• Nuclear Engineering and Technology
• /
• 제46권5호
• /
• pp.641-654
• /
• 2014

A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.

• Nuclear Engineering and Technology
• /
• 제45권6호
• /
• pp.745-752
• /
• 2013

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 요약집 춘계학술발표대회
• /
• pp.65.2-65.2
• /
• 2004

• Structural Engineering and Mechanics
• /
• 제45권4호
• /
• pp.439-454
• /
• 2013

Hollow cylindrical tubes with a prismatic sandwich lining designed to replace the solid cross-sections are studied in this paper. The sections are divided by a number of revolving periodic unit cells and three topologies of unit cells (Square, Triangle and Kagome) are proposed. Some types of multiple-topology designed materials are also studied. The feasibility and accuracy of a homogenization method for obtaining the equivalent parameters are investigated. As the curved elements of a unit cell are represented by straight elements in the method and the ratios of the lengths of the curved elements to the lengths of the straight elements vary with the changing number of unit cells, some errors may be introduced. The frequencies of the first five modes and responses of the complete and equivalent models under an internal static pressure and an internal step pressure are compared for investigating the scope of applications of the method. The lower bounds and upper bounds of the number of Square, Triangular and Kagome cells in the sections are obtained. It is shown that treating the multiple-topology designed materials as a separate-layer structure is more accurate than treating the structure as a whole.

• Nuclear Engineering and Technology
• /
• 제54권8호
• /
• pp.3106-3116
• /
• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

• 한국전산유체공학회지
• /
• 제16권3호
• /
• pp.95-103
• /
• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.

• 한국전산유체공학회지
• /
• 제14권2호
• /
• pp.59-67
• /
• 2009

An accurate prediction of the bypass flow is of great importance in the VHTR core design concerning the fuel thermal margin. Nevertheless, there has not been much effort in evaluating the amount and the distribution of the core bypass flow. In order to evaluate the behavior and the distribution of the coolant flow, a unit-cell experiment was carried out. Unit-cell is the regular triangular section which is formed by connecting the centers of three hexagonal blocks. Various conditions such as the inlet mass flow rate, block combinations and the size of bypass gap were examined in the experiment. CFD analysis was carried out to analyze detailed characteristics of the flow distribution. Commercial CFD code FLUENT 6.3 was validated by comparing with the experimental results. In addition, SST model and standard k-$\varepsilon$ model were validated. The results of CFD simulation show good agreements with the experimental results. SST model shows better agreement than standard k-$\varepsilon$ model. Results showed that block combinations and the size of the bypass gap have an influence on the bypass flow ratio but the inlet mass flow rate does not.

• Advances in Computational Design
• /
• 제3권2호
• /
• pp.165-190
• /
• 2018

In this article, the static behavior of non-prismatic sandwich beams composed of functionally graded (FG) materials is investigated for the first time. Two types of beams in which the variation of elastic modulus follows a power-law form are studied. The principle of minimum total potential energy is applied along with the Ritz method to derive and solve the governing equations. Considering conventional boundary conditions, Chebyshev polynomials of the first kind are used as auxiliary shape functions. The formulation is developed within the framework of well-known Timoshenko and Reddy beam theories (TBT, RBT). Since the beams are simultaneously tapered and functionally graded, bending and shear stress pushover curves are presented to get a profound insight into the variation of stresses along the beam. The proposed formulations and solution scheme are verified through benchmark problems. In this context, excellent agreement is observed. Numerical results are included considering beams with various cross sectional types to inspect the effects of taper ratio and gradient index on deflections and stresses. It is observed that the boundary conditions, taper ratio, gradient index value and core to the thickness ratio significantly influence the stress and deflection responses.

• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.248-257
• /
• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• 대한기계학회논문집
• /
• 제17권6호
• /
• pp.1412-1422
• /
• 1993

본 연구는 각주형 금형부품을 절삭가공할때 부품의 설계데이터로부터 부품의 형상적인 특징을 추출하여 공정설계시스템인 MOLDCAPP과 작업설계시스템인 COPS에서 사용할 수 있는 정보를 생성함으로써 CAD/CAM의 연결을 자동화할 수 있는 특징형상 인식 시스템을 개발하는 연구로, 특징형상 인식기법의 창안보다는, 가용한 기술의 장 점을 사용하여 각주형 공작물의 기계 절삭가공으로 생성될 수 있는 형상을 특징형상으 로 정의하고 ACIS로 설계된 CAD데이터로부터 정의된 특징형상을 추출하여 각 특징형상 들의 형상 데이터를 결정함으로써 MOLDCAPP, COPS 등의 공절설계시스템의 입력데이터 를 생성할 수 있도록 Fig.1과 같이 설계하였다. 특히 PART시스템과 같이 인식대상이 포괄적이지 않으나, 금형부품상의 특징형상으로 범위를 축소하고 금형부품의 가공특징 을 고려하여 인식규칙을 단순화함으로써 금형가공공정의 CAD/CAM연결에 이용될 수 있도록 하였다.