• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.28-36
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• 1993

Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.221-230
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• 2007

Wavelet theory was applied to detect a singularity in a reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, using wavelet transform after de-noising, singular points can easily be found. To test this theory, reactor power signals were generated using the HANARO(a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations contaminated with Gaussian noise. Wavelet transform decomposition and de-noising procedures were applied to these signals. It was possible to detect singular events such as a sudden reactivity change and abrupt intrinsic property changes. Thus, this method could be profitably utilized in a real-time system for automatic event recognition(e.g., reactor condition monitoring).

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.20-29
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• 1991

The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU longrightarrow FBR strategy is to be the best option for Korea. However, APWR, CANDU longrightarrow Passive Safe Reactor(PSR）longrightarrowFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.63-74
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• 2007

In this paper, a reconfigurable controller consisting of a normal controller and a standby controller is designed to control the thermoelectric (TE) power in the SP-100 space reactor. The normal controller uses a model predictive control (MPC) method where the future TE power is predicted by using support vector regression. A genetic algorithm that can effectively accomplish multiple objectives is used to optimize the normal controller. The performance of the normal controller depends on the capability of predicting the future TE power. Therefore, if the prediction performance is degraded, the proportional-integral (PI) controller of the standby controller begins to work instead of the normal controller. Performance deterioration is detected by a sequential probability ratio test (SPRT). A lumped parameter simulation model of the SP-100 nuclear space reactor is used to verify the proposed reconfigurable controller. The results of numerical simulations to assess the performance of the proposed controller show that the TE generator power level controlled by the proposed reconfigurable controller could track the target power level effectively, satisfying all control constraints. Furthermore, the normal controller is automatically switched to the standby controller when the performance of the normal controller degrades.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.96-104
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• 1995

EPRI URD requires that the reactor be capable of accommodating an unintended CEA drop without initiating a trip and operating at a reduced power with ay single CEA fully inserted. YGN 3 and 4 reactors have 12-finger CEAs, and the CPCS will trip the reactor due to their large reactivities when one of them is dropped at a high power. The ABB-CE reactor power cutback system has been proposed to be used against the 12-Finger CEA drop to avoid the reactor trips. The results of study show that the reactor power cutback can prevent the reactor trips of the 12-Finger CEA drop when the CPCS has enough operating thermal margin (more than 9％ for YGN 3&4 Cycle 1). It is noted, however, that the probability of a 12-Finger CEA drop is very low, less than one per 100 reactor years for YGN 3& and System 80$^{+}$ plants.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1047-1057
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• 2016

Design engineering process for field programmable gate array (FPGA)-based reactor trip functions are developed in this work. The process discussed in this work is based on the systems engineering approach. The overall design process is effectively implemented by combining with design and implementation processes. It transforms its overall development process from traditional V-model to Y-model. This approach gives the benefit of concurrent engineering of design work with software implementation. As a result, it reduces development time and effort. The design engineering process consisted of five activities, which are performed and discussed: needs/systems analysis; requirement analysis; functional analysis; design synthesis; and design verification and validation. Those activities are used to develop FPGA-based reactor bistable trip functions that trigger reactor trip when the process input value exceeds the setpoint. To implement design synthesis effectively, a model-based design technique is implied. The finite-state machine with data path structural modeling technique together with very high speed integrated circuit hardware description language and the Aldec Active-HDL tool are used to design, model, and verify the reactor bistable trip functions for nuclear power plants.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1972-1981
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• 2022

A typical ferritic/martensitic (F/M) steel sheet was subjected to pulsed laser surface remelting (LSR) and corrosion test in lead-bismuth eutectic (LBE) at 550 ℃. There present two modification zones with distinct microstructures in the LSRed specimen: (1) remelted zone (RZ) consisting of both bulk δ-ferrite grains and martensitic plates and (2) heat-affected zone (HAZ) below the RZ, mainly composed of martensitic plates and high-density precipitates. Martensitic transformation occurs in both the RZ and the HAZ with the Kurdjumov-Sachs and Nishiyama-Wassermann orientation relationships followed concurrently, resulting in scattered orientations and specific misorientation characteristics. Hardnesses of the RZ and the HAZ are 364 ± 7 HV and 451 ± 15 HV, respectively, considerably higher than that of the matrix (267 ± 3 HV). In oxygen-saturated and oxygen-depleted LBE, thicknesses of oxide layers developed on both the as-received and the LSRed specimens increase with prolonging corrosion time (oxide layers always thinner under the oxygen-depleted condition). The corrosion resistance of the LSRed F/M steel in oxygen-saturated LBE is improved, which can be attributed to the grain-refinement accelerated formation of dense Fe-Cr spinel. In oxygen-depleted LBE, the growth of oxide layers is very low with both types of specimens showing similar corrosion resistance.