• 한국태양에너지학회 논문집
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• 제28권6호
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• pp.33-39
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• 2008

The purpose of this study is not only to evaluate the stress behavior of tubes in superheater in HRSG during the load change operation but also to find root causes of failure from stress behavior. Firstly, temperature during operation was collected to perform stress analysis from actual HRSG. Part load and full load stress analysis which can be represented as the whole load change operations were performed using commercial finite element software. The possibility that can lead to tubes failure is found by stress analysis and its results is compared with metallurgical mircrostructure of failed tube which was taken from actual HRSG.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 하계학술대회 논문집 G
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• pp.2926-2928
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• 1999

원자력발전소 증기발생기 전열관 건전성 유지를 위하여 매 주기마다 실시하고 있는 비파괴검사의 일종인 와전류검사(ECT, Eddy Current Testing)의 주요 공정은 크게 3가지로 분류할 수 있다. 첫 번째는 전열관 상태검사를 위한 신호데이터 취득공정이고, 두 번째는 취득된 신효를 판독하여 전열관의 건전성 여부를 진단하는 평가공정, 세 번째는 평가공정에서 발생하는 데이터를 토대로 전열관 이력 및 상태를 유지관리하는 공정으로 구분할 수 있다. 본 논문에서는 위의 세 번째 공정결과 생성되는 전열관 이력 및 상태자료를 데이터베이스화하여 유지 관리하고, 데이터베이스화된 내용을 바탕으로 전열관 상태 변화추이를 파악하는 기능, 현재까지 비 체계화된 모든 전열관의 이력자료를 다양한 보고서 형태로 출력할 수 있는 기능 둥을 제공하기 위한 "인트라넷 증기발생기 전열관 이력관리시스템"의 설계 및 구현과정 을 정 리 하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기기기 및 에너지변환시스템부문
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• pp.50-52
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• 2005

In this paper, the analysis of the eddy current testing(ECT) signals under thc Influence of the ferromagnetic support plate was performed in steam generator(SG) tube of nuclear power plant. In order to remove the influence of the ferromagnetic support plate, a multi-frequency ECT was used. The models which was established for the analysis of the signals is calculated using numerical analysis of finite element method. Through the result of numerical analysis, improved signals is acquired considering the influence of the ferromagnetic support plate using mixing of multi-frequency This paper is presented the residual errors and the phase changes for analysis of the defect signals which should be considered when conducting a ECT using multi-frequency.

• 한국표면공학회지
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• 제34권4호
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• pp.289-296
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• 2001

Ni-P electroformed layers were investigated for developing a steam generator tube repair technology in PWRs. The effects of an additive, RPP (Reagent over Pitting Protection) and agitation on mechanical properties and microstructure of the layer were evaluated. The addition of the RPP showed to inhibit the formation of pores, to refine the grain size, and to increase the residual stress in the layer. However, the agitation of the solution during electroforming was observed to increase pores in local regions of the electroformed layer, resulting in decreasing its mechanical properties. The heat treatment of the layer at $343^{\circ}C$ for 1 hr. precipitated the very fine particles of Ni3P in the layer, which inhibited grain growth and increased microhardness.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.8-17
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• 2019

Perforated plates are used for the steam generator tube-sheet and the Reactor Vessel Closure Head in the Nuclear Power Plant. The ASME code, Section III Appendix A-8000, addresses the analysis of perforated plates, however, this analysis is only limited to the flat plate with a triangular perforation pattern. Based on the concept of the effective elastic constants, simulation of flat and spherical perforated plates and their equivalent solid plates were carried out using Finite Element Analysis (FEA). The isotropic material properties of the perforated plate were replaced with anisotropic material properties of the equivalent solid plate and subjected to the same loading conditions. The generated curves of effective elastic constants vs ligament efficiency for the flat perforated plate were in agreement with the design curve provided by ASME code. With this result, a plate with spherical curvature having perforations can be conveniently analyzed with equivalent elastic modulus and equivalent Poisson's ratio.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Journal of Radiation Protection and Research
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• 제35권2호
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• pp.57-62
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• 2010

