• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.2
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• pp.167-175
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• 2014

If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.838-848
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• 2015

One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented.SMCis ameansto control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.5 no.3
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• pp.21-29
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• 1985

An effective technique of sludge separation is required for concentrated activated sludge process. The dissolved-air flotation (DAF) has been shown to be efficient process for sludge separation. The factors affecting DAF process for activated sludge separation are type and concentration of sludge, air/solid ratio, ratio of pressurized water flow, pressure, sludge detention time, temperature, sludge and hydraulic loading rate, recycle flow rate of sludge and type and quantity of chemical aid. In order to study the optimal operation condition for sludge separation, the influence factors such as type and concentration of sludge, ratio of pressurized water flow and pressure are investigated by the batch and continuous reactor experiments of DAF and sedimentation test. By the experimental investigation, the results are as follows; 1. For the bulking and concentrated sludge, DAF is more effective than sedimentation for the sludge separation. 2. In DAF, the critical ratio of pressurized water flow exist. The critical value varies with the pressure in the tank. That is, according to the pressure changes from 3 to $5kg/cm^2$, the critical value varies from 0.25 to 0.67 accordingly. 3. Pressure affects the ratio of pressurized water flow, but it does not show any influence upon the DAF efficiency directly. 4. Continuous experimental results was not better than those of batch.

• Progress in Medical Physics
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• v.5 no.1
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• pp.75-85
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• 1994

As we enter the 2000s, there are four nuclear power units of the pressurized heavy water reactor-type in the commercial operation at the Wolsung Nuclear Power Plant(NPP) site where a larger amount of tritium ($\^$3/H) is released inevitably to the site environment. This radioctive nuclide is easily distributed throghout our environment because of its ubiquitous form as tritiated water (HTO) and its persistence in the environment. Tritum has certain characterisitics that present unique challenges for beta radiation dosimety and health risk assesment. In this paper, therefore, a variety of matters on tritium are considered and reviewed in terms of its characteristics and sources, metabolism and dosimetry, microdosimetry, radiobiology, risk assessment, and transport and cycling in the environment, etc.

• 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.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.668-674
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• 2017

Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit) would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS). To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR) cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ~2 and similar crystal morphology as the one (bonaccordite) reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of $5mL\;H_2/kg\;H_2O$, but absent when exposed under $75mL\;H_2/kg\;H_2O$ condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2977-2981
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• 2021

Corrosion fatigue crack growth (FCG) behavior of 316LN stainless steel was investigated in high-temperature pressurized water at different temperatures, load ratios (R = K_max/K_min) and rise times (t_R). The environmental assisted effect on FCG rate was observed when both the R and t_R exceeded their critical values. The FCG rate showed a linear relation with stress intensity factor range (ΔK) in double logarithmic coordinate. The environmental assisted effect on FCG rate depended on the ΔK and quantitative relations were proposed. Possible mechanisms of environmental assisted FCG rate under different testing conditions are also discussed.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1765-1775
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• 2019

Multiple spurious operations (MSOs) mean multiple fire induced circuit faults causing an undesired operation of one or more systems or components. The Nuclear Energy Institute (NEI) of the United States published NEI 00-01 as guidelines for solving MSOs. And this guideline includes MSO scenarios of pressurized water reactor (PWR) and boiling water reactor (BWR). Nuclear power plant operators in U.S. analyzed MSOs under MSO scenarios included in NEI 00-01 and operators of PWRs in Korea also analyzed MSOs under the scenarios of NEI 00-01. As there are no pressurized heavy water reactors (PHWRs) in the United States, MSO scenarios of PHWRs are not included in the NEI 00-01 and any feasible scenarios have not been developed. This paper developed MSO scenarios which can be applied to PHWRs by reviewing the 63 MSO scenarios included in NEI 00-01. This study found that seven scenarios out of the 63 MSO scenarios can be applied and three more scenarios need to be developed.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.