• Nuclear Engineering and Technology
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• 제51권1호
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• 대한기계학회논문집
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• 제14권1호
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• pp.112-118
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• 1990

본 연구에서는 ASTM-E24.01.06에서 제안하고 있는 실험방법을 응용하여 균열 정지 파괴인성값을 측정하였다.즉 쐐기와 분리형 부싱(wedge and split bushing)으 로 압축하중을 가함으로 균열선 웨지하중 시편[crack line wedge loaded specimen(CL- WL시편)]에 인장력을 발생시켜서 균열정지 응력확대계수( $K_{1a}$)를 결정하였다. 그리고 균열개시 응력확대계수가 균열정지 응력확대계수에 미치는 영향들을 여러가지 재료들에 대하여 체계적으로 검토하였다.다.

• 한국초전도ㆍ저온공학회논문지
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• 제13권1호
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• pp.41-45
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• 2011

Hybrid fault current limiters (FCL) have been researched at Yonsei University. The hybrid FCL has advantages such as having a rapid response to a sudden fault situation and a fast recovery time from a quench. It consists of an asymmetric HTS coil, a switching module, and a bypass reactor. The asymmetric HTS coil is wound with two different types of HTS wires in an opposite direction so that it has nearly zero inductance at the superconducting state. When the quench occurs at the fault state, a strong magnetic field is generated from the asymmetric coil because of different quench characteristics of two HTS wires, and then a repulsive force is induced in the switching module. The force opens the switch and the fault current is pushed into the bypass reactor. In this research, we analyzed the cause of the repulsive force and confirmed, experimentally and computationally, that the magnitude of a repulsive force is varied by changing the gap distance between the asymmetric coil and the switching module. By using the FEM simulation, we calculated the repulsive force with respect to the gap distance and verified that the effect of the gap distance. Then, short circuit test was carried out to confirm the correct operation of the fast switch.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3286-3297
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• 2024

Traditional text-based system engineering, which has been used in the design and application of passive residual heat removal system (PRHRS) for lead-cooled fast reactors, is prone to several problems such as low development efficiency, long iteration cycles, and model ambiguity. This study aims to effectively address the abovementioned problems by adopting a model-based system engineering (MBSE) method, which has been preliminarily applied to meet the design requirements of a PRHRS. The design process has been implemented based on the preliminary design of the system architecture and consists of three stages: top-level requirement analysis, functional requirements analysis, and design requirements synthesis. The results of the top-level requirements analysis and the corresponding use case diagram can determine the requirements, top-level use cases, and scenario flow of the system. During the functional requirements analysis, the sequence, activity, and state machine diagrams are used to develop the system function model and provide early confirmation. By comparing these sequence diagrams, the requirements for omissions and inconsistencies can be effectively checked. In the design requirements synthesis stage, the Analytic Hierarchy Process is used to conduct preliminary trade-off calculations on the system architecture, after which a white box model is established during the system architecture design. Through these two steps, the analysis and design of the system architecture are ultimately achieved. The resulting system architecture ensures the consistency of the design requirements. Ultimately, a functional hazard analysis was conducted for a specific incident to validate case requirements and further refine the system architecture. Future research can further reduce the design risk, improve the design efficiency, and provide a practical reference for the design and optimization of PRHRS in digital lead-cooled fast reactors.

• 청정기술
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• 제24권3호
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• pp.233-238
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• 2018

급속열분해를 통한 바이오-오일 생산 공정은 무산소 조건에서 바이오매스를 급속열분해하여 얻어진 열분해가스를 급속 냉각 시켜 열분해오일을 생산한다. 이에 공정 내부의 산소 농도를 0 ~ 3% 이하로 유지하기 위해 캐리어 가스로 질소를 사용한다. 그러나 공정의 규모가 커질수록 질소의 사용량이 증가하고, 이는 공정 운전비용 증감 및 지속적인 질소 가스 충전을 위한 설비비 증감 할 수밖에 없다. 이에 본 연구에서는 팜 부산물 열분해에서 질소 사용량 감소를 위해 미응축가스 재순환 공정을 적용하여, 가스재순환율에 따른 질소 사용량과 미응축가스의 가연성 성분의 농도 변화를 측정하고 이에 따른 바이오-오일의 품질 수율 변화를 측정하여 가스재순환 공정의 활용 가능성을 연구하였다.

• 전력전자학회논문지
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• 제26권1호
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• pp.25-31
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• 2021

This paper describes the design of a power cell-based pulsed power modulator with fast rise times. The pulse-generating section of the pulse power modulator is a series stack of power cells. Each power cell is composed of a storage capacitor, a pulse switch, and a bypass diode. When the pulse switches are turned on, the capacitors are connected in series and the sum of voltages is applied to the load. For output pulses with fast rise times, an IGBT with fast turn-on characteristics is adopted as a pulse switch and the optimized gate driving method is used. Pspice simulation is performed to account for the gate driving method. A 10 kV, 12-power cell-based pulsed power modulator is tested under resistive load and plasma reactor load. The rise times of output pulses less than 20 ns are confirmed, showing that the pulsed power modulator can be effectively applied to pulsed power applications with fast rise times.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4518-4526
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• 2023

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutron-moderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 춘계학술대회 논문집
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• pp.743-744
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• 2012

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.157.2-157
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• 2002