• Journal of the Korean Society for Nondestructive Testing
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• 제26권6호
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• pp.440-446
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• 2006

The reactor vessel body and closure head are fastened with the stud bolt that is one of crucial parts for safety of the reactor vessels in nuclear power plants. It is reported that the stud bolt is often experienced by fatigue cracks initiated at threads. Stud bolts are inspected by the ultrasonic technique during the overhaul periodically for the prevention of failure which leads to radioactive leakage from the nuclear reactor. The conventional ultrasonic inspection for stud bolts was mainly conducted by reflected echo method based on shadow effect. However, in this technique, there were numerous spurious signals reflected from every oblique surfaces of the thread. In this study, ultrasonic phased array technique was applied to investigate detectability of flaws in stud bolts and characteristics of ultrasonic images corresponding to different scanning methods, that is, sector and linear scan. For this purpose, simplified stud bolt specimens with artificial defects of various depths were prepared.

• IE interfaces
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• 제25권4호
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• pp.382-392
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• 2012

Reactor protection system to keep nuclear safety and operational economy of plants requires high reliability. Such a high reliability of the system can be achieved through the redundant design of components. However, common cause failures of components reduce the benefits of redundant design. Thus, the common cause failure analysis, to accurately calculate the reliability of the reactor protection system, is carried out using alpha-factor model. Analysis results to 24 operating months are that 1) the system reliability satisfies the reliability goal of EPRI-URD and 2) the common cause failure contributes 90% of the system unreliability. The uncertainty analysis using alpha factor parameters of 0.05 and 0.95 quantile values shows significantly large difference in the system unreliability.

• Proceedings of the KIEE Conference
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• 대한전기학회 2001년도 하계학술대회 논문집 B
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• pp.717-719
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• 2001

This paper deals with simulation of the three-phase dc reactor type fault current limiter(FCL). This is a preliminary step to develop the FCL's faculties for an application to high voltage transmission line. A three-phase dc reactor type FCL consists of transformers, diodes, and a superconducting coil. By this simulation for the short-circuit test we can investigate the safety of FCL's elements. And, result of simulation will contribute parameter toward optimal design.

• Proceedings of the KSME Conference
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• 대한기계학회 2002년도 학술대회지
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• pp.47-50
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• 2002

The numerical simulation of a 271-rod fuel assembly of nuclear Liquid-Metal Fast Reactor (LMFR) with an infernal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it potentially has very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within a assembly with practically no change in mixed mean temperature at the assembly exit.

• Proceedings of the KSME Conference
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.432-437
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• 2000

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The the radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively.

• Transactions of the Korean Society of Mechanical Engineers B
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• 제30권9호
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• pp.858-865
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• 2006

SMART Pilot is a multipurpose small capacity integral type reactor. Main coolant pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of $310^{\circ}C$ and 14.7MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present wort a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and lift-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP.

• Proceedings of the KIEE Conference
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• 대한전기학회 1997년도 하계학술대회 논문집 B
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• pp.600-602
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• 1997

The kalman filter which has good estimating capabilities by means of the recursive computation from the previously known of obtained data is usually used for the system estimation in the case of not being directly measurable. The best estimating technique is still open issues on the PWR reactor control system to increase operating contingencies and to predict the safety margins for safer reactor operation. This paper addressed its estimating technique using kalman filter for the more flexible reactor control and showed the reasonable approach for discretization of the continuos-time system for reduction of computation errors.

• Proceedings of the Korean Nuclear Society Conference
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.66-66
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• 2002

Under Atomic Energy Society of Japan (AESJ) and Japan Atomic Energy Research Institute (JAERO, a Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) has been set since March 1999 to review and investigate special subjects related to reactor physics research of Accelerator-Driven System (ADS). In the ADSWP, the extensive and aggressive activity is being made by 25 professional members in the field of reactor physics in Japan. The ADS is now studying three subjects related to subcriticality of ADS; (1) calculation accuracy of sub criticality on ADS, (2) critical safety issues of ADS, and (3) theoretical review of subcriticality and its measurement methods. This paper describes two topics related to the subjects (1) and (2); one is an analysis of maximum reactivity potentially inserted to a subcritical core and the other is a benchmark proposal for checking calculation accuracy of sub criticality on ADS. The full specification of the calculation benchmark will be supplied by June 2002. Researchers from overseas, especially from Korea, are welcome to join this benchmark

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.535-542
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• 2012

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

• 전기의세계
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• 제15권5호
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.