Monte Carlo Simulation of Nuclear Physics

[Rob Tayloe]

I realized that I have omitted the code package with which I have the greatest familiarity. That is the SCALE code package. I began using a very early implementation of these codes in 1980 when punch cards had to be submitted via a secure line to Oak Ridge and printout would come in a day or so. I thought that the dial-up modem (300 baud) was a great step forward and I used my Radio Shack TRS-80 model 1 computer to generate input decks to submit. The output was still printed the next day at work, but this still seemed like a great advance. I knew many of the early code developers and it was a treat to be able to interact with them periodically. I did take the formal SCALE training in the mid-90s; by that time one could use a personal computer to run most cases.

As with the LANL generated MCNP code, SCALE's use is regulated by the US Dept of Energy. The code package is free for most US university nuclear engineering programs and select government contractors. Others must apply to obtain and use the code package and pay a fee for distribution. The formal training courses are not free.

https://www.ornl.gov/onramp/scale-code-system
https://rsicc.ornl.gov/Default.aspx

The SCALE code system is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Nuclear Energy and Fuel Cycle Division (NEFCD) of the Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic solvers and three Monte Carlo radiation transport solvers selected based on the user’s desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s GUIs assist with accurate system modeling and convenient access to desired results.

[Rob Tayloe]

Earlier mention was made of the open source monte carlo radiation transport code OpenMC. I am repeating the main link below. Although most examples are given for eigenvalue (k-effective) calculations for nuclear reactors, there is an example given for calculating the gamma spectrum for a Cs-137 source with a NaI detector. The link to the discussion follows.

Main OpenMC link -
https://docs.openmc.org/en/stable/

Discussion of use of OpenMC to calculate the pulse height tally. An artificial 800 keV point source is used along with the 661 keV Cs-137 point source. [There seems to be a minor error in the write-up where the 800 keV source is called a 200 keV source].
https://github.com/openmc-dev/openmc-no ... ctor.ipynb

The calculated pulse height results are given, followed by some post processing to broaden the peaks to give results consistent with the detectors' resolution

[Rob Tayloe]

There are a great many technical papers published regarding use of monte-carlo, and in particular the MCNP code, to determine the response function or detection efficiency of radiation detectors. Many of these papers require access to a technical library which pays the considerable subscription fees. There are, however, some technical papers that do allow free public access. I have listed a few below that can be freely accessed (as of 21 Nov 2024); these pertain to use of MCNP for scintillation detector response or efficiency (usually NaI detectors are modeled).

Monte Carlo modelling of a NaI(Tl) scintillator detectors using MCNP simulation code 2017
https://www.researchgate.net/publicatio ... ation_code

Calculation of Intrinsic Efficiency of NaI(Tl) Detector Using MCNP Code 2005
https://ripublication.com/ijpap/1002.pdf

Estimation of Detection Efficiency for NaI detector using MCNP 2016
https://www.iosrjournals.org/iosr-jap/p ... 130133.pdf

Evaluation of the Nonlinear Response Function and Efficiency of a Scintillation Detector Using Monte Carlo and Analytical Methods 2014
http://ajesjournal.com/PDFs/2014-2/5_ev ... ciency.pdf

The Effect of Detector Dimensions on the NaI (Tl) Detector Response Function 2009
https://scialert.net/fulltext/?doi=jas.2009.2168.2173

Mathematical modeling utilizing the MCNP code and determination of response curves of a NaI(Tl) detector 2021
https://www.mittetecnologia.com.br/anai ... 0246-1.pdf

Precise Monte Carlo simulation of gamma-ray response functions for an NaI(Tl) detector 2002
https://www.researchgate.net/profile/Hu ... oad/3x.pdf

[Rob Tayloe]

The on-line calculator that was associated with the PNNL Compendium of Materials Compounds is no longer available. I did find an on-line (or downloadable) calculator (also from PNNL) that can be useful. Links are provided for the on-line version (as of 22 Nov 2024) and downloadable versions (and source code on a github link) -

https://pnnl-comp-mass-spec.github.io/M ... lator-VB6/

https://github.com/PNNL-Comp-Mass-Spec/ ... /tag/v6.50

https://github.com/PNNL-Comp-Mass-Spec/ ... ulator-VB6

The WISE Uranium Project website has several on-line calculators that may be of some interest. Previously, I have used the activity calculator when small samples were made radioactive with the OSU nuclear reactor. I found this activity calculator to be quite accurate for thermal neutron activation. I was able to create some interesting sources for students' gamma spec lab.
https://www.wise-uranium.org/calc.html

[Lposter]

Just to complete the listings...

Red Cullens TART2002, the competitor to MCNP once upon a time. As usual, hard to get. But interesting.

http://redcullen1.net/homepage.new/mc.htm

The CTBTO had a dedicated efficiency tool called VGSL which ran off a heavily modified MCNP4 but...unlike many others....included True Coincidence in a very thorough way. I dont know where you get it these days but it was very good.

EFFTRAN is free and very good for efficiency transfer. Find with Google.⅚

As is ETNA. http://www.lnhb.fr/home/rd-activities/s ... -software/

If you can get MCNP, then SuperSynth turns it into a very easy use dedicated gamma spectromtry work shop and really takes the donkey work out of it. But the databases are janky and need careful checking. Costs a few dollars but Id rather pay that than ruin my eyes editing .inp files or using VisEd.

http://www.mcnpvised.com/synth/synth.html