• Nuclear Engineering and Technology
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• 제27권3호
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• pp.336-343
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• 1995

적응제어의 한 방식인 자기동조제어(STR) 방식이 비선형 노심 모델의 출력 조정에 적용된다. 적응제어는 비선형, 시변 및 확률(Stochastic) 시스템을 위한 준최적 제어기를 설계하기 위한 적절한 제어 방식이다. 제어계통은 미지의 시변 파라메타를 갖는 3차 선형 모델에 기초한다. 파라메타는 가변 망각계수를 도입한 늑장 최소자승법에 의하여 매시간(Time Step) 순환적으로 평가된다. 평가된 파라메타를 이용하여 한 스텝 먼저 냉자재 평균온도가 예측되고 이 예측된 값과 Setpoint 값과의 차이를 최소화함은 물론, 제어봉의 움직임을 막고자 가중(Weighted) One-step-ahead 제어기가 설계된다. 또한 적분동작이 첨가되어 정상상태 에러가 제거된다. 넓은 운전영역을 포괄하는 비선형 PWR 모델이 원자로 출력 조정을 위한 본 제어기를 시뮬레이션하는데 이용되었다. 시뮬레이션 결과로부터 본 제어기의 성능이 우수한 것으로 판명되었다.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.615-624
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• 2008

In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1 %), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3648-3658
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• 2023

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly containing a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm² of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.

• 한국환경과학회지
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• 제27권6호
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• pp.447-456
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• 2018

In this research, the target process was a modified type of a conventional aeration tank with four different influent feeding points and alternated aeration to obtain nitrogen removal. For more accurate switching of influent feeding, the process was operated under a designed control strategy based on monitoring of $NH_4-N$ and $NO_X-N$ concentrations in the tank. However, the strategy did have some limitations. For example, it was not sensitive to detecting the end of each reaction when losing the balance between nitrification and denitrification of each opposite part of biological tank. To overcome the limitations of the existing control strategy, a diagnosis-based control strategy was suggested in this research using the diagnosis results classified as normal (N), ammonia accumulation (AA) and nitrate accumulation (NA). Using the pre-designed rules for control actions, the aeration and volume of the aerated part of the reactor could be increased or decreased at a fixed mode time. In simulations of the suggested diagnosis-based control strategy, the $NH_4-N$ and $NO_X-N$ removal rates in the reactor were maintained at higher levels than those of the existing control strategy.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• 대한전기학회논문지:시스템및제어부문D
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• 제52권8호
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• pp.452-461
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• 2003

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indicator system and its actual implementation in the existing nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indicator. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indicator as well. This paper investigates efficiency of the magnetic flux concentrator and the hysteresis using FEM and verified differences in physicals characteristics by comparing the results of FEM and those of the experiment. As a result, it is shown that the characteristics of prototype control rod position indicator have a good agreement with the results of FEM.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1254-1266
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• 2024

This study investigates the application of supercritical carbon dioxide (S-CO₂) direct-cycle micro modular reactors (MMRs) in primary frequency control (PFC), which is a scenario characterized by significant load fluctuations that has received less attention compared to secondary load-following. Using a modified GAMMA + code and a deep neural network-based turbomachinery off-design model, the authors conducted an analysis to assess the behavior of the reactor core and fluid system under different PFC scenarios. The results indicate that the acceptable range for sudden relative electricity output (REO) fluctuations is approximately 20%p which aligns with the performance of combined-cycle gas turbines (CCGTs) and open-cycle gas turbines (OCGTs). In S-CO₂ direct-cycle MMRs, the control of the core operates passively within the operational range by managing coolant density through inventory control. However, when PFC exceeds 35%p, system control failure is observed, suggesting the need for improved control strategies. These findings affirm the potential of S-CO₂ direct-cycle MMRs in PFC operations, representing an advancement in the management of grid fluctuations while ensuring reliable and carbon-free power generation.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• 융합정보논문지
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• 제9권2호
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• pp.34-42
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• 2019

본 연구에서는 자동제어시스템이 리액터 운전에 미치는 영향을 분석해 보았다. Propylene 리액터 공정은 복잡하게 구성되어 일반적으로 효율이 낮고, 생산성이 떨어지는 문제로 인해 경제적 손실이 많다. 본 연구에서는 장애요인을 보완하여 운전효율을 높이고, 생산성을 향상시키는 목적으로 연구 모델을 제시하였다. 기존 공정의 구성을 분석하여 하드웨어 시스템 및 소프트웨어 시스템을 독창성 있는 모델로 보완하였다. 설비의 구성을 연간 60 만톤/Year 프로플린을 생산하는 리액터 8기를 기준으로 적용하였다. 연구 모델 적용결과 운전 효율이 높아졌고, 안정성 91%과 더불어 생산량이 90~95% 으로 증가하였다. 향후 연구는 생산성 100% 향상을 위한 연구모델을 제시할 예정이다. 추가적으로 프로플린 부산물인 수소 생산공정인 PSA 시스템의 안정성과 생산성 향상 방안에 대해 연구하고자 한다.