94-Pu-241

 94-Pu-241 JAEA+      EVAL-JAN10 O.Iwamoto,N.Otuka,S.Chiba,+      
                      DIST-MAY10                       20100325   
----JENDL-4.0         MATERIAL 9443                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
06-09 Numbers of neutrons per fission were revised.               
07-05 Theoretical calculation was made with CCONE code.           
07-11 Fission cross section was revised with simultaneous         
      evaluation.                                                 
07-12 Fission cross section was revised with a result of new      
      simultaneous evaluation.                                    
08-01 Fission cross section was revised.                          
08-02 Fission cross section and nu-p were revised.                
      CCONE calculation was made with revised parameters.         
      Data were compiled as JENDL/AC-2008/1/                      
09-03 (1,452) and (1,455) were revised.                           
09-08 (MF1,MT458) was evaluated.                                  
09-09 fission and total sigs and URP were revised.                
09-10 fission cross section was revised.                          
09-12 New theoretical calculation was made with CCONE code.       
      Nu-d was revised.                                           
10-01 Data of prompt gamma rays due to fission were given.        
10-03 Covariance data were given.                                 
                                                                  
                                                                  
MF= 1 General information                                         
  MT=452 Number of Neutrons per fission                           
    Sum of MT's = 455 and 456.                                    
                                                                  
  MT=455 Delayed neutron data                                     
    Determined from nu-d of the following three nuclides and      
    partial fission cross sections calculated with CCONE code/2/. 
                                                                  
      Pu-242 = 0.0160   measured by Benedetti et al./3/           
      Pu-241 = 0.0064   data measured by Benedetti et al./3/      
                        reduced by a factor of 0.7                
      Pu-240 = 0.0046   data for Pu-239, reduced by a factor      
                        of 0.7                                    
                                                                  
    Decay constants were adopted from Ref./4/.                    
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Experimental data of Conde et al./5/, Frehaut et al./6/ and   
    D'yachenko et al./7/ were reproduced with two straight lines. 
    A thermal value of 2.929 evaluated by Holden and Zucker/8/    
    was also considered. They were re-normalized to the nu-p of   
    Cf-252 spontaneous fission of 3.756/9/.                       
                                                                  
  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/10/. 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 2000/11/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/12/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/13/ 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: below 300 eV)                
    Revised Derrien's evaluation/14/ below 20 eV was adopted.     
    Numerical data were taken from ENDF/B-VII.0/15/.              
                                                                  
    See Apendix A-1.                                              
                                                                  
  Unresolved resonance parameters (300 eV - 30 keV)               
    Parameters were determined with ASREP code/16/ to reproduce   
    capture cross section measured by Weston and Todd/17/,        
    fission cross section measured by Weston and Todd/17/,        
    Migneco et al./18/ and Gerasimov et al./19/, and total cross  
    section calculated with CCONE code.                           
                                                                  
    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           1386.67                                       
    elastic           11.26                                       
    fission         1012.34          567                          
    capture          363.06          180                          
   -------------------------------------------------------        
         (*) 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), fission (MT=18, 19, 20, 21, 38)  
  and capture cross sections were calculated with CCONE code/2/.  
  The model parameters were determined by considering integral    
  data as well as measured capture cross sections data and        
  fission cross section of JENDL-3.3.                             
                                                                  
  MT= 1 Total cross section                                       
    The cross section was calculated with CC OMP of Soukhovitskii 
    et al./20/.                                                   
                                                                  
  MT=2 Elastic scattering cross section                           
    From 300 eV to 30 keV, CCONE calculation was adopted.         
    Above 30 keV, it was calculated as (total)-(non-elastic       
    scattering cross sections).                                   
                                                                  
  MT=18 Fission cross section                                     
    From 300 eV to 10 keV, the data of Weston and Todd/17/,       
    Migneco et al./18/ and Gerasimov et al. /19/ were taken       
    into consideration.                                           
    Above 10 keV, experimental data measured after 1960 were      
    analyzed by simultaneous fitting of U-233, U-235, U-238,      
    Pu-239, Pu-240 and Pu-241 fission cross sections and their    
    ratios by the SOK code /21/.                                  
                                                                  
