5.1.8. Examples

The main problems solvable by the ORIGEN family of codes are enumerated below (with relevant components in parentheses).

  1. Decay (ORIGEN)

  2. Activation (ORIGEN)

  3. Fuel irradiation (ARP+ORIGEN or ORIGAMI)

  4. Emission spectra from decay (ORIGEN)

  5. Processing, including batch/continuous chemical removal, isotopic feed, and stream blending (ORIGEN)

  6. Unit conversions (OPUS)

Examples of the six variations above are contained in the following sections, except for fuel irradiation problems with ORIGAMI, as described in its own chapter.

5.1.8.1. Decay of 238U

Example 5.1.21 Decay of 238U
=origen
case{
    % use ENDF/VII-based decay library
    lib{ file="end7dec" }

    % create a material with 1 gram U-238
    mat{
        units=GRAMS
        iso=[u238=1.0]
    }
    time=[20L 1.0 1e9] %default units are days

    % save all information to f71
    save=yes
}
end

Example 5.1.21 illustrates using the “end7dec” binary decay library and decaying one gram of 238U for 109 days using the logarithmic array shortcut to put 20 logarithmically spaced values between 1.0 and 1e9 days.

5.1.8.2. 252Cf neutron Emission Spectrum

Example 5.1.22 252Cf neutron emission spectrum.
=origen

bounds{
    neutron=[ 2.000000e+07 6.376300e+06 3.011900e+06 1.826800e+06
            1.422700e+06 9.071800e+05 4.076200e+05 1.110900e+05 1.503400e+04
            3.035400e+03 5.829500e+02 1.013000e+02 2.902300e+01 1.067700e+01
            3.059000e+00 1.855400e+00 1.300000e+00 1.125300e+00 1.000000e+00
            8.000000e-01 4.139900e-01 3.250000e-01 2.250000e-01 1.000000e-01
            5.000000e-02 3.000000e-02 1.000000e-02 1.000000e-05 ]
}

case{
    title="Cf-252 decay"

    lib{ file="end7dec" pos=1 }

    time{
       units=YEARS
       t=[ 0.01 0.03 0.1 0.3 1 3 10 ]
    }

    mat{
       units=CURIES
       iso=[cf252=1.0]
    }

    %perform neutron calculation with defaults
    neutron=yes

    print{
       neutron{
           summary=yes
           spectra=yes
           detailed=yes
       }
    }
}
end

Example 5.1.22 illustrates decay of 1 Ci of 252Cf for ten years with calculation of the time- and energy-dependent neutron source. The case uses the binary decay library “end7dec.” The neutron energy group structure is defined in the bounds block with array “neutron.”

5.1.8.3. Simple Fuel Irradiation Plus Decay

Example 5.1.23 illustrates the input for irradiation of a fuel assembly for 200 days at 15 MW in case “irrad,” followed by decay for 5 years in the next case, “decay.” The ORIGEN library (f33) was prepared by ARP to have a single position with a burnup of 1500 MWd/MTU. The input material has been specified to have 1 MTU total with an enrichment of 4.0% in 235U. Note that there are consistency requirements between ARP and ORIGEN that cannot be checked by either module. Namely, the enrichment specified in ARP should be equivalent to the effective enrichment specified in the ORIGEN input, and the burnup-dependent cross section data on the ORIGEN library should be for the midpoint burnup of the case, which requires consistency between the operating history (power and time) and the initial isotopics heavy metal loading. In typical fuel depletion calculations, it is most convenient to specify a metric ton (106 grams) of initial heavy metal, such that the power in MW may be interpreted as MW/MTIHM. In the above example, the power history does not need to be constant but when combined with the time values should produce an average burnup of 1500 MWd/MTU in order that the cross sections interpolated by ARP in position 1 are valid.

