92-U -235

 92-U -235 JAEA+      EVAL-OCT09 O.Iwamoto,N.Otuka,S.Chiba,+      
                      DIST-MAY10                       20100331   
----JENDL-4.0         MATERIAL 9228                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
07-07 Calculation with CCONE code was performed.                  
07-09 Fission spectra up to 5 MeV were replaced with JENDL-3.3.   
07-11 Fission cross section was revised with simultaneous         
      evaluation.                                                 
07-12 Fission cross section was revised with new results of       
      simultaneous evaluation.                                    
08-01 Fission cross section was revised. New CCONE calculation    
      was adopted.                                                
08-02 Fission and capture cross sections, and nu-p were revised.  
      CCONE calculation was made with revised parameters.         
      Data were compiled as JENDL/AC-2008/1/.                     
09-08 (MF1,MT458) was evaluated.                                  
09-10 nu-p and fission cross section were revised.                
10-03 Covarinace data were given.                                 
                                                                  
                                                                  
MF= 1                                                             
  MT=452 Total number of neutrons per fission                     
    Sum of MT=455 and 456.                                        
                                                                  
  MT=455  Delayed neutron data                                    
    (same as JENDL-3.3/2/)                                        
    Evaluated by using the least-squares method on the basis of   
    the following experimental data in each energy region.        
                                                                  
    Thermal region: Keepin/3/, Conant/4/, Synetos/5/,             
                    Reeder/6/, Borzakov/7/                        
    50 keV - 7 MeV: Keepin/3/, Maksyutenko/8/, Masters/9/,        
                    Krick/10/, Evans/11/, Cox/12/,                
                    Besant/13/, Gudkov/14/, Loaiza/15/            
    14 - 15 MeV   : Keepin/16/                                    
                                                                  
    Decay constants at the thermal energy were adopted from       
    Keepin et al./17/                                             
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Below 500 eV, JENDL-3.3 was adopted, which was evaluated on   
    the basis of experimental data of Gwin et al./18,19/          
                                                                  
    Above 500 eV, experimental data were anlyzed by the GMA code  
    /20/ with Chiba-Smith approach/21/ for PPP minimization.      
    Experimental data are renormalized with nu-p of CF-252        
    spontaneous fission (3.756+/-0.031) reported by Vorobyev et   
    al./22/ if standards to derive original data were known.      
                                                                  
    Experimental data sets are summarized below.                  
    r: re-normalized by nu-p(252Cf spon) of A.S.Vorobyev et al.   
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
      r 12326.004  2.80E+5 - 1.45E+7  J.C.Hopkins+      /23/      
      r 12870.004  1.70E+7 - 1.96E+7  R.E.Howe          /24/      
      r 13101.003  5.00E+2 - 9.00E+6  R.Gwin+           /25/      
      r 20427.002  2.25E+5 - 1.36E+6  F.Kaeppeler+      /26/      
        21696.004  2.50E+6 - 1.41E+7  I.Johnstone       /27/      
      r 21785.003  1.14E+6 - 1.47E+7  J.Frehaut+        /28/      
      r 40262.002  8.60E+5 - 5.35E+6  M.V.Savin+        /29/      
      r 40493.002  1.98E+5 - 9.85E+5  M.V.Savin+        /30/      
        40785.002  1.43E+7            Ju.A.Vasilev+     /31/      
    --------------------------------------------------------------
                                                                  
    In the energy region from 1 keV to 1 MeV, GMA results were    
    increased by 0.2%.                                            
                                                                  
  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/32/. 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 2000/33/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/34/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/35/ 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 (RM: 1.0E-5 - 500 eV)             
    Adopted are parameters for Reich-Moore formula evaluated by   
    Leal et al./36/ In the present file, the upper boundary of    
    resolved resonance region is set to 500 eV.                   
                                                                  
    See Appendix A-1                                              
                                                                  
  Unresolved resonance parameters (500 eV - 30 keV)               
    Unresolved resonance parameters were determined with ASREP    
    code/37/ so as to reproduce the cross sections.               
    The parameters are used only for calculation of self-shielding
    factors.                                                      
                                                                  
     Thermal cross sections and resonance integrals (at 300K)     
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.(*)                 
                     (barns)       (barns)                        
    -------------------------------------------------------       
    total            698.90                                       
    elastic           15.12                                       
    fission          585.08           274                         
    capture           98.71           139                         
    -------------------------------------------------------       
         (*) In the energy range from 0.5 eV to 10 MeV.           
                                                                  
                                                                  
MF= 3 Neutron cross sections                                      
  Between 500 eV and 2.25 keV:                                    
    Cross sections were calculated with resonance parameters of   
    JENDL-3.3 which were taken from ENDF/B-VI.5/36/ and           
    broadened with a resolution function of R(E)=0.03*E.          
    The capture cross section was multiplied by ratios of average 
    capture cross sections of JENDL-3.2 and JENDL-3.3 to lower the
    cross sections to those of JENDL-3.2.                         
                                                                  
