Reconstitution of Reactor Pressure Boundary Components Design Stress Report for License Renewal

2010 ◽  
Author(s):  
Kirby Woods ◽  
Kenneth Thomas
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Cyclic Life ◽  
System Supplier ◽  
Asme Code ◽  
American Society ◽  
Renewal Period ◽  
License Renewal ◽  
Design Stress

The majority of United States Commercial Nuclear Power Plants (CNPP) within the next 10 years will reach the end of their license to operate and can be renewed per the “Atomic Energy Act” of 1954. This act allowed the commission to issue commercial electric power nuclear plants a license to operate (“licensed, but not to exceed 40 years, and maybe renewed upon the expiration of such period, (Chapter 10, Sec. 103(c))).” These CNPP licenses are also governed by the NSSS vendor specification requirements and by the American Society of Mechanical Engineers (ASME) design code standards. This connection is in the form of a stress report that defines the “cyclic life” adequacy criteria for this operational limit of 40 years. The license extension subsequently requires a reconstitution of the initial design stress report input parameters per ASME IWA-4120, 4223 & 4311 (e) for the renewal period extension. This requirement can entail an analysis of the operating conditions and cycles to demonstrate the material elasticity is maintained. The proposed approach for this reconstitution effort was a reanalysis in the form of a study of the Nuclear Steam System Supplier (NSSS) vendors’ original methodology to determine the NSSS vendor specification requirement for ASME code compliance and “cyclic life” adequacy. The information acquired from this evaluation has demonstrated the application to be a complex and simplistic approach. This effort to unravel the composite loading (thermal and pressure transients) condition in relation to specific plant and incident cycles provides both an understanding of component end-of-life limits and supports a comprehensive template for future fatigue life management programs. This paper summarizes this reconstituting effort that utilizes the original vessel stress analysis report to support the license renewal effort, provides a template for future fatigue management programs, demonstrates the conservatism of design, and offers an insight into the design philosophy revealing an elegant process that assures against failures.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

2014 ◽  
Author(s):  
Jason Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Diagnostic Testing ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Light Water ◽  
Water Reactor ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

scholarly journals ASME OM Code Implementation Lessons Learned for the AP1000® Plant

2017 ◽  
Author(s):  
Augustine A. Cardillo ◽  
Natalie M. Rodgers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Federal Regulations ◽  
Plant Design ◽  
Regulatory Requirements ◽  
Implementation Challenges ◽  
Service Testing ◽  
Asme Code ◽  
American Society

The Title 10 of the Code of Federal Regulations (10 CFR) Part 52 process and unique aspects of a passive plant design have presented new challenges for the development and implementation of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements. This paper will discuss lessons learned from the development and implementation of pre-service testing (PST) / in-service testing (IST) program plans for the AP1000® plant, for both international and domestic. Topics to be addressed include the following: • Level of detail in design certification • Treatment of unique passive plant features • Design certification commitments beyond Code requirements • Future regulatory requirements for high-risk non-safety component PST/IST • Implementation challenges for international plants Paper published with permission.

Combination Criterion for Multiple Laminar Flaws in Steel Components

2017 ◽  
Vol 741 ◽  
pp. 63-69
Author(s):  
Valéry Lacroix ◽  
Vratislav Mareš ◽  
Bohumír Strnadel ◽  
Kunio Hasegawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Rolling Direction ◽  
Separation Distance ◽  
Pressure Vessels ◽  
New Combination ◽  
Mechanical Interaction ◽  
Single Plane ◽  
Asme Code ◽  
American Society

