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THREE-DIMENSIONAL CORE DESIGN OF A SUPER FAST REACTOR WITH A HIGH POWER DENSITY

  • Cao, Liangzhi;Oka, Yoshiaki;Ishiwatari, Yuki;Ikejiri, Satoshi;Ju, Haitao
    • Nuclear Engineering and Technology
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    • v.42 no.1
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    • pp.47-54
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    • 2010
  • The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm3. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

Performance Analysis on Early Detection of Fault Symptom of a Pump with Abnormal Signals (오신호 입력에 따른 펌프의 고장징후 조기감지 성능분석)

  • Jung, Jae-Young;Lee, Byoung-Oh;Kim, Hyoung-Kyun;Kim, Dae-Woong
    • Journal of Power System Engineering
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    • v.20 no.2
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    • pp.66-72
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    • 2016
  • As a method to improve the equipment reliability, early warning researches that can be detected fault symptom of an equipment at an early stage are being performed out among developed countries. In this paper, when abnormal signal is input to actual normal signal of a pump, early detection studies on pump's fault symptom were carried out with auto-associative kernel regression as an advanced pattern recognition algorithm. From analysis, correlations among power of motor driving pump, discharge flow of pump, power output of pump, and discharge pressure of pump are exited. When the abnormal signal is input to one of those normal signals, the other expected values are changed due to the influence of the abnormal signal. Therefore, the fault symptom of pump through the early-warning index is able to detect at an early stage.

Application and optimal design of the bionic guide vane to improve the safety serve performances of the reactor coolant pump

  • Liu, Haoran;Wang, Xiaofang;Lu, Yeming;Yan, Yongqi;Zhao, Wei;Wu, Xiaocui;Zhang, Zhigang
    • Nuclear Engineering and Technology
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    • v.54 no.7
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    • pp.2491-2509
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    • 2022
  • As an important device in the nuclear island, the nuclear coolant pump can continuously provide power for medium circulation. The vane is one of the stationary parts in the nuclear coolant pump, which is installed between the impeller and the casing. The shape of the vane plays a significant role in the pump's overall performance and stability which are the important indicators during the safety serve process. Hence, the bionic concept is firstly applied into the design process of the vane to improve the performance of the nuclear coolant pump. Taking the scaled high-performance hydraulic model (on a scale of 1:2.5) of the coolant pump as the reference, a united bionic design approach is proposed for the unique structure of the guide vane of the nuclear coolant pump. Then, a new optimization design platform is established to output the optimal bionic vane. Finally, the comparative results and the corresponding mechanism are analyzed. The conclusions can be gotten as: (1) four parameters are introduced to configure the shape of the bionic blade, the significance of each parameter is herein demonstrated; (2) the optimal bionic vane is successfully obtained by the optimization design platform, the efficiency performance and the head performance of which can be improved by 1.6% and 1.27% respectively; (3) when compared to the original vane, the optimized bionic vane can improve the inner flow characteristics, namely, it can reduce the flow loss and decrease the pressure pulsation amplitude; (4) through the mechanism analysis, it can be found out that the bionic structure can induce the spanwise velocity and the vortices, which can reduce drag and suppress the boundary layer separation.

EXTENSION OF OPERATIONAL LIFE-TIME OF WWER-440/213 TYPE UNITS AT PAKS NUCLEAR POWER PLANT

  • Katona, Tamas Janos;Ratkai, Sandor
    • Nuclear Engineering and Technology
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    • v.40 no.4
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    • pp.269-276
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    • 2008
  • Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.

Low cycle fatigue properties of hydrogenated welding sheets of Zr-Sn-Nb alloy using funnel-shaped flat specimens

  • Lian-feng, Wei;Chen, Bao;Shi-zhong, Wang;Yong, Zheng;Meng-bin, Zhou
    • Nuclear Engineering and Technology
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    • v.52 no.8
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    • pp.1724-1731
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    • 2020
  • Low cycle fatigue tests on the hydrogenated welding seam of Zr-Sn-Nb alloy at room temperature and 360 ℃ had been carried out by using the funnel-shaped flat specimens. The relationships between nominal stress & strain directly measured across the funnel and local stress & strain at the root of the funnel are given by considering cyclic plasticity correction. The results show that the fatigue resistance of welding seam at room temperature is only slightly better than that at 360 ℃. Probabilistic fatigue life curves are obtained by using a two-parameter power function.

Abnormal Sound from Heat Exchanger of Condensate Water System at Nuclear Power Plant (원전 복수계통 열교환기의 이음 원인 분석)

  • Lee, Jun-Shin;Lee, Wook-Ryun;Kim, Tae-Ryong
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.26 no.4
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    • pp.469-474
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    • 2016
  • Abnormal sound was heard from a heat exchanger of condensate water system in a nuclear power plant, which was identified as impact sound of a loose part later. Nuclear power plants are normally equipped with loose part monitoring system for primary water system, but not for secondary water system. The abnormal sound was analyzed by using the impact signal-processing methodology based on the Hertz theory. The predicted results for impact location and size of the loose part showed good agreement with those of the actual loose part found during the overhaul period in the plant. So, this analysis methodology for the impact signal will be widely utilized for the primary and secondary side of the nuclear power plant.

A Quantitative Assessment of Organizational Factors Affecting Safety Using System Dynamics Model

  • Yu Jaekook;Ahn Namsung;Jae Moosung
    • Nuclear Engineering and Technology
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    • v.36 no.1
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    • pp.64-72
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    • 2004
  • The purpose of this study is to develop a system dynamics model for the assessment of the organizational and human factors in a nuclear power plant which contribute to nuclear safety. Previous studies can be classified into two major approaches. One is the engineering approach using tools such as ergonomics and Probability Safety Assessment (PSA). The other is the socio-psychology approach. Both have contributed to find organizational and human factors and to present guidelines to lessen human error in plants. However, since these approaches assume that the relationship among factors is independent they do not explain the interactions among the factors or variables in Nuclear Power Plants. To overcome these restrictions, a system dynamics model, which can show cause and effect relationships among factors and quantify the organizational and human factors, has been developed. Handling variables such as the degree of leadership, the number of employees, and workload in each department, users can simulate various situations in nuclear power plant organization. Through simulation, users can get insights to improve safety in plants and to find managerial tools in both organizational and human factors.

Critical Success Factors for Nuclear Power Plant Construction Projects (원전건설의 성공적인 시공을 위한 핵심성공요인(CSFs) 도출)

  • Shin, Gisung;Bang, Seongdeok;Kim, Kyeongseok;Kim, Hyoungkwan
    • Korean Journal of Construction Engineering and Management
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    • v.18 no.1
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    • pp.48-57
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    • 2017
  • The construction of the first nuclear power plant (Kori #1) in Korea started in 1971, Korea operates 24 nuclear power plants and is constructing 4 in 2015. During about 45 years of construction experience, insufficient studies have been investigated for the identification of critical success factors(CSFs) for nuclear power plant construction. Based on literature reviews and focus group interviews, this study presented a list of CSFs for construction of nuclear power plants. A survey for validating the results of CSFs was conducted with 164 experts. This study indicates that attentions should be placed upon the lowest price-based contract awarding policy, the need for reasonable pricing standard implementation, database development, and deployment for experienced nuclear power plant construction workers, identification and nurturing of competitive subcontractors, and minimization of lag times in construction activities.

Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

  • Alnaqbi, Jwaher;Hartanto, Donny;Alnuaimi, Reem;Imron, Muhammad;Gillette, Victor
    • Nuclear Engineering and Technology
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    • v.54 no.2
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    • pp.764-769
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    • 2022
  • The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.