92-U -238

 92-U -238 JAEA+      EVAL-NOV09 O.Iwamoto,N.Otuka,S.Chiba,+      
                      DIST-MAY10                       20100311   
----JENDL-4.0         MATERIAL 9237                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
06-07 (n,2n) cross section was revised.                           
06-10 nu-p was revised.                                           
07-04 Calculation with CCONE code was performed.                  
07-06 Fission spectra above 5.5 MeV were revised.                 
07-11 Fission cross section was revised with simultaneous         
      evaluation.                                                 
07-12 Fission cross section was revised with results of new       
      simultaneous evaluation. Unresolved resonance parameters    
      were revised.                                               
08-01 Fission cross section was revised.                          
08-03 Fission and capture cross sections, and nu-p were revised.  
      CCONE calculation was made with revised parameters.         
      Interpolation of (5,18) was changed.                        
      Data were compiled as JENDL/AC-2008/1/.                     
09-08 (MF1,MT458) was evaluated.                                  
09-10 fission cross section was modified slightly.                
09-11 Results of new CCONE calculation were adopted.              
10-03 Covariance 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/)                                        
    Experimental data of Krick and Evans /3/ were renormalized    
    to those of Meadows /4/, and the least-squares fitting was    
    carried out with the SOK code /5/.                            
    Decay constants were adopted from Keepin et al. /6/           
                                                                  
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Experimental data reported after 1960 were considered:        
      Butler et al./7/, Conde et al./8/, Asplund-Nilsson/9/,      
      Mather et al./10/, Vorob'jova et al./11/, Bao/12/,          
      Nurpeisov et al./13/, Frehaut et al./14,15/, Malinovskij et 
      al./16/, Boykov et al./17/, Smirenkin et al./18/.           
                                                                  
    Cf-252 nu-p of 3.756 /19/ was used. These experimental data   
    were fitted with two straight lines below and above 14 MeV.   
                                                                  
  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/20/. 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 2000/21/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/22/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/23/ 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 - 20 keV)             
    The data of ENDF/B-VII/24/ were adopted. The parameters were  
    analyzed by Derrien et al./25/ with SAMMY code /26/ and       
    experimental data up to 20 keV/27,28,29,30,31,etc./           
    The thermal capture cross section was adjusted to 2.683 b     
    evaluated by Trkov et al./32/                                 
                                                                  
  Unresolved resonance parameters (20 keV - 150 keV)              
    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             11.983                                      
    elastic            9.300                                      
    fission            1.68E-5          1.24                      
    capture            2.683          276                         
    -------------------------------------------------------       
         (*) 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 cross (MT=18, 19, 20,    
  21, 38) and capture cross sections were calculated with CCONE   
  code/33/.                                                       
                                                                  
  Model parameters for CCONE code were determined by considering  
  experimental data of total and (n,2n) cross sections, and       
  fission and capture cross sections of JENDL-3.3. OMPs were      
  based on those of Kunieda et al./34/, and adjusted to the total 
  cross section. The model parameters were further adjusted by    
  considering integral data.                                      
                                                                  
  MT= 1 Total cross section                                       
    Below about 12 MeV, cross section was calculated with CCONE   
    code. Above 12 MeV, a smooth cross-section curve was obtained 
    by spline fitting to the experimental data of Abfalterer et   
    al./35/                                                       
                                                                  
  MT=2 Elastic scattering cross section                           
    Calculated as total - non-elastic scattering cross sections   
    except for the 1.3-4.0MeV region where the differences among  
    adopted fission and calculated fission were reflected in the  
    inelastic scattering cross sections.                          
                                                                  
  MT=16 (n,2n) cross section                                      
    Calculated with CCONE code. Following experimental data were  
    considered for the determination of model parameters:         
      Frehaut et al./36,37/, Kornilov et al./38/, Karius          
      et al./39/, Raics et al./40/, Golovnya et al./41/,          
      Konno et al./42/, Filatenkov et al./43/, Veeser et al./44/, 
      Ryves et al./45/, Pepelnik et al./46/.                      
    The data of Frehaut et al. were multiplied by a factor of 1.1.
    The data of activation measurements were re-normalized by     
    adopting a intensity of 21.2% to the 208-keV gamma-rays from  
    Np-237.                                                       
                                                                  
  MT=18 Fission cross section                                     
    Below 400 keV, JENDL-3.3/2/ was adopted.                      
    Above 400 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    
    ratio by the SOK code/5/. Covariance matrix reported in       
    Manabe et al./47/ was also considered in the analysis.        
                                                                  
