• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.59-59
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• 2002

• Journal of Information Technology Applications and Management
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• 제11권4호
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• pp.35-47
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• 2004

The nuclear design analysis requires time-consuming and erroneous model-input preparation. code run. output analysis and quality assurance process. To reduce human effort and improve design quality and productivity. Innovative Design Processor (IDP) is being developed. Two basic principles of IDP are the document-oriented desigll and the web-based design. The document-oriented design is that. if the designer writes a design document called active document and feeds it to a special program. the final document with complete analysis. table and plots is made automatically. The active documents can be written with Microsoft Word or created automatically on the web. which is another framework of IDP. Using the proper mix-up of server side and client side programming under the LAMP (Linux/Apache/MySQL/PHP) environment. it e design process on the web is modeled as a design wizard style so that even a novice designer makes the design document easily. This automation using the IDP is now being implemented for all the reload design of Korea Standard Nuclear Power Plant (KSNP) type PWRs. The introduction of this process will allow large reduction in all reload design efforts of KSNP and provide a platform for design and R&D tasks of KNFC.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.187-194
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• 2011

Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

• Nuclear Engineering and Technology
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• 제21권3호
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• pp.186-192
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• 1989

14$\times$14형 교체노심용 핵연료집합체 상단고정체 설계의 일환으로 핵연료집합체 운송 및 취급시의 하중조건하에서 여러 상단고정체 구조모델들에 대한 구조해석을 ANSYS code를 사용하여 수행하였다. 해석 모델에서는 3차원 등매개변수(isoparametric) 요소를 사용하였으며 설계조건들을 위배하지 않으면서 Adapter plate에 있는 구멍들의 배치를 조정하는 한편 Adapter plate에 부착되는 두름판(Enclosure)의 부착 방법을 개선함으로서 상단고정체의 구조적 강도를 증가시켰다. 이러한 개념들은 14$\times$14 교체노심용 핵연료 설계에 채택되었다.

• Nuclear Engineering and Technology
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• 제13권2호
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• pp.63-72
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• 1981

핵연료 집합체의 예기치 않은 파손으로 인하여 설계 근거 재장전 방침이 변경되면 원자로심내에서의 출력 분포 비균형을 막기 위하여 파손된 핵연로 집합체 이외에 대칭 위치의 집합체도 제거되어야 할 경우가 있다. 이와 같은 때에 제거된 핵연료 집합체가 설계연소도에 미달되는 경우 이를 다시 사용하여 핵연료 이용률을 증진시키는 것이 연구되었다. TDCORE 코드가 노심ㆍ해석을 위해 이용되었으며, 최적장전모형을 찾는 코드로는 RELOAD-II가 이용되었다. 고리 1호기에 적용한 결과, 제 1주기말에 제거된 비교적 적게 연소된 4개의 핵연료집합체를 제 3주기에서 이용할 경우 주기 연소도가 l1648 MWD/MTU(가동율 : 80%)에 이를 수 있음을 알 수 있었으며 평형주기까지의 장전모형을 추적하였다.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.43-53
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• 2017

Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS) and Winfrith Improved Multigroup Scheme (WIMS) codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

• 한국통신학회논문지
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• 제35권3B호
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• pp.517-525
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• 2010

IPTV는 단순한 케이블 TV 대체를 넘어서서 다양한 양방향 서비스가 가능하게 되는 서비스이다. 증가되고 있는 서비스를 위해 IPTV 시스템을 하나의 서버가 모든 일을 처리하는 구조에서 점차 분산된 구조로 가야한다. 미디어 전달 방식을 중앙 집중 서버에서 분산된 릴레이와 피어를 사용하는 구조로 변경되어야 하며 IPTV 네트워크 제어 방식도 분산된 구조로 변경하여야 한다. 본 논문은 이러한 P2P 기술을 기반으로 한 실시간 IPTV 시스템 구조에 대한 논문으로 RELOAD를 이용하여 채널 검색을 하는 분산 채널 관리 계층과 트리 구조로 미디어 전송을 하는 분산 미디어 전송 계층으로 구성된 순수 P2P기반 IPTV 구조에 대해 제안한다. 또한 실제 시스템 구현을 통해 P2P IPTV 구조의 성능을 검증한다.

• 정보보호학회논문지
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• 제30권6호
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• pp.1237-1246
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• 2020

최근 마이크로 아키텍처의 취약점을 이용하여 내부의 비밀 정보를 노출시키는 캐시 부채널 공격들이 제안되었다. 캐시 부채널 공격 중 Flush+Reload 공격은 높은 해상도와 낮은 노이즈 특성으로 인해 여러 악의적 응용 공격에 활용되고 있다. 본 논문에서는 CPU 캐시 활동을 관측할 수 있는 PCM(Performance Counter Monitor) 기능을 이용하여 캐시 기반 부채널 공격을 찾아낼 수 있는 탐지기를 구현하였다. 특히, Spectre 공격과 AES 암호 연산 중 비밀 키 추출 공격이 발생했을 때를 가정하여 PCM 카운터 값의 변화를 관측하였다. 실험 결과, PCM의 4가지 카운터 특성이 캐시 부채널 공격에 크게 반응함을 확인하였고, SVM(Support Vector Machine). RF(Random Forest), MLP(Multi Level Perceptron)와 같은 머신 러닝 기반 검출기를 통해 높은 정확도로 캐시 부채널 공격을 탐지할 수 있었다.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.238-247
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• 2020

The Burnable Poison (BP) is very important for all Light Water Reactors in order to hold-down the initial excess reactivity and to control power peaking. The use of BP is even more essential as the excess reactivity increases significantly with a longer operation cycle. In this paper a feasibility study was conducted in order to investigate the benefits of a new combinational BP concept designed for 24-month cycle PWR core. The reference designs in this study are based on the two Korean fuel assemblies; 17 × 17 Westinghouse (WH) design and 16 × 16 Combustion Engineering (CE) design. A modification was done on these two designs to extend their cycle length from 18 months into 24 months. DeCART2D-MASTER code system was used to perform assembly and core calculations for both designs. A preliminary test was conducted in order to choose the best BP suitable for 24-month as a representative for single BP concept. The comparison between the results of two concepts (combinational BP concept and single BP concept) showed that the combinational BP concept can replace the single BP concept with better performance on holding down the initial excess reactivity without violating the design limitations.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.257-266
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• 1999

A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable.