MCNP - Tallies definition with "<"

In summary, the conversation is about the use of the "<" symbol in the MCNP manual for defining tallies. The person asking the question is confused about its meaning and is looking for clarification. The expert explains that "<" is used as a logical operator and is commonly used to specify elements in repeated structures built with universe/fill. They also mention that the manual provides a not-so-great explanation in section
  • #1
TL;DR Summary
Examples of tallies definition with logical operator "<" but I don't know what it means.
Hi everyone,

In MCNP manual there are often examples of Listing containing examples of tallies which have, in the definition of the cells/surfaces of the tally itself, the "<" symbol. I could not find in the document any reference to the use of logical expression in the definition of tallies (assuming "<" is actually used as a logical operator).

Could you please tell me what it means?

Referring to the last version of the manual (MCNP® Code Version 6.3.0 Theory & User Manual) the first example of use in a Listing is for Listing 5.13 but it appears in several other listing (e.g. Listing 5.51 or Listing 6.2).

Thanks a lot in advance.
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  • #2
I actually stared at the example for about 10 mins completely clueless, before realising I'd used these myself. That is a such an unhelpful fragment. < means within and [x y z], which you'd commonly see with it, means part of a lattice. They are for specifying elements in repeated structures built with universe/fill.

See " Repeated Structures Tallies" for not a great explanation.
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  • #3
Thanks a lot!
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1. What is MCNP?

MCNP (Monte Carlo N-Particle) is a computer code used for simulating the transport of particles through matter. It is commonly used in nuclear engineering and radiation physics applications.

2. What is the purpose of tallies in MCNP?

Tallies in MCNP are used to calculate various quantities of interest, such as neutron flux, energy deposition, and dose rates, at specific locations in a simulated system. They provide valuable information about the behavior of particles in a given system.

3. How are tallies defined in MCNP?

Tallies in MCNP are defined by specifying the type of tally (e.g. flux tally, surface tally), the particle type, the scoring region, and the scoring options (e.g. energy bins, angular bins). These parameters are input into the MCNP input file.

4. What is the significance of the "<" symbol in tallies definition in MCNP?

The "<" symbol in tallies definition in MCNP is used to specify the scoring region or surface for the tally. This allows for the calculation of quantities at specific locations in the system, such as at a particular surface or within a specific volume.

5. Are there any limitations to tallies in MCNP?

Yes, there are some limitations to tallies in MCNP. For example, tallies cannot be used to score quantities outside of the simulation geometry or to score particles that do not interact with the material in the system. Additionally, the accuracy of tallies may be affected by statistical uncertainties and the chosen scoring options.