• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.39-47
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• 2002

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock(PTS). The PTS event means an event or transient in pressurized water reactors(PWRs) causing severe overcooling(thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Environmental Engineering Research
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• v.16 no.2
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• pp.97-102
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• 2011

This study investigates the treatment of pigment wastewater using a sequential batch reactor (SBR) followed by dissolved ozone flotation-pressurized ozone oxidation treatement (DOF-$PO_2$). The process efficiency has been evaluated at the lab scale on the basis of water quality parameters. In addition, the effect of pure oxygen and air was investigated on the removal of COD, BOD, and TN in the SBR process. It was observed that under comparable conditions the removal efficiencies of these water quality parameters using pure oxygen and air were similar. The effect of the recycle rate was also investigated for its impact on the water quality parameters using different ozone dissolving pressures in a DOF process in order to optimise conditions. The results conclude that the use of an SBR and ozone contact by DOF-$PO_2$ is a highly effective treatment for pigment wastewater and aids in the achievement of effluent discharge criteria.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• Proceedings of the Society of Korea Industrial and System Engineering Conference
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• 2002.05a
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• pp.25-30
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• 2002

The problem of qualitative reliability system is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, extensive environment destruction, and fatal damage of human. Therefore this study is to develop the reliability evaluation model through the normalized scoring model by the quantitative and qualitative factors considering the advanced safety factors In the Advanced Pressurized water Reactor 1400MWe(APR 1400) under uncertainty Especially, the qualitative factors considering the information and human factors for the systematic and rational justification have been closely analyzed. The reliability evaluation model can be simply applied in real fields in order to minimize the industrial accident and human error in the digitalized nuclear power plant.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.474-477
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• 2002

Safety is very important to operate nuclear power plant. To guarantee the safety, nuclear power plant should be run without trouble. This paper presents the application of a failure diagnosis approach based on discrete event system theory to the pressurizer pressure control system for Pressurized Water Reactor. Also, this paper shows a scheme of failure diagnosis by distributed diagnoser.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2003.05a
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• pp.327-330
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• 2003

In this paper, a viscoelastic structural analysis for the spent pressurized water reactor(PWR) nuclear fuel disposal canister is carried out to predict the collapse of the canister while the canister is stored in a deep repository for long time. There may exist some subterranean heat in a deep repository while the nuclear fuel disposal canister is stored for long time. Then, a time-dependent viscoelastic structural deformation may occur in the canister due to the subterrnean heat Hence, the viscoelastic stress variation according to time should be computed to predict the structural strength of the canister. A viscoelastic material model is adopted. Analysis results show that even though some subterrnean heat may exist for quite a long time, the canister structure still endures stresses below the yield strength of the canister. Hence, some subterranean heat cannot seriously affect the structural strength of the canister.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.49-54
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• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics