• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1892-1900
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• 2023

A protective oxide layer forms on the material surfaces of a Nuclear Power Plant during operation due to high temperature. These oxides can host radionuclides, the activated corrosion products of fission products, resulting in decommissioning workers' exposure. These deposited oxides are iron oxides such as Fe₃O₄, Fe₂O₃ and mixed ferrites such as nickel ferrites, chromium ferrites, and cobalt ferrites. Developing a new chemical decontamination technology for domestic CANDU-type reactors is challenging due to variations in oxide compositions from different structural materials in a Pressurized Water Reactor (PWR) system. The Korea Atomic Energy Research Institute (KAERI) has already developed a chemical decontamination process for PWRs called 'HyBRID' (Hydrazine-Based Reductive metal Ion Decontamination) that does not use organic acids or organic chelating agents at all. As the first step to developing a new chemical decontamination technology for the Pressurized Heavy Water Reactor (PHWR) system, we investigated magnetite dissolution behaviors in various HyBRID inorganic acidic solutions to assess their applicability to the PHWR reactor system, which forms a thicker oxide film.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1291-1302
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• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.4
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• pp.427-433
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• 2007

A structural model of the spent nuclear fuel disposal canister for the pressurized water reactor(PWR) for about 10,000 years long term deposition at a 500m deep granitic bedrock repository has been developed. However this developed structural model of the spent nuclear fuel disposal canister is too heavy to handle without any structural safety problem. Hence a lighter structural model of the spent nuclear fuel disposal canister which is easy to handle has been tried to develop very much. One of the reasons which made the structural model heavy is considered to be due to the severe adaptation of the design conditions like external loads and safety factor etc. to the canister design. Hence a complementary analysis to alleviate such severe design conditions is required for the reduction of the canister weight. In this study, a complementary structural analysis for the spent nuclear fuel disposal canister is carried out changing the design conditions such as external loads and safety factors to recalculate the design parameters like diameter and thickness etc. of the canister. The complementary analysis results shows that the diameter of canister can be shortened from 122cm to 102cm to reduce the weight of the spent nuclear fuel disposal canister.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.594-597
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• 2001

As many industrial systems become more complex, they become extremely difficult to diagnose the cause of failures. The subject of this paper is ECCS(Emergency Core Cooling System) part of PWR(Pressurized Water Reactor). This paper presents modeling and diagnoser construction of ECCS based on discrete event system theory. Also, this paper presents that the ECCS system is diagnosible in our approach.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.479-489
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• 2019

Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer. Electrochemical analysis revealed that the oxide layers formed on the LCF crack surface of Alloy 690 had higher impedance and less defect density than those of 316 SS, which resulted in longer LCF life of Alloy 690 than 316 SS in a simulated PWR environment.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• 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.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.38 no.11
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• pp.861-869
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• 1989

The optimization problems, based on Pontryagin's Maximum Principle, for minimizing (damping) Xenon spatial oscillations in Load Following operations of Pressurized Water Reactor (PWR) is presented. The optimization model is formulated as an optimal tracking problem with quadratic objective functional. The oen-group diffusion equations and Xe-I dynamic equations are defined as equality constraints. By applying the maximum principle, the original problem is decomposed into a single time problem with no constraints. The resultant subproblems are optimized by using the conjugate Gradient Method. The computational results show that the Xenon spatial oscillation is minimized, and the reactor follows the load demand of the electrical power systems while maintaining the desired power distribution.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.941-952
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• 2017

Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.