• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.539-552
• /
• 2013

To establish an aging management plan considering seawater influx and changes in groundwater within nuclear power plant sites, the characteristics of groundwater flow must be understood. This study investigated the characteristics of groundwater flow within the site and analyzed groundwater level recorded by monitoring wells to evaluate groundwater flow characteristics and elements that affected these characteristics for supplying the information to conduct the appropriate aging management for ensuring the safety of the safety-related structures in Shin Kori Unit 1 and 2. The increase in groundwater level during the wet season results from high sea-level conditions and the large amount of precipitation. As a result of the analysis of groundwater distribution and change characteristics, the site could be divided into a rainfall-affected area and a tide-affected area. First, the rainfall-affected area can further be divided into areas that are affected simultaneously by excavation, backfill, and a permanent dewatering system. Secondly, areas that are not affected by excavation, or the dewatering system, or by structure arrangement and excavation. Analysis of the spectrum for wells affected by tides resulted in confirmation of the M2 component (12.421 hr) and S2 component (12.000 hr) of the semidiurnal tides, and the O1 component (25.819 hr) of the diurnal tides. In the cross-correlation results regarding tides and groundwater levels, the lag time occurred diversely within 1-3 hours by the effect of the well location from sea, the distribution of the backfill material with depth, and the concrete structure.

• Journal of Radiation Protection and Research
• /
• v.42 no.4
• /
• pp.222-228
• /
• 2017

Background: On June 18, 2017, Korea's first commercial nuclear reactor, the Kori Nuclear Power Plant No. 1, was permanently suspended, and the capacity of nuclear power generation facilities will be adjusted according to the governments denuclearization policy. In these circumstances, it is necessary to assess the quality of radiation safety management in nuclear power plants in Korea by evaluating the radiation dose associated with them. Materials and Methods: The average annual radiation dose per unit, the annual radiation dose per person, and the annual dose distribution were analyzed using the radiation dose database of nuclear reactors for the last 5 years. The results of our analysis were compared to the specifications of the Nuclear Safety Act and Medical Law in Korea. Results and Discussion: The annual average per unit radiation dose of global major nuclear power generation was 720 man-mSv, while that of Korea's nuclear power plants was 374 manmSv. No workers exceeded 50 mSv per year or 100 mSv in 5 years. The individual radiation dose according to occupational exposure was 0.59 mSv for nuclear workers, 1.77 mSv for non-destructive workers, and 0.8 mSv for diagnostic radiologists. Conclusion: The radiation safety management of nuclear power plants in Korea has achieved the best outcomes worldwide, which is considered to be the result of the as-low-as-reasonably-achievable (ALARA) approach and strict radiation safety management. Moreover, the occupational exposures were also very low.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.11 no.1
• /
• pp.52-60
• /
• 1991

The importance and role of preservice and inservice inspection(PSI/ISI) for nuclear power plant components are intimately related to plant design, safety, reliability and operation etc.. The Korea Atomic Energy Research Institute(KAERI) has been performing PSI/ISI in Korea since the PSI of Kori nuclear power plant, unit 1 had been performed in 1977. KAERI has localized PSI/ISI technology and has done much experience in ultrasonic flaw detection, evaluation and its excavation on the weldments of large pressure vessels. The results of flaw estimation using ultrasonic examination are compared with the actual flaw sizes revealed by field excavation. KAERI's experience regarding PSI/ISI was described and some discussions were added.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
• /
• v.24 no.4
• /
• pp.116-122
• /
• 2010

This paper describes the up-grade of the turbine governor for steam turbine due to its poor operation from long time use. The analog type governor of the unit 1 in Kori nuclear power plant in Korea was removed and the new digital type turbine governor was developed and installed. The procedure for the actual application, site adaptability test using dynamic simulator and the result of actual operation are described here. And the program for nuclear steam turbine is suggested here.

• Nuclear Engineering and Technology
• /
• v.14 no.4
• /
• pp.186-195
• /
• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.

