NRC Generic Letter 79-05, Information Relating to Categorization of Recent Regulatory Guides by the Regulatory Requirements Review Committee

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GL79005

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

January 26, 1979

Docket No.: 50-471

Mr. R. M. Butler Nuclear Projects Manager Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199

Dear Mr. Butler:

SUBJECT: INFORMATION RELATING TO CATEGORIZATION OF RECENT REGULATORY GUIDES BY THE REGULATORY REQUIREMENTS REVIEW COMMITTEE PILGRIM STATION, UNIT 2

We have recently advised utilities with plants in the post-CP phase of the reactor licensing process of the status of NRC staff review and use of recently-approved regulatory guides, and have indicated how these guides would be used in the Operating License review of their Final Safety Analysis Reports. Such information, while not directly applicable to you at this time, may nonetheless be useful to you for your future planning. The text of our letter to these utilities is the following:

"SUBJECT: IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - (Name of Plant) - OPERATING LICENSE REVIEW

During the last several years, we have reviewed and approved several new regulatory guides and branch technical positions or other modifications to existing staff positions. Our practice is that substantive changes in staff positions be considered by the NRC's Regulatory Requirements Review Committee (RRRC) which then recommends a course of action to the Director, Office of Nuclear Reactor Regulation (NRR). The recommended action includes an implementation schedule. The Director's approval then is used by the NRR staff as review guidance on individual licensing matters. Some of these actions will affect your application. This letter is intended to bring you up to date on these changes in staff positions so that you may consider them in your Final Safety Analysis Report (FSAR) preparation.

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"The RRRC applies a categorization nomenclature to each of its actions. (A copy of the summary of RRRC Meeting No. 31 concerning this categorization is attached as Enclosure 1.) Category 1 matters are those to be applied to applications in accordance with the implementation section of the published guide. We have enclosed lists of actions which are either Category 2 or Category 3, which are defined as follows:

Category 2: A new position whose applicability is to be determined on a case-by-case basis. You should describe the extent to which your design conforms, or you should describe an acceptable alternate, or you should demonstrate why conformance is not necessary.

Category 3: Conformance or an acceptable alternative is required. If you do not conform, or do not have an acceptable alternate, then staff-approved design revisions will be required.

"We believe that providing you with a list of the Category 2 and 3 matters approved to date will be useful in your FSAR preparation, and they will be an essential part of our operating license review. Enclosure 2 is a list of the Category 2 matters. Enclosure 3 is a list of the Category 3 matters.

"In addition to the RRRC categories, there also exists an NRR Category 4 list which are those matters not yet reviewed by the RRRC, but which the Director, NRR,, has deemed to have sufficient attributes to warrant their being addressed and considered in ongoing reviews. These Matters will be treated like Category 2 matters until such time as they are reviewed by the RRRC, and a definite implementation program is developed. A current list of Category 4 matters is attached (Enclosure 4). These also should be considered in your FSAR.

"In some instances the items in the enclosures may not be applicable to your application. Also, we recognize that your application may, in some instances, already conform to the stated staff positions, in your FSAR you should note such compliance.

"If you have any questions please let us know."

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For your information, I am enclosing a set of the enclosures that accompanied these individual letters. These enclosures list the present Category 1-4 matters discussed in the letter.

Sincerely,

Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation

Enclosures:

As stated

cc: See next page

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Mr. R. M. Butler

cc: Mr. William Griffin Charles Corkin, II, Esq. Project Engineer Assistant Attorney General Boston Edison Company Commonwealth of Massachusetts 800 Boylston Street One Ashburton Place, 19th Floor Boston, Massachusetts 02199 Boston, Massachusetts 03105

