92-U -234

 92-U -234 JAEA+      EVAL-JAN10 O.Iwamoto, T.Nakagawa, et al.    
                      DIST-JUL13                       20130624   
----JENDL-4.0u2       MATERIAL 9225                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
Update File Distribution                                          
Sep.14,2012 JENDL-4.0u1                                           
                                                                  
History                                                           
05-12 Fission cross section was evaluated with GMA code.          
06-06 Resonance parameters were modified.                         
07-05 Calculated with CCONE code.                                 
07-11 Resonance parameters were revised.                          
      Data were compiled as JENDL/AC-2008/1/.                     
09-03 Negative resonance, (1,452) and (1,455) were revised.       
09-08 (MF1,MT458) was evaluated.                                  
10-01 Data of prompt gamma rays due to fission were given.        
10-02 Covariance data were given.                                 
12-02 For MF1/MT458, E_nu and E_R were corrected.  As a result,   
      the Q-vaues (= E_R) were modified for MF3/MT18,19,20,21,38. 
      Re-compiled by K. Shibata.                                  
13-06 (MF2,MT151) LFW was corrected.                              
      (MF32,MT151) LFW and RP were corrected.                     
                                                                  
                                                                  
MF=1  General Information                                         
  MT=451   Descriptive data and dictionary                        
                                                                  
  MT=452   Number of neutrons per fission                         
    Sum of MT=455 and 456.                                        
                                                                  
  MT=455   Delayed neutrons per fission                           
    Determined from nu-d of the following three nuclides and      
    partial fission cross sections calculated with CCONE code/2/. 
                                                                  
      U -235 = 0.01053                                            
      U -234 = 0.0067                                             
      U -233 = 0.004998                                           
            The data for U-235 and U233is average of systematics  
            by Tuttle/3/, Benedetti et al./4/ and Waldo et al./5/ 
            For U-234, determined from experimental data of       
            Borzakov et al./6/, Conant et al./7/ and Keepin       
            et al./8/.                                            
                                                                  
    Six group decay constants were adopted from Brady and         
    England/9/.                                                   
                                                                  
  MT=456   Prompt neutrons per fission                            
    (same as JENDL-3.3)                                           
    Based on the experimental data by Mather et al./10/ Nu-p of   
    Cf-252 spontaneous fission was assumed to be 3.756.           
                                                                  
  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 (MLBW: 1.0-5 eV - 1.5 keV)        
    Dridi/15/ analyzed the capture cross section measured at      
    n_TOF with SAMMY code, and obtained the neutron and capture   
    widths of resonances up to 1.5 keV. Their parameters were     
    adopted in the present file. Since cross sections calculated  
    with MLBW formula were almost the same as those with RM       
    formula, MLBW formula was adopted.                            
    Fission widths were calculated from the neutron and capture   
    widths of Dridi et al. and fission areas calculated from the  
    resonance parameters of James et al./16/  For new             
    resonances measured by Dridi, fission width was not given.    
    The fission width of 5.17-eV resonance was determined so as   
    to reproduce the fission cross section measured by Heyse et   
    al./17/                                                       
                                                                  
    A negative resonance was assumed at -0.97 eV by Dridi/15/.    
    Its parameters were adjusted to the thermal cross sections.   
    Scattering radius of 9.6 fm was adopted, which was in good    
    agreement with CCONE calculation.                             
                                                                  
    The thermal cross sections to be reproduced:                  
      Fission = 0.064 +- 0.014 b                                  
         Wagemans et al./18/                                      
      Capture = 100.2 +- 1.0 b                                    
         Pomerance/19/, Lounsbury et al./20/, Cabell et al./21/   
         Bringer et al./22/                                       
                                                                  
  Unresolved resonance parameters (1.5 keV - 80 keV)              
    Parameters were determined with ASREP code/23/ so as to       
    reproduce cross sections in this energy region.               
    The parameters are used only for self-shielding calculations. 
                                                                  
     Thermal cross sections and resonance integrals (at 300K)     
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.(*)                 
                     (barns)       (barns)                        
    -------------------------------------------------------       
    total           118.18                                        
    elastic          17.82                                        
    fission           0.0670           5.20                       
    capture         100.29           610                          
    -------------------------------------------------------       
         (*) 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./24/                                                    
                                                                  
  MT=2 Elastic scattering cross section                           
    Calculated as total - non-elstic scattering cross sections    
                                                                  
  MT=18 Fission cross section (Above 1.5 keV)                     
    Above 1.5 keV, the following experimental data were analyzed  
    with the GMA code/25/:                                        
                                                                  
       Authors        Energy range     Data points  Reference     
       James+         1.40keV - 8.86MeV    2845      /16/         
       Behrens+       0.105 - 18.90 MeV     147      /26/(*1)     
       Meadows        0.600 - 9.91 MeV       56      /27/(*1)     
       Goverdovskiy+  16.0 MeV                1      /28/(*1)     
       Kanda+         0.490 - 6.97 MeV       30      /29/(*1)     
       Goverdovskiy+  4.91 - 10.4 MeV        33      /30/(*1)     
       Goverdovskiy+  0.210 - 0.997 MeV      22      /31/(*1)     
       Meadows        14.7 MeV                1      /32/(*1)     
       Fursov+        0.130 - 7.40 MeV       71      /33/(*1)     
                                                                  
