• 한국압력기기공학회 논문집
• /
• 제7권3호
• /
• pp.15-22
• /
• 2011

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

• 한국지능시스템학회:학술대회논문집
• /
• 한국퍼지및지능시스템학회 1993년도 Fifth International Fuzzy Systems Association World Congress 93
• /
• pp.1106-1108
• /
• 1993

This paper deals with an industrial application of a fuzzy feedback combined learning control to an industrial batch free radical polymerization reactor. As a result, the plant has reduced the batch reaction time by 50 minute and stabilized both by 40 percent reduction of the standard deviations of product qualities, such as the total solid content and the graft gum, and by 45 percent reduction of the standard deviation of the batch reaction end time.

• 한국전산유체공학회:학술대회논문집
• /
• 한국전산유체공학회 2005년도 추계 학술대회논문집
• /
• pp.45-47
• /
• 2005

We have investigated the mixing behavior of a tubular heat exchanger reactor using CFD and compared its mixing performance with different type of reactors such as jet mixer and continuous stirred tank reactor (CSTR). The mixedness in each reactor was quantified introducing a statistical average value, the coefficient of variation (CoV), which is a normalized standard deviation of concentration of a component over the whole fluid domain. Through the analysis of the flow pattern and turbulent energy distribution, we suggested a simple but effective way to improve the mixing performance of the tubular heat-exchanger reactor, which include the addition of the internals and/or the increase of the recycle flow rate. It was found that the CoV value of the tubular reactor could be nearly equivalent to that of CSTR by applying those two alternatives suggested here.

• Nuclear Engineering and Technology
• /
• 제52권9호
• /
• pp.2130-2137
• /
• 2020

The standard method for calculating anti-neutrino emissions from a reactor involves knowing the fractional fission rates for the most important fissioning nuclides in the reactor. To calculate these rates requires detailed reactor physics calculations based upon the reactor design, fuel design, burnup dependent fuel composition, location of specific fuel assemblies in the core and detailed operational data from the reactor. This has only been published for a few reactors during specific time periods, whereas to be of practical use for anti-neutrino reactor monitoring it is necessary to be able to predict these on the publicly available information from any reactor, especially if using these data to subtract the anti-neutrino signal from other reactors to identify an undeclared reactor and monitor its operation. This paper proposes a method to estimate the fission fractions for a specific reactor based upon publicly available information and provides a database based upon a series of spent fuel inventory calculations using the FISPIN10 code and its associated data libraries.

• Nuclear Engineering and Technology
• /
• 제34권3호
• /
• pp.259-267
• /
• 2002

NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2011년도 제42회 하계학술대회
• /
• pp.1814-1815
• /
• 2011

• 한국경영과학회:학술대회논문집
• /
• 한국경영과학회 2003년도 추계학술대회 및 정기총회
• /
• pp.267-270
• /
• 2003

The quality assurance activities are important to obtain the safety and reliability in decommissioning of research reactor. It is essential to establish and implement effective quality assurance program. Foreign state-of-the-art standards and practices of quality assurance are investigated and analyzed to select quality assurance requirements. In this paper, guidelines are offered a suggestion to establish optimal the technology standard and lay out the managerial control scheme of quality assurance for decommissioning on research reactor.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• pp.707-712
• /
• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.

• 소음진동
• /
• 제7권1호
• /
• pp.33-41
• /
• 1997

경수로형 원자력발전소 표준화 작업의 일환으로 만들어진 한국표준형 원자력 발전소는 그 건설부지를 한반도뿐만 아니라 인접 아시아국가의 여러곳을 목표로 하고 있으며 이와 관련하여 안전정지지진의 레벨을 0.3g로 증가시키려는 시도가 계획되고 있다. 본 연구에서는 이와 같은 지진레벨 증가가 기존의 0.2g로 설계된 원자로 내부 구조물과 핵연료집합체에 미치는 영향을 평가하였다. 운전기준지진 및 안전정지지진의 응답을 비교함으로써 비선형 응답특성을 조사하였고 한국표준형 원자력발전소의 원자로 내부구조물 및 핵연료집합체의 설계 타당성에 대하여 언급하였다.

• Nuclear Engineering and Technology
• /
• 제44권6호
• /
• pp.689-696
• /
• 2012

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.