    --------------------------------------------------------------
     Cross section                                                
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        30548.003  1.48E+7            N.A.Khan+         /22/      
        10636.002  5.00E+3 - 6.50E+4  G.W.Carlson+      /23/      
        20570.004  1.18E+6 - 2.63E+6  I.Szabo+          /24/      
        20567.004  3.50E+4 - 9.70E+5  I.Szabo+          /24/      
        20484.002  5.00E+3 - 2.98E+4  J.Blons+          /25/      
        40636.007  1.46E+7            M.I.Kazarinova+   /26/      
    --------------------------------------------------------------
                                                                  
     Ratio to U-235(n,f) cross section                            
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        40474.003  2.40E+4 - 7.40E+6  B.I.Fursov+       /27/      
        40474.003  1.27E+5 - 7.00E+6  B.I.Fursov+       /27/      
        10563.002  5.00E+3 - 2.96E+7  J.W.Behrens+      /28/      
        20364.002  1.37E+4 - 1.13E+6  F.Kaeppeler+      /29/      
    --------------------------------------------------------------
                                                                  
    Cross section was slightly modified in the energy region from 
    1 to 4 MeV, and from 7 to 8 MeV for JENDL-4.                  
                                                                  
  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 300 eV to 30 keV, the cross section was calculated from  
    the unresolved resonance parameters mentioned above. CCONE    
    calculation was adopted in the energy range above 50 keV.     
                                                                  
    The experimental data of Weston and Todd/17/ 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 were assumed in the laboratory system.
                                                                  
                                                                  
MF= 5 Energy distributions of secondary neutrons                  
  MT=18  Fission spectra                                          
    Calculated with CCONE code.                                   
                                                                  
  MT=455  Delayed neutron spectra                                 
     (Same as JENDL-3.3)                                          
    Results of summation calculation made by Brady and England/4/ 
    were adopted.                                                 
                                                                  
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code/2/.                                
    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./30/ 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 Number of delayed neutrons per fission                   
    Uncertainty of 5% /3/ was assumed in the energy region        
    below 5 MeV and 15% above 5 MeV.                              
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Covariance matrix was obtained by fitting to the experimental 
    data of nu-p.                                                 
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/31/ 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                            
    In the resolved resonance region, uncertainty of 5% was added 
    to the contributions from resonance parameter uncertainties.  
                                                                  
    Above 300 eV, covariances for CCONE calculation were adopted. 
                                                                  
 MT=18 Fission cross section                                      
    Error of 2 % was assumed in the resolved resonance region up  
    to 300 eV.                                                    
                                                                  
    In the energy range from 300 eV to 10 keV, the fission cross  
    section was determined from the experimetal data of Gerassimov
    et al./19/  Error of 5% was assumed in the region up to       
    9 keV.                                                        
                                                                  
    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                                    
    Error of 10 % was assumed in the resolved resonance region.   
                                                                  
    In the energy region from 300 eV to 2 keV, error of 15% was   
    assumed, and from 2 to 30 keV, error of 10%, by considering   
    dispersion of average cross sections of Weston and Todd/17/.  
                                                                  
    Above 30 keV, Covariance matrix was obtained with CCONE and   
    KALMAN codes/31/.                                             
                                                                  
                                                                  
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/32/                              
    * Global parametrization of Koning-Duijvestijn/33/            
      was used.                                                   
    * Gamma emission channel/34/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/35/ 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/36/. 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/37/,/38/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,2,3,9 (see Table 2)                      
  * optical potential parameters /20/                             
    Volume:                                                       
      V_0       = 49.682   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       = 17.45    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    = 0.24731                                         
      beta_4    = 0.05852                                         
      beta_6    = -0.02486                                        
                                                                  