Example 5.1.23 Simple fuel irradiation plus decay.
=arp
'library type
w17x17
'wt%
4.0
'number of cycles
1
'number of days per cycle
200.0
'cycle-average specific power (MW/MTU)
15.0
'number of interpolated cross section sets generated per cycle
1
'moderator density (g/cc)
0.723
'interpolated output ORIGEN library
w17x17_100d.f33
end
=origen
case(irrad){
    % use xs data at pos=1 corresponds to midpoint burnup (200 d * 15MW/MTU)/2
    lib {
         file="w17x17_100d.f33" pos=1
    }
    % 1 MT of enriched uranium
    mat {
        units=GRAMS
        iso=[u234=356 u235=40000 u236=184 u238=959460]
    }
    % power history (at least 4 steps for MATREX)
    time=[ 50 100 150 200 ] %default time in days
    power=[ 15 15 15 15 ] %power in MW
}
case(decay){
    time{
        units=YEARS
        start=0 %start time at 0 in this case for ease of input for t[]
        t=[0.1 0.3 0.9 1 2 3 4 5] %observe rule of threes
    }
    save{ file="discharge.f71" steps=[0 LAST] } %only save begin and end
}
end

It is important to note that with the MATREX solver, used by default, the recommendation for irradiation and decay of spent fuel is to use no fewer than four steps for the irradiation and begin the decay period with a time step on the order of weeks or a month, increasing the interval for each subsequent step by no more than a factor of three. In many continuation cases, such as the decay case described here, it is convenient to specify times starting from zero for the case with “start=0” in the time block.

5.1.8.4. Three Cycles of Irradiation Plus Decay

Example 5.1.24 is similar to the previous one, except there are three sets of burnup-dependent transition cross sections (positions) generated by ARP and used in ORIGEN, and there are three cases, one corresponding to each cycle. Neutron and gamma sources are generated and saved to the f71 file for the final decay case. The maximum burnup achieved is 60 GWd/MTIHM.

Example 5.1.24 Three cycles of irradiation plus decay
=arp
w17x17
4.0
3
500 500 500
40.0 40.0 40.0
1 1 1
0.723
ft33f001
end

=origen
bounds{ neutron="xn27g19v7.0"
        gamma=[1e+7 8e+6 6.5e+6 5e+6 4e+6 3e+6 2.5e+6 2e+6 1.66e+6 1.33e+6 1e+6
               8e+5 6e+5 4e+5 3e+5 2e+5 1e+5 5e+4 1e+4]}
case(c1){
    lib{ file="ft33f001" pos=1 }
    time=[8i 50 500]
    power=[10r40]
    mat{ iso=[u235=4e3 u238=960e3] }
}
case(c2){
    lib{ pos=2 }
    time=[8i 550 1000]
    power=[10r40]
}
case(c3){
    lib{ pos=3 }
    time=[8i 1050 1500]
    power=[10r40]
}
case(cool){
    time{ start=0 t=[20L 0.001 100] units=YEARS }
    save{ file="snf.f71" time_offset=1500 }
 gamma{ sublib=ALL brem_medium=UO2 }
 print{ neutron{ spectra=yes } }
 neutron{ alphan_medium=UO2 }
 }
 end

5.1.8.5. Load Isotopics from an f71 File

The “mat” block allows isotopics to be loaded from any position on an existing f71 file, as seen in Example 5.1.25. A table of contents is printed in the output file every time an f71 is read or written. Inspection of this table can help identify the appropriate position on the f71, as shown in Example 5.1.26, extracted from the output file.

Example 5.1.25 Load isotopics from an f71 file using the “mat” block and perform a follow-on decay.
'copy existing f71 file from input file directory to working directory
=shell
cp ${INPDIR}/discharge.f71 discharge2.f71
end

=origen
case(restart){
    % decay only library
    lib {
        file="end7dec" pos=1
    }
    % must know correct position
    mat {
        load{ file="discharge2.f71" pos=2 }
    }
    % continue timeline from previous case ending at 5 years
    time{
        units=YEARS
        start=5
        t=[10 20 40 80 160]
    }
    % append to file
    save{
        file="discharge2.f71" steps=[1 2 3 4 5]
    }
}
end
Example 5.1.26 Table of contents printed when accessing the f71 file.
==================================================================================
= Restart F71 File for case 'restart' (#1/1) =
----------------------------------------------------------------------------------
Data taken from position: 2
index time power flux fluence burnup libpos case step DCGNAB
    1 0.00000e+00 0.00000e+00 0.00000e+00 0.00000e+00 3.00000e+03 1 2 0 DC----
    2 1.57800e+08 0.00000e+00 0.00000e+00 0.00000e+00 3.00000e+03 1 2 8 DC----
D - state definition present
C - concentrations present
G - gamma emission spectra present
N - neutron emission spectra present
A - alpha emission spectra present
B - beta emission spectra present

