94-Pu-242

 94-Pu-242 JAEA+      EVAL-JAN10 O.Iwamoto, T.Nakagawa, Murata, + 
                      DIST-MAY10                       20100323   
----JENDL-4.0         MATERIAL 9446                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
History                                                           
06-08 Nu-p was revised.                                           
06-09 Resonance parameters were revised.                          
06-12 Fission cross section was revised.                          
07-05 Data were calculated with CCONE code.                       
      Data were compiled as JENDL/AC-2008/1/.                     
09-03 (1,452), (1,455) and (1,456) were revised.                  
09-08 (MF1,MT458) was evaluated.                                  
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=455 and 456                                         
                                                                  
  MT=455 Delayed neutrons                                         
    Determined from nu-d of the following three nuclides and      
    partial fission cross sections calculated with CCONE code/2/. 
                                                                  
      Pu-243 = 0.0153   an average of experimental data of Krick  
                        and Evans /3, 4/                          
      Pu-242 = 0.011    0.00160 /5/ was multiplied by 0.7.        
      Pu-241 = 0.0064   0.00911 /5/ was multiplied by 0.7.        
                                                                  
      Values for Pu-242 and 241 were multiplied by 0.7 to         
      reproduce the nu-d of 0.0113+-0.0009 at 14.7MeV/6/          
                                                                  
    Decay constants were evaluated by Brady and England/7/.       
                                                                  
  MT=456 Number of prompt neutrons per fission                    
    Least-squares fitting of a straight line to the experimental  
    data of Khokhlov et al./8/ Since their data were total        
    numbers of neutrons per fission, numbers of delayed neutrons  
    (MT=455) were subtracted.                                     
                                                                  
       nu-p = 2.8779 + 0.13755*E(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/9/. 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 2000/10/ does 
    not include average neutron energy values, the average values 
    were calculated using the formula shown in the report by      
    T.R. England/11/. The fractions of prompt energy were         
    calculated using the fractions of Sher's evaluation/12/ when  
    they were provided. When the fractions were not given by Sher,
    averaged fractions were used.                                 
                                                                  
                                                                  
MF= 2 Resonance parameters                                        
  MT=151                                                          
  Resolved resonance parameters (below 1 keV)                     
    Resonance parameters of JENDL-3.3 were modified:              
      * Upper boundary was decreased from 1.9 keV to 1 keV.       
      * Capture width of 2.67-eV resonances was changed from 22   
        meV to 26.8 meV.                                          
      * Fission width of 53.46-eV resonance was increased from    
        1.825 micro-eV to 36 micro-eV.                            
                                                                  
    Thermal capture cross section of 19.98+-0.66 b to be          
    reproduced was determined from Butler et al./13/, Durham      
    and Molson/14/ and Marie et al./15/.                          
                                                                  
  Unresolved resonance parameters (1 keV - 100 keV)               
    Parameters were estimated with ASREP code/16/ so as to        
    total, fission and capture cross sections in this energy      
    region. They are used only for self-shielding calculations.   
                                                                  
                                                                  
     Thermal cross sections and resonance integrals (at 300K)     
    -------------------------------------------------------       
                    0.0253 eV    reson. integ.(*)                 
                     (barns)       (barns)                        
    -------------------------------------------------------       
    total            28.213                                       
    elastic           8.326                                       
    fission           0.00244         4.36                        
    capture          19.885        1130                           
   -------------------------------------------------------        
         (*) In the energy range from 0.5 eV to 10 MeV.           
                                                                  
                                                                  
MF= 3 Neutron cross sections                                      
  Cross sections above the resolved resonance region except for   
  the elastic scattering (MT=2) and fission cross sections (MT=18,
  19, 20, 21, 38) were calculated with CCONE code/2/.             
                                                                  
  MT= 1 Total cross section                                       
    Calculated with CCONE code and modified below 500 keV by      
    multiplying an energy-dependent factor so as to reproduce     
    average total cross sections obtained from the data of Young  
    et al. /17/ The calculation was made with CC OMP of           
    Soukhovitskii et al./18/                                      
                                                                  
  MT= 2 Elastic scattering cross section                          
    Calculated as total cross section - sum of partial cross      
    sections.                                                     
                                                                  
  MT=18 Fission cross section                                     
    The following experimental data were analyzed in the energy   
    range from 1 keV to 20 MeV with the GMA code/19/:             
                                                                  
