ML20198Q151

Forwards Preliminary NRC Staff Views Re Criteria That Could Be Used for Determining What Constitutes Design Basis as Defined in 10CFR50.2

ML20198Q151
Person / Time
Issue date:	01/04/1999
From:	Matthews D NRC (Affiliation Not Assigned)
To:	Pietrangelo A NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT &
References
PROJECT-689 NUDOCS 9901070220
Download: ML20198Q151 (6)

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January 4, 1999 Mr. Anthony Pirtrangslo, Dirrctor Licensing Nuclear Energy Institute Suite 400 1776 i Street, NW

. Washington, DC 20006-3708

SUBJECT:

DRAFT CRITERIA FOR DETERMINING DESIGN BASIS INFORMATION AS DEFINED IN 10 CFR 50.2

Dear Mr. Pietrangelo:

This letter forwards preliminary NRC staff views regarding criteria that could be used for determining what constitutes design basis information as defined in 10 CFR 50.2. As we agreed during a meeting on September 19,1998, a clear, common, understanding of what constitutes design basis information as defined in 10 CFR 50.2 is important to the industry and the staff. I believe that sharing preliminary views on this subject is crucial to developing such an understanding and I hope that the enclosed draft criteria will prove helpful.

It is my understanding that the Nuclear Energy Institute (NEI) is working with the industry to update NEl 97-04, " Design Basis Program Guidelines," and that you intend to publish a revised i

version in January 1999. I encourage you to consider the staff's positions in your deliberations on this document.

I look forward to continuing our discussion of design bases issues and suggest that we set up a meeting for mid-January. Please feel free to call me or Stewart Magruder of my staff with any questions.

Sincerely, Original Signed By:

David B. Matthews, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation l

Project No. 689 7

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WASHINGTON. D.C. 2055:HX)01 January 4, 1999 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 i Street, NW Washington, DC 20006-3708

SUBJECT:

DRAFT CRITERIA FOR DETERMINING DESIGN BASIS INFORMATION AS DEFINED IN 10 CFR 50.2

Dear Mr. Pietrangelo:

Tnis letter forwards preliminary NRC staff views regarding criteria that could be used for determining what constitutes design basis information as defined in 10 CFR 50.2. As we agreed during a meeting on September 19,1998, a clear, common, understanding of what constitutes design basis information as defined in 10 CFR 50.2 is important to the industry and the staff. I believe that sharing preliminary views on this subject is crucial to developing such an understanding and I hope that the enclosed draft criteria will prove helpfu!.

It is my understanding that the Nuclear Energy institute (NEI) is working with the industry to update NEl 97-04, "Desiga Basis Program Guidelines," and that you intend to publish a revised version in January 1999. I encourage you to consider the staff's positions in your deliberations on this document.

I look forward to continuing our discussion of design bases issuas and suggest that we set up a meeting for mid January. Please feel free to call me or Stewart Magruder of my staff with any questions.

Sincerely, I

WIs&

David B. Matthews, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 689

Enclosure:

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b DRAFT December 10,1998 Criteria for determining design basis information as defined in 50.2 10 CFR 50.2 definition: Design Bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.

These values may be (1) restraints derived from generally accepted " state of the art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.

General Guidance:

The design basis information required by the definition in 10 CFR 50.2 is that information that describes:

1. The functions of a olant's structures, systems, and components (SSCs), where those functions and/or SSCs are required by NRC regulations, license condition, or order; and

2. The reference values to be used in the design process such that an SSC meets the functional requirements defined above under all conditions in which the SSC is required to function (e.g., capacity, rating, system output, limiting safety system settings).

Soecific Guidance:

a. Functional requirements are applicable to all conditions of plant operation during which an SSC may operate. These conditions include, in addition to accidents and anticipated operational occurrences: plant startup, normal operation, shutdown, emergency operation, special or infrequent operation, and system abnormal or emergency operation.

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b. Structure and component functions may be implicitly subsumed within UFSAR descriptions of j

system functions. Although every structure and component has a design basis, only those l

structure and component functions developed to meet requirements need to be explicitly l

described.

c. The design bases of a facility are a subset of the licensing basis and are required pursuant to 10 CFR 50.34(a)(3)(ii) and (b) to be included in the FSAR.

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d. Additional design information, including; design analyses, design output documents, and information regarding imptementation of the design bases is contained in other documents, some of which are docketed and some of which are retained by the licensee.

Enclosure DRAFT

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4 DRAFT December 10,1998 t

e. The design bases shou!d provide sufficient detail to describe why an SSC was designed as it was and how applicable requirements are met. Sufficient detail may require the inclusion of the following information:

1. A definition of those events, transients, and accidents for which the SSC must function as designed.

2. A definition of those events, transients, and accidents for which the SSC must be designed to withstand.

3. Codes and standards including the applicable issue and/or addenda.

4. Design conditions such as pressure, temperature, fluid chemistry and voltage.

5. Loads such as seismic, wind, thermal, and dynamic.

6. Environmental conditions anticipated during normal operation and during transient and accident events such as pressure, temperature, humidity, corrosiveness, nuclear radiation, electromagnetic radiation, water table fluctuations, and duration of exposure.

7. Interface requirements including definition of the functional and physicalinterfaces involving SSCs (e.g., electrical / mechanical system interface, safety /non-safety system interface)

8. Material requirements including such items as compatibility, electricalinsulation properties, protective coating, and corrosion resistance.

9. Mechanical requirements such as vibration, stress, shock, and reaction forces.

10. Structural requirements covering such items as equipment foundations, electrical raceway supports, and pipe supports.

11. Hydraulic requirements such as pump net positive suction head (NPSH), allowable pressure drops, allowable fluid velocities, maximum pump discharge pressure, and minimum or maximum flow rates necessary to assure safety functions are met.

12. Chemistry requirements such as limitations on water chemistry.

13. Electrical requirements such as capacity / capability of power supplies, power supply availability, maximum / minimum voltage, voltage quality, raceway cable fill, raceway ventilation, cable derating, and maximum / minimum load necessary to assure safety functions are met.

DRAFT

s.,.

DRAFT oecember 10,1998

14. Instrumentation and control requirements for normal operation, reactor protection, and transient and accident mitigation.

15. Requirements for redundancy, independence, and testability to meet single failure requirements.

16. Analysis and procedural requirements to demonstrate operability of SSCs that lack l

testability, l

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Nuclear Energy Institute Project No. 689 cc:

Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400 i

Suite 400

- 1776 i Street, NW 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708 Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy Institute ABB-Combustion Engineering, Inc.

Suite 400 12300 Twinbrook Parkway, Suite 330 1776 i Street, NW Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Engineering Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400

~1776 I Street, NW Washington, DC 20006-3708 Mr. Nicholas J. Liparuto, Manager Nuclear Safety and Regulatory Activities l

Nuclear and Advanced Technology Division l.

Westinghouse Electric Corporation i

P.O. Box 355 Pittsburgh, Pennsylvania 15230 i

l Mr. Jim Davis, Director Operations Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 r

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