93-Np-237

 93-Np-237 JAEA+      EVAL-JAN10 O.Iwamoto,T.Nakagawa,S.Chiba,+   
                      DIST-MAY10                       20100524   
----JENDL-4.0         MATERIAL 9346                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
06-02 Fission cross section was evaluated with GMA code.          
07-05 New calculation was made with CCONE code.                   
07-07 Re-calculation was made with CCONE code.                    
07-08 Fission cross section was revised.                          
07-11 Isomeric ratio of the (n,2n) reaction was given.            
07-12 Resonance parameters                                        
08-01 Fission cross section was revised.                          
08-02 Fission cross section and nu-p were revised.                
      CCONE calculation was made with revised parameters.         
08-03 Data were compiled as JENDL/AC-2008/1/.                     
09-03 Resonance parameters and fission cross section were         
      modified.                                                   
09-08 (MF1,MT458) was evaluated.                                  
10-01 Data of prompt gamma rays due to fission were given.        
10-02 Covariance data were given.                                 
                                                                  
                                                                  
MF= 1                                                             
  MT=452 Total neutron per fission                                
    Sum of MT=455 and 456.                                        
                                                                  
  MT=455 Delayed neutrons                                         
    (same as JENDL-3.3)                                           
    Nu-d was based on the experimental data of Saleh et al./2/,   
    Charlton et al./3/, Piksaikin et al./4/ and Zeinalov et       
    al./5/ Decay constants were detemined from the experimental   
    data of Piksaikin et al./4/.                                  
                                                                  
  MT=456 Prompt neutrons per fission                              
    The following data were fitted by GMA code /6/:               
    Nu-p measured by Veeser/7/, Frehaut et al./8/, Malinovskii    
    et al./9/, and nu-total by Boikov et al./10/, Thierens et     
    al./11/, Mueller et al./12/ and Khokhlov et al./13/           
                                                                  
  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/14/. 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 2000/15/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/16/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/17/ 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 500eV)                     
    The parameters given in JENDL-3.3 were modified:              
    Parameters of the resonances below 3.9 eV were modified so    
    as to reproduce the experimental data of Esch et al./18/ and  
    Tovesson and Hill/19/, and the thermal capture cross section  
    of 178.3 b and fission of 0.0202 b.                           
                                                                  
    Background cross section was given to the capture to get      
    agreement with the average cross section of Esch et al./18/   
                                                                  
    Thermal cross sections at 0.0253 eV were based on:            
      fission:                                                    
         Wagemans et al./20,21/, Kozharin et al./22/              
      capture:                                                    
         Kobayashi et al./23/, Katoh et al./24/, Harada et        
         al./25/, Bringer et al./26/, Esch et al./18/, and        
         others.                                                  
                                                                  
  Unresolved resonance parameters (500eV - 30keV)                 
    Parameters were determined with ASREP code/27/ so as to       
    reproduce the cross sections in the energy range from 500 eV  
    to 30 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           192.57                                        
    elastic          14.47                                        
    fission           0.0202           5.38                       
    capture         178.08            696                         
   -------------------------------------------------------        
         (*) 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/28/.            
  The model parameters were determined by considering integral    
  experimental data as well as measured cross-section data.       
                                                                  
  MT= 1 Total cross section                                       
    The cross section was calculated with CC OMP of Soukhovitskii 
    et al./29/                                                    
                                                                  
  MT= 2 Elastic scattering cross section                          
    Calculated as total - non-elastic scattering cross sections.  
                                                                  
  MT=16 (n,2n) cross section                                      
    Calculated with CCONE code. The experimental data of Gromova  
    et al./30/, Nishi et al./31/ and Landrum et al./32/ were      
    used to determine the model parameters for the CCONE          
    calculation.                                                  
                                                                  
  MT=18 Fission cross section                                     
    The following experimental data reported after 1982 were      
    analyzed with the GMA code /6/:                               
      Behrens+/33/, Cance+/34/, Alkhazov+/35/, Meadows/36/,       
      Wu+/37/, Garlea+/38/, Goverdovskij+/39/, Goverdovskij+/40/, 
      Zasadny+/41/, Goverdovskij+/42/, Kanda+/43/, Kovalenko+/44/,
      Alkhazov+/45/, Gul+/46/, Terayama+/47/, Meadows /48/,       
      Desdin+/49/, Merla+/50/, Garlea+/51/, Shcherbakov+/52/,     
      Furman+/53/, Baba+/54/, Tovesson+/55/.                      
                                                                  
    The data measured relatively to U235 fission were transformed 
    to cross sections by using the U235 fission cross section of  
    JENDL/AC-2008.                                                
                                                                  
    The results of GMA were used to determine the parameters in   
    the CCONE calculation.                                        
                                                                  
    Further modification was made for JENDL-4.0 in the energy     
    range from 500eV to 200keV, by eye-guiding the experimental   
    data of Tovesson and Hill /19/.                               
                                                                  
                                                                  
  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                                    
    Calculted with CCONE code. The experimental data of Esch et   
    al./56/, Kobayashi et al./57/, Weston and Todd/58/, Linder    
    et al./59/, and Buleeva et al./60/ 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                                           
    Calculated with CCONE code.                                   
                                                                  
  MT=455 Delayed neutrons                                         
    Calculated by Brady and England/61/.                          
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code.                                   
    Distributions from fission (MT=18) are not included.          
                                                                  
                                                                  
MF= 8 Radioactive Decay                                           
  MT=16                                                           
    Decay data of Np-236: taken from ENDF (as of 2007) /62/       
                                                                  
