• Nuclear Engineering and Technology
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• 제48권1호
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2600-2609
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• 2021

Low cycle fatigue behaviors of SA508-3 low-alloy steel were investigated in room-temperature air, high-temperature air and in light water reactor (LWR) water environments. The fatigue mean curve and design curve for the low-alloy steel are developed based on the fatigue data in room-temperature and high-temperature air. The environmental fatigue model for low-alloy steel is developed by the environmental fatigue correction factor (F_en) methodology based on the fatigue data in LWR water environments with the consideration of effects of strain rate, temperature, and dissolved oxygen concentration on the fatigue life.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Environmental Engineering Research
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• 제23권2호
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• pp.216-222
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• 2018

The growth characteristics of Spirulina platensis were investigated using four photo-bioreactors with $CO_2$-mixed air flows. Each reactor was operated under a specific condition: 3% $CO_2$ at 50 mL/min, 3% $CO_2$ at 150 mL/min, 6% $CO_2$ at 50 mL/min, and 6% CO2 at 150 mL/min. The 3% $CO_2$ at 150 mL/min condition produced the highest algal growth rate, while the 6% $CO_2$ at 150 mL/min conditioned produced the lowest. The algal growth performance was suitably assessed by the linear growth curve rather than the exponential growth. The medium pH decreased from 9.5 to 8.7-8.8 (3% $CO_2$) and 8.4-8.5 (6% $CO_2$), of which trends were predicted only by the pH-carbonate equilibrium and the reaction kinetics between dissolved $CO_2$ and $HCO_3{^-}$. Based on the stoichiometry between the nutrient amounts and cell elements, it was predicted that depleted nitrogen (N) at the early stage of the cultivation would reduce the algal growth rates due to nutrient starvation. In this study, use of the photobioreactors capable of good light energy distribution, proper ranges of $CO_2$ in bubbles and medium pH facilitated production of high amounts of algal biomass despite N limitation.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1109-1114
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• 2009

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) $UO_2$ fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the $UO_2$ matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1990년도 추계학술대회논문집; 한양대학교, 서울; 24 Nov. 1990
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• pp.113-124
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• 1990

Tap lines are small branch piping generally less than two inches in diameter. They typically branch off of header piping having a much larger diameter. An example of a common tap line is a 3/4 inch size high point vent or low point drain. Most tap lines have at least one valve near the header tap connection to provide isolation. Two valves are often required for double isolation. A light water reactor(LWR) nuclear power plant will have several hundred tap lines. These lines come in many sizes and shapes and serve numerous functions. A single process piping valve may have three different tap lines associated with it (figure 1). Table 1 delineates the different categories of tap lines. Vibration failures of tap lines are a common occurrence in all industrial plants including nuclear and fossil power plants. These types of failures constitute a significant percentage of all piping related failures. An unscheduled plant shutdown or outage resulting from the failure of a tap line decreases plant reliability and may have a detrimental effect on plant safety. Most tap line vibration failures can be avoided through the use of appropriate routing and support techniques. Standardized designs can be developed for use in a myriad of applications. These designs will not only minimize failures but will also reduce the necessary analysis and installation efforts.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.547-555
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• 2001

In pressurized light water reactor fuel assembly, spacer grids support nuclear fuel rods both laterally and vertically. The fuel rods are supported by spacer grid springs and grid dimples that are located in the grid cell. The support system allows for some thermal expansion and imbalance of the fuel rods. The imbalance is absorbed by elastic energy to prevent coolant flow- induced vibration damage. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. Since the optimization is carried out in the linear range of finite element analysis, the optimum solution is verified by nonlinear analysis. A good design is found and the final design is compared with the initial conceptual design. Commercial codes are utilized for structural analysis and optimization.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.336-347
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• 1997

During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe + $Al_2$O$_3$) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/$m^2$. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.238-247
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• 2020

The Burnable Poison (BP) is very important for all Light Water Reactors in order to hold-down the initial excess reactivity and to control power peaking. The use of BP is even more essential as the excess reactivity increases significantly with a longer operation cycle. In this paper a feasibility study was conducted in order to investigate the benefits of a new combinational BP concept designed for 24-month cycle PWR core. The reference designs in this study are based on the two Korean fuel assemblies; 17 × 17 Westinghouse (WH) design and 16 × 16 Combustion Engineering (CE) design. A modification was done on these two designs to extend their cycle length from 18 months into 24 months. DeCART2D-MASTER code system was used to perform assembly and core calculations for both designs. A preliminary test was conducted in order to choose the best BP suitable for 24-month as a representative for single BP concept. The comparison between the results of two concepts (combinational BP concept and single BP concept) showed that the combinational BP concept can replace the single BP concept with better performance on holding down the initial excess reactivity without violating the design limitations.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.38-47
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• 1986

경수로심의 열 여유도를 분석하기 위하여 반응표면 및 Monte Carlo 방법을 이용하는 통계적 분석 방법이 제시되었다. 통계적인 열 여유도 분석 방법은 입력변수들의 불확실도를 확률론적으로 처리함으로써 열 여유도의 최적 평가를 수행한다. 이 방법은 원자력 1호기 정상상태의 원자로심 분석에 응용되었으며 또한 종래의 결정론적 방법 및 웨스팅하우스의 개선된 열설계 방법과도 비교되었다. 본 연구를 통하여 반응표면 분석 방법은 통계적인 열 여유도 분석에 유용함을 알 수 있었으며, 이 방법을 통한 열 여유도의 증가도 확인되었다.