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Thermal hydraulic design of nuclear reactor core

  1. Apr 15, 2012 #1
    By a simple procedure , what should I do when I'm going through the thermal hydraulic design of nuclear reactor core...?
    ..
    I put initial guesses for the core dimensions ( fuel , clad , gap , length)
    initial guess for the fuel element pitch
    desired power ...
    I've found q'' critical heat flux based on the correlation in "Nuclear Systems I - Thermal Hydraulic Fundamentals - Todreas"
    I've found q'max ( based o DNB =1.3 )
    q'avg based on ( hot spot factor =2 )
    and based on q'max I've found the temperature distribution in one single fuel rod
    ...
    The results
    after assuming
    we have only 1 assembly inside the core that produces 17.6 MWth
    All fuel rods have the same enrichment
    the reflector will take care of the leakage

    to produce the desired power we need
    188 fuel rods inside the assembly
    Rf = 0.405 cm (fuel pellet radius)
    Tg = 0.005 cm (gap thickness)
    Tc = 0.05 cm (clad thickness)
    Lr = 360 cm (fuel rod length)
    fuel element pitch (p) = 1.3 cm
    Pressure 15.5 Mpa

    with coolant flux flow rate 436 kh/hr.cm2 in one single channel
    I've found that
    ΔTF ( temperature drop through the fuel pellet) = 1656.5 C
    ΔTg ( temperature drop through the gap) = 511.3 C
    ΔTC ( temperature drop through the clad) = 95 C
    ΔTC ( temperature drop through the coolant) = 18 C
    ..
    does this sounds correct because I sill don't have that sense of the numbers...!!
     
  2. jcsd
  3. Apr 15, 2012 #2

    Chronos

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    Gold Member

    Re: Reactor design

    I believe Westinghouse has some data you may find useful.
     
  4. Apr 15, 2012 #3

    Astronuc

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    Re: Reactor design

    188 fuel rods per assembly is close to the number for a 14x14 assembly, and a typical power level would be 13.6 MW/assembly for a 14x14 assembly, 14.5 to 25.7 MW/assembly for a 15x15 assembly, ~16.5 MW/assembly for a 16x16 assembly, and 17.7 to 18.6 MW/assembly for a 17x17 assembly. The smaller the diameter, the more limiting the heat flux for a given set of coolant flow rate, saturation temperature, coolant pressure and core hieght.

    Some published fuel design data are found in: http://www.neimagazine.com/journals/Power/NEI/September_2004/attachments/NEISept04p26-35.pdf [Broken]



    Some typical PWR fuel design numbers are:
    Code (Text):

    Lattice    LP   FR   GT  IT   FOD    GAP    CID    COD   PITCH  

     14x14   196  176  20   1  0.928  0.020  0.948  1.072  1.43
     15x15   225  204  20   1  0.928  0.020  0.948  1.072  1.43
     16x16   256  232  24   1  0.911  0.019  0.93   1.075  1.43
     17x17   289  264  24   1  0.819  0.017  0.836  0.95   1.263
     18x18   324  300  24   1  0.805  0.017  0.822  0.95   1.263
     
    LP = number of lattice positions
    FR = number of fuel rods
    GT = number of guide tubes
    IT = number of instrument tubes (optional) - one guide tube may serve as IT

    FOD = fuel pellet outer diameter (OD)
    GAP = diamteral fuel cladding gap
    CID = cladding inner diameter (ID)
    COD = cladding outer diameter (OD)
    PITCH = distance between centers of adjacent fuel rods. (need to verify for 16x16)

    For fuel rods, there is considerable variation for diametral/radial dimensions within the fuel rod, i.e., dimensions within the cladding envelope.

    Guide Tube OD ~ Pitch - spacer grid strip thickness

    Typical core height ~ 3.66 m (12 ft) or 4.27 m (14 ft)

    Plenum height ~ 18-20 cm for 3.66 m core

    See - EUR 20056 EN, Main Characteristics of Nuclear Power Plants in European Union and Candidate Countries
    ec.europa.eu/energy/nuclear/studies/doc/other/eur20056.pdf
    A1.3.6 Technical Data for REP900, page 84 of 163

    For REP 900 (3 coolant loops, 157 assembies/core)
    Code (Text):
    Core inlet temp         286°C
    Core outlet temp       323°C
    Tsat                       345°C
    Psat                       155 bar
    Core inlet pressure    157-158 bar
    Coolant mass flux      ~3500-3600 kg/m2-s
    Pressure drop - core  ~1.9 bar
     
    Typical fuel centerline temperature 900-1400°C
     
    Last edited by a moderator: May 5, 2017
  5. Apr 16, 2012 #4
    Re: Reactor design

    Thank you very much
     
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