How can you use MCNP to do time-dependent reactor calculations?

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To calculate fuel burn-up in a nuclear reactor, MCNPX can be utilized as it includes a built-in depletion mode. Alternatively, ORIGIN, part of the SCALE code package, can perform depletion calculations, which can then be coupled with MCNP for further analysis. There are automated codes available that facilitate this coupling process. Understanding these options is crucial for adjusting core composition based on fuel performance over time. Effective reactor design requires integrating these computational tools for accurate assessments.
mdcpablo
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The only thing I know how on the basis of nuclear reactor design is how to run kcode in MCNP and see if my theoretical reactor is critical. How would I be able to calculate how the fuel is burning in my reactor over some period of time, and change my core composition accordingly?
 
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There are two options. You can use MCNPX which has a depletion mode built into it, or you can use ORIGIN (part of the SCALE code package) to do depletion and then couple the results to MCNP (I believe there are codes which do the coupling automatically).
 
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