92-U -233

 92-U -233 SAEI+      Eval-Mar00 T.Mutsunobu, T.Kawano            
                      DIST-MAR02 Rev4-May00            20020214   
----JENDL-3.3         MATERIAL 9222                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
                                                                  
HISTORY                                                           
82-06 Evaluation for JENDL-2 was made by N. Asano (SAEI),         
      H. Matsunobu (SAEI) and Y.Kikuchi(JAERI).                   
87-03 Re-evaluation for JENDL-3 was made by H.Matsunobu (SAEI)    
      Main part of revision was the cross sections above 10 keV   
      and angular and energy distributions of neutrons.           
        Data were compiled by T. Nakagawa (JAERI).                
94-04 JENDL-3.2.                                                  
      Resonance parameters reevaluated by H.Derrien (JAERI)/1/    
      Cross sections       reevaluated by H.Matsunobu(SAEI)       
      Fission spectrum     reevaluated by T.Ohsawa(Kinki univ.)   
        Compiled by T.Nakagawa (NDC/JAERI)                        
                                                                  
     *****   Modified parts for JENDL-3.2   ********************  
      (2,151); New analysis with SAMMY                            
      MF=3   ; all data except fission cross section below 6.75   
               MeV and total cross section                        
      MF=4   ; for inelastic scattering                           
      (5,18)                                                      
     ***********************************************************  
                                                                  
00-02 JENDL-3.3. Evaluation was made by H. Matsunobu (SAEI)       
      and T.Kawano(Kyushu Univ.) and compiled by O.Iwamoto        
      (NDC/JAERI).                                                
                                                                  
     *****   Modified parts for JENDL-3.3   ********************  
     (1,452), (1,455), (1,456)                                    
     (2,151) unresolved resonance region                          
     (3,2), (3,16), (3,17), (3,18), (3,102)                       
     (5,16), (5,17), (5,91), (5,455)                              
     ***********************************************************  
                                                                  
02-01 Covariances were taken from JENDL-3.2 covariance file except
      for MF/MT=31/455, 31/456/, 33/18, and 33/102, which were    
      evaluated by Matsunobu./8/                                  
                                                                  
MF=2  Resonance Parameters                                        
  MT=151                                                          
                                                                  
     2200-m/s cross sections and calculated res. integrals        
                       2200 m/s      Res. Integ.                  
           total       588.38 b          -                        
           elastic      11.97 b          -                        
           fission     531.16 b         773 b                     
           capture      45.25 b         138 b                     
                                                                  
MF=3  Neutron Cross Sections                                      
 Smooth part (above 30 keV)                                       
                                                                  
  MT=18            Fission                                        
        Results of recent simultaneous evaluation of fission cross
        sections /2/ were adopted.                                
                                                                  
MF=5  Energy Distributions of Secondary Neutrons                  
  MT=16,17,91                                                     
      Calculated with pre-compound and multi-step evaporation     
      theory code EGNASH /3,4/.                                   
  MT=455  Delayed neutron spectrum.                               
      Summation calculation made by Brady and England /5/ was     
      adopted.                                                    
                                                                  
MF=31 Covariances of Average Number of Neutrons per Fission       
  MT=452                                                          
     Constructed from MT=455 and 456.                             
  MT=455                                                          
     Based on experimental data.  A chi-value was 0.13.           
  MT=456                                                          
     Based on experimental data.  A chi-value was 0.96.           
                                                                  
MF=32 Covariances of Resonance Paremeters (ref.6)                 
  MT=151                                                          
   Resolved resonance                                             
     Based on the SAMMY analysis by Derrien /1/.                  
   Unresolved resonance                                           
     The covariances were obtained by using kalman.               
                                                                  
MF=33 Covariances of Cross Sections (ref.6)                       
  MT=1                                                            
   Based on experimental data.  A chi-value was 0.244.            
  MT=2                                                            
   Constructed from mt=1, 4, 16, 17, 18, and 102.                 
  MT=4                                                            
   Based on experimental data.  A chi-value was 0.151.            
  MT=16                                                           
   Based on experimental data.                                    
  MT=17                                                           
   Taken from mt=16                                               
  MT=18                                                           
   The covariances were obtained by the simultaneous evaluation.  
   /2/                                                            
  MT=102                                                          
   Based on experimental data on alpha values.  A chi-value was   
   0.186.                                                         
                                                                  
