eVinci MCNP input file

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An urgent request for an MCNP input file for the Westinghouse eVinci microreactor was made, highlighting the need for proprietary information. It was suggested to contact Westinghouse directly, although it was noted that such details are unlikely to be publicly available. Participants discussed that Westinghouse may not use MCNP for their designs, as vendors often rely on proprietary codes. The NRC might utilize MCNP for verification of vendor analyses. Access to the input file is limited, and proprietary constraints are a significant barrier.
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TL;DR Summary
Dears,
Kindly, I need an MCNP input file for the Westinghouse eVinci microreactor, urgently
I appreciate all your efforts
Dears,
Kindly, I need an MCNP input file for the Westinghouse eVinci microreactor, urgently
I appreciate all your efforts
 
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Welcome to PF.

Can you just contact Westinghouse to request the file?
 
igomaa51 said:
TL;DR Summary: Dears,
Kindly, I need an MCNP input file for the Westinghouse eVinci microreactor, urgently
I appreciate all your efforts

Kindly, I need an MCNP input file for the Westinghouse eVinci microreactor, urgently
I appreciate all your efforts
Such details would be proprietary to Westinghouse, and they are unlikely to place such information in the public domain.

A non-proprietary summary presenation to the NRC. Note: the lack of detail.
https://www.nrc.gov/docs/ML2305/ML23053A351.pdf

Here is an image that one might be able to use.
https://spectrum.ieee.org/microreactor
 
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berkeman said:
Can you just contact Westinghouse to request the file?
Pretty sure @berkeman is being facetious. Westinghouse won't *give* you anything. If you buy an eVinci they might let you look at their code decks. In their offices.

I'm not sure that Westinghouse even uses MCNP. Maybe they do, but my experience (in other areas of reactor design) the reactor vendors use their own in-house codes. The NRC may use MCNP to back-check WEC analysis results.
 
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Likes Astronuc and berkeman
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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