The problem with this question is that there is no direct conversion from neutron flux to heat. Radiation heating when neutrons are around include: (1) the source of the neutrons, fission (f7 card), or (alpha,n) reactions in something like Be; (2) fast neutrons slowing down where the kinetic energy is transferred (f6:n card); and (3) neutron absorption resulting in nuclear reactions, where the energy released from the reaction needs to be followed. Hence there is no one card.

Hi Albert, I know that the people do , somehow, a conversion to transform the neutron flux into heat flux, on MCNP. Taking into account what you just said, how should I proceed to have the heat flux starting from the neutron one, on MCNP?

I don't know what other people do with that conversion. Be sure that when you do find out what other folks do with the conversion, make sure it is relevant to the problem on which you are working because MCNP will give you an answer.

yes, for sure. By now I'm only interested in the Neutron flux and I already have it. But power densities profile are also interesting for me and that's why I was asking, cause some codes ask us on the input to insert the q''(W/cm2) or q'''(W/cm3). I thought would be a card which we could straight find these variables from the neutron flux (F4).

Thank you.

Well, heat flux is another entity; however the f6 card gives you the heat deposition in MeV/g for a specific particle (f6:n, f6:p, etc). The +f6 card gives you energy deposition from all particles interactions in MeV/g. The neutron flux f4 can be converted to MeV/g with the fm card using the right constants. Converting MeV/g to W/cm3 is straight forward. Now, W/cm2 or heat flux (W/cm2-sec) depends on the thermal conductivity of the material(s), cooling, temperature distribution, etc; as such cannot be given by MCNP, if somebody does that is wrong.

What about the *Fmesh4, that is suppose to give us in Mev/cm2? I also can convert it into W/cm2 using the power normalization. Could you give a line example with the F4 and also a Fm to convert the neutron flux into W/cm3? Thank you.

try to get a grip of flux meaning flow, movement and flux as a population of particles in space. To have heat produced by particles means collisions and you don't have collisions at a zero thickness surface but particles crossing. Anything/cm2 in MCNP is related to movement of particles not heat or other related quantities. About the f4, please, look at the manual.

I know that, Itacil. You didn't get my point. Thank you anyway.

Quite an answer... Is there a point in that question?

For sure. Summing up, using the Fmesh4 I can calculate the neutron flux within some cells. putting this Fmesh4 together with a FM4 card, I am suppose to obtain the power density, for example, using:

FM4 C M R1 R2...

I was asking the values for this constants on the FM4 card (C, R1 and R2). That was the original question as long as I wasn't with the manual.

This is a different question...

Normally , what you need as input for PARET code are peaking factors. so we calculate this factors along the fuel elements by using this relation:

Local_FM4 / Average_FM4 (which is proportional to (local_power/Average_power))

how is defined the Avarege_FM4? is the average among the Fuel Elements only or considering the entire core (reflectors, etc)? How is that defined? Thanks. And one more question, if I have the flux already (Fmesh4), how would be the Fm4 card to convert ir into the power density (Mev/cm3)?

The average by considering only fuel elements which is considered as therm source (that generates a power)

for the second question you can use in MCNP the F6 card (Track length estimate of

energy deposition) as described in the following document. http://www.du.ac.ir/fa/contents/MCNP/SAH%20Feghhi/09.pdf

I've seen that one. But I really need to keep using Fmesh4, as long as I already have the flux profile using it. So, the only thing is missing is basically the Fm4 for it:

Fm4 C m -5 -6

Let's say that C= 180Mev/fission, m is the material when is occurring the heat generation, -5 is gamma ray production microscopic cross-section and -6 is the total fission microscopic cross-section. Doing this, it is only missing the fact that -5 and -6 are micro not macro. So, we still need to multiply it by the atomic density, N (That's suppose to be the atomic density of U-235, right?). Is any problem solving like this?

Calculating the power density q''', we need to follow the formula:

q'''(r)=G*integral(N_f (r)*Sigma_f(E)*phi(r,E)dE).

where G is ~180Mev/fission, N_f(r) is the atomic density of the fissionable isotope, Sigma_f is the microscopic fission cross section and phi, the neutron flux.

