• Title, Summary, Keyword: Fuel Rod

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Validation of the fuel rod performance analysis code FRIPAC

  • Deng, Yong-Jun;Wei, Jun;Wang, Yang;Zhang, Bin
    • Nuclear Engineering and Technology
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    • v.51 no.6
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    • pp.1596-1609
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    • 2019
  • The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

MODAL TESTING AND MODEL UPDATING OF A REAL SCALE NUCLEAR FUEL ROD

  • Park, Nam-Gyu;Rhee, Hui-Nam;Moon, Hoy-Ik;Jang, Young-Ki;Jeon, Sang-Youn;Kim, Jae-Ik
    • Nuclear Engineering and Technology
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    • v.41 no.6
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    • pp.821-830
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    • 2009
  • In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

Spacer Grid Assembly with Sliding Fuel Rod Support (삽입 및 이동 가능한 연료봉 지지부의 지지격자 형상)

  • Song, Kee-Nam;Lee, Sang-Hoon
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.34 no.7
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    • pp.843-850
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    • 2010
  • A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

MEASUREMENT OF NUCLEAR FUEL ROD DEFORMATION USING AN IMAGE PROCESSING TECHNIQUE

  • Cho, Jai-Wan;Choi, Young-Soo;Jeong, Kyung-Min;Shin, Jung-Cheol
    • Nuclear Engineering and Technology
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    • v.43 no.2
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    • pp.133-140
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    • 2011
  • In this paper, a deformation measurement technology for nuclear fuel rods is proposed. The deformation measurement system includes a high-definition CMOS image sensor, a lens, a semiconductor laser line beam marker, and optical and mechanical accessories. The basic idea of the proposed deformation measurement system is to illuminate the outer surface of a fuel rod with a collimated laser line beam at an angle of 45 degrees or higher. For this method, it is assumed that a nuclear fuel rod and the optical axis of the image sensor for observing the rod are vertically composed. The relative motion of the fuel rod in the horizontal direction causes the illuminated laser line beam to move vertically along the surface of the fuel rod. The resulting change of the laser line beam position on the surface of the fuel rod is imaged as a parabolic beam in the high-definition CMOS image sensor. An ellipse model is then extracted from the parabolic beam pattern. The center coordinates of the ellipse model are taken as the feature of the deformed fuel rod. The vertical offset of the feature point of the nuclear fuel rod is derived based on the displacement of the offset in the horizontal direction. Based on the experimental results for a nuclear fuel rod sample with a formation of surface crud, an inspection resolution of 50 ${\mu}m$ is achieved using the proposed method. In terms of the degree of precision, this inspection resolution is an improvement of more than 300% from a 150 ${\mu}m$ resolution, which is the conventional measurement criteria required for the deformation of neutron irradiated fuel rods.

Vibration Characteristics of a Nuclear Fuel Rod in Uniform Axial Flow (균일한 축방향 유동에 노출된 핵 연료봉의 진동특성 분석)

  • Jeon, Sang-Youn;Suh, Jung-Min;Kim, Kyu-Tae;Park, Nam-Gyu
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.16 no.11
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    • pp.1115-1123
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    • 2006
  • Nuclear fuel rods are exposed to axial flow in a reactor, and flow-induced-vibration due to the flow usually causes damage in the fuel rods. Thus a prior knowledge about dynamic behavior of a fuel rod exposed to the flow condition should be provided. This paper shows that dynamic characteristics of a nuclear fuel rod depend on axial flow velocity. Assuming small lateral displacement, the effects of uniform axial flow are investigated. The analytic results show that axial flow generally reduces fuel rod stiffness and raises its damping in normal condition. Also, the critical axial velocities which make the fuel rod behavior unstable were found. That is, solving generalized eigenvalue equation of the fuel rod dynamic system, the eigenvalues with positive real part are detected. Based on the simulation results, on the other hand, it turns out that the ordinary axial flow in nuclear reactors does not affect to stability of a nuclear fuel rod even in the conservative condition.

Modal Analysis and Testing for a Middle Spacer Grid of a Nuclear Fuel Rod (핵 연료봉 중간 지지격자의 모달 해석 및 실험)

  • Ryu, Bong-Jo;Koo, Kyung-Wan
    • The Transactions of The Korean Institute of Electrical Engineers
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    • v.61 no.12
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    • pp.1948-1952
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    • 2012
  • The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.

Development of the Spent Fuel Rod Cutting Device by Cutter Blade Method (Cutter blade 방식에 의한 사용후핵연료봉 절단 장치 개발)

  • 정재후;윤지섭;홍동회;김영환;김도우
    • Proceedings of the Korean Society of Precision Engineering Conference
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    • pp.393-396
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    • 2000
  • Spent fuel rod cutting device should cut a spent fuel rod to an optimal size in order to fast decladding operation. In this paper, for developing spent fuel rod cutting device with cutter blade, rod properties such as dimension and material of zircaloy tube and fuel pellet are investigated at first and then, various methods of existing cutting devices used commercially are investigated and their performance are analyzed and compared. This device is designed to be operated automatically via remote control system considering later use in Hot-Cell (radioactive area) and the mdularization in the structure of this device makes maintenance easy. SUS and Zircaloy-4 are selected as cut material used in the test of spent fuel rod cutting device by cutter blade. In order for constructing the high durable cutter blade, various materials are analyzed in terms of quality, shape, characteristic, and heat treatment, etc. and from these results, spent fuel rod cutting device is designed and manufactured based on the considerations of durability, round shape sustainability of rod cross-section, debris generation, and fire risk, etc.

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Development of Structural Analysis Modeling for KALIMER Fuel Rod

  • Kang, Hee-Young;Cheol Nam;Woan Hwang
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.175-180
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    • 1998
  • The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

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A Study on the Hydraulic Stability of Fuel Rod for the Advanced $16{\times}16$ Fuel Assembly Design ($16{\times}16$ 개량핵연료 연료봉의 수력적 안정성에 관한 연구)

  • Jeon Sang-Youn
    • Journal of the Computational Structural Engineering Institute of Korea
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    • v.18 no.4
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    • pp.347-360
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    • 2005
  • The fuel rod instability can be occurred because of the axial and cross flow due to the flow anomaly and/or flow redistribution in the lower core plate region of the pressurized water reactor. The fuel rod vibration due to the hydraulic instability is one of the root causes of fuel failure. The verification on the fuel rod vibration and instability is needed for the new fuel assembly design to verify the fuel rod instability. In this study, the fuel rod vibration and stability analyses were performed to investigate the effect of the grid height, fuel rod support condition, and span adjustment on the fuel rod vibration characteristics for the advanced $16{\times}16$ fuel assembly design. Based on the analysis results, the grid height and grid axial elevation of the advanced $16{\times}16$ fuel assembly design were proposed.

A Theoretical Analysis of the Acceptability of Design Stress Value for the Fuel Rod with Nonlinear Thermal Stresses (비선형 분포의 열응력이 작용하는 Fuel Rod에서 설계 응력값의 적합성여부에 대한 이론적 해석)

  • 호광일
    • Journal of Energy Engineering
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    • v.12 no.3
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    • pp.177-183
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    • 2003
  • The purpose of this paper is to verify that the design stress value of fuel rod for the irradiation test satisfies the structural design requirement. In this structural safety analysis thermal effect is the most severe element for the safety. The thermal effects are very complicated problem to be analyzed for the structural safety in short hand. By the application of theoretical analysis, the design margin of stress which was used in this fuel rod design was verified in the conservative point of view. In the future design of fuel rod, this analysis can be used as the theoretical method for the verification of safe design.