• Journal of Radiation Protection and Research
• /
• v.23 no.3
• /
• pp.175-184
• /
• 1998

In order to calculate dose rates from steam generators to be replaced from Kori unit 1 in 1998, radionuclide inventories inside steam generator were evaluated from smear test results and measured dose rates from S/G tubes withdrawn for the metallographical examination of damaged tubes. Based on the inventories, contact dose rates and dose rates at 1 m from the surface of a steam generator were calculated using the QAD-CG computer code. Contact dose rates ranged from 11.5 mR/hr at the bottom of channel head to 37.7 mR/hr at the middle of shell barrel, and showed no significant difference with dose rates at 1 m from the surface of steam generator. Shielding effects of lead and carbon steel were compared to provide basic shielding data. Lead shield showed excellent shielding effects. Dose rate at 1 m from the middle of S/G shell barrel decreased from 38.6 mR/hr to 15.5 mR/hr with the lead shield of 2 mm thickness. However, carbon steel showed a poor shielding effect even with the thickness of 2.0 cm. This can be explained with the great differences in the attenuation effect and buildup factor between lead and carbon steel for low energy photons.

• Journal of Radiation Protection and Research
• /
• v.28 no.2
• /
• pp.109-116
• /
• 2003

Based on the necessity to evaluate the activation products inventory during decommissioning lot domestic nuclear power plants, a preliminary estimation of the activation products inventory for Kori unit 1, which is getting close to the end of lifetime, was carried out with ANISN and ORIGEN2 code. In order to calculate neutron nux using ANISN code, the reactor was divided into 9 zones from core to bioshield concrete for radial direction. Also :he cross-section of main nuclides were calibrated with neutron flux in the reactor pressure vessel(RPV) region. The results showed that 95 % of tile total radioactivity in RPV from reactor shutdown to 10 years came from the nuclides of $^{55}Fe,\;^{59}Ni,\;^{63}Ni\;and\;^{60}Co$. And the total radioactivity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the total radioactivity of RPV reached to $5.25{\times}10^{6}GBq$ at shutdown time. As compared with the total radioactivity of bioshield concrete at reactor shutdown time, the radioactivity after tooling more than 10 years was below 1 %.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.2
• /
• pp.213-226
• /
• 2019

Regulatory agency and licensee are preparing for the site restoration of Kori unit 1, the first commercial NPP in Korea, scheduled for 2031. Developing regulatory guidelines and strategies is essential for effective restoration work. Unfortunately, Korea does not have experience of site restoration of commercial NPPs. Therefore, it is important to review cases from experienced countries to establish a strategy and regulatory standards. The U.S. has had numerous soil remediation experiences using RESRAD and MARSSIM. However, formalized evaluation methodologies for subsurface soil have not yet been established in MARSSIM. This survey focused on subsurface soil remediation by reviewing the five decommissioned NPPs under regulation of the US NRC. Overall process of remediating a contaminated subsurface soil and groundwater was reviewed to identify considerations and lessons that could be applicable in Korea. In addition, an applied methodology for evaluation of contaminated subsurface soil and related major issues between regulatory agency and licensees were reviewed in detail to support establishment of remediation strategy for Kori unit 1.

• 전기의세계
• /
• v.24 no.2
• /
• pp.84-91
• /
• 1975

This paper describes the results of a study for the system characteristics, especislly for the abnormal frequency drop of power system, when a large generation unit such as Kori Nuclear 1 (595MW) pulls out from the system. The automatic load shedding method now adopted in our system was re-studied to ameliorate the above problem. From the results of the study, a new under-frequency relay with an element for detecting the slope of frequency change and with time delay element to raise the lowered frequency to a desired value, was found to be effective, and should be purchased and utilized. By this study, an optimal and concrete load shedding method was recommended for reliable operation of power system.

• Proceedings of the Korea Society for Simulation Conference
• /
• 2004.05a
• /
• pp.94-100
• /
• 2004

KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulic simulation program (called ARTS-KORIl) based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 nuclear power plant simulator. To develop the RETRAN code as an NSSS T/H engine for the simulator, a number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made to satisfy the simulator requirements of robustness and real time calculation capability Some simplified models and a backup system were also developed to simulate some transients that cannot be efficiently calculated by the RETRAN part of ARTS-KORIl.