scholarly journals SMALL NUCLEAR POWER PLANTS. VOLUME ONE. DESIGN, CONSTRUCTION, AND OPERATING EXPERIENCE.

1966 ◽  
Author(s):  
S F Burton ◽  
A G Hosler
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Design Construction

Digital I&C Operating Experience in the US

2008 ◽  
Author(s):  
Bruce Geddes ◽  
Ray Torok
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Failure Modes ◽  
Operating Experience ◽  
Data Set ◽  
Common Cause ◽  
Hardware Failure ◽  
Common Cause Failure ◽  
Defensive Measures

The Electric Power Research Institute (EPRI) is conducting research in cooperation with the Nuclear Energy Institute (NEI) regarding Operating Experience of digital Instrumentation and Control (I&C) systems in US nuclear power plants. The primary objective of this work is to extract insights from US nuclear power plant Operating Experience (OE) reports that can be applied to improve Diversity and Defense in Depth (D3) evaluations and methods for protecting nuclear plants against I&C related Common Cause Failures (CCF) that could disable safety functions and thereby degrade plant safety. Between 1987 and 2007, over 500 OE events involving digital equipment in US nuclear power plants were reported through various channels. OE reports for 324 of these events were found in databases maintained by the Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). A database was prepared for capturing the characteristics of each of the 324 events in terms of when, where, how, and why the event occurred, what steps were taken to correct the deficiency that caused the event, and what defensive measures could have been employed to prevent recurrence of these events. The database also captures the plant system type, its safety classification, and whether or not the event involved a common cause failure. This work has revealed the following results and insights: - 82 of the 324 “digital” events did not actually involve a digital failure. Of these 82 non-digital events, 34 might have been prevented by making full use of digital system fault tolerance features. - 242 of the 324 events did involve failures in digital systems. The leading contributors to the 242 digital failures were hardware failure modes. Software change appears as a corrective action twice as often as it appears as an event root cause. This suggests that software features are being added to avoid recurrence of hardware failures, and that adequately designed software is a strong defensive measure against hardware failure modes, preventing them from propagating into system failures and ultimately plant events. 54 of the 242 digital failures involved a Common Cause Failure (CCF). - 13 of the 54 CCF events affected safety (1E) systems, and only 2 of those were due to Inadequate Software Design. This finding suggests that software related CCFs on 1E systems are no more prevalent than other CCF mechanisms for which adherence to various regulations and standards is considered to provide adequate protection against CCF. This research provides an extensive data set that is being used to investigate many different questions related to failure modes, causes, corrective actions, and other event attributes that can be compared and contrasted to reveal useful insights. Specific considerations in this study included comparison of 1E vs. non-1E systems, active vs. potential CCFs, and possible defensive measures to prevent these events. This paper documents the dominant attributes of the evaluated events and the associated insights that can be used to improve methods for protecting against digital I&C related CCFs, applying a test of reasonable assurance.

The Integrated Aging Management System Applied to Nuclear Power Plants in Korea

2013 ◽  
Author(s):  
Se-Youl Won ◽  
Kyeong-Soo Lee ◽  
Jae-Gon Lee
Keyword(s):  
Management System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Evaluation Report ◽  
Time Operation ◽  
Standard Guideline ◽  
Periodic Safety ◽  
Long Time

According to Post Fukushima action items in Korea, KHNP has established the integrated aging management system to reinforce aging management of system, structures, and components (SSCs) effectively for seven operating units, which are in service for more than twenty years, and for Kori Unit 1 and Wolsung Unit 1, which are subject to continued operation (CO) based on NUREG-1801 GALL report. KHNP’s integrated aging management programs (AMPs) focus on the establishment of aging management system for long-lived operation of nuclear power plants in Korea. The integrated aging management system consists of the integrated AMP standard guideline, operating guideline for each plant, individual AMPs of each plant, and AMP Data Base (DB) system including implementation results, basic DB information related to facilities operating in NPPs, and operating information such as operating experience and evaluation report. The integrated aging management system is importantly utilized for Periodic Safety Review (PSR) and the renewal of CO. Therefore, it will be largely contributed to keep NPPs the level of safety for long time operation through the effective aging management.

scholarly journals Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

2019 ◽  
Vol 5 (1) ◽  
pp. 67-74 ◽  
Author(s):  
Pavel L. Kirillov ◽  
Galina P. Bogoslovskaya
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Power ◽  
Operating Experience ◽  
Supercritical Pressure ◽  
Power Engineering ◽  
Fourth Generation ◽  
Supercritical Water Cooled Reactor

Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.

scholarly journals Outcomes of an international initiative for harmonization of low power and shutdown probabilistic safety assessment

2010 ◽  
Vol 25 (3) ◽  
pp. 222-228
Author(s):  
Giustino Manna ◽  
Irina Kuzmina ◽  
Jaroslav Holy
Keyword(s):  
Human Factors ◽  
Low Power ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Research Centre ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Many probabilistic safety assessment studies completed to the date have demonstrated that the risk dealing with low power and shutdown operation of nuclear power plants is often comparable with the risk of at-power operation, and the main contributors to the low power and shutdown risk often deal with human factors. Since the beginning of the nuclear power generation, human performance has been a very important factor in all phases of the plant lifecycle: design, commissioning, operation, maintenance, surveillance, modification, decommissioning and dismantling. The importance of this aspect has been confirmed by recent operating experience. This paper provides the insights and conclusions of a workshop organized in 2007 by the IAEA and the Joint Research Centre of the European Commission, on Harmonization of low power and shutdown probabilistic safety assessment for WWER nuclear power plants. The major objective of the workshop was to provide a comparison of the approaches and the results of human reliability analyses and gain insights in the enhanced handling of human factors.

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