• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.918-924
• /
• 2020

Film boiling is of great importance in nuclear safety as it directly influences the integrity of nuclear fuel in case of accidents involving loss of coolant. Recently, nuclear power plant safety under earthquake conditions has received much attention. However, to the best of our knowledge, there are no existing studies reporting film boiling in an oscillating system. Most previous studies for film boiling were performed on stationary systems. In this study, numerical simulations were performed for saturated film boiling of water on a horizontal surface under low frequencies to investigate the effect of system oscillation on film boiling heat transfer. A coupled level-set and volume-of-fluid method was used to track the interface between the vapor and liquid phases. With a fixed oscillation amplitude, overall, heat transfer decreases with oscillation frequency. However, there is a frequency region in which heat transfer remains nearly constant. This lock-on phenomenon occurs when the oscillation frequency is near the natural bubble release frequency. With a fixed oscillation frequency, heat transfer decreases with oscillation amplitude. With a fixed maximum amplitude of the additional gravity, heat transfer is affected little by the combination of oscillation amplitude and frequency.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.18-24
• /
• 2018

This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor ($k_{eff}$) versus burnup (B) are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B)-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE)14 $10{\times}10$ boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC)-68 storage cask. The results revealed that the curves of $k_{eff}$ versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of $k_{eff}$, ${\Delta}k$) in some compound effects was not a summation of the all ${\Delta}k$ resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of $k_{eff}$ versus B for both single and compound effects.

• Nuclear Engineering and Technology
• /
• 제48권6호
• /
• pp.1338-1348
• /
• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

• 대한기계학회논문집B
• /
• 제26권3호
• /
• pp.498-506
• /
• 2002

Improvements of fuel properties have become essential for exhaust emission reduction as well as for optimization of directly-related design factors and exhaust gas aftertreatment. In this study, the potential possibility of oxygenated fuel such as ethylene glycol mono-n-butyl ether(EGBE) was investigated for the sake of smoke reduction from diesel engine. Because EGBE include oxygen content approximately 27%, it is a kind of effective oxygenated fuel that the smoke emission of EGBE is reduced remarkably in comparison with commercial diesel fuel, that is, it can supply oxygen component sufficient at high load and speed in diesel engine. And, it was tried to analyze the quantities of the low and high boiling point hydrocarbon among the exhaust emissions in diesel engine. It have been investigated by the quantitative analysis of the hydrocarbon $C_1$~ $C_{6}$ using the gas chromatography. This study was carried out by comparing the chromatogram with diesel fuel and diesel fuel blended EGBE 20vo1-%. The results of this study show that the hydrocarbon $C_1$~C$_{6}$ among the exhaust emission of the mixed fuels are exhausted lower than those of the diesel fuel at the all load and speed. In particular, high boiling point hydrocarbons such as $C_{5}$ and $C_{6}$ were reduced remarkably at high speed and load.d.

• Nuclear Engineering and Technology
• /
• 제25권1호
• /
• pp.8-19
• /
• 1993

강제순환 냉각상실사고시 조밀화된 저장계통의 사용후 핵연료에서 생성된 붕괴열의 제거를 확인하기 위한 자연순환 해석모델이 개발되었다 채택된 수치기법은 ADI방법에 근거하였다. 사용후 핵연료의 붕괴열 생성율은 ANS-79 붕괴열 모델에 따라 계산되었으며, 보수적인 붕괴열 생성량 입력을 위해 chopped sine곡선에 따른 비균일 표면열속이 가정되었다. 저장조내 국부비등의 발생 가능성을 조사하기 위해서 민감도분석이 수행되었으며, 이는 핵연료간 거리 비, 열 생성량 및 핵연료 봉 반지름 등의 여러 변수를 변경시킴으로서 이루어졌다. 이 모델의 적용결과는 적절한 냉각시간 후의 조밀화된 사용후 핵연료 다발을 통한 자연대류 유량이 안전하고 효과적인 방식으로 저장조의 온도준위를 조절할 수 있음을 보여주고 있으며, 또한 사용후 핵연료봉 재배치를 위한 냉각시간에 관한 허용기준이 얻어졌다.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1602-1611
• /
• 2021

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

• 한국분무공학회지
• /
• 제5권1호
• /
• pp.30-40
• /
• 2000

The characteristics of liquefied butane spray are expected to be different from conventional diesel fuel spray, because a kind of flash boiling spray is expected when the back pressure is below the saturated vapor pressure of the butane(0.23MPa at 298K). The ambient pressure was held at a pressure above(0.37MPa) and below(0.15MPa) the fuel vapor pressure. The axial velocities, radial velocities, and size distributions in butane sprays were measured with PDPA(Phase Doppler Particle Analyzer) system. The PDPA measurement showed a smaller SMD at the 0.15MPa chamber pressure, compared to the 0.37MPa case. Log-hyperbolic density function for the droplets size distribution can be fitted to the experimental results of a liquefied butane spray.

• 한국자동차공학회논문집
• /
• 제11권5호
• /
• pp.22-28
• /
• 2003

Flash boiling mechanism in the injector interferes with fine fuel metering in a liquid phase LPG injection engine. This study presents a mathematical model to precisely predict an injection quantity. A calibration procedure of injection quantity, which is very prompt and precise in measuring, is developed using a gas analyzer. According to this procedure, injection quantity can be obtained under various fuel compositions, temperatures and injection pressures. The release pressure of liquid phase LPG is estimated based on these experimental data. Although the release pressure is much lower than the saturation pressure, it is linearly proportional to the saturation pressure.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.627-638
• /
• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Nuclear Engineering and Technology
• /
• 제52권11호
• /
• pp.2415-2421
• /
• 2020

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.