latory criteria or standards are being developed. For example, Criterion 1 on Quality Assurance states in part:

Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified.

Planned criteria and standards will specify in more detail these quality assurance require

ments.

Most of the General Design Criteria consist of safety performance requirements. The design requirements of the criteria are derived principally from those practices that have been regularly followed in design of recent reactors approved by the Atomic Energy Commission. In a few instances criteria have been included which are more stringent than have been generally required in the past.

Because the terminology of the General Design Criteria is deliberately phrased in terms of general principles of design, it is expected that these criteria will continue to serve their regulatory purpose for a considerable length of time before becoming obsolete or requiring extensive revision. The terminology is such as to leave considerable room for flexibility in interpretation and judgment in determining the appropriate manner in which they should be applied to individual reactor cases. In particular, where uniform requirements could not be established, specific requirements are not defined in these criteria. Rather, they have been left for later development in more detailed criteria and standards, or in the framework of the licensing process where the judgment of those in the AEC's licensing staff, the ACRS, Hearing Boards, reactor designers, and reactor operators can be brought to bear.

(b) Engineering Codes Developed by Professional Societies With AEC Participation

Engineering codes and standards which have been developed with industry through individual membership of AEC personnel in code committees include:

(i) Section III, "Nuclear Vessels," of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. This code, which was approved by ASME in August of 1964 and republished in July 1968, includes detailed requirements concerning materials, design, fabrication, inspection, and testing of pressure vessels for use in nuclear powerplants. Appendix IX of this code, which was issued in December of 1967, includes requirements for quality assurance control programs, nondestructive examination methods, and the qualification of nondestructive examination procedures, equipment, and personnel in order to assure compliance with the Nuclear Vessel Code.

(ii) Section B31.7, "Nuclear Power Piping," of the USA Standards Institute Power Piping Code. This section was published by ASME for trial use and comment in February 1968. The requirements of this code parallel those of the ASME Nuclear Vessel Code.

(iii) Institute of Electrical and Electronic Engineers (IEEE 279) Criteria for Nuclear Power Plant Protection Systems. These criteria are an industry consensus of an acceptable approach to assessing the adequacy of the functional performance of nuclear powerplant protection systems and their reliability in meeting design requirements. Work is in progress on a number of supporting standards and guidelines to supplement IEEE 279, including (1) equipment qualification testing, (2) periodic testing, and (3) numerical reliability analyses techniques.

Criteria Now Being Developed

(a) Supplementary Design Criteria for Pressurized-Water Reactors (PWRs) and Boiling-Water Reactors (BWRs)

In March 1966, the AEC started two projects for developing design criteria jointly with industrial groups. The purposes of these projects

are:

(i) To develop criteria, supplementary to the AEC General Design Criteria, and repre

resenting more definitive and specific application of these criteria to the systems and components of pressurized-water and boilingwater reactors.

(ii) To enlist the expert assistance of additional technical specialists from industry to speed the development of more detailed criteria for these reactor types.

One of the joint efforts is to develop supplementary criteria for PWRs. It was initiated with the former N.6 Reactor Safety Standards Committee of the American Standards Association, and is presently being carried out by the System Engineering Subcommittee of the American Nuclear Society, with support by the AEC staff.

The second joint effort has been undertaken to develop similar supplementary criteria for BWRS. These efforts have been coordinated with the concurrent revision of the General Design Criteria. The Supplementary Design Criteria will be more specific than the General Design Criteria; less so than engineering codes and standards.

(b) Regulatory Criteria for Nuclear Pressure Vessels of Reactor Coolant Systems and Associated Critical Systems

The AEC "Tentative Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels" are being developed in a joint working arrangement between AEC staff personnel and the Advisory Committee on Reactor Safeguards. An ad hoc working group, the AEC-ACRS Primary System Review Group was formed in early 1967. Several members of this Review Group are also members of related ASME code committees. The initial objective was to compile a list of requirementsabove those presently required by section III of the ASME code for nuclear pressure vesselswhich are currently being followed as a matter of good practice, or which it is believed should be followed in the future. The technical requirements compiled as a result of these efforts are being incorporated into, and are consistent with, the AEC standards to be applied to nuclear pressure vessels for the AEC reactor develop、ment program. These technical requirements

95

were made available in August of 1967 to engineering code groups and to others in the nuclear industry in the form of tentative criteria for the design, fabrication, and inspection of pressure vessels for licensed nuclear power reactors. As a result, approximately 20 percent of the criteria have been adopted by the ASME Nuclear Vessel Code Committee and included in the July 1968, revision to section III of the ASME Boiler and Pressure Vessel Code, "Nuclear Vessels," and, consequently, have been deleted from the AEC regulatory criteria. The remainder of the criteria are being revised to reflect comments received from industry prior to publication in the "Federal Register" as a Notice of Proposed Rule Making.

