• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2704-2710
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• 2024

Nuclear archaeology research provides scientific methods to reconstruct the operating histories of fissile material production facilities to account for past fissile material production. While it has typically focused on analyzing material in permanent reactor structures, spent fuel or high-level waste also hold information about the reactor operation. In this computational study, we explore a Bayesian inference framework for reconstructing the operational history from measurements of isotope ratios from a sample of nuclear waste. We investigate two different inference models. The first model discriminates between three potential reactors of origin (Magnox, PWR, and PHWR) while simultaneously reconstructing the fuel burnup, time since irradiation, initial enrichment, and average power density. The second model reconstructs the fuel burnup and time since irradiation of two batches of waste in a mixed sample. Each of the models is applied to a set of simulated test data, and the performance is evaluated by comparing the highest posterior density regions to the corresponding parameter values of the test dataset. Both models perform well on the simulated test cases, which highlights the potential of the Bayesian inference framework and opens up avenues for further investigation.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.92-99
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• 2020

High-temperature mechanical behaviors of Type 316L stainless steel (SS), which is considered as one of the major structural materials of Generation-IV nuclear reactors, were investigated through the tension and creep tests at elevated temperatures. The tension tests were performed under the strain rate of 6.67×10^-4 (1/s) from room temperature to 650℃, and the creep tests were conducted under different applied stresses at 550℃, 600℃, 650℃, and 700℃. The tensile behavior was investigated, and the modeling equations for tensile strengths and elongation were proposed as a function of temperature. The creep behavior was analyzed in terms of various creep equations: Norton's power law, modified Monkman-Grant relation, damage tolerance factor(λ), and Z-parameter, and the creep constants were proposed. In addition, the tested tensile and creep strengths were compared with those of RCC-MRx. Results showed that creep exponent value decreased from n=13.55 to n=7.58 with increasing temperature, λ = 6.3, and Z-parameter obeyed well a power-law form of Z=5.79E52(σ/E)^9.12. RCC-MRx showed lower creep strength and marginally different in creep strain rate, compared to the tested results. Same creep deformation was operative for dislocation movement regardless of the temperatures.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1330-1338
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• 2018

High-density UMo/Zr monolithic nuclear fuel plates have a promising application prospect in high flux research and test reactors. The solid state welding method called co-rolling is used for their fabrication. Hot co-rolling simulations for the composite blanks of UMo/Zr monolithic nuclear fuel plates are performed. The effects of coupon sizes and mechanical property parameters on the contact pressures between the to-be-bonded surfaces are investigated and analyzed. The numerical simulation results indicate that 1) the maximum contact pressures between the fuel coupon and the Zircaloy cover exist near the central line along the plate length direction; as a whole the contact pressures decrease toward the edges in the plate width direction; and lower contact pressures appear at a large zone near the coupon corner, where de-bonding is easy to take place in the in-pile irradiation environments; 2) the maximum contact pressures between the fuel coupon and the Zircaloy parts increase with the initial coupon thickness; after reaching a certain thickness value, the contact pressures hardly change, which was mainly induced by the complex deformation mechanism and special mechanical constitutive relation of fuel coupon; 3) softer fuel coupon will result in lower contact pressures and form interfaces being more out-of-flatness.

• Journal of Radiation Protection and Research
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• v.35 no.2
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• pp.57-62
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• 2010

Maintenance on the water chamber of steam generator, the change of pressurizer heater, the removal of pressure tube feeder, and so on during outage in nuclear power plants (NPPs) has a likelihood of high radiation exposure to whole body of workers even short time period due to the high radiation exposure rates. In particular, it is expected that hands would receive the highest radiation exposure because of its contact with radiation materials. In this study, field tests on extremity dose assessment of radiation workers for contact works with high radiation exposure were conducted during the maintenance periods in Korean pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). In this field test, radiation workers were required to wear additional TLDs on the back and wrist, and an extremity dosimeter on fingers including a main TLD on the chest, while performing maintenance. As a result, it was found that the equivalent dose for fingers was distributed in the fixed range of deep dose and the equivalent dose for wrists.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.41-51
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• 2004

The cost-effective reduction of occupational radiation dose (ORD) at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORD data of existing plants. Through the data analysis, it is required to identify what are the jobs of repetitive high ORD at the nuclear power plant. In this study, Percentile Rank Sum Method (PRSM) is proposed to identify repetitive high ORD jobs, which is based on non-parametric statistical theory. As a case study, the method is applied to ORD data of maintenance and repair jobs at Kori units 3 and 4 that are pressurized water reactors with 950 MWe capacity and have been operated since 1986 and 1987, respectively in Korea. The results was verified and validated, and PRSM has been demonstrated to be an efficient method of analyzing the data.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.1
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• pp.1-8
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• 2010

The ultrafashionable machinery that require high quality electricity power has been daily come into being. Because domestic power system has been larger and more complicated in accordance with raising power demand by power market requirement. Because of these power market situations, The FACTS (Flexible AC Transmission System) which is power transmission system for the next generation to meet flexible supply the power and reliability has been applied. If they, compensators and FACTS, are used inter-efficiently in range that does not affect the stability and a badly influence the security, they might be increase in the voltage stability of system, supply reliability and also achieve the voltage control in a suddenly changed power system. Therefore we describe and suggest on this treatise that a plan for coordination control between UPFC, Shunt elements (Sh. Capacitors & Sh. Reactors) among compensators and also describe the method to keep or control the voltage of power system in allowable ranges. The method follows that, we used characteristics of each equipment, UPFC would be also settled to keep the identified voltage range in change of load bus, Shunt elements also would be settled to supply the reactive power shortage in out of operating range of UPFC to cope actively with change of the power system. As the result of simulation, it is possible to keep the load bus voltage in limited range in spite of broad load range condition. This helps greatly for the improvements of supply reliability and voltage stability.

• The Transactions of the Korean Institute of Electrical Engineers P
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• v.51 no.2
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• pp.60-67
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• 2002

This paper presents a control approach for designing a self tuning fuzzy controller for a synchronous generator excitation and SVC system. A combination of thyristor-controlled reactors and fixed capacitors (TCR-FC) type SVC is recognized as having the most flexible control and high speed response, which has been widely utilized in power systems, is considered and designed to improve the response of a synchronous generator, as well as controlling the system voltage. The proposed parameter self tuning algorithm of fuzzy controller is based on the steepest decent method using two direction vectors which make error between inference values of fuzzy controller and output values of the specially selected PSS reduce steepestly. Using input-output data pair obtained from PSS, the parameters in antecedent part and in consequent part of fuzzy inference rules are learned and tuned automatically using the proposed steepest decent method. The related simulation results show that the proposed fuzzy controller is more powerful than the conventional ones.

• The Transactions of the Korean Institute of Power Electronics
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• v.22 no.6
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• pp.484-495
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• 2017

In HVDC thyristor valves, more than 95% of heat loss occurs in snubber resistors and valve reactors. In order to dissipate the heat from the valves and to suppress the electrolytic current, water with a high heat capacity and a low conductivity of less than 0.2 uS/cm must be used as a refrigerant of the heat sink. The cooling parts must also be arranged to reduce the electrolytic current, whereas the pipe that supplies water to the thyristor heat sink must have the same electric potential as the valve. Corrosion is mainly caused by electrochemical reactions and the influence of water quality and leakage current. This paper identifies the refrigerants involved in the ionization, electrical conductivity, and corrosion in HVDC thyristor valves. A method for preventing corrosion is then introduced. The design of the heat sink with an excellent heat radiation is also analyzed in detail.