  Above 2.25 keV:                                                 
    Cross sections except for the total (M=1), elastic scattering 
    (MT=2), fission (MT=18, 19, 20, 21, 38) and capture cross     
    sections were calculated with CCONE code/38/.                 
    The model parameters were determined by considering integral  
    experimental data as well as measured cross-section data.     
                                                                  
  MT= 1 Total cross section                                       
    In the energy range from 2.25 to 500 keV, cross section was   
    calculated with CCONE code/38/.                               
    Above 500 keV, the cross section was determined by spline     
    fitting to experimental data of Schwartz et al./39/, Poenitz  
    et al./40,41/, Harvey et al./42/, Cabe and Cance/43/, and     
    Uttley et al./44/ These experimental data were used also      
    for adjustment of the OMP of Soukhovitskii et al./45/         
                                                                  
  MT=2 Elastic scattering cross section                           
    Calculated as total - non-elstic scattering cross section     
                                                                  
  MT=16 (n,2n) cross section                                      
    Calculated with CCONE code. The experimental data of Frehaut  
    et al./46/ were considered to determine the model parameters  
    of CCONE calculation.                                         
                                                                  
  MT=18 Fission cross sections                                    
    From 2.25 keV to 10 keV, JENDL-3.3/2/ was adopted. The data   
    of JENDL-3.3 were based on the experimental data of Weston    
    and Todd/47/.                                                 
                                                                  
    Above 10 keV, experimental data measured after 1980 were      
    analyzed by simultaneous fitting of U-233, U-235, U-238,      
    Pu-239, Pu-240 and Pu-241 fission cross section and its ratio 
    by the SOK code /48/.                                         
                                                                  
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        41112.002  1.88E+6 - 2.37E+6  V.A.Kalinin+      /49/      
        22304.006  2.60E+6 - 1.47E+7  K.Merla+          /50/      
        22304.002  4.45E+6 - 1.88E+7  K.Merla+          /50/      
        12924.002  1.07E+6 - 5.99E+6  R.G.Johnson+      /51/      
        40969.011  6.24E+5 - 7.85E+5  N.N.Buleeva+      /52/      
        22091.002  1.35E+7 - 1.49E+7  T.Iwasaki+        /53/      
        30721.002  1.42E+7            J.W.Li+           /54/      
        12877.004  5.05E+3 - 2.05E+5  L.W.Weston+       /47/      
        10987.002  3.10E+5 - 2.82E+6  A.D.Carlson+      /55/      
        30634.002  1.47E+7            J.W.Li+           /56/      
        12826.002  1.46E+7            M.Mahdavi+        /57/      
        10971.002  1.41E+7            O.A.Wasson+       /58/      
        21620.002  2.50E+6 - 4.45E+6  M.Cance+          /59/      
        21777.002  5.40E+3 - 8.25E+4  F.Corvi+          /60/      
        10950.002  2.45E+5 - 1.20E+6  O.A.Wasson+       /61/      
        40601.002  6.00E+3 - 4.50E+4  A.A.Bergman+      /62/      
        -----.---  3.03E+6 - 2.96E+7  A.D.Carlson+      /63/      
    --------------------------------------------------------------
                                                                  
    The obtained cross section from 1 to 4 MeV and from 7 to 8    
    MeV was slightly modified for JENDL-4.                        
                                                                  
    The fission cross section of JENDL-3.3 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).                                
                                                                  
  MT=102 Capture cross section                                    
    From 2.25 keV to 1 MeV, experimental data measured after 1970 
    were analyzed by the GMA code/20/ with the Chiba and Smith    
    approach/21/ for PPP minimization/64/. All experimental       
    data are given in the form of alpha-value (=ratios to the     
    U-235(n,f) cross section), which were normalized to absolute  
    cross section by the JENDL-3.3 U-235(n,f) cross section. Data 
    points of Kononov et al./65/ below 20 keV were not            
    considered because their systematic error was very large in   
    this energy region.                                           
                                                                  
