• 대한전기학회:학술대회논문집
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• 대한전기학회 1995년도 추계학술대회 논문집 학회본부
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• pp.172-174
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• 1995

About a half of Electric power is generated by nuclear power plants in korea. So, the stable operation of nuclear power plant is very important for suppling the essential national electric power. A S/G(Steam Generator) level control is the most difficult system in PWR(Pressurized Water Reactor) nuclear power plant. Because of the non-linear and the non-nominal response of S/G level control, it Is very difficult to control the level by automatic mode or manual mode. The goal of this study is to establish and verify a advanced control algorithm by analyzing, modelling, stability calculation, controller parameter calculation, simulation for S/G level control system.

• 한국재료학회지
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• 제1권1호
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• pp.29-36
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• 1991

Pressurized Water Reactor (PWR) pressure vessels are made of forged low alloy steel plates internally clad with an austenitic stainless steel by welding to improve anti-corrosion properties. They display a characteristic behavior of dissimilar metal weld interface during post weld heat treatment (PWHT) and service at high temperature and pressure. In this Study, Metallugical structure of weld interface of SA 508 Class 3 forged steel clad with 309L, Austenitic stainless steel after PWHT was investigated. To estimate the width of the carburized/decarburized bands quantitatively, a model for carbon diffusion was proposed and a theoretical equation was derived.

• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1993년도 추계학술대회논문집
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• pp.21-30
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• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1120-1125
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• 2004

Welding characteristics of STS304 stainless steel and Inconel 600 using a continuous wave Nd:YAG laser are experimentally investigated. Alloy 600 being used in steam generator tubing of pressurized water reactor(PWR) exposed to some corrosion environment, stress corrosion cracking can occur on this material. Presented here are the results from a series of experiments in which dissimilar metal welds were made using the gas tungsten arc welding process with pure argon shielding gas. But It is well known that solidification cracking susceptibility of austenitic alloys depends on the solidification temperature range and amount/distribution of solute rich liquid that exists at the terminal stages of solidification. An experimental study was conducted to determine effects of welding parameters and to optimize those parameters that have the most influence on eliminating or reducing the extent welding zone formation at dissimilar metal welds.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1135-1142
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• 2017

The applicability of the algebraic collapsing acceleration (ACA) technique to the method of characteristics (MOC) in cases with scattering anisotropy and/or linear sources was investigated. Previously, the ACA was proven successful in cases with isotropic scattering and uniform (step) sources. A presentation is first made of the MOC implementation, available in the DRAGON5 code. Two categories of schemes are available for integrating the propagation equations: (1) the first category is based on exact integration and leads to the classical step characteristics (SC) and linear discontinuous characteristics (LDC) schemes and (2) the second category leads to diamond differencing schemes of various orders in space. The acceleration of these MOC schemes using a combination of the generalized minimal residual [GMRES(m)] method preconditioned with the ACA technique was focused on. Numerical results are provided for a two-dimensional (2D) eight-symmetry pressurized water reactor (PWR) assembly mockup in the context of the DRAGON5 code.

• 한국정밀공학회지
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• 제21권11호
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• pp.53-60
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• 2004

Laser welding techniques have been characterised for various materials. In this paper, the laser weldability of STS304 stainless steel and Inconel 600 at dissimilar metal welds using a continuous wave Nd:YAG laser are experimentally investigated. Inconel 600 is being used in a steam generator tubing of pressurized water reactor(PWR) exposed to some corrosion. Stress corrosion cracking can occur on this material. An experimental study was conducted to determine effects of welding parameters, on eliminating or reducing the extent welding zone formation at dissimilar metal welds and to optimize those parameters that have the most influence parameters such as focus length, power, beam speed, shielding gas, and wave length of laser were tested.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2874-2879
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• 2020

Due to excess reactivity, fresh nuclear fuel often contains burnable poisons. This research looks at six different burnable poisons and their impacts on reactivity, material attractiveness, and waste management. An MCNP simulation of a PWR fuel pin was performed with a fuel burnup of 60 GWd/MTHM to determine when each burnable poison fuel type would decrease below a k_∞ of 1. For determining the plutonium material attractiveness in each burnable poison fuel type, the plutonium isotopic content of the used fuel was evaluated using Bathke's Figure of Merit formula. For the waste management analysis, the thermal output of each burnable poison fuel type was determined through ORIGEN decay simulations at 100 and 300 years after being discharged from the core. The performance of all six burnable poisons varied over the three criteria considered and no single burnable poison performed best in all three considerations.

• Journal of Mechanical Science and Technology
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• 제15권3호
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• pp.327-338
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• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell, and lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and high swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear structural analysis. Canister types studied hear are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.73-82
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• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3665-3674
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• 2021

One of the important parts of the in-core fuel management is loading pattern optimization (LPO). The loading pattern optimization as a reasonable design of the in-core fuel management can improve both economic and safe aspects of the nuclear reactor. This work proposes the hybrid of fuzzy logic controller with harmony search algorithm (HS) for loading pattern optimization in a pressurized water reactor. The music improvisation process to find a pleasing harmony is inspiring the harmony search algorithm. In this work, the adjustment of the harmony search algorithm parameters such as the bandwidth and the pitch adjustment rate are increasing performance of the proposed algorithm which is done through a fuzzy logic controller. Hence, membership functions and fuzzy rules are designed to improve the performance of the HS algorithm and achieve optimal results. The objective of the method is finding an optimum core arrangement according to safety and economic aspects such as reduction of power peaking factor (PPF) and increase of effective multiplication factor (Keff). The proposed approach effectiveness has been tried in two cases, Michalewicz's bivariate function problem and NEACRP LWR core. The results show that by using fuzzy harmony search algorithm the value of the fitness function is improved by 15.35%. Finally, with regard to the new solutions proposed in this research it could be used as a trustworthy method for other optimization issues of engineering field.