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 92-U -237 BRC,+      EVAL-NOV04 LOPEZ JIMENEZ, MORILLON, ROMAIN  
                      DIST-JAN09                     20090105     
----JEFF-311          MATERIAL 9234                               
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
***************************  JEFF-3.1.1  *************************
**                                                              **
**         Original data taken from:  JEFF-3.1                  **
**                                                              **
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*****************************  JEFF-3.1  *************************
**                                                              **
**         Original data taken from:  New evaluation            **
**                                                              **
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  05-01 NEA/OECD (Rugama) 8 delayed neutron groups                
  Jefdoc-976 (Wilson and England, Prog Nucl Eng 41,71(2002)       
                                                                  
                                                                  
*****************************  JEFF-3.1  *************************
**                                                              **
**         Original data taken from:  New evaluation            **
**                                                              **
******************************************************************
                                                                  
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  JEFF-3.1 evaluation above the unresolved resonance region       
  based on model calculations, from 10 keV to 30 MeV.             
      M-J. Lopez-Jimenez, B. Morillon, P. Romain,                 
                    J-Ch. Sublet                                  
              CEA/DAM Bruyeres-le-Chatel                          
                  CEA/DEN Cadarache                               
                                                                  
MF=1 General Information                                          
                                                                  
   The prompt fission neutron multiplicity and spectra            
   are calculated using the BRC improved Los Alamos model from    
   Vladuca and Tudora [1]. The model parameters are slightly      
   different from those adopted in [1]. The prompt fission        
   neutron multiplicity is obtained from an energetic balance     
   ratio. The available energy (the average fission energy        
   released minus the average fission fragment kinetic energy     
   minus the average prompt gamma ray energy) is divided by the   
   energy carry away by the neutron (the average fission          
   fragment neutron separation energy plus the average            
   center-of-mass energy of the emitted neutrons). The main       
   improvement is the dependence of the average total             
   fission-fragment kinetic energy and the average gamma energy   
   on neutron incident energy.                                    
                                                                  
                                                                  
   MT=452 Total Nubar. Sum of MT=455 and 456                      
   MT=455 Delayed Neutron Yields.                                 
          05-01 NEA/OECD (Rugama) 8 delayed neutron groups.       
   MT=456 Prompt Neutron Yields.                                  
   Vladuca and Tudora BRC improved Madland-Nix model              
   MT=458 not given                                               
                                                                  
                                                                  
MF=2  Resonance Parameters                                        
                                                                  
   MT=151   JENDL3.3                                              
                                                                  
 [*******   JENDL3.3                                              
   1) Resolved Resonance Parameters: MLBW (1.0e-5 - 200 ev)       
      below 45 ev, hypothetical resonances were generated from    
      fission width of 0.004 ev, s0 of 1.0e-4 and level spacing of
      3.5 ev, and adjusted to reproduce thermal cross sections.   
      above 46 ev, parameters were estimated from fission-area    
      data measured by MCNALLY et al.[2]                          
   2) Unresolved Resonance Parameters: 200 ev - 30 kev            
      obtained by fitting to capture and fission cross sections   
      with ASREP[3].                                              
         S0 and S2 = (0.97 - 1.02)e-4, S1 = (1.95 - 2.04)e-4,     
         Gamma-f = (0.006 - 0.070) ev, Gamma-g = 0.035 ev         
         R = 9.668 fm                                             
                                                                  
      calculated thermal cross sections and res. integral (barns) 
                     0.0253 ev      resonance integral            
                      (barns)          (barns)                    
        Total         478.50               -                      
        Elastic        24.39               -                      
        Fission         1.70             48.7                     
        Capture       452.40           1080.0                     
                                                       *******]   
                                                                  
                                                                  
   Unresolved Resonance Range  10 keV to 30 keV :                 
   The four energy dependant widths parameters originally         
   described in JENDL-3.3 have been removed to account for        
   direct interaction on the first inelastuc level                
                                                                  
                                                                  
                                                                  
MF=3  Reaction Cross-sections                                     
                                                                  
   From the energy of 1 keV up to 200 MeV, eigth states (ground-  
   sate rotationnal band {1/2+,3/2+,5/2+,7/2+,9/2+} and octupolar 
   band {1/2- (540.62 keV), 3/2- (554.98 keV), 5/2- (578.01 keV)})
   Coupled Channel Calculations are performed using the ECIS95[4] 
   code which also provides compound nucleus cross sections and   
   transmission coefficients used in pre-equilibrium/evaporation  
   emission treated in the exciton and Hauser-Feshbach models     
   implemented in the Bruyeres-le-Chatel modified version of the  
   GNASH code[5]. This reaction code has been modified to include 
   width fluctuation factors, relativistic kinematics, and a more 
   realistic treatment of the fission process. A new fission [6,7]
   penetrability model taking into account Triple Humped Fission  
   Barrier (THFB) has been developed, explicitly coupling class   
   I, II and III states while damping those of class II and III.  
   Emission of light hadrons up to He4 are explicitly treated in  
   the model calculations. Fission decay of associated residual   
   nuclei is also treated. However, none of these emissions and   
   fission cross-sections, are yet explicitly provided in this    
   file.                                                          
                                                                  