When we have a fuel material made with difference elements, N_f is the atomic density of the fissionable one, for example, using fuels like U3Si2-Al, in general, only U235 is fissionable by thermal neutrons, so, N_f = N_f from U235.

However, on MCNP, de microscopic cross section is the total one, not only that one from U235. So, Do I need to enter as the N_f the U235 atomic density or the U3Si2-Al atomic density? Does MCNP know which material is fissionable? That's the question I have right now.

Thanks!

Very confusing, right? If you read the manual: fm4 N n -6 -8 where N is the number density, n the material number, -6 total fission cross section, -8 fission Q (MeV/fission) (space between -6 and -8 makes the code to multiply them) this gives you the fission energy deposition(f7 card) and it is not exactly f6. Making simpler fm4 N n 1 -4, where "1" is the total cross section; -4 is the average heating number => this associated with f4:n duplicates the f6 card.

I've put to run: Fm4 C 21 -6 -8

where C has the factor to convert the MCNP's F4 into real flux (n/cm2.s, basically a power normalization), plus the factor to convert Mev/s into W. And now, finally, I'm suppose to have the W/cm3 generated in the region I'm interested in, right?

No, read above. -6 -8 is not the total heat deposited and your normalization does not make sense.

The normalization was correct and it's not the problem (we have from MCNP particle/cm2 as the result from Fmesh4 and, once we know the reactor power (in this case 5MW), we can figure out the fission rate and convert particle/cm2 into n/cm2.s ( on average). Might be another way to do that straight from MCNP (using another FM card), but I don't know yet. Could you please then just write down the Fm4 card to get the Fmesh4 particle/cm2 result and change it into the power density ( in Watts/cm3)? it Is all I need by now and I'm not sure whether I'm doing correctly on MCNP or not (quite sure I'm not). I will appreciate the help.

C=N*density*conversion(MeV to Joules)

N=atoms/g*1e-24

density=density of the material (g/cm**3)

conversion=conversion factor MeV to Joules

fm4 C n 1 -4

The power is usually normalized later. In your case, as Boulaich said, you probably use the ratio to the average, then the normalization goes away.

I hope that helps.

yes, did help a lot. In my case the fuel is U3Si2-Al (4.8gU/cm3), so N is the U-235 atomic density (or MCNP somehow knows which isotope is fissile), The Material density is the density of the entire U3Si2-Al, right? Thank you, it was what I was looking for.

Note that what I call number density is N*density what is atoms/cm-barn.

N is Avogadro number divided by the atomic mass of the material

yes, got it. the atoms/cm3 = density*N_Avog/Atomic-Mass.

both density and the atomic mass are from the entire material, U3Si2_Al, right?

M = M_U+M_Si+M_Al

and

Density = density of U3Si2-Al.

I guess I can also do:

Fm4 -M M 1 -4, where -M makes the code to multiply by the atomic density of the material M (that is suppose to be N*density) .

WOW! References recommended: Avogadro number, number of atoms in a chemical compound and in an alloy, MCNP material card and how it relates to atomic density in the cell card, etc.

Finally, do you work with the Multiproposito reactor?

No, I don't, but have colleagues working with. I've been abroad for a while working with another kind of project. In short, I had only a brief introduction to MCNP during the classes (which was nothing) and then had to keep going by myself. Now I'm learning a bit more about the tallies. I've performed the flux profile calculations and it was ok. The problem came out when I was trying to get the power density to go forward into the TH analysis (I knew only the F4 card at the moment), that's why I've decided asking here to get an already done FM card to do that and avoid mistakes.

Adding an Asterisk befor your tally (*Fn) changes the unit into an energy tally. for example *F4 ---> [MeV /cm²] rather than [particles / cm²]. these can enhance your normalized values.

In MCNP manual in the test problems look for the Kcode test case and use the formula there to find out numer of fission required to get 1 watt and then use it as ur multiplier

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