• 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.

• Tunnel and Underground Space
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• v.29 no.2
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• pp.75-88
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• 2019

With Kori nuclear power plant unit 1 as a beginning in April 1978, 24 nuclear power plants have been operated in Korea and two more plants are under construction. As the nuclear power plants being operated, radioactive wastes from the plants have been accumulated so that various methods for disposing them have been proposed. In Korea, researches have been conducted, being focused on DGD (Deep Geological Disposal), however, DBD (Deep Borehole Disposal) method needs considering as an alternative. In this technical note, element technologies for DBD were analyzed by compiling previous researches and their applicability on domestic cases were investigated. Conceptual studies regarding relevant designs were conducted and finally, technical challenges for actual disposal were described.

• International conference on construction engineering and project management
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• 2017.10a
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• pp.8-13
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• 2017

Nuclear power plant (NPP) constructions are large scale projects that are executed for several years, and schedule control utilizing various schedules is a critically important factor. Recently Korea independently developed the Advanced Power Reactor (APR) 1400 and is building nuclear facilities applying this new reactor type. The construction of Shin-Kori NPP (SKN) Unit 3, which adopted the APR1400, was completed and commercial operation has begun, while, SKN 4, Shin-Hanul NPP (SHN) Units 1&2, and SKN 5&6 are currently under construction. Prior to the development of the APR1400, Korea built 24 reactors and accumulated the schedule data of various reactor types which provided the foundation for schedule reduction to be possible. However, as there is no schedule development and review system established based on the standard schedule data (standard activities, durations, etc.) by reactor type, the process for developing the schedule for new builds is low in efficiency consuming much time and manpower. Also all construction data has been accumulated based on schedule activities. But because the connectivity of activities between projects is low, it is difficult to utilize such accumulated data (causes for schedule delay, causes for design changes, etc.) in new build projects. Due to such reasons, issues continue to arise in the process of developing standard schedule activities and a standard schedule for nuclear power plant construction. In order to develop a standard schedule for NPP construction, i) the development of an NPP standard schedule activity list, ii) development of the connection logic of NPP standard schedule activities, iii) development of NPP standard schedule activity resources and duration, and iv) integration of schedule data need to be performed. In this paper, an analysis was made on the coherence of schedule activity descriptions of existing NPPs by applying the probabilistic methodology on activities with low connectivity due to the utilization of the numbering system of four APR1400 reactors (SHN 1&2 and SKN 3&4).This study also describes the method for developing a standard schedule activity list and connectivity measures by extracting same and/or similar schedule activities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.441-454
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• 2018

Since the permanent shutdown of Kori Unit 1 in 2017, a full-scale decommissioning project for a commercial nuclear reactor has been approaching. It is estimated that about 160,000 t of low-activity concrete waste will be produced from decommissioning of one unit of this commercial nuclear power reactor. Accordingly, it is necessary to review whether the effectiveness of the current regulatory framework for clearance waste (i.e. waste stream that meets activity concentration guidelines or dose criteria for clearance set forth in NSSC Notice No. 2017-65) can be maintained for the clearance of a bulk amount of concrete waste. In this regard, the IAEA SRS No. 44, which was used as a basis for revision of the Korean clearance regulations, is thoroughly analyzed and the radiological effects from four different clearance scenarios, along with input values and parameters derived from industrial practices in Korea, were evaluated. Though it is shown that the maximum annual dose from most recycling scenarios will be less than the clearance dose criterion for the normal scenario (i.e. an order of magnitude of $0.01mSv{\cdot}y^{-1}$), the radiation dose, estimated with conservative assumptions for the banking scenario, may exceed the above clearance dose criteria. Therefore, for safe and sustainable clearance of the bulk amount of concrete waste, it is required to diversify the concrete waste processors, perform more detailed site-specific assessment, and apply limiting conditions to the banking scenario.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.337-344
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• 1994

Enthalpy rise hot channel factor($F_{\Delta{H}}$$^{N}$) is one of the most limiting constraints in determining the fuel loading pattern(LP) for PWR's. In order to enhance the LP design flexibility without any changes of not only basic fuel specifications but also Technical Specifications and Operation Procedures, we apply the enrichment zoning concept to Westinghouse designed PWR's to flatten the rod power distributions within the fuel assembly and thus to reduce $F_{\Delta{H}}$$^{N}$. Enrichment zoning is described that each assembly consists of two different enrichment fuels ; the lower enriched fuels are located in positions which are expected to have the higher rod power and vice versa for the higher enriched fuels. As a result of unit assembly calculations to flatten the rod power distribution within the assembly, the appropriate enrichment difference is found to be 0.3~0.4w/o. Through core depletion calculations for the 18-month cycle of Kori Unit 4, the $F_{\Delta{H}}$$^{N}$ behavior in core with the enrichment zoning concept is investigated. A comparison with the reference case without the enrichment zoning results in a reduction in $F_{\Delta{H}}$$^{N}$ of approximately 1.5%.TEX>H/$^{N}$ of approximately 1.5%.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.1-9
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• 1989