The simple input in Example 5.1.27 can be used to print contents of an existing f71 file. The file is renamed so that it does not have extension “.f71.” This prevents the automatic copy back from the working directory to the input file directory. Note that the same is true with any ORIGEN libraries that have the extension “.f33.” There is a rule in place that any file in the working directory with extension “.f71” or “.f33” is copied back to the input file’s directory. To prevent this from occurring, the “.f71” or “.f33” extensions must not be used in the filename, or unneeded files with a shell command at the end must be explicitly deleted.

Example 5.1.27 Isotopics from an f71 or f33 file.
'avoid .f71 extension to prevent automatic copy back
=shell
cp /path/to/unknown.f71 f71
end

=origen
case(test){
    lib{ file="end7dec" pos=1 }
    mat{ load{ file="f71" pos=1 } }
    time=[1]
}
end

'remove any *.f71 or *.f33 in the working directory to prevent automatic copy back
'(not necessary in this example, but for reference)
=shell
rm -f *.f71
rm -f *.f33
end

5.1.8.6. Continuous Feed and Removal

Both continuous feeding of nuclides into a system and chemical removal of elements from a system are required to simulate molten fuel systems such as the molten salt reactor (MSR). Simulation of removal can be used to represent other physical processes such as purification systems (i.e., removal of chemical species by filtration or ion-exchange columns) and ventilated systems in which the removal can be represented using a rate constant (1/s). Example 5.1.28 applies to simulation of a molten salt reactor system, and it uses both continuous feed of 232Th and simultaneous removal of fission products according to their removal process. To actually simulate this system properly, an ORIGEN reactor library appropriate for the MSR must be generated.

There are 11 groups of nuclides in this example, each with the same removal constant. The groups, by element and removal rate in units of (1/s), are

  • Group 1: Ca to As (3.37 \(\times\) 10-9);

  • Group 2: Y, La, Ce, Pr, Nd, Pm, Sm, Gd (2.31 \(\times\) 10-7);

  • Group 3: Eu (2.31 \(\times\) 10-8 );

  • Group 4: Se, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Sb, Te (5 \(\times\) 10-2);

  • Group 5: Zr, Cd, In, Sn (5.79 \(\times\) 10-8) ;

  • Group 6: Kr, Xe (5 \(\times\) 10-2);

  • Group 7: Br, I (1.93 \(\times\) 10-7);

  • Group 8: Rb, Sr, Cs, Ba (3.37 \(\times\) 10-9);

  • Group 9: Th, Li, Be, F (3.37 \(\times\) 10-9);

  • Group 10: Pa (3.86 \(\times\) 10-6); and

  • Group 11: Np, Pu, Am, Cm, Bk, Cf (1.98 \(\times\) 10-9).

A 232Th feed rate of 2.0 \(\times\) 10-2 grams/s is used.

Example 5.1.28 Demonstration of continuous feed and removal within an ORIGEN irradiation case.
=origen
case{
    title="Single fluid MSR depletion calculation"

    lib{
       file="msr.f33"
       pos=1
    }
    time{
        units=YEARS
        t=[ 8i 0.05 1.0 ]
    }

    power=[ 10r30 ] %30 MW

    %initial material
    mat{
       %FLiBe with pure Li7 and 1 MTIHM loading
       units=GRAMS
       iso =[f=1e7 li7=5e6 be=1e6 th232=9.5e5 u233=0.5e5]

