need help with COUPLE

[Peter Wolniewicz]

Dear Forum.

I would like to use COUPLE to specify a neutron spectrum to irradiate some material. 

I have briefly studied this thread and also the SCALE 6.2 manual:

https://groups.google.com/g/scale-users-group/c/zY8JZFJWBAA/m/uK0HroAQAgAJ

I am still confused about the FIDO input. Could you help me with examples on how to specify a neutron flux spectrum in COUPLE?

My secondary goal is to be able to sample changes in spectrum and see how that changes activation in different materials in ORIGEN. First step I think it is to specify a spectrum and later move on to something similar to SAMPLER+COUPLE.

If there are some study materials on how COUPLE works it would be much appreciated!

Thank you!

Peter Wolniewicz

SKB AB, Sweden

[Willem van Rooijen]

Your request is unclear. What exactly are you trying to do?

If you wish to create specific cross sections for a depletion calculation, it is (probably) easier to use TRITON to specify the reactor composition. A library with effective cross sections for ORIGEN is then also created. Copy the file with cross section data to your working directory at the end of your calculation. Then you can use ORIGEN with the cross sections from TRITON and perform a depletion calculation.

In most cases, it is not really helpful to only specify a weight spectrum for ORIGEN, since the self-shielding depends on the specific composition and spatial arrangement of the materials.

Let me know whether this helps.

Van Rooijen

2022年11月11日金曜日 20:13:00 UTC+9 Peter Wolniewicz:

[Peter Wolniewicz]

Thank you Willem,

I have previously modelled reactor internals in TRITON and produced f33/f71 and then used these in ORIGEN for activation and decay calculations. What I want to do now, is to activate materials when the neutron flux spectrum is known from a code that does not produce the f71/f33 binary files that ORIGEN needs. The only way to go in this case, as I understand it, is to use COUPLE where the irradiation spectrum can be specified. My question is: How do I specify the neutron flux spectrum using COUPLE so that I dont have to model everything in TRITON?

For example:

energy bins in eV:

0.01 1E4 1E6

neutron flux:

1E12 1E10 1E7

I would like to do somthing like the above in ORIGEN, so I guess COUPLE is the way to go. But how?

Peter Wolniewicz

[Willem van Rooijen]

Hello Peter,

Are you using the "new" ORIGEN-S (SCALE 6.2 and up) or the "old" ORIGEN-S (SCALE 6.0 and earlier)?

In (very) early versions of SCALE, ORIGEN-S was available with 3-group cross sections, the so-called "thermal", "resonance", and "fast" cross sections (THERM, RES, FAST in the ORIGEN manual). The user can "mix" these three cross sections by inputting a flux value for each of the three groups. I think one can set the three flux values in the ORIGEN-S input (at least in the "old" input format). The COUPLE code was traditionally used for two tasks: to translate the (ASCII formatted) "card image libraries" to binary (with 3 energy groups), or to update an existing binary library with data from an AMPX working library.

In the "new" ORIGEN-S it appears to be possible to manually input a flux spectrum (data block 1, 1$$ a18 NWGT). The manual states: "The flux spectrum must be given in order of descending neutron energy according to the convention that group 1 is the highest energy group. The group structure (number of groups and group boundaries) must be consistent with the ORIGEN reaction resource (Block1 0$$ JD)". I am unfamiliar with the new-fangled terminology (0$$ a3 JD "unit number of ORIGEN reaction resource (80)").

Bottom line: are you using "new" or "old" ORIGEN-S. If "new", it seems to be possible to create a custom ORIGEN-S library with user-supplied cross sections.

Cheers,

Van Rooijen

2022年11月17日木曜日 0:27:05 UTC+9 Peter Wolniewicz:

[Peter Wolniewicz]

Willem,

I am using SCALE 6.2.3 and =origen without the "s". 