Dale G. Stoodley, Counsel Henry Herrman, Esq. Boston Edison Company 151 Tremont Street, 27K 800 Boylston Street Boston, Massachusetts 02111 Boston, Massachusetts 02199 Mr. & Mrs. Alan R. Cleeton George H. Lewald, Esq. 22 Mackintosh Street Ropes & Gray Franklin, Massachusetts 02038 225 Franklin Street Boston, Massachusetts 02110 W. M. Sides Quality Assurance Manager William S. Abbott Boston Edison Company Attorney & Counsellor at Law 800 Boylston St. 50 Congress Street, Suite 925 Boston, Massachusetts 02199 Boston, Massachusetts 02109 Mr. R. A. Fortney B. N. Pushek EDS Nuclear Bechtel Power Corp. 220 Montgomery Street P. O. Box 3695 San Francisco, California 94104 San Francisco, California 94119 Edward Luton, Esq., Chairman John D. Fassett Atomic Safety and Licensing Board Vice President and General U. S. Nuclear Regulatory Commission Counsel Washington, D. C. 20555 United Illuminating Company 80 Temple Street Dr. Dixon Callihan New Haven, Connecticut 06506 Union Carbide Corporation P. O. Box Y Mr. R. Newman Oak Ridge, Tennessee 37830 Combustion Engineering, Corp. 1000 Prospect Hill Road Dr. Richard F. Cole Windsor, Connecticut 06095 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission W. C. Tallman, President Washington, D. C. 20555 Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105

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Mr. R. M. Butler

cc: Richard S. Salzman, Esq., Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Michael C. Farrar, Esq. Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

SEP 24 1975

Lee V. Gossick Executive Director for Operations

REGULATORY REQUIREMENTS REVIEW COMMITTEE MEETING NO. 31, JULY 11, 1975

1. The Committee discussed issues related to the implementation of Regulatory Guides on existing plants and the concerns expressed in the June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject:

REGULATORY GUIDE IMPLEMENTATION, and made the following recommendations and observations:

a. Approval of new Regulatory Guides and approval of revisions of existing guides should move forward expeditiously in order that the provisions of these regulatory guides be available for use as soon as possible in on-going or future staff reviews of license applications. The Committee noted that over the recent past, the approval of proposed regulatory guides whose content is acceptable for these purposes has experienced significant delays in RRRC review pending the determination of the applicability of the guide to existing plants, often requiring significant staff effort. To avoid these delays, the Committee concluded that, henceforth, approval of proposed regulatory guides should be uncoupled from the consideration of their backfit applicability.

b. The implementation section of new regulatory guides should address, in general, only the applicability of the guide to applications in the licensing review process using, in so far as possible, a standard approach of applying the guide to those applications docketed 8 months after the issuance date of the guide for comment. Exceptions to this general approach will be handled on a case-by-case basis.

c. The regulatory position of each approved proposed guide (or proposed guide revision) will be characterized by the Committee as to its backfitting potential, by placing it in one of three categories:

Category 1 - Clearly forward fit only. No further staff consideration of possible backfitting is required.

ENCLOSURE 1

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Category 2 - Further staff consideration of the need for backfitting appears to be required for certain identified items of the regulatory position--these individual issues are such that existing plants need to be evaluated to determine their status with regard to these safety issues in order to determine the need for backfitting.

Category 3 - Clearly backfit. Existing plants should be evaluated to determine whether identified items of the regulatory position are resolved in accordance with the guide or by some equivalent alternative.

From time to time, for a specific guide, there will probably be some variation among these categories or even within a category, and these three broad category characterizations will be qualified as required to meet a particular situation.

d. It is not intended that the Committee categorization appear in the guide itself. The purpose of the categorization is to indicate those items of the regulatory position for which the Committee can make a specific backfit recommendation without additional staff work (Categories 1 and 3), and to indicate those items for which additional staff work is required in order to determine backfit considerations (Category 2).

e. The Committee recommends that for approved guides in Category 2, staff efforts be initiated in parallel with the process leading to publication of the guide in order that specific backfit requirements for existing plants be determined within a reasonable period of time after publication of the guide.

f. The Committee observed that more attention needs to be given to the identification of acceptable alternatives to the positions outlined in the guides in order to provide additional options and flexibility to applicants and licensees, with the possible benefits of additional innovation and exploration in the solution of safety issues.

2. The Committee reviewed the proposed Regulatory Guide 1.XX: THERMAL OVERLOAD PROTECTION FOR MOTORS ON MOTOR-OPERATED VALVES and recommended approval. This guide was characterized by the Committee as Category 1

- no backfitting, with the stipulation that as an appropriate occasion presented itself in conjunction with the review of some particular aspect of existing plants, the thermal overload protection provisions be audited.

ENCLOSURE 1 (CONT'D)

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3. The Committee reviewed the proposed Regulatory Guide I.XX: INSTRUMENT SPANS AND SETPOINTS and recommended approval subject to the following comment:

Paragraph 5 of Section C (page 4 of the proposed Guide) should be reworded in light of Committee comments, to the satisfaction of the Director, Office of Standards Development. This guide was characterized by the Committee as Category 1 - no backfit.

