• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.229-239
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• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.63-77
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• 1994

The objective of this work is to investigate fluid mixing phenomena related to pressurized thermal shock(PTS) in a pressurized water reactor(PWR) vessel downcomer during transient cooldown with direct vessel injection(DVI) using test models. The test model designs were based on ABB Combustion Engineering(C-E) System 80＋ reactor geometry. A cold leg small break loss-of-coolant accident(LOCA) md a main steam line teak were selected as the potential PTS events for the C-E System 80＋. This work consist of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluid and existing coolant in the downcomer region, and the second part is to compare the results of thermal mixing tests with DVI in the other test model. Row visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small break LOCA Measured transient temperature profiles agree well with the predictions by the REMIX code for a small break LOCA and with the calculations by the COMMIX-1B code for a steam line break event.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.4
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• pp.216-222
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• 2004

Study was focused on the quenching embrittlement property of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment in terms of high temperature oxidation and phase transformation. Property in LOCA condition of advanced cladding that contained Nb element was also investigated. Claddings were oxidized at given temperature and given time followed by water quenching. The results showed that ${\beta}$ phase which formed at quenching stage has an influence on cladding property. In case of advanced cladding, Nb retards cladding oxidation, thus enhances quenching resistance.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2295-2303
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• 2021

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 ㎛. After high temperature oxidation under steam at 1200 ℃ with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.

• Journal of ILASS-Korea
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• v.8 no.1
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• pp.23-28
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• 2003

When a cold HPSI (High Pressure Safety Injection) fluid associated with a design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena may arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated collant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• 한국전산유체공학회:학술대회논문집
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• 1998.05a
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• pp.75-81
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• 1998

A new advanced integral reactor of 330 MWt capacity named SMART(System-integrated Modular Advanced ReacTor) is currently under development at KAERI(Korea Atomic Energy Research Institute). One of the major design features of the integral reactor is locating the steam generators(SG) inside reactor vessel and eliminating the possibility of LB LOCA(large Break Loss of Coolant Accident). Orifices are fitted at the low part of steam generator cassette to stabilize and balance coolant flow distribution in the MCP (Main Circulation Pump) pressure header. A sensitivity analysis is conducted to determine the optimum orifice size using computer code 'CFX'.

• 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.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.3
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• pp.343-351
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• 2011

Nuclear containment building is the last barrier for being secure from any nuclear power plant accident. Therefore, it is very important to understand the ultimate capacity of nuclear containment building to loads associated with severe accidents. LOCA (loss of coolant accident) is considered as the basic accidental load and CANDU-type containment building is considered as a target structure in order to conduct the numerical analysis for the structural safety of a containment building. The CANDU-type containment building is a prestressed concrete shell structure which has the dome and the cylindrical wall and is reinforced with bonded tendons. In this paper, the evaluation of ultimate internal pressure capacity was carried out by nonlinear analysis of a prestressed concrete containment building using 3-dimensional structural analysis system.