    --------------------------------------------------------------
     Cross section                                                
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        13586.011  1.00E+7            J.W.Meadows+      /48/      
        22304.003  4.80E+6 - 1.88E+7  K.Merla+          /49/      
        30669.002  4.00E+6 - 5.50E+6  J.X.Wu+           /50/      
        20779.003  1.39E+7 - 1.46E+7  M.Cance+          /51/      
        40547.007  1.48E+7            V.M.Adamov+       /52/      
        40483.002  1.60E+5 - 1.55E+6  P.E.Vorotnikov+   /53/      
        40081.002  2.50E+6            I.M.Kuks+         /54/      
        21209.002  1.27E+7 - 1.94E+7  B.Adams+          /55/      
        22565.002  1.45E+7            G,Winkler+        /56/      
    --------------------------------------------------------------
                                                                  
     Ratio to U-233(n,f) cross section                            
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        10422.006  1.00E+6 - 2.85E+7  J.W.Behrens+      /57/      
    --------------------------------------------------------------
                                                                  
     Ratio to U-235(n,f) cross section                            
    --------------------------------------------------------------
        EXFOR      Energy range (eV)  Authors           Reference 
    --------------------------------------------------------------
        10635.002  1.51E+5 - 2.39E+7  F.C.Difilippo+    /58/      
        41455.003  5.77E+5 - 2.94E+7  O.A.Shcherbakov+  /59/      
        30722.002  1.47E+7            J.W.Li+           /60/      
        13134.007  1.47E+7            J.W.Meadows       /61/      
        40831.004  1.38E+7 - 1.48E+7  A.A.Goverdovskij+ /62/      
        40831.003  5.44E+6 - 1.04E+7  A.A.Goverdovskij+ /62/      
        30588.002  1.35E+7 - 1.48E+7  M.Varnagy+        /63/      
        40506.002  9.81E+5 - 7.00E+6  B.I.Fursov+       /64/      
        10653.004  1.44E+5 - 2.92E+7  J.W.Behrens+      /65/      
        20870.002  2.65E+6 - 7.01E+6  M.Cance+          /66/      
        20869.002  4.67E+6 - 8.85E+6  C.Nordborg+       /67/      
        20409.002  1.37E+6 - 2.96E+7  S.Cierjacks+      /68/      
        10506.002  5.33E+6 - 1.04E+7  J.W.Meadows       /69/      
        10504.002  1.09E+6 - 3.03E+6  J.W.Meadows       /70/      
        10237.003  8.99E+5 - 5.15E+6  J.W.Meadows       /71/      
        10232.006  2.00E+6 - 3.00E+6  W.P.Poenitz+      /72/      
        22282.006  1.35E+7 - 1.49E+7  F.Manabe+         /47/      
        -----.---  8.33E+5 - 2.96E+7  P.W.Lisowski+     /73/      
    --------------------------------------------------------------
                                                                  
    The cross section was slightly modified in the energy region  
    from 1 to 4 MeV and from 7 to 8 MeV.                          
                                                                  
  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                                    
    Based on JENDL-3.3/2/ data which were evaluated as follows:   
      Below 300 keV, evaluation was mainly based on the data      
      measured by Kazakov et al./74/.  Above 300 keV, data were   
      taken from JENDL-2 which were determined mainly from the    
      measurements by Poenitz/75/, Panitkin and Sherman/76/,      
      Moxon/77/, Fricke et al./78/ and Menlove and Poenitz/79/.   
      Above 1 MeV, statistical model calculation was made, and    
      direct and semi-direct capture cross section was calculated 
      with DSD code/80/.                                          
    They 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 neutrons                                   
    Below 5 MeV, data of JENDL-3.3/2/ were adopted.               
    Comment of JENDL-3.3:                                         
    * Distributions were calculated with the modified Madland-Nix 
      model/81,82/. The compound nucleus formation cross          
      sections for fission fragments (FF) were calculated using   
      Bechetti-Greenlees potential/83/. Up to 3rd-chance-fission  
      were considered at high incident neutron energies.          
        Parameters adopted:                                       
           Total average FF kinetic energy = 167.41 - 172.65 MeV  
           Average energy release          = 186.115 - 186.364 MeV
           Average mass number of light FF = 99 - 111             
           Average mass number of heavy FF = 128 - 140            
           Level density of the light FF   = 10.106 - 10.963      
           Level density of the heavy FF   = 11.441 - 7.811       
           Ratio of nuclear temperature                           
                     for light to heavy FF = 1.0                  
        Note that the parameters vary with the incident energy    
        within the indicated range.                               
                                                                  
