Monte Carlo & Radiation
#1
01-11-2007, 01:28 PM
Monte Carlo & Radiation
I was just googling around for information about something known as 'monte carlo codes', and found this forum. I guess i'm in the right area (off topic) because the monte carlo I know of involves radiation detection and dosimetry.
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
#2
01-11-2007, 04:04 PM
Monte Of The Month -- February 2008
Join Date: Aug 2006
Location: Born in East LA
Posts: 4,286
RE: Monte Carlo & Radiation
ORIGINAL: Dosimetry
I was just googling around for information about something known as 'monte carlo codes', and found this forum. I guess i'm in the right area (off topic) because the monte carlo I know of involves radiation detection and dosimetry.
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
I was just googling around for information about something known as 'monte carlo codes', and found this forum. I guess i'm in the right area (off topic) because the monte carlo I know of involves radiation detection and dosimetry.
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
#4
01-11-2007, 07:06 PM
RE: Monte Carlo & Radiation
ORIGINAL: Dosimetry
I was just googling around for information about something known as 'monte carlo codes', and found this forum. I guess i'm in the right area (off topic) because the monte carlo I know of involves radiation detection and dosimetry.
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
I was just googling around for information about something known as 'monte carlo codes', and found this forum. I guess i'm in the right area (off topic) because the monte carlo I know of involves radiation detection and dosimetry.
I just went on one of these course for MCNP training. It's basically a piece of software thats used to simulate various situations involving radiation and how it interacts with materials. MCNP stands for Monte Carlo N Particle.
Pretty different to the monte carlo stuff here or what!! Hope you find it as interesting as I do
Exactly what is dosimetry and why is is the N Particle called Monte Carlo? Is the Principality of Monte Carlo where it comes from?
#5
01-11-2007, 08:42 PM
Monte Of The Month -- February 2008
Join Date: Aug 2006
Location: Born in East LA
Posts: 4,286
RE: Monte Carlo & Radiation
I found this on the Los Alamos website:
[blockquote]MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, Detector Design and analysis, nuclear oil well logging, Accelerator target design, Fission and fusion reactor design, decontamination and decommissioning. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori.
Pointwise cross-section data typically are used, although group-wise data also are available. For neutrons, all reactions given in a particular cross- section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorbtion in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields.
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data.
MCNP5 contains numerous flexible tallies: surface current & flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height taly for energy or charge deposition, mesh tallies, and radiography tallies.
[blockquote][/blockquote][/blockquote]
[blockquote]MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, Detector Design and analysis, nuclear oil well logging, Accelerator target design, Fission and fusion reactor design, decontamination and decommissioning. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori.
Pointwise cross-section data typically are used, although group-wise data also are available. For neutrons, all reactions given in a particular cross- section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(alpha,beta) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorbtion in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields.
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data.
MCNP5 contains numerous flexible tallies: surface current & flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height taly for energy or charge deposition, mesh tallies, and radiography tallies.
[blockquote][/blockquote][/blockquote]
#7
01-14-2007, 03:43 PM
RE: Monte Carlo & Radiation
ORIGINAL: 04 Intimidator
Umm... just keep it away from my car.
Umm... just keep it away from my car.
[sm=funnypostabove.gif] Yeah...I still don't understand what it is. But then again, Chemistry was my worst Science subject.
Thread
Thread Starter
Forum
Replies
Last Post
JD1979
New Member Area
2
11-03-2020 07:58 AM
uortiz714
Off Topic
3
05-04-2012 08:12 AM
Dovas Monte
New Member Area
14
02-01-2011 01:24 PM
team monte carlo ss
Off Topic
3
08-11-2008 06:02 PM