• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.299-304
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• 1997

A neuro-fuzzy control algorithm is applied for xenon spatial oscillations in a pressurized water reactor. The consequent and antecedent parameters of the fuzzy rules are tuned by the gradient descent mettled. The reactor model used for computer simulations is a two-point xenon oscillation model. The reactor core is axially divided into two regions and each region has one input and one output and is coupled with the other region. The interaction between the regions of the reactor core is treated by a decoupling scheme. This proposed control of mettled exhibits very fast responses to a step or a ramp change of target axial offset without any residual flux oscillations.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.98-109
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• 1983

A two-step alternating direction explicit method is developed to solve the space-dependent reactor kinetics equations in two space dimensions. As a special case in the general class of alternating direction implicit methods, this method is analysed for accuracy and stability. To test the validity of this method it is compared with the implicit-difference method used in the TWIGL program. It is shown that the two methods are closely related. The time dependent neutron fluxes of the pressurized water reactor (PWR), during control rod insertion, and, of the CANDU-PHW reactor, in case of postulated loss of coolant accident, are obtained from the numerical calculation results.

• Journal of Electrical Engineering and information Science
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• v.2 no.3
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• pp.66-70
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• 1997

The steam generator level is susceptible to the nonminimum phase in dynamics due to the thermal reverse effects known as "shrink and swell" in a pressurized water reactor. A state feedback assisted control concept is presented for the change of dynamic performance to the minimum phase the concept incorporates a nonlinear digital observer as a part of the control system. The observer is deviced to estimate the state variables that provide the true indication of water inventory by compensating for shrink and swell effects. The concept is validated with implementation into the steam generator simulation model.

• Journal of Welding and Joining
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• v.34 no.5
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• pp.13-18
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• 2016

Toshiba has developed a laser peening system for PWRs(pressurized water reactors) as well after the one for BWRs(boiling water reactors), and applied it for BMI(bottom-mounted instrumentation) nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.206-213
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• 2003

This study focuses on an understanding of the environmental effect on delayed hydride cracking velocity (DHCV) of CANDU Zr-2.5Nb tubes. To simulate DHC susceptibility of the Zr-2.5Nb tubes in reactor operating conditions, DHC tests were successfully carried out in pressurized water at 180 and $250^{\circ}C$ using a self-designed autoclave for the first time. Using 17 mm compact tension specimens electorlytically charged to 34 and 60 ppm H, 3 to 7 DHCV data were determined in water at both temperatures and compared to those determined in air that were already confirmed to be valid through a round robin test on DHCV of Zr-2.5Nb tubes sponsored by a IAEA coordinated research program. The pressurized water environment has little effect on DHCV of Zr-2.5Nb tube in water at both temperatures even though DHCV is slightly lower in water than that in air. The lower DHCV of the Zr-2.5Nb tube during short-term tests is discussed in viewpoint of the cooling rate from the peak temperature to the test temperature.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.87-94
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• 2009

Tritium is the one of the dominant contributors to the internal radiation exposure of workers at pressurized heavy water reactors (PHWRs). This nuclide is likely to release to work places as tritiated water vapor (HTO) from a nuclear reactor and gets relatively easily into the body of workers by inhalation. Inhaled tritium usually reaches the equilibrium of concentration after approximately 2 hours inside the body and then is excreted from the body with a half-life of 10 days. Because tritium inside the body transports with body fluids, a whole body receives radiation exposure. Internal radiation exposure at PHWRs accounts for approximately 20-40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Thus, tritium is an important nuclide to be necessarily monitored for the radiation management safety. In this paper, metabolism for tritium is established using its excretion rate results in urine samples of workers at PHWRs and an effective half-life, a key parameter to estimate the radiation exposure, was derived from these results. As a result, it was found that the effective half-life for workers at Korean nuclear power plants is shorter than that of International Commission on Radiological Protection guides, a half-life of 10 days.

• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.138-145
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• 2012

As of the end of 2010, 21 nuclear power reactors were operating in Korea. Radioactive effluents from nuclear power plants (NPPs) had been increased continuously and the radioactivity of effluents released in 2010 was 547.12 TBq. From 2001 to 2010, the annual average radioactivity of gaseous and liquid effluents per reactor was 11.61 TBq for pressurized water reactor (PWR) plants and 118.12 TBq for PHWR (pressurized heavy water reactor) plants. Most of the radioactivity from gaseous and liquid effluents was came from $^3H$. Based on the results of release trends and analysis, effluents characteristics was suggested for the management of radioactive effluents from NPPs.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.396-410
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• 2018

To assist operators to properly assess the current situation of the plant, accurate fault diagnosis methodology should be available and used. A reliable fault diagnosis method is beneficial for the safety of nuclear power plants. The major idea proposed in this work is integrating the merits of different fault diagnosis methodologies to offset their obvious disadvantages and enhance the accuracy and credibility of on-line fault diagnosis. This methodology uses the principle component analysis-based model and multi-flow model to diagnose fault type. To ensure the accuracy of results from the multi-flow model, a mechanical simulation model is implemented to do the quantitative calculation. More significantly, mechanism simulation is implemented to provide training data with fault signatures. Furthermore, one of the distance formulas in similarity measurement-Mahalanobis distance-is applied for on-line failure degree evaluation. The performance of this methodology was evaluated by applying it to the reactor coolant system of a pressurized water reactor. The results of simulation analysis show the effectiveness and accuracy of this methodology, leading to better confidence of it being integrated as a part of the computerized operator support system to assist operators in decision-making.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.89-103
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• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.