Operational Impacts and Consequences of Piping Component Failure: A Review of Operating Experience Data As Recorded in CODAP

2018 ◽  
Author(s):  
Braedon Carr ◽  
Bengt Lydell ◽  
Jovica R. Riznic
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling Water ◽  
Operating Experience ◽  
Transport Systems ◽  
Volume Control ◽  
Operational Experience ◽  
Component Failure ◽  
Plant Safety ◽  
Operational Impacts

Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2]. In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.

scholarly journals Ion chromatographic analysis of aqueous iodine species by conductivity and radioactivity measurements

2005 ◽  
Vol 93 (9-10) ◽  
Author(s):  
Dorothea Schumann ◽  
R. Grasser ◽  
R. Dressler ◽  
H. Bruchertseifer
Keyword(s):  
Chromatographic Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Radioactive Iodine ◽  
Cooling Water ◽  
Molecular Iodine ◽  
New Device ◽  
Iodine Species ◽  
Radioactivity Measurements

SummaryA new device was developed for the identification of several iodine species in aqueous solution using ion chromatography. Iodide, iodate and molecular iodine can be determined. (The equipment allows both conductivity and radioactivity detections.) The method is applicable for the determination of radioactive iodine contaminations in the cooling water of nuclear power plants.

Energy in the 1980s - Operational experience in nuclear power stations

1974 ◽  
Vol 276 (1261) ◽  
pp. 571-586
Keyword(s):  
Nuclear Power ◽  
Nuclear Energy ◽  
Large Scale ◽  
Electrical Power ◽  
Operating Experience ◽  
Operational Experience ◽  
Original Design ◽  
Safety Requirements ◽  
Design Concepts ◽  
Power Stations

From the first self-sustaining nuclear reaction to the present day represents a span of three decades: within that time large-scale generation of electrical power from nuclear energy has become acknowledged as economic, safe and environmentally acceptable. Within the U .K . 10% of electricity consumed is of nuclear origin. Some of the C.E.G.B. reactors have been in service for over 10 years. The operating experience that has been gained shows how the original design concepts have been ultimately developed. Some of the difficulties encountered and the engineering solutions are presented. Operating experience feeds back to the design philosophy and safety requirements for future nuclear plant. In this way a foundation is provided for the further exploitation of what must become a major source of energy in the next decade.

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.

Explore the Self-Governance on Nuclear Safety Emergency - A Case Study of the 3rd Nuclear Power Plant

2012 ◽  
Vol 260-261 ◽  
pp. 103-106
Author(s):  
Yi Chun Lin ◽  
Yung Nane Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Alternative Energy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Security Issue ◽  
Safety Issues ◽  
Plant Safety

The ripples of the tsunami crisis in Japan triggered introspections of nuclear plant safety issues in the worldwide. Many countries have claimed the suspension of nuclear power plants. However, some countries such as Taiwan, under nearly 99% energy is exported, the disasters force government and citizen to face the importance of nuclear safety, especially the neighborhoods nearby the nuclear power plants. We have to face the nuclear safety since there is no other alternative energy presently. The 3rd nuclear power plant located in the south of Taiwan, which has the same geographic features with Fukushima, Japan. Presently, there is no precedent in Taiwan of precaution and rescue team and civil supervised mechanic on nuclear security issue. This paper will review according to transparent information, public participation and cross-organization cooperation to propose the execution and work division principles, including information monitor, educational propagation, hide and evacuation, emergence aid and care, rear and refuge service. The ultimate target is to establish self-governance inside nearby neighborhood to confront nuclear disaster at the critical moment.

Operational Experience of TN 24 Dual Purpose Spent Fuel Storage Casks at the DOEL Nuclear Power Plants in Belgium

2001 ◽  
Author(s):  
C. Baroux ◽  
M. Detrilleaux ◽  
G. Demazy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
High Capacity ◽  
Radiation Shielding ◽  
Ring Closure ◽  
Operational Experience ◽  
Spent Fuel ◽  
Dual Purpose ◽  
New Members

