• Nuclear Engineering and Technology
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• 제46권6호
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• pp.737-752
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• 2014

As condition-based maintenance (CBM) has risen as a new trend, there has been an active movement to apply information technology for effective implementation of CBM in power plants. This motivation is widespread in operations and maintenance, including monitoring, diagnosis, prognosis, and decision-making on asset management. Thermal efficiency analysis in nuclear power plants (NPPs) is a longstanding concern being updated with new methodologies in an advanced IT environment. It is also a prominent way to differentiate competitiveness in terms of operations and maintenance costs. Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.896-904
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• 2017

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.855-865
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• 2024

Commercial deployment of nuclear microreactors presents an opportunity for the industry to rethink its approach to manufacturing, siting, operation and maintenance, and fuel cycle management as certain principles used in grid-scale nuclear projects are not applicable to a decentralized microreactor economy. The success of this nascent industry is dependent on its ability to reduce infrastructure, logistical, regulatory and lifecycle costs. A utility-like 'Central Facility' that consolidates the services required and responsibilities borne by vendors into one or a few centralized locations will be necessary to support the deployment of a fleet of microreactors. This paper discusses the requirements for a Central Facility, its implications on the cost structures of owners and suppliers of microreactors, and the impact of the facility for the broader microreactor industry. In addition, this paper discusses the pre-requisites for eligibility as well as the opportunities for a Central Facility host site. While there are many suitable locations for such a capability across the U.S., this paper considers a facility co-located with the Vogtle Nuclear Power Plant and Savannah River Sites to illustrate how a Central Facility can leverage the existing infrastructure and stimulate a local ecosystem.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2888-2898
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• 2021

Vital area identification (VAI) is an essential procedure for the design of physical protection systems (PPSs) for nuclear power plants (NPPs). The purpose of PPS design is to protect vital areas. VAI has been improved continuously to overcome the shortcomings of previous VAI generations. In first-generation VAI, a sabotage fault tree was developed directly without reusing probabilistic safety assessment (PSA) results or information. In second-generation VAI, VAI model was constructed from all PSA event trees and fault trees. While in third-generation VAI, it was developed from the simplified PSA event trees and fault trees. While VAIs have been performed for NPPs in full-power operations, VAI for NPPs in low-power and shutdown (LPSD) operations has not been studied and performed, even though NPPs in LPSD operations are very vulnerable to sabotage due to the very crowded nature of NPP maintenance. This study is the first to research and apply VAI to LPSD operation of NPP. Here, the third-generation VAI method for full-power operation of NPP was adapted to the VAI of LPSD operation. In this study, LPSD VAI for a few plant operational states (POSs) was performed. Furthermore, the operation strategy of vital areas for both full-power and LPSD operations was discussed. The LPSD VAI method discussed in this paper can be easily applied to all POSs. The method and insights in this study can be important for future LPSD VAI that reflects various LPSD operational states. Regulatory bodies and electric utilities can take advantage of this LPSD VAI method.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.621-631
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• 2017

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

• 한국국방경영분석학회지
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• 제16권2호
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• pp.14-28
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• 1990

The objective of this study is to propose how to acquire through international cooperation the sensitive nuclear technology, so called reprocessing technology. In spite of the need to reuse spent fuel, the transfer of the sensitive technology has been tightly controlled by the nuclear advanced countries due to the fear of nuclear proliferation and, in fact, it would be impossible to secure it by the economic means. In this regard, as a means of acquiring the sensitive nuclear technology, this study proposes the following; 1) President's declaration concerning the peaceful uses of nuclear energy, 2) the establishment and maintenance of national basis through inter-ministerial cooperation, 3) as a confidence building measure, the efforts to strengthen our role in the international nuclear community, and 4) the establishment of the synthetic feedback system to efficiently coordinate. In line with those stated above, this study suggests that it be necessary to invest consistently for developing new technologies and cultivating human resources. Furthermore, this study proposes the necessity to resolve the problems lying ahead by the national consensus achieved through the discussions among the public concerning the sensitive nuclear technology.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2011년도 춘계학술발표대회
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• pp.547-550
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• 2011

