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.

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]

Impact of Undetected Fabrication Flaws on LBB Risk

2017 ◽  
Author(s):  
Robert Kurth ◽  
Cédric Sallaberry ◽  
Elizabeth Kurth ◽  
Frederick Brust
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Leak Detection ◽  
Nuclear Regulatory Commission ◽  
Acceptance Testing ◽  
Analysis Tool ◽  
Risk Of Rupture ◽  
Asme Code ◽  
Regulatory Commission ◽  
The Impact

On-going assessments of the impact of active degradation mechanisms in US nuclear power plants previously approved for leak before break (LBB) relief are being performed with, among other methods, the extremely low probability of rupture (xLPR) code being developed under a memorandum of understanding between the US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) [1]. This code finished with internal acceptance testing in July of 2016 and is undergoing sensitivity and understanding analyses and studies in preparation for its general release. One of the key components of the analysis tool is the integration of inspection methods for damage and the impact of leak detection on the risk of a pipe rupture before the pipe is detected to be leaking. While it is not impossible to detect a crack or defect that is less than 10% of the pipe wall thickness current ASME code does not give credit for inspections identifying a crack of this size. This study investigates the impact of combining the probabilistic analysis results from xLPR with a pre-existing flaw to determine the change, if any, to the risk. Fabrication flaws were found to have a statistically significant impact on the risk of rupture before leak detection.

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.

Technical Basis for Proposed Fourth Revision to ASME Code Case N-513

2014 ◽  
Author(s):  
Robert O. McGill ◽  
Guy DeBoo ◽  
Russell C. Cipolla ◽  
Eric J. Houston
Keyword(s):  
Nuclear Regulatory Commission ◽  
Short Term ◽  
Safety Issues ◽  
Evaluation Procedures ◽  
The Us ◽  
Bent Pipe ◽  
Asme Code ◽  
Moderate Energy ◽  
Regulatory Commission ◽  
Technical Basis

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been three revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the three revisions have been previously published. There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fourth revision. These changes include addressing the current condition on the Code Case acceptance by the US Nuclear Regulatory Commission (NRC), clarification of the Code Case applicability limits and expansion of Code Case scope to additional piping components. New flaw evaluation procedures are given for through-wall flaws in elbows, bent pipe, reducers, expanders and branch tees. These procedures evaluate flaws in the piping components as if in straight pipe by adjusting hoop and axial stresses to account for the geometry differences. These changes and their technical bases are described in this paper.

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.

Quantifying Detection Capabilities of Detrimental Conditions in High-Density Polyethylene Butt-Fusion Joints

2010 ◽  
Author(s):  
Caleb J. Frederick
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Fusion ◽  
Nuclear Regulatory Commission ◽  
High Density ◽  
Full Spectrum ◽  
Asme Code ◽  
Regulatory Commission ◽  
Joint Integrity

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) piping in non-safety-related and safety-related systems. HDPE has been chosen in limited quantity to replace carbon steel piping as it does not support rust, rot, or biological growth. However, due to its relatively short nuclear service history, developing a way to accurately evaluate joint integrity in HDPE is crucial to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate using ultrasonic Phased Array to quantify detection of flaws and detrimental conditions in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Currently the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, there is the appearance of a good, quality joint. However, the fabricated joint does not have the required strength as the co-crystallization along the pipe faces has not occurred. Therefore, performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate assurance of joint integrity. As a potential solution, volumetric examination is being considered by the USNRC to safeguard against this and other types of detrimental conditions. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive correlation with ultrasonic data.

Status of Code Cases Related to Application of Risk-Informed Inservice Inspection in ASME Code Section XI

1998 ◽  
Vol 120 (4) ◽  
pp. 438-440
Author(s):  
O. F. Hedden
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Extensive Discussion ◽  
Managerial Support ◽  
Piping Systems ◽  
Asme Code ◽  
Sources Of Support ◽  
Favorable Reaction ◽  
Regulatory Commission

ASME Code Section XI Cases N-577 and N-578, for application of risk-informed technology to examination of piping systems in nuclear power plants, are proceeding, with review and acceptance by ASME Board on Nuclear Codes and Standards and by the U.S. Nuclear Regulatory Commission remaining before implementation. Sources of support for a favorable reaction by NRC will be reviewed, starting with developmental research sponsored by NRC in the late 1980s. Extensive discussion in the engineering community as exemplified by forums presented by ASME PVP in 1994 and 1995 will be cited. Recent academic and NRC managerial support for risk-informed performance-based regulation will also be cited. The expressed need for risk neutrality will then be addressed.

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.

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