS DEFINED BY THE TRANSITION STATE  
           SPECTRA AT INNER AND OUTER BARRIER HUMPS, DETERMINED   
           FOR AM-242M                                            
           WG-WITH ACCOUNT OF FISSION COMPETITION =0.050 EV
                                                                  
 MF=3   NEUTRON CROSS SECTIONS                                    
                                                                  
   MT=1,4,51-75,91,102.  TOTAL, ELASTIC AND INELASTIC             
           SCATTERING, CAPTURE CROSS SECTION                      
           TOTAL,DIRECT ELASTIC AND DIRECT INELASTIC FOR MT=53,57,
           66 AND OPTICAL TRANSMISSION COEFFICIENTS FROM          
           COUPLED CHANNELS CALCULATIONS.                         
           THE DEFORMED OPTICAL POTENTIAL USED:                   
           VR=(46.10-0.3*E) MEV    RR=1.26 FM  AR=0.615 FM        
           WD=(3.53+0.4*E)  MEV  E <  10 MEV    RD=1.24 FM        
           WD= 7.50 MEV          E=>  10 MEV    AD=0.5 FM         
           VSO=6.2 MEV RSO=1.12 FM ASO=0.47 FM  B2=0.206 B4=0.092 
           FOUR LEVELS OF GROUND STATE ROTATIONAL BAND            
           ARE COUPLED.                                           
           CAPTURE,COMPOUND ELASTIC AND INELASTIC BY STATISTICAL  
           MODEL, SEE MT=18-21.                                   
           ABOVE NEUTRON ENERGY 5 MEV CAPTURE IS ASSUMED TO BE    
           CONSTANT.                                              
           ADOPTED LEVEL SCHEME OF AM-242 FROM NUCLEAR DATA       
           SHEETS /7/.                                            
                                                                  
             No       ENERGY(MEV)     SPIN-PARITY   K             
                                                                  
            g.s.       0.00000           1   -      0             
             1         0.04410           0   -      0             
             2         0.04863           5   -      5             
             3         0.05290           3   -      0             
             4         0.07580           2   -      0             
             5         0.11400           6   -      5             
             6         0.14000           6   -      6             
             7         0.14800           5   -      0             
             8         0.14990           4   -      0             
             9         0.19000           7   -      5             
            10         0.19760           3   -                    
            11         0.21700           7   -      6             
            12         0.22000           1   -      1             
            13         0.23050           2   +      1             
            14         0.24200           2   -      1             
            15         0.24410           3   -      3             
            16         0.26300           6   -      0             
            17         0.26310           7   -      0             
            18         0.27010           2   +                    
            19         0.27500           3   -      1             
            20         0.28330           3   +                    
            21         0.28840           4   -      3             
            22         0.29180           2   -      2             
            23         0.30500           8   -      6             
            24         0.30690           3   -                    
            25         0.31900           4   -      1             
                                                                  
                                                                  
          OVERLAPPING LEVELS ARE ASSUMED ABOVE 0.320 MEV          
          LEVEL DENSITY PARAMETERS: SEE MT 18-21                  
   MT=16,17.  (N,2N) AND (N,3N) CROSS SECTION                     
          FROM STATISTICAL MODEL CALCULATIONS /8/  WITH THE       
          ACCOUNT OF PRE-EQUILIBRIUM NEUTRON EMISSION:SEE MT=18-21
   MT=18,19,20,21.  FISSION CROSS SECTION IS CALCULATED WITHIN    
          STATISTICAL MODEL /9,10/. THE FISSION BARRIER PARAMETERS
          ARE FIXED BY FITTING MEASURED DATA FOR AM-242M BY       
          BROWNE  ET AL./5/ AND FURSOV ET AL./11/.                
          THE FIRST CHANCE FISSION MT=19 IS CALCULATED WITH       
          THE CONTRIBUTION OF EMISSIVE FISSION TO TOTAL FISSION   
          CROSS SECTION ACCORDING TO /9,12/.                      
                                                                  
 MF=4   ANGULAR DISTRIBUTIONS OF SECONDARY NEUTRONS               
                                                                  
        FOR MT=2, 53, 57, 66 FROM COUPLED CHANNELS CALCULATIONS   
        WITH ADDED ISOTROPIC STATISTICAL CONTRIBUTION.            
                                                                  
        MT=16,17,18-21,51-52,54-56,58-65,67-75,91 - ISOTROPIC     
                                                                  
 MF=5   ENERGY DISTRIBUTIONS OF SECONDARY NEUTRONS                
                                                                  
        ENERGY DISTRIBUTIONS FOR MT=16,17 WERE                    
        CALCULATED BY STATISTICAL MODEL OF CASCADE NEUTRON        
        EMISSION TAKING INTO ACCOUNT THE HISTORY OF THE DECAY     
        WITH THE ALLOWANCE OF PRE-EQUILLIBRIUM EMISSION OF        
        THE FIRST NEUTRON /4/                                     
        ENERGY DISTRIBUTIONS FOR MT=18,19,20,21 WERE ADOPTED TO   
        BE IDENTICAL TO THAT OF AM-242M, WHICH WERE CALCULATED    
        BY MADLAND-NIX MODEL /2/ WITH ACCOUNT OF COMPETITION      
        BETWEEN MULTIPLE-CHANCE FISSION PROCESSES UP THROUGH      
        THIRD-CHANCE FISSION WITH THE ALLOWANCE OF PRE-EQUILIBRIUM
        EMISSION OF THE FIRST NEUTRON /4,8/                       
                                                                  
REFERENCES                                                        
                                                                  
 1. Brady M.C., Wright R.Q., England T.R., Report                 
    ORNL/CSD/TM-226(1991), IAEA-NDS-102, 1992.                    
 2. Madland D.G., Nix J.R., Nucl. Sci. Engng. 81, 213 (1982).     
 3. Howe R.E., Browne J.C., Dougan R.J., Dupzyk R.J., Landrum J.H.
    Nucl. Sci. Eng.,77, 454 (1984).                               
 4. Maslov V.M., Porodzinskij Yu.V., Sukhovitskij E.Sh., Proc.    
    Int. Conf. on Neutron Physics, 14-18 Sept., Kiev, USSR,       
    V.1, p.413, 1988.                                             
 5. Browne J.C., White R.M., Howe R.E. et al., Phys. Rev. C,      
    29, 2188 (1984).                                              
 6. Porodzinskij Yu.V., Sukhovitskij E.Sh., Nuclear Constants,    
    4, p.27, 1987 (in Russian).                                   
 7. ENSDF, 1995.                                                  
 8. Ignatjuk A.V., Maslov V.M., Pashchenko A.B. Sov. J. Nucl.     
    Phys. 47, 224 (1988).                                         
 9. Ignatjuk A.V., Maslov V.M., Proc. Int. Symp. Nuclear Data     
    Evaluation Methodology, Brookhaven, USA, October 12-16, 1992, 
    p.440, World Scientific, 1993.                                
10. Maslov V.M. Sov. J. At. Energy 64, 478 (1988).                
11. Fursov B.I., Samylin B.F., Smirenkin G.N., Polynov V.N.,      
    Nuclear Data for Science and Technology, Proc. Int. Conf.,    
    Gatlinburg, Tennessee, USA, May 9-13, 1994, v.1,p.269.        
12. Maslov V.M., Kikuchi Y. JAERI-Research 96-030, 1996.          
 ========== Description given in Maslov's data (end) =============