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Jaakko Leppänen
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Welcome to the Serpent discussion forum!

Post by Jaakko Leppänen » Sat Mar 20, 2010 1:15 am

This forum is intended for students and professionals in the nuclear field, who are using or interested in the Serpent Monte Carlo reactor physics burnup calculation code (see http://montecarlo.vtt.fi). Unfortunately I had to disable guest posting to reduce spamming, so you need to register in order to participate in the discussion.

Notice that this is an open forum, and anything you post here can be viewed by anybody. Make sure you do not disclose any export-controlled, sensitive or confidential material. If you need to make references in the Serpent source code, please do not attach more than a few lines of code in your posts. Distributing complete subroutines violates the conditions of the license agreement.

Mosebetsi

Re: Welcome to the Serpent discussion forum!

Post by Mosebetsi » Wed May 26, 2010 3:36 pm

Dear Jaako

I have just installed Serpent and its is the first time I use it and any other MCNP type of the code for that matter. That being the case, I do not know where to start with the input and running the code.

Do you perhaps have training manuals or presentation preferably with worked-out examples that I can follow to get myself of the ground. I can unfortunately not get help from anyone around here, nobody knows how to use it.

Even if it is not neccessarily from you personaaly Jaako, i will be gratefull for any help.


Kind regards

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Jaakko Leppänen
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Posts: 2179
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Location: Espoo, Finland
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Re: Welcome to the Serpent discussion forum!

Post by Jaakko Leppänen » Wed May 26, 2010 4:55 pm

Mosebetsi wrote:Dear Jaako

I have just installed Serpent and its is the first time I use it and any other MCNP type of the code for that matter. That being the case, I do not know where to start with the input and running the code.

Do you perhaps have training manuals or presentation preferably with worked-out examples that I can follow to get myself of the ground. I can unfortunately not get help from anyone around here, nobody knows how to use it.

Even if it is not neccessarily from you personaaly Jaako, i will be gratefull for any help.


Kind regards
Probably the best way to get started is to look at one of the examples found at the website:

http://virtual.vtt.fi/virtual/montecarlo/examples/

The BWR case, for example, is a 2D model of an asymmetric boiling water reactor fuel assembly in an infinite lattice. The input file can be roughly divided into geometry and material definitions, along with some additional parameters for the calculation. After the title, the geometry definition begins with the fuel pins:


% --- Fuel Pin definitions:

pin 1
fuel1 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 2
fuel2 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 3
fuel3 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 4
fuel4 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 5
fuel5 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 6
fuel6 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

pin 7
fuel7 4.33500E-01
void 4.42000E-01
clad 5.02500E-01
cool

% --- Empty lattice position:

pin 9
cool


Each pin, numbered from 1 to 9 consist of concentric annular regions with different materials. Pins 1-7 contain different fuels (fuel1 to fuel7), surrounded by gas gap, cladding and coolant water. Pin 9 is an empty lattice position, containing only water.

The fuel pins are put into a lattice, defined as:


% --- Lattice (type = 1, pin pitch = 1.295):

lat 10 1 0.0 0.0 12 12 1.295
9 9 9 9 9 9 9 9 9 9 9 9
9 1 2 3 5 5 5 5 5 3 2 9
9 2 3 5 6 6 6 6 7 5 4 9
9 3 5 7 6 7 6 6 6 6 5 9
9 5 6 6 6 6 6 6 7 6 6 9
9 5 6 7 6 9 9 9 6 7 6 9
9 5 6 6 6 9 9 9 6 6 6 9
9 5 6 6 6 9 9 9 6 6 6 9
9 5 7 6 7 6 6 6 7 6 5 9
9 3 5 6 6 7 6 6 6 6 5 9
9 2 4 5 6 6 6 6 5 5 3 9
9 9 9 9 9 9 9 9 9 9 9 9


The numbers in the layout correspond to the pin numbers and the parameters determine the lattice size (12 by 12 pins) and pitch (1.295 cm).