       %continuous feed of th-232
       feed=[th232=2e-2] %g/s
    }

    %continuous removal by element using atomic numbers
    processing{ removal{rate=3.37e-9 ele=[12i 20 33]}
                removal{rate=2.31e-7 ele=[39 57 58 59 60 61 62 64]}
                removal{rate=5.79e-8 ele=[40 48 49 50]}
                removal{rate=5e-2 ele=[36 54]}
                removal{rate=1.93e-7 ele=[35 53]}
                removal{rate=3.37e-9 ele=[37 38 55 56]}
                removal{rate=3.37e-9 ele=[90 3 4 9]}
                removal{rate=3.86e-6 ele=[91]}
                removal{rate=1.98e-9 ele=[93 94 95 96 97 98]}
    }

    print{
         cutoffs=[ GRAMS=0.1 ] %do not show grams < 0.1% of total
         nuc{ units=[GRAMS] total=yes }
         kinf=yes   %print k-infinity summary
         absfrac_sublib=LT %absorption fractions for light nuclides in FLiBe
    }
} %end case
end

The absorption rates, fission rates, and k-infinity values are printed during irradiation, which can be used to evaluate the influence of the feed rates and removal constants on the time-dependent reactor performance. For the MSR in particular, the ability to self-sustain can be assessed from the “k-infinity” summary output, enabled by “kinf=yes” in print block (as well as the absorption fractions in the light nuclides in FLiBe), and enabled by “absfrac_sublib=LT” in the print block (Example 5.1.29). This example uses pure Li-7, whereas if natural Li is used, one will see a much larger fraction of absorptions in Li-6 and much lower “k-infinity.”

Example 5.1.29 Continuous feed and removal–history overview.
=============================================================================================
= History overview for case '1' (#1/1)
= Single fluid MSR depletion calculation
---------------------------------------------------------------------------------------------
   step          t0          t1          dt           t        flux     fluence       power      energy
    (-)         (y)         (y)         (s)         (s)   (n/cm2-s)     (n/cm2)        (MW)       (MWd)
      1      0.0000      0.0500  1.5780E+06  1.5780E+06  1.9782E+13  3.1217E+19  3.0000E+01  5.4792E+02
      2      0.0500      0.1556  3.3313E+06  4.9093E+06  2.0112E+13  9.8216E+19  3.0000E+01  1.7046E+03
      3      0.1556      0.2611  3.3313E+06  8.2407E+06  2.0575E+13  1.6676E+20  3.0000E+01  2.8613E+03
      4      0.2611      0.3667  3.3313E+06  1.1572E+07  2.1057E+13  2.3691E+20  3.0000E+01  4.0181E+03
      5      0.3667      0.4722  3.3313E+06  1.4903E+07  2.1557E+13  3.0872E+20  3.0000E+01  5.1748E+03
      6      0.4722      0.5778  3.3313E+06  1.8235E+07  2.2075E+13  3.8226E+20  3.0000E+01  6.3315E+03
      7      0.5778      0.6833  3.3313E+06  2.1566E+07  2.2613E+13  4.5759E+20  3.0000E+01  7.4882E+03
      8      0.6833      0.7889  3.3313E+06  2.4897E+07  2.3171E+13  5.3478E+20  3.0000E+01  8.6449E+03
      9      0.7889      0.8944  3.3313E+06  2.8229E+07  2.3750E+13  6.1390E+20  3.0000E+01  9.8016E+03
     10      0.8944      1.0000  3.3313E+06  3.1560E+07  2.4351E+13  6.9502E+20  3.0000E+01  1.0958E+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)
=============================================================================================


=============================================================================================
=   Overall neutron balance for case '1' (#1/1)
=   Single fluid MSR depletion calculation
---------------------------------------------------------------------------------------------
                       0.0E+00y    5.0E-02y    1.6E-01y    2.6E-01y    3.7E-01y    4.7E-01y
n-production           2.9579E+18  2.9272E+18  2.9110E+18  2.9123E+18  2.9138E+18  2.9154E+18
n-absorption           2.6018E+18  2.6178E+18  2.6858E+18  2.7707E+18  2.8583E+18  2.9489E+18
k-inf                  1.1369E+00  1.1182E+00  1.0838E+00  1.0511E+00  1.0194E+00  9.8864E-01
=============================================================================================

