Activation in a Co array

[Roberto]

Hi, 

I am really novice in SCALE usage, so my question is very basic. I would like to know how should I proceed to calculate a Co alloy activity device by knowing only average fast and thermal neutron flux (not a multigroup energy spectrum) in it for different reactor positions.

I imagine that, at least, I will need to generate one o more libraries so I´m interested in which steps should I follow to generate these libraries for using ORIGEN as much for SCALE 6.2.4, as for SCALE 6.3.1.

Thank you very much in advance, regards.

Roberto

[Willem van Rooijen]

Hello Roberto,

Your question is impossible to answer. What is a "Co alloy activity device"? It may be clear to you but not to others.

What kind of calculation are you trying to do, and what kind of boundary conditions apply? You mention that you "guess" that certain steps are needed, but without a detailed explanation of your goals it is impossible to help you.

Shinichi

2024年12月15日日曜日 1:40:48 UTC+9 Roberto:

[Roberto]

Hello Shinichi (o Willem),

Thank you for taking the time to give me an answer. I will try to clarify my needs.

I would like to calculate activation in a Co bar (my device) which could be put  in different reactor positions and I have neutron flux estimations (fast and thermal ones, only two values for position) for that Co bar when is put in those positions.

Let´ suppose that the Co bar weight 500 g, how should I calculate activity after an irradiation of 180 days, and then a decay of 30 days?

I have found an example in SCALE manual (https://scale-manual.ornl.gov/origen/origen-examples.html#create-an-origen-activation-library) quite similar, irradiating Fe, but based in a library (n200.data.reaction) already available in SCALE database, I suppose that this example could be useful for my case, but would this calculation be correct or I would need to develop an ad hoc library for my case? I have no experience in this field.

Thank you again.

Roberto

[Willem van Rooijen]

Hello Roberto,

I realize that it may be unexpected, but I would suggest a hand calculation.

Natural cobalt has only one isotope, Co-59. The main activation reaction is Co-59 -> Co-60. Under neutron irradiation the reaction Co-60 -> Co-61 will also occur, but given the half-life for Co-60 my physical intuition is that this is negligible.

In fact, you can set up the depletion equations and solve by hand for 4 to 5 isotopes. It is not so difficult. Alternatively you can use the general analytical solution for the general depletion problem, the so-called Bateman equation. Another option could be to use numerical software like SciPy to solve the ODEs numerically.

Obviously, you need to know the cross sections and the half-lives. In both cases you can use the JANIS software:

https://www.oecd-nea.org/jcms/pl_39910/janis

With JANIS you can select an isotope and calculate group XS. This calculation uses the same approach as NJOY, with an assumption for the flux. JANIS has cross sections at 293 K. You can request that JANIS makes 1 cross section for the entire energy range, or you can request 2 cross sections, one for the fast range and one for the thermal range. JANIS also contains information on radioactive decay.

There are also ways to do the work with SCALE but to be honest, I'd try the analytical solution + JANIS first.

Shinichi / Willem (I am technically Japanese and I have a Japanese name (Shinichi). I thought I had all settings in Google set up correctly but for some reasons sometimes the old settings, with my old name, take precedence.)

2024年12月20日金曜日 9:43:29 UTC+9 Roberto:

[Roberto]

Hello  Shinichi,

Thank you very much for your reply so descriptive and complete, I will have to go over a number of concepts previously to put it in practice, mainly as a verification exercise for a simple case; however I would like to learn how to do it properly with ORIGEN, mainly because what I´ve mentioned as a general case (a Co bar) it is a cluster of Co bars with different contents and neutron fluxes, so there are many cases to calculate/ analyze. 

Thanks again, best regards,

Roberto

[Willem van Rooijen]

Even if the problem concerns different geometrical entities, with different compositions and flux levels, the ODE does not change; only the initial conditions and the value of the flux change. Thus, if you write down the analytical general solution, all you need to do is substitute the relevant numerical values.

That being said, the description of your problem is still too vague to provide much guidance. It is possible to set up ORIGEN for your case, you can set the compositions either in terms of atomic densities, or number of moles, or even absolute mass, as long as you scale the value of the flux accordingly.

