ENE6953A A brief Serpent tutorial

Transcription

1 1 ENE6953A A brief Serpent tutorial

2 2 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:

3 3 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

4 Serpent tutorial: input example 4

5 Serpent tutorial 5

6 6 Isotopic identifications: Isotopes are given through ZAId numbers ZZZAAA.xxc numbers ZZZAAA: 6-digits ZZZ: atomic number Z AAA: atomic mass A Ex: U235=> , or simply : 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

7 Isotopic identifications: isotopic densities 7

8 Isotopic identifications: mass fractions 8 Put (-) before mass fraction

9 Isotopic identifications: thermal data 9 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)

10 Doppler broadening 10 Temperature cross-sections can be adjusted using the keyword tmp Always use a value above the initial data:

11 11

12 Geometry: pin cell example 12

13 13 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.

14 Geometry definition in Serpent: 14 Geometry is defined in terms of regions (or volumes) bounded by surfaces (plans, quadratic, etc.)

15 Geometry definition in Serpent: 15 Some useful surfaces for reactor calculations:

16 16 Geometry definition in general MC codes: Similar to MCNP5, KENO6, etc. Some MCNP5 surface definition:

17 17 Geometry definition: 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)

18 Geometry definition: 18 Example of simple cell definitions

19 Geometry definition: 19 Example of simple cell definitions

20 Geometry definition: 20 SERPENT provides derived surface types for lattice calculations: Pin cards Lattices: regular, hexagonal, cluster To be used for the assignment (see the documentation!)

21 Simulation: 21 Similar to the KCODE card in MCNP Main parameters: number of neutrons/cycle, number of cycle, number of skipped cycles 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

22 22 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.)

23 Output file: 23

24 Output file: 24

25 Output file: 25

26 Output file: 26