                                                                  
MF= 9 Multiplicities for Production of Radioactive Nuclides       
  MT=16                                                           
    Meta-stable state (T-1/2 =22.5H) production was assumed to    
    be 70% from the cross sections measured around 14 MeV.        
                                                                  
                                                                  
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./63/ 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 of 4% and 10 % was assumed below 2 MeV and above 6 MeV, 
    respectively. Between 2 and 6 MeV, 7% was assumed./64/        
                                                                  
  MT=456                                                          
    Covariance was obtained by GMA fitting to the experimental    
    data (see MF1,MT456). Obtained standard deviation was         
    multiplied by a factor of 3, because it was too small.        
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Format of LCOMP=0 was adopted.                                
                                                                  
    Standard diviations of resonance parameters were taken from   
    JENDL-3.3 covariance file /64/, which were estimated from     
    errors reported in Refs./65,66,67,68/.                        
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/69/ 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 resonance region (below 500 eV), standard deviation    
    (SD) of 8 % was added to the contributions from resonance     
    parameters.                                                   
                                                                  
    Above 500 eV, covariances were obtaibed with CCONE and        
    KALMAN codes, and experimental data.                          
                                                                  
  MT=18 Fission cross section                                     
    SD of 4% was added in the energy region up to 1 eV, and 15%   
    from 1 eV to 500 eV.                                          
                                                                  
    Above 500 eV, covariances were obtained with GMA code/6/.     
    SD was multiplied by a factor of 2.0.                         
                                                                  
  MT=102 Capture cross section                                    
    Additional SD of 4% was given from 0.1 eV 500 eV.             
                                                                  
                                                                  
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/28/ 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/70/                              
    * Global parametrization of Koning-Duijvestijn/71/            
      was used.                                                   
    * Gamma emission channel/72/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/73/ 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/74/. 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/75/,/76/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,3,5,7 (see Table 2)                      
  * optical potential parameters /29/                             
    Volume:                                                       
      V_0       = 49.8581  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       = 1.25     fm                                     
      a_v       = 0.57     fm                                     
    Surface:                                                      
      W_0       = 17.1839  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    = 0.212764                                        
      beta_4    = 0.066                                           
      beta_6    = 0.0015                                          
                                                                  
  * Calculated strength function                                  
    S0= 0.99e-4 S1= 2.49e-4 R'=  9.76 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of Np-237                                   
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000  5/2 +  *                                           
   1  0.03320  7/2 +  *                                           
   2  0.05954  5/2 -                                              
   3  0.07592  9/2 +  *                                           
   4  0.10296  7/2 -                                              
   5  0.13000 11/2 +  *                                           
   6  0.15851  9/2 -                                              
   7  0.19146 13/2 +  *                                           
   8  0.22596 11/2 -                                              
   9  0.26754  3/2 -                                              
  10  0.26990 15/2 +                                              
  11  0.28135  1/2 -                                              
  12  0.30506 13/2 -                                              
  13  0.31680  7/2 +                                              
  14  0.32442  7/2 -                                              
  15  0.33236  1/2 +                                              
  16  0.34850 17/2 +                                              
  17  0.35970  5/2 -                                              
  18  0.36859  5/2 +                                              
  19  0.37093  3/2 +                                              
  20  0.39552 15/2 -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 3. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Np-238 18.3635  0.0000  2.2742  0.3206 -0.9887  1.4167         
   Np-237 18.2971  0.7795  2.4371  0.3964 -0.9743  3.1574         
   Np-236 18.2307  0.0000  2.1332  0.3000 -0.7998  1.1669         
   Np-235 18.1643  0.7828  2.2924  0.3974 -0.9420  3.1307         
   Np-234 18.0979  0.0000  2.1332  0.2845 -0.6773  1.0000         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
   Np-238   6.199   0.460   5.848   0.370                         
   Np-237   6.250   0.950   5.200   0.600                         
   Np-236   5.500   0.600   5.200   0.400                         
   Np-235   6.250   0.950   5.200   0.600                         
   Np-234   6.200   0.460   5.850   0.370                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Np-238 20.5728  0.0000  2.6000  0.3280 -2.4114  2.0000         
   Np-237 20.4984  0.9094  2.6000  0.3286 -1.5020  2.9094         
   Np-236 17.3240  0.0000  2.6000  0.3613 -2.5262  2.0000         
   Np-235 20.3497  0.9133  2.6000  0.3299 -1.4981  2.9133         
   Np-234 20.2753  0.0000  2.6000  0.3306 -2.4114  2.0000         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Np-238 20.5728  0.0000  0.2800  0.3661 -1.7021  2.0000         
   Np-237 20.4984  0.9094  0.2400  0.3674 -0.7921  2.9094         
   Np-236 17.3240  0.0000  0.2000  0.4067 -1.7431  2.0000         
   Np-235 20.3497  0.9133  0.1600  0.3699 -0.7870  2.9133         
   Np-234 20.2753  0.0000  0.1200  0.3712 -1.6997  2.0000         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for Np-238                   
  --------------------------------------------------------        
  K0 = 1.300   E0 = 4.500 (MeV)                                   
  * E1: ER = 10.98 (MeV) EG = 2.17 (MeV) SIG = 309.44 (mb)        
        ER = 14.08 (MeV) EG = 4.66 (MeV) SIG = 540.00 (mb)        
  * M1: ER =  6.62 (MeV) EG = 4.00 (MeV) SIG =   2.08 (mb)        
  * E2: ER = 10.17 (MeV) EG = 3.25 (MeV) SIG =   6.65 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
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