• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.1-17
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• 2018

This article reviews the multiscale nature of stress corrosion cracking (SCC) observed by high-resolution characterizations in austenite stainless steels and Ni-base superalloys in light water reactors (including boiling water reactors, pressurized water reactors, and supercritical water reactors) with related opinions. A new statistical summary and comparison of observed degradation phenomena at different length scales is included. The intrinsic causes of this multiscale nature of SCC are discussed based on existing evidence and related opinions, ranging from materials theory to practical processing technologies. Questions of interest are then discussed to improve bottom-up understanding of the intrinsic causes. Last, a multiscale modeling and simulation methodology is proposed as a promising interdisciplinary solution to understand the intrinsic causes of the multiscale nature of SCC in light water reactors, based on a review of related supporting application evidence.

• Nuclear Engineering and Technology
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• v.40 no.7
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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.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.603-616
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• 2009

The past two decades were mainly devoted to model validation and computer code verification against global corium experiments, code application to reactor situations, and investigation of the role of melt properties in steam explosion energetics. Corium data were essentially provided by JRC-Ispra in the FARO and KROTOS facilities and by KAERI in the TROI facility. Verification of code applicability to reactor situations was performed essentially in the frame of the international OECD/SERENA programme. The paper makes a synthesis of the findings made during the above-mentioned period and expresses a personal view of the author with respect to the progress made and expected for the resolution of the steam explosion issue for light water reactors.

• Composites Research
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• v.35 no.3
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• pp.161-174
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• 2022

SiC fiber-reinforced SiC matrix composite is a promising accident-tolerant fuel cladding material to improve the safety of light water nuclear reactors. Compared to the current zirconium alloy fuel cladding as well as metallic accident-tolerant fuel cladding, SiC composite fuel cladding has exceptional accident-tolerance such as excellent structural integrity and extremely low corrosion rate during severe accident of light water nuclear reactors, which reduces reactor core temperature and delays core degradation processes. In this paper, we introduce the concept, technical issues, and properties of SiC composite accident-tolerant fuel cladding during operation and accident scenarios of light water nuclear reactors.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Journal of Engineering Education Research
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• v.15 no.1
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• pp.45-52
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• 2012

The success history of Light Water Reactors (PWR and BWR) showed how a dominant technology could be shaped in a political and economical context. The american nuclear politics, the interest of american nuclear industry, and the accumulated technological know-hows made it possible that the not inherently safe reactor-Light Water Reactor- became a prominent reactor model. The path dependency of reactor technology on LWR kept the engineers from developing a new safer reactor, even if the severe reactor accidents occurred. In oder to increase safety of nuclear power system, we should understand the social shaping process of nuclear technology.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.259-270
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• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.44-44
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• 2002

• Nuclear Engineering and Technology
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• v.41 no.8
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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.