       Authors        Energy range     Data points  Reference     
       Bulter         141 keV - 1.66 MeV     65      /20/         
       Fomushkin+     14.5 MeV                1      /21/         
       Bergen+        0.1 - 2.96 MeV        141      /22/(*1)     
       Auchampaugh+   0.95 keV - 3.99 MeV  3102      /23/         
       Meadows        0.397 - 9.92 MeV       49      /24/(*2)     
       Behrens        97.2 keV - 20.0 MeV   133      /25/(*2)     
       Kuprijanov+    0.127 - 7.4 MeV        71      /26/(*2)     
       Cance+         2.47 MeV                2      /27/         
       Alkhazov+      14.7 MeV                1      /28/         
       Weigmann+      0.3 - 9.7 MeV         222      /29/         
       Arlt+          14.7 MeV                1      /30/         
       Meadows        14.7 MeV                1      /31/(*2)     
       Iwasaki+       0.597 - 6.76 MeV       17      /32/(*2)     
       Staples        0.514 - 19.5 MeV      124      /33/(*2)     
                                                                  
        *1) only the data above 100 keV were used.                
        *2) ratio to U-235 fission cross section                  
                                                                  
    The results of GMA were used to determine the parameters in   
    the CCONE calculation.                                        
                                                                  
  MT=19, 20, 21, 38 Multi-chance fission cross sections           
    Calculated with CCONE code, and renormalized to the total     
    fission cross section (MT=18).                                
                                                                  
  MT=102 Capture cross section                                    
    Calculated with CCONE code. The experimental data of Wisshak  
    and Kaeppeler /34,35/ and Hockenbury et al./36/ were used to  
    determine the parameters in the CCONE calculation.            
                                                                  
                                                                  
MF= 4 Angular distributions of secondary neutrons                 
  MT=2 Elastic scattering                                         
    Calculated with CCONE code.                                   
                                                                  
  MT=18 Fission                                                   
    Isotropic distributions in the laboratory system were assumed.
                                                                  
                                                                  
MF= 5 Energy distributions of secondary neutrons                  
  MT=18  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/7/ 
    were adopted.                                                 
                                                                  
MF= 6 Energy-angle distributions                                  
    Calculated with CCONE code.                                   
    Distributions from fission (MT=18) are not included.          
                                                                  
                                                                  
MF=12 Photon production multiplicities                            
  MT=18 Fission                                                   
    Calculated from the total energy released by the prompt       
    gamma-rays due to fission given in MF=1/MT=458 and the        
    average energy of gamma-rays.                                 
                                                                  
                                                                  
MF=14 Photon angular distributions                                
  MT=18 Fission                                                   
    Isotoropic distributions were assumed.                        
                                                                  
                                                                  
MF=15 Continuous photon energy spectra                            
  MT=18 Fission                                                   
    Experimental data measured by Verbinski et al./37/ for        
    Pu-239 thermal fission were adopted.                          
                                                                  
                                                                  
MF=31 Covariances of average number of neutrons per fission       
  MT=452 Number of neutrons per fission                           
    Combination of covariances for MT=455 and MT=456.             
                                                                  
  MT=455                                                          
    Error of 10% was assumed below 5 MeV and above 5 MeV,         
    respectively by comparing with experimental data/4, 6/        
                                                                  
  MT=456                                                          
    Covariance was obtained by fitting a linear function to the   
    experimental data of Khokhlov et al./8/(see MF1,MT456).       
    Variances were multiplied by a factor of 2.                   
                                                                  
                                                                  
MF=32 Covariances of resonance parameters                         
    Format of LCOMP=0 was adopted.                                
                                                                  
    Standard deviations of resonance energy, neutron and capture  
    widths were taken from Mughabghab /38/ Those of fission       
    width were based on the data of fission area reported by      
    Weigmann et al./29/ and Auchampaugh et al./23/.               
                                                                  
    If no information was available, uncertainties were assumed.  
                                                                  
                                                                  
MF=33 Covariances of neutron cross sections                       
  Covariances were given to all the cross sections by using       
  KALMAN code/39/ and the covariances of model parameters         
  used in the theoretical calculations.                           
                                                                  
  For the following cross sections, covariances were determined   
  by different methods.                                           
                                                                  
  MT=1, 2 Total and elastic scattering cross sections             
    In the resonance region (below 1 keV), uncertainty of 8 %     
    was added.                                                    
                                                                  
    Above 1 keV, covariance matrix was obtained with CCONE and    
    KALMAN codes/39/.                                             
                                                                  
  MT=18 Fission cross section                                     
    In the resonance region from 10 to 1000 eV, addtional error   
    of 50% was given.                                             
                                                                  
    Above the resonance region, cross section was evaluated with  
    GMA code/19/. Standard deviation obatianed was multiplied     
    by a factor of 2.0.                                           
                                                                  
  MT=102 Capture cross section                                    
    In the resonance region from 10 to 1000 eV, addtional error   
    of 10% was given.                                             
                                                                  
    Above 1 keV, covariance matrix was obtained with CCONE and    
    KALMAN codes/39/.                                             
                                                                  
