• Title/Summary/Keyword: Cladding System

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Study on Microstructure and Corrosion Characteristics of Zr-0.8Sn-xNb Ternary Alloys (Nb 첨가량에 따른 Zr-0.8Sn-xNb 3원계 합금의 미세조직 및 부식특성 연구)

  • Kim, Hyeon-Gil;Jeong, Yong-Hwan;Wi, Myeong-Yong
    • Korean Journal of Materials Research
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    • v.9 no.5
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    • pp.452-459
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    • 1999
  • To develop advanced cladding materials, the effect of Nb addition on the microstructure and corrosion characteristics of Zr-0.8Sn-xNb alloys was investigated. As the Nb content increased, the grain size decreased and the volume fraction of precipitates increased. It was observed from the corrosion test at $360^{\circ}C$ that the corrosion resistance increased with decreasing Nb content. The best corrosion resistance was obtained in Zr-0.8Sn-0.2Nb alloy with high volume fraction of tetra-$ZrO_2$in the oxide. Therefore, it is suggested that Nb in theZr-0.85Sn-xNb system should be added within the solubility limit of Nb from the viewpoint of alloy design.

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WASTE CLASSIFICATION OF 17×17 KOFA SPENT FUEL ASSEMBLY HARDWARE

  • Cho, Dong-Keun;Kook, Dong-Hak;Choi, Jong-Won;Choi, Heui-Joo
    • Nuclear Engineering and Technology
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    • v.43 no.2
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    • pp.149-158
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    • 2011
  • Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

Investigation of a Hydrogen Mitigation System During Large Break Loss-Of-Coolant Accident for a Two-Loop Pressurized Water Reactor

  • Dehjourian, Mehdi;Sayareh, Reza;Rahgoshay, Mohammad;Jahanfarnia, Gholamreza;Shirani, Amir Saied
    • Nuclear Engineering and Technology
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    • v.48 no.5
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    • pp.1174-1183
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    • 2016
  • Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

Analysis of Three-Dimensional Cracks in Inhomogeneous Materials Using Fuzzy Theory

  • Lee, Yang-Chang;Lee, Joon-Seong
    • International Journal of Fuzzy Logic and Intelligent Systems
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    • v.5 no.2
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    • pp.119-123
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    • 2005
  • This paper describes a fuzzy-based system for analyzing the stress intensity factors (SIFs) of three-dimensional (3D) cracks. 3D finite element method(FEM) was used to obtain the SIF for subsurface cracks and surface cracks existing in inhomogeneous materials. A geometry model, i.e. a solid containing one or several 3D cracks is defined. Several distributions of local node density are chosen, and then automatically superposed on one another over the geometry model by using the fuzzy theory. Nodes are generated by the bucketing method, and ten-noded quadratic tetrahedral solid elements are generated by the Delaunay triangulation techniques. The singular elements such that the mid-point nodes near crack front are shifted at the quarter-points, and these are automatically placed along the 3D crack front. The complete FE model is generated, and a stress analysis is performed. The SIFs are calculated using the displacement extrapolation method. The results were compared with those surface cracks in homogeneous materials. Also, this system is applied to analyze cladding effect of surface cracks in inhomogeneous materials.

Margin Benefit Assessment of A Digital Monitoring System for Existing Analog Plants (기존 아날로그 발전소를 위한 디지탈 감시계통의 여유도 잇점평가)

  • Auh, Geun-Sun;Yoon, Tae-Young
    • Nuclear Engineering and Technology
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    • v.26 no.2
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    • pp.294-299
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    • 1994
  • Margin benefits are quantatively assessed when a Digital Monitoring System(DMS) is assumed to be installed to an operating Westinghouse analog type plant. Applied plant and cycle is YongGwang Unit 1 Cycle 6. The referenced digital monitoring system is the COLSS (Core Operating Limit Supervisory System) of ABB-CE. Considered fuel design limits are DNBR and LDCA Fq. 2003-D Power distributions within the present CAOC (Constant Axial Offset Control) limits are calculated for the analysis. The most limiting DNB prevention event of CEA Withdrawal is analyzed with the ROPM (Required OverPower Margin) concept of ABB-CE. The result show that the DMS can bring around 7% more margins for both DNB and LOCA Fq standpoints of view. The DMS can also monitor the PCI (Pellet-Cladding Interaction) limits.

