05/01017 Lessons learned with vibration monitoring systems in German nuclear power plants

2005 ◽  
Vol 46 (3) ◽  
pp. 156
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Vibration Monitoring ◽  
Monitoring Systems

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.

A 12 MM/S RMS Screening Vibration Velocity for Pipes Using ANSI-OM3 Standard and Regulatory Design Rules

2005 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Yannick Pons ◽  
Pierre Moussou ◽  
Michae¨l Gaudin
Keyword(s):  
Correction Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Vibration Monitoring ◽  
Vibration Velocity ◽  
Concentrated Mass ◽  
Design Rules ◽  
Velocity Relationship ◽  
Regulatory Design

ASME ANSI-OM3 standard is dedicated to the assessment of piping vibrations for nuclear power plants. It provides an allowable zero-to-peak velocity, which is derived from a stress/velocity relationship, where corrections factors (C1, C2K2, C3, C4 and C5) and an allowable stress σal are introduced. In the ANSI-OM3 standard, the C4 correction factor depends on the pipe layout and on its boundary conditions, and is calculated for a few cases. In a former work, it was proposed to extend this factor to a larger number of pipe setups. Besides, the correction factor C1, which stands for the effect of concentrated mass, is established on a given set-up: a clamped-clamped straight pipe span on its first vibrating mode. C1 is then supposed to be conservative on any piping layout. Finally, allowable velocities derived from the ANSI-OM3 stress/velocity relationship may be very conservative. One way to reduce this conservatism is to introduce regulatory design rules. For a larger set of pipe geometries, a new set of C1 and C4 correction factors are computed using weight and pressure designs. Using these numerical results, allowable velocities can be calculated. Then, we propose here to check if a screening vibration velocity of 12 mm/s rms is fulfilled. For the 181 geometries on 3708, which do not meet the criterion, a seismic design checking is applied. Finally, by this way, 99.7% of the tested geometries, which are supposed to be acceptable with respect to static and seismic designs, display allowable velocities above 12 mm/s rms and the minimum allowable vibration velocity is 11.2 mm/s. This screening vibration velocity of 12 mm/s commonly used for vibration monitoring of piping systems in EDF nuclear power plants is then supported.

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.

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.

Overview of Seismic Issues in Design Certification and Combined License Application Reviews

2009 ◽  
Author(s):  
Jim Xu ◽  
Sujit Samaddar
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Site Specific ◽  
License Application ◽  
Standard Design

The U.S. Nuclear Regulatory Commission (NRC) established a new process for licensing nuclear power plants under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which provides requirements for early site permits (ESPs), standard design certifications (DCs), and combined license (COL) applications. In this process, an application for a COL may incorporate by reference a DC, an ESP, both, or neither. This approach allows for early resolution of safety and environmental issues. The COL review will not reconsider the safety issues resolved by the DC and ESP processes. However, a COL application that incorporates a DC by reference needs to demonstrate that pertinent site-specific parameters are confined within the safety envelopes established by the DC. This paper provides an overview of site parameters related to seismic designs and associated seismic issues encountered in DC and COL application reviews using the 10 CFR Part 52 process. Since DCs treat the seismic design and analysis of nuclear power plant (NPP) structures, systems, and components (SSC) as bounding to future potential sites, the design ground motions and associated site parameters are often conservatively specified, representing envelopes of site-specific seismic hazards and parameters. For a COL applicant to incorporate a DC by reference, it needs to demonstrate that the site-specific hazard in terms of ground motion response spectra (GMRS) is enveloped by the certified design response spectra of the DC. It also needs to demonstrate that the site-specific seismic parameters, such as foundation-bearing capacities, soil profiles, and the like, are confined within the site parameter envelopes established by the DC. For the noncertified portion of the plant SSCs, the COL applicant should perform the seismic design and analysis with respect to the site-specific GMRS and associated site parameters. This paper discusses the seismic issues encountered in the safety reviews of DC and COL applications. Practical issues dealing with comparing site-specific features to the standard designs and lessons learned are also discussed.

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