• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.289.1-289
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.109.2-109
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• 1999

• IE interfaces
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• v.22 no.2
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• pp.104-115
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• 2009

The objective of the study is to evaluate the effectiveness of Crew Resource Management (CRM) training program for operators in the Main Control Room (MCR) simulator of APR-1400 Nuclear Power Plant. The experiments were conducted for two different crews of operators performing six different emergency operating scenarios during four-week period. Each crew consisted of the five operators: senior reactor operator, safety technical advisor, reactor operator, turbine operator, and electric operator. All crews (Crew A and B) participated in the training program for the technical knowledge and skills which were required to operate the simulator of the MCR during the first week. To verify the effectiveness of the CRM training program; however, only Crew A was selected to attend the CRM training after the technical knowledge and skills training. The results of the experiments showed that the CRM training program improved the individual attitudes of Crew A significantly. Team skills of Crew A were found to be significantly better than those of Crew B. The CRM training did not have positive effects on enhancing the individual performance of Crew A; however, as compared to that of Crew B. Implication of these findings was discussed further in detail.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• Journal of the Ergonomics Society of Korea
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• v.26 no.1
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• pp.93-97
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• 2007

As many as 111 reactor trips have occurred for recent 5 years('01-'05), and 26 cases of them have occurred due to human error. The trend of human error rate didn't decrease in 2004, so KHNP started to make efforts to decrease human errors. In 2006 KHNP bench marked excellent foreign nuclear power plants and introduced human error prevention tools. In addition, KHNP created as many as 40 posters for human performance improvement. The posters are based on the about 500 real incident reports collected through K-HPES from 1995. Therror preventive tols for the poster. This paper explains design of the posters and their application.

• Journal of the Korea Society for Simulation
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• v.15 no.1
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• pp.11-17
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• 2006

This paper shows a case study of designing a PLC logic simulator that was developed to simulate and verify PLC control programs for nuclear plant systems. The nuclear control system requires strict restrictions rather than normal process control system does, as it works with a high-risky and dangerous nuclear plant. One is that it should assure the safeness of the control programs by exploiting severe testing. The other restriction is that the control programs should be executed fast enough such that they could control multi devices concurrently in real-time. To cope with these restrictions, we devised a logic compiler which generates C-code programs from given PLC logic programs. Once the logic program was translated into C-code, the program could be analyzed by conventional software analysis tools and could be used to construct a fast logic simulator after cross-compiling, in fact, that is a kind of compiled-code simulator.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.263-263
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.222-222
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.269.2-269
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• 2002