Abstract Spent nuclear fuel has been stored at the DOEL power station in Belgium in dual-purpose metal casks since 1995. The casks were procured from TRANSNUCLEAIRE by SYNATOM to meet the operational demands for on-site dry storage solutions for fuel arising from the four PWR reactors at DOEL. The TN 24 type of cask was chosen and a range of different cask types were developed. The initial requirement was for dual purpose cask to contain fuel from the DOEL units 3 and 4, these having similar fuel types but different lengths, and thus two new members of the TN 24 family were developed; the TN 24 D and TN 24 XL with capacities of 28 and 24 SFA’s. These casks were licensed as B(U) fissile packagings with approval certificates granted by the French and validated by the Belgium competent authorities for the transport configurations. Both cask designs were also analyzed by TRANSNUCLEAIRE in their storage configurations to ensure that the criteria for safe interim storage could be met. Since 1995, a total of 18 TN 24 D and TN 24 XL casks have been loaded with spent fuel assemblies with an average burn-up of 40,000 MWd/tU. SYNATOM subsequently decided to purchase further casks for DOEL 3 and 4 fuels with higher enrichments, higher burn-ups and shorter cooling times. TRANSNUCLEAIRE developed the TN 24 DH and TN 24 XLH casks within the similar envelope size and weight limits. The increase in performance was achieved by an in-depth optimization of each design in terms of radiation shielding, heat transfer and criticality safety. This paper shows how this optimization process was undertaken for the TN 24 DH and TN 24 XLH casks, 16 of which have been ordered by SYNATOM. DOEL 1 and 2 units use much shorter PWR fuel and it was decided to ship the fuel to unit 3 with an internal transfer cask because the handling limitations in the DOEL 1 and 2 pool prohibited the loading of a high capacity dual purpose transport/storage cask. The TN 24 SH cask was subsequently designed for DOEL 1 and 2 PWR fuel with a capacity of 37 assemblies and nine of there casks have been ordered by SYNATOM. The casks are fitted with monitoring devices to detect any change in the performance of the double metal O ring closure system and none of the casks has shown any deterioration in leaktightness. This paper examines the operation experience of loading and storing more than 30 TN 24 dual purpose casks and compares the performance with design expectations.

Study on a Methodology of Human Factor Engineering Operating Experience Review for Nuclear Power Plant

2010 ◽  
Author(s):  
Danying Gu ◽  
Shuhui Zhang ◽  
Zhonghe Ning
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Factor ◽  
Operating Experience ◽  
Sources Of Information ◽  
Nuclear Facilities ◽  
Human Factor Engineering ◽  
Design Procedures ◽  
Function Allocation ◽  
Review Criterion

The reviewing of operating experience at nuclear power plants (NPP) is not only critically important to safe and reliable operations, but also useful to guide the design of new plants which are similar to the current one under review. How to identify and analyze the safety-related operating experience and then implement a more extensive review is a vital and challengeable issue. In this paper, a methodology of human factor engineering (HFE) operating experience review (OER) is proposed for NPP. The need for the application of HFE in the life cycle activities of NPP and other nuclear facilities has been demonstrated by plant operating histories and regulatory and industry reviews. As a very important element of HFE, the OER is performed from the beginning of the design process. The main purpose of performing an OER is to verify that the applicant has identified and analyzed HFE-related safety problems and issues in previous designs that are similar to the current one. In this way, negative features associated with predecessor designs may be avoided in the current NPP design while retaining positive features. The research of OER concentrates on the aspect of review criterion, scope and implementation procedure of the HFE-related operating experience. As the NRC requirement, the scope of operating experience can be divided into six types in accordance with sources of information. The implementation procedures of USA and China are introduced, respectively. The resolution of HFE OER issues involve function allocation, changes in automation, HSI equipment design, procedures, training, and so forth. The OER conclusions can contribute to other HFE activities and improve the safety, reliability and usability of the HSI design in NPP.

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.

Operational Experience Data Analysis on Events Related to Maintenance at Nuclear Power Plants

2016 ◽  
Vol 184 (4) ◽  
pp. 575-583 ◽  
Author(s):  
Antonio Ballesteros ◽  
Radian Sanda ◽  
Michael Maqua ◽  
Jean-Luc Stephan
Keyword(s):  
Data Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operational Experience

Study on PRA Procedure Considering Combination of Multiple Events Using DQFM Methodology

2018 ◽  
Author(s):  
Hirohisa Yamakawa ◽  
Hitoshi Muta
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Event ◽  
Seismic Events ◽  
Power Station ◽  
Component Failure ◽  
Multiple Events ◽  
Fukushima Daiichi ◽  
Tsunami Event ◽  
Direct Quantification

Fukushima Daiichi Nuclear Power Station accident occurred by the Great East Japan Earthquake on March 11, 2011. After that, continuous enhancement of nuclear safety is being required in Japan. The accident of Fukushima was caused by the seismic induced tsunami event, namely, multiple events. The other examples of multiple events due to the seismic event are such as internal fire and internal flooding in the nuclear power plants. In addition, structures, such as a building, and piping might be damaged by the seismic event, which could impact component failure dependently. In order to consider these kinds of events, the development of PRA procedures for multiple events caused by the seismic events will be highly demanded. So, we developed a basic PRA methodology for seismic induced tsunami events using “Direct Quantification of Fault Tree using Monte Carlo simulation (DQFM) methodology”. And we verified its applicability through the evaluation.

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