  * Calculated strength function                                  
    S0= 1.19e-4 S1= 2.62e-4 R'=  9.47 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of Pu-241                                   
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000  5/2 +  *                                           
   1  0.04197  7/2 +  *                                           
   2  0.09578  9/2 +  *                                           
   3  0.16131 11/2 +  *                                           
   4  0.16168  1/2 +                                              
   5  0.17094  3/2 +                                              
   6  0.17505  7/2 +                                              
   7  0.22299  5/2 +                                              
   8  0.23194  9/2 +                                              
   9  0.23500 13/2 +  *                                           
  10  0.24489  7/2 +                                              
  11  0.30117 11/2 +                                              
  12  0.33700  1/2 +                                              
  13  0.33714  9/2 +                                              
  14  0.37300 11/2 +                                              
  15  0.37600  1/2 -                                              
  16  0.38500 13/2 +                                              
  17  0.40445  9/2 -                                              
  18  0.40890  7/2 -                                              
  19  0.44600 11/2 -                                              
  20  0.47300  1/2 -                                              
  21  0.49500  7/2 +                                              
  22  0.50300 13/2 +                                              
  23  0.51881  5/2 -                                              
  24  0.53420  7/2 +                                              
  25  0.56142  7/2 -                                              
  26  0.57000 15/2 -                                              
  27  0.61484  9/2 -                                              
  28  0.64500 13/2 -                                              
  29  0.68100  1/2 -                                              
  30  0.75517  1/2 +                                              
  31  0.76927  1/2 -                                              
  32  0.77915  3/2 -                                              
  33  0.78415  3/2 +                                              
  34  0.80044  3/2 +                                              
  35  0.80048  5/2 +                                              
  36  0.81095  5/2 -                                              
  37  0.83159  5/2 +                                              
  38  0.83330  7/2 -                                              
  39  0.83484  3/2 +                                              
  40  0.84196  1/2 -                                              
  -------------------                                             
  *) Coupled levels in CC calculation                             
                                                                  
Table 3. Level density parameters                                 
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Pu-242 18.5806  1.5428  2.4520  0.3710  0.0263  3.6230         
   Pu-241 19.1301  0.7730  2.1853  0.3372 -0.4308  2.4675         
   Pu-240 18.5012  1.5492  2.1440  0.3871 -0.0917  3.7899         
   Pu-239 18.4349  0.7762  1.8503  0.3560 -0.5001  2.5655         
   Pu-238 18.3685  1.5557  1.9652  0.3804  0.0287  3.6608         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
   Pu-242   6.172   0.998   4.674   0.600                         
   Pu-241   5.795   0.580   5.487   0.520                         
   Pu-240   6.250   1.040   4.920   0.600                         
   Pu-239   6.050   0.700   5.700   0.600                         
   Pu-238   6.000   1.040   4.800   0.600                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Pu-242 20.4971  1.7999  2.6000  0.3433 -0.8376  3.9999         
   Pu-241 20.4242  0.9018  2.6000  0.3647 -2.0579  3.4018         
   Pu-240 20.3513  1.8074  2.6000  0.3300 -0.6156  3.8074         
   Pu-239 20.2784  0.9056  2.6000  0.3523 -1.8394  3.2056         
   Pu-238 20.2054  1.8150  2.6000  0.3313 -0.6081  3.8150         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Pu-242 20.4971  1.7999  0.5000  0.3933 -0.2466  4.1999         
   Pu-241 20.4242  0.9018  0.4600  0.3804 -0.9744  3.1018         
   Pu-240 20.5363  1.8074  0.4200  0.3796 -0.0661  4.0074         
   Pu-239 20.2784  0.9056  0.3800  0.3901 -1.0534  3.2056         
   Pu-238 20.2054  1.8150  0.3400  0.3694  0.1086  3.8150         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for Pu-242                   
  --------------------------------------------------------        
  K0 = 2.481   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 =  6.58 (MeV) EG = 4.00 (MeV) SIG =   3.87 (mb)        
  * E2: ER = 10.11 (MeV) EG = 3.21 (MeV) SIG =   6.78 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
References                                                        
 1) O.Iwamoto et al.: J. Nucl. Sci. Technol., 46, 510 (2009).     
 2) O.Iwamoto: J. Nucl. Sci. Technol., 44, 687 (2007).            
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 4) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989).     
 5) H.Conde et al.: J. Nucl. Energy, 22, 53 (1968).               
 6) J.Frehaut et al.: CEA-R-4626 (1974).                          
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 8) N.E.Holden, M.S.Zucker: Nucl. Sci. Eng., 98, 174 (1988).      
 9) A.S.Vorobyev et al.: 2004 Santa Fe, Vol.1, p.613 (2004).      
10) G.Audi: Private communication (April 2009).                   
11) J.Katakura et al.: JAERI 1343 (2001).                         
12) T.R.England et al.: LA-11151-MS (1988).                       
13) R.Sher, C.Beck: EPRI NP-1771 (1981).                          
14) H.Derrien et al.: Nucl. Sci. Eng., 150, 109 (2005).           
15) M.B.Chadwick et al.: Nucl. Data Sheets, 107, 2931 (2006).     
16) Y.Kikuchi et al.: JAERI-Data/Code 99-025 (1999) in Japanese.  
17) L.W.Weston, J.H.Todd: Nucl. Sci. Eng., 65, 454 (1978).        
18) E.Migneco et al.: 1970 Helsinki, p.437 (1970).                
19) V.F.Gerasimov et al.: JINR-E3-97-213, p.348 (1997).           
20) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).     
21) T.Kawano et al.: JAERI-Research 2000-004 (2000).              
22) N.A.Khan et al.: Nucl. Instrum. Methods, 173, 137 (1980).     
23) G.W.Carlson et al.: Nucl. Sci. Eng., 63, 149 (1977).          
24) I.Szabo et al.: 1976 ANL, p.208 (1976).                       
25) J.Blons et al.: 1971 Knoxville, Vol.2, p.836 (1971).          
26) M.I.Kazarinova et al.: Sov. J. At. Energy, 8, 125 (1961).     
27) B.I.Fursov et al.: At. Energy, 44, 236 (1978).                
28) J.W.Behrens et al.: Nucl. Sci. Eng., 68, 128 (1978).          
29) F.Kaeppeler et al.: Nucl. Sci. Eng., 51, 124 (1973).          
30) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).            
31) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in         
    Japanese.                                                     
32) C.Kalbach: Phys. Rev. C33, 818 (1986).                        
33) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).     
34) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).     
35) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).                   
36) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).            
37) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990).                
38) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).     
                                                                  
