• Nuclear Engineering and Technology
• /
• 제35권3호
• /
• pp.251-259
• /
• 2003

The Taguchi method was applied to investigate the effect of design factors on the performance of the shutdown cooling heat exchanger in the SMART-P. This method provided the simulation matrix for the KDESCENT program and an efficient tool for analyzing the simulation results. Levels of the design factors were selected by the effectiveness-NTU method. From 18 runs with the KDESCENT program, it was found that the performance of the system was greatly influenced by the inlet temperature at the shell side and the mass flow rate of the reactor coolant at the tube side. After applying the Taguchi method, we identified the important design factor that should be controlled and designed carefully. This method provides an efficient way to estimate the influence of each design factor on a system performance.

• 대한기계학회논문집B
• /
• 제32권5호
• /
• pp.400-405
• /
• 2008

Air-operated valve is one of principal valves that are used to control fluid flow in nuclear power plants. It is necessary to evaluate periodically the performance of the AOV(Air-Operated Valve) which is used for controlling flow and pressure in nuclear power plants. In this study, we developed the integrated program HAES including hardware system to evaluate the performance of AOV. HAES(HANVIT AOV Evaluation System) consists of many modules such as valve information, thrust and torque analysis, actuator analysis, weak analysis and margin analysis. We applied with the test valve and showed that results evaluated by HAES is similar to those offered in the plant.

• Nuclear Engineering and Technology
• /
• 제40권2호
• /
• pp.107-116
• /
• 2008

The present paper describes the preliminary compilation, assessment and examination of the supercritical heat transfer(SCHT) database. The availability and reliability of the SCHT data are discussed. Similarities in thermodynamic supercritical properties and SCHT behaviour of water, $CO_{2}$ and R-134a have been examined and some tentative conclusions are made. Finally, the future experimental and analytical program at the University of Ottawa is described.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2007년도 제38회 하계학술대회
• /
• pp.612-613
• /
• 2007

The line impedance is important data that are applied in all analysis fields of electric power system such as power flow, fault current, stability and relay calculation etc. Usually, the impedance can be accurately calculated in case of overhead line. However, in case of power cables or combined transmission lines, the impedance can not be accurately calculated because cable systems have the sheath, grounding wires, and earth resistances. Therefore, if there is a fault in cable system, these terms will severely be caused many errors for calculating impedance. In this paper, the line impedance is measured in a power system of underground cables, and is analyzed by a generalized circuit analysis program, EMTP(Electromagnetic Transient Program), for comparison with the measured value. These analysis results are considered to become foundation of impedance calculation for underground cables.

• Nuclear Engineering and Technology
• /
• 제39권4호
• /
• pp.237-248
• /
• 2007

Sodium-cooled fast reactor is considered in France as a potential candidate for a prototype of 4th generation system to be built by 2020. A detailed working program has been launched recently to identify by 2012 the potential improvement tracks for later industrial development of these reactors. The goals for innovation are first identified: Progress of the safety with a special attention to severe accidents risk minimization and mitigation (defense in depth approach); Economic competitiveness of the system mainly by reducing the capital cost, the investment risks by enhancing in service inspection and repair capacities, and raising the availability; Sustainability with fissile material management while reducing the proliferation risk; capacity for long-lived waste transmutation.

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.3085-3099
• /
• 2021

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to numerically assess the damage and vibrations of NPP buildings subjected to aircrafts crash. In Part I of present paper, two shots of reduce-scaled model test of aircraft impact on NPP were conducted based on the large rocket sled loading test platform. In the present part, the numerical simulations of both scaled and prototype aircraft impact on NPP buildings are further performed by adopting the commercial program LS-DYNA. Firstly, the refined finite element (FE) models of both scaled aircraft and NPP models in Part I are established, and the model impact test is numerically simulated. The validities of the adopted numerical algorithm, constitutive model and the corresponding parameters are verified based on the experimental NPP model damages and accelerations. Then, the refined simulations of prototype A380 aircraft impact on a hypothetical NPP building are further carried out. It indicates that the NPP building can totally withstand the impact of A380 at a velocity of 150 m/s, while the accompanied intensive vibrations may still lead to different levels of damage on the nuclear related equipment. Referring to the guideline NEI07-13, a maximum acceleration contour is plotted and the shock damage propagation distances under aircraft impact are assessed, which indicates that the nuclear equipment located within 11.5 m from the impact point may endure malfunction. Finally, by respectively considering the rigid and deformable impacts mainly induced by aircraft engine and fuselage, an improved Riera function is proposed to predict the impact force of aircraft A380.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1436-1445
• /
• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.

• Nuclear Engineering and Technology
• /
• 제55권6호
• /
• pp.2215-2221
• /
• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.464-470
• /
• 1997

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks(LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code)computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanism of calandria vessel and containment. In addition, some insights for accident management program(AMP) are also given.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.611-616
• /
• 1997

The Level 1 probabilistic safety assessment(PSA) for Wolsong(WS) 2/3/4 nuclear power plant(NPPs) in design stage is performed using the methodologies being equivalent to PWR PSA. Accident sequence evaluation program(ASEP) human reliability analysis(HRA) procedure and technique for human error rate prediction(THERP) are used in HRA of WS 2/3/4 NPPs PSA. The purpose of this paper is to introduce the procedure and methodology of HRA in WS 2/3/4 NPPs PSA. Also, this paper describes the interim results of importance analysis for human actions modeled in WS 2/3/4 PSA and the findings and recommendations of administrative control of secondary control area from the view of human factors.