ML20136E816
| ML20136E816 | |
| Person / Time | |
|---|---|
| Issue date: | 10/20/1983 |
| From: | Harold Denton NRC |
| To: | Stello V NRC |
| Shared Package | |
| ML20136E769 | List: |
| References | |
| CON-IIT07-659-91, CON-IIT7-659-91 83-095, 83-95, NUREG-1455, NUDOCS 8403140316 | |
| Download: ML20136E816 (9) | |
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I
. O I October 20, 1983 bNDI b00$
NOTE TO VICTOR STELLO 0. O -09s< ,
l l Enclosed for the information of the CRGR are final drafts of two NRC l letters concerning "important to safety" and " safety-related". The first, prepared by and concurred in by ELD, is a statement of the staff j position on the definition of these two terms and the legal basis for l 1 that position. This proposed reply to the counsel for a Utility l Classification Group has been concurred in by all affected NRR Divisions,
- IE and Research, i
i The second final draft, which has also been concurred in by IE, Research, I ELD and affected NRR Divisions, is a generic letter to all licensees and '
, pennittees which sets forth the staff position on this issue and, based r i on that position, advises pennittees and licensees of their responsibility
} under the current regulations to develop and implement quality assurance t programs for both safety-related equipment and equipment important to l safety. The generic letter also requests comments on the need for and
- methods of developing additional guidance on suitable quality assurance programs for plant equipment important to safety not covered by Appendix B.
Since neither of the two replies establishes new requirements for permittees
, or licensees, but rather informs them of their responsibility to comply !
j with existing requirements, tht.re appears to be no reason for CRGR review. [
{ If, however, additional regulatory guidance is subsequently developed as ,
- a result of comments received in response to the generic letter, that new
. guidance would be subject to CRGR review. l The letters are now being prepared in final form for signature by the
! Director, NRR and the Director, DL, Office of Nuclear Reactor Regulation, i respectively. In response to the ED0's request, they will be reviewed by '
him prior to dispatch. ;
- j. Should you believe that a briefing of CRGR is desirable, we will be pleased l to comply. ,,,
. , . AM -f f d/HaroldR.Denton '
i
Enclosures:
! 1. Final Df t Ltr to T. S. Ellis, Counsel
! to the Utility Classification Group
- 2. Final Dft Generic Ltr for all Permittees
! and Licensees 4
l cc: See Next Page '
+ . ,
s Victor Stello October 20, 1983 cc: Guy Cunningham ELD Bill Olmstead, ELD R. DeYoung, IE Jim Taylor, IE R. B. Minogue, RES Guy Arlotto, RES D. Eisenhut, NRR R. Mattson, NRR R. Vollmer, NRR T. Spets, NRR
DRAFT T. S. Ellis, III, Esq.
Hunton & Williams 707 East Main Street P. O. Box 1535 Richmond, Virginia 23212
Dear Mr. Ellis:
The Executive Director for Operations has asked me to respond to your letter of August 26, 1983, in which you express concern, on behalf of the Utility Safety Classification Group, over the NRC use of the terms "important to safety" and
" safety-related." Your concern appears to be principally derived from recent licensing cases in which the meaning of these terms in regard to NRC quality assurance requirements has been at issue, and my memorandum to NRR personnel of November 20, 1981.
I agree that the use of these terms in a variety of contexts over the past several years has not always appeared to be consistent. In recognition of this
-. problem I attempted in my 1981 memorandum to NRR personnel to set forth definitions of these terms for use in all future regulatory documents and staff testimony before the adjudicatory boards. As you are aware, the position taken in that memorandum was that "important to safety" and " safety-related" are not syn'onymous terms as used in the Commission regulations. The fomer encompasses the broad scope of equipment covered by Appendix A to 10 CFR Part 50, the General Casign Criteria, while the latter refers to a narrower subset of this class of equipment defined in Appendix A to 10 CFR Part 100 Section VI(a)(1) and, more recently, in 10 CFR 50.49(b)(1). Based on such a distinction between these tems, it has been and would continue to be staff practice to apply the quality i
DRAFT assurance requirements of Appendix B to 10 CFR Part 50 only to the narrower class of " safety-related" equipment, absent a specific regulation directing otherwise.
More importantly, however, this does not mean that there are no existing NRC quality standards or quality assurance requirements for the broader class of equipment which does not meet the definition of " safety-related." General Design Criterion 1 requires quality standards and a quality assurance program for all structures, systems and components "important to safety." These requirements, like those of Appendix B to 10 CFR Part 50, are " graded" in that GDC-1 mandates the application of quality standards and programs " commensurate with the importance of the safety functions to be performed," and expressly allows the use of " generally recognized codes and standards" where applicable and sufficient. Pursuant to our regulations, permittees or licensees are responsible for developing and implementing quality assurance programs for plant design and construction or for plant operation which meet the more general requirements of GDC-1 for plant equipment "important to safety " and the more prescriptive requirements of Appendix B for " safety-related" plant equipment.