The rest of the geometry is defined using surfaces and cells:


% --- Outer channel (assembly pitch = 15.375):

surf 1 sqc 0.0 0.0 6.70000
surf 2 sqc 0.0 0.0 6.93000
surf 3 sqc -0.233 -0.233 7.68750

% --- Channel inside assembly:

surf 4 sqc 0.6475 0.6475 1.6742
surf 5 sqc 0.6475 0.6475 1.7445

% --- Cell definitions:

cell 1 0 moder -4 % Water inside moderator channel
cell 2 0 box 4 -5 % Moderator channel walls
cell 3 0 fill 10 -1 5 % Pin lattice
cell 4 0 box 1 -2 % Channel box wall
cell 5 0 moder 2 -3 % Water outside channel box
cell 99 0 outside 3 % Outside world


The construction is a bit more complicated, but the idea is that the cells 1-5 and 99 are defined by the intersections of surfaces 1-5. Cell 2, for example, contains material "box", located outside surface 4 (positive sign) and inside surface 5 (negative sign). Both surfaces are square cyliders, infinite in the axial direction. Unlike the other cells, cell 3 doesn't contain a material, but is filled with universe 10, which corresponds to the lattice defined earlier.

The material definitions are:


% --- Fuel materials:

mat fuel1 -10.424
92235.09c -0.015867
92238.09c -0.86563
8016.09c -0.1185

mat fuel2 -10.424
92235.09c -0.018512
92238.09c -0.86299
8016.09c -0.1185

mat fuel3 -10.424
92235.09c -0.022919
92238.09c -0.85858
8016.09c -0.1185

mat fuel4 -10.424
92235.09c -0.026445
92238.09c -0.85505
8016.09c -0.1185

mat fuel5 -10.424
92235.09c -0.029971
92238.09c -0.85153
8016.09c -0.1185

mat fuel6 -10.424
92235.09c -0.032615
92238.09c -0.84888
8016.09c -0.1185

% --- Fuel with Gd:

mat fuel7 -10.291
92235.09c -3.13109E-02
92238.09c -8.14929E-01
64152.09c -6.70544E-05
64154.09c -7.13344E-04
64155.09c -5.06012E-03
64156.09c -7.08860E-03
64157.09c -5.43718E-03
64158.09c -8.64341E-03
64160.09c -7.69426E-03
8016.09c -1.19056E-01

% --- Cladding and channel box wall:

mat clad -6.55
40000.06c -0.98135
24000.06c -0.00100
26000.06c -0.00135
28000.06c -0.00055
50000.06c -0.01450
8016.06c -0.00125

mat box -6.55
40000.06c -0.98135
24000.06c -0.00100
26000.06c -0.00135
28000.06c -0.00055
50000.06c -0.01450
8016.06c -0.00125

% --- Coolant (40% void fraction):

mat cool -0.443760 moder lwtr 1001
1001.06c 0.66667
8016.06c 0.33333

% --- Moderator:

mat moder -0.739605 moder lwtr 1001
1001.06c 0.666667
8016.06c 0.333333


The number after the name is material density. Negative numbers refer to mass densities (in g/cm3) and positive numbers to atomic densities (in 1/barn*cm). Followed by that is the isotopic composition. All nuclides are in format "<Z><A>.<id>", where <Z> is the element number and <A> is the isotope mass. Uranium-235, for example is 92235. Zero mass numbers refer to elements in their natural isotopic composition. 40000, for example, is natural zirconium. The id number determines the library. In the data files distributed with the Serpent installation package, the library id's are related to temperature (03c = 300K, 06c = 600K, etc.). Following nuclide identifier is the nuclide fraction. Again, negative numbers refer to mass fractions and positive numbers to atomic fractions.

The additional parameters determine things like the number of neutron histories run, the type of boundary conditions, cross section library file path, etc., etc.

If you are starting from scratch, learning the input format may take some time. The basic features are described in the User's Manual, found at the website: http://montecarlo.vtt.fi, but in my opinion the best way to learn is to start from an existing case and simply play around with the values to see what happens. All examples at the website produce png-format plot files that show the geometry. Since Serpent uses a universe-based geometry, it might be helpful to look into some MCNP tutorials as well.

I hope this helps. Let me know if you have any questions or if you get stuck with some problem.

Ps. Questions about the input are perhaps better suited under the "users" topic.
- Jaakko

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