=============================================================================================
=   Fraction of absorption rate for light elements for case '1' (#1/1)
=   Single fluid MSR depletion calculation
---------------------------------------------------------------------------------------------
            0.0E+00y    5.0E-02y    1.6E-01y    2.6E-01y    3.7E-01y    4.7E-01y
f-19        9.5276E-02  9.4189E-02  9.2293E-02  9.0506E-02  8.8782E-02  8.7113E-02
be-9        9.4833E-02  9.3751E-02  9.1862E-02  9.0083E-02  8.8366E-02  8.6704E-02
li-7        6.3293E-02  6.2572E-02  6.1312E-02  6.0125E-02  5.8980E-02  5.7871E-02
li-6        0.0000E+00  5.6409E-04  1.7104E-03  2.8000E-03  3.8355E-03  4.8190E-03
he-3        0.0000E+00  2.9340E-08  7.1009E-07  3.0708E-06  7.9374E-06  1.5939E-05
o-16        0.0000E+00  4.3552E-08  1.3502E-07  2.2606E-07  3.1678E-07  4.0728E-07

*{list continues}*

5.1.8.7. Calculate Fuel \(\left(\alpha,n \right)\) Emissions in a Glass Matrix

Batch processing options are provided to separate various components of the nuclide compositions into different streams and to recombine the streams to form new compositions. This example applies to the irradiation of typical commercial fuel and subsequent storage of separated fission products and actinides from fuel in a glass matrix. The matrix is important in determining the \(\left(alpha,n\right)\) component of the neutron source because the alpha particles interact with the light element constituents in the matrix, with \(\left(alpha,n\right)\) yields corresponding to the medium containing the \(\alpha\)-emitting nuclides. Therefore, an accurate calculation of the neutron source in a glass matrix requires combining the oxide fuel compositions after irradiation with the defined glass matrix. The calculation could be performed by creating a case by manually entering the calculated nuclide activities and the matrix composition. However, this is only practical if the number of source nuclides is small. This example applies batch processing and blending options to generate the required compositions.

The blending option is used to combine two streams: one stream from irradiated fuel, and the other stream defining the glass matrix composition.

An irradiation case—”irrad”—is performed first to generate spent nuclear fuel compositions. The next case “spent” decays the results for one year. At the start of the decay, only selected elements are retained in the stream by performing processing with the “retained” array. In this example, all elements are removed except for Se (99.8%); Rb, Sr, Te, Cs, Ba, Dy (77.8%); and U, Np, Pu, Am, Cm (1%).

The glass matrix compositions are then defined in the third case, “glass.”

The final case—”blend”—blends 10% of each the last step’s isotopics from the “spent” and “glass” cases. To test the dependence on the time when the blend is performed, the blend can be changed to “blend=[spent(N)=0.1 glass=0.1 ],” where N is the index of the step from which to take isotopics from the spent case.

Example 5.1.30 Continuous feed and removal–blending option.
=arp
w17x17
3
1
360
40
1
0.723
fuellib
end

=origen
bounds{
    neutron=[ 1.00E-05 1.00E-02 3.00E-02 5.00E-02 1.00E-01 2.25E-01
              3.25E-01 4.14E-01 8.00E-01 1.00E+00 1.13E+00 1.30E+00
              1.86E+00 3.06E+00 1.07E+01 2.90E+01 1.01E+02 5.83E+02
              3.04E+03 1.50E+04 1.11E+05 4.08E+05 9.07E+05 1.42E+06
              1.83E+06 3.01E+06 6.38E+06 2.00E+07 ]
}
case(irrad){
    title="Fuel Stream 1 Irradiation"
    lib{ file="fuellib" pos=1 }
    time=[ 8I 36 360 ]
    power=[ 10r40 ]
    mat{
        units=GRAMS
        iso=[u234=534 u235=60000 u236=276 u238=939190]
    }
}
case(spent){
    title="Fuel Stream 1 Decay"
    time{
        t=[ 0.1 0.3 1 3 10 30 100 300 360 ] start=0 %enter times from 0
    }
    processing {
        retained=[se=0.998 rb=0.778 sr=0.778 te=0.778 cs=0.778 ba=0.778
                  dy=0.998 u=0.010 np=0.010 pu=0.010 am=0.010 cm=0.010]
    }
}
case(glass){
    title="100 kg glass" time{ t=[ 1 ] start=0 }
    mat{
        units=GRAMS
        iso=[li=2.18e3 b=2.11e3 o=46.4e3 f=0.061e3 na=7.65e3 mg=0.49e3 al=2.18e3
             si=25.4e3 cl=0.049e3 ca=1.08e3 mn=1.83e3 fe=8.61e3 ni=0.70e3
             zr=0.88e3 pb=0.049e3 ]
    }
}
case(blend){
    title="final blended case"  time=[ 1.01 3 10 30 100 ] %continue previous
    mat{
        blend=[ spent=0.1 glass=0.1 ] %blend factors of 0.1 for each
    }
 neutron{ alphan_medium=CASE } %use this case's isotopics
 print{ ele{ total=yes units=[GRAMS] }
        neutron{ summary=yes spectra=yes detailed=yes }
 }
}
end