MF=34 Covariances of Angular Distributions (ref.6)                
  MT=2                                                            
   The covariances of p1 coefficients were obtained by using      
   kalman.  A chi-value was 1.474.                                
                                                                  
MF=35 Covariances of Energy Distributions                         
  MT=18                                                           
   The covariances were obtained by using kalman./7/              
                                                                  
                                                                  
References                                                        
 1) Derrien H.: J. Nucl. Sci. Technol., 31, 379 (1994)            
 2) Kawano T. et al.: JAERI-Research 2000-004 (2000).             
 3) Yamamuro N.: JAERI-M 90-006 (1990).                           
 4) Young P.G. and Arthur E.D.: LA-6947 (1977).                   
 5) Brady M.C. and England T.R.: Nucl. Sci. Eng., 103, 129(1989). 
 6) Shibata K. and Hasegawa A.: JNC TJ9400 2000-004 (2000)        
    [in Japanese].                                                
 7) Kawano T. et al.: JAERI-Research 99-009 (1999) [in Japanese]. 
 8) Matsunobu H.: Private communication (2001).                   
                                                                  
                                                                  
******************** JEDL-3.2 ************************************
 92-U -233 SAEI+      Eval-Mar87 H.Matsunobu,Y.Kikuchi,T.Nakagawa 
                      Dist-Sep89 Rev2-Apr94                       
History                                                           
82-06 Evaluation for JENDL-2 was made by N. Asano (SAEI),         
      H. Matsunobu (SAEI) and Y.Kikuchi(JAERI).                   
87-03 Re-evaluation for JENDL-3 was made by H.Matsunobu (SAEI)    
      Main part of revision was the cross sections above 10 keV   
      and angular and energy distributions of neutrons.           
        Data were compiled by T. Nakagawa (JAERI).                
94-04 JENDL-3.2.                                                  
      Resonance parameters reevaluated by H.Derrien (JAERI)/1/    
      Cross sections       reevaluated by H.Matsunobu(SAEI)       
      Fission spectrum     reevaluated by T.Ohsawa(Kinki univ.)   
        Compiled by T.Nakagawa (NDC/JAERI)                        
                                                                  
     *****   Modified parts for JENDL-3.2   ********************  
      (2,151); New analysis with SAMMY                            
      MF=3   ; all data except fission cross section below 6.75   
               MeV and total cross section                        
      MF=4   ; for inelastic scattering                           
      (5,18)                                                      
     ***********************************************************  
                                                                  
                                                                  
MF=1  General Information                                         
  MT=451  Comments and dictionary                                 
  MT=452  Nu-total                                                
      Sum of Nu-d and Nu-p                                        
  MT=455  Nu-d                                                    
      Below 4 MeV                                                 
         Nu-d = 0.0075094 + 4.627E-5*ln(E(MeV))                   
      Between 4 and 20 MeV                                        
         Based on the data of Masters et al. /2/ and Evans et al. 
         /3/.                                                     
  MT=456  Nu-p                                                    
      Renormalization was made to 3.756 of Cf-252.                
      Below 1 MeV                                                 
         Nu-p = 2.486 + 0.1121*(E-DE),                            
         where DE is difference of average fragment kinetic energy
         between incident and thermal neutron energies.  It was   
         taken from data of Boldeman et al. /4/.                  
      Between 1 and 2.73 MeV                                      
         Nu-p = 2.436 + 0.1279*E                                  
      Between 2.73 and 7.47 MeV                                   
         Nu-p = 2.327 + 0.1678*E                                  
      Above 7.47 MeV                                              
         Nu-p = 2.857 + 0.09689*E                                 
                                                                  
MF=2  Resonance Parameters                                        
  MT=151                                                          
  a) Resolved resonance region ( 1 eV to 150 eV)                  
      Resolved resonance parameters for the Reich-Moore formula   
      were obtained by using SAMMY/5/.  Details are given in      
      Appendix.                                                   
  b) Unresolved resonance region ( 0.15 keV to 30 keV)            
      Parameters were deduced with ASREP code /6/ so as to        
      reproduce the evaluated cross sections in this energy       
      region.                                                     
                                                                  
     2200-m/s cross sections and calculated res. integrals        
                       2200 m/s      Res. Integ.                  
           total       588.38 b          -                        
           elastic      11.97 b          -                        
           fission     531.16 b         774 b                     
           capture      45.25 b         138 b                     
                                                                  
MF=3  Neutron Cross Sections                                      
 Smooth part (above 30 keV)                                       
                                                                  
