• Nuclear Engineering and Technology
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• 제27권3호
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• pp.385-392
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• 1995

월성원자력 2, 3, 4호기 safety-critical 소프트웨어에 대한 규제 기관의 요구사항을 만족시키기 위하여 소프트웨어 엔지니어링 절차가 개발되었다. 본 논문에서는 그 중에서 검증시험절차에 대하여 기술하였는데, 검증시험이란 설계그룹에서 개발된 소프트웨어가 독립된 기능그룹에서 부여한 요구사항을 모두 만족하는지를 확인하는 것이다. 이 검증시험을 수행하기 위하여 시험설비와 시험용 소프트웨어가 개발되었으며, 검증시험은 기능시험, 성능시험 및 자기점검시험 등으로 구성되었다. 시험결과를 분석하여, 불만족한 경우는 설계그룹에 통보되어 소프트웨어가 수정되었고, 최종결과는 보고서로 작성되어 규제기관에 제출될 것이다. 개발된 검증시험 방법과 절차는 효율적이고 성공적이었으며, 시험결과는 소프트웨어가 기능사양서를 충분히 만족시킨다는 것을 성공적으로 검증함을 보여주었다. 본 시험방법은 다른 safety-critical 소프트웨어 검증에도 적용될 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.17-26
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• 1996

프레팅마모 기인 연료봉 손상을 방지할 수 있는 노내 연료봉 지지조건은 잔여 지지격자스프링 변위량 또는 연료봉 /지지격자 갭에 의해 평가될 수 있다. 핵연료 설계 인자들이 프레팅마모 손상에 미치는 영향을 평가하기 위해 연소도의 함수로서 노내 연료봉 지지조건을 모사할 수 있는 방법론을 사용하여 GRID-FORCE프로그램을 개발하였다. 이 프로그램에서는 노내 연료봉 지지조건에 영향을 주는 주요 인자로서 피복관 크립, 초기 스프링 변위, 초기 스프링힘 그리고 스프링힘 조사이완이 고려된다. 이 주요 인자들에 대한 민감도 분석 결과, 초기 스프링 변위, 스프링힘 조사이완, 피복관 크립 순으로 노내 연료봉 지지조건에 영향을 주는 것으로 나타났다. 이 프로그램을 실제 노내에서 발생한 프레팅마모 기인 연료봉 손상에 적용한 결과를 토대로 판단해 볼 때 이 프로그램을 새로 개발된 피복관 재질 및 /또는 새로 개발된 지지격자 설계가 프레팅마모 기인 연료봉 손상을 방지할 수 있는 설계여유도를 효과적으로 평가할 수 있음을 알 수 있다.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.260-267
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• 2016

Background: This study analyzes the differences in the annual averaged atmospheric dispersion factor and ground deposition factor produced using two classification methods of atmospheric stability, which are based on a vertical temperature difference and the standard deviation of horizontal wind direction fluctuation. Materials and Methods: Daedeok and Wolsong nuclear sites were chosen for an assessment, and the meteorological data at 10 m were applied to the evaluation of atmospheric stability. The XOQDOQ software program was used to calculate atmospheric dispersion factors and ground deposition factors. The calculated distances were chosen at 400 m, 800 m, 1,200 m, 1,600 m, 2,400 m, and 3,200 m away from the radioactive material release points. Results and Discussion: All of the atmospheric dispersion factors generated using the atmospheric stability based on the vertical temperature difference were shown to be higher than those from the standard deviation of horizontal wind direction fluctuation. On the other hand, the ground deposition factors were shown to be same regardless of the classification method, as they were based on the graph obtained from empirical data presented in the Nuclear Regulatory Commission's Regulatory Guide 1.111, which is unrelated to the atmospheric stability for the ground level release. Conclusion: These results are based on the meteorological data collected over the course of one year at the specified sites; however, the classification method of atmospheric stability using the vertical temperature difference is expected to be more conservative.

• 한국전산구조공학회논문집
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• 제24권4호
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• pp.449-455
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• 2011

최근 가압중수형 원전 격납건물의 내압능력 및 비선형 거동에 관한 실증실험과 해석코드에 대한 검증을 위하여 인도의 BARC 주관으로 가압중수형 격납건물 1/4 축소모델을 건설하였고, 내압성능평가를 위한 국제공동연구가 수행되었다. 이 논문은 가압중수형 1/4 축소모델 격납건물에 대한 내압성능과 비선형 거동을 예측하기 위하여 유한요소해석을 수행하였고 그 결과를 도출하였다. 대상 격납건물은 기초매트와 원통형 벽체 및 돔으로 구성되어 있고, 수평 텐던의 정착을 위하여 4개의 부벽(buttress)을 가지고 있다. 유한요소해석을 위하여 ABAQUS를 이용하였고 콘크리트, 철근 및 텐던에 대한 유한요소 모델을 작성하여 극한내압해석을 수행하였다. 유한요소해석결과 콘크리트의 초기 균열은 $1.6P_d$(design pressure)에서 발생하였고, 철근의 항복은 $3.36P_d$ 그리고 극한내압능력은 $4.0P_d$ 수준으로 나타났다.