The section in the 6.2.3 manual that contains "The flux spectrum must be given in order of descending neutron energy according to the convention that group 1 is the highest energy group. The group structure (number of groups and group boundaries) must be consistent with the ORIGEN reaction resource (Block1 0$$ JD)"  is related to COUPLE.

I find section "5.1.3.1 COUPLE Module" in the manual hard to follow and understand, and therefore I created this thread = ) 

In short, How do I use COUPLE to specify a flux spectrum? 

[Pavlo Ivanusa]

Attached is an example file of using COUPLE with a custom flux. It is using the 238 group structure with a user specified flux.

[Steve Skutnik]

One thing to add to Pavlo's good example; when using COUPLE, your energy group structure for the flux weighting is based on the energy groups of the multi-group library you are using for cross-sections. So for example, in Pavlo's example, he was using the ENDF-VII 238-group library. For SCALE-6.2, this has been superseded by a newer 252-group library (also based on ENDF/B-VII.1). The AMPX multi-group library used is referred to by a unit number; previously unit 80 pointed to the 238-group library; in 6.2, it's unit 74 that points to the 252-group library. (You can find these unit aliases in ${SCALE}/data/origen_filenames.) Going back to Pavlo's example, that means switching out the "80" in the 0$$ block with 74.

For the energy group boundaries for the 238-group and 252-group libraries, see Tables 10.1.8 and 10.1.9 in the SCALE manual; these list out the upper energy bounds for each group on the respective libraries.

[Peter Wolniewicz]

Thank you very much. I hope I understand things better now.

If I know the neutron flux spectrum in 252 groups, would the below adjusted code be fine to use? It does not give any error when I run it. I have changed from 238 to 252 groups in  block 2, 9** array, and changed to unit 74 in 0$$ array. 

/Peter

=couple

% All unit numbers are give in the ORIGEN Data Resoures chapter

% Unit 21 = END7DEC

% Each block contains one or more arrays followed by 't'

% Unit numbers are written with only a "number"

% Block 1 = Titles, Unit numbers and case controls

% 0$$ and 1$$ below belongs to Block1

% a3=JD, Unit number of origen reaction resource (80),

% a4=ND, Input ORIGEN binary file (21)

% a6=MD, Unit number output ORIGEN library file (33)