5.1.8.8. Create an ORIGEN Decay Library from a Decay Resource

The following ORIGEN input (Example 5.1.31) is all that is required to produce a binary decay library from the ORIGEN decay resource file. Note that ORIGEN decay resources can also be raw ENDF formatted data files.

Example 5.1.31 Creation of an ORIGEN decay library from the decay resource
 =origen

 build_lib("origen.end7dec") {
   nuclide {
     type=NAMED_SET
     list="complete_v6.2"
   }
   decay {
     type=ENDF_DECAY
     resource="${DATA}/origen_data/origen.rev03.decay.data"
   }
 }

 end

5.1.8.9. Create an ORIGEN Reaction Library

Example 5.1.32 creates an ORIGEN reaction library from an AMPX library created using the SCALE t-depl-1d sequence. This particular sequence is used to create the ORIGEN library which contains all possible reactions and decays between the standard set of nuclides used for all Polaris and TRITON calculations. For this reason, allow_zero=yes is important because it makes sure no reactions are eliminated. The spectrum in this case is not important because Polaris and TRITON will overwrite the values for each depletable zone at each time step.

Example 5.1.32 Creation of an ORIGEN reaction library.
 =t-depl-1d parm=(addnux=4,bonami)
 PWR pincell calculation to get transition definition skeleton
 v7-56

 read comp
 'Fuel
 uo2  1 den=10.412  1  900 92234 0.04 92235 4.11  92238 95.85   end
 wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600  end
 h2o  26 den=0.6798  1  593 end
 wtptbor 26 0.6798 1 5000 100 500e-6 593 end
 end comp

 read celldata
   latticecell squarepitch hpitch=.83116 26 fuelr=0.47815 1 cladr=0.5588 25 end
 end celldata

 read depletion
   1
 end depletion

 read burndata
   power=40  burn=0 end
 end burndata

 read keep
   microlibfile
 end keep

 read model
 Infinite lattice PWR pin cell
 read parm
  prtflux=no collapse=yes prtmxsec=no
  sn=4 inners=5 outers=100 epsouter=1e-6
  epsglobal=1e-6
 end parm
 read materials
  mix=1 pn=0 com='4.11% enriched fuel' end
  mix=25 pn=0 com='cladding' end
  mix=26 pn=0 com='water' end
 end materials
 read geom
   geom=cylinder leftBC=refl rightBC=white
   zoneIDs 1 25 26 end zoneIDs
   zoneDimensions .47815 .55880 .83116 end zoneDimensions
   zoneIntervals 3r8 end zoneIntervals
 end geom
 end model
 end

 =origen

 build_lib("origen.transition.def") {

   nuclide {
     type=NAMED_SET
     list="complete_v6.2"
   }

   decay {
     type=ENDF_DECAY
     resource="${DATA}/origen_data/origen.rev03.decay.data"
   }

   neutron(1) {
     type=ENDF_ENERGY_DEPENDENT
     reaction_resource="${DATA}/origen.rev01.jeff56g"
     fp_yield_resource="${DATA}/origen_data/origen.rev05.yield.data"

