• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.108-113
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• 2015

Plastic injection moulding machine is widely used for many industrial field. It is classified into mandatory safety certification machinery in Industrial Safety and Health Act because of its high hazard. In order to prevent industrial accidents by plastic injection moulding machine, it is necessary for designer to identify hazardous factors and assess the failure modes to mitigate them. This study tabulates the failure modes of main parts of plastic injection moulding machine and how their failure has affect on the machine being considered. Failure Mode & Effect Analysis(FMEA) method has been used to assess the hazard on plastic injection moulding machine. Risk and risk priority number(RPN) has been calculated in order to estimate the hazard of failures using severity, probability and detection. Accidents caused by plastic injection moulding machine is compared with the RPN which was estimated by main regions such as injection unit, clamping unit, hydraulic and system units to find out the most dangerous region. As the results, the order of RPN is injection unit, clamping unit, hydraulic unit and system units. Barrel is the most dangerous part in the plastic injection moulding machine.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• The KSFM Journal of Fluid Machinery
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• v.19 no.5
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• pp.50-60
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• 2016

A design methodology of the modeled test facility to conserve an injection performance of a passive safety injection system is proposed. This safety injection system is composed of a core makeup tank and a safety injection tank. Individual tanks are connected with pressure balance line on the top side and injection line on the bottom side. It is important to conserve the scaled initial injection flow rate and total injection time since this system can be operated by small gravity head without any active pumps. Differential pressure distribution of the injection line induced by the gravity head is determined by the vertical length and elevation of each tank. However, the total injection time is adjustable by the flow resistance coefficient of the injection line. The scaling methodology for the tank and flow resistance coefficient is suggested. A key point of this test facility design is a scaling analysis for the flow resistance coefficient. The scaling analysis proposed on this paper is based on the volume scaling law with the same vertical length to the prototype and can be extended to a model with a reduced vertical length. A set of passive injection test were performed for the tanks with the same volume and the different length. The test results on the initial flow rate and total injection time showed the almost same injection characteristics and they were in good agreement with the design values.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.41-58
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• 2016

Reliability Centered Maintenance (RCM) introduces a systematic method and decision logic tree for utilizing previous operating experience focused on reliability and optimization of maintenance activities. In this paper RCM methodology is applied on safety injection system for APR-1400. Functional Failure Mode Effects and Criticality Analysis (FME&CA) are applied to evaluate the failure modes and the effect on the component, system and plant. Logic Tree Analysis (LTA) is used to determine the optimum maintenance tasks. The results show that increasing the condition based maintenance will reduce component failure and improve reliability and availability of the system. Also the extension of the surveillance test interval of Safety Injection Pumps (SIPs) would lead to an improved pump's availability, eliminate the unnecessary maintenance tasks and this will optimize maintenance activities.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.54-70
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• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.395-400
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• 1996

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are peformed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSI pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SIT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA.

• Proceedings of the Korean Information Science Society Conference
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• 1998.10b
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• pp.487-489
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• 1998

원자력 발전소와 같이 환경적으로 큰 재난을 가져 올 수 있는 시스템에서는 전체 시스템을 구축하기 전에 구축하고자 하는 시스템의 안전성을 보장할 수 있는지의 여부와 그러한 시스템의 조작자들의 훈련을 위해 실시간 시뮬레이션이 반드시 필요하다. 본 논문에서는 원자력 발전소의 SIS(Safety Injection System)를 실시간 객체 TMO(Time-triggered Message Triggered Model)를 이용 모델링하는 기법과, 분산 실시간 객체 플랫폼인 WTMOS위에서 구현된 SIS 시뮬레이션 시스템에 대해 기술하였다.