• Nuclear Engineering and Technology
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• 제44권4호
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• pp.369-378
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• 2012

A load-following operation in APR+ nuclear plants is necessary to reduce the need to adjust the boric acid concentration and to efficiently control the control rods for flexible operation. In particular, a disproportion in the axial flux distribution, which is normally caused by a load-following operation in a reactor core, causes xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. A model predictive control (MPC) method was used to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for APR+ nuclear plants. Some tracking controllers employ the current tracking command only. On the other hand, the MPC can achieve better tracking performance because it considers future commands in addition to the current tracking command. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at the current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The support vector regression (SVR) model that is used widely for function approximation problems is used to predict the future outputs based on previous inputs and outputs. In addition, a genetic algorithm is employed to minimize the objective function of a MPC control algorithm with multiple constraints. The power level and ASI are controlled by regulating the control banks and part-strength control banks together with an automatic adjustment of the boric acid concentration. The 3-dimensional MASTER code, which models APR+ nuclear plants, is interfaced to the proposed controller to confirm the performance of the controlling reactor power level and ASI. Numerical simulations showed that the proposed controller exhibits very fast tracking responses.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2215-2221
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• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2007년도 추계학술대회 논문집
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• pp.233-234
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• 2007

We investigated the quench characteristics in integrated three-phase flux-lock type superconducting fault current limiter (SFCL), which consisted of three-phase flux-lock reactor wound on one iron core with the same turn's ratio between coil 1 and coil 2 for each single phase. To study the quench characteristics of the SFCL, the experiments was performed on various fault type such as the single line-to-ground fault, the double line-to-ground fault, the triple line-to-ground fault. From the experimental results, the generated point of element resistances was different on various fault type.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2006년도 춘계학술대회 논문집
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• pp.467-470
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• 2006

Superconducting fault current(SFCLs) are expected to improve not only reliability but also stability of real power systems. The analysis on the single line-to-ground fault current of the integrated three phase flux-lock type SFCL, which consists of three flux-lock reactor wound on an iron core in each single phase and three YBCO thin films, was investigated in current limiting operating characteristics. We compared additive polarity winding with the subtractive one in the flux lock reactor. Its turns ratio each phase between the primary and the secondary coils is 63:42. When a single line-to-ground fault occurred in any phase, the peak value of line current in the fault phase of the additive polarity winding increased up to 31.44[A] during first-half cycle. On the other hand, the peak value in the subtractive polarity winding increased up to 81.77[A] under the same conditions. This is because the current flow between the primary and the secondary windings becomes to be additive or subtractive in each winding direction. We confirmed that the current limiting behavior in the additive polarity winding was more effective for a single-line-to ground fault.

• 시스템엔지니어링학술지
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• 제14권1호
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• pp.63-71
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• 2018

This paper discusses decommissioning procedure requirements management using requirement engineering to systematically manage the technical requirements and criteria that are required in decontamination and decommissioning activities, and the regulatory requirements that should be complied with in a decommissioning strategy for research reactors and nuclear power plants. A schema was designed to establish the traceability and change management related to the linkage between the regulatory requirements and technical criteria after classifying the procedures into four groups during the full life-cycle of the decommissioning. The results confirmed that the designed schema was successfully traced in accordance with the regulatory requirements and technical criteria required by various fields in terms of decontamination and decommissioning activities. In addition, the changes before and after the revision of the Nuclear Safety Act were also determined. The dismantling procedure requirement management system secured through this study is expected to be a useful tool in the integrated management of radioactive waste, as well as in the dismantling of research reactor and nuclear facilities.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2005년도 추계학술대회 논문집 Vol.18
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• pp.165-166
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• 2005

In this paper, we investigated the quench generation of HTSC elements in fault types according to inductance variation in the integrated three-phase flux-lock type SFCL. The integrated three-phase flux-lock type SFCL was the upgrade version of the single-phase flux-lock type SFCL. The structure of the integrated three-phase flux-lock type SFCL consisted of three-phase flux-lock reactor wound on an iron core with the ratio of the same turn between coil 1 and coil 2 in each phase. When the SFCL is operated under the normal condition, the flux generated in the iron core is zero because the flux generated between two coils of each single phase is canceled out. Therefore, the SFCL's impedance is zero, and the SFCL has negligible influence on the power system. However, if a fault occurs in any single-phase among three phases, the flux generated in the iron core is not zero any more. The flux makes HTSC elements of all phases quench irrespective of the fault type, which reduces the current of fault phase as well as the current of sound phase. It was observed that the fault current limiting characteristics of the suggested SFCL were dependent on the quench characteristics of HTSC elements in all three phases.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2005년도 하계학술대회 논문집 Vol.6
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• pp.54-56
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• 2005

In this paper, we investigated the quench characteristics of HTSC elements in the integrated three-phase flux-lock type SFCL according to fault types such as the single-line-to-ground fault, the double-line-to-ground fault, the line-to-line fault and the three-line-to-ground fault. The integrated three-phase flux-lock type SFCL was the upgrade version of the single-phase flux-lock type SFCL. The structure of the integrated three-phase flux-lock type SFCL consisted of three-phase flux-lock reactor wound on an iron core with the ratio of the same turn between coil 1 and coil 2 in each phase. When the SFCL is operated under the normal condition, the flux generated in the iron core is zero because the flux generated between two coils of each single phase is canceled out. Therefore, the SFCL's impedance is zero, and the SFCL has negligible influence on the power system. However, if a fault occurs in any single-phase among three phases, the flux generated in the iron core is not zero any more. The flux makes HTSC elements of all phases quench irrespective of the fault type, which reduces the current of fault phase as well as the current of sound phase. It was observed that the fault current limiting characteristics of the suggested SFCL were dependent on the quench characteristics of HTSC elements in all three phases.

• 공업화학
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• 제26권4호
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• pp.495-504
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• 2015

본 연구에서는 충진층 플라즈마 반응기의 특성 및 에틸렌을 분해하는데 있어서 플라즈마 반응기의 직렬 및 병렬 배열에 따른 영향에 대해 조사하였다. 플라즈마 방전 개시 전의 반응기 유효 커패시턴스는 ${\gamma}$-알루미나 펠릿이 충진된 경우가 충진되지 않은 경우보다 컸으나, 일단 플라즈마 방전이 개시되고 나면 ${\gamma}$-알루미나 충진 여부와 관계없이 유효 커패시턴스가 유사하였다. 플라즈마 상태에서 생성되는 전자의 에너지는 전기장세기에 크게 의존하며, 0~20%(v/v) 범위의 산소농도(질소 : 80~100% (v/v))에서는 기체조성에 크게 영향을 받지 않는 것으로 나타났다. 플라즈마 상태에서 생성되는 여러 활성 성분들 중 바닥상태의 산소원자 및 오존이 산화 반응에 주로 관여하며, 전기장세기가 높아질수록 산소원자가 상대적으로 감소하는 대신 질소원자의 분율이 급격히 증가한다. 플라즈마 공정에서 전압, 방전 전력, 기체 유량, 체류시간 등 반응기의 성능에 영향을 주는 여러 가지 파라미터들이 있지만, 모든 파라미터들이 비에너지밀도 하나로 통합될 수 있음을 확인하였으며, 직렬 및 병렬로 연결된 반응기의 성능도 비에너지밀도만의 함수로 간주할 수 있으므로 반응기 설계 과정이 크게 단순화될 수 있다. 비에너지밀도의 함수로 나타낸 반응속도상수를 이용하여 계산한 결과도 실험데이터를 잘 예측할 수 있었다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.