• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.250.2-250
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• 2001

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.219-228
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• 1983

The blowdown history of the TMI-2 accident up to the isolation of the relief valve associated with a small break LOCA is reviewed briefly. An analysis is made to determine what instruments should be added in the core in order to prevent core damage in the case of the TMI-2 accident. With the added instruments a procedure is presented on how to predict the uncovered level of the core and how to calculate operator time margin. Sample calculations are done for the TMI-2 accident to determine the uncovered level and operator time margin. Finally, the map to show the uncovered level of the core and operator time margin is drawn with measurable parameters by the above methods.

• Journal of Radiation Protection and Research
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• v.20 no.1
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• pp.53-64
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• 1995

The radiation fields following the large loss of coolant accident (LOCA) have been assessed for the vital areas in the service building of Wolsong 2, 3, and 4 nuclear power plants. The ORIGEN2 code was used in calculating the fission product inventories in the fuel. The source terms were based upon the activity released following the dual failure accident scenario, i.e., a LOCA followed by impaired emergency core cooling (ECC). Configurations of the reactor building, the service building, and the ECC system were constructed for the QAD-CG calculations. The dose rates and the time-integrated doses were calculated for the time period of upto 90 days after the accident. The results showed that the radiation fields in the vital access areas were found to be sufficiently low. Some areas however showed relatively high radiation fields that may require limited access.

• Journal of the Korea Safety Management & Science
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• v.7 no.4
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• pp.85-99
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• 2005

The main objective of this paper is to search whether containment vessel's best pressure may increase until how long when loss of coolant accident (LOCA) happened in containment vessel of Ulchin nuclear power plant 1 and 2. Another goal of this research is to find the influential factors that increase containment vessel pressure. Model for this research is Ulchin nuclear power plant 1 with 10 cycles. Data were collected by simulator of Ulchin nuclear power plant 1 and design of experiment was used for data analysis. For the experiment, seven factors that are going to influence in containment vessel pressure were chosen. It was found that fatter which influences in early rise of containment vessel pressure after LOCA is only explosion size. Also, containment vessel's best pressure (3.74 bar.a) was much lower than limit (4.86 bar.a) of FSAR (Final Safety Analysis Report).

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.141-154
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• 1984

An analysis is presented of key phenomena and scenario which imply some general trends for beyond design-basis-accident in Kori-1 PWR dry containment. The study covers a wide range of severe accident sequences initiated by small break LOCA. The MARCH computer code, with KAERI modifications was used in this analysis. The major emphasis of the paper are two folds, 1) the phenomenologic understanding of severe accident and 2) a study of H2 combustion and debris/ water interactions in a specific small break LOCA for Kori-1 plant. The sensitivity studies for the specific plant data and thermal interaction modelings used in the SASA were performed. The results show that if hydrogen burning does occur at low concentration, the resulting peak pressure does not exceed the design value, while the lower concentration assumption results in repeated burning due to the continuing H$_2$ generation. For debris/water interaction, the particle size has no effect on the magnitude of peak pressure for the amount of water assumed to be in the reactor cavity. But, the occurrence of peak pressure is considerably delayed in case of using the dryout correlation. The peak containment pressure predicted from the hydrogen combustion and steam pressure spite during full core meltdown scenario does not present a severe threat to the containment integrity.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.218-227
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• 1986

Studies were conducted to determine the extent of oxidation and same of the mechanical property changes of Zircaloy-4 fuel cladding after it was exposed to hot steam environment. The purpose of these tests was to provide some informations on the embrittlement behavior of CANDU type fuel cladding, which could be experienced under the loss-of-coolant accident conditions. The Zircaloy fuel cladding tubes were exposed in a steam environment at the temperature of 90$0^{\circ}C$, 1,00$0^{\circ}C$. The growth of the ZrO$_2$ layer combined with an oxygen rich $\alpha$-phase layer into the Zircaloy tube material was found as a function of time t and temperature of steam exposure, E=1.1√Dt+0.002 where D is a temperature dependent diffusion coefficient. The tensile strength of the specimens exposed for a short period increased but decreased continuously with further exposure. The circumferential elongation was drastically changed with the exposure time while the hoop strength did't decrease greatly. The X-ray measurement of preferred orientation of the Zircaloy tube material indicated that grains in the as received tube were oriented such that the poles of the basal (0001) planes were predominantly radial, while the poles of the basal plane in the tube materials heattreated at 1,00$0^{\circ}C$ were oriented tangentially. It appears that this reoriented texture may contribute to lessening the decrease of the hoop strength of the heat treated Zircaloy tube material.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.259-266
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• 1989

An improved version of RELAP5/MOD2 Cycle 36.04 code is assessed for LOFI LBLOCA Test L2-5. Minor modifications to the original version have been done to avoic reflood related errors. Based on the modified version, one base case and two cases for sensitivity study on downcomer and core channel modelling are calculated. The calculation results are compared with the experimental data for primary system pressure, break mass How rate and cladding temperature at hot spot According to the comparison, it is found that the hydraulic system behaviors are well predicted, excessive core cooling exist in blowdown phase for a single core channel and a combined downcomer case, and a better result can be obtained for a two core channel case.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.645-660
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• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.

• Journal of Energy Engineering
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• v.17 no.2
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• pp.67-76
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• 2008

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). Recently several integral effect tests for the reflood period of a LBLOCA (Large Break LOss of Coolant Accident) of the APR1400 have been performed with the ATLAS. In the APR1400 a high flow condition is changed to a low flow condition due to an fluidic device during an operation of the SIT. As the self-controlled fluidic device was not installed in the ATLAS, a set of characterization tests was performed to simulate its injection capability from the SIT for the APR1400 simulation. In the ATLAS the required SIT flow rate in the high flow condition was acquired by installing orifices with an optimized flow area to throttle the SIT discharge line and the low flow condition was achieved by changing the opening of the flow control valve in the SIT injection line. The test results showed that the safety injection systems of the ATLAS could simulate the required high and low flow rates of the SIT for the APR1400 simulation efficiently.

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.3
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• pp.10-21
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• 2014

In Fukushima accident, when the severe accident such as a natural disaster happens, it is impossible to monitor the plant status due to a extreme environment and station blackout and most I&C systems break downs. Finally, these cause the loss of emergency cooling function and thus results in a hydrogen explosion and radiation leak. In this paper, the emergency response system is introduced that monitors and controls properly when the sever accidents like Fukushima accident happen, And the performance requirements of a wireless communication system used in the emergency respons system is described and the performance of emergency communication system is analyzed using the markov model.