                                                                  
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/40/                              
    * Global parametrization of Koning-Duijvestijn/41/            
      was used.                                                   
    * Gamma emission channel/42/ was added to simulate direct     
      and semi-direct capture reaction.                           
                                                                  
  3) Hauser-Feshbach statistical model                            
    * Moldauer width fluctuation correction/43/ 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/44/. 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/45/,/46/       
      was used. The prameters are shown in Table 7.               
                                                                  
                                                                  
------------------------------------------------------------------
                              Tables                              
------------------------------------------------------------------
                                                                  
Table 1. Coupled channel calculation                              
  --------------------------------------------------              
  * rigid rotor model was applied                                 
  * coupled levels = 0,1,2,3,4 (see Table 2)                      
  * optical potential parameters /18/                             
    Volume:                                                       
      V_0       = 49.97    MeV                                    
      lambda_HF = 0.01004  1/MeV                                  
      C_viso    = 15.9     MeV                                    
      A_v       = 12.04    MeV                                    
      B_v       = 81.36    MeV                                    
      E_a       = 385      MeV                                    
      r_v       = 1.2568   fm                                     
      a_v       = 0.633    fm                                     
    Surface:                                                      
      W_0       = 17.2     MeV                                    
      B_s       = 11.19    MeV                                    
      C_s       = 0.01361  1/MeV                                  
      C_wiso    = 23.5     MeV                                    
      r_s       = 1.1803   fm                                     
      a_s       = 0.601    fm                                     
    Spin-orbit:                                                   
      V_so      = 5.75     MeV                                    
      lambda_so = 0.005    1/MeV                                  
      W_so      = -3.1     MeV                                    
      B_so      = 160      MeV                                    
      r_so      = 1.1214   fm                                     
      a_so      = 0.59     fm                                     
    Coulomb:                                                      
      C_coul    = 1.3                                             
      r_c       = 1.2452   fm                                     
      a_c       = 0.545    fm                                     
    Deformation:                                                  
      beta_2    = 0.23892                                         
      beta_4    = 0.04807                                         
      beta_6    = -0.02376                                        
                                                                  
  * Calculated strength function                                  
    S0= 0.98e-4 S1= 2.93e-4 R'=  9.32 fm (En=1 keV)               
  --------------------------------------------------              
                                                                  
Table 2. Level Scheme of Pu-242                                   
  -------------------                                             
  No.  Ex(MeV)   J PI                                             
  -------------------                                             
   0  0.00000   0  +  *                                           
   1  0.04454   2  +  *                                           
   2  0.14730   4  +  *                                           
   3  0.30640   6  +  *                                           
   4  0.51810   8  +  *                                           
   5  0.77860  10  +                                              
   6  0.78045   1  -                                              
   7  0.83230   3  -                                              
   8  0.86500   3  +                                              
   9  0.92700   5  -                                              
  10  0.95600   0  +                                              
  11  0.99250   2  +                                              
  12  1.01950   3  -                                              
  13  1.03920   1  +                                              
  14  1.06400   4  -                                              
  15  1.08440  12  +                                              
  16  1.09210   6  +                                              
  17  1.10200   2  +                                              
  18  1.12200   5  -                                              
  19  1.15100   2  -                                              
  20  1.15450   3  -                                              
  -------------------                                             
  *) 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-243 18.6999  0.7698  2.4578  0.3280 -0.3352  2.3315         
   Pu-242 18.6337  1.5428  2.4520  0.3701  0.0291  3.6198         
   Pu-241 18.5675  0.7730  2.1853  0.3473 -0.4715  2.5167         
   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         
  --------------------------------------------------------        
                                                                  
Table 4. Fission barrier parameters                               
  ----------------------------------------                        
  Nuclide     V_A    hw_A     V_B    hw_B                         
              MeV     MeV     MeV     MeV                         
  ----------------------------------------                        
   Pu-243   5.750   0.680   5.520   0.520                         
   Pu-242   6.100   1.000   4.850   0.600                         
   Pu-241   5.950   0.580   5.480   0.520                         
   Pu-240   6.250   1.040   4.920   0.600                         
   Pu-239   6.050   0.700   5.700   0.600                         
  ----------------------------------------                        
                                                                  
Table 5. Level density above inner saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Pu-243 20.5699  0.8981  2.6000  0.3633 -2.0616  3.3981         
   Pu-242 20.4971  1.7999  2.6000  0.3503 -0.9450  4.0999         
   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         
  --------------------------------------------------------        
                                                                  
Table 6. Level density above outer saddle                         
  --------------------------------------------------------        
  Nuclide      a*    Pair  Eshell       T      E0  Ematch         
            1/MeV     MeV     MeV     MeV     MeV     MeV         
  --------------------------------------------------------        
   Pu-243 20.9439  0.8981  0.5400  0.3740 -0.9758  3.0981         
   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         
  --------------------------------------------------------        
                                                                  
