Simulation of Pu-238 Decay

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The discussion focuses on simulating the decay of Pu-238 in RTGs using MCNP6.1.1b, specifically addressing issues with the F6 tally output. The user reports a discrepancy between the calculated decay heat (22 watts) and the expected output (62.5 watts) from a Pu-238 pellet. A suggestion is made to remove the ACT option and adjust the PAR setting to potentially resolve the issue. Another user notes that the PAR=sa setting is new and may not function as expected, indicating that using PAR=a with specific energy settings yields correct results. The conversation emphasizes troubleshooting the simulation parameters to align the output with known decay heat values.
omarkhairallah
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Hi everyone, In my graduation project i am trying to simulate the decay of Pu-238 (in RTGs) and the dose from it on mcnp6.1.1b. I used the F6 tally and (PAR=sa) in the SDEF card to simulate the decay of alphas and also set the mode to ( MODE a e p ) to put the gammas and betas in consideration ( if any interactions could produce one of them and contribute to the total dose).

But i am facing some complexities, when using the F6 tally, the output is 0.21667 Mev per gram, which is when multiplied by the grams of the cell (145.6) and converting to watts, it's only 22 watts while the pellet of Pu-238 should produce 62.5 watts (In one RTG pellet) of decay heat, i want someone to help me with this because i dont know what's wrong.

You can find my input file attached.
 

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Hi @omarkhairallah,

Welcome to PhysicsForums.

I don't know what par=sa actually does. It's new in 6.1.1 and it looks up alpha decay information from decay tables and CINDER. I don't understand ACT either. When I use par=a erg=5.5 and remove ACT to make it run on my version the tally when multiplied by the mass is 5.4999 MeV. Which is trivially what it should be. Try removing ACT and if that doesn't fix it then the spontaneous alpha option isn't doing what you think it is.
 
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Thank you @Alex A .
 
Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information(power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.) for Material 2 is zero, while Material 1...

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