4. The Committee reviewed Proposed Regulatory Guide 1.97: INSTRUMENTATION FOR LIGHT WATER COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT and deferred further consideration to a later meeting in order to permit incorporation of recent comments by the Division of Technical Review.

Edson G. Case, Chairman Regulatory Requirements Review Committee

ENCLOSURE 1 (CONT'D)

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September 15, 1978

CATEGORY 2 MATTERS

Document Number Revision Date Title

RG 1.27 2 1/76 Ultimate Heat Sink for Nuclear Power Plants

RG 1.52 1 7/76 Design, Testing, and Maintenance Criteria for Engineered-Safety-

Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants (Revision 2 has been published but the changes from Revision 1 to Revision 2 may, but need not, be considered.

RG 1.59 2 8/77 Design Basis Floods for Nuclear Power Plants

RG 1.63 2 7/78 Electric Penetration Assemblies in Containment Structures for Light Water Cooled. Nuclear Power Plants

RG 1.91 1 2/78 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites

RG 1.102 1 9/76 Flood Protection for Nuclear Power Plants

RG 1.105 1 11/76 Instrument Setpoints

RG 1.108 1 8/77 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants

RG 1.115 1 7/77 Protection Against Low-Trajectory Turbine Missiles

RG 1.117 1 4/78 Tornado Design Classification

RG 1.124 1 1/78 Service Limits and Loading Combinations for Class 1 Linear Type Component Supports

RG 1.130 0 7/77 Design Limits and Loading Combinations for Class 1 Plate-

and Shell-Type-Component Supports

(Continued)

ENCLOSURE 2

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CATEGORY 2 MATTERS (CONT'D)

Continued

Document Number Revision Date Title

RG 1.137 0 1/78 Fuel Oil Systems for Standby Diesel Generators (Paragraph C.2)

RG 8.8 2 3/77 Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as Low as is Reasonably, Achievable (Nuclear Power Reactors)

BTP ASB Guidelines for Fire Protection 9.5-1 1 for Nuclear Power Plants (See Implementation Section, Section D)

BTP MTEB 5-7 4/77 Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping

RG 1.141 0 4/78 Containment Isolation Provisions for Fluid Systems

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ENCLOSURE 2 (CONT'D)

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September 15, 1973

CATEGORY 3 MATTERS

Document Number Revision Date Title

RG 1.99 1 4/77 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel materials (Paragraphs C.1 and C.2.

RG 1.101 1 3/77 Emergency Planning for Nuclear Power Plants

RG 1.114 1 11/76 Guidance on Being Operator at the Controls of a Nuclear Power Plant

RG 1.121 0 8/76 Bases for Plugging Degraded PWR Steam Generator Tubes

RG 1.127 1 3/78 Inspection of Water-Control Structures Associated with Nuclear Power Plants

RSB 5-1 1 1/78 Branch Technical Position:

Design Requirements of the Residual Heat Removal System RSB 5-2 0 3/73 Branch Technical Position:

Reactor Coolant System Overpressurization Protection (Draft copy attached)

RG 1.97 1 8/77 Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Paragraph C.3 - with additional guidance on paragraph C.3.d to be provided later)

RG 1.68.2 1 7/78 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants

RG 1.56 1 7/78 Maintenance of Water Purity in Boiling Water Reactors

Attachment:

BTP RSB 5-2 (Draft)

ENCLOSURE 3

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BRANCH TECHNICAL POSITION RSB 5-2 OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORS WHILE OPERATING AT LOW TEMPERATURES

A. Background

General Design Criterion 15 of Appendix A. 10 CFR 50, requires that "the Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."

Anticipated operational occurrences, as defined in Appendix A of 10 CFR 50, are "those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.

Appendix G of 10 CFR 50 provides the fracture toughness requirements for reactor pressure vessels under all conditions. To assure that the Appendix G limits of the reactor coolant pressure boundary are not exceeded during any anticipated operational occurrences, Technical Specification pressure-temperature limits are provided for operating the plant.

The primary concern of this position is that during startup and shutdown conditions at low temperature, especially in a water-solid condition, the reactor coolant system pressure might exceed the reactor vessel pressure-temperature limitations in the Technical Specifications established for protection against brittle fracture. This inadvertent overpressurization could be generated by any one of a variety of mal-

functions or operator errors. Many incidents have occurred in operating plants as described in Reference 1.