    Above 5.5 MeV, the spectra calculated with CCONE code /33/    
    were adopted.                                                 
                                                                  
  MT=455 Delayed neutrons                                         
    (same as JENDL-3.3)                                           
    Taken from Brady and England/84/. Group abundace parameters   
    were adjusted so as to reproduce total delayed neutron        
    emission rate measured by Keepin/6/, and East et al./85/.     
                                                                  
                                                                  
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 of U-235   
    measured by Verbinski et al./86/ was adopted for the whole    
    energy region. The intensity of photon below 0.14 MeV, where  
    no data were given, was assumed to be the same as that        
    between 0.14 and 0.3 MeV.                                     
                                                                  
    Data were extended up to 20 MeV for JENDL-4.0.                
                                                                  
                                                                  
MF=14 Angular Distributions of Photons                            
   Isotropic distributions were assumed for all sections.         
                                                                  
                                                                  
MF=15 Continuous Photon Energy Spectra                            
  MT=18   Fission                                                 
    (same as JENDL-3.3)                                           
    U-235 spectra measured by Verbinski et al./86/.               
                                                                  
                                                                  
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                                                          
    Covariance was obtained by fitting to the experimental        
    data (see MF1,MT456).                                         
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Format of LCOMP=2 was adopted.                                
                                                                  
    The covariance matrix was given to the resonance parameters   
    up to 2 keV, and the energy range was set to 1.0e-5 eV to     
    1.5 keV. The covariance matrix of resonance parameters was    
    taken from ORNL evaluation /87,88/.                           
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/89/ 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 1.5 to 20 keV, uncertainty of 5% was 
    assumed. Above 20 keV, covariance of the CCONE calculation    
    was adopted.                                                  
                                                                  
  MT=18 Fission cross section                                     
    In the energy range below 400 keV, uncertainty of 80% was     
    assumed.                                                      
                                                                  
    Above 400 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 1.5 to 20 keV, uncertainty of 10%    
    was assumed. Above 20 keV, the covariance matrix was taken    
    from JENDL-3.3.                                               
                                                                  
                                                                  
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/33/ 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/90/                              
    * Global parametrization of Koning-Duijvestijn/91/            
      was used.                                                   
    * Gamma emission channel/92/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/93/ 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/94/. 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/95/,/96/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,2,3,4,7 (see Table 2)                    
  * optical potential parameters /97/                             
    Real:                                                         
      VRO       = -42.0786 MeV                                    
      VR1       = 0.027                                           
      VR2       = 0.00012  1/MeV                                  
      VR3       = 3.5e-07  1/MeV^2                                
      VRLA      = 96.2445  MeV                                    
      ALAVR     = 0.00386032 1/MeV                                
    Imaginary-surface:                                            
      WD0       = 0        MeV                                    
      WD1       = 0                                               
      WDA1      = 0                                               
      WDBW      = 14.5489  MeV                                    
      WDWID     = 12.0143  MeV                                    
      ALAWD     = 0.013353 1/MeV                                  
    Imaginary-volume:                                             
      WC0       = 0        MeV                                    
      WC1       = 0                                               
      WCA1      = 0                                               
      WCBW      = 17       MeV                                    
      WCWID     = 105      MeV                                    
      BNDC      = 0        MeV                                    
    Spin-orbit:                                                   
      VS        = 6.634    MeV                                    
      ALASO     = 0.005    1/MeV                                  
      WSO       = 0        MeV                                    
      WS1       = 0                                               
      WSBW      = -3.1     MeV                                    
      WSWID     = 160      MeV                                    
    Radius and diffuseness:                                       
      RR        = 1.23906  fm                                     
      RRBWC     = 0        fm                                     
      RRWID     = 0        MeV                                    
      PDIS      = 2                                               
      AR0       = 0.650409 fm                                     
      AR1       = 0        fm                                     
      RD        = 1.21933  fm                                     
      AD0       = 0.66086  fm                                     
      AD1       = 0        fm/MeV                                 
      RC        = 1.21     fm                                     
      AC0       = 0.685    fm                                     
      AC1       = 0        fm/MeV                                 
      RW        = 0        fm                                     
      AW0       = 0        fm                                     
      AW1       = 0        fm/MeV                                 
      RS        = 1.0751   fm                                     
      AS0       = 0.59     fm                                     
      AS1       = 0        fm/MeV                                 
      RZ        = 1.264    fm                                     
      RZBWC     = 0        fm                                     
      RRWID     = 0        MeV                                    
      AZ        = 0.341    fm                                     
      CCOUL     = 0.9      MeV                                    
      ALF       = 0                                               
    Coulomb correction:                                           
      CISO      = 24.3     MeV                                    
      WCISO     = 18       MeV                                    
    Deformation:                                                  
      beta_2    = 0.223632                                        
      beta_4    = 0.09                                            
      beta_6    = -0.0031                                         
                                                                  