    Experimental data sets are summarized below.                  
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        12424.003  3.47E+3 - 2.56E+4  J.B.Czirr+        /66/      
        20158.002  8.50E+3 - 5.50E+4  R.E.Bandl+        /67/      
        20880.002  1.06E+4 - 1.96E+5  H.Beer+           /68/      
        20880.003  1.04E+5 - 3.07E+5  H.Beer+           /68/      
        20880.004  1.70E+4 - 6.40E+4  H.Beer+           /68/      
        20880.005  3.79E+5 - 4.81E+5  H.Beer+           /68/      
        21777.004  2.50E+3 - 8.25E+4  F.Corvi+          /60/      
        40412.002  2.04E+4 - 1.10E+6  V.N.Kononov+      /65/      
        40502.002  3.49E+3 - 8.89E+3  Ju.V.Ryabov       /69/      
        40581.002  2.50E+3 - 4.50E+4  G.V.Muradyan+     /70/      
        40609.004  2.45E+4            V.P.Vertebnyy+    /71/      
        12409.003  2.00E+5 - 6.00E+5  G.de Saussure+    /72/      
        12407.002  1.23E+4 - 6.90E+5  L.W.Weston+       /73/      
        12331.005  3.00E+4 - 1.00E+6  J.C.Hopkins+      /74/      
        12416.002  1.78E+5 - 1.01E+6  B.C.Diven+        /75/      
    --------------------------------------------------------------
                                                                  
    The results of the GMA were used to determine the parameters  
    in the CCONE calculation.                                     
                                                                  
    Above 1.3 MeV, results of the CCONE calculation were adopted. 
                                                                  
    The present results were slightly modified by considering     
    integral data.                                                
                                                                  
                                                                  
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 neutron spectra                            
    Below 5 MeV, spectra given in JENDL-3.3/2/ were adopted.      
    Relevant part of JENDL-3.3 comments:                          
    *DISTRIBUTIONS WERE CALCULATED WITH A MODIFIED MADLAND-NIX    
     MODEL WITH CONSIDERATION FOR MULTIMODAL NATURE OF THE FISSION
     PROCESS/76,77/.  THE COMPOUND NUCLEUS FORMATION CROSS SEC-   
     TIONS FOR FISSION FRAGMENTS WERE CALCULATED USING BECCHETTI- 
     GREENLEES POTENTIAL/78/.  THE IGNATYUK FORMULA/79/ WERE      
     USED TO GENERATE THE LEVEL DENSITY PARAMETERS. UP TO         
     3rd-CHANCE-FISSION WERE CONSIDERED AT HIGH INCIDENT NEUTRON  
     ENERGIES.                                                    
                                                                  
       PARAMETERS ADOPTED FOR THERMAL-NEUTRON FISSION:            
          (S1: STANDARD-1, S2: STANDARD-2, SL: SUPERLONG MODES)   
          TOTAL AVERAGE FRAGMENT KINETIC ENERGY                   
                                          = 187 MEV FOR S1        
                                          = 167 MEV FOR S2        
                                          = 157 MEV FOR SL        
           AVERAGE ENERGY RELEASE                                 
                                          = 194.49 MEV FOR S1     
                                          = 184.86 MEV FOR S2     
                                          = 190.95 MEV FOR SL     
           AVERAGE MASS NUMBER OF LIGHT FF =  96                  
           AVERAGE MASS NUMBER OF HEAVY FF = 140                  
           LEVEL DENSITY OF THE LIGHT FF   = 10.31(S2), 11.43(S1) 
           LEVEL DENSITY OF THE HEAVY FF   = 8.89(S1), 13.25(S2)  
           MODE BRANCHING RATIO = 0.18342(S1), 0.81589(S2),       
                                  0.00069(SL)                     
       NOTE THAT THE PARAMETERS VARY WITH THE INCIDENT ENERGY     
       WITHIN THE INDICATED RANGE.                                
                                                                  
    Above 5.5 MeV, calculated with CCONE code/38/.                
                                                                  
  MT=455 Delayed neutron spectra                                  
    (same as JENDL-3.3)                                           
    Taken from Brady and England/80/. Group abundace parameters   
    were adjusted so as to reproduce total delayed neutron        
    emission rate measured by Keepin/17/, Piksaikin/81/ and       
    East/82/.                                                     
                                                                  
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code.                                   
    Distributions from fission (MT=18) are not included.          
                                                                  
                                                                  
MF=12 Photon Production Multiplicities (option 1)                 
  MT=18 Fission                                                   
    (same as JENDL-3.3)                                           
    The thermal neutron-induced fission gamma spectrum measured   
    by Verbinski et al./83/ was adopted.                          
                                                                  
                                                                  
MF=14  Photon Angular Distributions                               
  MT=18                                                           
    (same as JENDL-3.3)                                           
    Isotropic distributions were assumed.                         
                                                                  
                                                                  
MF=15  Continuous Photon Energy Spectra                           
  MT=18                                                           
    (same as JENDL-3.3)                                           
    Experimental data by Verbinski et al./83/ 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                                                          
     (Same as JENDL-3.3/2/)                                       
                                                                  
  MT=456                                                          
     Below 500 eV, the covariance of JENDL-3.3 was adopted.       
     Above 500 eV, it was obtained by fitting to the experimental 
     data described above. The error of nu-p was multiplied by    
     a factor of 2.0.                                             
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Compact format (LCOMP=2, NDIGIT=5) was adopted.               
                                                                  