Additional discussion on the background of this position is contained in Reference 1.

ENCL 3 (CONT)

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B. Branch Position

1. A system should be designed and installed which will prevent exceeding the applicable Technical Specifications and Appendix G limits for the reactor coolant system while operation at low temperatures. The system should be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to satisfy the Technical Specification limits, particularly while the reactor coolant system is in a water-solid condition.

2. The system must be able to perform its function assuming any single active component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most limiting single active failure. The cause for initiation of the event, e.g., operator error, component malfunction, will not be considered as the single active failure. The analysis should assume the most limiting allowable operating conditions and systems configuration at the time of the postulated cause of the overpressure event. All potential overpressurization events must be considered when establishing the worst case event. Some events may be prevented by protective interlocks or by locking out power. These events should be reviewed on an individual basis. If the interlock/power lockout is acceptable, it can be excluded from the analyses provided the controls to prevent the event are in the plant Technical Specifications.

3. The system must meet the design requirements of IEEE 279 (see Implementation). The system may be manually enabled, however, the electrical instrumentation and control system must provide alarms to alert the operator to:

a. properly enable the system at the correct plant condition during cooldown,

b. indicate if a pressure transient is occurring.

4. To assure operational readiness, the overpressure protection system must be tested in the following manner:

a. A test must be performed to assure operability of the system electronics prior to each shutdown.

b. A test for valve operability must as a minimum be conducted as specified in the ASME Code Section XI.

c. Subsequent to system, valve, or electronics maintenance, a test on that portion(s) of the system must be performed prior to declaring the system operational.

ENCL 3 (CONT)

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5. The system must meet the design requirements of regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants" and Section III of the ASME Code.

6. The overpressure protection system must be designed to function during an Operating Basis Earthquake. It must not compromise the design criteria of any other safety-grade system with which it would interface, such that the requirements of Regulatory Guide 1.29, "Seismic Design Classification" are met.

7. The overpressure protection system must not depend on the availability of offsite power to perform its function.

8. Overpressure protection systems which take credit for an active component(s) to mitigate the consequences of an overpressurization event must include additional analyses considering inadvertent system initiation/actuation or provide justification to show that existing analyses bound such an event.

C. Implementation

The Branch Technical Position, as specified in Section B, will be used in the review of all Preliminary Design Approval (PDA), Final Design Approval (FDA), Manufacturing License (ML), Operating License (OL), and Construction Permit (CP) applications involving plant designs incorporating pressurized water reactors. All aspects of the position will be applicable to all applications, including CP applications utilizing the replication option of the Commission's standardization program, that are docketed after March 14, 1978. All aspects of the position, with the exception of reasonable and justified deviations from IEEE 279 requirements, will be applicable to CP, OL, ML, PDA, and FDA applications docketed prior to March 14, 1978 but for which the licensing action has not been completed as of March 14, 1978. Holders of appropriate PDA's will be informed by letter that all aspects of the position with the exception of IEEE 279 will be applicable to their approved standard designs and that such designs should be modified, as necessary, to conform to the position. Staff approval of proposed modifications can be applied for either by application by the PDA-holder on the PDA-docket or by each CP applicant referencing the standard design on its docket.

The following guidelines may be used, if necessary, to alleviate impacts on licensing schedules for plants involved in licensing proceedings nearing completion on March 14, 1978:

ENCL 3 (CONT)

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1. Those applicants issued an OL during the Period between March 14, 1978 and a date 12 months thereafter may merely commit to meeting the position prior to OL issuance but shall, by license condition, be required to install all required staff-approved modifications prior to plant startup following the first scheduled refueling outage.

2. Those applicants issued an OL beyond March 14, 1979 shall install all required staff-approved modifications prior to initial plant startup.

3. Those applicants issued a CP, PDA, or ML during the period between March 14, 1978 and a date 6 months thereafter may merely commit to meeting the position but shall, by license condition, be required to amend the application, within 6 months of the date of issuance of the CP, PDA, or ML, to include a description of the proposed modifications and the bases for their design, and a request for staff approval.

4. Those applicants issued a CP, PDA, or ML after September 14, 1978 shall have staff approval of proposed modifications prior to issuance of the CP, PDA. or ML.

D. References

1. NUREG-0138, Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3, 1976 Memorandum from Director, NRR, to NRR Staff.

ENCL 3 (CONT)