Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics.
A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR'. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. Preliminary testing of the code was performed using a 3x3 PWR pin-cell problem. The preliminary results are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k{sub eff} and the power profile was observed. Increased computational capabilities and improvements in computational methods have accelerated interest in high fidelity modeling of nuclear reactor cores during the last several years. High-fidelity has been achieved by utilizing full core neutron transport solutions for the neutronics calculation and computational fluid dynamics solutions for the thermal-hydraulics calculation. Previous researchers have reported the coupling of 3D deterministic neutron transport method to CFD and their application to practical reactor analysis problems. One of the principal motivations of the work here was to utilize Monte Carlo methods to validate the coupled deterministic neutron transport and CFD solutions. Previous researchers have successfully performed Monte Carlo calculations with limited thermal feedback. In fact, much of the validation of the deterministic neutronics transport code DeCART in was performed using the Monte Carlo code McCARD which employs a limited thermal feedback model. However, for a broader range of temperature/fluid applications it was desirable to couple Monte Carlo to a more sophisticated temperature fluid solution such as CFD. This paper focuses on the methods used to couple Monte Carlo to CFD and their application to a series of simple test problems.
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- DE-AC02-06CH11357
- OSTI ID:
- 991607
- Report Number(s):
- ANL/NE/CP-58661; TRN: US1007517
- Resource Relation:
- Conference: Joint International Topical Meeting on Mathematics & Computations and Supercomputing in Nuclear Applications (M&C + SNA 2007); Apr. 15, 2007 - Apr. 19, 2007; Monterey, CA
- Country of Publication:
- United States
- Language:
- ENGLISH
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Related Subjects
AUDITS
COMPUTER CODES
COMPUTERIZED SIMULATION
CROSS SECTIONS
FEEDBACK
FLUID MECHANICS
MONTE CARLO METHOD
NEUTRON TRANSPORT
POWER DENSITY
PROGRAMMING LANGUAGES
REACTOR PHYSICS
REACTORS
SIMULATION
TEMPERATURE DISTRIBUTION
TESTING
THERMAL HYDRAULICS
TRANSPORT
VALIDATION