                                                                  
Appendix A-1: Resonance parameters                                
----------------------------------------------------------------  
  Reevaluation of rhe resonance parameters in the energy range up 
to 20 eV - Colaboration ORNL(USA)-CEN/CAD(FRANCE)                 
  H. Derrien and L.C Leal,  ORNL                                  
  A. Courcelle and A. Santamarina,  CEN/CAD                       
(submitted to NSE, 2003)                                          
                                                                  
                                                                  
Cross section integrals in the energy range 0.0021 eV to 20.0 eV  
                                                                  
   -------------------------------------------------------------  
   Energy Range       Capture b-eV             Fission b-eV       
        eV          1993  Present             1993  Present       
   -------------------------------------------------------------  
   0.0021-0.020    12.25   12.51  2.1%       31.06   31.43  1.2%  
   0.0200-0.030     3.67    3.72  1.4%       10.24   10.36  1.2%  
   0.0300-0.100    15.28   15.59  2.0%       49.02   49.14  0.2%  
   0.1000-0.500   110.58  117.40  6.2%      262.76  263.84  0.4%  
   0.5000-1.000     5.90    6.03  2.2%       17.93   17.89 -0.2%  
   1.0000-3.000     7.30    7.16 -2.0%       54.88   52.89 -3.8%  
   3.0000-20.00  1213.   1234.    1.7%     3039.   3026.   -0.4%  
   -------------------------------------------------------------  
                                                                  
The cross sections at 0.0253 eV are the following:                
                                                                  
                   Total   1386.5 b                               
                   Fission 1012.2 b                               
                   Capture  363.0 b                               
                                                                  
very close to the current standard.                               
                                                                  
The fit of the experimental data (Young total, Wagemans fission,  
Weston capture and Weston fission ) were performed by assuming    
that young total and Wagemans fission were OK with the standard   
and that Weston original data(from the EXFOR file)needed a renor- 
zation to agree with the standard. A consistent normalization     
was obtained by a SAMMY fit leading to a decrease of 4.5%         
in the original Weston fission and a decrease of 3.0% in the      
original Weston capture.The differences between the 1988, 1993    
and the present results are consistent with the experimental      
errors given by Weston for the measured fission (2-3%) and        
capture (10-15%). The default of normalization of the Weston      
fission,due to severe experimental errors in the thermal range,   
was responsible for the large uncertainty in the capture results.