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A Group Effort that Grew PUBLIC ACCESS

As the Industry's Needs Expanded, So Did the Scope of ASME's Nuclear Codes and Standards.

Mechanical Engineering 136(05), 42-45 (May 01, 2014) (4 pages) Paper No: ME-14-MAY3; doi: 10.1115/1.2014-May-3

This article discusses why it is essential to develop new codes and standards for nuclear power industry. The reason for developing new codes for nuclear power sector is simple to understand. It has to do with the fundamental purpose of all standards: Standards exist to serve not only all the stakeholders in an industry – manufacturers, regulators, insurers, operators of equipment, but also the members of the general public who happen to be in the neighborhood. Standards support prosperity and, more important than that, they maintain public safety. Presently, different committees are working on the next generation of their standards in the nuclear power sector. They are incorporating recent experience and integrating new technologies, from materials to theoretical tools. The ASME/ANS Nuclear Risk Management Committee is currently expanding the scope of the standard to cover risk at shutdown and include long-term maintenance of containment and releases to the public after an accident. In addition, requirements for advanced reactors and the lessons learned from the Fukushima accident are under development.

When the first edition of Section III of the ASME Boiler and Pressure Vessel Code appeared 50 years ago, it provided rules for three classes of pressure vessels for nuclear power plants.

This was, however, the birth of an industry, an entire supply chain that would eventually provide one-fifth of the electricity consumed annually by the United States. So it quickly became apparent that the industry needed to address many issues besides the design and construction of the reactor vessels. More guidance was needed—and welcomed—by the industry and other stakeholders.

Section III eventually grew to encompass rules governing the construction and inspection during the building of storage tanks, piping, pumps, valves, containments, and other components of nuclear power plants. The code also addresses containment systems for storage and transport packaging of spent fuel and high-level radioactive material and waste.

Section XI of the BPVC concerns inservice inspection of critical components in nuclear power plants. It includes provisions for repair or replacement of components, and procedures for evaluating plant operating events.

Fifty years ago, the new Code helped give birth to an industry.

Over the years, new committees formed to address issues ranging from quality assurance to risk management.

In contrast to the Boiler Code rules for nuclear components, which were essentially incorporated by reference into the U.S. Nuclear Regulatory Commission's regulations and therefore required for licensing, several standards were developed for voluntary adoption by the users. However, most of them have been identified by the Nuclear Regulatory Commission as acceptable methods for meeting its requirements and have frequently been treated as requirements. Regulatory endorsement of standards is consistent with government policy to prefer to endorse standards that represent the affirmation of all affected and knowledgeable interests, such as those developed by ASME Standards and Certification, rather than issue government rules and guidance.

Today, ASME's Board on Nuclear Codes and Standards oversees six committees, in addition to Section III and Section XI. Together the eight committees have issued 22 nuclear codes and standards.

There are almost 1,300 volunteers participating in the various nuclear committees, subcommittees, and working groups. They include approximately 100 international participants from eight countries besides the United States.

The following are the additional nuclear codes and standards committees and their products.

In the late 1960s, a high-level committee under the aegis of the American National Standards Institute reviewed the status of nuclear standards. After consulting with the former Atomic Energy Commission, ANSI concluded that there was a need for standards to assure that the quality of plant construction was acceptable.

As a result, ANSI asked its committee N45, which was managed by ASME, to develop a series of quality assurance and control standards. The first standard to be issued covered a general quality assurance program. Later, the group issued several construction-phase control and inspection standards.

After the Board on Nuclear Codes and Standards was created, this activity was transferred to a newly formed ASME Committee on Nuclear Quality Assurance and the individual standards were consolidated. They currently reside, along with requirements developed over that past 40 years, in a single document, NQA-1 Quality Assurance Requirements for Nuclear Facility Applications.

The original scope of the QA standard covered nuclear plants and was recognized by the Nuclear Regulatory Commission for meeting its requirements. Subsequently, it was expanded to cover other nuclear facilities and also adopted by the U.S. Department of Energy for its facilities.

