Could a nuclear battery (RTG) be simulated by MCNP5?

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Discussion Overview

The discussion centers around the feasibility of simulating a nuclear battery, specifically a radioisotope thermoelectric generator (RTG) using plutonium-238, with the MCNP5 software. Participants explore the capabilities of MCNP5 in modeling energy deposition and the challenges associated with radioactive decay and thermal calculations.

Discussion Character

  • Exploratory, Technical explanation, Debate/contested

Main Points Raised

  • One participant inquires about the possibility of simulating an RTG with MCNP5 and asks about specific features related to thermoelectric conversions.
  • Another participant suggests that it should be possible to calculate energy deposition using the F6 tally in MCNP5.
  • A later reply questions the appropriateness of MCNP5 for modeling radioactive decay, suggesting that while energy deposition can be modeled, thermal source calculations may require additional methods.
  • It is proposed that for temperature distribution calculations, a finite-element code might be necessary to handle heat transfer, using the thermal source derived from the decay of Pu-238.

Areas of Agreement / Disagreement

Participants express differing views on the capabilities of MCNP5 for this application, with some believing it can handle certain aspects while others argue it may not be suitable for modeling radioactive decay or thermal distributions directly.

Contextual Notes

Participants note limitations in MCNP5's ability to model radioactive decay and suggest that additional calculations or software may be required for accurate thermal modeling.

Al-mutawakel
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I am a graduate nuclear engineer(no Master yet), and I ask if nuclear battery(radioisotope thermoelectric generator, that use plutonium-238 as heat source) can be simulate by MCNP5. And is their any card in MCNP5 treat or deal with thermoelectric converions. Thanks
 
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Hi
I think it must be possible, because you can calculate the energy deposition (F6 tally)
 
thanks for replied.
 
What are you trying to model exactly? MCNP is used to model the transport of neutrons and gamma's, but it isn't the right tool to model radioactive decay.

Are you trying to calculate the thermal source? If so, it is a fairly easy hand calculation, you just need to know the the mass of Pu-238, the half-life, and the amount of energy released per decay. You might also have to include the decay of any daughter products from the decay of Pu-238.

If you want to actually calculate a temperature distribution, you would probably need to use a finite-element code that performed heat transfer. You would enter the heat source (calculated by the decay of Pu-238), and then the code could calculate the temperatures of the surrounding medium, including the thermal couples.
 

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