   The Unresolved Resonance Range end at 10 KeV and the model     
   calculations data are implemented from 10 KeV.                 
                                                                  
   MT=1    calculation from BRC deformed optical potential        
           over the whole energy range 1 keV-200 MeV.             
   MT=2    calculation from BRC deformed optical potential        
   MT=3    calculation from BRC deformed optical potential        
   MT=4    calculation from BRC deformed optical potential        
           sum of mt=51-91.                                       
   MT=16   (n,2n) cross section                                   
   MT=17   (n,3n) cross section                                   
   MT=18   (n,F) calculation with BRC modified GNASH code, with   
           a triple humped fission barrier penetration model      
   MT=19-21(n,f),(n,nf),(n,2nf) calculation with BRC modified     
           GNASH code, with a triple humped fission barrier       
           penetration model.                                     
   MT=37   (n,4n) cross-section                                   
   MT=38   (n,3nf)calculation with BRC modified GNASH code,       
           with a triple humped fission barrier penetration       
           model. In fact this cross section include more         
           complex processes thus as : (n,4nf),(n,pf),(n,df),     
           (n,tf),(n,He-3f),(n,He-4f),(n,pnf), ...                
   MT=51-84(n,n') cross-section for 1st-34th excited states       
   MT=91   (n,n') continuum cross-section                         
   MT=102  (n,g) cross-section                                    
                                                                  
MF=4   Angular Distributions of Secondary Particles               
                                                                  
   MT=2     elastic angular distribution, given up to 30 MeV      
   MT=18    fission given up to 30 MeV (assumed isotropic)        
   MT=51-84 inelastic levels, 1st-34th excited states             
                                                                  
   With a uniform number of angular points (91), equal values     
   of the tabulated probability distributions may occur.          
                                                                  
MF=5   Energy Distributions of Secondary Particles                
                                                                  
   MT=18   Vladuca and Tudora BRC improved Madland-Nix model      
   MT=455  extended NEA/OECD data                                 
                                                                  
MF=6  Products Energy-angle Distributions                         
                                                                  
   MT-16   pre-ENDF/B-VII (237l)                                  
   MT=17   pre-ENDF/B-VII                                         
   MT=37   pre-ENDF/B-VII                                         
   MT=91   pre-ENDF/B-VII                                         
                                                                  
MF=12  Photon Production Multiplicities                           
                                                                  
   MT=18   pre-ENDF/B-VII                                         
   MT=102  pre-ENDF/B-VII                                         
                                                                  
MF=13  Photon Production Cross-section                            
                                                                  
   MT=3    pre-ENDF/B-VII                                         
                                                                  
MF=14  Photon Angular Distribution                                
                                                                  
   MT=3    pre-ENDF/B-VII                                         
   MT=18   pre-ENDF/B-VII                                         
   MT=102  pre-ENDF/B-VII                                         
                                                                  
MF=15  Continuous Photon Energy Spectra                           
                                                                  
   MT=3    pre-ENDF/B-VII                                         
   MT=18   pre-ENDF/B-VII                                         
   MT=102  pre-ENDF/B-VII                                         
                                                                  
                                                                  
                                                                  
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 References                                                       
                                                                  
[1] G. Vladuca and A. Tudora, Ann. Nuc. Energy. 28, 689 (2001).   
                                                                  
[2] J.H.Mcnally et al: phys. rev., c9, 717 (1974).                
                                                                  
[3] Y.Kikuchi: unpublished.                                       
                                                                  
[4] J. Raynal, "Code ECIS95" CEA report N-2772, (1994).           
                                                                  
[5] P.G. Young, E.D. Arthur and M. B. Chadwick, Workshop on       
    Nuclear Reaction Data and Nuclear Reactors, Trieste,          
    Italy (1996).                                                 
                                                                  
[6] M-J. Lopez-Jimenez, B. Morillon and P. Romain "Triple humped  
    fission barrier model for a new 238U neutron cross-section    
    evaluation and first validation with TRIPOLI code", to be     
    published, ANE, (2004).                                       
                                                                  
[7] A.J. Koning, M.C. Duijvestijn and M-J. Lopez-Jimenez, "Data   
    Evaluation up to 200 MeV for Fe, Pb and U", NRG Report,       
    20567/03.56876/P, (2003).                                     
                                                                  
                                                                  
                                                                  
                                                                  
                                                                  
                                                                  
                                                                  
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