• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.981-988
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• 2018

An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.919-924
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• 2005

Fretting-wear caused by turbulence excitation for KSNP(Korea standard nuclear power plant) steam generator is investigated numerically. Secondary sides density and normal velocity are obtained by the thermal-hydraulic data of the steam generator. Because nonlinear finite element analysis is complex and time consuming, work rate is estimated by using linear analysis for simple straight 2-span tube. Wear volume and depth by using work rate calculation are estimated. Span length, secondary side fluid density and normal velocity are adopted to study the effects on the fretting-wear by turbulence excitation. When secondary sides density and normal velocity is increased, It turns out that secondary side density and normal gap velocity are very important paramater for fretting-wear phenomena of the steam generator.

• Proceedings of the KWS Conference
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• 2006.10a
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• pp.26-28
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• 2006

When materials such as metal are into the secondary side of steam generator, they, so called foreign objects, may have influences on the integrity of the steam generator tubes. They cause the tube wear due to the relative motion between the tubes and foreign objects and the tube impact due to flow. The best way to avoid the effects is to remove all the foreign objects. However, it is not easy to remove the foreign materials thoroughly due to their condition such as the location. Considering the wear and impact by the foreign materials, KEPRI(Korea Electric Power Research Institute) developed the methodology to evaluate the foreign materials analytically. This methodology was described with a computer program in order to obtain the fast results. The program informs whether the tubes have the structural integrity when the foreign material strikes the tubes. Moreover, this gives us the remaining life of the steam generator tubes. In this paper, the program, which evaluates the effects of the foreign objects in the secondary side of steam generator, is introduced.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.4
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• pp.2043-2048
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• 2013

OPR-1000 CE Steam Generator, of which tube material is composed of Alloy-600 HTMA in nuclear power plant, secondary side is generated ODSCC(Outside Diameter Stress Corrosion Cracking) due to the accumulated sludge. ODSCC is centered around the tube sheet and is being affected depending on the height of the sludge. Chemical cleaning was carried out for a top-of-the-tube sheet(TTS) of Steam Generator in order to decrease corrosive condition of the secondary side of Steam Generator tubes and suppress the occurrence of stress corrosion cracking. The amount of sludge removal was 259.2kg. The height of the accumulated sludge was reduced from 0.71 to 0.34 inches. Corrosion rate as the maximum 2.34 mils was satisfied to within EPRI (Electric Power Research Institute) recommendation(10 mils).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.96-99
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• 2014

The aim of the study was to evaluate the water treatment of pressurized water reactor secondary side by the mixed amine of ammonia and ethanolamine, from the standpoint of corrosion control, as compared with all volatile treatment of ammonia. The pressurized water reactor systems have switched a secondary side pH control agent to minimize the corrosion in the moisture separator/reheater and feedwater heater systems and the transport of corrosion products into steam generator. As results of field test, pH was increased in the steam generator and the wet steam area of moisture separator/reheater and the concentration of Fe were decreased by more than 50% as compared with water treatment of ammonia.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.1029-1040
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• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.841-848
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• 2009

A simple model to analyze non-linear density-wave instability in a sodium-cooled helically coiled steam generator is developed. The model is formulated with three regions with moving boundaries. The homogeneous equilibrium flow model is used for the two-phase region and the shell-side energy conservation is also considered for the heat flux variation in each region. The proposed model is applied to the analysis of two-phase instability in a JAEA (Japan Atomic Energy Agency) 50MWt No.2 steam generator. The steady state results show that the proposed model accurately predicts the six cases of operating temperatures on the primary and secondary sides. The sizes of three regions, the secondary side pressure drop according to the flow rate, and the temperature variation in the vertical direction are also predicted well. The temporal variations of the inlet flow rate according to the throttling coefficient, the boiling and superheating boundaries and the pressure drop in the two-phase and superheating regions are obtained from the unsteady analysis.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.870-880
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• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.05a
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• pp.138-143
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• 1994

A simplified steam generator model has been developed for the simulation of the operational transients of CANDU nuclear power plant. For the analysis of the secondary side, a control volume approach is used and the flow conservation equations are applied for each control volume. The typical steam generator control logic such as the level control and the pressure control are incorporated into the steam generator model with appropriate interface conditions. The steam line including ASDV, CSDV, and governor valve also has been modeled. Test results for typical operational transient case show reasonable transient behavior of steam generator in a real time basis, which is promising for a CANDU engineering simulator.