원전 계획예방정비기간 증기발생기 수실작업, 가압기 전열관교체 또는 압력관피더 제거작업 지역 등은 높은 방사선량률을 보이는 지역으로, 짧은 시간 동안의 작업으로 작업종사자는 높은 피폭을 받을 가능성이 있다. 특히, 방사성물질과 접촉하는 손 부위는 높은 피폭이 일어날 수 있다. 이런 점을 고려하여 국내 가압경수로원전과 가압중수로원전의 계획 예방정비기간 중 증기발생기 수실 노즐댐 설치와 제거작업, 원자로 냉각재펌프 보수작업, 원자로헤드 보수 및 검사작업 등과 같은 고피폭 접촉작업에서 방사선작업종사자의 말단선량을 측정하기위한 현장시험을 실시하였다. 여기에 참여한 작업종사자는 가슴과 등 부위에 일상적인 절차에 따른 복수선량계를 패용한 것 이외에 손목에 개인선량계를 추가로 패용하였고, 손가락 부위에는 말단선량계 (Extremity dosimeter)를 패용하였다. 그 결과, 손가락이 받는 등가선량은 각각 손목이 받는 등가선량 및 가슴 또는 등 부위가 받는 등가선량과 일정한 비율로 평가됨을 확인하였다.

• 비파괴검사학회지
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• 제20권1호
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• pp.1-9
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• 2000

본 연구에서는 유도초음파를 이용하여 열 교환기와 증기발생기 튜브의 결함을 비파괴적으로 탐상하고 그 크기를 산정하였다. 이론적인 해석을 위해 인코넬 (Inconel) 튜브에 대한 위상 및 군속도 분산선도를 Longitudinal 모드와 Flexural 모드에 대해 구하였다. 튜브의 원주방향 레이저노치와 튜브 지지대 하단의 방전가공결함(EDM wear)을 각각 비대칭 및 대칭 탐촉자 세트를 사용하여 탐상하였다. 실험결과 방전가공결함은 L(0, 2), L(0, 3), L(0, 4) 모드로 탐상되었으며, 그 중 L(0, 4) 모드가 결함으로부터 가장 잘 반사되었다. 레이저노치의 경우에는 L(0, 1) 모드 주변의 Flexural 모드가 결함을 탐상하고 크기를 산정하는데 사용될 수 있음을 보였다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제34회 추계학술대회 개최
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• pp.296-303
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• 2001

In nuclear power steam generators, high flow rates can induce vibration of the tubes resulting in fretting wear damage due to contacts between the tubes and their supports. In this paper the fretting wear tests and the sliding wear tests were performed using the steam generator tube materials of Inconel 690 against STS 304. Sliding tests with the pin-on-disk type tribometer were done under various applied loads and sliding speeds at air and water environment. Fretting tests were done under various vibrating amplitudes, applied normal loads and various temperatures. From the results of sliding and fretting wear tests, the wear of Inconel 690 can be predictable using the work rate model. Depending on normal loads and vibrating amplitudes, distinctively different wear mechanisms and often drastically different wear rates can occur. At room temperature, the wear coefficient K of Inconel 690 is 7.57${\times}$10$\^$13/Pa$\^$1/ in air and it is 1.93${\times}$10$\^$13/Pa$\^$1/ in water. At room temperature, it is found that the wear volume in air is more than in water. In water, the wear coefficient K at 50$^{\circ}C$ and 80$^{\circ}C$ is 4.35${\times}$10$\^$-13/Pa$^1$ and 5.81${\times}$10$\^$-13/Pa$^1$ respectively, Therefore, it is found that the wear volume extremely increases by increasing on temperature in water. This study shows that the dissolved oxygen with temperature increment increases and the wear due to fluidity is severe.