Commercial-Grade Item Dedication for Nuclear Safety-Related Application

2010 ◽  
Author(s):  
Yao-Tung Hsu ◽  
Ming-Huei Chen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Personal Responsibility ◽  
Control Device ◽  
Project Planning ◽  
Nuclear Safety ◽  
Lessons Learned ◽  
Commercial Grade ◽  
Equipment Qualification

This paper describes the general practice and lessons learned from Commercial-Grade Item dedication for nuclear safety-related system applications in Taiwan. The dedication process qualified the commercial off-the-shelf components to be applied as basic components. In past fifteen years, Institute of Nuclear Energy Research (INER) has actively performed the dedication service to help local nuclear power plants solve their procurement problems of nuclear grade items, due to reduced availability of qualified suppliers and/or obsolete issues of qualified components. The Scope of dedication includes material, electrical and mechanical components located in mild and harsh environment. Thousands of components such as piping, fitting, breaker, relay, motor, and control device etc., have already been dedicated to and successfully used in local nuclear power plants. The Commercial-Grade Item dedication process is based on EPRI documentations. Besides, the technical evaluation and equipment qualification are included during the dedication process. The requirements for equipment qualification are described in IEEE standards. Although the codes and standards for dedication in Taiwan refer to those in USA, the challenges may happen due to different regulators, utility, manufacture’s quality culture, and personal responsibility. The key to successful dedication will be dependent on the well-defined component requirements and good project planning. This paper introduces the self-reliant experiences in dedication and economic benefit to local nuclear power plants.

Benefits of Digital Equipment Generic Qualification Activities

2002 ◽  
Author(s):  
James E. Thomas ◽  
Samuel C. Steiman
Keyword(s):  
Nuclear Power ◽  
Electromagnetic Compatibility ◽  
Nuclear Regulatory Commission ◽  
Project Planning ◽  
Nuclear Safety ◽  
Lessons Learned ◽  
Digital Platforms ◽  
Digital Devices ◽  
Commercial Grade ◽  
Digital Equipment

As a result of nuclear power plant instrumentation and control obsolescence issues, there have been numerous activities during recent years relating to the qualification of digital equipment. Some of these activities have been “generic” in nature in that the qualification was not limited to plant specific applications, but was intended to cover a broad base of potential applications of the digital equipment. These generic qualifications have been funded by equipment manufacturers and by utility groups and organizations. The generic activities sponsored by the Electric Power Research Institute (EPRI) have been pilot projects for an overall generic qualification approach. The primary benefit resulting from the generic qualification work to date is that a number of digital platforms and digital devices are now available for use in various nuclear safety-related applications. Many of the tests and evaluations necessary to support plant specific applications have been completed. The amount of data and documentation that each utility must develop on a case by case basis has been significantly reduced. There are also a number of additional benefits resulting from these industry efforts. The challenges and difficulties in qualifying digital equipment for safety-related applications are now more clearly understood. EPRI has published a lessons learned document (EPRI Report 1001452, Generic Qualification of Commercial Grade Digital Devices: Lessons Learned from Initial Pilots, which covers several different qualification areas, including device selection, project planning, vendor surveys and design reviews, and electromagnetic compatibility (EMC) qualification. Application of the experience and lessons learned from the EPRI pilot activities should help reduce the effort and cost required for future qualification work. Most generic qualification activities for commercial equipment have been conducted using the approach of EPRI TR-106439, Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications, and the supplemental guidance of EPRI TR-107339, Evaluating Commercial Digital Equipment for High Integrity Applications. The U. S. Nuclear Regulatory Commission (NRC) has reviewed some of the qualification activities and reports and issued a number of Safety Evaluation Reports (SERs). Although some additional work may be required to support specific applications, the generic qualification activities and NRC reviews have provided significant clarification and further stabilized the regulatory environment for use of digital equipment in safety-related applications.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Design and Analysis of a Reliable Communication System in Nuclear Safety Instrument & Control System

2021 ◽  
Author(s):  
Le Li ◽  
Zhihui Zhang ◽  
Chao Gao ◽  
Fei Zhou ◽  
Guangqiang Ma
Keyword(s):  
Communication Networks ◽  
Communication System ◽  
Communication Systems ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Safety Communication ◽  
Digital Control Systems

Abstract With the development of digital instrument and control technology for nuclear power plants in recent decades, communication networks have become an important part of safety digital control systems, which takes charge in data exchange between the various sub-systems, and extremely impact on the reliability and safety of the entire I&C system. Traditional communication systems where some special features, such as reliability, safety, real-time, certainty, and independence are not strictly required are various illustrated. However, how to implement a communication system in a safety I&C system is rarely stated in current research. In this research, a reliable safety communication system applied in nuclear power plants is designed and analyzed. The five key characteristics of nuclear safety communication networks are explained, followed by explanation of how to achieve these characteristics. The analysis and verification of the designed system are also stated in this paper, which contributes to proving that the designed nuclear safety communication system could applied in the nuclear power plants.

scholarly journals Reactor performance, system reliability, instrumentation and control

2020 ◽  
Vol 6 ◽  
pp. 43
Author(s):  
Andreas Schumm ◽  
Madalina Rabung ◽  
Gregory Marque ◽  
Jary Hamalainen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Software Systems ◽  
Reactor Performance ◽  
Horizon 2020 ◽  
Term Operation ◽  
In The Beginning ◽  
Electrical Cables

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

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.