    Below 200 eV: covariance matrix of resonance parameters       
                  obtained with SAMMY code /84/ was given.        
    200 - 500 eV: covariance matrices were given in MF33.         
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/85/ 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 energy range from 200 to 500 eV, uncertainty of 2%     
    was assumed for the total cross section, and 4% for the       
    elastic scattering. From 0.5 to 2.25 keV, uncertainty of 5%   
    was assumed. Above 2.25 keV, covariance of the CCONE          
    calculation was adopted.                                      
                                                                  
  MT=18 Fission cross section                                     
    In the energy range from 200 to 500 eV, uncertainty of 2%     
    was assumed. From 500 eV to 9 keV, uncertainty was assumed    
    to be 5%.                                                     
    Above 9 keV, covariance matrix was obtained by simultaneous   
    evaluation among the fission cross sections of U-233, U-235,  
    U-238, Pu-239, Pu-240 and Pu-241(See MF=3, MT=18, and /1/).   
    Since the variances are very small, they were adopted by      
    multiplying a factor of 2.                                    
                                                                  
  MT=102 Capture cross section                                    
    In the energy range from 200 to 500 eV, uncertainty of 3%     
    was assumed.                                                  
    From 500 eV to 2.25 keV, uncertainty was assumed to be 10%.   
    In the energy region from 2.25 keV to 1 MeV, capture cross    
    section was obtained with GMA code/64/. Its covariance matrix 
    was obtained simultaneously.                                  
    Above 1 MeV, covariance of the CCONE calculation was adopted. 
                                                                  
                                                                  
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 5 MeV, based on the covarinaces given in JENDL-3.3.     
    Above 5 MeV, estimated with CCONE and KALMAN codes.           
                                                                  
                                                                  
***************************************************************** 
  Calculation with CCONE code                                     
***************************************************************** 
                                                                  
  Models and parameters used in the CCONE/38/ 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/86/                              
    * Global parametrization of Koning-Duijvestijn/87/            
      was used.                                                   
    * Gamma emission channel/88/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/89/ 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/90/. 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/91/,/92/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,3,6,9 (see Table 2)                        
  * optical potential parameters /45/                             
    Volume:                                                       
      V_0       = 49.8613  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       = 0.61     fm                                     
    Surface:                                                      
      W_0       = 17.1117  MeV                                    
      B_s       = 11.19    MeV                                    
      C_s       = 0.01361  1/MeV                                  
      C_wiso    = 23.5     MeV                                    
      r_s       = 1.1803   fm                                     
      a_s       = 0.61     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    = 0.201044                                        
      beta_4    = 0.11                                            
      beta_6    = 0.0015                                          
                                                                  
