14-Si- 30 LANL,ORNL EVAL-JUN97 M.B.CHADWICK,P.G.YOUNG,D.HETRICK 
 Ch97,Ch99,He97       DIST-AUG99 REV1-                 19990907   
----ENDF/B-VI         MATERIAL 1431         REVISION 1            
-----INCIDENT NEUTRON DATA                                        
------ENDF-6 FORMAT                                               
******************************************************************
Oct 22,2004, Holly Trellue remade mf6mt5 using corrected gnash    
code, to fix an earlier bug. Impact is reduced sec. particle prod.
for new ENDF/B-VII release                                        
                                                                  
 ENDF/B-VI MOD 2 Evaluation, June 1996, M.B. Chadwick and         
           P.G. Young, LANL                                       
                                                                  
 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code     
   in cooperation with ECN Petten.                                
                                                                  
  This evaluation provides a complete representation of the       
nuclear data needed for transport, damage, heating,               
radioactivity, and shielding applications over the incident       
neutron energy range from 1.0E-11 to 150 MeV.  The discussion     
here is divided into the region below and above 20 MeV.           
                                                                  
INCIDENT NEUTRON ENERGIES < 20 MeV                                
                                                                  
  Below 20 MeV the evaluation is based completely on the ENDF/B-  
VI.5 (Release 5) evaluation by D. Hetrick, N. Larson, D. Larson,  
L. Leal, and S. Epperson.                                         
                                                                  
INCIDENT NEUTRON ENERGIES > 20 MeV                                
                                                                  
  The ENDF/B-VI Release 5 evaluation extends to 20 MeV and        
includes cross sections and energy-angle data for all             
significant reactions.  The present evauation utilizes a more     
compact composite reaction spectrum representation above 20 MeV   
in order to reduce the length of the file. No essential data for  
applications is lost with this representation.                    
  The evaluation above 20 MeV utilizes MF=6, MT=5 to represent    
all reaction data.  Production cross sections and emission        
spectra are given for neutrons, protons, deuterons, tritons,      
alpha particles, gamma rays, and all residual nuclides produced   
(A>5) in the reaction chains. To summarize, the ENDF sections     
with non-zero data above En = 20 MeV are:                         
                                                                  
  MF=3 MT=  1  Total Cross Section                                
       MT=  2  Elastic Scattering Cross Section                   
       MT=  3  Nonelastic Cross Section                           
       MT=  5  Sum of Binary (n,n') and (n,x) Reactions           
                                                                  
  MF=4 MT=  2  Elastic Angular Distributions                      
                                                                  
  MF=6 MT=  5  Production Cross Sections and Energy-Angle         
               Distributions for Emission Neutrons, Protons,      
               Deuterons, Tritons, and Alphas; and Angle-         
               Integrated Spectra for Gamma Rays and Residual     
                                                                  
               Nuclei That Are Stable Against Particle Emission   
                                                                  
  The evaluation is based on nuclear model calculations that      
have been benchmarked to experimental data, especially for n +    
Si28 and p + Si28 reactions [Ch97]. We use the GNASH code system  
[Yo92], which utilizes Hauser-Feshbach statistical,               
preequilibrium and direct-reaction theories. Spherical optical    
model calculations are used to obtain particle transmission       
coefficients for the Hauser-Feshbach calculations, as well as     
for the elastic neutron angular distributions.                    
  Cross sections and spectra for producing individual residual    
nuclei are included for reactions. The energy-angle-correlations  
for all outgoing particles are based on Kalbach systematics       
[Ka88].                                                           
  A model was developed to calculate the energy distributions of  
all recoil nuclei in the GNASH calculations [Ch96]. The recoil    
energy distributions are represented in the laboratory system in  
MT=5, MF=6, and are given as isotropic in the lab system. All     
other data in MT=5,MF=6 are given in the center-of-mass system.   
This method of representation utilizes the LCT=3 option approved  
at the November, 1996, CSEWG meeting.                             
                                                                  
