Abstract
The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a three-dimensional multi-group nodal diffusion code developed for the steady and transient states using microscopic cross-sections generated by the STREAM code for 37 isotopes. A depletion chain containing 22 actinides and 15 fission products and burnable absorbers was solved using the Chebyshev rational approximation method. A simplified one-dimensional single-channel thermal-hydraulic calculation was performed with various values for the thermal conductivity. Advanced features such as burnup adaptation and CRUD modeling capabilities are implemented for the multi-cycle analysis of commercial reactor power plants. The performance of RAST-K v2 has been validated with the measured data of PWRs operating in Korea. Furthermore, RAST-K v2 has been coupled with a sub-channel code (CTF), fuel performance code (FRAPCON), and water chemistry code for multiphysics analyses. In this paper, the calculation models and engineering features implemented in RAST-K v2 are described, and then the application status of RAST-K v2 is presented.
Highlights
Nodal diffusion codes are generally employed in the second part of the conventional two-step approach in the traditional procedure for nuclear reactor core analysis and design
In contrast with existing nodal diffusion codes, it is easy to implement new engineering features in RAST-K v2 owing to the fact that we developed the source as well as the executable
STREAM-spent nuclear fuel (SNF) was employed for a code-to-code comparison, which showed that the results of the isotope inventory obtained using RAST-K v2 agree with those obtained using the STREAM-SNF within a relative error of 5%
Summary
Nodal diffusion codes are generally employed in the second part of the conventional two-step approach in the traditional procedure for nuclear reactor core analysis and design. The RAST-K v1 [9,10] was developed for evaluating the static and dynamic control abilities of control rods during dynamic control rod measurements (DCRMs) It was employed in a real-time core simulator for 21 pressurized water reactors (PWRs) operating in South Korea. In. 2006, it was licensed by the Korean regulatory body as a core transient analysis code to be used in the zero-power condition of PWRs. The RAST-K v1 uses a non-linear coarse mesh finite difference method with a nodal expansion method (NEM) or analytic nodal method (ANM) as a neutronic solver. In contrast with existing nodal diffusion codes, it is easy to implement new engineering features in RAST-K v2 owing to the fact that we developed the source as well as the executable This supports convenient core design, analysis, and coupling with other physics codes and platforms.
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