Table 7. Gamma-ray strength function for Pu-243                   
  --------------------------------------------------------        
  K0 = 2.100   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 =  6.57 (MeV) EG = 4.00 (MeV) SIG =   3.39 (mb)        
  * E2: ER = 10.10 (MeV) EG = 3.19 (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).            
 3) M.S.Krick, A.E.Evans: Nucl. Sci. Eng., 47, 311 (1972).        
 4) A.E.Evans  et al.: Nucl. Sci. Eng., 50, 80 (1973).            
 5) G.Benedetti et al.: Nucl. Sci. Eng., 80, 379 (1982).          
 6) E.Yu.Bobkov et al.: At. Energy, 67, 408 (1989).               
 7) M.C.Brady, T.R.England: Nucl. Sci. Eng., 103, 129 (1989).     
 8) Yu.A.Khokhlov et al.: 1994 Gatlinburg, Vol.1, p.272 (1994).   
 9) G.Audi: Private communication (April 2009).                   
10) J.Katakura et al.: JAERI 1343 (2001).                         
11) T.R.England et al.: LA-11151-MS (1988).                       
12) R.Sher, C.Beck: EPRI NP-1771 (1981).                          
13) J.P.Butter et al.: Canadian J. Phys., 35, 147 (1957).         
14) R.W.Durham, F.Molson: Canadian J. Phys., 48, 716 (1970).      
15) F.Marie et al.: Nucl. Instrum. Methods, A556, 547 (2006).     
16) Y.Kikuchi et al.: JAERI-Data/Code 99-025 (1999) in Japanese.  
17) T.E.Young et al.: Nucl. Sci. Eng., 43, 341 (1971).            
18) E.Sh.Soukhovitskii et al.: Phys. Rev. C72, 024604 (2005).     
19) W.P.Poenitz: BNL-NCS-51363, Vol.I, p.249 (1981).              
    S.Chiba, D.L.Smith: ANL/NDM-121 (1991).                       
20) D.K.Butler: Phys. Rev., 117, 1305 (1960).                     
21) E.F.Fomushkin et al.: Sov. J. Nucl. Phys., 5, 689 (1967).     
22) D.W.Bergen, R.R.Fullwood: Nucl. Phys., A163, 577 (1971).      
23) G.F.Auchampaugh et al.: Nucl. Phys., A171, 31 (1971).         
24) J.W.Meadows: Nucl. Sci. Eng., 68, 360 (1978).                 
25) J.W.Behrens et al.: Nucl. Sci. Eng., 66, 433 (1978).          
26) V.M.Kupriyanov et al.: Sov. Atomic Energy, 46, 35 (1979).     
27) M.Cance, G.Grenier: 1982 Antwerp, p.51 (1982).                
28) I.D.Alkhazov et al.: 1983 Moskova, vol.2, p.201 (1983).       
29) H.Weigmann et al.: Nucl. Phys., A438, 333 (1985).             
30) R.A.Arlt et al.: KE, 24, 48 (1981).                           
31) J.W.Meadows: Ann. Nucl. Energy, 15, 421 (1988).               
32) T.Iwasaki et al.: J. Nucl. Sci. Technol., 27, 885 (1990).     
33) P.Staples, K.Morley: Nucl. Sci. Eng., 129, 149 (1998).        
34) K.Wisshak, F.Kaeppeler: Nucl. Sci. Eng., 69, 39 (1979).       
35) K.Wisshak, F.Kaeppeler: Nucl. Sci. Eng., 66, 363 (1978).      
36) R.W.Hockenbury et al.: 1975 Washington, Vol.2, p.584(1975).   
37) V.V.Verbinski et al.: Phys. Rev., C7, 1173 (1973).            
38) S.F.Mughabghab: "Atlas of Neutron Resonances," Elsevier       
    (2006).                                                       
39) T.Kawano, K.Shibata, JAERI-Data/Code 97-037 (1997) in         
    Japanese.                                                     
40) C.Kalbach: Phys. Rev. C33, 818 (1986).                        
41) A.J.Koning, M.C.Duijvestijn: Nucl. Phys. A744, 15 (2004).     
42) J.M.Akkermans, H.Gruppelaar: Phys. Lett. 157B, 95 (1985).     
43) P.A.Moldauer: Nucl. Phys. A344, 185 (1980).                   
44) D.L.Hill, J.A.Wheeler: Phys. Rev. 89, 1102 (1953).            
45) J.Kopecky, M.Uhl: Phys. Rev. C41, 1941 (1990).                
46) J.Kopecky, M.Uhl, R.E.Chrien: Phys. Rev. C47, 312 (1990).