Solving Neutron Problems with Commercial Finite Element Method Codes

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SUMMARY

The discussion centers on the challenges of solving neutron problems in nuclear reactors using commercial finite element method codes, specifically questioning the applicability of ANSYS for this purpose. It is established that neutron transport calculations for fuel assemblies and neutron diffusion calculations for the entire reactor core are typically performed using codes like CASMO and SIMULATE from STUSVIK. While ANSYS offers good expansibility for high-performance computing (HPC), there is skepticism regarding its effectiveness for neutronics applications, with alternatives like DENOVO and MPACT being developed for 3D finite element analysis (FEA) based neutronics.

PREREQUISITES
  • Understanding of neutron transport and diffusion calculations
  • Familiarity with CASMO and SIMULATE codes from STUSVIK
  • Knowledge of high-performance computing (HPC) principles
  • Awareness of finite element analysis (FEA) methodologies
NEXT STEPS
  • Research the capabilities of DENOVO and MPACT for neutronics applications
  • Explore the integration of cross-section libraries in reactor modeling
  • Investigate the limitations of ANSYS in neutron problem-solving
  • Study the development of 3D FEA-based neutronics codes
USEFUL FOR

Nuclear engineers, reactor physicists, and computational scientists interested in advanced methods for neutron problem-solving and those evaluating the use of commercial software in nuclear applications.

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As I know, the method to solve neutron problem is divided into two steps now, neutron transport calculation for fuel assemblies and neutron diffusion calculation for whole reactor core, both using specified code such as CASMO and SIMULATE from STUSVIK. I want to know whether the commercial finite element method code such as ANSYS can be edited to solve neutron problem by one step, using fine energy group structure. For the code like ANSYS has very good expansibility and is very suitable for the HPC calculation, we may have a code not very fast but can work on HPC cluster to get accurate solution for neutron problem. Is there any efforts on such work had been done by any institute or company all over the world?
 
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I don't believe anyone is seriously proposing to solve fuel rod/assembly/core neutronics with ANSYS. However, there are efforts to build 3D FEA-based neutronics codes, e.g., DENOVO and MPACT (derived from DeCART).

One needs a cross-section library for each assembly lattice as a function of burnup. CASMO is the cross-section library for SIMULATE. SIMULATE has a simple fuel rod model for calculating fuel temperature and a relatively simple thermal-hydraulics model for developing the coolant (moderator) state. I don't see the value in doing this with ANSYS.
 


Yes, I can't see ANSYS being useful in this application.
 

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