Simulation of the RPI Reactor Critical Facility Core Using High-Fidelity Neutronics Code PROTEUS

Journal Article · Sat Jul 01 04:00:00 EDT 2017 · Transactions of the American Nuclear Society

OSTI ID:23050314

Lee, C. H.; Jung, Y. S.; Shehemon, E. R. ^[1]; O'Grady, D. J. ^[2]

---

1. Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL, 60439 (United States)
2. University of Illinois, 104 South Wright Street, Urbana, IL, 61801 (United States)

The PROTEUS code is a high-fidelity three-dimensional (3D) deterministic neutron transport code developed by ANL under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The code contains the SN2ND (second order discrete ordinates method) and two MOC transport solvers, MOCFE and MOCEX (first order methods of characteristics), based on unstructured finite element meshes which allow users to model complex or unconventional geometry reactor problems. Among the two MOC solvers which normally perform better than SN for heterogeneous geometry problems, MOCEX is based on a rigorous formulation using the two-dimensional (2D) MOC radially and the discontinuous Galerkin finite element method axially to overcome the drawbacks of MOCFE with the full 3D MOC representation which requires enormous memory and computational efforts. The Reactor Critical Facility (RCF), an open tank research reactor at the Rensselaer Polytechnic Institute (RPI), became operational in 1956 under the American Locomotive company and has been operating using low-enriched uranium fuel pins, which makes it the only university research facility to use fuel rods similar to operating commercial light water reactors (LWRs). This paper presents modeling and simulation results of the RCF using PROTEUS for code verification and validation which also supports an ongoing NEUP project in which a set of benchmark validation experiments for multiphysics coupling is being developed.

• OSTI ID: 23050314
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 116; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BENCHMARKS
COMPUTERIZED SIMULATION
DISCRETE ORDINATE METHOD
ENRICHED URANIUM
FINITE ELEMENT METHOD
FUEL PINS
FUEL RODS
GEOMETRY
NEUTRON TRANSPORT
NUCLEAR ENERGY
RESEARCH REACTORS
THREE-DIMENSIONAL CALCULATIONS
TWO-DIMENSIONAL CALCULATIONS
VALIDATION
VERIFICATION
WATER COOLED REACTORS
WATER MODERATED REACTORS