• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1126-1139
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• 2016

In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor ($K_{eff}$) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

• Journal of Electrical Engineering and Technology
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• 제9권4호
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• pp.1309-1314
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• 2014

The measurement of stator core loss for an induction motor at each manufacturing process is carried out in this paper. Iron loss in the stator core of induction motor changes after each manufacturing process due to the mechanical stress, which can cause the deterioration of the magnetic performances. This paper proposes a new iron loss measuring system of the stator core in an induction motor, which can be applied to the case when the distribution of magnetic flux density is not uniform along the magnetic flux path. In the system, the iron loss is calculated based on the induced voltage of the B-search coil and exciting current.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• 한국초전도ㆍ저온공학회논문지
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• 제12권4호
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• pp.24-30
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• 2010

The renewable energy source is considered as a good measure to cope with the global warming problem and the fossil energy exhaustion. The construction of electric power plant such as an offshore wind farm is rapidly increasing and this trend is expected to be continued during this century. The bulky and long distance power transmission media is essential to support and promote the sustainable expansion of renewable energy source. DC power cable is generally considered as the best solution and the demand for DC electric power has been rapidly increasing. Especially, the high temperature superconducting (HTS) DC cable system begins to make a mark because of its advantages of huge power transmission capacity, low transmission loss and other environmental friendly aspects. Technical contents of DC HTS cable system are very similar to those of AC HTS cable system. However the DC HTS cable can be operated near its critical current if the heat generation is insignificant, while the operating current of AC HTS cable is generally selected at about 50~70% of the critical current because of AC loss. We chose the specifications of the cable core of 'Tres Amigas' project as an example for our study and investigated the heat generation when the DC HTS cable operated near the critical current by some electric and thermal analyses. In this paper, we listed some technical issues on the design of the DC HTS cable core and described the process of the cable core design. And the results of examination on the current capacity, heat generation, harmonic loss and current distribution properties of the DC HTS cable are introduced.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1871-1885
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• 2019

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic P_N scattering, and the multi-group cross-section generation model was performed on the k_eff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a k_eff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• KSII Transactions on Internet and Information Systems (TIIS)
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• 제14권8호
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• pp.3366-3383
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• 2020

The Poisson point process (PPP) is widely used in wireless network modeling and performance analysis because it can provide tractable results for heterogeneous cellular networks (HetNets) analysis. However, it cannot accurately reflect the spatial distribution characteristics of the actual base stations (BSs). Considering the fact that the distribution of macro base stations (MBSs) is exclusive, the deployment of MBSs is modeled by the Matérn hard-core point process (MHCPP), and the deployment of pico base stations (PBSs) is modeled by PPP. This paper studies the performance of multiantenna HetNets and improves the energy efficiency (EE) of HetNets by optimizing the transmit power of PBSs. We use a simple approximate method to study the signal-to-interference ratio (SIR) distribution in two-tier MHCPP-PPP HetNets and derive the coverage probability, average data rate and EE of HetNets. Then, an optimization algorithm is proposed to improve the EE of HetNets. Finally, three transmission technologies are simulated and analyzed. The results show that multiantenna transmission has better system performance than single antenna transmission and that selecting the appropriate transmit power for a PBS can effectively improve the EE of the system. In addition, two-tier MHCPP-PPP HetNets have higher EE than two-tier PPP-PPP HetNets.

• 대한기계학회논문집B
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• 제37권9호
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• pp.855-862
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• 2013

개방 노심 열적여유도 해석 코드에 입력으로 제공되는 APR+ (Advanced Power Reactor Plus)의 수력학적 특징을 결정하기 위해 일련의 1/5 축소 원자로 유동분포 시험이 수행되었다. 본 연구에서는 원자로 내부 유동 계산시 다공성 모델을 사용한 전산유체역학의 적용성을 평가하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX V.14를 사용하여 계산을 수행하였다. 결론적으로 본 연구에서 사용한 일부 원자로 내부 구조물에 대한 다공성 영역 처리방식을 통해 원자로 내부의 유동 특성을 정성적으로 적절히 파악할 수 있을 것으로 판단된다. 만일 충분한 계산 자원이 확보된 조건인 경우라면 노심 입구 상류에 위치한 원자로 내부 구조물의 실제 기하 형상을 고려함으로써 노심 입구 유량분포를 보다 정확하게 예측할 수 있을 것으로 예상된다.

• 대한전기학회논문지:시스템및제어부문D
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• 제54권6호
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• pp.409-416
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• 2005

For broadband high-data-rate power line communication with the allocated frequency bandwidth from 2 to 30 MHz on medium voltage (MV) distribution power lines, a signal coupling unit is developed. The coupling unit is composed of a coupling capacitor for coupling communication signal, a drain coil, and an impedance matching part. The coupling capacitor made of ceramic capacitor is designed for transmission property of better than 1 dB in the frequency range. The drain coil is used for preventing low frequency high voltage from junction of medium voltage power line in case that a coupling capacitor is not working properly any more. Also, using ferrite core, a novel broadband impedance matching transformer is developed. A complete coupling unit with a coupling capacitor, a drain coil, and a matching transformer is housed by polymer for good isolation and distinguishing from high voltage electric facilities. Each is fabricated and its frequency behavior is tested. Finally, complete signal couplers are equipped in a MV PLC test bed and their performance are measured. The measurement shows that the coupling capacitor works excellently.