• Structural Engineering and Mechanics
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• 제2권2호
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• pp.157-171
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• 1994

A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced form earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed.

• 대한기계학회논문집A
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• 제27권3호
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• pp.424-431
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• 2003

The spacer grid is one of the main structural components in the fuel assembly, which supports the fuel rods and maintains coolable geometry from an external load. A vibration test and a finite element analysis using ABAQUS on a dummy fuel rod continuously supported by Optimized H type(OHT) and New Doublet (ND) spacer grids arc performed to obtain the vibration characteristics such as natural frequencies and mode shapes an(1 to verify a finite element model. The results from the test and the finite element analysis are compared by modal assurance criteria (MAC) values. It is resulted that MACs for the first, the third and the fifth mode shapes are relatively good as compared with those of the second an(1 fourth ones. The natural frequency differences between two methods as well as the mode comparison results for the rod with OHT spacer grid are better than those with ND spacer grid. It is judged that the FE model for the ND spacer grid spring should be modified to consider the long contact length which actually happen when the spring supports the rod.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.292-296
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• 2018

Recent events in the nuclear industry have resulted in a movement towards increased seismic and LOCA excitations and requirements that challenge current fuel designs. AREVA NP's GAIA fuel design introduces unique and robust characteristics to resist the effects of seismic and LOCA excitations. For demanding seismic and LOCA scenarios, fuel assembly spacer grids can undergo plastic deformations. These plastic deformations must not prohibit the complete insertion of the control rod assemblies and the cooling of the fuel rods after the accident. The specific structure of the GAIA spacer grid produces a unique and stable compressive deformation mode which maintains the regular array of the fuel rods and guide tubes. The stability of the spacer grid allows it to absorb a significant amount of energy without a loss of load-carrying capacity. The GAIA-specific grid behavior is in contrast to the typical spacer grid, which is characterized by a buckling instability. The increased mechanical robustness of the GAIA spacer grid is advantageous in meeting the increased seismic and LOCA loadings and the associated safety requirements. The unique GAIA spacer grid behavior will be incorporated into AREVA NP's licensed methodologies to take full benefit of the increased mechanical robustness.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.547-555
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• 2001

In pressurized light water reactor fuel assembly, spacer grids support nuclear fuel rods both laterally and vertically. The fuel rods are supported by spacer grid springs and grid dimples that are located in the grid cell. The support system allows for some thermal expansion and imbalance of the fuel rods. The imbalance is absorbed by elastic energy to prevent coolant flow- induced vibration damage. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. Since the optimization is carried out in the linear range of finite element analysis, the optimum solution is verified by nonlinear analysis. A good design is found and the final design is compared with the initial conceptual design. Commercial codes are utilized for structural analysis and optimization.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.303-303
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• 2004

• 대한기계학회논문집A
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• 제35권2호
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• pp.201-206
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• 2011

지지격자체는 경수로 핵연료집합체의 특성과 성능에 영향을 주는 가장 중요한 핵심 구조부품 중에 하나이다. 지지격자체 설계시의 우선적으로 고려해야할 사항은 핵연료가 원자로에 장전되어 있는 동안 내내 연료봉이 기계적인 원인에 의해 손상되지 않도록, 즉 연료봉의 기계적 지지건전성이 유지되도록 설계하는 것이다. 연료봉이 유동기인진동에 의해서 진동할 때 연료봉과 연료봉 지지부 사이에서 상대변위 발생을 완화해 줌으로서 연료봉의 프레팅 마모 손상 가능성이 감소될 수 있는 것으로 알려져 있다. 본 연구에서는 이동 가능한 연료봉 지지부로 구성된 새로운 지지격자체 형상을 제안하였고, 제안된 이동 가능 지지부의 연료봉 지지특성을 유한요소해석을 통해 분석하였다.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1119-1126
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• 2021

Spacer grid with mixing vane had been widely used in nuclear reactor core. One of the main feather of spacer grid with mixing vane was that strong swirl flow was formed after the spacer grid. The swirl flow not only changed the bubble generation in the near wall field, but also affected the bubble behaviors in the center region of the subchannel. The interaction between bubble and the swirl flow was one of the basic phenomena for the two phase flow modeling in fuel assembly. To obatin better understanding on the bubble behaviors in swirl flow, full three dimension numerical simulations were conducted in the present paper. The swirl flow was assumed in the cylindral calculation domain. The bubble interface was captured by Volume Of Fluid (VOF) method. The properties of saturated water and steam at different pressure were applied in the simulation. The bubble trajectory, motion, shape and force were obtained based on the bubble parameters captured by VOF. The simulation cases in the present study included single bubble with different size, at different angular velocity conditions and at different pressure conditions. The results indicated that bubble migrated to the center in swirl flow with spiral motion type. The lateral migration was mainly related to shear stress magnitude and bubble size. The bubble moved toward the center with high velocity when the swirl magnitude was high. The largest bubble had the highest lateral migration velocity in the present study range. The effect of pressure was small when bubble size was the same. The prelimenery simulation result would be beneficial for better understanding complex two phase flow phenomena in fuel assembly with spacer grid.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.21-36
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• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• 소음진동
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• 제6권6호
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• pp.827-834
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• 1996

본연구에서는 핵연료집합체의 검증계획의 일환으로 2차측 배관파단의 영향을 조사하였다. 원자로노심의 상세모델을 이용한 동적해석으로 배관파단에 의한 응답을 구하였다. 파단적 누설개념의 적용으로 10인치 이상의 고에너지 배관에 대하여 양단 파단이 설계에서 배제됨에 따라 본 연구에서는 주증기관과 급수관의 파단을 가정 하였다. 핵연료 집합체의 전단력, 굽힘모우멘트, 변위 및 지지격자체의 충격하중에 대하여 자세히 고찰하였고 이들 동적해석 결과를 이용하여 핵연료집합체의 구조적 건전성을 평가하였으며 사고조건에서 2차측 배관파단이 핵연료집합체의 구조적 건전성 에 미치는 영향을 검토하였다.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.