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A Study on the Design Characteristics of Steel Frame in Modern Architecture (근대건축과 철구조의 디자인특성에 관한 연구)

  • 이정욱
    • Korean Institute of Interior Design Journal
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    • no.6
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    • pp.53-62
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    • 1995
  • This research aims at proving the fact that the forms, spaces and many other design concepts of Modernism are much related with the changes of materials, struc-tures, and the way of construction as well as the idealis-tic and aesthestic things through the history of steel, one of the most important materials of the style. The meaning steel has in the modern architecture can be studied in the structure and industrial production system. 1) Steel frame broadened the range of understanding the space and created the new form through the skeleton/skin structure by reinterpreting the existing space fac-tors while it was being adopted to the architecture. Walls could be freed from the traditional function of bear-ing wall and roofs gave the transparancy to the interior by being linked with the glass. Posts lost the function which confines the space in the frame of the grid system and gave the flexibility to the interior due to the economical materials. These changes made the movable partition, screen with various materials and the system furniture which divides the space more important. 2) In the aspect of the industrialized architecture, it be-came the moment that the most of the archtectural com-posing parts were in mass production as they were standarized, high qualified, and generalized by the indus-trial characteristics of steel, and the specialization of structure and cladding, but the neither of the efforts to make the building itself by mass production or to standarize it was fulfilled. The high-tech architecture which borrows its archtectural manifestation from the high technology, however, is consistently paying efforts on such industrialization.

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Impact of Multi-dimensional Core Thermal-hydraulics on Inherent Safety of Sodium-Cooled Fast Reactor (다차원 노심열수력 현상이 소듐고속로 고유안전성에 미치는 영향)

  • Kwon, Young-Min;Jeong, Hae-Yong;Ha, Kwi-Seok
    • Proceedings of the KSME Conference
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    • 2008.11b
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    • pp.3175-3180
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    • 2008
  • A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

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A System Engineering Approach to Predict the Critical Heat Flux Using Artificial Neural Network (ANN)

  • Wazif, Muhammad;Diab, Aya
    • Journal of the Korean Society of Systems Engineering
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    • v.16 no.2
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    • pp.38-46
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    • 2020
  • The accurate measurement of critical heat flux (CHF) in flow boiling is important for the safety requirement of the nuclear power plant to prevent sharp degradation of the convective heat transfer between the surface of the fuel rod cladding and the reactor coolant. In this paper, a System Engineering approach is used to develop a model that predicts the CHF using machine learning. The model is built using artificial neural network (ANN). The model is then trained, tested and validated using pre-existing database for different flow conditions. The Talos library is used to tune the model by optimizing the hyper parameters and selecting the best network architecture. Once developed, the ANN model can predict the CHF based solely on a set of input parameters (pressure, mass flux, quality and hydraulic diameter) without resorting to any physics-based model. It is intended to use the developed model to predict the DNBR under a large break loss of coolant accident (LBLOCA) in APR1400. The System Engineering approach proved very helpful in facilitating the planning and management of the current work both efficiently and effectively.

Experimental and theoretical justification of passive heat removal system for irradiated fuel assemblies of the nuclear research reactor in a spent fuel pool

  • Ta Van Thuong;O.L. Tashlykov;S.M. Glukhov;D.E. Shumkov;Yu.V. Volchikhina
    • Nuclear Engineering and Technology
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    • v.55 no.6
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    • pp.2088-2095
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    • 2023
  • The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

Analysis of LOFT LP-02-6 Experiment Using RELAP5/MOD3.2

  • Park, Tong-Soo;Lee, Jae-Hoon;Park, Byung-Suh;Cho, Chang-Sok
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.11a
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    • pp.357-362
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    • 1996
  • LOFT LBLOCA test, LP-02-6 was analyzed using RELAP5/MOD3.2. It has a distinguished thermal-hydraulic phenomenon of a positive bottom-up core flow in tile blowdown phase. A modified nodalization which is based on that used in LP-LB-1 calculation by Lubbesmeyer was used in the calculation. RELAP5/MOD3.2 predicted overall system hydraulic behavior relatively well. However, the bottom-up quenching in the middle part of the core was not predicted sufficiently. It was demonstrated also that the peak cladding temperature can be predicted well by adjusting a discharge coefficient. But more improvements are needed in order to apply this code to actual plants with less user dependency.

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