• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.689-698
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• 2014

Reactor internals are sensitive to dynamic loads such as earthquakes and flow induced vibration. Thus, it is essential to identify the dynamic characteristics to evaluate the seismic integrity of the structures. However, a full-sized system is too large to perform modal experiments, making it difficult to extract data on its modal characteristics. In this research, we constructed a finite element model of the APR1400 reactor internals to identify their modal characteristics. The commercial reactor was selected to reflect the actual boundary conditions. Our FE model was constructed based on scale-similarity analysis and fluid-structure interaction investigations using a fabricated scaled-down model.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.171-186
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• 2007

Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.

• 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.50 no.6
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3326-3333
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• 2023

The results of the analysis of a mathematical modeling for the IBR-2M pulsed reactor dynamics for a transition from a power pulse repetition frequency of 5 Hz-10 Hz are presented. The change in the amplitude response of the reactor for variable pulse delayed neutron fraction was studied. We used a set of power feedback parameters determined experimentally in 2021 at an energy output of 1820 MW·day. At a pulse repetition frequency of 10 Hz, the amplitude of pulse energy oscillations significantly depends on the value of the delayed neutron fraction in pulse β_p. Depending on β_p both suppression and amplification of reactor power fluctuations in the frequency ranges of 0.05-0.20 and 1.25-5.00 Hz can be realized.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1495-1502
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• 2020

A new method of X-ray source spectrum estimation based on compressed sensing is proposed in this paper. The algorithm K-SVD is applied for sparse representation. Nonnegative constraints are added by modifying the L₁ reconstruction algorithm proposed by Rosset and Zhu. The estimation method is demonstrated on simulated spectra typical of mammography and CT. X-ray spectra are simulated with the Monte Carlo code Geant4. The proposed method is successfully applied to highly ill conditioned and under determined estimation problems with a good performance of suppressing noises. Results with acceptable accuracies (MSE < 5%) can be obtained with 10% Gaussian white noises added to the simulated experimental data. The biggest difference between the proposed method and the existing methods is that multiple prior knowledge of X-ray spectra can be included in one dictionary, which is meaningful for obtaining the true X-ray spectrum from the measurements.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1814-1820
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• 2023

Safety classification of systems, structures, and components (SSC) is an essential activity for nuclear reactor design and operation. The current regulatory trend is to require risk-informed safety classification that considers first, the severity, but also the frequency of SSC failures. While safety classification for nuclear power plants is covered in many regulatory and scientific publications, research reactors received less attention. Research reactors are typically of lower power but, at the same time, are less standardized i.e., have more variability in the design, operational modes, and operating conditions. This makes them more challenging when considering safety classification. This work presents the Integrated Risk-Informed Safety Classification (IRISC) procedure which is a novel extension of the IAEA recommended process with dedicated probabilistic treatment of research reactor designs. The article provides the details of probabilistic analysis performed within safety classification process to a degree that is often missing in most literature on the topic. The article presents insight from the implementation of the procedure in the safety classification for the MARIA Research Reactor operated by the National Center for Nuclear Research in Poland.

• Proceedings of the Korean Institute of Intelligent Systems Conference
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• 1993.06a
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• pp.1062-1065
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• 1993

Nuclear reactor operation is a human intensive task; one of the features of a problem for which fuzzy controllers present the most suitable solution. The performance of the fuzzy controllers can further be improved through tuning. In this work, application of a fuzzy controller in real-time control of a nuclear reactor is presented. The fuzzy controller is tuned on-line using direct gradient search method.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1338-1348
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• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.