• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1443-1451
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• 2020

The objective of this study is to design a robust power control system for a small pressurized water reactor (PWR) to achieve stable power operations under conditions of external disturbances and internal model uncertainties. For this purpose, the multiple-input multiple-output transfer function models of the reactor core at five power levels are derived from point reactor kinetics equations and the Mann's thermodynamic model. Using the transfer function models, five local reactor power controllers are designed using an H infinity (H_∞) mixed sensitivity method to minimize the core power disturbance under various uncertainties at the five power levels, respectively. Then a multimodel approach with triangular membership functions is employed to integrate the five local controllers into a multimodel robust control system that is applicable for the entire power range. The performance of the robust power system is assessed against 10% of full power (FP) step load increase transients with coolant inlet temperature disturbances at different power levels and large-scope, rapid ramp load change transient. The simulation results show that the robust control system could maintain satisfactory control performance and good robustness of the reactor under external disturbances and internal model uncertainties, demonstrating the effective of the robust power control design.

• Nuclear Engineering and Technology
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• 제45권1호
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• pp.61-66
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• 2013

Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as $^{129}I$ is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for $^{129}I$ as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding $^{125}I$ as a radio-isotopic tracer ($t_{1/2}$ = 60.14 d) to the simulation sample, in order to measure the activity concentration of $^{129}I$ in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, $Cl^-$ form) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of $^{129}I$ was examined, as was the effect of $^3H$ on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous $^3H$ presence was found with activity concentrations of $^3H$ lower than 50 Bq/mL, and with a boron concentration of less than 2,000 ${\mu}g/mL$.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• 자원ㆍ환경경제연구
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• 제27권2호
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• pp.261-286
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• 2018

본 논문은 제 7차 전력수급기본계획에서 제시한 신규 원자력발전, 석탄 발전, 그리고 LNG 복합 발전의 균등화발전비용과, 고리 1호기(가압형 경수로, PWR) 및 월성 1호기(가압형 중수로, PHWR)의 계속운전 기간별(10년과 20년) 균등화발전비용을 추정하여 비교해서 원전 계속운전의 노형별 및 계속운전 기간별 경제성을 평가하였다. 균등화발전비용을 이용한 원자력 발전의 계속운전 경제성은 노형, 계속운전기간, 할인율, 이용률 등으로부터 영향을 받는다. 분석결과에 따르면 가압형 경수로(고리 1호)는 가압형 중수로(월성 1호)보다 경제성이 높다. 원자력발전의 계속운전과 다른 전원의 경제성 비교 결과를 보면 가압형 경수로(고리 1호)의 경우에 20년 계속운전이 신규 원자력 발전 및 석탄발전보다 경제적이다. 그러나 가압형 중수로(월성 1호)의 경우에 20년 계속운전은 LNG 복합 발전보다 경제적이지만, 신규 원전 및 신규 석탄발전보다 비경제적이다. 원자력발전의 계속운전에서 보면 20년 계속운전이 경제적이며, 특히 가압형 경수로는 다른 전원보다 비용효율적이다. 원자력발전의 계속운전 정책은 모든 원전을 폐로하기 보다는 안전성과 경제성을 동시에 고려하는 선별적 접근 방식이 유효하다.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.929-944
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• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• 한국정밀공학회지
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• 제32권9호
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• pp.815-824
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• 2015

In this study, we conducted the approximate multi-objective optimization of gap sizes of pressurized-water reactor (PWR) annular fuels. To determine the contacting tendency of the inner-outer gaps between the annular fuel pellets and cladding, thermoelastic-plastic-creep (TEPC)analysis of PWR annular fuels was performed, using in-house FE code. For the efficient heat transfer at certain levels of stress, we investigated the tensile, compressive hoop stress and temperature, and optimized the gap sizes using the non-dominant sorting genetic algorithm (NSGA-II). For this, response surface models of objective and constraint functions were generated, using central composite (CCD) and D-optimal design. The accuracy of approximate models was evaluated through $R^2$ value. The obtained optimal solutions by NSGA-II were verified through the TEPC analysis, and we compared the obtained optimum solutions and generated errors from the CCD and D-optimal design. We observed that optimum solutions differ, according to design of experiments (DOE) method.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1995년도 봄 학술발표회 논문집
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• pp.99-106
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• 1995

The seismic responses of a base Isolated Pressurized Water Reactor(PWR) are investigated using a mathematical model which expresses the superstructure as a linear lumped mass-spring and the seismic Isolator as an equivalent spring-damper. Time history analyses are performed for the 1940 El Centre earthquake with linear amplification. In the analysis 5% of structural damping is used for the superstructure. The effects of high damping rubber bearing on seismic response of the superstructure in base isolated system are evaluated for four stiffness model types. The acceleration responses in base isolated PWR superstructure with high damping rubber bearings are much smaller than those in fixed base structure. In the higher strain region where stiffness behaves non-linearly, the acceleration responses modelled by one equivalent stiffness are smaller than those in nonlinear spring model, and the higher stiffness spring model of isolator exhibits larger peak acceleration response at superstructure in the frequency range above 2.0 Hz. when subjected to linearly amplified 1940 El Centre earthquake.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.1-8
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• 1999

In this paper, we propose a new type of power control system for the rod drive coil of the CEDM of the PWR NPP in order to supply more reliable DC power The electrical modelling of the controlled rod drive coil was done by referring related documentations. The design of the proposed system is based on this electric81 model satisfying the existing specification. A high power DC-DC converter scheme is adopted utilizing the SMPS technique in the design of the proposed system. In order to show the effectiveness of the proposed system, an experimental system with the capability of 3.2 K Watt was set up for a rod with four cores and some computer simulations and experimentations were carried out. The result shows a very similar tracking performance with that of the existing system to the driving command. As a result of this, the proposed method can be applied to the power control system for the rod drive coil of the CEDM of the PWR NPP.