• 시스템엔지니어링학술지
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• 제17권2호
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• pp.69-81
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• 2021

The human performance issues and increased automation issues in advanced Small Modular Reactors (SMRs) are critical to numerous stakeholders in the nuclear industry, due to the undesirable implications targeting the Man Machine Interface Systems (MMIS) complexity of (Generation IV) SMRs. It is imperative that the design of future SMRs must address these problems. Nowadays, Multi Agent Systems (MAS) are used in the industrial sector to solve multiple complex problems; therefore incorporating this technology in the proposed innovative SMR (I-SMR) design will contribute greatly in the decision making process during plant operations, also reduce the number MCR operating crew and human errors. However, it is speculated that an increased level of complexity will be introduced. Prior to achieving the objectives of this research, the tools used to analyze the system for complexity reduction, are the McCabe's Cyclomatic complexity metric and the Henry-Kafura Information Flow metric. In this research, the systems engineering approach is used to guide the engineering process of complexity reduction concept of the system in its entirety.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1349-1359
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• 2016

Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX ($T_{co}$) and the hot outlet temperature of the intermediate heat exchanger ($T_{ho2}$) by manipulating the hot-side flow rates of the heat exchangers ($F_h/F_{h2}$) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX ($T_{co}$) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.838-848
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• 2015

One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented.SMCis ameansto control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• 방사성폐기물학회지
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• 제21권4호
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.445-450
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• 1997

Self-Powered Neutron Detectors(SPNDs) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability of beta particle as a function of their birth position fur the typical geometry of rhodium-based SPNDs. Also, a simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation and the numerical method, the burn-up profile of rhodium density and the neutron sensitivity are calculated as a function of burn-up time in the reactor. The sensitivity of the SPNDs decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• 국제학술발표논문집
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• The 7th International Conference on Construction Engineering and Project Management Summit Forum on Sustainable Construction and Management
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• pp.8-13
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• 2017

Nuclear power plant (NPP) constructions are large scale projects that are executed for several years, and schedule control utilizing various schedules is a critically important factor. Recently Korea independently developed the Advanced Power Reactor (APR) 1400 and is building nuclear facilities applying this new reactor type. The construction of Shin-Kori NPP (SKN) Unit 3, which adopted the APR1400, was completed and commercial operation has begun, while, SKN 4, Shin-Hanul NPP (SHN) Units 1&2, and SKN 5&6 are currently under construction. Prior to the development of the APR1400, Korea built 24 reactors and accumulated the schedule data of various reactor types which provided the foundation for schedule reduction to be possible. However, as there is no schedule development and review system established based on the standard schedule data (standard activities, durations, etc.) by reactor type, the process for developing the schedule for new builds is low in efficiency consuming much time and manpower. Also all construction data has been accumulated based on schedule activities. But because the connectivity of activities between projects is low, it is difficult to utilize such accumulated data (causes for schedule delay, causes for design changes, etc.) in new build projects. Due to such reasons, issues continue to arise in the process of developing standard schedule activities and a standard schedule for nuclear power plant construction. In order to develop a standard schedule for NPP construction, i) the development of an NPP standard schedule activity list, ii) development of the connection logic of NPP standard schedule activities, iii) development of NPP standard schedule activity resources and duration, and iv) integration of schedule data need to be performed. In this paper, an analysis was made on the coherence of schedule activity descriptions of existing NPPs by applying the probabilistic methodology on activities with low connectivity due to the utilization of the numbering system of four APR1400 reactors (SHN 1&2 and SKN 3&4).This study also describes the method for developing a standard schedule activity list and connectivity measures by extracting same and/or similar schedule activities.