  * Calculated strength function                                  
    S0= 0.93e-4 S1= 2.11e-4 R'=  9.50 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of U-235                                    
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000  7/2 -  *                                           
   1  0.00008  1/2 +                                              
   2  0.01304  3/2 +                                              
   3  0.04621  9/2 -  *                                           
   4  0.05171  5/2 +                                              
   5  0.08174  7/2 +                                              
   6  0.10304 11/2 -  *                                           
   7  0.12930  5/2 +                                              
   8  0.15047  9/2 +                                              
   9  0.17071 13/2 -  *                                           
  10  0.17139  7/2 +                                              
  11  0.19712 11/2 +                                              
  12  0.22542  9/2 +                                              
  13  0.24913 15/2 -                                              
  14  0.25900  7/2 +                                              
  15  0.29114 11/2 +                                              
  16  0.29467 13/2 +                                              
  17  0.33285  5/2 +                                              
  18  0.33852 17/2 -                                              
  19  0.35730 15/2 +                                              
  20  0.36707  7/2 +                                              
  21  0.36900 13/2 +                                              
  22  0.39322  3/2 +                                              
  23  0.41478  9/2 +                                              
  24  0.42675  5/2 +                                              
  25  0.43860 19/2 -                                              
  26  0.44572  7/2 +                                              
  27  0.45450 15/2 +                                              
  28  0.47130 11/2 +                                              
  29  0.47430  7/2 +                                              
  30  0.48500 17/2 +                                              
  31  0.50992  9/2 +                                              
  32  0.53240 13/2 +                                              
  33  0.53323  9/2 +                                              
  34  0.55040 21/2 -                                              
  35  0.56800 19/2 +                                              
  36  0.58780 11/2 +                                              
  37  0.60808 11/2 +                                              
  38  0.61640 15/2 +                                              
  39  0.63317  5/2 -                                              
  40  0.63781  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         
  --------------------------------------------------------        
    U-236 18.6157  1.5623  2.7551  0.3856 -0.1537  3.8909         
    U-235 18.4419  0.7828  2.6265  0.3721 -0.7434  2.8828         
    U-234 18.4800  1.5689  2.5578  0.3899 -0.1502  3.9076         
    U-233 18.4122  0.7861  2.4694  0.3817 -0.8188  2.9881         
    U-232 18.3442  1.5757  2.6095  0.3885 -0.1133  3.8795         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
    U-236   6.201   1.040   5.417   0.550                         
    U-235   5.700   0.400   5.600   0.300                         
    U-234   6.050   1.040   5.400   0.600                         
    U-233   5.970   0.800   5.450   0.520                         
    U-232   5.800   1.040   5.100   0.600                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-236 20.9712  1.8226  2.6000  0.3244 -0.5720  3.8226         
    U-235 20.3497  0.9133  2.6000  0.3515 -1.8195  3.2133         
    U-234 20.2753  1.8304  2.6000  0.3522 -0.9023  4.1304         
    U-233 20.2008  0.9172  2.6000  0.3312 -1.4942  2.9172         
    U-232 20.1263  1.8383  2.6000  0.3319 -0.5731  3.8383         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-236 21.5182  1.8226  0.1400  0.3654  0.0498  3.9226         
    U-235 20.3497  0.9133  0.1000  0.4065 -1.2054  3.4133         
    U-234 20.2753  1.8304  0.0600  0.3939 -0.1191  4.1304         
    U-233 20.2008  0.9172  0.0200  0.3732 -0.7817  2.9172         
    U-232 20.1263  1.8383 -0.0200  0.3745  0.1401  3.8383         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for  U-236                   
  --------------------------------------------------------        
  K0 = 1.500   E0 = 4.500 (MeV)                                   
  * E1: ER = 10.90 (MeV) EG = 2.50 (MeV) SIG = 303.18 (mb)        
        ER = 13.80 (MeV) EG = 4.70 (MeV) SIG = 450.00 (mb)        
  * M1: ER =  6.63 (MeV) EG = 4.00 (MeV) SIG =   2.69 (mb)        
  * E2: ER = 10.19 (MeV) EG = 3.28 (MeV) SIG =   6.51 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
References                                                        
 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009).     
 2) T.Kawano et al.: JAERI-Research 2003-026 (2003).              
 3) G.R.Keepin et al.: J. Nucl. Ener, 6, 1 (1957).                
 4) J.F.Conant, P.F.Palmedo: Nucl. Sci. Eng., 44, 173 (1971).     
 5) S.Synetos, J.G.Williams: INDC(NDS)-107, 183 (1979).           
 6) P.L.Reeder, J.G.Williams: Phys. Rev., 28C, 1740 (1983).       
 7) S.B.Borzakov et al.: Atomnaya Energiya, 79, 231 (1995).       
 8) B.P.Maksyutenko: Sov. Phys. JETP, 8, 565 (1959).              
 9) C.F.Masters et al.: Nucl. Sci. Eng., 36, 202 (1969).          
10) M.S.Krick, A.E.Evans: Nucl. Sci. Eng., 47, 311 (1972).        