The purpose of this paper is to verify and validate the maintenance tasks of the construction of a nuclear facility using a digital mock-up and simulation technology instead of a physical mock-up. Prior to the construction of a nuclear facility, a remote simulator that provides the opportunity to produce a complete digital mock-up of the PRIDE (Pyroprocess Integrated Inactive DEmonstration Facility) region and its remote handling equipment, including operations and maintenance procedures has been developed. In this paper, the system architecture and graphic user interface of a remote simulator that coincides with the extraordinary nature of a nuclear fuel cycle facility is introduced. In order to analyze the remote accessibility of a remote manipulator, virtual prototyping that was performed it by using haptic device of external input devices under a 3D full-scale digital mock-up is explained.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2262-2273
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• 2020

Digital twin technology can provide significant value for the prognostics and health management (PHM) of critical plant components by improving insight into system design and operating conditions. Digital twinning of systems can be utilized for anomaly detection, diagnosis and the estimation of the system's remaining useful life in order to optimize operations and maintenance processes in a nuclear plant. In this regard, a conceptual framework for the application of digital twin technology for the prognosis of Control Element Drive Mechanism (CEDM), and a data-driven approach to anomaly detection using coil current profile are presented in this study. Health management of plant components can capitalize on the data and signals that are already recorded as part of the monitored parameters of the plant's instrumentation and control systems. This work is focused on the development of machine learning algorithm and workflow for the analysis of the CEDM using the recorded coil current data. The workflow involves features extraction from the coil-current profile and consequently performing both clustering and classification algorithms. This approach provides an opportunity for health monitoring in support of condition-based predictive maintenance optimization and in the development of the CEDM digital twin model for improved plant safety and availability.

• 전력전자학회:학술대회논문집
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• 전력전자학회 2018년도 전력전자학술대회
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• pp.177-179
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• 2018

After Fukushima nuclear accident at 2011, nuclear industrial has been focused on operation and maintenance phase, not design and construction phase. Continued good operating performance of nuclear power plants has been the best critical issue to nuclear utilities. Replacement for complete components as well as parts of components is being procured because nuclear utilities must maintain safety and reliability of operating nuclear power plants. However, many suppliers and manufacturers are giving up a nuclear quality assurance program under reduction in new construction of nuclear power plants. It is able to be increased difficulty in procuring spare parts to support operations and maintenance of nuclear power plants. Over 20% of nuclear power plant equipment in some countries is obsolete. Owing to obsolescence of nuclear safety-related items and/or withdrawing a nuclear quality assurance program of suppliers and manufactures, some replacement item and part might be procured to the item not covered by appendix B to USNRC 10 CFR Part 50. Under various methods of the nuclear repair and replacement methodology, utilities are supposed to establish a typical program for a repair and replacement of an electrical equipment and its parts in conjunction with a nuclear quality assurance. Concerning this typical program, this study suggests the repair and replacement methodology of electrical equipments used in nuclear power plants by procurement of a power supply, based on nuclear regulations, codes, standards, guidelines, specific and general technical information, etc..

• Journal of Radiation Protection and Research
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• 제34권4호
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• pp.176-183
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• 2009

원전 계획예방정비기간 증기발생기 수실작업 등은 매우 높은 방사선량율을 보이는 지역으로, 짧은 시간 동안 작업으로 종사자는 높은 피폭을 받을 가능성이 있다. 특히, 방사성물질과 접촉작업을 하는 손 부위에서 고피폭이 예상된다. 이런 점을 고려하여 2004년 수행된 국내 원전의 복수선량계 알고리즘 적용성 시험의 TLD 판독결과를 이용하여 고피폭 접촉 작업의 방사선장을 분석하였다. 그 결과, 원전 고피폭 접촉작업의 입사방사선장은 고에너지 광자(High Energy Photon Field)에 의한 피폭으로 해석되었다. 한편 2009년 울진 4호기 계획예방정비기간 S/G 정비작업과 월성 1호기 압력관 교체작업에 참여한 방사선작업종사자에 대해 말단선량 현황과 방사선장을 분석하기위한 현장시험을 실시하였다. 그 결과 입사방사선장은 고에너지 방사선장으로 확인되었다.