  * Calculated strength function                                  
    S0= 1.10e-4 S1= 1.81e-4 R'=  9.48 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of U-238                                    
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.04492   2  +  *                                           
   2  0.14838   4  +  *                                           
   3  0.30718   6  +  *                                           
   4  0.51810   8  +  *                                           
   5  0.68011   1  -                                              
   6  0.73193   3  -                                              
   7  0.77590  10  +  *                                           
   8  0.82664   5  -                                              
   9  0.92721   0  +                                              
  10  0.93055   1  -                                              
  11  0.95012   2  -                                              
  12  0.96613   2  +                                              
  13  0.96631   7  -                                              
  14  0.99723   0  +                                              
  15  0.99758   3  -                                              
  16  1.02800   4  -                                              
  17  1.03725   2  +                                              
  18  1.05638   4  +                                              
  19  1.05773   3  +                                              
  20  1.05966   3  +                                              
  21  1.06027   2  +                                              
  22  1.07670  12  +                                              
  23  1.10571   3  +                                              
  24  1.12884   2  -                                              
  25  1.13075   4  +                                              
  26  1.13570   7  -                                              
  27  1.15070   9  -                                              
  -------------------                                             
  *) 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-239 17.1586  0.7762  3.0776  0.3946 -0.8805  3.0192         
    U-238 19.7315  1.5557  3.0121  0.3609 -0.0539  3.7504         
    U-237 18.2087  0.7795  2.7455  0.3666 -0.6759  2.7868         
    U-236 20.1615  1.5623  2.7551  0.3852 -0.3630  4.1437         
    U-235 18.5328  0.7828  2.6265  0.3874 -0.9246  3.1046         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
    U-239   6.452   0.702   5.574   0.474                         
    U-238   6.331   0.950   4.952   0.600                         
    U-237   6.036   0.650   5.641   0.500                         
    U-236   6.220   1.040   5.034   0.550                         
    U-235   5.790   0.400   5.470   0.300                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-239 20.6471  0.9056  2.6000  0.3418 -1.7200  3.1056         
    U-238 20.5728  1.8150  2.6000  0.3324 -0.6607  3.8750         
    U-237 20.4984  0.9094  2.6000  0.3286 -1.5020  2.9094         
    U-236 20.4241  1.8226  2.6000  0.3293 -0.5887  3.8226         
    U-235 20.3497  0.9133  2.6000  0.3299 -1.4981  2.9133         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
    U-239 22.3136  0.9056  0.2600  0.3564 -0.8646  3.0056         
    U-238 20.2054  1.8150  0.2200  0.4201 -0.4806  4.5150         
    U-237 21.5966  0.9094  0.1800  0.3569 -0.7807  2.9094         
    U-236 20.4241  1.8226  0.1400  0.3693  0.1220  3.8226         
    U-235 20.3497  0.9133  0.1000  0.3706 -0.7868  2.9133         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for  U-239                   
  --------------------------------------------------------        
  K0 = 2.946   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 =  6.61 (MeV) EG = 4.00 (MeV) SIG =   4.47 (mb)        
  * E2: ER = 10.15 (MeV) EG = 3.24 (MeV) SIG =   6.50 (mb)        
  --------------------------------------------------------        
                                                                  
                                                                  
References                                                        
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 2) T.Kawano et al.: JAERI-Research 2003-026 (2003).              
 3) M.S.Krick, A.E.Evans: Nucl. Sci. Eng., 47, 311 (1972).        
 4) J.W.Meadows: ANL-NDM-18 (1976).                               
 5) T.Kawano et al.: JAERI-Research 2000-004 (2000)               
 6) G.R.Keepin et al.: Phy. Rev., 107, 1044 (1957).               
 7) D.Butler et al : 1961 Vienna, p.125 (1961).                   
 8) H.Conde, N.Starfelt: Nucl. Sci. Eng., 11, 397 (1961).         
 9) I.Asplund-Nilsson et al.: Nucl. Sci. Eng., 20, 527 (1964).    
10) D.S.Mather et al.: Nucl. Phys., 66, 149 (1965).               
11) V.G.Vorob'jova et al.: YK-15, 3 (1974).                       
12) Z.Bao et al.: At. Energy Sci. Technol., 9, 362 (1975).        
13) B.Nurpeisov et al.: Sov. At. Energy, 39, 807 (1975).          
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