• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.117-123
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• 2015

To remove heat generated during a burn-up test of nuclear fuels, the heat generation rate of nuclear fuels should be calculated accurately, and a coolant should be circulated in the test loop at an adequate flow rate. HANARO is an open pool-type reactor with an independent test loop for the burn-up test of nuclear fuels. A test rig is installed in the test loop, and a coolant is circulated through the test loop to maintain the temperature of the nuclear fuel rods within a desired temperature during an irradiation test. The components and sensors in the test rig can be broken or malfunction owing to the flow-induced vibration. In this study, a coolant flow simulation system was developed to verify and confirm the soundness of components and sensors assembled in the test rig with a high flow rate of the coolant.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.855-864
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• 2017

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.28-36
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• 1993

Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.

• Journal of the Korean Society of Safety
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• v.35 no.5
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• pp.115-120
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• 2020

From 2009 when the CNS facility was installed, the number of reactor failures due to abnormal CNS facility system has increased significantly. Of the total of 19 nuclear reactor shutdowns over the six years from 2009 to 2019, there were 10 nuclear reactor shutdowns associated with the CNS facility, which are very numerous. Therefore, this report intends to analyze the history of nuclear reactor shutdowns due to CNS facility system failure in detail, and to present the root cause and solution to problems. As a result of FMEA implementation of CNS facility system, a total of 76 SPVs were selected. In addition, 10 cases of reactor shutdown history due to CNS facility system abnormalities were analyzed in detailed, and improvement plans for solving the root cause and problem were suggested for each trip history. The results of this study are expected to be able to operate the domestic research reactor and CNS facilities more stably by providing effective measures to prevent recurrence of CNS facilities and reactor trips.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.39 no.2
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• pp.113-118
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• 2016

Project management is a tool for smooth operation during a full cycle from the design to normal operation including the schedule, document, and budget management, and document management is an important work for big projects such as the JRTR (Jordan Research and Training Reactor). To manage the various large documents for a research reactor, a project management system was resolved, a project procedure manual was prepared, and a document control system was established. The ANSIM (Advanced Nuclear Safety Information Management) system consists of a document management folder, document container folder, project management folder, organization management folder, and EPC (Engineering, Procurement and Construction) document folder. First, the system composition is a computerized version of the Inter-office Correspondence (IOC), the Document Distribution for Agreement (DDA), Design Documents, and Project Manager Memorandum (PM Memo) works prepared for the research reactor design. Second, it reviews, distributes, and approves design documents in the system and approves those documents to register and supply them to the research reactor user. Third, it integrates the information of the document system-using organization and its members, as well as users' rights regarding the ANSIM document system. Throughout these functions, the ANSIM system has been contributing to the vitalization of united research. Not only did the ANSIM system realize a design document input, data load, and search system and manage KAERI's long-period experience and knowledge information properties using a management strategy, but in doing so, it also contributed to research activation and will actively help in the construction of other nuclear facilities and exports abroad.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3566-3570
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• 2024

The production of the radioisotope molybdenum-99 in the WWR-K research reactor is achieved through the activation method ⁹⁸Mo(n,γ)⁹⁹Mo, utilizing a target of natural molybdenum trioxide irradiated under standard conditions (thermal neutron spectra and water environment). Under such conditions, the maximum specific activity of molybdenum-99 reaches (2.3 ± 0.3) Ci/g Mo after 7 d of irradiation. However, the escalating demand for molybdenum-99 and the need to reduce its production cost, necessitates urgent and increased productivity. This study aims to optimize the irradiation conditions for molybdenum powder in the WWR-K reactor to increase the specific activity of molybdenum-99. For this purpose, we evaluated various irradiation capsule designs comprised various neutron moderator materials and thicknesses. Through extensive modeling calculations, we obtained an optimal capsule design that increases the specific activity of molybdenum-99 to 3.31 Ci per 1 g of Mo.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.