• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3106-3116
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• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.19-26
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• 1988

저누출 장전 모형은 새 연료를 안에서부터 넣는 in-out 형태를 취하여 격납 용기의 fluence를 줄이고 중성자 경제성을 높이고자 하는 것으로, 이 경우에는 노심내의 전체적인 중성자 경제성은 좋아지지만 노심 중앙부에서의 새연료의 과다 반응도 때문에 안전성 여유를 줄이게 되므로 많은 수의 가연성 독붕봉을 사용하여 첨두 인자를 조절해야만 한다. 본 논문에서는 가연성 독붕봉 연소에 따른 출력 변화를 섭동으로 취급하며, 이를 출력감도 계수(Power Sensitivity Coefficient)로 표시한다. 최적화된 가연성 독붕봉의 분포를 구하기 인하여 알고 있는 주기말상태로부터, 노심 내의 출력과 과다 반응도를 제어하면서 주기초로 추적해 나가는 역연소법(Reverse Depletion Method)의 도입에 대한 타당성을 출력 민감도 계수개념과 선형 계획법을 이용하여 원자력 7호기 제1주기에 응용하여 검증했으며, 가연성 독붕봉의 추정량과 실제량과의 차이는 최대 4.5%의 오차를 보였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 제36회 하계학술대회 논문집 A
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• pp.329-331
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• 2005

This paper proposes a line-to-ground cable fault location method for underground distribution system. The researched cable is composed of core and sheath. And underground cabke system has been analyzed using Distributed Parameter Circuit. The effectiveness of proposed algorithm has been verified through EMTDC simulations.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1429-1435
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• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

• 유변학
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• 제11권2호
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• pp.159-167
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• 1999

샌드위치 사출성형 공정은 기존의 사출성형 공정이 가지지 못하는 여러 장점들로 인해 최근 산업적으로 주목 받고 있는 고분자 가공 공정이다. 이 공정의 해석적인 접근은 거의 불가능하므로, 본 연구에서는 수치모사를 통해서 샌드위치 사출성형의 충전 공정을 연구하였다. 수치모사는 기본적으로 유한요소법을 사용하였고 Flow Analysis Network(FAN)/관할체적(Control Volume)법 등을 함께 이용하였다. 그리고 skin polymer의 선단을 확인할 수 있는 기존의 충전율 변수와 함께 skin polymer와 core polymer의 경계를 표시하는 새로운 충전율 변수를 도입하였고 이것을 이용하여 core polymer의 선단을 추적하였다. 새로운 충전율 변수는 두께 방향으로 온도장을 풀기 위해 나눈 각 층에서 정의되었다. 수치모사에 사용된 skin polymer와 core polymer로는 물성이 다른 두 고분자 물질을 주입시켜서 나타나는 충전 형태를 비교했다. 즉, 점도 상수, power-law 지수 등과 같은 유변 물성이 다른 두 고분자 물질을 충전시키기 위해 공정상 필요한 입구에서의 압력 등을 계산했으며 나중에 들어가게 되는 core polymer의 충전 완료 후 금형 내에서의 두께 방향과 흐름 방향으로의 분포 등을 구하였다. 또한 실제 공정 상에서 가공조건에 해당되는 switchover time과 벽 온도 등의 조건을 바꿔가면서 수치모사를 진행하였다. 사례 연구를 통하여 얻어진 물성과 가공 조건에 따른 core polymer의 충전 형태와 입구에서의 압력 등은 샌드위치 사출성형의 산업적 이용에 매우 유용하게 사용될 수 있다.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.538-544
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• 1996

This paper presents the results of the core follow analysis for Yonggwang Unit 3 Cycle 1. The values of peaking factors (Fxy, Fq, Fr anf Fz) and core power distribution measured and processed by CECOR code［1］ are compared with those predicted by ROCS code［2］, The measured boron rundown is also compared with the predicted values. As results, the comparison of peaking factors, radial and axial power distributions and boron rundown between the measured and the predicted show good agreement throughout the cycle. Additionally, assembly burnup differences between CECOR and ROCS at EOC1 (13650 MWD/MTD are within 5% of core average burnup.

• 전력전자학회:학술대회논문집
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• 전력전자학회 2006년도 전력전자학술대회 논문집
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• pp.470-474
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• 2006

This dissertation discuses about a ferrite core plasma source using low operating frequency without sputtering problem by the stored electric field. Compared with the conventional RF power system with 13.56MHz switching frequency, the proposed plasma power system is only separated at 400kHz, so that it makes possible to use of low cost switching elements, PWM control and soft switching. Moreover, it could improve the coupling efficiency for plasma and antenna by using the ferrite core in order to transfer the energy of the load This dissertation tried to analyze new plasma generation method for the plasma generation system by modeling the plasma load and grafting the concept of impedance matching in order to interpret it with the formula This dissertation verified the ferrite core inductive coupling plasma source authorized for 400kHz of low frequency power by applying to the semi-conductor ash process thru the measurement of ash capacity and uniformed plasma distribution on the actual wafer.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.335-341
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• 1997

The generic analysis method for core flow instability is suggested to confirm that the core flow instability would not occur on PWR conditions. For the confirmation, the stability criteria of each fuel type are provided. Instability investigations in various accident conditions prove that the locked rotor accident is the most limiting case to instability. Parametric Effects are surveyed and in good agreement with available studies. The effects of heat flux distribution become negligible as the subcooling number is decreased. The power margin to instability is calculated quantitatively in various accident conditions.