MCNP5 vs MCNP6.2 syntax difference

  • Thread starter werling3
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Summary:

input can run on MCNP5 but not on MCNP6

Main Question or Discussion Point

So I am getting a fatal error on a tally that contains a universe definition in it.

f12:n (10<(u=1))
sd12 1.0
e12 0.1 20.0

the error specifically says "fatal error. invalid universe format in f card bin 1 tally 2."
However, this input is able to run on MCNP5 without experiencing a fatal error.
Does anyone know if this is simply a syntax problem & how it could be fixed.

Thanks.
 

Answers and Replies

  • #2
Astronuc
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One may wish to look at the format of MCNPX.

MCNP6 represents the culmination of a multi-year effort to merge the MCNP5® [X-503] and MCNPX® [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams.

https://rsicc.ornl.gov/codes/ccc/ccc8/ccc-810.html
The initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. So, syntax may have changed. I'll ask some colleagues.
 
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DEvens
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Could you possibly post a complete model that shows the problem?
 

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