Temperature in MCNP - Using Library ENDF7 for Research Reactor

In summary: TMP card is used for the energy of thermal neutrons so you have to accommodate the material temperatures.
  • #1
chivasorn
22
0
Hi there,
I want to know that how can involve the temperature in mcnp code.
for example; the library endf7 for mcnp has five certain temperature:300 kelvin, 600 kelvin, 900 kelvin, 1200 kelvin & 1500 kelvin. if I want to calculate flux distribution in 330 kelvin for a research reactor, how can I?

best regards
 
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  • #2
Is there no temperature dependence functions.

One could interpolate, or given than 330 K is ~ 300 K, simply use 300 K. The Doppler broadening shouldn't be too significant going from 300 K to 330 K. The density changes are not very significant either.
 
  • #3
Hi Astronuc.
I have another question; TMP card in MCNP that can be entered at cell card have not any effect in result of calculations?
Please, more explain about that, I need.

have best time.
 
  • #4
TMP card is used for the energy of thermal neutrons so you have have to accommodate the material temperatures
 
  • #5
Hi..

for temperature, there is a nuclear code to produce the library for it...
an NJOY, but it some kind of rare code...

by MODUL ACER, you can create the applicable data for any temperature...and it is also applicable for WIMS...

but, I do agree with Mr. Astronuc...
 
  • #6
If the value of TMP parameter (or its default value if not given explicitly) in a cell card differs from temperature of nuclide in cross-section file, then MCNP modifies the nuclide cross-section using free-gas treatment (you can see a warning message during execution and in output file). It resets temperature to the room temp. and then (if a TMP value given) to the requested temperature. I haven't dug into this problem, but from what I heard, the free-gas treatment is not very precise procedure. The treatment happens even if you have difference in the last digit, i.e. negligible. This can potentially produce unnecessary discrepancy in your results. Please, correct me if it is wrong.

chivasorn said:
Hi Astronuc.
I have another question; TMP card in MCNP that can be entered at cell card have not any effect in result of calculations?
Please, more explain about that, I need.

have best time.
 
Last edited by a moderator:

1. What is MCNP?

MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo simulation code used for simulating the transport of particles through matter.

2. What is the importance of temperature in MCNP?

Temperature is an important parameter in MCNP simulations, as it affects the material properties used in the simulations, such as cross sections and scattering probabilities.

3. What is ENDF7 library?

The ENDF7 (Evaluated Nuclear Data File) library is a database of evaluated neutron cross sections used in nuclear engineering and research.

4. How is the ENDF7 library used in MCNP?

The ENDF7 library can be used in MCNP to provide accurate and up-to-date nuclear data for materials used in simulations, including temperature-dependent cross sections.

5. How does temperature affect the results of MCNP simulations?

The temperature used in MCNP simulations can significantly impact the results, as it affects the material properties and neutron behavior. It is important to carefully consider and accurately input the temperature in order to obtain reliable results.

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