ene6953a a brief serpent tutorial - polytechnique montréal · 3d monte carlo code specially...

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ENE6953A: méthodes numériques pour le génie nucléaire

ENE6953A

A brief Serpent tutorial

Serpent tutorial

3D Monte Carlo code specially designed for lattice physics calculations Homogenization Burnup calculation

Similar in some aspects to many Monte Carlo codes point-wise cross sections (ACE format) Universe-based geometry model (KENO/MCNP/TART…) Classical lattices (PWR/BWR, cluster, hexagonal) are

directly available

See the website for examples, documentation, forum, etc: http://montecarlo.vtt.fi/


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Serpent tutorial

Analog Monte Carlo simulation: sufficient in most of reactor physics application

Delta-tracking method combined with the regular ray tracing method (parametrized with a threshold DT in [0,1])

Collision estimator for detector calculations

Keff estimates: Analog (generation-wise) Collision Absorption=implicit


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Serpent tutorial: input example


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Serpent tutorial


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Isotopic identifications: Isotopes are given through ZAId numbers

ZZZAAA.xxc numbers

ZZZAAA: 6-digits ZZZ: atomic number Z AAA: atomic mass A Ex: U235=> 092235, or simply 92235 : H1 => 1001

xx is corresponding to the temperature at which the cross sections were produced

03.c: 300 K, 06.c: 600 K, 09c: 900 K, etc…

See the document SSS_ACE_JEFF31 for assign #3 ENE6953A: méthodes numériques pour le génie nucléaire

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Isotopic identifications: isotopic densities


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Put (-) before mass fraction


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Isotopic identifications: mass fractions

When atoms (hydrogen in water, deuterium in heavy water, etc.) are molecularly bound, binding affects energy loss in collision experienced by slow neutrons.

Special cross-sections are required ( S(α,β) data)

See the document SSS_THERMAL for thermal data (JEFF3.1 section for assign #3)


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Isotopic identifications: thermal data


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Doppler broadening

• Temperature cross-sections can be adjusted using the keyword “tmp”

• Always use a value above the initial data:


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Geometry: pin cell example


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Geometry definition:

SERPENT geometric model is a classic in Monte Carlo codes

Universe-based model permits to divide in several level the geometry

Example: 1. Universe (1): pin cell 2. Universe (2): assembly fuel lattice, filled by universe (1) 3. Universe (3): lattice of assembly fuels, filled by (2) 4. Real reactor : (3)+reflector+ baffle, etc.


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Geometry definition in Serpent:

Geometry is defined in terms of regions (or volumes) bounded by surfaces (plans, quadratic, etc.)


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Geometry definition in Serpent:

Some useful surfaces for reactor calculations:


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Geometry definition in general MC codes: Similar to MCNP5, KENO6, etc.

Some MCNP5 surface definition:


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Cells are then defined by a set of positive and negative surfaces Positive surface number: cell is located “inside” Negative surface number: neutorn is located “outside”

In MCNP: logical operator permit to define complicated shapes (union, intersection, complementary)


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Example of simple cell definitions


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Example of simple cell definitions


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SERPENT provides derived surface types for lattice calculations:

Pin cards

Lattices: regular, hexagonal, cluster

To be used for the assignment (see the documentation!)


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Simulation:

Similar to the KCODE card in MCNP

Main parameters: number of neutrons/cycle, number of cycle, number of skipped cycles


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Bc 2: reflective bc

gcu: universe for group constant generation (spatial homogenization)

Sym: symetry for group constant generation

Nfg: energy directive for energy condensation

Pop: MC parameters

Plot & mesh : picture of the geometry

Running SERPENT:

In doppler/boltzmann computer

./rserpent input_file

Output files: .png (geometry and mesh) Det.m (if detector card defined) .out : check of the geometry and materials .seed: seed used by the RNG Input_file.m: contains the important results (Keff, group

constants, etc.)


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Output file:


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Output file:


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Output file:


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Output file:


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ene6953a a brief serpent tutorial - polytechnique montréal · 3d monte carlo code specially...

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