  MT=1    Total                                                   
     Based on the data of Poenitz /7,8/. Between 10 and 48        
     keV, cross-section curve calculated with the statistical-    
     model code CASTHY /9/ and the coupled-channel theory code    
     ECIS /10/ was normalized at 48 keV.                          
  MT=2    Elastic                                                 
     Obtained by subtracting non-elastic scattering cross section 
     from the total cross section.                                
  MT=4 and 51-64,91  Inelastic scattering                         
     Calculated with CASTHY /9/ and ECIS /10/. Coupled levels were
     first three levels.  Deformed OMP recommended by Haouat et   
     al. /11/, was slightly modified so as to reproduce the       
     experimental data of Smith et al. /12/, and spherical OMP was
     the same as that used for JENDL-2.  In the energy range above
     8.25 MeV, the cross section was approximated by using an     
     extponential-type fuction, because the cross section curve   
     obtained by CASTHY and ECIS showed large fluctuation.        
                                                                  
      Deformed OMP                                                
         V =46.4-0.3*E            , Ws=3.5 +0.4*E , Vso=6.2 (MeV) 
         r0=1.26                  , rs=1.26       , rso=1.12 (fm) 
         a0=0.63                  , b =0.52       , aso=0.47 (fm) 
         Beta-2=0.20, Beta-4=0.074                                
      Spherical OMP                                               
         V =41.8-0.20*E+0.008*E**2, Ws=6.50-0.15*E, Vso=6.0 (MeV) 
         r0=1.31                  , rs=1.36       , rso=1.32 (fm) 
         a0=0.57                  , b =0.44       , aso=0.50 (fm) 
                                    (dir. W.S.)                   
                                                                  
     Level scheme was taken from Ref. /13/.                       
             No.      Energy(MeV)  Spin-Parity                    
             g.s.      0.0           5/2 +      *                 
              1        0.04035       7/2 +      *                 
              2        0.0922        9/2 +      *                 
              3        0.1551       11/2 +      *                 
              4        0.29882       5/2 -                        
              5        0.31191       3/2 +                        
              6        0.3208        7/2 -                        
              7        0.34047       5/2 +                        
              8        0.3537        9/2 -                        
              9        0.397        11/2 -                        
             10        0.39849       1/2 +                        
             11        0.41576       3/2 +                        
             12        0.5039        7/2 -                        
             13        0.5467        5/2 +                        
             14        0.5971        7/2 +                        
     Above 0.6 MeV, assumed to be overlapped. Levels with asterisk
     were coupled in the ECIS calculation.                        
                                                                  
  MT=16,17  (n,2n) and (n,3n)                                     
     Calculated by using the EGNASH-2 code /14/. The (n,2n) cross 
     section was normalized to fission-spectrum-averaged value of 
     0.00408 b measured by Kobayashi et al./15/.  The same        
     normalization factor was also applied to the (n,3n) cross    
     section.                                                     
  MT=18   Fission                                                 
     Based on the experimental data of Gwin et al. /16/, Carlson  
     et al. /17/, Manabe et al. /18/, Kanda et al. /19/, Iwasaki  
     et al. /20/, Meadows /21/, Lisowski et al./22/ and the       
     fission cross section of U-235 obtained by the simultaneous  
     evaluation /23/ and measured by Carlson et al./24/ between   
     13.25 and 20 MeV.                                            
  MT=102  Capture                                                 
     In the energy range from 30 keV to 1 MeV, the alpha values   
     measured by Hopkins and Diven /25/ were multiplied by the    
     fission cross section. In the high energy region, values     
     calculated with CASTHY and ECIS were normalized to 0.0578 b  
     at 1 MeV.                                                    
  MT=251  Mu-bar                                                  
     Calculated with CASTHY and ECIS.                             
                                                                  
MF=4  Angular Distributions of Secondary Neutrons                 
  MT=2, 51-64 and 91                                              
     Calculated with CASTHY and ECIS.                             
  MT=16,17 and 18                                                 
     Assumed to be isotropic in the Lab system.                   
                                                                  