11) A.E.Evans, M.M.Thorpd: Nucl. Sci. Eng., 50, 80 (1973).        
12) S.A.Cox: ANL/NDM-5 (1974)                                     
13) C.B.Besant et al.: British Nucl. Ener. Soc., 16, 161 (1977).  
14) A.N.Gudkov et al.: Atomnaya Energiya,  66, 100 (1989).        
15) D.Loaiza et al.: ANS, 76, 361 (1997).                         
16) G.R.Keepin: LA-4320, (1969).                                  
17) G.R.Keepin et al.: Phys. Rev., 107, 1044 (1957).              
18) R.Gwin et al.: Nucl. Sci. Eng., 87, 381 (1984). and           
    EXFOR 12906.003                                               
19) R.Gwin: ORNL-TM-6246 (1978).                                  
20) W.P.Poenitz: BNL-NCS-51363, Vol.I, p.249 (1981).              
21) S.Chiba, D.L.Smith: ANL/NDM-121 (1991).                       
22) A.S.Vorobyev et al.: 2004 Santa Fe, p.613 (2004).             
23) J.C.Hopkins et al.: Nucl. Phys. 48, 433 (1963).               
24) R.E.Howd: Nucl. Sci. Eng. 86, 157 (1984).                     
25) R.Gwin et al.: Nucl. Sci. Eng. 94, 365 (1986).                
26) F.Kaeppeler et al.: 1975 Washington, Vol.2, p.549 (1975).     
27) I.Johnstond: AERE-NP/R-1912 (1956).                           
28) J.Frehaut et al.: 1982 Antwerp, p.78 (1982).                  
29) M.V.Savin et al.: Sov. Nucl. Phys. 16, 638 (1973).            
30) M.V.Savin et al.: YFI-27, p.4 (1979).                         
31) Ju.A.Vasilev et al.: Sov. Phys. 11, 483 (1960).               
32) G.Audi: Private communication (April 2009).                   
33) J.Katakura et al.: JAERI 1343 (2001).                         
34) T.R.England et al.: LA-11151-MS (1988).                       
35) R.Sher, C.Beck: EPRI NP-1771 (1981).                          
36) L.C.Leal et al.: Nucl. Sci. Eng., 131, 230 (1999).            
37) Y.Kikuchi et al.: JAERI-Data/Code 99-025 (1999) in Japanese.  
38) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007).            
39) R.B.Schwartz et al.: Nucl. Sci. Eng. 54, 322 (1974).          
40) W.P.Poenitz,J.F.Whalen: ANL-NDM-80 (1983).                    
41) W.P.Poenitz et al.: Nucl. Sci. Eng. 78, 333 (1981).           
42) J.A.Harvey et al.: 1988 Mito, p.115 (1988).                   
43) J.Cabe, M.Cancd: CEA-R-4524 (1973).                           
44) C.A.Uttley et al.: 1966 Paris, Vol.1, p.165 (1966).           
45) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).     
46) J.Frehaut et al.: Nucl. Sci. Eng., 74, 29 (1974).             
47) L.W.Weston, J.H.Todd: Nucl. Sci. Eng., 88, 567 (1984).        
48) T.Kawano et al.: JAERI-Research 2000-004 (2000).              
49) V.A.Kalinin et al.: At. Energy, 71, 181 (1991).               
50) K.Merla et al.: 1991 Juelich, p.510 (1991).                   
51) R.G.Johnson et al.: Private communication (1991).             
52) N.N.Buleeva et al.: At. Energy, 65, 348 (1988).               
53) T.Iwasaki et al.: 1988 Mito, p.87 (1988).                     
54) J.W.Li et al.: INDC(CPR)-009, 3 (1986)                        
55) A.D.Carlson et al.: Private communication (1984).             
56) J.W.Li et al.: 1982 Anterp, p.55 (1982).                      
57) M.Mahdavi et al.: 1982 Antwerp, p.58 (1982).                  
58) O.A.Wasson et al.: Nucl. Sci. Eng., 80, 282 (1982).           
59) M.Cance et al.: CEA-N-2194 (1981).                            
60) F.Corvi et al.: NEANDC(E)-232, 5 (1982).                      
61) O.A.Wasson et al.: Nucl. Sci. Eng., 81, 196 (1982).           
62) A.A.Bergman et al.: 1980 Kiev, Vol.3, p.54 (1980)             
63) A.D.Carlson et al.: Private communication (2007).             
64) N.Otuka et al.: J. Nucl. Sci. Technol., 44, 815 (2007).       
65) V.N.Kononov et al.: At. Energy, 38, 81 (1975).                
66) J.B.Czirr et al.: 1970 Helsinki, Vol.1, p.331(47 (1970).      
67) R.E.Bandl et al.: Nucl. Sci. Eng., 48, 324 (1972).            
68) H.Beer et al.: PSB-BER-624 (1978).                            
69) Ju.V.Ryabov: At. Energy, 40, 339 (1976).                      
70) G.V.Muradyan et al.: 1980 Kiev, Vol.2, p.119 (1980).          
71) V.P.Vertebnyy et al.: 1980 Kiev, Vol.2, p.254 (1980).         
72) G.de Sasussure et al.: 1966 Paris, Vol.2, p.233(48)(1966).    
73) L.W.Weston et al.: Nucl. Sci. Eng., 20, 80 (1964).            
74) J.C.Hopkins et al.: Nucl. Sci. Eng., 12, 169 (1962).          
75) B.C.Diven et al.: Phys. Rev., 109, 144 (1958).                
76) D.G.Madland, J.R.Nix J.R.: Nucl. Sci. Eng., 81, 213 (1982).   
77) T.Ohsawa et al.: to be published. See also. T.Ohsawa et al.,  
    Nucl. Phys. A653, 17 (1999)                                   
78) F.D.Becchetti Jr., G.W.Greenlees: Phys. Rev., 182, 1190       
     (1969).                                                      
79) A.V.Ignatyuk: Sov. J. Nucl. Phys., 29, 450 (1979).            
80) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989).     
81) V.M.Piksaikin: private communication (1997).                  
82) L.V.East et al.: LA-4605-MS (1970).                           
83) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).            
84) L.C.Leal et al.: private communication (2008).                
85) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in         
    Japanese.                                                     
86) C.Kalbach: Phys. Rev. C33, 818 (1986).                        
87) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).     
88) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).     
89) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).                   
90) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).            
91) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990).                
92) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).     
                                                                  