Sidney Bernsen

The ASME Operation and Maintenance Standards Committee was established in the late 1970s with the direction from the Board on Nuclear Codes and Standards to develop standards and guides for testing of nuclear power plant components in operating plants. The primary focus of the committee was testing of pumps, valves, and snubbers to replace similar sections in place at that time contained in Section XI of the Code.

To date, the committee has developed a document comprising three divisions for industry use and regulatory endorsement. The latest approved version is the 2012 edition, OM-2012a Operation and Maintenance of Nuclear Power Plants.

Division 1 contains Code requirements for testing nuclear power plant safetyrelated pumps, valves, and snubbers. It also incorporates risk-informed testing for components and provides for upgrades necessary for new generation construction. Division 2 consists of standards, and Division 3 has guides. Both provide testing methodologies for other nuclear plant components such as heat exchangers and diesel generators.

The U.S. Nuclear Regulatory Commission has endorsed Division 1 with minor exceptions. Other international regulatory bodies have also endorsed this standard.

The work of this committee has spawned nuclear industry users groups as well as a triennial NRC/ASME symposium on pump, valve, and snubber testing.

John Zudans

In 1971, the ANSI N45.8 Committee was organized to develop standards for highreliability air cleaning equipment for nuclear facilities and corresponding tests to confirm performance of the equipment. Two standards, ANSI/ASME N509-1976 and ANSI/ ASME N510-1975 were published.

In 1976, the committee was reorganized as the ASME Committee on Nuclear Air and Gas Treatment. The scope of responsibility increased to include the development of standards for design, fabrication, inspection, and testing of air cleaning and conditioning components used in nuclear facilities. ASME AG-1 was the new standard resulting from the increased scope. The standard contains requirements, specific prohibitions, and guidance for construction activities.

The first edition of AG-1 Code on Nuclear Air and Gas Treatment was approved by the American National Standards Institute in 1985. The current version is AG-1-2012. This Code provides requirements for the performance, design, fabrication, installation, inspection, acceptance testing, and quality assurance of equipment used in nuclear power plants. In addition the committee has produced ASME-N511-2007 In-Service Testing of Nuclear Air Treatment, Heating, Ventilating, and Air-Conditioning Systems.

Thomas Vogan

Federal regulations require measures to ensure that key equipment in nuclear power plants operates as specified under extreme environmental and emergency conditions. These include seismic and other conditions occurring in the event of accidents. While the NRC published rules and guides, the development of qualification standards was assumed by the nuclear standards community.

In the early 1970s, initial development of qualification standards was assigned to the ANSI N45 Committee. This committee initiated separate working groups, one on valves and one on pumps, to prepare standards to ensure that pumps and valves used in nuclear plants would function as specified. Subsequently these groups became separate subcommittees, reporting to different standards committees. After the ASME Board on Nuclear Codes and Standards was formed, the mechanical engineering scope of these standards was assigned to a new ASME Committee on Qualification of Mechanical Equipment, and the Institute of Electrical and Electronics Engineers assumed the responsibility for environmental qualification standards.

The development of standards never ends. Committees are already working on the next generation.

The first qualification standard to be issued for valves was ANSI N278.1-1975, which covered the preparation of functional specifications. Subsequently ANSI B16.41 was issued to cover functional qualification requirements for power-operated active valve assemblies for nuclear power plants. In 1994 the QME committee published QME-1 1994 that included seismic and functional qualification of active mechanical equipment, including pumps and valves. This standard replaced ANSI 278.1-1975 and has been revised several times to incorporate experienced based information and analytical techniques. The 2007 version has been endorsed by the U.S. Nuclear Regulatory Commission.

An updated version of the standard was issued in 2012 and includes new sections on standardization of experienced-based seismic equipment qualification and the qualification of dynamic restraints. It also requires that users of the standard must provide a Qualification Specification.

