Addition of Central Control Rod in Small Modular Reactor Geometry Using Serpent

AI Thread Summary
To create a small modular reactor geometry in Serpent code, users need to define the reactor's geometry accurately, including the placement of a control rod at the center. It is essential to incorporate the correct code scripts in the geometry section to facilitate this setup. For a hexagonal lattice configuration, specific parameters must be adjusted to ensure proper representation of the reactor's structure. Users are encouraged to refer to Serpent documentation for guidance on scripting and geometry definitions. Proper implementation will enable accurate modeling of reactor behavior with the added control rod.
emilmammadzada
Messages
129
Reaction score
19
TL;DR Summary
Addition of Central Control Rod in Small Modular Reactor Geometry Using Serpent
Dear Experts I would like to create a small modular reactor-like geometry for Serpent code (example: ABV reactor). How can I add a control rod to the center like in the picture below? Which code scripts should I add in the geometry part. And how can I do this in Hexagonal laticce?
 

Attachments

  • Fuel-assembly-of-ABV-reactor-IAEA-International-Atomic-Energy-Agency-2007_Q320.jpg
    Fuel-assembly-of-ABV-reactor-IAEA-International-Atomic-Energy-Agency-2007_Q320.jpg
    27.5 KB · Views: 30
  • serpent-input.txt
    serpent-input.txt
    9.2 KB · Views: 29

Thread 'MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly'

Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information（power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.） for Material 2 is zero, while Material 1...
View full post »

Thread 'Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)'

Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6? If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning. I'd be really...
View full post »

Similar threads

Hexagonal fuel arrays (VVER and fast reactor fuel)
Replies
0
Views
2K
Fukushima Candu Reactor compared k_eff values with Serpent
Replies
0
Views
2K
Design error in reactor nuclear simulation in serpent code
Replies
1
Views
3K
Basic understanding of Graphite-Water reactors
Replies
1
Views
2K
Reactor Channel Layout: RBMK, PWR, CANDU Comparison
Replies
1
Views
4K
B How Does an IEC Fusion Reactor Work?
Replies
19
Views
5K
Automatic Window Lift: Get Help Setting Up Low-Cost, Efficient System
Replies
4
Views
2K
Fixing Error: (vlog-7) Failed to Open Design Unit File
Replies
4
Views
7K
MATLAB Matlab Derivative block analysis and filter design
Replies
11
Views
3K
Requesting suggestions for languages, libraries, and architectures for parallel (and sometimes non parallel) numerical and scientific computations
Replies
8
Views
4K

Hot Threads

Recent Insights

Back
Top