• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• 한국유체기계학회 논문집
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• 제13권5호
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• pp.5-10
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• 2010

5MW, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1431-1441
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• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.157-164
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• 2015

Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

• 한국안전학회지
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• 제16권3호
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.877-887
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• 2022

A new reactor concept is described that directly couples a supercritical CO₂ (sCO₂) power cycle with a CO₂-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO₂ concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (~20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1303-1314
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• 2016

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical $CO_2$ cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 하계학술대회 논문집 D
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• pp.2252-2254
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• 2002

The Reactor internal structures which consist of many complex components are subjected to flow-induced vibration due to high temperature and pressure in Reactor coolant system. The above flow-induced vibration causes degradation of structural integrity of the Reactor and may result in loosing mechanical binding component which might impact other equipment and component or cause flow blockage. It is important to analyze reactor noise signal for the early detection of potential problem or failure in order to diagnosis reactor integrity in the point of view of safety and plant economics. Detailed composition of diagnostic system reactor internal structures using neutron noise(RIDS).