Recent content by mark_bose

First of all thank you for the help. All the resources you linked seem very interesting. I was looking for the resources mentioned by @Astronuc and i found this interactive database from IAEA called THERMPRO (maybe it is linked to the THERSYST document?)...

Hello, I am looking for a comprehensive resource (paper, book or any kind of document) that contains the most relevant properties of most nuclear fuels. To be more specific, the information i am looking for are thermophysical properties, chemical compatibility with other materials and...

I'm using Serpent and the assumption of @rpp about the procedure is correct. For what concern temperature profiles I'm using an exact proportionality with power in each ring of the core (labeled as "B" "C" "D" ...). Initial and final T. profiles are taken starting from power distribution and...

Hi all, I'm trying to derive fuel temperature coefficient in a TRIGA reactor using a monte carlo code. When i do that, if i assume a radial temperature profile along the core, i obtain smaller value (-7pcm/K) than the one achieved with uniform temperature (-9pcm/K). More in detail: in my case...

Thank you all for the suggestions. Probably i will start with the simplifications you suggest. It seems that "Theory and Methodology" chapter in OpenMC manual could be useful.

Thank you all for your replies, Thank you, i heard about MCNP but i don't have a licence. However I have some experience with a similar code named Serpent. But my goal is to write my own. I tried to do so but unfortunately i wasn't able to find anything useful about this specific topic. Yes...

Hi, i would like to write my own MC code in order to simulate the transport of Neutrons in Nuclear reactors. I know the basics of MC and i have already written a code for homogeneus reactors, my problem is the generalization to more complex geometries made of different materials, such as fuel...

Hi all, I'm a Nuclear engineering student from Italy.