How to Determine Fission Rate in MCNP6?

[J Chancleto]

TL;DR Summary

Fission rate

Hi!

First of all, thank you for your time.

I am simulating a nuclear engine for space applications. I want to know the fission rate of the engine but i dont know how. I am using xming to plot the fmesh 4 and the tally is:

fmesh4:n geom=xyz origin= -50. -50. -50.
imesh= 50. iints=100
jmesh= 50. jints=100
kmesh= 50. kints=100
OUT=NONE
fm4 -1. 0 -6

I open MCNP6 console, enter mcnp6 i=name.txt, and i can obtain keff. Then, i use r=runtpe z, then fmesh 4 and i can see the plot. In the output file i dont find the fission rate value. How can i obtain it? Thank you so much

 

[Alex A]

Welcome to physcisforums!

At criticality the reactor fission rate is whatever you want it to be. Each fission produces around 200MeV, alternately each gram of U-235 produces about 1 MW.Day of energy. So I guess the question is really, 'how much power do you need?'.

All the tally results I know in MCNP are per source particle, so they need to be multiplied by the total neutron activity to produce a real answer. Turn the energy you need into the fission rate. Then for a reactor that is on but not exploding keff=1.0, so that is the same as the neutron source activity.

You may have better ways of working this out, and a BURN run might tell you more details but I don't actually know if this will calculate energy produced by all the side reactions and tell you a result, or if it all feeds correctly into the deposited energy tallies (You will still need to choose a power output at the start). You will want to run a BURN anyway I suspect because it will tell you how long the engine can run for given use cases. When the reactor reaches keff < 1.0 it stops. BURN will not stop automatically, keeping the reactor critical is an assumption of the method and someone else's problem.

This sounds like fun, good luck!

 

[rpp]

What exactly are you trying to do? Do you want to produce a pretty picture? Calculate the total fission rate for the reactor? Or calculate the fission rates in particular parts of the core?

While fmesh is very nice for making a plot, I've found that it isn't very useful for extracting real data unless you are lucky enough where the mesh lines up with your geometry exactly (for example a lattice). I don't think this is the case for you.

If you want the total fission rate in the reactor, this is going to be normalized to the core power. The core power will give you the total fission rate, but you'll need cross sections and energy per fission edits.

If you want the fission distribution on a non-uniform mesh, you will need to set up multiple other tallies, most likely "f4" tallies to tally the fission rates in different cells.

 

[J Chancleto]

What exactly are you trying to do? Do you want to produce a pretty picture? Calculate the total fission rate for the reactor? Or calculate the fission rates in particular parts of the core?

While fmesh is very nice for making a plot, I've found that it isn't very useful for extracting real data unless you are lucky enough where the mesh lines up with your geometry exactly (for example a lattice). I don't think this is the case for you.

If you want the total fission rate in the reactor, this is going to be normalized to the core power. The core power will give you the total fission rate, but you'll need cross sections and energy per fission edits.

If you want the fission distribution on a non-uniform mesh, you will need to set up multiple other tallies, most likely "f4" tallies to tally the fission rates in different cells.

The final result that i want to obtain is the core power. I thought that to obtain that i needed the total fission rate in the reactor. Sorry if my comments are a bit "silly", I am new to MCNP, I have to use it for my thesis and at work they can't give me a hand. In the original message there is a part where I attach the fmesh4 and "f4" that I use. Would it work as it is? And, how could I get the value of the total fission rate or core power? Thank you very much for everything.

 

[Alex A]

If you simulate a car, not in mcnp but generally, that could tell you everything going on inside the car. It's not a meaningful question to ask the program how fast the car is going. That is set by the driver of the car. That would need to be something you tell the program, not something the program tells you.

You might be able to work it out, say you know the RPM, then from the gearing and making the assumption that the tires are gripping the road that would tell you the speed of the car. It could be 5 miles an hour in reverse or 150 miles an hour forward the car can do either and everything in between. If you want to know how fast it can go your simulation needs to include parameters that will fail and you can test them.

A reactor is similar, it can run at any power output and any neutron flux density. Real world limits have to do with the thing melting/exploding etc. The power output of a reactor is something that it is designed for, but what it runs at is set by the operator.

MCNP only simulates the neutronics, so it doesn't know when a reactor would melt. If your parameter is neutron flux across a surface you can simulate that and then work out the total power. If your parameter is hydrogen being heated from 30K to 3000K then you need a mass flow so you know the total power needed.

So the question is what do you know about the reactor?