                                                                  
***************************************************************** 
  Appendix A-1 from ENDF/B-VI.8                                   
***************************************************************** 
 File 2                                                           
   MT=151  Resonance parameters, from a new analysis by Leal      
     et al. [LE97], using the multilevel R-matrix analysis code   
     SAMMY [LA96].  Energy range for U235 is 0 to 2.25 keV.       
     For the first time, integral data were fitted during the     
       analysis process:  Thermal cross sections (fission,        
       capture, and elastic), Westcott g-factors (fission and     
       absorption) are from the ENDF/B-6 standards [CA93], and    
       the K1 value is from Hardy [HA79].                         
     Thermal parameters obtained in the present evaluation,       
       first using the microscopic experimental data only, and    
       second including the integral data as well, are compared   
       to the SAMMY input in the following Tabld:                 
                                                                  
       Parameter        SAMMY input       Fit to    Fit to diff.  
                          value          diff data    & integ.    
                                          alone         data      
       -----------    -----------------  --------   ------------  
       Fission        584.25 +/- 1.11    582.28      584.88       
       Capture         98.96 +/- 0.74     99.18       98.66       
       Scattering      15.46 +/- 1.06     15.44       15.12       
       Westcott gf    0.9771 +/- 0.0008    0.9743      0.9764     
       Westcott ga    0.9790 +/- 0.0008    0.9774      0.9785     
       Westcott gg                         0.9956      0.9910     
       K1(barn)       722.70 +/- 3.90    717.48      722.43       
                                                                  
     The final adjustment of nu by SAMMY to the recommended K1    
       value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission   
       and absorption cross sections calculated from the final    
       resonance parameters.                                      
     In the following Tables, the fission and capture cross       
       sections calculated in this evaluation with the code       
       SAMMY are compared with experimental data.                 
                                                                  
       Experimental and calculated total cross sections.          
                                                                  
       Energy Range    Calculated  Schrack   Weston   Weston      
           (eV)          (b.eV)     (b.eV)   (b.eV)   (b.eV)      
      ---------------  ----------  -------   ------   ------      
         0.5 -   20.0    910.4      929.9                         
        20.0 -   60.0   1867.8     1882.8    1869.9               
        60.0 -  100.0    954.0      968.0     954.2               
       100.0 -  200.0   2032.7     2092.7    2089.5   2073.9      
       200.0 -  300.0   2062.2     2007.0    2060.0   2054.6      
       300.0 -  400.0   1280.8     1321.6    1297.1   1292.9      
       400.0 -  500.0   1333.2     1391.5    1351.8   1347.9      
       500.0 -  600.0   1489.2     1467.9    1499.2   1494.3      
       600.0 -  700.0   1126.6     1156.4    1134.1   1132.6      
       700.0 -  800.0   1088.7     1085.8    1093.3   1075.7      
       800.0 -  900.0    797.6      784.0     813.0    804.9      
       900.0 - 1000.0    724.4      723.9     738.2    721.4      
      1000.0 - 2000.0   7036.1                        7054.2      
                                                                  
       Experimental and calculated capture cross sections.        
                                                                  
       Energy Range     Calculated   De Saussure     Perez        
           (eV)           (b.eV)       (b.eV)       (b.eV)        
      ---------------   ----------   -----------    ------        
        0.5  -   20.0      653.5         647                      
       20.0  -   60.0     1066.1        1084         1057         
       60.0  -  100.0      490.2         477          504         
       100.0 -  200.0     1158.8        1148         1138         
       200.0 -  300.0      907.8         904          940         
       300.0 -  400.0      660.2         658          642         
       400.0 -  500.0      495.9         506          478         
       500.0 -  600.0      533.3         506          562         
       600.0 -  700.0      494.8         481          449         
       700.0 -  800.0      490.1         513          475         
       800.0 -  900.0      439.8         444          397         
       900.0 - 1000.0      504.2         542          482         
      1000.0 - 1100.0      509.6         522          463         
      1100.0 - 1200.0      413.7         395          332         
      1200.0 - 1300.0      340.4         372          267         
      1300.0 - 1400.0      304.1         304          225         
      1400.0 - 1500.0      355.7         301          254         
      ---------------   ----------   -----------    ------        
        20.0 - 1500.0     9164.7        9046         8665         
                                                                  