%

********************************************************************************

* cross sections from 238-group JEFF-3.1/A (unit 80) *

* collapsed using pwr fuel spectrum *

0$$ a3 74 a4 21 a6 33 e

1$$ a14 1 a18 252 e t

9**

3.1525E-05 8.3597E-05 1.4740E-04 1.7630E-04 5.4766E-04 9.7021E-03

2.7468E-02 9.5828E-02 2.6843E-01 1.4730E-01 7.3400E-01 5.4884E-01

1.6204E-01 6.9092E-01 6.1186E-01 2.0753E-01 9.1868E-02 7.4367E-02

1.4195E-01 1.2150E-01 2.4007E-01 1.9634E-01 2.1505E-01 5.6029E-02

7.5633E-02 4.5249E-02 1.5134E-01 2.9142E-01 2.9291E-01 3.7304E-02

2.9624E-01 1.1673E-01 1.0070E-01 2.2345E-01 1.2447E-01 1.0557E-01

6.5585E-02 7.8661E-02 3.5196E-01 3.6737E-01 4.7374E-01 3.8046E-01

1.8310E-01 2.8898E-01 1.6246E-01 3.3643E-02 8.2745E-02 2.4071E-02

1.7154E-01 1.1961E-01 3.2347E-02 8.4367E-02 3.0866E-01 1.3131E-01

2.6350E-01 1.7961E-01 2.0346E-01 1.0664E-01 1.8185E-01 2.6273E-01

2.4915E-02 1.3303E-01 8.6700E-02 7.2824E-02 2.2994E-02 1.1839E-01

8.7029E-02 1.7970E-02 1.5502E-01 1.0885E-01 1.8767E-01 1.0376E-02

1.1134E-01 3.2956E-01 3.3995E-02 9.5719E-02 7.1399E-02 2.5031E-03

4.2052E-02 1.5228E-02 2.2995E-01 1.3273E-02 1.3515E-02 3.3859E-02

3.4211E-02 5.5852E-02 4.9036E-02 9.4138E-03 2.7298E-02 2.8193E-02

3.3527E-02 1.0281E-02 3.3214E-02 1.7369E-02 5.1380E-02 1.3614E-02

1.4168E-02 1.4606E-02 9.6655E-03 1.4030E-02 2.0326E-02 1.2590E-02

1.8554E-02 1.6982E-02 1.7778E-02 6.1553E-03 1.2982E-02 7.4895E-03

8.8957E-03 1.1435E-02 1.1477E-02 6.9187E-03 2.2432E-02 7.8742E-03

2.0149E-02 4.3478E-02 4.7941E-02 5.1822E-02 2.3877E-02 8.9530E-03

2.1381E-02 1.2224E-02 3.8842E-02 2.8281E-02 2.7097E-02 1.9252E-02

2.0951E-02 3.0660E-02 3.7085E-02 1.5260E-02 6.4829E-02 4.4455E-02

5.0560E-02 5.2112E-02 6.9493E-03 4.4716E-03 4.7571E-04 7.9955E-03

1.3543E-02 3.3722E-02 2.7006E-02 2.0779E-02 7.1341E-02 2.9921E-02

2.6940E-02 4.4967E-02 1.3751E-02 7.0729E-03 4.5784E-03 1.5270E-02

1.5819E-02 1.5780E-02 1.6305E-02 1.7488E-02 1.6186E-02 1.4810E-02

1.7409E-02 1.8246E-02 2.5423E-02 1.3212E-02 1.8263E-02 2.1549E-02

2.2529E-02 2.3717E-02 2.4882E-02 1.4290E-02 1.4672E-02 1.4985E-02

1.5159E-02 1.5379E-02 7.7446E-03 7.7478E-03 7.6773E-03 7.4247E-03

2.8015E-03 2.6426E-03 2.4316E-03 2.1719E-03 1.8837E-03 1.6016E-03

1.3625E-03 1.1889E-03 1.0943E-03 1.0776E-03 1.1404E-03 1.2872E-03

1.5172E-03 1.8164E-03 2.1559E-03 6.7858E-03 8.2451E-03 9.1156E-03

9.7165E-03 2.0923E-02 2.2692E-02 2.4469E-02 2.6392E-02 2.8522E-02

1.5162E-02 1.5843E-02 3.3944E-02 3.7243E-02 4.1119E-02 4.5901E-02

2.4865E-02 2.6304E-02 2.7596E-02 3.0678E-02 3.4956E-02 4.9090E-02

6.8504E-02 9.4054E-02 1.2663E-01 1.6712E-01 2.1550E-01 2.7457E-01

1.3004E-01 1.3925E-01 1.4689E-01 1.5226E-01 1.5315E-01 1.4871E-01

1.3634E-01 5.7846E-02 1.3741E-01 1.2579E-02 8.9174E-03 2.5301E-03

1.3004E-01 1.3925E-01 1.4689E-01 1.5226E-01 1.5315E-01 1.4871E-01

1.3634E-01 5.7846E-02 1.3741E-01 1.2579E-02 8.9174E-03 2.5301E-03

1.8929E-03 7.1402E-04 5.5535E-04 4.0332E-04 1.7220E-04 8.6956E-05

7.8990E-05 5.0103E-05 3.7590E-05 1.4610E-06 3.7590E-05 1.4610E-06 e t

done

end

=shell

cp ft33f001 "${OUTDIR}/customspectrum.f33"

end

[Steve Skutnik]

Peter,

Yes, it looks like you have everything together; if you know your spectrum in the 252-group structure, everything you have there should be good (i.e., I see 252 entries for you 9** array and you've specified the correct library & number of groups).

