• Nuclear Engineering and Technology
• /
• 제44권2호
• /
• pp.151-160
• /
• 2012

The advantages for using Monte Carlo methods to analyze full-core reactor configurations include essentially exact representation of geometry and physical phenomena that are important for reactor analysis. But this substantial advantage comes at a substantial cost because of the computational burden, both in terms of memory demand and computational time. This paper focuses on the challenges facing full-core Monte Carlo for keff calculations and the prospects for Monte Carlo becoming a routine tool for reactor analysis.

• Nuclear Engineering and Technology
• /
• 제55권5호
• /
• pp.1687-1707
• /
• 2023

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann's model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon's 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

• Nuclear Engineering and Technology
• /
• 제55권6호
• /
• pp.2230-2245
• /
• 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
• /
• 제52권12호
• /
• pp.2699-2708
• /
• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 지구물리와물리탐사
• /
• 제23권1호
• /
• pp.1-12
• /
• 2020

섭입 및 열개와 같이 대변형을 수반하는 지구동역학적 현상 발생은 암석권의 국지적인 약대의 발달이 필요하다. 이러한 약화 기작 중 하나인 전단열은 암석권의 온도를 국부적으로 높여 강도를 낮추는 역할을 하여 암석권 파괴를 촉진시킬 수 있다. 본 연구에서는 전단열에 대한 정량적인 분석을 위하여 2차원 탄소성 인장 분지 모형을 제작하여 기존 수치 모사 연구를 벤치마크하였다. 암석권의 항복강도, 인장 속도, 변형량- 및 온도-의존성 약화 현상 등을 조절하여 전단열 발생량에 미치는 영향을 분석하였다. 실험 결과, 약화를 고려하지 않은 경우 전단열의 발생량은 암석권의 항복강도 및 인장 속도와 양의 상관관계가 있는 것으로 나타났다. 기준 모형인 항복강도 100 MPa, 인장 속도 2 cm/yr로 설정된 경우, 총 20 km 인장된 시점(0.025의 변형률)에서 ~ 50 K의 온도 상승을 보여주었다. 소성 변형 및 온도에 따른 약화가 포함된 경우에는, 더 효율적인 약화 기작이 더 강한 전단열의 생성으로 이어지는데 이러한 현상은 약화 기작과 전단열 발생 사이에 양성되먹임이 작용함을 지시한다. 또한 변형 초기에 급격한 전단열 발생량을 보여주지만, 변형이 지속되어 암석권의 강도가 약화되면 전단열 발생 속도가 최대 ~ 80% 감소했다. 이는 약화 기작이 포함된 경우 전단열은 비교적 손상되지 않은 상태인 암석권의 강도에 큰 영향을 미침을 시사한다.