  Preequilibrium corrections were performed in the course of the  
GNASH calculations using the exciton model of Kalbach [Ka77,      
Ka85], validated by comparison with calculations using Feshbach,  
Kerman, Koonin (FKK) theory [Ch93].  Discrete level data from     
nuclear data sheets were matched to continuum level densities     
using the formulation of Ignatyuk [Ig75] and pairing and shell    
parameters from the Cook [Co67] analysis. Neutron and charged-    
particle transmission coefficients were obtained from the         
optical potentials, as discussed below.  Gamma-ray transmission   
coefficients were calculated using the Kopecky-Uhl model [Ko90].  
                                                                  
DETAILS OF THE n + SI-30 ANALYSIS                                 
                                                                  
  The neutron total cross section above 20 MeV was obtained by    
evaluating the nat-Si experimental data, with a particular        
emphasis on the high-accuracy Los Alamos measurements by Finlay   
[Fi93]. The Si-28 total cross section, based on nat-Si data, was  
scaled by 1.047 (an A**2/3 factor) for Si-30.                     
  The Madland global medium-energy optical potential [Ma88] was   
used for neutrons above 46 MeV, and the Wilmore-Hodgson           
potential was used for lower neutron energies.  The Madland       
global medium-energy optical potential was used for protons       
above 28 MeV, and the Beccheti-Greenlees potential was used for   
lower proton energies. In both cases the transition region to     
the Madland potential was chosen to approximately give            
continuity in the reaction cross section.  For deuterons, the     
Perey global potential was used; for alpha particles the Moyen    
potential was used; and for tritons the Beccheti-Greenlees        
potential was used.                                               
  While the above optical potentials did describe the             
experimental proton nonelastic cross section data fairly well,    
we modified the theoretical predictions slightly to better agree  
with the measurements, and renormalized the transmission          
coefficients accordingly. In addition to using Si nonelastic      
proton cross section measurements, we also were guided by p+Al    
nonelastic data, scaled by A**2/3. The Si-30 nonelastic cross     
section was taken by scaling the evaluated Si-28 value by 1.047   
(an A**2/3 factor).                                               
  Inelastic scattering to the 2+ (2.24 MeV) and 4+ (5.95 MeV)     
states in 30-Si was determined using a coupled-channel ECIS       
calculation.  To produce continuity in the calculated inelastic   
cross sections up to 150 MeV, we performed an oblate rotational   
band (0+, 2+, 4+) coupled channel calculation using the Madland   
medium energy potential (with its imaginary potential reduced by  
20%, to approximately account for the coupling). Deformation      
parameters were chosen to reproduce the JENDL-3 evaluation at 20  
MeV (H.Kitazawa et al.). The resulting deformation parameters     
(beta_2=-0.33, beta_4=0.20) were close to those used for Si-28.   
  The same preequilibrium input parameters were used as for Si-   
28, which was benchmarked against (n,xz) data from the Louvain    
group at 63 MeV, and against unpublished (n,xp) data by Haight    
et al. for neutrons up to 50 MeV. See our ENDF file-1 for n+28Si  
for more details.                                                 
                                                                  
                                                                  
****************************************************************  
                                                                  
 ENDF/B-VI MOD 1 Evaluation, May, 1996                            
           D.M. Hetrick, D.C. Larson, N.M. Larson, L.C. Leal,     
           S.J.Epperson (ORNL)                                    
                                                                  
    Complete documentation for this evaluation is found in        
 Hetrick, et al. [He97].                                          
    This work employed nuclear model codes including the          
 Distorted Wave Born Approximation (DWBA) program DWUCK [Ku72]    
 and the Hauser-Feshbach code TNG [Fu88],[Fu80],[Sh86].  The TNG  
 code provides energy and angular distributions of particles      
 emitted in the compound and pre-compound reactions, ensures      
 consistency among all reactions, and maintains energy balance.   
 Evaluation work was completed in 1989, except for MT=3,          
 MF=1,2,3,102 which were completed May 1996.  The latter results  
 Incorporated a new resonance region analysis.  Uncertainty files 
 are provided for all FILE 3 cross sections using methodology     
 described in Hetrick et al. [He91].                              
                                                                  
