지원사업
학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
커뮤니티
연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2023.6
- 수록면
- 2,230 - 2,245 (16page)
- DOI
- 10.1016/j.net.2023.02.025
이용수
초록· 키워드
오류제보하기
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 cal culations 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 bench marks 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.
목차
등록된 정보가 없습니다.
참고문헌 (0)
참고문헌 신청함께 읽어보면 좋을 논문
논문 유사도에 따라 DBpia 가 추천하는 논문입니다. 함께 보면 좋을 연관 논문을 확인해보세요!
-
Solution of OECD/NEA PWR MOX/UO2 benchmark with a high-performance pin-by-pin core calculation code
Nuclear Engineering and Technology
2024 .09
-
Development and verification of PWR core transient coupling calculation software
Nuclear Engineering and Technology
2021 .11
-
Approximate core sizing of PWR600 MWe based on APR1400
대한기계학회 춘추학술대회
2016 .12
-
The JFNK method for the PWR's transient simulation considering neutronics, thermal hydraulics and mechanics
Nuclear Engineering and Technology
2020 .01
-
Review on Revision History of ANSI/ANS-18.1 and PWR-GALE Along With Its Potential Impact to Anticipated Radioactive Source Terms to Environment From PWRs
한국방사성폐기물학회 학술대회
2019 .01
-
Design and analysis of RIF scheme to improve the CFD effi ciency of rod-type PWR core
Nuclear Engineering and Technology
2021 .10
-
Creep Model for PWR Spent Nuclear Fuel
한국방사성폐기물학회 학술대회
2015 .01
-
Verification and Validation of STREAM/RAST-K for PWR Analysis
Nuclear Engineering and Technology
2019 .01
-
Development and Validation of a Fast Sub-channel Code for LWR Multi-physics Analyses
Nuclear Engineering and Technology
2019 .01
-
Design of the flexible switching controller for small PWR core power control with the multi-model
Nuclear Engineering and Technology
2021 .03
-
PWR core calculation based on pin-cell homogenization in three-dimensional pin-by-pin geometry
Nuclear Engineering and Technology
2024 .06
-
Transportation Analysis of the 14OFA PWR Fuel Assembly
한국방사성폐기물학회 학술대회
2019 .01
-
Fast Neutron Detection for Spent PWR Fuel in Dry Storage
한국방사성폐기물학회 학술대회
2017 .01
-
Preliminary Calculation of CRUD Source-term With Modified Corrosion Product Transport Model for PWR Primary Circuit
한국방사성폐기물학회 학술대회
2018 .01
-
Development of supporting platform for the fine flow characteristics of reactor core
Nuclear Engineering and Technology
2024 .05
-
Preparation of the Pin-loading Tension Test for Evaluation of Fracture Toughness in PWR Cladding Tubes
한국방사성폐기물학회 학술대회
2019 .01
-
CTF/DYN3D multi-scale coupled simulation of a rod ejection transient on the NURESIM platform
Nuclear Engineering and Technology
2017 .01
-
Preliminary Study on the Development of a Horizontal Disposal Concept for PWR Spent Fuels
한국방사성폐기물학회 학술대회
2019 .01
-
High performance 3D pin-by-pin neutron diffusion calculation based on 2D/1D decoupling method for accurate pin power estimation
Nuclear Engineering and Technology
2021 .11
-
Analysis on Disassembling Methods of Nozzle for PWR Spent Fuel
한국방사성폐기물학회 학술대회
2017 .01
이 논문의 저자 정보
최근 본 자료
댓글(0)
0