Tom Ruggiero

The ASME Standards Committee on Cranes for Nuclear Facilities was established in 1976, shortly after the U.S. Nuclear Regulatory Commission issued NUREG-0554, written guidelines for safety-critical or single-failure-proof cranes (i.e., cranes that cannot drop a load with the failure of a single component). While the NUREG provided guidelines, there was need for standards to capture, maintain, and extend the guidelines to appropriate design requirements and details as well as manufacturing, storage, erection, inspection, and testing. In 1980, the committee's scope was broadened from nuclear power plants to include other critical load-handling facilities.

The committee maintains two ANSI approved standards, NOG-1 Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) and NUM-1 Rules for Construction of Cranes, Monorails, and Hoists (with Bridge or Trolley or Hoist of the Underhung Type). Both the standards can be applied to cranes at facilities other than nuclear plants, where enhanced crane safety may be required.

The committee's membership is derived from the nuclear power industry, aerospace industry, crane manufacturers, crane suppliers, the U.S. Navy, the U.S. Department of Energy, and the U.S. Nuclear Regulatory Commission. The standards have been adopted for critical load handling by industry and government participants.

Aaron Kureck

Standards support prosperity and, more important than that, they maintain public safety.

Probabilistic risk analysis was introduced to nuclear power plant safety evaluations in the 1970s. It was refined to evaluate plants after the Three Mile Island accident, and use of risk information was introduced into the ASME Codes and Standards in the late 1990s by providing risk-informed alternative rules for inservice inspection and testing and for selected operation and maintenance activities.

Probabilistic risk analyses evolved over a period of more than 30 years from relatively simple, limited-scope evaluations to rather complex, full-scope modeling of the plant. During this time, there was no standard to determine the technical capability, fidelity, and adequacy of the risk assessment. As codes, standards, and regulatory applications evolved, there was clearly a need to develop standards for risk analysis to support these applications.

The ASME Board on Nuclear Codes and Standards in 1998 initiated a project to develop an appropriate risk analysis standard. The initial standard, ASME RA-S-2002 Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications was published in 2002, with Addenda issued in 2003 as ASME RA-Sa-2003.

The initial scope of this standard covered currently operating light water nuclear power plants at power, since practices for this scope were the most highly developed and supported most of the current applications. In parallel, the American Nuclear Society undertook efforts to develop standards to cover internal and external conditions, long-term release probability, and offsite consequences that were not addressed by the ASME standard. These areas required more time and technical effort since their considerations were not as highly developed or as well understood at the time. ANS eventually published standards covering internal fires and external events, such as earthquakes, flooding, and tornadoes.

Now several of these standards have been combined into a single document, ASME/ANS RA-S-2008. The standard is the responsibility of one committee, which reports to the ASME Board on Nuclear Codes and Standards and to the American Nuclear Society's Standards Board. The latest approved version of the standard is Addendum B (ASME/ANS RA-Sb-2013).

This joint committee has now prepared and issued, or will shortly issue for trial use, PRA standards for advanced light water reactors and non-light water reactors, and extensions to evaluate the risk from long-term containment failures and off-site consequences.

Sidney Bernsen

The development of standards never ends. Committees are already working on the next generation of their standards. They are incorporating recent experience and integrating new technologies, from materials to theoretical tools.

The ASME/ANS Nuclear Risk Management Committee, for example, is currently expanding the scope of the standard to cover risk at shutdown and from long-term maintenance of containment and releases to the public after an accident. In addition, requirements for advanced reactors and the lessons learned from the Fukushima accident are under development.

The reason that the work must continue is simple to understand. It has to do with the fundamental purpose of all standards: Standards exist to serve all the stakeholders in an industry—manufacturers, regulators, insurers, operators of equipment, but also the members of the general public who happen to be in the neighborhood. In short, standards support prosperity and, more important than that, they maintain public safety.

Copyright © 2014 by ASME
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