 DESCRIPTION OF FILES                                             
 MF=2  RESONANCE PARAMETERS                                       
   MT=151 Resonance parameters from a new analysis for all three  
      isotopes of silicon, by Leal et al. [Le96], using the       
      multilevel R-matrix analysis code SAMMY [La96].  Energy     
      range 1.0E-5 eV to 1.50 MeV.                                
      The following data were included in the analysis:           
      (1) Total cross section data of Perey et al. [Pe72] for     
        natural silicon, measured on the 47-m flight path at the  
        Oak Ridge Electron Linear Accelerator (ORELA)             
      (2) Transmission data of Harvey et al. [Ha96a] for natural  
        silicon measured on the 200-m flight path at ORELA        
      (3) Total cross section data of Larson et al. [La76],       
        measured on the 80- and 200-m flight paths at ORELA       
      (4) 29Si-oxide [Ha96b) and 30Si-oxide [Ha83] transmission   
        data of Harvey et al., measured on the 80-m ORELA flight  
        path                                                      
      (5) Elastic scattering thermal cross section for 30Si       
        2.490+-0.169 barns from Mughabghab [Mu81].  Capture       
        thermal cross section 0.107+-0.003 barns from Raman et    
        al. [Ra92].  Values given by resonance parameters are     
        2.500 and 0.107 barns respectively.                       
                                                                  
 MF=3 CROSS SECTIONS                                              
    MT=1  Total cross section 1.E-5 eV to 1.50 MeV taken from     
       resonance parameters evaluation.  From 1.50 to 20 MeV,     
       based on 200-m data [Ha96a] for natural silicon.           
       Thermal value = 2.597 barns from (exp.) and 2.590 from     
       resonance parameters.                                      
    MT=2   Elastic scattering cross sections were obtained by     
       subtracting the nonelastic from the total.                 
    MT=3   Nonelastic cross section: sum of MT=4,16,22,28,102,103,
       104, and 107.                                              
    MT=4   Total inelastic cross section: sum of MT=51-56 and 91. 
    MT=16  (n,2n) cross sections calculated by the TNG code       
       [Fu88],[Fu80],[Sh86].                                      
    MT=22  (n,na) cross sections calculated by the TNG code.      
       No data available other than alpha emission.               
    MT=28  (n,np) cross sections calculated by the TNG code.      
       No data available other than proton emission.              
    MT=51-56  Inelastic scattering excited levels; results are    
       taken from the code TNG [Fu88],[Fu80],[Sh86] which includes
       direct interaction cross sections from DWUCK [Ku72]        
       calculations (when data available).                        
    MT=91  Inelastic scattering exciting the continuum calculated 
       by code TNG with guidance from (n,xn) emission data.       
    MT=102 (n,g) capture cross section, 1.e-5 eV to 1.50 MeV      
       taken from resonance parameters.  From 1.5o to 20 MeV,     
       evaluation based on TNG calculation, normalized to         
       available data at 14.7 MeV.  Thermal value = 0.107 barns   
       [Ra92].                                                    
    MT=103 (n,p) cross sections were calculated by the code TNG   
       [Fu88],[Fu80],[Sh86] from 9.0 to 20.0 MeV.  From threshold 
        to 9.0 MeV, taken from ENDF/B-V evaluation.               
    MT=107 (n,a) cross sections were calculated by the code TNG   
       [Fu88],[Fu80],[Sh86].                                      
    MT=600-605 (n,p) reaction to discrete levels calculated by    
       code TNG [Fu88],[Fu80],[Sh86].                             
    MT=649 (n,p) continuum protons calculated by code TNG [Fu88], 
       [Fu80],[Sh86].                                             
    MT=800-811 (n,a) reaction to discrete levels calculted by the 
       code TNG [Fu88],[Fu80],[Sh86].                             
    MT=849 (n,a) continuum alphas calculated by the code TNG      
       [Fu88],[Fu80],[Sh86].                                      
                                                                  
 MF=4 ANGULAR DISTRIBUTIONS                                       
    MT=2   Angular distributions of secondary neutrons from       
       elastic scattering are taken from the ENDF/B-V evaluation. 
                                                                  