    The fission and capture resonance integral calculated from    
 the present evaluation are 276.04 b and 140.49 b, respectively,  
 giving a capture-to-fission ratio (alpha value) of 0.509 in      
 excellent agreement with the value obtained from integral        
 measurements.                                                    
    The following energy-differential data were included in the   
 analysis:                                                        
  (1) Transmission data of Harvey et al. [HA86] on the ORELA      
      18-meter flight path, with sample thickness of 0.03269      
      atoms/barn, cooled to 77 K (0.4 to 68 eV).                  
  (2) Transmission data of Harvey et al. [HA86] on the ORELA      
      80-meter flight path, with sample thickness of 0.00233      
      atoms/barn, cooled to 77 K (4 to 2250 eV).                  
  (3) Transmission data of Harvey et al. [HA86] on the ORELA      
      80-meter flight path, with sample thickness of 0.03269      
      atoms/barn, cooled to 77 K (4 to 2250 eV).                  
  (4) Fission data of Schrack [SC88] on the RPI Linac at 8.4      
      meter flight path (0.02 to 20 eV).                          
 (5,6) Fission and capture data of de Saussure et al. [DE67]      
      on the ORELA 25.2-meter flight path (0.01 to 2250 eV).      
 (7,8) Fission and capture data of Perez et al. [PE73] on the     
      ORELA 39-meter flight path (0.01 to 100 eV).                
  (9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter  
      flight path (0.01 to 20 eV).                                
 (10) Transmission data of Spencer et al. [SP84] on the ORELA     
     ORELA 18-meter flight path, sample thickness of 0.001468     
     atom/barn (0.01 to 1.0 eV).                                  
 (11) Fission data of Wagemans et al. [WA88] on the Geel 18-      
      meter flight path (0.001 to 1.0 eV)                         
 (12,13) Absorption and fission data of Gwin [GW96] at ORELA      
      (0.01 to 4.0 eV).                                           
 (14) Fission data of Weston and Todd [WE84] on the ORELA         
      18.9-meter flight path (14 to 2250 eV).                     
 (15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to    
      1.0 eV).                                                    
 (16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to      
      0.15 eV).                                                   
 (17) Fission data of Weston and Todd [WE92] on the ORELA         
      86.5-meter flight path (100 to 2000 eV).                    
 (18) Fission yield data of Moxon et al. [MO92] at ORELA          
      (0.01 to 50.0 eV).                                          
                                                                  
 ---------------------------------------------------------------- 
 REFERENCES FOR RESOLVED RESONANCE REGION                         
                                                                  
 [CA93] A. Carlson, W.P. Poenitz, G.M. Hale et al., "The ENDF/B-6 
    Neutron Cross Section Measurements Standards," National       
    Institute of Standards and Technology report NISTIR-5177      
    (1993)                                                        
 [DE67] G. de Saussure, R. Gwin, L.W. Weston, and R.W. Ingle,     
    "Simultaneous Measurements of the Neutron Fission and Capture 
    Cross Section for 235U for Incident Neutron Energy from       
    0.04 eV to 3 keV," Oak Ridge National Laboratory report       
    ORNL/TM-1804 (1967)                                           
 [GW84] R. Gwin, R.R. Spencer, R.W. Ingle, J.H. Todd, and S.W.    
    Scoles, Nuc.Sci.Eng. 88, 37 (1984)                            
 [GW96] R. Gwin, To be published in Nuclear Science Engineering   
 [HA79] J. Hardy, Brookhaven National Laboratory, report          
    BNL-NCS-51123 [ENDF-300] (1979) Sec. B.1                      
 [HA86] J.A. Harvey, N.W. Hill, F.G. Perey et al., Nuclear Data   
    for Science and Technology, Proc. Int. Conf. May 30-June 3,   
    1988, Mito, Japan. (Saikon Publishing, 1988) p. 115           
 [LA96] N.M. Larson, "Updated Users' Guide to SAMMY" report       
    ORNL/TM-9179/R3 (1996)                                        
 [LE97] L.C. Leal, H. Derrien, N.M. Larson, R.Q. Wright,          
    "R-Matrix Analysis of 235U Neutron Transmission and Cross     
    Sections in the Energy Range 0 eV to 2.25 keV," Oak Ridge     
    National Laboratory report ORNL/TM-13516 (1997).              
 [MO92] M.C. Moxon, J.A. Harvey, and N.W. Hill, private           
    communication, Oak Ridge National Laboratory (1992).          
 [PE73] R.B. Perez, G. de Saussure, and E.G. Silver, Nucl.Sci.    
    Eng. 52, 46 (1973)                                            
 [SC88] R.A. Schrack, "Measurement of the 235U(n,f) Reaction from 
    Thermal to 1 keV," Nuclear Data for Science and Technology,   
    Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon           
    Publishing, 1988) p. 101                                      
 [SP84] R.R. Spencer, J.A. Harvey, N.W. Hill, and L. Weston,      
    Nucl.Sci.Eng. 96, 318 (1987)                                  
 [WA87] J.A. Wartena, H. Weigmann, and C. Burkholz, report  IAEA  
    Tecdoc 491 (1987) p.123                                       
 [WA88] C. Wagemans, P. Schillebeeckx, A.J. Deruyter, and R.      
    Barthelemy, "Subthermal Fission Cross Section Measurements    
    for 233U and 239Pu," Nuclear Data for Science and Technology, 
    Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon           
    Publishing, 1988) p. 91                                       
 [WE84] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 88, 567 (1984)   
 [WE90] H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach,    
    and J.A. Wartena, The Physics of Reactors, Proc. Int.  Conf., 
    Marseille, 1990, Vol.1 (1990) p. 133                          
 [WE92] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992)