'This distinction between the terms "important to safety" and " safety-related" has been accepted in two recent adjudicatory decisions where the issue was squarely faced. In the Matter of Metropolitan Edison Company, et. al. (Three Mile Island Nuclear Station, Unit 1), ALAB-729, ?!RC (tiay 26,1983):
In the Matter of Long Island Lighting Company (Shoreham Nuclear Power Station, Unit 1),LBP , NRC ( ). Moreover, the Commission itself recognized
DRAFT 4
and endorsed a distinction between the terms in promulgating the Seismic and Geologic Siting Criteria for Nuclear Power Plants (see Section VI(a)(1) and VI(a)(2) of Appendix A to 10 CFR Part 100) and the Environmental Qualification Rule (see Supplementary Infomation and 10 CFR 50.49(b)). Also, in preparing this response, members of the licensing staff and legal staff reviewed all of the material on this subject provided by your letter, and have also reviewed
] numerous other regulatory documents, including both staff and Commission issuances over the past several years in which the tems " safety-related" and "important to safety, are used. Based on these reviews, while it is apparent that some confusion continues to exist with regard to the distinction between the terms, the staff is convinced that the position it has previously taken l
remains correct. .
i The final point which I considered in responding to your letter is the consistency 1 ,
- of NRC staff practice over the years with our position on this issue, and the technical basis for that practice. While previous staff licensing reviews were not specifically directed towards determining whether in fact permittees or licensees have implemented quality assurance programs which adequately address
, all" structures, systems, and components important to safety, this was not because of any concern over lack of regulatory requirements for this class of equipment.
! , Rather, our practice was based upon: (1) the staff view that nomal industry i
l practice is generally acceptable for most, if not all, equipment not covered by i
- Appendix 8 within this class; and (2) the absence of indications that licensees or pemittees were not implementing adequate quality assurance program for plant
- equipment important to safety or were disputing the coverage of such equipment to i
i i-
" DRAFT 4
by GDC-1. Our experience has indicated that both of these premises should be reconsidered. Accordingly, we plan to use the distinction in the meaning of the two terms discussed above in our future regulatory reviews of rermittee's and
, licensee's committments to develop and implement those quality assurance programs required by the regulations, and to request industry views on the need for and method of developing additional guidance concerning acceptable quality assurance programs for those structures, systems and components important to safety not covered by Appendix B to 10 CFR Part 50.
The intent of the generic letter in preparation, which prompted your letter to Mr. Dircks, was to clarify the meaning of the terms and, more importantly, to inform permittees and licensees of their responsibility under the regulations to develop and implement quality assurance programs consistent with these terms. Your latter has served to highlight the need for prompt NRC action on this subject, and the generic letter to which your referred is being issued contemporaneously with this response. A copy of that letter is enclosed for
, your information.
Sincerely, Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated
DRAFT 10/17/83 Generic Letter to All Nuclear Facilities Holding an Operating License or Construction Pemit 5
The primary purposes of this letter are: (1) to bring to the attention of holders of construction permits their obligation and responsibility under the Commission's regulations to develop and implement quality assurance programs for the design, fabrication, erection and testing of structures, systems and components "important to safety"N, as well as for " safety-related" structures, systems and componentsU; (2) to bring to the attention of holders of operating licenses their obligation and responsibility to develop and implement managerial and administrative controls to assure safe operation (i.e., cuality assurance programs for operation) of structures, systems and
- components "important to safety"N, as well as for " safety-related" structures,
- systems and componentsU; and (3) to inform both permittees and licensees of the definitions of the tems " structures, systems and components important to safety" and " safety-related structures, systems and components" as they apply to the quality assurance requirements of the Commission regulations.
The letter is prompted by indications that some permittees and licensees may not be defining these terms in a manner consistent with the definitions established by NRC regulations, and, as a result, may not be properly
- mplementing the Commission's regulations.
U Criterion 1 of the General Design Criteria for Nuclear Power Plants (Appendix A to 10 CFP. Part 50)
U Quality Assurance Criteria for Nuclear Power Plants ( Appendix B to 10 CFR Part 50)
. DRAFT 2
i The term structures, systems and components important to safety refers to the broad scope of plant equipment covered by the General Design Criteria for Nuclear Power Plants (Appendix A to Part 50) and is defined in Appendix A to be " structures, systems and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public."
- The term safety-related structures, systems, and components refers to a narrower class of equipment defined in Appendix A to 10 CFR Part 100, and, more recently, in 10 CFR 50.49(b)(1) to be that relied upon to remain functional during and following design basis events to ensure "... (i) the integrity of the reactor coolant pressure boundary; (ii) the capability to shut dona the reactor and maintain it in a safe shutdown condition, and (iii) -the capability to prevent or mitigate the consequences of accidents ..."
j In 10 CFR 50.49, the Commission also made it clear that it considered equipment I that is " safety-related" to be a subset of equipment "important to safety."
l Thus, the quality assurance requirements of General Design Criterion 1 must
! be applied to a broader set of structures, systems and components than those which are defined as " safety-related." Accordingly, permittees and licensees are expected to develop and implement quality assurance programs pursuant to GDC-1 for the broader set of structures, systems and components defined to be important to safety, and to continue to develop and implement quality assurance programs pursuant to Appendix B to 10 CFR Part 50 for the narrower set of
r . .
O DRAFT plant equipment defined to be safety-related. Future regulatory reviews will be directed towards determining whether permittees and licensees have made a suitable committment to develop and implement those quality assurance programs required by the regulations.
A secondary purpose of this letter is to help us determine whether there is a need for the industry and/or the staff to develop more specific guidance concerning the quality assurance programs appropriate for plant structures, systems and components which are important to safety, but not safety-related, as well as guidance for identifying which plant features fall within the broader class of equipment. Any views you may have on whether guidance in either area is needed and suggestions on how it should be developed would be welcomed and may be sent to the Director, Division of Quality Assurance, Safeguards and Inspection Programs, Office of Inspection and Enforcement.
We have received a recent letter from a utility safety classification group that deals with the subject of "important to safety" and " safety-related." Enclosed, at the Group's request, is a copy of that letter and the flRC reply.
Sincerely, I
Darrell G. Eisenhut, Director Division of Licensing i Office of Nuclear Reactor Regulation Enclosure :
As stated
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