As far as cross sections are concerned, your description is difficult to understand. You say you have two values for the flux (presumably a "fast" flux and a "thermal" flux) and you need corresponding cross sections for each isotope. How exactly do you want to use those two cross sections? ORIGEN uses only oen cross section (i.e. "the" cross section) to solve the depletion problem. If you want to solve the depletion problem explicitly using two energy groups, you'll have to write your own solver.

See you,

Shinichi

2024年12月20日金曜日 21:31:02 UTC+9 Roberto:

[Roberto]

Shinichi,

As I´ve already commented, I will try an analytical solution for a simple case, I will need some time to go ahead in that direction.

My problem used to be solved in previous SCALE/ ORIGEN versions (6.0, 6.1 I guess), by using spectral parameters (THERM, RES and FAST), so I calculated these parameters with a cell/ supercell code, then I calculated thermal neutron flux with a core/ reactor code, in the needed positions and with only 4 values the ORIGEN activation problem was solved.

My compositions are in %w/w, that´s not a problem.

Yes, by following the process above mentioned, I have calculated cross sections for my Co device for two energy groups (above and below 0.625 eV), and with a core code I have calculated neutron fluxes in that device (2 average values, again above and below 0.625 eV).  

That being said, I hoped to apply a process using my results from cell/ supercell and core calculations that through ORIGEN let me calculate activity in my device, but it seems much more complex if I cannot use a pre-generated SCALE library.

Thank you again for your good will and time to help me.

Regards,

Roberto

[Willem van Rooijen]

Feel free to send a message if you need more help.

Good luck,

Shinichi

2024年12月22日日曜日 2:30:33 UTC+9 Roberto:

[Steve Skutnik]

Hello Roberto,

In the example you provide, ORIGEN is using the Library Builder capability (previously provided by COUPLE) to define a new neutron energy group spectrum to generate a new activation library. Here, the first 20 groups of the 200 group library are given a flat (uniform) profile, and the rest are zero, effectively approximating a uniform fast flux across the first 20 energy groups exclusively.

So there are a few ways to attack your own problem. One way would be is if you know the relative ratio of fast, epithermal, and thermal flux, you could coarsely approximate this using one of the MG libraries and define a flat profile within the energy bounds of the group structure. So let's say you have a ratio of 1 : 0..5 : 2 of fast : epithermal : thermal neutrons. Then you would find the group boundaries corresponding to your fast, epithermal, and thermal regions and create a "rough" spectrum like so:

build_lib("ff.f33") { 

  neutron(1) { 

  type=ENDF_ENERGY_DEPENDENT reaction_resource="n252.reaction.data"

  spectrum { type=MULTIGROUP flux=[71R 1.0 160R 0.5 21R2]  }

  }

}

Note that this is an extremely coarse approximation; you would be better served if you know the actual multi-group (e.g., 200-group, 252-group) energy spectrum.

An alternative would be to create a pin cell model using something like CSAS and create an AMPX working library from this; then you can use the calculated MG flux off of the library directly to build your F33 library for activation; see the training presentation on creating activation libraries with ORIGEN for more details.

Regards,

-Steve

[Roberto]

Hello Steve,

I really appreciate your complete reply with such a descriptive and clear approach for the approximate variant. It seems to me that the mentioned approach has similarities with calculations carried out with previous ORIGEN versions, based on spectral parameters: FAST, RES and THERM. I used to calculated these parameters with a cell/ supercell code: DRAGON and then they were applied in an ORIGEN calculation by knowing neutron thermal flux in the activate component. In fact I have currently calculated these parameters for my devices with Co, using a 172 energy groups WIMS library, thinking that they would be useful for this case.

I am also thankful for your description of a more precise alternative, with the material about activation libraries for ORIGEN, I really haven´t seen it before.

I hope to be able to realize my calculations in the new ORIGEN with all the information that you and other colleagues have send to me in reply to my questions. 

Thank you again, and happy seasons all of you!
Regards,

Roberto