An analytic method for quantitative comparisions between the alternatives for radiation protection optimization is required to aid the decision making process. This paper introduces the dynamic Markov model to evaluate the effect of inservice inspection, testing, and repair activities of the plant on radiation protection. In the example to put the Markov model into practice, the steam generator inspection intervals which minimize expected cost and total exposure dose were determined using the data for Kori-2 unit and foreign plants. The results show that the effect of the radiation exposure on the steam generator inspection interval is determined by the cost rather than the radiation exposure. The Markov model used in the example can be applied easily to the domestic NPPs by replenishing the data and also can be used in evaluating the comparative priority between various alternatives for radiation protection optimization.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• Korean System Dynamics Review
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• v.16 no.3
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• pp.5-29
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• 2015

The importance of the concept of safety culture has increased in the security of high-risk facility after Chernobyl accident in 1986. This paper elaborated the concept of safety culture and its main factors by Causal Loop Diagram. Due to the decline of safety culture, the occurrence of incidents and accidents require more and more corrective actions to the members of high-risk facilities and thereby increasing their workloads. Employees who must complete the task within the given time have to have time pressures and don't comply with the rules and procedures. Also, a schedule pressure is a big stress for employees, causing mistakes in precision work. In order to improve these problems, CLD of the safety culture in this paper suggests hiring more workers, re-allocation of given workloads and strengthen the learning, communication capabilities and safety leadership. In addition, the two real accident cases were analyzed to test the feasibility of the System Dynamic simulation model through the process of structuring the fault trees on the stationary black out accident in Kori unit 1 in South Korea and Kleen Energy power station explosion in US. The simulation results show that the various safety factors cause the serious accident combined with mechanical failure and safety culture will reduce the possibility of the accidents in these high-risk organizations. This simulation model can contribute to analyzing the impact of the organizational and human factors of safety culture and can provide the alternatives in high-risk facilities.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4809-4818
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• 2022

The decommissioning project of a nuclear facility is a large-scale process that is expected to take about 15 years or longer. The range of risks to be considered is large and complex, then, it is expected that various risks will arise in decision-making by area during the project. Therefore, in this study, the risk family derived from the Decommissioning Risk Management (DRiMa) project was reconstructed into a decommissioning project risk profile suitable for the Kori Unit 1. Two criteria of uncertainty and importance are considered in order to prioritize the selected 26 risks of decommissioning project. The uncertainty is scored according to the relevant laws and decommissioning plan preparation guidelines, and the project importance is scored according to the degree to which it primarily affects the triple constraints of the project. The results of risks are divided into high, medium, and low. Among them, 10 risks are identified as medium level and 16 risks are identified as low level. 10 risks, which are medium levels, are classified in five categories: End state of decommissioning project, Management of waste and materials, Decommissioning strategy and technology, Legal and regulatory framework, and Safety. This study is a preliminary assessment of the risk of the decommissioning project that could be considered in the preparation stage. Therefore, we expect that the project risks considered in this study can be used as an initial data for reevaluation by reflecting the detail project progress in future studies.

• Journal of Radiation Industry
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• v.17 no.4
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• pp.509-518
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• 2023

The Kori Unit 1 nuclear power plant, which is planned to be dismantled after permanent shutdown, is expected to generate a large amount of various types of radioactive waste during the dismantling process. For the disposal of Very-low-level waste, which is expected to account for the largest amount of generation, the Korea Radioactive waste Agency (KORAD) is in the process of detailed design to build a 3-phase landfill disposal facility in Gyeongju. In addition, a large container is being developed to efficiently dispose of metal and concrete waste, which are mainly generated as Very low-level waste of decommissioning. In this study, based on the design characteristics of the 3-phase landfill disposal facility and the large container under development, radiation exposure dose evaluation was performed considering the normal and accident scenarios of radiation workers during operation. The direct exposure dose evaluation of workers during normal operation was performed using the MCNP computer program, and the internal and external exposure dose evaluation due to damage to the decommissioning waste package during a drop accident was performed based on the evaluation method of ICRP. For the assumed scenario, the exposure dose of worker was calculated to determine whether the exposure dose standards in the domestic nuclear safety act were satisfied. As a result of the evaluation, it was confirmed that the result was quite low, and the result that satisfied the standard limit was confirmed, and the radiational disposal suitability for the 3-phase landfill disposal facility of the large container for dismantled radioactive waste, which is currently under development, was confirmed.