• Nuclear Engineering and Technology
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• 제51권2호
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• pp.546-555
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• 2019

Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

• 기술혁신연구
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• 제24권4호
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• pp.221-248
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• 2016

본 연구는 원자력기술의 혁신특성을 심층조사 함으로써 4세대 원자력에너지기술의 전환 시 발생 가능한 이슈를 이해하고 한국의 정책현황을 분석하여 4세대 원자력에너지기술의 성공적 안착을 위한 시사점을 도출하였다. 이는 현재 세계 5강의 원자력에너지기술 강국인 한국에서 4세대 원자력에너지기술의 도래에도 지속가능한 경쟁력을 유지하기 위한 중요한 정책적인 시사점으로 활용할 수 있을 것으로 사료된다. 정성적 문헌연구방법을 통해 원자력기술혁신에 관한 문헌들을 조사한 결과, 4세대 원자력 에너지기술의 전환에 영향을 미치는 요인은 크게 4가지로 확인할 수 있었다. 이는 장기간의 안정적인 자원 할당, 혁신을 위한 이해당사자 간의 지속적인 상호작용, 완전한 시스템을 위한 기술과 노하우의 축적, 적용 및 실증을 위한 정책적인 시장이었다. 이를 한국의 4세대 원자력에너지기술과 연관한 정책을 대상으로 적용하여 사례분석을 한 결과, 현재 4세대 기술이 기술개발의 초기단계에서 실증단계로 넘어가는 시점임을 감안하더라도 연관한 정책은 실증과 운영을 위한 전문인력의 체계적인 양성 방안, 사회적 수용성과 저항에 대한 대응, 실증에 대한 구체적인 계획 수립, 4세대 원자력시스템을 적용하기 위한 정책적인 시장을 제안하는 장기적인 노력, 이해당사자들 간의 구체적이고 지속적인 상호작용을 적극 장려하는 것이 체계적으로 필요함을 제시하였다.

• 한국기술혁신학회:학술대회논문집
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• 한국기술혁신학회 2003년도 추계학술대회
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• pp.359-368
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• 2003

21세기를 맞이하면서, 국제 원자력계는 원자력의 새로운 방향을 활발하게 모색하였고, 새로운 방향의 하나로서 혁신 개념의 원자로 개발이 필요하다는 데에 공통된 인식을 형성하고 있었다. 혁신 원자로 개발의 효과적 달성을 위하여 미국과 유럽, 일본 등과 우리나라를 포함한 원자력 선진국들은 에너지안정공급 능력이 우수하고 국민수용이 가능하며 절대 안전성의 확보 및 경제적으로 경쟁력이 우수한 원자력시스템 개발을 위한 노력을 활발하게 진행하고 있다. 본 연구에서는 미래형 혁신 원자력시스템의 기술 목표를 제4세대 원자력시스템을 기준으로 살펴보고, 현재 국제 공동연구로 개발을 추진 중인 제4세대 원자력시스템의 기술적 특성에 대하여 기술하였다.

• 전기학회논문지
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• 제63권4호
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• pp.455-460
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• 2014

South Korea's power consumption is increasing every year. For stable electric power supply, more generation facilities are needed. But it is not easy to build nuclear power generation facilities, so provision of renewable energy is thought of as the solution. For the system's stable management, practical use of energy storing system is needed. Currently, pumping up electric power station is considered most useful. In this study, we have calculated the least amount of energy storing device by considering the renewable energy, HVDC, and change in power for the appliance of ESS in Jeju system, according to The 6th Basic Plan for Long-term Electricity Supply and Demand. Also we have calculated the amount of the battery and about the load equalizing effect to use battery as power storing device. Finally, we have calculated the reduction of electricity generation and the reduction of $CO_2$ emission with this study.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.237-248
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• 2007

Sodium-cooled fast reactor is considered in France as a potential candidate for a prototype of 4th generation system to be built by 2020. A detailed working program has been launched recently to identify by 2012 the potential improvement tracks for later industrial development of these reactors. The goals for innovation are first identified: Progress of the safety with a special attention to severe accidents risk minimization and mitigation (defense in depth approach); Economic competitiveness of the system mainly by reducing the capital cost, the investment risks by enhancing in service inspection and repair capacities, and raising the availability; Sustainability with fissile material management while reducing the proliferation risk; capacity for long-lived waste transmutation.

• 한국수소및신에너지학회논문집
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• 제22권1호
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• pp.118-127
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• 2011

An IGCC was evaluated as one of the next generation technologies that would be able to substitute for coal-fired power plants. According to "The 4th Basic Plan of Long-term Electricity Supply & Demand" which is developed by the Electricity Business Acts, the first IGCC will be operated at 2015. Like other new and renewable energy such as solar PV, Fuel cell, The IGCC is considered as non-competitive generation technology because it is not maturity technology. Before the commercial operation of an IGCC in our electricity market, its economic feasibility under the Korean electricity market, which is cost-based trading system, is studied to find out institutional support system. The results of feasibility summarized that under the current electricity trading system, if the IGCC is considered like a conventional plant such as nuclear or coal-fired power plants, it will not be expected that its investment will be recouped within life-time. The reason is that the availability of an IGCC will plummet since 2016 when several nuclear and coal-fired power plants will be constructed additionally. To ensure the reasonable return on investment (NPV>0 IRR>Discount rate), the availability of IGCC should be higher than 77%. To do so, the current electricity trading system is amended that the IGCC generator must be considered as renewable generators to set up Price Setting Schedule and it should be considered as pick load generators, not Genco's coal fired-generators, in the Settlement Payment.

• 한국전산유체공학회지
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• 제21권1호
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.513-518
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• 1996

This report presents a demonstration of application of realistic evaluation methodology to a posturated cold leg large break LOCA in a Westinghouse three-loop pressurized water reactor with 17$\times$17 fuel. The new method of this analysis can be divided into three distinct step: 1) Best Estimate Code Validation and Uncertainty Quantification 2) Realistic LOCA Calculation 3) Limiting Value LOCA Calculation and Uncertainty Combination RELAP5/MOD3/K ［1］, which was improved from RELAP5/MOD3.1, and CONTEMPT4/MOD5 code were used as a best estimate thermal-hydraulic model for realistic LOCA calculation. The code uncertainties which will be determined in step 1) were quantified already in previous study ［2］, and thus the step 2) and 3) for plant application were presented in this paper. The application uncertainty parameters are divided into two categories, i.e. plant system parameters and fuel statistical parameters. Single parameter sensitivity calculations were performed to select system parameters which would be set at their limiting value in Limiting Value Approach (LVA) calculation. Single run of LVA calculation generated 27 PCT data according to the various combinations of fuel parameters and these data provided input to response surface generation. The probability distribution function was generated from Monte Carlo sampling of a response surface and the upper 95$^{th}$ percentile PCT was determined. Break spectrum analysis was also made to determine the critical break size. The results show that sufficient LOCA margin can be obtained for the demonstration NPP.