Update on Code Case N-755 Revision Class 3 Polyethylene Piping

2009 ◽  
Author(s):  
Matthew G. Golliet
Keyword(s):  
United States ◽  
Nuclear Regulatory Commission ◽  
Water Systems ◽  
Polyethylene Pipe ◽  
Nuclear Plants ◽  
Service Water ◽  
Asme Code ◽  
Pipe Replacement ◽  
Regulatory Commission ◽  
Essential Service

Nuclear plants have received United States Nuclear Regulatory Commission (NRC) approval to use polyethylene pipe in safety related systems such as essential service water systems. ASME Code Case N-755 is the basis for the utility requests. ASME Nuclear Code committees are developing a revision to provide clarification to the code case requirements and address NRC concerns. Polyethylene pipe replacement projects provide a solution to the age degradation issues such as corrosion and fouling of buried carbon steel pipe.

Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities

2008 ◽  
Author(s):  
Frank J. Schaaf
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Regulatory Review ◽  
Plastic Pipe ◽  
Service Water ◽  
Division 1 ◽  
Minimum Requirements ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Applications

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

Background to Recent Revision of the Section III Seismic Piping Rules

2002 ◽  
Author(s):  
John Minichiello ◽  
Ernest B. Branch ◽  
Timothy M. Adams ◽  
Yasuhide Asada ◽  
Richard W. Barnes
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Experimental Program ◽  
Japanese Industry ◽  
Regulatory Commission ◽  
Special Working ◽  
Special Working Group ◽  
Considerable Discussion

The new rules for seismic piping design in Section III that were developed and included in the requirements in 1994 Addenda of the ASME Boiler and Pressure Vessel Code (B&PV Code) generated considerable discussion within the industry and from the United States Nuclear Regulatory Commission, (USNRC). The USNRC initiated a review of the results of the previous EPRI/NRC experimental program and the Japanese industry started its own experimental program. To accommodate and address developments resulting from these efforts, the ASME, B&PV Code established a Special Working Group (SWG) to continue the review and study of the questions and information generated. This paper reports on the efforts of this SWG which resulted in refinements of the revised rules. These refinements have been accepted for inclusion in Section III of the ASME, B&PV Code.

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.

Non-Proprietary Application of ASME Code Section III, Appendix N, to SONGS Replacement Steam Generators

2017 ◽  
Author(s):  
R. D. Blevins
Keyword(s):  
Elastic Stability ◽  
Nuclear Regulatory Commission ◽  
Stability Diagram ◽  
Design Flow ◽  
Mode Shapes ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Public Data ◽  
Asme Code ◽  
Regulatory Commission

Flow-induced vibration analysis of the San Onofre Nuclear Generating Station (SONGS) Replacement Steam Generators is made using non-proprietary public data for these steam generators on the Nuclear Regulatory Commission public web site, www.NRC.com. The analysis uses the methodology of Appendix N Section III of the ASME Boiler and Pressure Vessel Code, Subarticle N-1300 Flow-Induced Vibration of Tubes and Tube Banks. First the tube geometry is assembled and overall flow and performance parameters are developed at 100% design flow, then analysis is made to determine the flow velocity in the gap between tubes and tube natural frequencies and mode shapes. Finally, the mass damping and reduced velocity for tubes on the U bend are assembled and plotted on the ASME code Figure N-11331-4 fluid elastic stability diagram.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.

Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials

2005 ◽  
Author(s):  
V. N. Shah ◽  
B. Shelton ◽  
R. Fabian ◽  
S. W. Tam ◽  
Y. Y. Liu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Fissile Material ◽  
Radioactive Materials ◽  
Asme Code ◽  
Material Transportation ◽  
Accident Conditions ◽  
Regulatory Commission ◽  
High Level

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.

Management of Mixed Hazardous and Radioactive Waste in the United States

2001 ◽  
Author(s):  
William Greenman ◽  
Kimberly Cole
Keyword(s):  
United States ◽  
Regulatory Authority ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Slow Development ◽  
Treatment Facilities ◽  
Mixed Waste ◽  
Treatment And Disposal ◽  
Regulatory Commission ◽  
The U.S

Abstract In the United States, mixed-waste is typically defined as waste that contains both radioactive constituents and non-radioactive constituents that pose a threat to human health or the environment (hazardous waste). Prior to 1986 the U.S. Nuclear Regulatory Commission (NRC) had sole regulatory authority over mixed-waste because of its radioactive constituents. In 1986, however, the U.S. Environmental Protections Agency (EPA) was granted regulatory authority over the hazardous constituents in mixed-waste; and, a system of dual regulation was created. Dual regulation of mixed-waste by the EPA and the NRC has caused significant problems for the regulated community. The burden of dual regulation has contributed to the slow development of treatment technologies, and to the overall lack of treatment capacity available to U.S generators of mixed-waste. This paper reviews the requirements that the EPA and the NRC mandate with regard to mixed-waste generation, treatment and disposal; and it explores technical impacts of those requirements as they relate to generators, treatment facilities and the public.

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