• Environmental and Resource Economics Review
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• v.27 no.2
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• pp.261-286
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• 2018

This paper evaluated the economic feasibility of the life extension of Kori unit 1 and Wolsong unit 1 according to the types of the nuclear power plants (NPPs) and the life extension period comparing to the levelized costs of energy (LCOE) of the new NPPs, coal-fired plants (CFPs), and combined cycle gas turbine (CCGTs) which proposed in the $7^{th}$ Basic Plan for Electricity Supply and Demand. The economic feasibility of the life extension of NPPs using LCOE method is affected by the types of NPPs, lifetime extension periods, discount rate, and capacity factor. According to the analysis results, the pressurized light water reactor (PWR) is more economical than the pressurized heavy water reactor (PHWR). Comparing the economical efficiency between the life extension of NPPs and other alternatives, the operation of the PWR for 20 years is more economical than the one of new NPPs and CFPs. However, 20 years of life extension of PHWR is more economical than the CCGTs, but less economical than new NPPs and CFPs. In summary, the 20 years of life extension of the NPPs seems to be more, especially for the PWR, which is more cost effective than other generation alternatives. Therefore, the government policy of the life extension of NPPs need to be a selective approach that simultaneously considers both safety and economics rather than closing all NPPs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.75-87
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• 2020

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS⁺), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS⁺ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10^-4 mSv·yr^-1, primarily from ¹²⁹I. The proposed KRS⁺ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv·yr^-1.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.91-98
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• 2009

The "3+3 Process" for safety critical software for nuclear power plants' I&C (Instrumentation and Control system) has been developed in this work. The main idea of the "3+3 Process" is both to simplify the software development and safety analysis in three steps to fulfill the requirements of a software safety plan [1]. The "3-Step" software development process consists of formal modeling and simulation, automated code generation and coverage analysis between the model and the generated source codes. The "3-Step" safety analysis consists of HAZOP (hazard and operability analysis), FTA (fault tree analysis), and DV (design validation). Put together, these steps are called the "3+3 Process". This scheme of development and safety analysis minimizes the V&V work while increasing the safety and reliability of the software product. For assessment of this process, validation has been done through prototyping of the SDS (safety shut-down system) #1 for PHWR (Pressurized Heavy Water Reactor).

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Nuclear Engineering and Technology
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• v.11 no.4
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• pp.263-274
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• 1979

A statistical approach is used to investigate the relative economic advantages of pressurized water reactor (PWR) and pressurized heavy water reactor (PHWR-CANDU) nuclear power plants for hypothetical 900Mwe systems with the throwaway fuel cycle to be built in the Republic of Korea. Power cost is decomposed into the cost components related to the plant capital, operation and maintenance, working capital requirements and fuel cycle operation. The calculation of construction cost is performed with the modified version of computer code ORCOST, and the modified POWERCO-50 is used to evaluate the cost components. Most of economic parameters are treated as statistical variables, each being given with a certain range. Through a random sampling procedures. the probability histograms on unit plant construction costs and power generating costs are obtained. The power cost probability histograms of the PWR and the PHWR plants overlap considerably, and the power costs of two systems appear to be almost same with the PHWR power cost being 0.4mil1/kwh lower compared with 39.4 mills/kwh for the PWR plant (July 1986 US-dollars). When a construction period of PHWR plant is longer by one year than that of PWR plant, there is no difference in the unit power cost of two plants. This comparison leads to no definite conclusion on the cost advantage of the PWR plant versus the PHWR plant. We conclude that the selection issue of nuclear power plants in Korea still remains an open question and that future effort to solve this question should be made toward economic quantification of those factors such as technology transfer and localization.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1540-1554
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• 2023

To use nuclear energy sustainably, spent nuclear fuel, classified as high-level radioactive waste and inevitably discharged after electricity generation by nuclear power plants, must be managed safely and isolated from the human environment. In Korea, the land area is limited and the amount of high-level radioactive waste, including spent nuclear fuels to be disposed, is relatively large. Thus, it is particularly necessary to maximize disposal efficiency. In this study, a high-efficiency deep geological repository concept was developed to enhance disposal efficiency. To this end, design strategies and requirements for a high-efficiency deep geological repository system were established, and engineered barrier modules with a disposal canister for pressurized water reactor (PWR)-type and pressurized heavy water reactor type Canada deuterium uranium (CANDU) plants were developed. Thermal and structural stability assessments were conducted for the repository system; it was confirmed that the system was suitable for the established strategies and requirements. In addition, the results of the nuclear safety assessment showed that the radiological safety of the new system met the Korean safety standards for disposal of high-level radioactive waste in terms of radiological dose. To evaluate disposal efficiency in terms of the disposal area, the layout of the developed disposal areas was assessed in terms of thermal limits. The estimated disposal areas were 2.51 km² and 1.82 km² (existing repository system: 4.57 km²) and the excavated host rock volumes were 2.7 Mm³ and 2.0 Mm³ (existing repository system: 4.5 Mm³) for thermal limits of 100 ℃ and 130 ℃, respectively. These results indicated that the area and the excavated volume of the new repository system were reduced by 40-60% compared to the existing repository system. In addition, methods to further improve the efficiency were derived for the disposal area for deep geological disposal of spent nuclear fuel. The results of this study are expected to be useful in establishing a national high-level radioactive waste management policy, and for the design of a commercial deep geological repository system for spent nuclear fuels.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.40-44
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• 2015

Sensitivity on the reactor safety was evaluated for the safety margin and time delay applied to the opening pressure of liquid relief valve(LRV) of the primary heat transport system(PHTS) in the pressurized heavy water reactor(PHWR) type nuclear power plant. Since the LRV is the pressure boundary for the PHTS in the safety analysis, the operating of LRV has a significant effect on the safety analysis results. Therefore it is required during the regulatory review of Wolsong Unit 1 safety analysis to find the safety effect of the application of safety margin and time delay to the LRV opening pressure for the safety analysis of PHTS pressurizing events.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.2
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• pp.136-141
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• 2004

A torsional guided wave mode was applied to detect a crack in a pipe. An array of electromagnetic acoustic transduce. (EMAT that can generate and receive torsional guided ultrasound with the frequency of 200kHz was designed and fabricated for testing a pipe of 2.5 inch diameter Artificial notches with various depths were fabricated in a bent feeder pipe mock-up and the detectability was examined from the distance of 2m of the specimen. The axial notches with the depth of 5% of wall thickness were successfully detected by a torsional mode (T(0,1)) generated by the EMAT However, it was found that the depth of defects was not related to the signal amplitude.

• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.33-45
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• 2013

Systems Engineering Approach was applied to the development of operator-support core management system based on the on-site operation experience and document of core management procedures, which is for enhancing operability and safety in PHWR (Pressurized Heavy Water Reactor) operation. The dissertation and definition of the system were given on th basis of investigating and analyzing the core management procedures. Fuel management, detector calibration, safety management, core power distribution monitoring, and integrated data management were defined as main user's requirements. From the requirements, 11 upper functional requirements were extracted by considering the on-site operation experience and investigating documents of core management procedures. Detailed requirements of the system which were produced by analyzing the upper functional requirements were identified by interviewing members who have responsibility of the core management procedures, which were written in SRS (Software Requirement Specification) document by using IEEE 830 template. The system was designed on the basis of the SRS and analysis in terms of nuclear engineering, and then tested by simulation using on-site data as a example. A model of core power monitoring related to the core management was suggested and a standard process for the core management was also suggested. And extraction, analysis, and documentation of the requirements were suggested as a case in terms of systems engineering.