• 한국소음진동공학회논문집
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• 제16권9호
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• pp.904-911
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• 2006

A modified energy method for the fretting wear of the steam generator tube is proposed to calculate the wear-out depth between the nuclear steam generator tube and its support. Estimation of fretting-wear damage typically requires a non-linear dynamic analysis with the information of the gap velocity and the flow density around the tube. This analysis is very complex and time consuming. The basic concept of the energy method is that the volume wear rate due to the fretting-wear phenomena Is related to work rate which is time rate of the product of normal contact force and sliding distance. The wearing motion is due to dynamic interaction between vibrating tube and its support structure, such as tube support plate and anti-vibration bar. It can be assumed that the absorbed work rate would come from turbulent flow energy around the vibrating tube. This study also numerically obtains the wear-out depth with various wear topologies. A new dissection method is applied to the multi-span tubes to represent the vibrational mode. It turns out that both the secondary side density and the normal gap velocity are important parameters for the fretting-wear phenomena of the steam generator tube.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 1999년도 춘계학술대회 논문집
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• pp.163-169
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• 1999

Residual stresses causing stress corrosion cracking (SCC) of thin-walled steam generator U tubes were investigated. The residual stresses were measured by hole drilling methods, and the applied stresses resulting from the internal pressure and the temperature gradient in the steam generator were estimated theoretically. In U-bent regions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319MPa in axial direction at $\phi$= $0^{\circ}$ in position. Maximum tensile residual stress of 170MPa was found to be at the flank side at position of $\phi$= $90^{\circ}$, i.e., at apex region. Hoop stress due to the pressure and temperature differences between primary and secondary side were analyzed to be 76 MPa and 45 MPa, respectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.135.2-135
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• 1999

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2008년도 춘계학술발표대회 개요집
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• pp.10-10
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• 2008

• 한국수소및신에너지학회논문집
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• 제22권4호
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• pp.546-551
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• 2011

The corrosion on steam generator tubes of the secondary side of pressurized water reactor inhibits heat transfer. One of the most efficient techniques improving the heat transfer performance of a nuclear electric generation is a corrosion control. The environmental parameters mostly affecting corrosion are materials and chemical additives. It seems that no further corrosion occurs in steels with Polyacrylic acid polymer dispersant treatment. Polyacrylic acid forms a protective coating with uniform thickness on metal surface. Polyacrylic treatment appears to be the most convenient way to enhance the thermal performance by the thermal conductivity improvement in steam generators.

• 비파괴검사학회지
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• 제36권4호
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• pp.295-299
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• 2016

증기발생기 2차측(관판 상단 및 관 지지판)에 잔류하는 이물질은 전열관에 마모를 야기하여 누설을 발생시킬 가능성이 있으며, 마모는 전열관 재료의 부식 특성과는 별개로 이물질의 재료, 형태, 크기 등에 따라 큰 영향을 받는다. 관판 상단 및 관 지지판 부위의 이물질 여부는 원격육안검사(FOSAR) 방법과 와전류탐상검사(ECT)를 통해 확인하고 있다. 증기발생기 2차측의 잔류 이물질은 그 재질이나 전열관과의 접촉상태 등에 따라 검출하는데 제한성이 나타난다. 따라서, 본 연구에서는 다양한 형태와 재료의 이물질을 관판 상단에서 수직 및 수평한 방향으로 이동시켰을 때, 이물질에 대한 보빈신호의 검출 감도를 측정하기 위해 데이터를 수집하고 분석하였다.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.61-66
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• 2011

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.812-824
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• 2021

In order to reduce risks from the Steam Generator Tube Rupture (SGTR) accident and to meet safety targets, various measures have been analyzed to minimize the amount of fission product (FP) release. In this paper, we propose an introduction of a Containment Filtered Venting System (CFVS) connected to the steam generator secondary side, which can reduce the amount of FP release while minimizing adverse effects identified in the previous studies. In order to compare the effect of new equipment with the existing strategy, accident simulations using MELCOR were performed. As a result of simulations, it is confirmed that CFVS operation lowers FP release into the environment, and the release fractions are lower (minimum 0.6% of the initial inventory for Cs) than that of the strategy which intends to depressurize the primary system directly (minimum 15.2% for Cs). The sensitivity analyses identify that refill of the CFVS vessel is a dominant contributor reducing the amount of FP released. As the new strategy has the possibility of hydrogen combustion and detonation in CFVS, the installation of an igniter inside the CFVS vessel may be considered in reducing such hydrogen risk.