-Steve

[Peter Wolniewicz]

Thanks for all the help.

I am a bit curious about the 3-group cross section. Has this feature been removed in Origen 6.2 and up? 

I am also a bit confused regarding the use of the words "cross section" vs "spectrum":

When I use couple with a custom flux spectrum, what is the unit of the entries - is it neutrons/cm2/s where the sum of all entries is equal to 1? The cross sections for every nuclide in the library are only produced after COUPLE has run and created the binary file?

What does "weighting" in "weighting spectrum" mean? 

Sorry about my confusion = )

[Steve Skutnik]

Hi Peter,

In SCALE 6.1 and up, the 3-group cross-sections have been supplanted by the 238-group and 252-group AMPX MG libraries based on the JEFF-3.1/A activation libraries.

With respect to COUPLE: what you are providing is a set of relative weights for the energy bins. So in effect this is unitless; all we are doing is assigning the relative weights for a flux-weighted collapse of multi-group cross-sections, i.e.:

\frac{  \sum{ \sigma_g \phi_g } }{ \sum{ \phi_g } }

When you use COUPLE, what you are doing is pulling in MG reaction cross-sections from the ORIGEN reaction resource data (i.e., the JEFF/3.1-A library). COUPLE uses this MG data with your provided weight factors (i.e., the "weighting spectra") to produce one-group, flux-weighted cross-sections that give you an equivalent reaction rate as a function of scalar flux.

-Steve

[Willem van Rooijen]

Perhaps a bit more background might be helpful.

In the multi-group method (MG), in the preparation of the MG cross section, the reaction rate (sigma * flux) is integrated over the energy range of one energy group. At this stage of the calculation the actual neutron flux spectrum is unknown, and one has to make an "educated guess". This educated guess is often called the "weight function". Thus the group cross section is defined as :

sigma_group = (integral of sigma(E) * weight_function(E)) / (integral of weight_function(E))

In effect, you are calculating a weighted average of the cross section. In SCALE, the AMPX software provides the weight function (as a function of energy) and the group-integrated values of the weight function are stored in the MG library (IIRC under MT1099, "group integral of the weight function"). The MG cross sections from AMPX are thus prepared with an "educated guess" and are  not specific to a particular problem.

For an actual reactor problem, nowadays you can use CENTRM to calculate a weight function as a function of energy for the specific problem at hand. Then, the MG cross sections from AMPX are replaced by the "actual" MG cross sections from CENTRM (modules WORKER, CRAWDAD, etc). Basically, you start with "MG cross sections prepared by AMPX with an educated guess for the flux", then perform a flux calculation in continuous energy, then re-calculate the MG cross sections. The updated MG cross sections are thus specific to one particular problem.

The MG theory thus has a "weighting function" to create the MG cross sections (in 238 or 252, or any other number of groups). One can also "collapse" such a set of MG cross sections. In this collapse, the MG flux is often referred to as "weighting spectrum", "weighting flux", etc. Thus in ORIGEN, the MG libraries are collapsed to one group, using a MG "weighting spectrum", which is calculated for one specific problem.

The flux spectrum in SCALE is normalized to "unit fission source", i.e. the summation over all groups of nu*sigma_f*flux = 1.

Regards,

Van Rooijen

2023年2月9日木曜日 22:40:33 UTC+9 Peter Wolniewicz:

[Peter Wolniewicz]

Thank you! Where can I see the energy group structure (energy bins) for the various libraries? I found some information online about "Scale 252 group structure" with energy bins but I dont know if it is correct or not.

Regards,

Peter Wolniewicz

[Steve Skutnik]

Peter,

Tables 10.1.8 and 10.1.9 in the SCALE manual have what you are looking for. Also note that the SCALE 6.3 manual is now online and can be viewed as a searchable web page (i.e., the tables you need can be found here)

)-Steve