• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1264-1272
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• 2016

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.781-786
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• 2017

The presence of crust on the inner walls of metallic ducts impairs transportation because crust completely or partially hinders the passage of fluid to the processing unit and causes damage to equipment connected to the production line. Its localization is crucial. With the development of the electronic imaging system installed at the Argonauta/Nuclear Engineering Institute (IEN)/National Nuclear Energy Commission (CNEN) reactor, it became possible to visualize crust in the interior of metallic piping of small diameter using real-time neutron radiography images obtained with a low neutron flux. The obtained images showed the resistance offered by crust on the passage of water inside the pipe. No discrepancy of the flow profile at the bottom of the pipe, before the crust region, was registered. However, after the passage of liquid through the pipe, images of the disturbances of the flow were clear and discrepancies in the flow profile were steep. This shows that this technique added the assembled apparatus was efficient for the visualization of the crust and of the two-phase flows.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.579-588
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• 2022

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for the formulation of severe accident mitigation measures, the article select three experiments to evaluate the updated severe accident models of MIDAC. Among them, QUENCH-06 is the international standard No.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod melting experiment recently carried out by Xi'an Jiaotong University. The heating and melting model with lumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarev correlations combination in MIDAC could better meet the needs of severe accident analysis. Although the influence of nitrogen still need to be further improved, the air oxidation model with NUREG still has the ability to provide guiding significance for engineering practice.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1080-1082
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• 2003

To develop a reliability analysis program applied to the diagnosis for air operated valve's integrity. we collected, analyzed AOV failure data from foreign and domestic nuclear power plants, and classified whole subjects of this program into several groups according to type and size. We established a theoretical basis using Lognormal Distribution and Bayesian Theory to develop analysis methodology. The result of this program was applied to the calculation of operational unavailability of AOV, and the effect of AOV's failure. Also this program can be applied to the development of diagnostic technique considering AOV environment (temperature, pressure), and setting-up maintenance cycle.

• Nuclear Engineering and Technology
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• 제38권7호
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• pp.591-618
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• 2006

An integrated R&D program is being conducted to study, qualify, and in some cases, develop materials with required properties for the reactor systems being developed as part the U.S. Department of Energy's Generation IV Reactor Program. The goal of the program is to ensure that the materials research and development (R&D) needed to support Gen IV applications will comprise a comprehensive and integrated effort to identify and provide the materials data and its interpretation needed for the design and construction of the selected advanced reactor concepts. The major materials issues for the five primary systems that have been considered within the U.S. Gen IV Reactor Program-very high temperature gas-cooled, supercritical water-cooled, gas-cooled fast spectrum, lead-cooled fast spectrum, and sodium-cooled fast spectrum reactors-are described along with the R&D that has been identified to address them.

• 시스템엔지니어링학술지
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• 제12권1호
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• pp.73-87
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• 2016

Combining power generation and water production by desalination is economically advantageous. Most desalination projects use fossil fuels as an energy source, and thus contribute to increased levels of greenhouse gases. Environmental concerns have spurred researchers to find new sources of energy for desalination plants. The coupling of nuclear power production with desalination is one of the best options to achieve growth with lower environmental impact. In this paper, we will per-form a sensitivity study of coupling nuclear power to various combinations of desalination technology: {1} thermal (MSF [Multi-Stage Flashing], MED [Multi-Effect Distillation], and MED-TVC [Multi-Effect Distillation with Thermal Vapour Compression]); {2} membrane RO [Reverse Osmosis]; and {3} hybrid (MSF-RO [Multi-Stage Flashing & Reverse Osmosis] and MED-RO [Multi-Effect Distillation & Reverse Osmosis]). The Korean designed reactor plant, the APR1400 will be modeled as the energy production facility. The economical evaluation will then be executed using the computer program DEEP (Desalination Economic Evaluation Program) as developed by the IAEA. The program has capabilities to model several types of nuclear and fossil power plants, nuclear and fossil heat sources, and thermal distillation and membrane desalination technologies. The output of DEEP includes levelized water and power costs, breakdowns of cost components, energy consumption, and net saleable power for any selected option. In this study, we will examine the APR1400 coupled with a desalination power plant in the Kingdom of Saudi Arabia (KSA) as a prototypical example. The KSA currently has approximately 20% of the installed worldwide capacity for seawater desalination. Utilities such as power and water are constructed and run by the government. Per state practice, economic evaluation for these utilities do not consider or apply interest or carrying cost. Therefore, in this paper the evaluation results will be based on two scenarios. The first one assumes the water utility is under direct government control and in this case the interest and discount rate will be set to zero. The second scenario will assume that the water utility is controlled by a private enterprise and in this case we will consider different values of interest and discount rates (4%, 8%, & 12%).

