Depletion of Spent Fuel

[Saehyun Choi]

Hi, I'm new to the SCALE and trying to teach myself.

I'm trying to calculate the actinide composition, k infinite, and maximum burnup of a single spent fuel bundle.

I was able to calculate the amounts of actinides (Pu, U) in the spent fuel bundle using ORIGAMI.

But I have no clue how to proceed to compute the multiplication factor and burnup.

The conditions are:

  1. UO2 fuel with different U isotope composition (U234, U235, U236, U238)

  2. Typical LWR (or PWR)

  3. Irradiate the fuel until the k∞ reaches 1.0 (In other words, determine the burnup when the fuel is discharged.)

The values I'm looking for are:

  1. The amount of (or the ratio of) actinides (Pu, U) in the spent fuel bundle.

  2. The value of the initial k infinite (k∞).

  3. The burnup (GWD/MTU) when the fuel is discharged.

Any help will be greatly appreciated. Thank you!

-----------------------------------------------------------------ORIGAMI input code--------------------------------------------------------------------------------------

=origami

% Case and identifier information

  title="RepU 1"

  prefix= RepU1

  

% Parameter options

  options{ 

           mtu=1.0

   stdcomp=yes

   decayheat=yes

   ft71=all

  }

% Array containing ORIGEN library names

  libs=[ "w17x17" ]

  

% Fuel composition (UO2-based fuel)

  fuelcomp{

    stdcomp(RepU){

      base=uo2

      iso[92234=0.088 92235=3.52 92236=1.14 92238=95.25]

      }

    mix(1){ comps[ RepU=100.0 ] }

  }

% Axial variation in water density (g/cc) corresponding to the axial power zones

  modz=[ 0.7332 ]

% Axial (Z) power shaping factors / fractional power distribution for the assembly

  pz=[ 1.0 ]

% Power history

% Does not need to have cycles. Can be just one cycle and 5-year cool down.

% Specific power (MW/MTU) for each cycle / Cycle length (days) / Number of libraries per cycle / downtime between cycles (days)

% Cooling time (days)

  hist[

    cycle{ power=40 burn=333.33 nlib=4 down=16.67 }

    cycle{ power=40 burn=333.33 nlib=4 down=16.67 }

    cycle{ power=40 burn=333.33 nlib=4 down=0 }

    cycle{ down=1825}

  ]

  

% Output edit options

  print{

        nuc{ units=[grams watts g-watts] sublibs=[ac] }

}

% Nuclides included in comp file (OPTIONAL: overrides default)

  nuccomp=[

     92232 92233 92234 92235 92236 92237 92238 92239 92240 92241

94236 94237 94238 94239 94240 94241 94242 94243 94244 94246

  ]

end

----------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

[Steve Skutnik]

Hello Saehyun,

Unfortunately there's not a built-in option for the neutron balance (k-inf) output in ORIGAMI; however, this feature is present in the more general Origen input. ORIGAMI is primarily designed to be a specialized interface to ORIGEN for spent fuel assembly depletion problems (i.e., representing axial and radial burnup gradients).

For Origen, you simply need to specify "kinf=yes" in the print options block, i.e.:

print{
   kinf=yes
}

As for the burnup, generally speaking both Origen and ORIGAMI will output the calculated burnups in the output automatically. In the case of ORIGEN, this is a bit more explicit; in the output, you'll see a table like this for each case:

=========================================================================================================================
=   History overview for case 'c01_down' (#3/8)                                                                         =
=   Cycle 1-->Cycle 2: Downtime (30 days)                                                                               =
-------------------------------------------------------------------------------------------------------------------------
   step            t0            t1            dt             t          flux       fluence         power        energy
    (-)           (d)           (d)           (s)           (s)     (n/cm2-s)       (n/cm2)          (MW)         (MWd)
      1    540.000000    540.001526  1.296000E+02  4.665613E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      2    540.001526    540.004578  2.678400E+02  4.665640E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      3    540.004578    540.013672  7.862400E+02  4.665718E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      4    540.013672    540.041199  2.376000E+03  4.665956E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      5    540.041199    540.123474  7.110720E+03  4.666667E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      6    540.123474    540.370422  2.133216E+04  4.668800E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      7    540.370422    541.111084  6.399648E+04  4.675200E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      8    541.111084    543.333313  1.919981E+05  4.694400E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
      9    543.333313    550.000000  5.760029E+05  4.752000E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
     10    550.000000    570.000000  1.728000E+06  4.924800E+07  0.000000E+00  6.664200E+21  0.000000E+00  1.080000E+04
          
              step - step index within this case
                t0 - time at beginning-of-step in input units
                t1 - time at end-of-step in input units
                dt - length of step in seconds
                 t - end-of-step cumulative time in seconds
              flux - flux in neutrons/cm^2-sec (CALCULATED)
           fluence - cumulative end-of-step fluence in neutrons/cm^2 (CALCULATED)
             power - power in mega-watts (INPUT)     
            energy - cumulative end-of-step energy released in mega-watt-days (INPUT)     
=========================================================================================================================

Here, "energy" is the cumulative energy release from the mass of material; the burnup would therefore just be this value divided by the material mass. (i.e., if you are using 1 MTU as your mass basis, then this quantity would directly represent the cumulative burnup in MWd/MTU. To get to GWd/MTU, just divide by 1000.)

Finally, as for calculating the inventories of U and Pu, there are a couple of ways you can go about this. One of the easiest ways would be to use the OPUS module to output the U and Pu masses following your depletion case. OPUS reads the saved concentrations from the binary F71 file in Origen & ORIGAMI (which is generated by specifying the "save" option). So for example, after your input, you'd have the following block:

=opus
data="name_of_your_FT71_file"
units=grams
libtype=act
symnuc=u pu end
end

Alternatively, if you save the F71, you can open it up interactively in the Fulcrum graphical user interface; this will allow you to view the concentrations for all times, including both as graphs and tables.