• Journal of Ginseng Research
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• 제42권3호
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• pp.379-388
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• 2018

Background: Ginsenosides are the main ingredients of Korean Red Ginseng. They have extensively been studied for their beneficial value in neurodegenerative diseases such as Parkinson's disease (PD). However, the multitarget effects of Korean Red Ginseng extract (KRGE) with various components are unclear. Methods: We investigated the multitarget activities of KRGE on neurological dysfunction and neurotoxicity in a 1-methyl-4-phenyl-1,2,3,6-tetrahydropyridine (MPTP)-induced mouse model of PD. KRGE (37.5 mg/ kg/day, 75 mg/kg/day, or 150 mg/kg/day, per os (p.o.)) was given daily before or after MPTP intoxication. Results: Pretreatment with 150 mg/kg/day KRGE produced the greatest positive effect on motor dysfunction as assessed using rotarod, pole, and nesting tests, and on the survival rate. KRGE displayed a wide therapeutic time window. These effects were related to reductions in the loss of tyrosine hydroxylase-immunoreactive dopaminergic neurons, apoptosis, microglial activation, and activation of inflammatory factors in the substantia nigra pars compacta and/or striatum after MPTP intoxication. In addition, pretreatment with KRGE activated the nuclear factor erythroid 2-related factor 2 pathways and inhibited phosphorylation of the mitogen-activated protein kinases and nuclear factor-kappa B signaling pathways, as well as blocked the alteration of blood-brain barrier integrity. Conclusion: These results suggest that KRGE may effectively reduce MPTP-induced neurotoxicity with a wide therapeutic time window through multitarget effects including antiapoptosis, antiinflammation, antioxidant, and maintenance of blood-brain barrier integrity. KRGE has potential as a multitarget drug or functional food for safe preventive and therapeutic strategies for PD.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.24-35
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• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• 정보처리학회논문지D
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• 제11D권6호
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• pp.1319-1326
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• 2004

본 논문에서는 모델링과 전산코드 실행 등에 시간과 노력이 많이 드는 핵설계 업무를 자동화하기 위하여 $IDP^{TM}$(Innovative Design Processor)를 개발하였다. IDP의 기본 원리는 문서중심 설계와 웹기반 설계이다. 문서중심 설계란 프로그래머가 아닌 일반 설계자가 동적문서(active document)라는 문서를 작성하여 이를 특수한 프로그램이 파싱후 실행하도록 하면 해석결과와 표 및 그림 둥이 담긴 완전한 설계문서를 자동적으로 얻게 됨을 말한다. 동적문서는 일반 HTML 또는 XML 편집기를 이용하여 작성될 수 있고 웹에서 또한 작성이 가능하다. LAMP(Linux/Apache/MySQL/PHP)의 웹기반 설계 환경하에서 설계마법사 형태의 서버 및 클라이언트 프로그램을 통해 설계 초보자도 쉽게 설계를 할 수 있게 되어 있다.

• Nuclear Engineering and Technology
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• 제21권1호
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• pp.40-47
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• 1989

중성입자입사 장치의 효율적인 빔형성 구조를 목적으로 정전기장 내에서 하전 입자의 움직임을 시간의 흐름에 따라 계산해 볼 수 있는 프로그램을 만들어 입자 모사 모형을 찾았다. 가속관 내의 입자의 움직임은 일정 시간 간격으로 계산하였고 전위는 유한차분법에 의해 Poisson 방정식에서 구하였다. 행렬식은 반복해법인 successive overrelaxation법을 사용하였고 전하밀도와 임자에 미치는 전기장의 힘을 구할 때는 cloud-in-cell모델을 사용하였다. 이 전자계산 코드를 사용하여 가속관 내 전극의 여러 조건들을 변화시켜가면서 빔형성 구조의 최적 설계를 수행하였다. 중성자 입사 장치의 가속관에서 가속 감속-전극간의 간격변화, 감속전극의 두께 변화, 가속 전극의 형태변화 등을 통하여 이들이 빔의 모양에 끼치는 영향을 조사하여 몇 가지 경우에 있어서 일정한 시간 간격으로 나타나는 입자들의 움직임을 예시하였다. 이 입자 모사모형을 통하여 가속전극의 형태가 빔 퍼짐에 가장 주요한 역할을 하는 것을 알았다.

• 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.