MF=5  Energy Distributions of Secondary Neutrons                  
  MT=16,17,91                                                     
     Calculated with PEGASUS /26/.                                
  MT=18  Fission spectrum                                         
     Distributions were calculated with the modified Madland-Nix  
     model/27,28/.  The compound nucleus formation cross sections 
     for fission fragments (FF) were calculated using Becchetti-  
     Greenlees potential/29/.  Up to 4th-chance-fission were      
     considered at high incident neuttron energies.  The Ignatyuk 
     formula/30/ were used to generate the level density          
     parameters.                                                  
       Parameters adopted:                                        
           Total average FF kinetic energy = 172.311-0.0212*E(MeV)
           Average energy release          = 188.438 MeV          
           Average mass number of light FF =  95                  
           Average mass number of heavy FF = 139                  
           Level density of the light FF   =  9.999- 10.094       
           Level density of the heavy FF   = 11.89 - 12.20        
       Note that the parameters vary with the incident energy     
       within the indicated range.                                
  MT=455  Delayed neutrons                                        
     Recommendation by Saphier et al. /31/ was adopted.           
                                                                  
References                                                        
 1) Derrien H.: to be published in J. Nucl. Sci. Technol. (1994)  
 2) Master C.F. et al.: Nucl. Sci. Eng., 36, 202 (1969).          
 3) Evans A.E. et al.: Nucl. Sci. Eng., 50, 80 (1973).            
 4) Boldeman J.W. et al.: Nucl. Phys., A265, 337 (1976).          
 5) Larson N.: ORNL/TM-9179/R1 (1985).                            
 6) Kikuchi Y.: to be published.                                  
 7) Poenitz W.P. et al.: Nucl. Sci. Eng., 78, 333 (1981).         
 8) Poenitz W.P. et al.: ANL/NDM-80 (1983).                       
 9) Igarasi S. and Fukahori T.: JAERI 1321 (1991).                
10) Raynal J.: ECIS.                                              
11) Haouat G., et al.: Nucl. Sci. Eng., 81, 491 (1982).           
12) Smith A.B. et al.: 1982 Antwerp, p.039 (1982).                
13) Lederer D.G. and Shirley V.S.: Table of Isotopes, 7th Ed.     
    (1978).                                                       
14) Yamamuro N.: JAERI-M 90-006 (1990).                           
15) Kobayashi K.: J. Nucl. Sci. Technol., 10, 668 (1973).         
16) Gwin R. et al.: Nucl. Sci. Eng., 59, 79 (1976).               
17) Carlson G.W. and Behrens J.W.: Nucl. Sci. Eng., 66, 205 (1978)
18) Manabe F., et al. : 1987 Annual Meeting of Atomic Energy      
    Society of Japan, Nagoya, p.167 (1987) in Japanese.           
19) Kanda K., et al.: 1985 Santa-Fe, p.569 (1985).                
20) Iwasaki T., et al.: private communication (1987).             
21) Meadows J.W.: Nucl. Sci. Eng., 54, 317 (1974).                
22) Lisowski P.W.: 1991 Juelich, p.732 (1992).                    
23) Kanda Y. et al.: 1985 Santa Fe, 2, 1567 (1986).               
24) Carlson A.D. et al.: 1991 Juelich, p.518 (1992).              
25) Hopkins J.C. and Diven B.C.: Nucl. Sci. Eng., 12, 169 (1962). 
26) Iijima S, et al.: JAERI-M 87-025, p.337 (1987).               
27) Madland D.G. and Nix J.R.: Nucl. Sci. Eng., 81, 213 (1982).   
28) Ohsawa T. and Shibata T.: 1991 Juelich Conf., 965 (1992).     
29) Becchetti Jr.F.D. and Greenlees G.W.: Phys. Rev., 182, 1190   
    (1969).                                                       
30) Ignatyuk A.V.: Sov. J. Nucl. Phys., 29, 450 (1979).           
31) Saphier D., et al.: Nucl. Sci. Eng., 62, 660 (1977).          
                                                                  
                                                                  
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 Appendix   RESONANCE DATA ,JAERI DECEMBRE 1992                   
***************************************************************** 
                                                                  