Studies on Safety Assessment Method for Operating License Extension of Nuclear Power Plants in China

2017 ◽  
Author(s):  
Zhilin Chen ◽  
Ping Huang ◽  
Chunhui Wang ◽  
Zhiyuan Chi ◽  
Fangjie Shi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Management Systems ◽  
Effect Evaluation ◽  
Environment Effect ◽  
Assessment Environment

It’s the trend to extend the operating license time, called Operating License Extension (OLE) in China, of nuclear power plants (NPPs) in the future. It needs to be adequately demonstrated by licensees and approved by the regulator to gain an extended license time, such as 20 years. The demonstration methods for OLE are different among countries due to the different management systems for NPPs. Safety assessment, environment effect evaluation and update of the final safety analysis report (FSAR) will be the main aspects during OLE demonstration of NPPs in China according to the technical policy issued by National Nuclear Safety Administration (NNSA). Technical methods for scoping and screening, aging management review and time-limited aging analyses, which are the main contents of safety assessment are established based on the technical policy drafted by NNSA and international experiences in order to assist the operators to implement the safety assessment for OLE of NPP.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

A Multi-Dimensional Approach to Design and Execution of Modifications, Refurbishments, and Beyond for Existing Nuclear Power Plants

2020 ◽  
Author(s):  
Omid Malekzadeh ◽  
Matthew Monid ◽  
Michael Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Three Dimensional ◽  
Spatial Models ◽  
Technical Information ◽  
Lessons Learned ◽  
Piping System ◽  
Full Potential ◽  
Cad Models

Abstract Three-Dimensional (3D) CAD models are utilized by many designers; however, they are rarely utilized to their full potential. The current mainstream method of design process and communication is through design documentation. They are limited in depth of information, compartmentalized by discipline, fragmented into various segments, communicated through numerous layers, and finally, printed onto an undersized paper by the stakeholders and end-users. Large nuclear projects, such as refurbishments and decommissioning, suffer from spatial, interface, and interreference challenges, unintentional cost and schedule overruns, and quality concerns that can be rooted to the misalignments between designed and in-situ or previously as-built conditions that tend to stem from inaccessibility and lack of adequate data resolution during the transfer of technical information. This paper will identify the technologies and the methodology used during several piping system modifications of existing nuclear power plants, and shares the lessons learned with respect to the benefits and shortcomings of the approach. Overall, it is beneficial to leverage available multi-dimensional technologies to enhance various engineering and execution phases. The utilization and superposition of various spatial models into 3D and 4D formats, enabled the modification projects to significantly reduce in-person plant walkdown efforts, provide highly accurate as-found data, and enable stakeholders of all disciplines and trades to review the as-found, as-designed, and simulated as-installed modification; including the steps in between without requiring significant plant visits. This approach will therefore reduce the field-initiated changes that tend to result in design/field variations; resulting in less reliance on Appendix T of ASME BPVC Section III, reduction in the design registration reconciliations efforts, and it aligns with the overarching goal of EPRI guideline NCIG-05. Beyond the benefits to design and execution, the multidimensional approach will provide highly accurate inputs to some of the nuclear safety’s Beyond Design Basis Assessments (BDBA) and allowed for the incorporation of actual design values as input and hence removing the unnecessary over-conservatisms within some of the inputs.

Qualification of Motor Actuated Valve Assemblies for Safety Related Nuclear Applications: Lessons Learned

2018 ◽  
Author(s):  
Ronald Farrell ◽  
L. Ike Ezekoye
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Decision Processes ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
Missed Opportunities ◽  
Ieee Standards ◽  
Electrical Components ◽  
Nuclear Applications

Safety related valves in nuclear power plants are required to be qualified in accordance with the ASME QME-1 standard. This standard describes the requirements and the processes for qualifying active mechanical equipment that are used in nuclear power plants. It does not cover the qualification of electrical components that are addressed using IEEE standards; however, QME-1 recognizes that both mechanical and electrical components must be qualified when they are interfaced as an assembly. Qualifying both mechanical and electrical valve assemblies can be challenging. Considerable amount of judgment is used when developing the plan to qualify any valve with an electric motor actuator. If the wrong steps are taken in planning the tests, the results from the tests may not be useful thus triggering the need to perform additional tests to comply with QME-1 requirements. This paper presents lessons learned in the process of qualifying valve assemblies to meet QME-1 requirements. The lessons include the decision processes associated with planning and executing valve testing, analysis of the valve assemblies for natural frequency determination, and missed opportunities to capture relevant test data during the tests. Finally, the paper will discuss challenges associated with justifying the tests and extending the results of the tests to cover untested valve assemblies.

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