• Corrosion Science and Technology
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• 제18권4호
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• pp.138-147
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• 2019

Pipes and components of the secondary system in the pressurized water reactor (PWR) are mainly comprised of manufactured carbon steel. Thus, the generated carbon steel corrosion products are transported into the steam generator and deposited, thereby deteriorating the integrity of the steam generator. Environmental condition in the secondary system of the PWRs differs across different locations. So, the corrosion rate and types of corrosion products depend on specific locations in the secondary system. In this study, the quantity and chemical compositions of corrosion products generated in various locations that vary in different temperatures and chemistry conditions were investigated. As a result of evaluating the PWR "Unit A" that is in current operation, the amount of corrosion products generated in the section of high temperature feedwater system was identified as the largest source in the secondary system. Major components of corrosion products were iron oxides such as magnetite, hematite, and lepidocrocite.

• 방사성폐기물학회지
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• 제14권1호
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• pp.35-44
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• 2016

Leading national R&D project to design a PWR spent nuclear fuel interim dry storage system that has been under development since mid-2009, which consists of a dual purpose metal cask and concrete storage cask. To ensure the safe operation of dry storage systems in foreign countries, major confinement monitoring techniques currently consist of pressure and temperature measurement. In the case of a dual purpose metal cask, a pressure sensor is installed in the interspace of bolted double lid(primary and secondary lid) in order to measure pressure. A concrete storage cask is a canister based system made of double/redundant welded lid to ensure confinement integrity. For this reason, confinement monitoring method is real time temperature measurement by thermocouple placed in the air flow(air intake and exit) of the concrete structure(over pack and module). The use of various monitoring technologies and operating experiences for the interim dry storage system over the last decades in foreign countries were analyzed. On the basis of the analysis above, development of the confinement monitoring technology that can be used optimally in our system will be available in the near future.

• Nuclear Engineering and Technology
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• 제50권6호
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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.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.431-438
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• 2014

The effect of various pH agents on the corrosion behavior of carbon steel was investigated under a simulated secondary water chemistry condition of a pressurized water reactor (PWR) in a laboratory, and the steel's corrosion performance was compared with the field data obtained from Uljin NPP unit 2 reactor. All tests were carried out at temperatures of $50^{\circ}C-250^{\circ}C$and pH of 8.5 - 10. The pH at a given temperature was controlled by adding different agents. Laboratory data indicate that the corrosion rate of carbon steel decreased as the pH increased under the test conditions and the highest corrosion rate was measured at $150^{\circ}C$. This high corrosion rate may be related to high dissolution and instability of Fe oxide ($Fe_3O_4$) at $150^{\circ}C$. It was also found that an addition of ethanolamine (ETA) to ammonia was more effectivefor anticorrosion than ammonia alone, and that mixed treatment reduced 50% of iron or more at pHs of 9.5 or higher, especially in the steam generator (SG) and the moisture separator & re-heater (MSR).

• 한국전산유체공학회지
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• 제20권3호
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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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• 제51권4호
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• 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.

• Corrosion Science and Technology
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• 제10권5호
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• pp.175-179
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• 2011

The transportation and deposition of iron corrosion products are important elements that affect both the steam generator (SG) integrity and secondary system in pressurized water reactor (PWR) nuclear power plants. Most of iron corrosion products are generated on carbon steel materials due to flow accelerated corrosion (FAC). The several parameters like water chemistry, temperature, hydrodynamic, and steel composition affect FAC. It is well established that the at-temperature pH of the deaerated water system has a first order effect on the FAC rate of carbon steels through nuclear industry researches. In order to reduce transportation and deposition of iron corrosion products, increasing pH(t) tests were applied on secondary system of A, B units. Increasing pH(t) successfully reduced flow accelerated corrosion. The effect of increasing pH(t) to inhibit FAC was identified through the experiment and pH(t) evaluation in this paper.

• Corrosion Science and Technology
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• 제12권5호
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• pp.227-232
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• 2013

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1173-1183
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• 2018

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.