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.555-561
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• 1994

핵연료 건전성 점검을 위하여 다중채널분석기로 얻은 감마선 스펙트럼을 자동으로 빨리 분석하는 프로그램을 개발하였다. 핵연료의 건전성은 실시간 감시와 주기적인 시료 분석을 통한 원자로냉각재내의 방사선준위로 확인된다. 영광 3·4 호기의 경우, 실시간 감시 계통인 프로세스 방사선 감시 계통(PRMS)이 핵연료의 건전성을 확인한다. 현재, PRMS의 스펙트로미터 채널의 신호처리기는 단일채널 분석기이어서 오직 하나의 방사성핵종만을 파악할 수 있다. 따라서 PRMS를 개선하기 위해서는 단일채널분석기를 다중채널분석기로 대치하여야 한다. 이 프로그램은 실시간 모드와 수동모드로 실행되며, 모든 과정을 자동으로 수행한다. 미 국가표준국의 혼합 표준 선원에 대한 시험 결과는 상용 다중채널분석기인 Canberra System 100의 결과와 잘 일치하였다. 결론적으로 개발된 프로그램은 원자력발전소의 감마선 감시에 사용할 수 있을 것으로 보인다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.555-561
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• 1995

이 논문에서는 영광 3호기와 같이 노심내 핵계장 장치를 갖고 있는 가압형 원자력 발전소용 노심출력 분포 감시프로그램의 개발에 수정형 Borresen 모형의 응용 타당성을 검토해 보았다 이를 위해 수정형 Borresen 방정식의 소격모형해를 핵계장장치의 측정치를 경계조건으로 하여 풀었으며, 이로부터 영광 3호기 첫주기 노심의 3차원 출력분포를 계산하였다. 그 결과는 현재 영광 3호기의 축방향 출력분포 감시프로그램으로 활용되고 있는 COLSS 예측치와 비교하였으며, 이를 통하여 수정형 Borresen 모형으로 제안한 방법이 COLSS보다 축방향 출력분포를 실제에 더 가깝게 모사할 수 있음을 보였다. 노심 출력거동에 대한 예측능력이 있고 또한 전산속도면에서의 이점이 있어서 제안된 수정형 Borresen 방법이 노심출력분포 감시프로그램 개발에 유용하게 활용될 수 있다고 결론을 내렸다.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• 제1권1호
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• pp.9-27
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• 2013

Nagra was established in 1972 by the Swiss nuclear power plant operators and the Federal Government to implement permanent and safe disposal of all types of radioactive waste generated in Switzerland. The Swiss Nuclear Energy Act specifies that these shall be disposed of in deep geological repositories. A number of different geological formations and sites have been investigated to date and an extended database of geological characteristics as well as data and state-of-the-art methodologies required for the evaluation of the long-term safety of repository systems have been developed. The research, development, and demonstration activities are further supported by the two underground research facilities operating in Switzerland, the Grimsel Test Site and the Mont Terri Project, along with very active collaboration of Nagra with national and international partners. A new site selection process was approved by the Federal Government in 2008 and is ongoing. This process is driven by the long-term safety and feasibility of the geological repositories and is based on a step-wise decision-making approach with a strong participatory component from the affected communities and regions. In this paper a brief history and the current status of the Swiss radioactive waste management program are presented and special characteristics that may be useful beyond the Swiss program are highlighted and discussed.