A laminar flaw is a planar subsurface flaw parallel to the rolling direction of the plate, where the applied stress is typically parallel to the rolling direction. The laminar flaw oriented within 10 degree of a plane parallel to the component surface is defined as a laminar flaw, in accordance with the definition of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section XI. The ASME Code provides combination criterion for multiple laminar flaws. If there are two or more laminations, these laminations are projected to a single plane and, if the separation distance of the projected laminations is less than or equal to 25.4 mm, the laminations shall be combined into a single large laminar flaw in the assessment. The combination criterion was established on the basis of the non-destructive examination capabilities in the 1970’s. However, this methodology did not consider the offset distance of the laminations nor the mechanical interaction between the flaws. Therefore that combination methodology is not suited in case of a large number of laminar flaws. This may occur e.g. in case of hydrogen flaking in steel forging components. Actually, when multiple discrete laminar flaws are close to each other, interaction between the flaws has to be taken into account and these flaws shall be combined to a single laminar flaw for assessment. Stress intensity factor interactions for inclined laminar flaws were analyzed in the frame of hydrogen flaking issue in reactor pressure vessels of Doel 3 and Tihange 2 Belgian nuclear power plants. Based on the mechanical interaction between flaws, new combination criterion was developed and was presented in this paper.

Regulatory Requirements and Perspectives on the Use of Inservice Testing Software in the Nuclear Industry

2014 ◽  
Author(s):  
Jason B. Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Regulatory Requirements ◽  
Operational Readiness ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing (IST) and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses the use of the ASME OM Code in 10 CFR 50.55a(b)(3) . This paper focuses on applicable regulatory requirements and regulatory perspectives associated with the use of IST software in the nuclear industry. Paper published with permission.

scholarly journals Application of the Advanced Surface Modification Process to the ASME Code Case for Sections III and XI of Nuclear Power Plants

Metals ◽  
2020 ◽  
Vol 10 (2) ◽  
pp. 210
Author(s):  
Sungwoo Cho ◽  
Hyun-Uk Hong ◽  
Nicholas Mohr ◽  
Marc Albert ◽  
John Broussard ◽  
...  
Keyword(s):  
Surface Modification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cost Savings ◽  
Performance Criteria ◽  
Laser Peening ◽  
Modification Process ◽  
Asme Code ◽  
American Society

The advanced surface modification process is known as a promising solution to improve the performance of machine components and systems, especially for vehicles, nuclear power plants, biomedical device, etc. There have been several successful applications of water jet peening and underwater laser peening to nuclear components in Japan since 2001 which resulted in inspection and repair cost savings. The prerequisite condition for the application of the advanced surface modification process to nuclear power plants is the approval of the American Society of Mechanical Engineers (ASME) Code Case, so performance criteria and requirements (PCRs) in the ASME Code Case for repair and maintenance of nuclear power components are explained. A challenging project to apply advanced surface modification processes, such as ultrasonic nanocrystal surface modification and air laser peening to new nuclear power plants and new canisters, was created with the goal to develop a technical basis and the PCRs for ASME Section III (New Manufacturing). The results of this work will be an ASME Section III Code Case which is currently in progress. An initial draft of the new Code Case with the intermediate results of this work is introduced. Four kinds of advanced surface modification processes are explained and compared briefly.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

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, Analysis and System-Level Modelling of a Single Axis Capacitive Accelerometer

2016 ◽  
Author(s):  
Zakriya Mohammed ◽  
Owais Talaat Waheed ◽  
Ibrahim (Abe) M. Elfadel ◽  
Aveek Chatterjee ◽  
Mahmoud Rasras
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noise Analysis ◽  
Linear Displacement ◽  
Operating Conditions ◽  
System Level ◽  
Complete Analysis ◽  
Capacitive Accelerometer ◽  
Static Capacitance ◽  
Cross Axis

The paper demonstrates the design and complete analysis of 1-axis MEMS capacitive accelerometer. The design is optimized for high linearity, high sensitivity, and low cross-axis sensitivity. The noise analysis is done to assure satisfactory performance under operating conditions. This includes the mechanical noise of accelerometer, noise due to interface electronics and noise caused by radiation. The latter noise will arise when such accelerometer is deployed in radioactive (e.g., nuclear power plants) or space environments. The static capacitance is calculated to be 4.58 pF/side. A linear displacement sensitivity of 0.012μm/g (g = 9.8m/s2) is observed in the range of ±15g. The differential capacitive sensitivity of the device is 90fF/g. Furthermore, a low cross-axis sensitivity of 0.024fF/g is computed. The effect of radiation is mathematically modelled and possibility of using these devices in radioactive environment is explored. The simulated noise floor of the device with electronic circuit is 0.165mg/Hz1/2.

Sign in / Sign up

Export Citation Format

Share Document