     %%%%%%%%%%%%%%
     allow_zero=yes
     %%%%%%%%%%%%%%

     spectrum {
       type=MULTIGROUP
       flux=[56R 1.0]
     }

     xs_update {
       type=AMPX_LIBRARY
       file="sysin.microWorkLib_0.f44"
       mixture=1
     }

   }
 }

 end

 =shell
 cp origen.transition.def ${OUTDIR}/pwr.rev04.orglib
 end

5.1.8.10. Create an ORIGEN Activation Library

Activation calculations typically do not require self-shielding, so an ORIGEN “activation” library is a means to refer to a library that is using infinitely dilute cross sections from the reaction resource. An ORIGEN activation library can be created very easily as long as the flux spectrum is available in one of the reaction resource group structures.

In the example below, we create a fast spectrum library based on the 200-group reaction resource and irradiate one gram of iron for 10 days at 1e15 flux.

Example 5.1.33 Creation of an ORIGEN activation library.
 =origen
 build_lib("ff.f33") {
     neutron(1) {
         type=ENDF_ENERGY_DEPENDENT
         reaction_resource="n200.reaction.data"
         spectrum {
           type=MULTIGROUP
           flux=[20r 1.0 180r 0.0]
         }
     }
 }
 case {
   lib {file="ff.f33"}
   mat { iso=[fe=1.0] units=GRAMS }
   time=[10]
   flux=[1e15]
 }
 end

5.1.8.11. Create an ORIGEN Library with User-Supplied Cross Sections

It is possible to change specific cross sections on ORIGEN libraries. The following input uses all default values for all data (the flux spectrum from the reaction resource will be used) but changes the U-238 fission (18) and n,gamma (102) reactions to 20 and 10 barns, respectively.

Example 5.1.34 Creation of an ORIGEN library with user-supplied cross sections.
 =origen

 build_lib("my.f33") {
   neutron(1) {
     type=ENDF_ENERGY_DEPENDENT
     reaction_resource="n252.reaction.data"
     coeff_update [
         u238 18  20
         u238 102 10
     ]
   }
 }

 end

5.1.8.12. Printing library cross-section values

5.1.8.13. Ranking Contribution to Toxicity

The OPUS input in Example 5.1.36 creates a plot of the volume of various nuclides with the maximum permissible concentration (MPC) in water using the “snf.f71” produced by the “Three cycle plus decay” example (Example 5.1.24). Only positions 6–16 are plotted using “minposition” and “maxposition.” Three nuclides (242Cm, 137mBa, and 99Tc ) are forced to be included via the “symnuc” list, with 17 more nuclides (for total “nrank=20”) included according to their average rank in terms of the MPC in water. The total number of nuclides requested is 20. The x-axis label is set to “TIME (YEARS),” and the title is “SPENT FUEL AT 60 GWD/MTHM.”

Example 5.1.36 Using OPUS to produce a ranked contribution to radiotoxicity
=shell
  cp ${INPDIR}/snf.f71 f71
end

=opus
data="f71"
typarams=nucl
units=h2om**3
time=year
libtype=fisact
minposition=6  maxposition=16   nrank=20
title="SPENT FUEL AT 60 GWD/MTIHM "
xlabel="TIME (YEARS)"
symnuc=cm-242 ba-137m tc-99 end
end

5.1.8.14. Spectrum plots with OPUS

5.1.8.14.1. Photon Spectrum Plot

The OPUS input shown in Example 5.1.37 creates a plot of the photon spectrum for all times between 1 and 5 years, using “tmin=1” and “tmax=5” with “time=years.”

Example 5.1.37 Plotting the photon spectrum using OPUS.
=shell
cp ${INPDIR}/snf.f71 f71
end

=opus
data="f71"
typarams=gspectrum
units=intensity
tmin=1 tmax=5 time=years
end

Example 5.1.38 shows the output of the time-dependent gamma spectrum extracted from the f71 file by OPUS.