      The Reich-Moore R-matrix resonance parameters were obtained 
from sequencial Bayesian fits of selected experimental total, fis-
sion and capture cross sections performed with the computing code 
SAMMY. The selected experimental data were the following:         
 1) total cross sections measured by Pattenden/1/, Moore/2/,      
    Kolar/3/, Harvey/4/;                                          
 2) fission cross sections measured by Weston/5,6/, Blons/7/,     
    Deruyter/8/, Wagemans/9/.                                     
 3) capture cross sections measured by Weston/5,6/.               
      In the low energy range the data of Pattenden,Moore,Harvey, 
Kolar, Weston, Deruyter and Wagemans were considered.In the energy
range from 15 eV to 150 eV, only the data from Kolar, Blons and   
Weston could be analysed owing to the poor resolution of the other
data. Some of the data were renormalized to the Axton standard    
/10/ at 0.0253 eV. The fission cross section measurement of Weston
/5/ available in the energy range above 1 eV was renormalised to  
the data of Wagemans/9/ over the energy range from 1 eV to 20 eV, 
resulting in an increase of the cross section by 2.4% compared to 
the original EXFOR file. The fission cross section measurement of 
Blons available in the energy range above 15 eV was renormalized  
to the the data of Weston over the energy range from 15 eV to 150 
eV, resulting in an increase of the cross section by 2.9% compared
to the data in the original EXFOR file. The total cross section of
Kolar needed a background correction of (2.3-0.038E) barn, E in   
eV, in the energy range from 77 eV to 150 eV, corresponding to 0  
to 1.5% of the measured transmission.  These renormalizations and 
background corrections were performed after preliminary fits of   
the data available from the EXFOR file in order to realize a      
consistency of +-2% among the cross section of the experimental   
data base. The final SAMMY fits were performed without renormali- 
zation and background correction parameters                       
    The transmission data of Pattenden, Moore, Harvey and Kolar   
are not available from the EXFOR file and were not requested from 
the authors. The SAMMY fits were performed on the experimental ef-
fective total cross sections using the sample thicknesses and the 
experimental resolution to calculate the theoritical effective    
cross sections. Enough informations were found in the publications
by the authors to ensure the accuracy of the calculations.        
    Due to the high resolution of the transmission measurements of
Kolar(100 m flight path) and of the fission measurements of Blons 
(50 m flight path and sample cooled down at liquid nitrogen tempe-
rature) the analysis could be performed up to 150 ev neutron ener-
gy. The high resolution fission cross section of Cao/11/ were not 
included in the experimental data base owing to a severe problem  
of the renormalization of the data.                               
    The capture cross sections of Weston were included in the fits
below 30 eV only. Above 30 eV the statistical accuracy of the data
was too poor and the try and error method was used in a prelimina-
ry work to obtain the capture width of some strong capture reson- 
ances. The capture width of the other resonances was kept at a    
constant value of 41 meV close to the average value obtained by   
fitting the energy range below 30 eV. Some resonances not pertain-
ing to 233U were disclosed in the experimental data and were iden-
tified as 195Pt resonances. The experimental data of Table 2 were 
roughly corrected for these resonances.                           
    The values of the cross sections obtained by Axton at 0.0253  
eV were included in all the experimental data available in the    
thermal range with the small error bars obtained by Axton, in     
order to ensure the best agreement between the calculated and the 
evaluated thermal values. The values calculated from the resonance
parameters are the followings:                                    
                                                                  
                Calculation          Axton                        
           (RESENDD 0.1% 300 K)    Evaluation                     
                                                                  
   Fission       531.29 b        530.70+-1.34 b                   
   Capture        45.27 b         45.62+-0.70 b                   
   Scattering     11.99 b         12.19+-0.67 b                   
                                                                  
                                                                  
                                                                  
      Tables 1 and 2 show the average cross sections calculated   
from the resonance parameters compared with the average experimen-
tal data and with average JENDL-3, ENDF/B-VI and JEF-2 data.      
                                                                  
                                                                  
   Table 1     Fission Cross Sections                             
                                                                  
-------------------------------------------------------------     
    Energy   Wagem Deruy Westo Blons Calcu JENDL ENDF6 JEF2       
  Range(eV)                                                       
-------------------------------------------------------------     
 0.021-0.031 525.6 526.7 526.8       526.5 528.8 523.9 520.6      
 0.031-0.082 362.9 363.5 363.9       361.7 363.9 361.9 359.7      
 0.082-1.000 151.3 150.9 150.6       150.1 149.4 149.0 148.6      
 1.000-2.100 388.8 387.7 391.7       387.9 383.0 378.9 382.1      
 2.100-2.750 204.4 204.6 207.5       204.6 205.9 198.1 198.8      
 2.750-3.000  50.1  53.4  51.9        50.2  52.9  50.6  50.8      
 3.000-15.00 106.2 105.6 104.9       104.3 103.6 101.2 101.5      
                                                                  