 MF=6 PRODUCT ENERGY-ANGLE DISTRIBUTIONS                          
    MT=16  (n,2n) reaction includes simple constant yields for    
       the neutron and 20Si residual, and energy dependent gamma- 
       ray yield based on TNG calculated gamma-ray spectra.  TNG  
       calculated normalized distributions are given for each     
       product.  (Angular distributions are given only for the    
       outgoing neutron).                                         
    MT=22  (n,na) reaction includes simple constant yields for    
       the neutron, alpha, and 26Mg residual, and energy-         
       dependent gamma-ray yield based on TNG calculated gamma-   
       ray spectra.  Calculated normalized distributions are      
       given for each product.  (Angular distributions are given  
       only for the outgoing neutron; isotropy is assumed for the 
       alpha and residual).                                       
    MT=28  (n,np) reaction includes simple constant yields for    
       the neutron and 29Al residual, and energy dependent gamma- 
       ray yield based on TNG calculated gamma-ray spectra;       
       calculated normalized distributions are given for each     
       product.  (Angular distributions are given only for the    
       for the outgoing neutron).                                 
    MT=51-56 Inelastic scattering exciting levels; each section   
       includes simple constant yields for the neutron and 30Si   
       residual; angular distributions are given for the outgoing 
       neutron (Legendre coefficients computed by DWUCK [Ku72]    
       and TNG [Fu88],[Fu80],[Sh86]).                             
    MT=91  Inelastic scattering exciting the continuum includes   
       simple constant yields for the neutron and 30Si residual   
       and energy dependent gamma-ray yield based on TNG          
       calculated gamma-ray spectra; TNG calculated normalized    
       distributions are given for each.  (Angular distributions  
       are given only for the outgoing neutron).                  
    MT=600-605 (n,p) proton angular distributions to discrete     
       states taken from the TNG calculation.                     
    MT=649 (n,p) continuum includes simple constant yields for    
       protons and 30Al residual, and energy-dependent gamma-ray  
       yield based on TNG calculated gamma-ray spectra; calculated
       normalized distributions are given for each product.       
    MT=800-811 (n,a) alpha angular distributions to discrete      
       states taken from the TNG calculation.                     
    MT=849 (n,a) continuum: includes simple constant yields for   
       alphas and 27Mg residual, and energy-dependent gamma-ray   
       yield based on calculated gamma-ray spectra -- calculated  
       normalized distributions are given for each product.       
                                                                  
 MF=12 PHOTON PRODUCTION MULTIPLICITIES                           
    MT=51-67 Branching ratios for the levels are given.  These    
       ratios were used in the TNG calculation (i.e., the         
       continuum and discrete gammas are separated).              
    MT=102 (n,g) capture spectra based on thermal measurements of 
       Raman et al. [Ra92].  Yields are given for all observed    
       gamma rays.  The thermal results are used from 1.e-5 eV to 
       20 MeV.  Some small renormalizations were used on large    
       yields to ensure energy balance.                           
    MT=601-605 Branching ratios for the levels are given.   These 
       ratios were used in the TNG calculation (i.e., the         
       continuum and discrete gammas are separated).              
    MT=801-811 Branching ratios for the levels are given.  These  
       ratios were used in the TNG calculation (i.e., the         
       continuum and discrete gammas are separated).              
                                                                  
 MF=14 PHOTON ANGULAR DISTRIBUTIONS                               
    MT=51-67,102,601-605,801-811  All gamma ray angular           
       distributions assumed to be isotropic.                     
                                                                  
 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS                      
    Uncertainty files are given for all FILE 3 reactions, using   
      the methodology described in Hetrick et al. [He91], and     
      used for ORNL ENDF/B-VI evaluations for isotopes of Cr, Fe, 
      Ni, Cu, and Pb.  LB=0,1,8 sub-subsections are utilized.     
                                                                  