Example 5.1.38 Photon spectrum plot from OPUS– output of the time-dependent gamma spectrum.
            1/(s.MeV) |        1.25y       2.15y       3.73y
----------------------+-------------------------------------
 1.00e+01 -- 8.00e+00 |   2.7961e+11  3.0661e+03  3.0661e+03
 8.00e+00 -- 6.50e+00 |   7.1201e+13  1.9436e+04  1.9436e+04
 6.50e+00 -- 5.00e+00 |   8.9725e+15  2.5300e+10  5.3041e+09
 5.00e+00 -- 4.00e+00 |   3.7075e+16  9.1643e+12  1.9001e+12
 4.00e+00 -- 3.00e+00 |   1.1734e+17  1.3562e+14  2.8824e+13
 3.00e+00 -- 2.50e+00 |   3.4321e+17  5.5945e+15  4.8216e+15
 2.50e+00 -- 2.00e+00 |   5.2847e+17  1.1338e+16  5.9433e+15
 2.00e+00 -- 1.66e+00 |   8.3528e+17  3.3015e+16  1.8382e+16
 1.66e+00 -- 1.33e+00 |   1.8060e+18  2.7233e+17  2.4658e+17
 1.33e+00 -- 1.00e+00 |   2.4156e+18  1.4238e+17  9.2264e+16
 1.00e+00 -- 8.00e-01 |   5.0662e+18  3.2193e+17  2.6845e+17
 8.00e-01 -- 6.00e-01 |   6.4083e+18  2.1010e+18  1.8903e+18
 6.00e-01 -- 4.00e-01 |   7.1803e+18  1.1822e+18  1.0301e+18
 4.00e-01 -- 3.00e-01 |   8.7297e+18  1.7856e+18  1.6414e+18
 3.00e-01 -- 2.00e-01 |   1.9481e+19  1.0658e+19  9.7018e+18
 2.00e-01 -- 1.00e-01 |   2.9105e+19  1.5198e+19  1.3996e+19
 1.00e-01 -- 5.00e-02 |   6.2160e+19  9.1756e+18  8.3611e+18
 5.00e-02 -- 1.00e-02 |   1.1219e+20  3.4628e+19  3.1458e+19

5.1.8.14.2. Neutron Spectrum Plot

The OPUS input in Example 5.1.39 creates a plot of the total neutron spectrum at all times.

Example 5.1.39 Neutron spectrum plot using OPUS
=opus
data="f71"
typarams=nspe
units=intensity
end

5.1.8.15. Isotopic Weight Percentages for Uranium and Plutonium During Decay

The isotopic distributions in uranium and plutonium, in weight percent, may be plotted with the OPUS input in Example 5.1.40.

Example 5.1.40 Isotopic weight percentages for uranium and plutonium during decay.
=opus
data="f71"
units=wpel
libtype=act
symnuc=u pu end
end

5.1.8.16. User-Specified Response Function in OPUS

In Example 5.1.41, response conversion factor data are entered in the RESPONSE= array as nuclide-response factor pairs, with the nuclide id being the legacy nuclide ID (ZZAAAI), with ZZ two digits of atomic number, AAA three digits of mass number, and I one digit of isomeric state. For example, 235mU would be given as 922351 and H-1 as 10010. In this example, it is assumed that the response factors are activity based (response/Bq of nuclide), so the units of Becquerels are requested using the UNITS keyword. The user-supplied response conversion factors are applied to the Becquerel units for all nuclides in the RESPONSE= array, and the results for any nuclide for which no response factors are provided are zeroed.

Example 5.1.41 Example of a user-specified response function in OPUS.
=opus
data="f71"
units=becq
time=year
typarams=nucl
response=
 270600 1.92027E+00  962460 2.27711E-01  481151 1.64222E-01
 822100 5.19246E-12  962470 1.75728E-09  501230 3.81237E-02
 832101 9.43341E-09  962480 1.05672E+00  511240 1.53438E+02
 882260 1.51600E-05  962500 1.60583E+02  511250 1.81023E-04
 892270 1.67080E-05  982490 2.21420E-04  511260 5.13360E+00
 902280 2.06832E-02  982500 2.87199E+02  end
end