Temperature in MCNP - Using Library ENDF7 for Research Reactor

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To involve temperature in MCNP code, users can interpolate between the available temperatures in the ENDF-7 library, with 300 K being a suitable approximation for 330 K due to minimal Doppler broadening effects. The TMP card is crucial as it adjusts thermal neutron energy based on material temperatures, but discrepancies between TMP values and cross-section file temperatures can lead to inaccuracies due to the free-gas treatment applied by MCNP. This treatment can reset temperatures to room temperature, potentially causing discrepancies in results if not managed carefully. For more precise temperature data, tools like NJOY and MODUL ACER can be utilized to create applicable libraries. Understanding these nuances is essential for accurate flux distribution calculations in research reactors.
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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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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.
 
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.
 
TMP card is used for the energy of thermal neutrons so you have have to accommodate the material temperatures
 
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...
 
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.
 
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