Field Experience With Code Case N-659: Ultrasonic Examination in Lieu of Radiography

2007 ◽  
Author(s):  
Douglas O. Henry
Keyword(s):  
Nuclear Power ◽  
Cost Benefit ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Material Transport ◽  
Ultrasonic Examination ◽  
Key Factor ◽  
Asme Code ◽  
Regulatory Commission ◽  
Sample Set

Code Case N-659 Revision 0 was approved in 2002 to allow ultrasonic examination (UT) an alternative to radiography (RT) for nuclear power plant components and transport containers under Section III of the ASME Code. The Nuclear Regulatory Commission has not approved N-659 and its subsequent revisions (currently N-659-2) for general use, but they have been used on a case-by-case basis mainly where logistic problems or component configuration have prevented the use of radiography. Like the parallel Code Case 2235 for non-nuclear applications under Section I and Section VIII, Code Case N-659 requires automated, computerized ultrasonic systems and capability demonstration on a flawed sample as a prerequisite for using UT in lieu of RT. Automated ultrasonic examination can be significantly more expensive than radiography, so a cost-benefit evaluation is a key factor in the decision to use the Code Case. In addition, the flaw sample set has recently become an issue and a topic of negotiation with the NRC for application of the Case. A flaw sample set for a recent radioactive material transport cask fabrication project was successfully negotiated with the NRC. The Code Case N-659 approach has been used effectively to overcome barriers to Code required radiography. Examples are examination of welds in an assembled heat exchanger and in a radioactive material transport cask assembly where internal shielding prevented radiography of the weld. Future development of Code Case N-659 will address sample set considerations and application-specific Code Cases, such as for storage and transport containers, will be developed where NRC concerns have been fully addressed and regulatory approval can be obtained on a generic basis.

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.

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.

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.

Potential Effects of Historic Rail Accidents on the Integrity of Spent Nuclear Fuel Transportation Packages

2009 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hazardous Materials ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Radioactive Material ◽  
Material Transport ◽  
Long Duration ◽  
The Us ◽  
Regulatory Commission

The US Nuclear Regulatory Commission (NRC) completed an analysis of historical rail accidents (from 1975 to 2005) involving hazardous materials and long duration fires in the United States. The analysis was initiated to determine what types of accidents had occurred and what impact those types of accidents could have on the rail transport of spent nuclear fuel. The NRC found that almost 21 billion miles of freight rail shipments over a 30 year period had resulted in a small number of accidents involving the release of hazardous materials, eight of which involved long duration fires. All eight of the accidents analyzed resulted in fires that were less severe than the “fully engulfing fire” described as a hypothetical accident condition in the NRC regulations for radioactive material transport found in Title 10 of the Code of Federal Regulations, Part 71, Section 73. None of the eight accidents involved a release of radioactive material. This paper describes the eight accidents in detail and examines the potential effects on spent nuclear fuel transportation packages exposed to the fires that resulted from these accidents.

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.

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.

Proposed License Structure of Small Modular Reactor in China

2018 ◽  
Author(s):  
Zheng Hua ◽  
Wei Shuhong
Keyword(s):  
Nuclear Power ◽  
Modular Design ◽  
Nuclear Regulatory Commission ◽  
Light Water Reactor ◽  
Research Progress ◽  
Structure Characteristics ◽  
Common Structure ◽  
Small Modular Reactor ◽  
Regulatory Commission ◽  
Design Construction

Small Modular Reactor (SMR) is getting more and more attention due to its safety and multi-purpose application. License structure is an important issue for SMR licensing. Modular design, construction and operation, shared or common structure, system and components (SSC) challenge existing large light water reactor license structure. Existing nuclear power plant license structure, characteristics of SMR and its effect on license structure, and research progress of U.S Nuclear Regulatory Commission (NRC) are analyzed, SMR license structure in China are proposed, which can be used as a reference for SMR R&D, design and regulation.

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