• Nuclear Engineering and Technology
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• 제43권4호
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• pp.329-334
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• 2011

Two conceptual flowsheets were developed for recycling used nuclear fuel. One flowsheet was developed for recycling used oxide nuclear fuel from light water reactors while the other was developed for recycling used metal fuel from fast spectrum reactors. Both flowsheets were developed from a set of design principles including efficient actinide recovery, nonproliferation, waste minimization and commercial viability. Process chemistry is discussed for each unit operation in the flowsheet.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.104-115
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• 1999

The turbulent mixing rate is a very important variable in the thermal-hydraulic design of nuclear reactors. In this study, the turbulent mixing rate the fluid flows through rod bundles is estimated with the scale analysis on the flow pulsation phenomenon. Based upon the assumption that the turbulent mixing is composed of molecular motion, isotropic turbulent motion (turbulent motion without the flow pulsation), and How pulsation, the scale relation for the mixing is derived as a function of P/D, Re, and Pr. The derived scale relation is compared with published experimental results and shows good agreements. Since the scale relation is applicable to various Prandtl number fluid flows, it is expected to be useful for the thermal-hydraulic analysis of liquid metal coolant reactors as well as of moderate Prandtl number coolant reactors.

• Nuclear Engineering and Technology
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• 제12권3호
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• pp.194-213
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• 1980

The paper describes and quantifies the wastes arising in the nuclear fuel cycle for Light Water Reactors, Heavy Water Reactors and Fast Breeder Reactors. The management and disposal technologies are indicated, together with their environmental impacts. Both once-through and uranium-plutonium recycle systems are evaluated, and comparisons are made on the basis of tingle reference technologies for waste management, and for one gigawatt/year of electricity generation. Environmental impacts are assessed, particularly that of health and safety, and a reference costing system is applied purely as a basis for comparing the fuel cycles. From this study it call be concluded generally that the relative differences of the impacts of waste management and disposal between the selected fuel cycles are not decisive factors in choosing a fuel cycle. Employing the technologies assumed, the radioactive wastes from any of the fuel cycles studied can be managed and disposed of with a high degree of safety and without undue risk to man or the environment. The cost of waste management and disposal is only a few percent of the value of the electricity generated and does not vary greatly between fuel cycles.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.91-96
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• 1995

In this paper, research efforts to develop computer code system for analysis of the transmutation reactors at KAIST are described Especially the computer code HANCELL for assembly calculation of fast reactors is mainly described. Features and function of the code are identified md current status of the code development is provided

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.199-206
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• 2012

Using fast neutron Generation IV reactors, recycling of americium and curium may become feasible. The detrimental impact of americium on safety parameters has recently been quantified in terms of a power penalty for surviving a given set of transients in sodium fast reactors. In the present paper, a review of the physical reasons for the adverse effect of americium is provided, and different Gen-IV technologies are assessed with respect to their capability of hosting americium in the fuel.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.80-87
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• 2018

Consideration is given to the possibility to use changes in buoyancy as a negative reactivity feedback mechanism during temperature transients in heavy liquid metal fast reactors. It is shown that by the proper use of heavy pellets in the fuel elements, fuel rods could be endowed with a passive self-ejection mechanism and then with a negative feedback. A first estimate of the feasibility of the mechanism is calculated by using a simplified geometry and model. If in addition, a neutron poison pellet is introduced at the bottom of the fuel, then when the fuel element is displaced upward by buoyancy force, the reactivity will be reduced not only by disassembly of the core but also by introducing the neutron poison from the bottom. The use of induced buoyancy opens up the possibility of introducing greater amounts of actinides into the core, as well as providing a palliative solution to the problem of positive coolant temperature reactivity coefficients that could be featured by the heavy liquid metal fast reactors.

• 한국세라믹학회지
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• 제42권8호
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• pp.575-581
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• 2005

SiC-based ceramics are considered as candidate materials for the advanced nuclear energy systems such as the generation IV reactors and the fusion reactors due to their excellent high-temperature strength and irradiation resistance. The advanced nuclear energy systems and their main components adopting ceramic composites were briefly reviewed. A novel fabrication method of $SiC_f/SiC$ composites by introducing SiC whiskers was also described. In addition, the charged-particle irradiation ($Si^{2+}$ and $H^{+}$ ion) into CVD SiC was carried out to simulate the severe environments of the advanced nuclear reactors. SiC whiskers grown in the fiber preform increased the matrix infiltration rate by more than $60\%$ compared to the conventional CVI process. The highly crystalline and pure SiC showed little degradation in hardness and elastic modulus up to a damage level of 10 dpa at $1000^{\circ}C$.