Evolution of Section XI of the ASME Boiler and Pressure Vessel Code

2000 ◽  
Vol 122 (3) ◽  
pp. 234-241 ◽  
Author(s):  
Owen F. Hedden
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Operating Conditions ◽  
Nondestructive Examination ◽  
The Us ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Year 2000

This article will describe the development of Section XI from a pamphlet-sized document to the lengthy and complex set of requirements, interpretations, and Code Cases that it has become by the year 2000. Section XI began as a set of rules for inservice inspection of the primary pressure boundary system of nuclear power plants. It has evolved to include other aspects of maintaining the structural integrity of safety class pressure boundaries. These include procedures for component repair/replacement activities, analysis of revised and new plant operating conditions, and specialized provisions for nondestructive examination of components and piping. It has also increased in scope to cover other Section III construction: Class 2, Class 3 and containment structures. First, to provide a context for the discussions to follow, the differences in administration and enforcement between Section XI and the other Code Sections will be explained, including its dependence on the US Nuclear Regulatory Commission. The importance of interpretations and Code Cases then will be discussed. The development of general requirements and requirements for each class of structure will be traced. The movement of Section XI toward a new philosophy, risk-informed inspection, will also be discussed. Finally, an annotated bibliography of papers describing the philosophy and technical basis behind Section XI will be provided. [S0094-9930(00)01703-0]

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Cycle Fatigue ◽  
Structural Integrity ◽  
Fatigue Cracking ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Boiling Water ◽  
Cycle Fatigue ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

scholarly journals Statistical Analysis of Loss of Offsite Power Events

2016 ◽  
Vol 2016 ◽  
pp. 1-9 ◽  
Author(s):  
Andrija Volkanovski ◽  
Antonio Ballesteros Avila ◽  
Miguel Peinador Veira
Keyword(s):  
Statistical Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Reporting System ◽  
Nuclear Regulatory Commission ◽  
The Us ◽  
Mode Of Operation ◽  
Order Of Magnitude ◽  
The One ◽  
Regulatory Commission

This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC).

Practical Considerations of Site Clearance

2003 ◽  
Author(s):  
Thomas S. LaGuardia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Site Specific ◽  
Exposed Individual ◽  
The Us ◽  
Pathways Analysis ◽  
Residual Radioactivity ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) established criteria for acceptable residual radioactivity related to decommissioning nuclear power plants in the US [1]. A level of 25 mRem per year to the maximum exposed individual by site-specific pathways analysis, with ALARA is acceptable to the NRC. Systems and structures containing very low levels of radioactivity that meet this criteria are deemed acceptable to abandon in place as part of the decommissioning process and termination of the license. Upon license termination by the NRC, the owner may then demolish and remove remaining structures. In practice, site-specific criteria imposed by local state mandates, company management decisions, real estate value, and long-term liability potential have driven nuclear plant licensees to adopt an alternative disposition for these materials. Although the reasons are different at each site, the NRC’s criteria of 25 mRem per year are not the controlling factor. This paper will describe the regulatory process for termination of the license, and the other factors that drive the decision to remove radioactive and non-radioactive material for decommissioning. Several case histories are presented to illustrate that the NRC’s criteria for license termination are not the only consideration.

Codes, Standards and Regulatory Topics for Nuclear Facilities

2008 ◽  
Author(s):  
Taunia Wilde ◽  
Shannan Baker ◽  
Gary M. Sandquist
Keyword(s):  
Quality Control ◽  
Quality Assurance ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Quality Assurance Program ◽  
Nuclear Facilities ◽  
The Us ◽  
Regulatory Commission

The design, construction, operation, maintenance, and decommissioning and decontamination of nuclear infrastructure particularly nuclear power plants licensed in the US by the US Nuclear Regulatory Commission (NRC) or operated by the US Department of Energy (DOE) or the US Department of Defense (DOD) must be executed under a rigorous and documented quality assurance program that provides adequate quality control and oversight. Those codes, standards, and orders regulate, document and prescribe the essentials for quality assurance (QA) and quality control (QC) that frequently impact nuclear facilities operated in the US are reviewed and compared.

A Study on Maintenance Rule Application in VVER NPP

2017 ◽  
Author(s):  
Ma Chao ◽  
Deng Wei ◽  
An Jin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Power Production ◽  
Nuclear Regulatory Commission ◽  
Application Process ◽  
The Us ◽  
Rule Application ◽  
Maintenance Activities ◽  
Regulatory Commission ◽  
Wide Usage

Maintenance effectiveness is important for the safety and power production of Nuclear Power Plants (NPP). U.S. Nuclear Regulatory Commission (NRC) Maintenance Rule (10CFR50.65, MR) became effective in 1996, and it is mandatory for all the US plants to use Maintenance Rule in their daily maintenance activities. With the development and wide usage of Probabilistic Risk Assessment (PRA) technique in China, China regulator and utilities are trying to adopt MR in maintenance activities. Brief study on application of MR in some VVER-typed China nuclear plant is carried out and some main results are shown. All the application process and results will be useful for later official application of MR in China.

scholarly journals Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

2010 ◽  
Author(s):  
Joseph Braverman ◽  
Richard Morante ◽  
Charles Hofmayer ◽  
Robert Roche-Rivera ◽  
Jose Pires
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Gas Generation ◽  
Design Basis ◽  
And Performance ◽  
Regulatory Commission ◽  
Technical Basis

Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 [1] and US NRC Standard Review Plan, Section 3.8) [2]; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 [3] and 10 CFR 50 [1]); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 [3] as well as SECY 90–016 [4], SECY 93–087 [5], and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

Technical Basis for Acceptance of Flaws in Moderate Energy Class 2 and 3 Vessels and Tanks

2004 ◽  
Author(s):  
Pat L. Strauch ◽  
Warren H. Bamford ◽  
Sushil K. Daftuar
Keyword(s):  
Structural Integrity ◽  
Growth Curves ◽  
Nuclear Regulatory Commission ◽  
Operating Conditions ◽  
Energy Class ◽  
Asme Code ◽  
Moderate Energy ◽  
Leakage Monitoring ◽  
Regulatory Commission ◽  
Technical Basis

New procedures and acceptance criteria for the evaluation of degradation, including through-wall flaws, in moderate energy Class 2 and 3 vessels and tanks have been prepared for implementation within Section XI of the ASME Code. The provisions are contained in a proposed Code Case and are focused on the structural integrity margin of the vessel or tank against gross failure. The assessment of the degraded condition is based on the flaw evaluation procedures already established in ASME Section XI. Additional provisions for periodic inspection and leakage monitoring are included to assure that analysis assumptions are conservative for the operating conditions. The precedent for permitting operation with degraded components was established in United States Nuclear Regulatory Commission (NRC) Generic Letter 90-05 and Code Case N-513-1 for piping, as well as several NRC-accepted plant-specific relief requests associated with leaking tanks. The technical basis for the procedures is presented, and the objectives and scope of its application are explained. The basis for the analytical procedures follows from evaluation rules contained in ASME Section XI, Appendix A. Other issues regarding consequences of leakage, growth of degradation, and augmented inspections and surveillance are also addressed, as well as reference crack growth curves for stress corrosion cracking for conditions appropriate for application of these procedures.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

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.

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