(c) Design Access and Inservice Structural In

tegrity Inspection Requirements for Reactor Coolant Systems and Associated Critical Systems

The AEC-ACRS Primary System Review Group has also developed criteria for design of access and inservice structural integrity inspection requirements for reactor coolant systems and associated critical systems which are considered essential to permit the periodic assessment of the structural integrity and operational safety of water-cooled nuclear power plants. A draft of these criteria was approved by the ad hoc working group. This draft was made available to the USASI/ASME N.45 subcommittee on inservice inspection. This subcommittee is comprised of members from utilities, nuclear component manufacturers, nuclear systems designers, architect engineers, and insurance inspection agencies. It had already started development of similar criteria, and using these and the information contained in the proposed AEC criteria, the USASI/ASME subcommittee prepared and published for committee use, "Recommended Rules for Inservice Inspection of Nuclear Power Unit Components." It was agreed that a member of the ad hoc group from the AEC would participate with the subcommittee in revising the USASI/ASME document for publication. Work on these criteria is continuing.

(d) Regulatory Criteria for Nuclear Power

Piping of Reactor Coolant Systems and
Associated Critical Systems

Upon issuance of the Nuclear Piping Code as a section of the USASI Code for Pressure Piping, USAS B31.7, the AEC-ACRS Primary System Review Group started working on development of AEC criteria beyond those required by USAS B31.7, similar to the AEC criteria covering nuclear pressure vessels. Work on these criteria is still in progress. It is expected that some of these criteria will be adopted by the ASME Nuclear Piping Code Committee and incorporated into USAS B31.7.

(e) Standard for Auxiliary Electrical Power Systems for Nuclear Power Plants

This standard is being prepared by the Institute of Electrical and Electronic Engineers Group, SC-4, Auxiliary Power Supplies, IEEE/Nuclear Science Group/Technical Committee for Standards (TCS) with active support and participation of the AEC. The latest draft of the proposed standard was issued on January 22, 1968.

(f) Supplementary Criteria for Instrumenta

tion and Controls for Nuclear Power Plants These AEC criteria will provide detailed safety criteria for instrumentation and control systems deemed essential to protect the health and safety of the public and are intended to supplement the General Design Criteria for Nuclear Power Plants. A draft of these criteria, prepared by the AEC has been furnished to the IEEE which is developing similar criteria. (g) Supplementary Criteria for Electrical

Power Systems for Nuclear Power Plants These criteria are being developed to ensure that power will be available under postaccident conditions for the safe shutdown of the reactor. Criteria are being included for:

(i) System design to ensure adequate power for essential functions under normal and postaccident conditions.

(ii) Operating procedures of on-site power sources to minimize the probability of a complete loss of pumping power.

(iii) Testing of installed power systems to periodically confirm that their performance is adequate.

Copies of the initial draft of these criteria dated March 1, 1968, have been supplied to the IEEE Technical Committee for Standards and the American Nuclear Society (ANS) Systems Engineering Subcommittee for review and

comment.

(h) Quality Assurance Requirements

Experience to date with design, construction, and operation of nuclear power plants has made it abundantly clear that there is an urgent need for augmented efforts by the AEC, the utilities, the plant suppliers, and the various technical societies to develop and apply improved quality assurance practices. To this end, the following effort is in progress to provide detailed quality assurance requirements to implement and supplement Criterion 1 of the General Design Criteria:

(i) Quality assurance requirements for specific components and systems are being included in the criteria and codes applicable to the components and systems. For example, as indicated previously, Appendix IX of the ASME Nuclear Vessel Code, which was issued in December of 1967, includes requirements for quality control programs, nondestructive examination methods, and the qualification of nondestructive examination procedures, equipment, and personnel in order to ensure compliance with the code.

(ii) The AEC-ACRS Primary System Review Group is preparing requirements for a quality assurance system at the construction site of a nuclear power plant. These requirements will apply to the administrative and technical aspects of the construction phase and will include provisions for effective coordination and execution of responsibilities shared by individual contractors involved during the entire construction phase. Included in the quality assurance system requirements are provisions to ensure ready detection of nonconforming practices and timely and positive corrective action.

(iii) The USASI Nuclear Standards Board N.45 Committee (Reactor Plants and