• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.455-467
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• 2002

RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.5
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• pp.329-337
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• 2016

As a turbulence-enhancing device, a mixing vane, which is installed at a spacer grid of the fuel assembly, plays an important role in improving convective heat transfer by generating either swirl flow in the subchannels or cross flow between the fuel rod gaps. Therefore, both the geometric configuration and the arrangement pattern of a mixing vane are important factors in determining the performance of a mixing vane. In this study, in order to examine the flow-distribution features inside a $5{\times}5$ fuel assembly with split-type mixing vanes, which was used in the benchmark calculation of the OECD/NEA, we conduct simulations using the commercial computational fluid dynamics software, ANSYS CFX R.14. We compare the predicted results with measured data obtained from the MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, we discuss the effect of the split-type mixing vanes on the flow pattern inside the fuel assembly.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1998.05a
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• pp.235-242
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• 1998

Aiming at one of decisive alternatives for long term aspect of nuclear power concerns, an integral and closed nuclear system, AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor) concept is under development. The AMBIDEXTER complex essentially comprises two mutually independent loops of the radiation/material transport and the heat/energy conversion, centered at the integrated reactor assembly, which enables one to utilize maximum benefits of nuclear energy under minimum risks of nuclear radiation. And it provides precious radioisotopes and radiation sources from its waste stream. Also the reactor operates at very low level of fission products inventory throughout its lifetime. The nuclear and thermalhydraulic characteristics of the molten TH/$^{233}$ U fuel salt extend the capability of the self-sustaining AMBIDEXTER fuel cycle to enhance resource security and safeguard transparency. The reactor system is consisted of a single component module of the core, heat exchangers and recirculation pumps with neither pipe connections nor active valves in between, which will significantly improve inherent features of nuclear safety. States of the core technologies associated with designing and developing the AMBIDEXTER concept are mostly available in commercialized form and thus demonstration of integral aspects of the concept should be the prime area in future R＆D programs.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.488-493
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• 2018

The Firebrick Resistance-Heated Energy Storage (FIRES) concept developed by the Massachusetts Institute of Technology aims to enhance profitability of the nuclear power industry in the next decades. Studies carried out at Massachusetts Institute of Technology already provide estimates of the potential revenue from FIRES system when it is applied to industrial heat supply, the likely first application. Here, we investigate the possibility of operating a power plant (PP) with a fluoride-salt-cooled high-temperature reactor and a closed Brayton cycle. This variant offers features such as enhanced nuclear safety as well as flexibility in design of the PP but also radically changes the way of operating the PP. This exploratory study provides estimates of the revenue generated by FIRES in addition to the nominal revenue of the stand-alone fluoride-salt-cooled high-temperature reactor, which are useful for defining an initial design. The electricity price data is based on the day-ahead markets of Germany/Austria and the United States (Iowa). The proposed method derives from the equation of revenue introduced in this study and involves simple computations using MatLab to compute the estimates. Results show variable economic potential depending on the host grid but stress a high profitability in both regions.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.3975-3984
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• 2022

Distributed control system architecture is adopted for I&C systems of Prototype Fast Breeder Reactor, where the geographically distributed control systems are connected to centralized servers & display stations via switched Ethernet networks. TCP/IP communication plays a significant role in the successful operations of this architecture. The communication tasks at control nodes are taken care by TCP/IP offload modules; local area switched network is realized using layer-2/3 switches, which are finally connected to network interfaces of centralized servers & display stations. Safety, security, reliability, and fault tolerance of control systems used for safety-related applications of nuclear power plants is ensured by indigenous design and qualification as per guidelines laid down by regulatory authorities. In the case of commercially available components, appropriate suitability analysis is required for getting the operation clearances from regulatory authorities. This paper details the proposed approach for the suitability analysis of TCP/IP communication nodes, including control systems at the field, network switches, and servers/display stations. Development of test platform using commercially available tools and diagnostics software engineered for control nodes/display stations are described. Each TCP link behavior with impaired packets and multiple traffic loads is described, followed by benchmarking of the network switch's routing characteristics and security features.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.48-57
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• 2018

Configuration management (CM) is the process of identifying and documenting characteristics of plant structures, systems and components (SSCs), and of ensuring that changes to these characteristics are properly assessed, approved, implemented, verified and recorded. The purpose of this study is to examine regulation and technical standards developed under different concepts and level of depth for classification of nuclear SSCs as an essential prerequisite of the CM. In this context, main contents of currently adopted NSSC Notice 2016-10 are reviewed and compared with those in recently published ANSI/ANS 58.14 and IAEA SSG-30. The technical standards were prototypically used for classification of O-rings in two nuclear systems. It is found that ANSI/ANS 58.14 results in different categories taking into account specific features while IAEA SSG-30 leads to same categorization of the O-rings. Key findings will be summarized for Korean regulatory amendment in the future.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.32-37
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• 2009

The emergency diesel generator (EDG) in a nuclear power plant (NPP) shall start within 10 secondss and supply electrical power to engineered safety features within one minute and less if a loss of offsite power (LOOP), A design-basis event, or their combination occur. Each NPP has an EDG set consisting of two diesel generators for redundancy. In addition to the EDG set, an alternate Alternating Current Diesel Generator (AAC DG) is installed and shared by several units to cope with a station black out (SBO), i.e., loss of the offsite power concurrent with reactor trip and unavailability of the EDG set. The objective of this study is to analyze the failure data of emergency diesel generators reported in overseas nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.493-498
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• 1996

A PC window-based computer code, CONPAS(CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree(CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree(LSET) for its detailed quantification. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, and sensitivity analysis, reporting aspects including tabling and graphic, and user-friend interface.

• Journal of the Ergonomics Society of Korea
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• v.27 no.1
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• pp.21-27
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• 2008

An increased use of teams of actors within complex systems has led to the emergence of various approaches for the assessment of different features associated with team performance. Over the last two decades, the performance of teams in complex systems has received considerable attention from the human factors community, and a number of methods have been developed in order to assess and evaluate team performance. The purpose of this paper is to propose a methodology for measuring team efficiency of operators in the advanced main control room of Korean nuclear power plant. Team efficiency is an index which is estimated of gabs between individual performances and team performance. The index was developed to compare among teams through past all performance measurements.