?Learning MCNP for Research Reactor Analysis

In summary: The problem is that you are not understanding the meaning of R0. R0 is the reference point for the flux calculation. Without knowing the R0, it is difficult to calculate the flux.
  • #1
ISAAC BAIDOO
5
0
I just started learning MCNP for research reactor analysis. Can anybody advise where and how I could easily get material to read and practice
 
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  • #3
Ibrahim Hany said:
For an easy and quick start: Start learning from the Primer.

For a real study with the details, head to the 3 volumes of the Manual. (Start with Vol 2)

Ibrahim Hany said:
For an easy and quick start: Start learning from the Primer.

For a real study with the details, head to the 3 volumes of the Manual. (Start with Vol 2)

Check this: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/mcnp5.shtml

Check this: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/mcnp5.sht
thanks a lot
I have just started and it getting interesting but I tried to run some basic example but i could not reference/direct the MCNP to the inp. file.
for example, how do I start when the mcnp command prompt shows: c:\MCNP>_
 
  • #4

You can install the VisEd, it is a good visualizing tool and also you may run on it.

Or you can start a batch file ( in the folder of the input file ) with the command:

MCNP5 (or whatever the version you are using) I=name of the input file with extension O=name of the output file with extension
 
  • #5
Dear Sir,
Thank you so much for your eagerness to help.
kindly explain to me the following:
  1. Can you tell me what LCS: error in MCNPX
I got the LCS error prompt from running the input file below:
example 4-3, repeated structure, two cylinders
C cell cards
1 1 9.90605e-2 -1 -4 5 u=3 imp:n=1 $ solution
2 0 -1 4 u=3 imp:n=1 $ void region
3 2 -2.7 1:-5 u=3 imp:n=1 $Al container
4 0 -2 -3 6 fill=3 imp:n=1
5 like 4 but trcl (17 0 0) imp:n=1
6 3 -1.0 10 -11 8 -9 7 -3 #4 #5 imp:n=1
7 0 -10:11:-8:9:-7:3 imp:n=0
C surface cards
1 cz 6.35 $ solution radius
2 cz 6.50 $
3 pz 80.0 $ top of container
4 pz 70.2 $ top of solution
5 pz 0.0
6 pz -0.15
7 pz -20.15 $ bottom of tank
c sides of tank
8 px -16.5
9 px 43.5
10 py -26.5
11 py 26.5
c data cards
c materials cards
m1 1001.62c 6.2210e-2 8016.62c 3.3621e-2
9019.62c 2.5161e-2 92235.66c 1.1760e-2
92238.66c 8.2051e-5
mt1 lwtr.60t
m2 13027.62c 1.0
m3 1001.62c 2 8016.62c 1
mt3 lwtr.60t
c control cards
kcode 5000 1.0 50 250
ksrc 0 0 35 17 0 35

I was trying this example from the mcnp primer, but whiles running, it stopped at the 192nd cycle and indicated “LCS error”
  1. I would want to know how I can execute basic plot command.
The primer indicates I could execute plot command by “mcnp inp=filename ip” or mcnp ip inp=filename. I expect to see the “The default plot is a PX slice centered at (0, 0, 0) with an extent of -100 cm to 100 cm on the Y-axis and -100 cm to 100 cm on the Z-axis” as indicated in the primer. Instead, the is the error comment I get, “bad trouble in mcnpx in routine unique cannpt create outp
In fact, the main problem is how to use the command to display the plot for now before I begin to think about how to make in unique.
Best Regards
Isaac Kwasi Baidoo
 
  • #6
can i ask you sir??
 
  • #7
Dear Sir,

Sorry sir, i found problem in solving detector flux using tally F5. For tally F5 there is R0 for point and ring detector. I still confuse what’s meaning of R0 ? can you help me to give me some example for make me easy to undestand it?thank you sir,

I hope you help me to solving the problem
 

Related to ?Learning MCNP for Research Reactor Analysis

1. What is MCNP and why is it used for research reactor analysis?

MCNP (Monte Carlo N-Particle) is a computer code used for simulating and analyzing various nuclear systems. It uses statistical methods to simulate the movement and interactions of particles within a system, making it useful for studying the behavior of nuclear reactors. It is commonly used for research reactor analysis due to its accuracy and ability to handle complex geometries and materials.

2. How can I learn MCNP for research reactor analysis?

There are several ways to learn MCNP, including attending workshops, taking online courses, and self-study using resources such as user manuals and tutorial materials provided by the developers. It is also helpful to have a background in nuclear engineering or physics.

3. Is MCNP difficult to learn?

Learning MCNP can be challenging, especially for those without a strong background in nuclear engineering or physics. However, with dedication and practice, it is possible to become proficient in using the code for research reactor analysis.

4. Are there any limitations to using MCNP for research reactor analysis?

While MCNP is a powerful and widely used code, it does have some limitations. It may not be suitable for very large or complex systems, and it requires a significant amount of computational resources and time to run simulations. Additionally, it relies on accurate input data and assumptions, so it is important for users to have a good understanding of the physics and phenomena being modeled.

5. Can MCNP be used for other types of nuclear systems besides research reactors?

Yes, MCNP can be used for a variety of nuclear systems, including power reactors, medical applications, and nuclear safeguards. It is a versatile code that can be adapted for different types of simulations with appropriate input data and parameters.

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