• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.41-51
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• 1988

In order to develop a food-chain computer code suitable to the environmental conditions of Korea, the FOOD III code was partially modified. The excution results for Korean male-adult were compared to those from the Canadian version of FOOD III to deduce a more realistic approach in dose assessment. The amounts of Mn-54, Co-58, Co-60, I-131 and I-132 released from Kori unit 1 in1984 were used as the source terms for the sample calculation. The maximum atmospheric dispersion factor(X/Q) value on the site boundary was applied. Through the code modification, organ doses decreased by about $20{\sim}70%$ and the effective committed dose equivalent by about 40% to be $7.935{\times}10^{-6}Sv/y$ which is 0.16% of the ICRP limit, $5{\times}10^{-3}Sv/y$.

• Nuclear Engineering and Technology
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• v.14 no.2
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• pp.86-91
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• 1982

In 1977, spent fuel storage capacity of Ko-Ri Unit 1 was raised to contain 4-2/3 core, by reducing the center-to-center spacing between fuel assemblies from 53.34cm to 36cm. In this paper the adequacy is discussed in detail by examining the previous design analysis report. According to the analytic method presented by Core Performance Branch, study on credible abnormal moderator density condition is performed by using KENO-IV for the redesigned spent fuel storage facility. Result shows that 36cm for the center-to-center spacing between fuel assemblies is not enough to keep the storage safe at water density of 0.1143g/㎤, which gives the maximum $K_{eff}$ 0.9958$\pm$0.0048, which exceeds the CPB regulation limit 0.98. From sensitivity study regarding to the center-to-center spacing, it should be maintained to space greater than 43cm in order to meet the CPB requirements.s.

• Journal of Radiation Protection and Research
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• v.10 no.1
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• pp.50-63
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• 1985

A derivation of new release limit, named Derived Release Limit(DRL), into the atomsphere from a reference nuclear power plant has been performed on the basis of the new system of dose limitation recommended by the ICRP, instead of the (MPC)a limit which has been currently used until now as a general standard for radioactive effluents in Korea. In DRL Calculation, a Concentration Factor Method was applied, in which the concentrations of long-term routinely released radionuclides were in equilibrium with dose in environment under the steady state condition. The analytical model used in the exposure pathway analysis was the one which has been suggested by the USNRC and the exposure limits applied in this analysis were those recommended by the USEPA lately. In the exposure pathway analysis, all of the pathways are not considered and some may be excluded either because they are not applicable or their contribution to the exposure is insignificant compared with other pathways. In case, the environmental model developed in this study was applied to the Kori nuclear power plant as the reference power plant, the highest DRL value was calculated to be as $9.10{\times}10^6Ci/yr$ for Kr-85 in external whole body exposure from the semi-infinite radioactive cloud, while the lowest DRL value was observed 3.64Ci/yr for Co-60 in external whole body exposure from the contaminated ground, by the radioactive particulates. The most critical exposure pathway to an individual in the unrestricted area of interest (Kilchun-Ri, 1.3 km to the north of the release point) seems to be the exposure pathway from the contaminated ground and the most critical radionuclide in all pathways appears to be Co-60 in the same pathway. When comparing the actual release rate from KNU-l in 1982 with the DRL's obtained here the release of radionuclides from KNU-1 were much lower than the DRL's and it could be conclued that the exposure to an individual had been kept below the exposure limits recommended by the USEPA.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.98-103
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• 2009

The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.355-367
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• 2017

The decommissioning of Kori unit 1, which was permanently shut down in June of 2017, will be the first instance of the dismantling of a commercial nuclear power plant in Korea. The disposal of waste during the dismantling process accounts for a large part of the total decommissioning cost. Therefore, structures consisting of activated and contaminated concrete must be economically and safely dismantled by establishing a proper dismantling strategy. This study focuses on optimized dismantling and disposal scenarios pertaining to a biological shield. Several dismantling cases, regulations and technologies related to waste treatment as these practices pertain to nuclear power plants are analyzed. To minimize the amount of waste from the biological shield dismantling process, an optimized dismantling scenario is presented and disposal alternatives for dismantled concrete waste are proposed.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.355-366
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• 1994

The USNRC issued a revised ECCS rule that allows the use of best estimate computer codes for safety analysis. The rule also requires an estimation of uncertainty in calculated system response when applying the best estimate computer codes. A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the ECCS rule has been developed and this paper describes the application of new realistic evaluation methodology to large break LOCA for, the demonstration of the new methodology. The computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/MOD3.1, was used as the best estimate code in the application. The uncertainty of the code was evaluated by assessing several separate and integral effect tests, and for the application to actual plant Kori 3 & 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by random sampling or Monte-Carlo method for each response surface. Final uncertainties were quantified at 95% probability level and safety margins for large break LOCA were discussed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.145-150
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• 2005

To dispose of radwaste in a repository, the safety of disposal must be ensured. This study developed a program for estimating radionuclide accumulation of radwaste, based on the material balance method, one of the indirect methods, and performed application evaluation during the 9th preventive maintenance period of Gori Plant 4, one of the commercial power plants in Korea. First of all, to ensure the technique developed in this study is assessed accurately, this study utilized the data regarding the radionuclide removal in the purification system during the shutdown water chemistry control, and a related estimation technique called SCALP. The target nuclide was Co-60, and it turned out that the relative error was less than $1\%$. The estimation result was compared with the result of direct measurement of the radwaste during the corresponding period as presented by commercial power plants. The result showed that the quantity of Co-60 measured by the direct method was about $50\%$ less than that calculated by the technique developed in this study.

• Journal of the Korean Recycled Construction Resources Institute
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• v.7 no.4
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• pp.316-322
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• 2019

A probabilistic approach considering uncertainties was employed to investigate the effects on the material characteristics and strength of nuclear bio-logical shield concrete walls, when exposed to long-term radiation during the entire service life. Time-dependent compressive and tensile strengths were estimated by conducting the neutron fluence analysis. For the contaminated concrete, individual compressive and tensile failure probabilities can be possibly evaluated by not only establishing limit-state function withthe predefined critical values but also performing Monte Carlo Simulation. Nuclear power plant types similar to the Kori Unit 1, which was shut off permanently in 2017 after the 40-year operation, were herein selected for an illustrative purpose. Consequently, the probability-based performance assessment and prediction of contaminated concrete walls were well demonstrated.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.16 no.9
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• pp.6297-6304
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• 2015

Nuclear power plant decommissioning has attracted attention according to the shutdown decision of Kori 1 which is Korea's first nuclear power plant. Nuclear power plant decommissioning is the one who never experienced ever in our country. So, its process is difficult and time-consuming. In addition, it is difficult to determine the decommissioning quantity. This study proposed the plan that can be used in quantity take-off for nuclear power plant decommissioning using BIM technology being utilized in recent construction industry. As a result, we suggested the method of BIM-based quantity take-off such as the selection decommissioning method and process, setting up of BIM modeling environment, establishment of OBS & WBS, integrated BIM modeling, the definition of quantity property. The proposed plan can be utilized usefully from when permanent stopping nuclear power plant occurs intensively. Furthermore, the overseas nuclear power plant decommissioning project order also are expected through technology securement based on this plan.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.