5.1.8. Examples
The main problems solvable by the ORIGEN family of codes are enumerated below (with relevant components in parentheses).
Decay (ORIGEN)
Activation (ORIGEN)
Fuel irradiation (ARP+ORIGEN or ORIGAMI)
Emission spectra from decay (ORIGEN)
Processing, including batch/continuous chemical removal, isotopic feed, and stream blending (ORIGEN)
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
=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
=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.
=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.
=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.
'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
==================================================================================
= 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.
'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.
=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.”
=============================================================================================
= 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.
=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.
=origen
build_lib("origen.end7dec") {
nuclide {
type=NAMED_SET
list="complete_v6.2"
}
decay {
type=ENDF_DECAY
resource="${DATA}/origen_data/origen.rev03.decay.data"
}
}
end
5.1.8.9. Create an ORIGEN Reaction Library
Example 5.1.32 creates an ORIGEN reaction library from an
AMPX library created using the SCALE t-depl-1d sequence. This particular
sequence is used to create the ORIGEN library which contains all possible
reactions and decays between the standard set of nuclides used for all
Polaris and TRITON calculations. For this reason, allow_zero=yes is
important because it makes sure no reactions are eliminated. The spectrum
in this case is not important because Polaris and TRITON will overwrite the
values for each depletable zone at each time step.
=t-depl-1d parm=(addnux=4,bonami)
PWR pincell calculation to get transition definition skeleton
v7-56
read comp
'Fuel
uo2 1 den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end
wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600 end
h2o 26 den=0.6798 1 593 end
wtptbor 26 0.6798 1 5000 100 500e-6 593 end
end comp
read celldata
latticecell squarepitch hpitch=.83116 26 fuelr=0.47815 1 cladr=0.5588 25 end
end celldata
read depletion
1
end depletion
read burndata
power=40 burn=0 end
end burndata
read keep
microlibfile
end keep
read model
Infinite lattice PWR pin cell
read parm
prtflux=no collapse=yes prtmxsec=no
sn=4 inners=5 outers=100 epsouter=1e-6
epsglobal=1e-6
end parm
read materials
mix=1 pn=0 com='4.11% enriched fuel' end
mix=25 pn=0 com='cladding' end
mix=26 pn=0 com='water' end
end materials
read geom
geom=cylinder leftBC=refl rightBC=white
zoneIDs 1 25 26 end zoneIDs
zoneDimensions .47815 .55880 .83116 end zoneDimensions
zoneIntervals 3r8 end zoneIntervals
end geom
end model
end
=origen
build_lib("origen.transition.def") {
nuclide {
type=NAMED_SET
list="complete_v6.2"
}
decay {
type=ENDF_DECAY
resource="${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.
=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.
=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
5.1.8.12.1. Print the Cross Section Values on an ORIGEN Library with OBIWAN
The obiwan utility was introduced in SCALE 6.3 to help users interrogate
the data contained in ORIGEN binary files. From the command line, one
can view all coefficients on the library with the following. See
Sect. 5.2 for details.
obiwan view my.f33
5.1.8.12.2. Print the Cross Section Values on an ORIGEN Library in ORIGEN
Cross sections can also be printed when the library is used in ORIGEN by setting “print_xs=yes” in the options block, shown as Example 5.1.35.
=shell
cp ${INPDIR}/my_rx.f33 ft33f001
end
=origen
options{ print_xs=yes }
case{
lib{ file="ft33f001" pos=1 }
time=[1] %dummy time
mat{ iso=[u235=1] } %dummy iso
}
end
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.”
=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.”
=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.
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.
=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.
=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.
=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