***************************************************************** 
                                                                  
REFERENCES                                                        
                                                                  
[Be69]  F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182,   
   1190 (1969)                                                    
[Be71]  F.D. Becchetti, Jr., and G.W. Greenlees (Ed: H.H.         
   Barschall and W. Haeberli, The University of Wisconsin Press,  
   1971) p.682                                                    
[Ch93]  M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 
[Ch96]  M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J.      
   Koning, "High-Energy Nuclear Data Libraries for Accelerator-   
   Driven Technologies:  Calculational Method for Heavy Recoils," 
   Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation    
   Technology and Applications, Kalmar, Sweden, 3-7 June 1996.    
[Ch97]  M.B. Chadwick and P.G. Young, "GNASH Calculations of      
   n,p + Si isotopes and Benchmarking of Results" in APT PROGRESS 
   REPORT: 1 June - 1 July 1997, internal Los Alamos National     
   Laboratory memo T-2-97/MS-43, 7 July 1997 from R.E. MacFarlane 
   to L. Waters.                                                  
[Ch99]  M.B. Chadwick, P.G. Young, G.M. Hale, et al., Los Alamos  
  National Laboratory report, LA-UR-99-1222 (1999)                
[Fi93]  R.W. Finlay, W.P. Abfalterer, G. Fink, et al., Phys.Rev.  
   C 47, 237 (1993)                                               
[Fu80]  C.Y. Fu, Symp. on Neutron Cross Sections from 10 to 50 MeV
   Upton, NY, May 1980, Brookhaven National Laboratory report     
   BNL-NCS-51425, Vol. 2 (1980) p.675                             
[Fu88]  C.Y. Fu, Oak Ridge National Laboratory report             
   ORNL/TM-7042 (1980); see also Nucl.Sci.Eng. 100, 61 (1988)     
[Ha83]  J.A. Harvey, W.M. Good, R.F. Carlton, et al., Phys.Rev.   
   C 28, 24 (1983)                                                
[Ha96a] J.A. Harvey, private communication (1996)                 
[Ha96b] J.A. Harvey, private communication (1996)                 
[He91]  D.M. Hetrick, D.C. Larson, and C.Y. Fu, Oak Ridge         
   National Laboratory report ORNL/TM-11763 [ENDF-350] (1991)     
[He97]  D.M. Hetrick, D.C. Larson, N.M. Larson, et al., Oak Ridge 
   National Laboratory report ORNL/TM-11825 (1997)                
[Ig75]  A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J.    
   Nucl.Phys. 21, 255 (1975)                                      
[Ka77]  C. Kalbach, Z.Phys.A  283, 401 (1977)                     
[Ka85]  C. Kalbach, Los Alamos National Laboratory report         
   LA-10248-MS (1985)                                             
[Ka88]  C. Kalbach, Phys.Rev. C 37, 2350 (1988); see also         
   C. Kalbach and F.M. Mann, Phys.Rev. C 23, 112 (1981)           
[Ko90]  J. Kopecky and M. Uhl, Phys.Rev. C 41, 1941 (1990)        
[Ku72]  P.D. Kunz, "Distorted wave code DWUCK72," Univ. of        
   Colorado, unpublished (1972).                                  
[La76]  D.C. Larson, C.H. Johnson, J.A. Harvey, and N.W. Hill,    
   Oak Ridge National Laboratory report ORNL/TM-5618 (1976)       
[La96]  N.M. Larson, Oak Ridge National Laboratory report         
   ORNL/TM-9179 (1984); see also ORNL/TM-9179/R1 (1985),          
   ORNL/TM-9179/R2 (1989), and ORNL/TM-9179/R3 (1996)             
[Le96]  L.C. Leal, N.M. Larson, D.C. Larson, D.M. Hetrick,        
   "R-matrix analysis of Si28, Si29, and Si30 up to 1.8 MeV,"     
   to be published.                                               
[Ma88]  D.G. Madland, Proc. OECD/NEANDC Specialist's Meeting on   
   Preequilibrium Nuclear Reactions, Semmering, Austria, Feb.     
   1988, NEANDC-245 'U' (1988) p.103                              
[Mu81]  S.F. Mughabghab, M. Divadeenam, N.E. Holden, Neutron      
   Cross Sections, Vol. 1, Neutron resonance parameters and       
   thermal cross sections, Part A: Z=1-60, (Academic Press,       
   NY, 1981)                                                      
[Pe63]  C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963)      
[Pe72]  F.G. Perey, T.A. Love, W.E. Kinney, Oak Ridge National    
   Laboratory report ORNL-4823 [ENDF-178] (1972)                  
[Ra92]  S. Raman, et al., Phys.Rev. C 46, 972 (1992)              
[Sh86]  K. Shibata and C.Y. Fu, Oak Ridge National Laboratory     
   report ORNL/TM-10093 (1986)                                    
[Wi64]  D. Wilmore and P.E. Hodgson, Nucl.Phys. 55, 637 (1964)    
[Yo92]  P.G. Young, E.D. Arthur, and M.B. Chadwick, report        
   LA-12343-MS (1992)                                             
                                                                  
