• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.4
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• pp.187-195
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• 2013

In this paper, firstly, acceleration-time histories were generated by varying strong motion duration in the frequency domain for application to a seismically isolated nuclear power structure, so as to examine the effects of strong motion duration on the behavior of the structure. Secondly, real recorded earthquakes were modified to match the target response spectrum based on the revised SRP 3.7.1(2007) and the modified time histories were applied to the analysis of a seismically isolated nuclear power structure. The obtained values of acceleration and displacement responses of the structure were, finally, compared with the values obtained in case of applying acceleration-time histories generated in the frequency domain to the structure.

• Corrosion Science and Technology
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• v.10 no.5
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• pp.175-179
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• 2011

The transportation and deposition of iron corrosion products are important elements that affect both the steam generator (SG) integrity and secondary system in pressurized water reactor (PWR) nuclear power plants. Most of iron corrosion products are generated on carbon steel materials due to flow accelerated corrosion (FAC). The several parameters like water chemistry, temperature, hydrodynamic, and steel composition affect FAC. It is well established that the at-temperature pH of the deaerated water system has a first order effect on the FAC rate of carbon steels through nuclear industry researches. In order to reduce transportation and deposition of iron corrosion products, increasing pH(t) tests were applied on secondary system of A, B units. Increasing pH(t) successfully reduced flow accelerated corrosion. The effect of increasing pH(t) to inhibit FAC was identified through the experiment and pH(t) evaluation in this paper.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.141-145
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• 2008

Due to the uncertainties resulting from cost projection, evaluation over long term period, and adequacy of applied discount rate, the economic assessment for back-end fuel cycle is different from each organizations or individuals. In this paper, the features and limitations of some noticeable economic evaluations were investigated and analysed to contribute for the public participation and back-end fuel cycle policy related researches. As a result of analysis, we found that the reprocess and recycling is more economical than direct disposal option, but the result includes high uncertainty that depends on the input parameters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.289-295
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• 2016

Kori Unit 1, the first PWR (Pressurized Water Reactor) plant in Korea, began its commercial operation in 1978 and will permanently shut down on June 18, 2017. After moving the spent fuels to SFP (Spent Fuel Pool) system, Kori Unit 1 will perform a full system chemical decontamination to reduce radiation levels inside the various plant systems. This paper will describe the operation concept of the full system chemical decontamination for Kori Unit 1 based on experiences overseas.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.77-86
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• 2008

This paper describes the nonlinear analyses of a 1:4 scale model of a prestressed concrete containment vessel (PCCV) using an axisymmetric model and a three-dimensional model. These two models are refined by comparison of the analysis results and with testing results. This paper is especially focused on the analysis of behavior under pressure and the temperature effects revealed using an axisymmetric model. The temperature-dependent degradation properties of concrete and steel are considered. Both geometric and material nonlinearities, including thermal effects, are also addressed in the analyses. The Menetrey and Willam (1995) concrete constitutive model with non-associated flow potential is adopted for this study. This study includes the results of the predicted thermal and mechanical behaviors of the PCCV subject to high temperature loading and internal pressure at the same time. To find the effect of high temperature accident conditions on the ultimate capacity of the liner plate, reinforcement, prestressing tendon and concrete, two kinds of analyses are performed: one for pressure only and the other for pressure with temperature. The results from the test on pressurization, analysis for pressure only, and analyses considering pressure with temperatures are compared with one another. The analysis results show that the temperature directly affects the behavior of the liner plate, but has little impact on the ultimate pressure capacity of the PCCV.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.975-983
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• 2020

Since the Fukushima Daiichi accident in 2011, concerns for the safety of multi-unit Nuclear Power Plant (NPP) sites have risen. This is because more than 70% of NPP sites are multi-unit sites that have two or more NPP units and a multi-unit accident occurred for the first time. After this accident, Probability Safety Assessment (PSA) has been considered in many countries as one of the tools to quantitatively assess the safety for multi-unit NPP sites. One of the biggest concerns for a multi-unit accident such as Fukushima is that the consequences (health and economic) will be significantly higher than in the case of a single-unit accident. However, many studies on multi-unit PSA have focused on Level 1 & 2 PSA, and there are many challenges in terms of public acceptance due to various speculations without an engineering background. In this study, two kinds of multi-unit Level 3 PSA for multi-unit site have been carried out. The first case was the estimation of multi-unit risk with conservative assumptions to investigate the margin between multi-unit risk and QHO, and the other was to identify the effect of time delays in releases between NPP units on the same site. Through these two kinds of assessments, we aimed at investigating the level of multi-unit risk and understanding the characteristics of risk in a multiunit context.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.303-309
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• 2019

Tritium is an important nuclide that must be monitored for radiation safety management. In this study, HTO was orally administered to rats at the level of 37 kBq ($1{\mu}Ci$) or 370 kBq ($10{\mu}Ci$) to examine tissue distribution and excretion levels. After sacrifice, wet and dry tissue samples were weighed and analyzed for tissue free-water tritium (TFWT) and organically bound tritium (OBT). The mean tissue concentrations of TFWT (OBT) were 30.9 (17.8) and 4.4 (8.1) Bq/g on days 7 and 13 at the 37 kBq level and 30.8 (64.6) Bq/g on day 17 at the 370 kBq level. To assess the cytogenetic damage due to tritium exposure, a cytokinesis-blocked micronucleus (MN) assay was performed in blood samples from rats exposed to HTO for 14 and 21 days after oral administration. There was no significant difference in the MN frequencies between the control and exposed rats.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3949-3956
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• 2022

Kori Unit 1 is planning a system decontamination project to reduce radiation exposure of decommissioning workers, prevent the spread of contamination and down-grade the level of classification of radioactive waste. The system decontamination range for Kori Unit 1 will be the entire primary system, including RCS, CVCS and RHRS. Some system design modifications are required for the system decontamination operation. In this paper, major system design modifications were evaluated based on the conditions that system restoration is needed after completion of system decontamination. The major system design modifications are CIDF connection location to system, system decontamination operating pressure control, RCP seal water injection and formation of letdown flow. It was evaluated that there was no negative effect on the system due to the system design modifications. However, as the RCP seal water is injected into the system in the oxidation process, the concentration of the oxidizing agent is diluted. Therefore, the oxidizing agent injection and system decontamination operation procedures should be developed to address the dilution effect of the oxidizing agent. The system design modifications dealt in this paper will be finally confirmed through on-site investigation in the future, and if necessary, the system design modifications will be re-evaluated.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.901-910
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• 2021

KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that a fuel bundle can stay in the CANDU reactor without a time limit. However, the regulator requested traditional design values as well as the burnup limit reflecting the computer code uncertainty. The method for the PWR burnup limit was not applied to the CANDU fuel bundle. Since there was no approved methodology to build the burnup limit with uncertainties, KHNP introduced a Monte-Carlo method coupled with a 95/95 approach to determine the conservative burnup limit from the viewpoint of the centerline temperature, internal pressure, strain measurement deviation. Moreover, to consider the uncertainties of various computing models, a converted power uncertainty was introduced. This paper presents the methodology and puts forward the limits on burnup, evaluated for each of the existing and modified fuel bundles, in consideration of the pressure tube aging condition.