• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.695-700
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• 1997

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of .each strategy are also investigated.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.29 no.1 s.232
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• pp.46-54
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• 2005

Film cooling characteristics has been examined numerically for the height variation of a stepped slot exit. In this study, the upstream wall height of the stepped slot exit varies from -2d (d = slot width) to 3d, blowing ratio ranges from 0.5 to 3, and injection angles are $15^{\circ},\;30^{\circ},\;and\;45^{\circ}$. The results showed that film cooling performance was mainly subjected to the magnitude of recirculation region near the downstream-side slot exit as well as the magnitude and the distribution region of turbulent kinetic energy due to the local velocity and momentum differences between the coolant and the main flow near the slot exit. The up-1d type slot at higher blowing ratios over 2 and the flat type slot at lower blowing ratios below 1 have the best film cooling performances, in case of the injection angles of $30^{\circ},\;and\;45^{\circ}$, respectively. Compared with the other injection angles, in case of the injection angles of $15^{\circ}$, the best film cooling performances was shown in even a higher upstream wall (up-3d) at higher blowing ratio like 3 by the gradual reduction of the coolant velocity which minimizes the local velocity differences between the coolant and the main flow near the slot exit.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.16 no.5
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• pp.105-112
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• 2017

In this study, plastics were deformed after molding due to the characteristics of the material. The Taguchi experimental design method was utilized to find the molding conditions that minimized deformation of the plastic bezel to be assembled in an automotive steering wheel. The injection molding conditions applied to the experimental design method are the melt temperature, cavity plate coolant temperature, core plate coolant temperature, and cooling time. Each condition was divided into five levels, and a total of 25 experiments were planned. However, instead of performing 25 actual molding experiments, the injection molding analysis was performed using the Moldflow program, and the deformation values for each molding analysis were obtained. The optimal molding conditions were obtained from these deformation values. The actual injection molding experiment using optimal molding conditions was compared with the deformation amount of the current molded product. The deformation was measured using a precise 3D scanner. The deformation amount of the molded product under optimal molding conditions was 16.1% lower than the deformation amount of the current molded product.

• Journal of Energy Engineering
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• v.26 no.3
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• pp.71-77
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• 2017

In the event of loss of coolant accident (LOCA) and station black out (SBO) in the primary system of a nuclear reactor, the coolant water should be injected to reactor coolant system (RCS) without any intervention of operators or active components. To satisfy the requirements, hybrid safety injection tank (Hybrid SIT) was suggested by Korea Atomic Energy Research Institute (KAERI). The pressure equalizing time of Hybrid SIT is an important parameter to determine the timing of coolant injection. To predict the pressure equalizing time of the Hybrid SIT, a separate effect test facility was constructed and sensitivity tests were conducted in various conditions. The main parameter determining the pressure equalizing time was obtained from separate effect test (SET) results. The wall of condensation on the inner wall of SIT and direct contact condensation on the water surface affected to the pressure equalizing time very much. In this study, the effect of each condensation phenomena on pressure equalizing time was quantitatively analyzed from results of SET and a prediction method of pressure equalizing time was proposed.

• 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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• 2002.10a
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• pp.353.2-353
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• 2002

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.1833-1838
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• 2003

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

• Journal of the Korea Institute of Military Science and Technology
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• v.9 no.1 s.24
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• pp.117-124
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• 2006

Experiments on film cooling were performed with a small scale rocket engine homing liquid oxygen (LOx) and Jet A-1(jet engine fuel). Film coolants(Jet A-1 and water) were injected through the film cooling injector. Film cooled length and the outside wall temperature of the combustor were determined for chamber pressure, and the different geometries(injection angle) with the flow rates of film coolant. The loss of characteristic velocity due to film cooling was determined for the case of film cooling with water and Jet A-1. As the coolant flow increases, the outside wall temperatures decrease but the decrease in the outside wall temperatures reduced over the 8 percent film coolant flow rate. The efficiency of characteristic velocity was decreased with the Increase of the film coolant flow rate.

• Proceedings of the KSME Conference
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• 2000.11a
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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.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.