 ************************ C O N T E N T S *********************** 
                                1        451        523          2
                                2        151         36          1
                                3          1        755          2
                                3          2        755          2
                                3          3         77          2
                                3          4         14          1
                                3          5         57          2
                                3         16          7          1
                                3         22          6          1
                                3         28          6          1
                                3         51         12          1
                                3         52         11          1
                                3         53         11          1
                                3         54         11          1
                                3         55         10          1
                                3         56         10          1
                                3         91         10          1
                                3        102         11          1
                                3        103         10          1
                                3        107         14          1
                                3        600          8          1
                                3        601          8          1
                                3        602          7          1
                                3        603          7          1
                                3        604          7          1
                                3        605          7          1
                                3        649          7          1
                                3        800         10          1
                                3        801          9          1
                                3        802          9          1
                                3        803          8          1
                                3        804          8          1
                                3        805          7          1
                                3        806          7          1
                                3        807          7          1
                                3        808          7          1
                                3        809          7          1
                                3        810          7          1
                                3        811          7          1
                                3        849          6          1
                                4          2       1524          2
                                6          5      11736          2
                                6         16        216          1
                                6         22        212          1
                                6         28        166          1
                                6         51         69          1
                                6         52         60          1
                                6         53         63          1
                                6         54         66          1
                                6         55         55          1
                                6         56         56          1
                                6         91        925          1
                                6        600         39          1
                                6        601         36          1
                                6        602         35          1
                                6        603         36          1
                                6        604         33          1
                                6        605         31          1
                                6        649        342          1
                                6        800         62          1
                                6        801         54          1
                                6        802         50          1
                                6        803         47          1
                                6        804         41          1
                                6        805         39          1
                                6        806         42          1
                                6        807         40          1
                                6        808         38          1
                                6        809         35          1
                                6        810         38          1
                                6        811         36          1
                                6        849        380          1
                               12         51          3          1
                               12         52          3          1
                               12         53          3          1
                               12         54          3          1
                               12         55          4          1
                               12         56          3          1
                               12        102        103          1
                               12        601          3          1
                               12        602          3          1
                               12        603          3          1
                               12        604          3          1
                               12        605          3          1
                               12        801          3          1
                               12        802          3          1
                               12        803          3          1
                               12        804          3          1
                               12        805          4          1
                               12        806          3          1
                               12        807          3          1
                               12        808          3          1
                               12        809          3          1
                               12        810          4          1
                               12        811          4          1
                               14         51          1          1
                               14         52          1          1
                               14         53          1          1
                               14         54          1          1
                               14         55          1          1
                               14         56          1          1
                               14        102          1          1
                               14        601          1          1
                               14        602          1          1
                               14        603          1          1
                               14        604          1          1
                               14        605          1          1
                               14        801          1          1
                               14        802          1          1
                               14        803          1          1
                               14        804          1          1
                               14        805          1          1
                               14        806          1          1
                               14        807          1          1
                               14        808          1          1
                               14        809          1          1
                               14        810          1          1
                               14        811          1          1
                               33          1         20          1
                               33          2         14          1
                               33          3         16          1
                               33          4          9          1
                               33         16         12          1
                               33         22         14          1
                               33         28         12          1
                               33         51         16          1
                               33         52         16          1
                               33         53         16          1
                               33         54         16          1
                               33         55         16          1
                               33         56         16          1
                               33         91         14          1
                               33        102         19          1
                               33        103         14          1
                               33        107         16          1
                               33        600         14          1
                               33        601         14          1
                               33        602         14          1
                               33        603         14          1
                               33        604         14          1
                               33        605         14          1
                               33        649         14          1
                               33        800         16          1
                               33        801         14          1
                               33        802         16          1
                               33        803         14          1
                               33        804         14          1
                               33        805         14          1
                               33        806         14          1
                               33        807         14          1
                               33        808         14          1
                               33        809         14          1
                               33        810         14          1
                               33        811         14          1
                               33        849         14          1