• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• 공업화학
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• 제24권4호
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• pp.337-343
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• 2013

멤브레인 반응기는 멤브레인과 반응기를 결합하여 반응과 분리의 단위공정을 하나로 결합함으로써 전체공정을 단순화하고 반응효율을 높이고자 하는 혁신 기술로써, 멤브레인을 이용한 생성물의 선택적 제거를 통해 열역학적 평형을 뛰어넘는 전환율, 부반응물 생성 억제에 의한 반응 효율 및 선택성을 향상시킬 수 있다. 특히 이온전도성 세라믹을 이용한 멤브레인 반응기는 연료전지의 개발, 고순도 산소/수소의 분리/정제, 이산화탄소의 전환 및 다양한 화학제품제조에 까지 응용될 수 있기 때문에 시장의 확대와 더불어 크게 발전할 수 있을 것으로 기대된다. 본 총설에서는 수소이온 전도성 세라믹 멤브레인 반응기에 대한 연구동향과 다양한 응용분야 및 향후 전망 등에 고찰해 보고자 한다.

• 에너지공학
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• 제24권2호
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• pp.154-166
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• 2015

The in-core instrumentation measures core power distribution and coolant temperature in local regions of the core in pressurized water reactors. The installation types are distinguished by the designs of routing paths that exit either through reactor bottom mounted instrument nozzles or through reactor top mounted instrument nozzles. Although each type has unique advantages, it is generally known that top mounted design is more competitive with respect to emphasizing nuclear safety issues and ability to cope with severe accidents. The international nuclear vendors have provided various types of reactors with top mounted design. Nuclear power reactors in Korea, however, only have been designed to be applicable to the use of bottom mounted design, and it has been pointed out that the capabilities of Korean reactors against severe accidents should be further enhanced. The paper deals with technical issues on reactor internal and external design, in-core instrumentation, support assembly, sealing mechanism with nozzles, handling, and analytical issues in order to establish the ways of development.

• Nuclear Engineering and Technology
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• 제32권1호
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.561-570
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• 2008

When reactivity insertion such as refueling occurs in CANDU reactors, the power and the water levels are tilted in the upper outer zone of the LZCS (Liquid Zone Control System) and fluctuate unstably for a certain period of time (1-5 days). The instability described above is observed in most CANDU reactors in service around the world, but its root cause is unidentified and no solutions to this problem have been established. Therefore, this study attempted to prove experimentally and analytically that the root cause lies in the hold-up of light water on the top of the TSP (Tube Support Plate) due to the mismatch between net volumetric flow rate of light water and helium crossing the narrowed porous TSP installed within the LZCS compartment. Our method was to perform a hydrodynamic simulation of in/outflow of light water and helium. Two solutions for the aforementioned instability of LZCS are suggested. One is to regulate the compartment for both inflowing helium gas and outflowing light water; the other is to enlarge the flow paths of helium and light water within TSP. The former may be applicable to nuclear reactors in service and the latter to those planned for construction.

• 상하수도학회지
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• 제18권2호
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• pp.181-187
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• 2004

The purpose of this study is to investigate the biofilm reactors capable of doing high efficiency treatment. Vertical fluidized bed biofilm reactor(VFBBR) and spiral fluidized bed biofilm reactor(SFBBR) was used for their performence in biodegradation of artificial sewage. The factors influencing the efficiency of those reactors were compared with difference of physical condition. They had same size but different structure to gain access of its unique characteristics. When recycle solution with flow rate of 22 mL/min and artificial sewage with flow rate of 2~10 mL/min were fed into two reactors in aerobic state, the average $COD_{cr}$, removal rate for biodegradation of SFBBR was greater than VFBBR. After reactor feed sewage was constantly maintained as flow rate of 4 mL/min and the recycle solution were changed to 10~32 mL/min respectively, the average $COD_{cr}$ removal rate of artificial sewage in SFBBR was greater than VFBBR. In this experiment for addition of support media into two reactors SFBBR was 4.1% excellent than VFBBR. Above all, SFBBR excelled VFBBR in boidegradation of organic matter in sewage.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2005년도 하계학술대회 논문집 Vol.6
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• pp.296-297
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• 2005

We have investigated the quench performance of shunt reactors in the parallel connection of resistive type superconducting fault current limiter (SFCL) components based on YBCO films. To increase voltage rating, components are connected in series and to increase current level, they are connected in parallel. This method has cauesd the unbalanced quench between each components. To improve the problem, we have compared the quench properties between the current limiting components without and with shunt reactors connected in parallel. To improve the quench performance, across individual SFCL components connected the shunt reactor in parallel. The components with shunt reactors successfully produced simultaneous quench, resulting from the bypass of the fault current in the direction of the shunt reactor.

• Nuclear Engineering and Technology
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• 제37권1호
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• 방사성폐기물학회지
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• 제17권1호
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• pp.47-58
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• 2019

The paper presents the results of a multi-criteria comparative evaluation of 12 feasible Russian nuclear energy deployment scenarios with thermal and fast reactors in a closed nuclear fuel cycle. The comparative evaluation was performed based on 6 performance indicators and 5 different MCDA methods (Simple Scoring Model, MAVT / MAUT, AHP, TOPSIS, PROMETHEE) in accordance with the recommendations elaborated by the IAEA/INPRO section. It is shown that the use of different MCDA methods to compare the nuclear energy deployment scenarios, despite some differences in the rankings, leads to well-coordinated and similar results. Taking into account the uncertainties in the weights within a multi-attribute model, it was possible to rank the scenarios in the absence of information regarding the relative importance of performance indicators and determine the preference probability for a certain nuclear energy deployment scenario. Based on the results of the uncertainty/sensitivity analysis and additional analysis of alternatives as well as the whole set of graphical and attribute data, it was possible to identify the most promising nuclear energy deployment scenario under the assumptions made.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1244-1249
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• 2023

Recently, the Kingdom of Saudi Arabia (KSA) announced the development of first-of-a-kind(FOAK) and most advanced futuristic vertical city and named as 'The LINE'. The project will have zero carbon dioxide emissions and will be powered by clean energy sources. Therefore, a study was designed to understand which clean energy sources might be a better choice. Because of its nearly carbon-free footprint, nuclear energy may be a good choice. Nowadays, the development of very small modular reactors (vSMRs) is gaining attention due to many salient features such as cost efficiency and zero carbon emissions. These reactors are one step down to actual small modular reactors (SMRs) in terms of power and size. SMRs typically have a power range of 20 MWe to 300 MWe, while vSMRs have a power range of 1-20 MWe. Therefore, a study was conducted to discuss different vSMRs in terms of design, technology types, safety features, capabilities, potential, and economics. After conducting the comparative test and analysis, the fuel cycle modeling of optimal and suitable reactor was calculated. Furthermore, the levelized unit cost of electricity for each reactor was compared to determine the most suitable vSMR, which is then compared other generation SMRs to evaluate the cost variations per MWe in terms of size and operation. The main objective of the research was to identify the most cost effective and simple vSMR that can be easily installed and deployed.