N11 / PHYS820 Radiation Shielding 2023 Practical 2
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N11 / PHYS820
Radiation Shielding 2023
Practical 2
PRACTICAL 2:
MCNP simulation of a detector rig
The objective of this practical is to perform a validation of the MCNP code against the experimental measurements you have analysed in practical 1. You will require the geometry of the two setups to complete this practical.
Part A [20 marks]
(1) Read through the short guide to MCNP, the accompanying video that is provided and familiarise yourself with the example program “he3demo” . I have included the code from this program on the next page for reference. This program defines the geometry illustrated in figure 1 below.
Particle terminated Detector
Particle tallied
Source
Moderator
Figure 1: The geometry defined in he3demo, the blue circle is the source location.
(2) Now type in the program code “he3demo” into an appropriate text editor. You may choose to use “Notepad” on Windows 10. Save the program as “he3demo” in the documents folder in your account. When saving the file from Notepad, put quotation marks around the file name to avoid the program adding a .txt extension. Remember you are limited to 7 character filenames! I would suggest you create a “mcnp” folder to put these files in to keep things neat.
(3) Once you have saved the program leave the Notepad window open, you may need it to debug the code!
MCNP example code for part A
he3demo - demonstration he-3 detector
C
C material densities
C 1 = -0.96 2 = 3.940e-4
C
C ######################## Define Cells ############################# C
1 2 4.925e-5 -1 -2 3 imp:n 1 $Detector
100 1 -0.96 -100 -2 3 #(-1 -2 3) imp:n 1 $Moderator
200 0 200 imp:n 0 $Define world
201 0 -200 #(-100 -2 3) imp:n 1 $Define world
C
C ######################## Define Surfaces ########################## C
1 cy 2.5 |
$Detector cylinder |
2 py 50 |
$limits (right) |
3 py -50 |
$limits (left) |
100 cy 10.5 |
$Moderator |
200 sx -20 70 |
$Define world |
C
C ######################## Define Source #############################
C Source AM/Be energy definition. Point source. Isotropic.
sdef pos=-11 0 0 erg=d1
si1 h 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 7.5 8 8.5
9 9.5 10 10.5 11 11.5 12
sp1 d 0 0.15 0.08 0.09 0.06 0.100 0.15 0.107 0.101 0.11 0.102 0.15
0.07 0.12 0.05 0.08 0.10 0.05 0.02 0.03 0.04 0.02 0 0 0
C
C ######################## Define Tallies ############################ C
f14:n 1 $Tally neutrons in cell 1
e14 0.5e-6 1e-6 0.5e-5 1e-5 0.5e-4 1e-4 0.5e-3 1e-3
0.5e-2 1e-2 0.5e-1 1e-1 0.5 1 5 10 50 100
C
C
f24:n 1 $Tally neutrons in cell 1
e24 0.5e-6 1e-6 0.5e-5 1e-5 0.5e-4 1e-4 0.5e-3 1e-3
0.5e-2 1e-2 0.5e-1 1e-1 0.5 1 5 10 50 100
fm24 -1 2 103 $simulate (n,p) interaction
C
C ######################## Define Materials ########################## C
m1 6000.60C 1 1001.60C 2 $Polythene
m2 2003.60C 1 $He-3
mt1 poly.01t $thermal treatment card
C
C ######################## Define Run Time ########################### C
ctme 1
(4) Now open a Command Prompt in Windows and navigate to the folder you saved your “he3demo” file in. You may need to use the “cd” change directory command to reach the location [Example: if you saved you file in documents/mcnp, you will type “cd documents/mcnp <rtn>”]. You can check you are in the correct location by typing “dir” and it should list the contents of the folder.
(5) Now type “mcnp ip n=he3demo <rtn>”, this will open the MCNP program and the geometry plotter. Use the “px”, “py”, “pz”, “scale” and “extent” commands to look at the 3D world you have created. [Take a screenshot of this and include it in your writeup].
(6) You can debug if needed at this point, if not you should proceed to run the Monte Carlo. You first need to delete your output files, these are the he3demoo and he3democ files in this case. Do this either from the cmd prompt or from the Windows Explorer.
(7) Now type “mcnp ixr n=he3demo <rtn>”, this will run the Monte Carlo code. It will take as long as you have specified in the input, my example has 1 minute. Once complete a dialogue box will appear to indicate this.
(8) You can now open your output file in notepad and look through it. Use my video to help you understand what you see. Now extract the two relevant tally outputs into Excel.
(9) Convert the tally outputs from MCNP probability units into neutrons per second. Remember to do this you need to consider the units of the F14 and F24 tallies which are n/cm2/n and n/cm3/n, respectively. You should assume the neutron yield from the source is 2.5x106n/s. You should now:
a. Tabulate your data with appropriate errors, don’t forget to include only the statistically relevant number of decimal places. [5]
b. Plot the F14 tally, the number of incident neutrons vs energy. [5]
c. Do the same for the F24, number of detected neutrons vs energy. [5]
d. Finally, plot the intrinsic efficiency of the detector vs energy. [5]
Note, you may want to consider using a log scale to show the results properly. Be sure to include the units on both axes and error bars where appropriate. Provide a brief narrative describing your findings.
Part B [55 marks]
You will now modify the geometry of the he3demo program to match the detector setup used in practical 1. The initial geometry you should use is illustrated in figure 2. You will need to add:
(1) The new geometry. Note the cadmium sheet is 1m x 1m x 1mm, the polythene moderator sheets are 1m x 1m x 2cm and the borated polyethene sheets are 1m x 1m x 1cm.
(2) The new material definitions for water, boron, cadmium and any other relevant material such as concrete if needed. Remember to justify your cross-section choice
(3) Don’t forget you need thermal treatment cards for any moderating material.
(4) Be sure to add both neutron and photon tallies.
Figure 2: A simplification of the neutron rig geometry used in experiment 1
You should now:
(1) Complete any necessary changes to your MCNP code and debug it! You should include a copy of your code, an example geometry plot and a justification of the included materials, densities and the selected code run time per moderator thickness. [10]
(2) Simulate the response of the He-3 detector as a function of moderator thickness for the four scenarios in practical 1. In each case calculate both the number of neutrons incident and those detected. Tabulate your results. Don’t forget errors. [10]
(3) Now produce the plots vs moderator thickness separated into thermal, high energy and total energy bins as a function of moderator thickness. You should overlay your experimental results on the relevant plots. Don’t forget errors. [20]
(4) Plot an example photon energy spectrum for the rig, comment on what it shows. [10]
(5) Comment on the level of agreement or not for your neutron results, did you achieve validation? [5]
2023-02-07