       (*1) Relative to U-235 fission. Data were converted to     
            cross sections using JENDL-3.3 data.                  
                                                                  
    The results of GMA were used to determine the parameters in   
    the CCONE calculation.                                        
                                                                  
  MT=19, 20, 21, 38 Multi-chance fission cross sections           
    Calculated with CCONE code, and renormalized to the total     
    fission cross section (MT=18).                                
                                                                  
                                                                  
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 fission neutrons                                   
    Calculated with CCONE code.                                   
                                                                  
  MT=455 Delayed neutrons                                         
    Taken from Brady and England /9/.                             
                                                                  
                                                                  
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        
    U-235 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 15% was assumed.                                     
                                                                  
  MT=456                                                          
    Covariance of obtained by fitting a stlight line to           
    data points with errors of 2% at 1 MeV and 5% at 10 MeV.      
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Format of LCOMP=0 was adopted. Errors of neutron and          
    capture widths were adopted from Ref./35/ Those of fission    
    widths were assumed to be the same as relative error of       
    fission widths given by James et al./16/                      
                                                                  
    Further errors of 10% were added to the total, elastic        
    scattering, fission and capture cross sections in the         
    resonance region.                                             
                                                                  
                                                                  
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=18 Fission cross section                                     
    Above 1.5 keV, evaluated with GMA code/25/. Variances         
    obtained by GMA were multiplied by a factor of 1.5.           
                                                                  
  MT=102 Capture cross section                                    
    Above 1.5 keV, 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                                           
    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 = 0,1,2,3,4 (see Table 2)                      
  * optical potential parameters /24/                             
    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= 0.94e-4 S1= 2.42e-4 R'=  9.58 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of U-234                                    
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.04350   2  +  *                                           
   2  0.14335   4  +  *                                           
   3  0.29607   6  +  *                                           
   4  0.49704   8  +  *                                           
   5  0.74120  10  +                                              
   6  0.78629   1  -                                              
   7  0.80988   0  +                                              
   8  0.84930   3  -                                              
   9  0.85170   2  +                                              
  10  0.92674   2  +                                              
  11  0.94785   4  +                                              
  12  0.96260   5  -                                              
  13  0.96860   3  +                                              
  14  0.98945   2  -                                              
  15  1.02370   4  +                                              
  16  1.02380  12  +                                              
  17  1.02383   3  -                                              
  18  1.04453   0  +                                              
  19  1.06930   4  -                                              
  20  1.08530   2  +                                              
  21  1.09090   5  +                                              
  22  1.09590   6  +                                              
  23  1.12527   7  -                                              
  24  1.12668   2  +                                              
  25  1.12760   5  -                                              
  26  1.15000   3  +                                              
  27  1.16520   3  +                                              
  28  1.17210   6  +                                              
  29  1.17420   1  +                                              
  30  1.19473   6  -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 3. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-235 18.5328  0.7828  2.6265  0.3874 -0.9246  3.1046         
    U-234 18.4650  1.5689  2.5578  0.3902 -0.1511  3.9087         
    U-233 18.3972  0.7861  2.4694  0.3819 -0.8199  2.9895         
    U-232 18.3293  1.5757  2.6095  0.3887 -0.1141  3.8805         
    U-231 18.2614  0.7895  2.6793  0.4123 -1.1756  3.4264         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
    U-235   5.790   0.600   5.600   0.320                         
    U-234   6.220   1.040   5.000   0.600                         
    U-233   5.970   0.800   5.400   0.520                         
    U-232   5.800   1.040   5.100   0.600                         
    U-231   6.000   0.800   5.600   0.520                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-235 20.3497  0.9133  2.6000  0.3445 -1.7123  3.1133         
    U-234 20.2753  1.8304  2.6000  0.3306 -0.5810  3.8304         
    U-233 18.0364  0.9172  2.6000  0.3977 -2.2218  3.5172         
    U-232 20.1263  1.8383  2.6000  0.3319 -0.5731  3.8383         
    U-231 20.0518  0.9211  2.6000  0.3325 -1.4903  2.9211         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-235 20.8948  0.9133  0.1000  0.3865 -1.0315  3.2133         
    U-234 20.2753  1.8304  0.0600  0.4009 -0.2030  4.2304         
    U-233 16.2328  0.9172  0.0200  0.5015 -1.6781  3.9172         
    U-232 20.1263  1.8383 -0.0200  0.3745  0.1401  3.8383         
    U-231 20.0518  0.9211 -0.0600  0.3758 -0.7765  2.9211         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for  U-235                   
  --------------------------------------------------------        
  K0 = 1.501   E0 = 4.500 (MeV)                                   
  * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 300.00 (mb)        
        ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)        
  * M1: ER =  6.64 (MeV) EG = 4.00 (MeV) SIG =   2.68 (mb)        
  * E2: ER = 10.21 (MeV) EG = 3.29 (MeV) SIG =   6.52 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
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