• Proceedings of the Computational Structural Engineering Institute Conference
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• 2001.04a
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• pp.259-266
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• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell and the lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these large pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear static structural analysis. Canister types studied here are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.163-168
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• 1996

Simulations of Wolsong-1 Phase-B commissioning measurements have been performed, as part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. A required component of these simulations is the calculation of incremental cross sections representing reactivity control devices in the reactor. The incremental cross section properties of the Wolsong-1 adjusters, Mechanical Control Absorbers (MCA) and liquid Zone Control Units (ZCU) are based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods.

• Fire Science and Engineering
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• v.30 no.4
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• pp.20-26
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• 2016

When the fire takes place in Nuclear Powr Plants(NPPs), the reactor should achieve and maintain safe shut-down conditions and minimize the radioactive material released to the environment. The U.S. Nuclear Regulatory Commission (NRC) has issued numerous generic communications related to fire protection over the past 20 years, after it issued its requirements in the Fire Protection Rule set forth in Title 10, Section 50.48 of the Code of Federal Regulations (10 CFR 50.48) and Appendix R to the 10 CFR 50. The and Nuclear Energy Institute (NEI) has developed a Methodology for Risk Informed Fire Safe-Shutdown Analysis, which is related to the Deterministic Method for Multiple Spurious Operations solutions. The aim of this study was to identify, achieve, and maintain Post-Fire Safe-Shutdown of the Main Control Room (MCR) of the CANDU reactor, even though one train of the multiple Safety Structures, Systems, and Components (SCCs) fail by the technical specification and analysis method.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1892-1900
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• 2023

A protective oxide layer forms on the material surfaces of a Nuclear Power Plant during operation due to high temperature. These oxides can host radionuclides, the activated corrosion products of fission products, resulting in decommissioning workers' exposure. These deposited oxides are iron oxides such as Fe₃O₄, Fe₂O₃ and mixed ferrites such as nickel ferrites, chromium ferrites, and cobalt ferrites. Developing a new chemical decontamination technology for domestic CANDU-type reactors is challenging due to variations in oxide compositions from different structural materials in a Pressurized Water Reactor (PWR) system. The Korea Atomic Energy Research Institute (KAERI) has already developed a chemical decontamination process for PWRs called 'HyBRID' (Hydrazine-Based Reductive metal Ion Decontamination) that does not use organic acids or organic chelating agents at all. As the first step to developing a new chemical decontamination technology for the Pressurized Heavy Water Reactor (PHWR) system, we investigated magnetite dissolution behaviors in various HyBRID inorganic acidic solutions to assess their applicability to the PHWR reactor system, which forms a thicker oxide film.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.601-610
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• 2024

This paper gives an extensive analysis on the characterization of xenon isotopic ratios for various nuclear reactors and nuclear explosions through neutronic depletion codes. The results of the characterization can be used for discriminating the sources of the xenon isotopes' release among the nuclear explosions and nuclear reactors. The considered sources of the xenon radionuclides do not only include PWR, CANDU, and nuclear explosions using uranium and plutonium bombs, but also IRT-200 and 5MWe Yongbyon (MAGNOX reactor) research reactors operated in North Korea. A new data base (DB) on xenon isotopic activity ratios was produced using the results of the characterization, which can be used in discrimination of the sources of xenon isotopes. The results of the study show that 5MWe Yongbyon reactor has quite different characteristics in ¹³⁵Xe/¹³³Xe ratio from the PWRs and the nuclear reactors have different characteristics in ¹³⁵Xe/¹³³Xe ratios from the nuclear explosions.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.13-19
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• 2009

As the operating time in nuclear power plants (NPPs) increases, the integrity of nuclear components may be continually degraded due to aging effects of systems, structures and components. Recently, a number of NPPs are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. Therefore, it is beneficial to build a monitoring system to measure an aging status. In this paper, the Aging Monitor (AM) based on lots of aging database obtained from the operating plants and research results on the aging effects was developed to monitor, manage and evaluate the aging phenomena systematically and effectively in NPPs. The AM for the CANDU is divided into 6 modules: (1) Aging Alarm/Coloring Monitor, (2) Aging Database, (3) Aging Document, (4) Real-time Integrity Monitor, (5) Surveillance and Inspection Management System, and (6) Continued Operation and Periodic Safety Review (PSR) Safety Evaluation. The proposed system is expected to provide the integrity assessment for the major mechanical components of an NPP under concurrent working environments.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.468-481
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• 2002

There are 4 CANDU6 type reactors operating at Wolsong site. For fuelling operation of certain fuel channels (with flow less than 21.5 kg/s) a FARE flow Assist Ram Extension) device is used. During the refuelling operation, two remote controlled F/Ms (Fuelling Machines) are attached to a designated fuel channel and carry out refuelling job. The upstream F/M inserts new fuel bundles into the fuel channel while the downstream F/M discharges spent fuel bundles. In order to assist fuelling operation of channels that has lower coolant How rate, the FARE device is used instead of F/M C-ram to push the fuel bundle string. The FARE device is essentially a How restricting element that produces enough drag force to push the fuel bundle string toward downstream F/M. Channels that require the use of FARE device for refuelling are located along the outside perimeter of reactor. This paper presents the FARE device design feature, steady state hydraulic and operational characteristics and behavior of the device when coupled with fuel bundle string during fuelling operation. The study showed that the steady state performance of FARE device meets the design objective that was confirmed by downstream F/M C-ram force to be positive.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.38-43
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• 2011

For the efficient and safe operation of nuclear power plant, evaluating quantitatively aging phenomenon of major components is necessary. Especially, typical aging parameters such as stresses and cumulative usage factors should be determined accurately to manage the lifetime of the plant facility. The 3-D finite element(FE) model is generated to calculate the aging parameters. Mechanical and thermal transfer functions called Green's functions are developed for the FE model with standard step input. The stress results estimated from transfer functions are verified by comparing with 3-D FE analyses results. Lastly, we suggest an effective fatigue evaluation methodology by using the transfer functions. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system.

• Nuclear Engineering and Technology
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• v.36 no.4
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.