 0.021-15.00 134.2 133.5 133.5       132.5 131.7 129.1 129.5      
                                                                  
 15.00-30.00             95.03 94.60 95.51 96.49 91.80 92.69      
 30.00-50.00             40.13 40.30 40.27 40.19 38.85 39.16      
 50.00-75.00             40.66 40.49 40.53 40.79 35.80 39.90      
 75.00-100.0             35.57 36.70 36.03 36.58 33.36 32.74      
 100.0-125.0             36.84 36.89 36.97 31.78 29.94 28.94      
 125.0-150.0             21.29 20.29 20.78 16.30 22.10 26.25      
                                                                  
 15.00-150.0             41.39 41.48 41.45 39.91 38.49 40.05      
--------------------------------------------------------------    
                                                                  
                                                                  
  Table 2   Capture Cross Sections                                
                                                                  
-------------------------------------------------------------     
    Energy        Weston     Calcul  JENDL3  ENDF6   JEF2.2       
  Range(eV)                                                       
-------------------------------------------------------------     
 0.021-0.031      45.17       44.90   44.82   45.40   45.54       
 0.031-0.082      32.51       32.57   32.45   32.58   32.68       
 0.082-1.000      14.06       14.44   13.99   13.24   13.13       
 1.000-2.100      66.83       66.54   70.54   67.46   67.31       
 2.100-2.750     111.83      110.67  106.25  112.04  110.80       
 2.750-3.000       7.50        8.25    8.85    7.53    5.74       
 3.000-15.00      17.43       17.61   19.51   17.66   17.02       
                                                                  
 0.021-15.00      24.85       24.97   26.57   24.22   24.43       
                                                                  
 15.00-30.00      13.25       13.97   11.92   13.27   12.67       
 30.00-50.00       5.21        5.81    4.85    5.47    5.00       
 50.00-75.00       4.91        5.38    4.42    3.80    5.25       
 75.00-100.0       8.71        9.07    5.39    4.30    5.33       
 100.0-125.0       5.37        6.01    3.55    3.88    4.63       
 125.0-150.0       3.38        3.78    2.12    3.54    4.12       
                                                                  
 15.00-150.0       6.39        6.90    4.91    5.16    5.73       
--------------------------------------------------------------    
    The experimental capture data of Weston was increased by      
a background correction according to the evaluation of Reynolds   
et al./12/ and renormalized to the original value of Weston in    
the energy range 1.0 eV to 2.75 eV.                               
                                                                  
                                                                  
              RESONANCE INTEGRAL FISSION                          
                                                                  
    Energy range   this work  JENDL-3   ENDF/B-6   JEF-2          
   ------------------------------------------------------         
    0.5 eV-150 eV    710.34    710.53    691.08    697.18         
    150 eV-20 MeV               64.25     63.34     65.29         
    0.5 eV-20 MeV              774.79    754.43    762.47         
   ------------------------------------------------------         
                                                                  
                                                                  
              RESONANCE INTEGRAL CAPTURE                          
                                                                  
    Energy range   this work  JENDL-3   ENDF/B-6   JEF-2          
   ------------------------------------------------------         
    0.5 eV-150 eV    131.92    131.77    128.79    127.51         
    150 eV-20 MeV                6.65      7.58      7.24         
    0.5 eV-20 MeV              138.42    136.37    134.75         
   ------------------------------------------------------         
                                                                  
References of Appendix                                            
 1) PATTENDEN et al.: Nucl. Sci. Eng., 17, 404 (1963)             
 2) MOORE et al.: Phys. Rev., 118, 714 (1960)                     
 3) KOLAR et al.: 1970 Helsinki, Vol.I, 387 (1970)                
 4) HARVEY et al.: 1979 Knoxville, p.690 (1979)                   
 5) WESTON et al.: Nucl. Sci. Eng., 34, 1 (1968)                  
 6) WESTON et al.: Nucl. Sci. Eng., 42, 143 (1970)                
 7) BLONS et al.: Nucl. Sci. Eng., 51, 130 (1973)                 
 8) DERUYTER et al.: Nucl. Sci. Eng., 54, 423 (1974)              
 9) WAGEMANS et al.: 1988 Mito, 91 (1988)                         
10) AXTON et al.: BCMN Report, GE/PH/01/86 (1986)                 
11) CAO et al.: 1970 Helsinki. Vol.I, 419 (1970)                  
12) REYNOLDS et al.: KAPL-M-7323 (1973)                           
                                                                  
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