ML20132C610
| ML20132C610 | |
| Person / Time | |
|---|---|
| Issue date: | 12/20/1984 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML20132C615 | List: |
| References | |
| FOIA-85-385, TASK-PII, TASK-SE SECY-84-476, NUDOCS 8501070245 | |
| Download: ML20132C610 (17) | |
Text
{{#Wiki_filter:- _ _ _ _ _ _ _ _ _ _ _ 1 E v /pa % l L i k POLICY ISSUE i December 20, 1984 (InfOrrnatIOn) SECY-84 476 For: The Commis',ioners ) t From: Williar J. Dircks Execut.se Director for Operations
Subject:
STAFF PLA'45 CONCERNING IMPLEMENTATION OF THE COMMISSION'S DIRECTIVES IN THE AREA 0F EQUIPMENT IMPORTANT TO SAFETY "i
Purpose:
To inform the Commission of the steps the staff is taking l to implement the Commission's directives in the area of i equipment important to safety.
Background:
In its Shorehaml decision the Commission stated: l "The material already in the record of this proceeding shows that the issue presented by Question 12 requires further consideration in a forum broad enough to encom-pass the far-reatning ramifications of any decision on this issue. As the Appeal Board found, the history of the use of the terms 'important to safety' and ' safety related' is tortuous and somewhat inconsistent. A comprehensive analysis of this history will be more accurate if it has the benefit of the institutional memories of as many individuals as possible. The application of such an analysis could result in a decision having :,ignificant consequences for the NRC's reculatory program. This potential for significant decirion warrants broad public participation. Accordingly, the Commission will initiate a rulemaking proceeding on this issue." 1 In the Matt 3r of Long Island Lighting Company (Shoreham Nuclear Power Station, =.- l Unit 1) CL1-84-9, 19 NRC 1323 (June 5, 1984). 3 Questior. 1: Are the terms "important to safety" and " safety-related" to be ~ deemed synonymous for the purpose of establishing an acceptable quality assurance program in accordance with GDC 1 of Appendix A and Appendix B to 10 CFR Part 507
Contact:
G. Ted Ankrum, IE e5010702"3 e41220 492-4774
- 5. M. Goldberg, IE e4,47; d (P 492-4968 hm j
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u The Comissioners Discussion: NRC has met with interested parties, including industry, regarding the important-to-safety subject and various approaches to resolve this issue. Based on these discussions, the staff concluded that the involved industry groups varied in their positions on how to best resolve the matter short of rulemaking. AdJitional discussions are planned next month between the staff and interested industry groups to inform the staff of industry's near-term plans. Pending the results of those further discussions, the staff is planning to go forward on a Notice of Proposed Rulemaking. to the Comission for its decision in early 1985. The Notice of Proposed Rulemaking is expected to propose specific guidance for equipment important to safety but not safety-related, Timilar to that published for public coment for non-safety-related I.TWS equipment (See Enclosure 1). The delay in sending the proposed rule to the Comission is to allow the staff to have further discussions with industry in January and to consider the public comments received on (A guidance for non-safety-related ATWS equipment. The public coment period for ATWS OA guidance closed on December 10, 1984 Rulemaking would be ir. tended to resolve two issues.
- First, significant differences exist between the industry and the staff concerning the meaning of a significant body of NRC regulations, regulatory guidance, and related correspondence in which the term "important to safety" is used.
The absence of a comon understanding cannot be allowed to continue if the NRC is to have a viable regulatory process and comuni-cate effectively. Enclosure 2 is a description of the different points of view and history of this issue.
- Second, overall piant safety should be improved by increasing the reliability of equipment important to safety but not safety-related.
This equipment has been observed as a frequent cause of problems such as plant trips and consequent challenges to safety systems. At the November 20, 1984 Comission meeting to review the AE00 program, the AE00 Director stated: "The dominant cause of the plant scrams seems to be coming from balance-of-plant (or non-safety-related) systems, primarily the feedwater system... Fifty percent of these scrams were above eighty percent power." i
The Commissioners In publishing specific guidance for equipment important to safety. but not safety-related, the staff's intent is to provide'a framework for improved plant operation by the following actions: (1) we would issue formal guidance in the form of a rule for this class of equipment, and (2) we would inspect by direct observation, to ensure that the -guidance is being properly implemented. This is clearly not an urgent safety issue, as evidenced by our case-by-case corrective action over the past fifteen years. However, the staff is convinced that plant operation and safety will be improved by adopting formal guidance for this entire class of equipment. Due to the lesser safety significance of this equipment, the guidance is expected to be less stringent than the requirements of 10 CFR Part 50 Appendix B for safety-related structures, systems, and components. The proposed QA guidance for non-safety-related ATWS equipment is indicative of the less stringent requirements currently envisioned. The Notice of Proposed Rulemaking is expected to contain guidance similar to that developed ft. the ATWS rule (after considering public connents) and also an enumeration of the basic issues intended to be addressed by the proposed rulemaking. The staff intends to hold a~ series of regional workshops on this topic shortly after publication of the proposed rule to further refine the staff's approach. Following an enhanced public coment process, a final rule will be submitted to the Commission for its decision toward the end of FY85. I will keep the Commission informed of our progress to resolve this issue. Vil ircks Executive Director for Operations
Enclosures:
l 1. Federal Register Notice QA Guidance for Non-Safety-Related ATWS Equipment 2. Description of Different Points of View and History of the Important-to-Safety Issue i l
[7590-01) NUCLEAR REGULATORY COMMISSION !ssuance for Comen,t of Quality Assurance (0A) Guidance Related To Anticipated Transients Without Scram (ATWS) Equipment That is Not Safety-Related AGENCY: Nuclear Regulatory Comission. ACTION: Request for Coment.
SUMMARY
The NRC staff is issuing a proposed generic letter to all interested parties, including licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This generic letter provides quality assurance (0A) guidance for non-safety-related equipment that is ~ associated with 10 CFR 50.62, " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants." The final 0A guidance is expected to be issued in february 1985 and is considered the reference date that initiates the schedule in 10 CFR 50.62 (d). DATES: The comment period expires December 10, 1984 Comments received after this date will be considered if it is practical to do so. ADDRESSES: Send coments to: Mr. Stephen M. Goldberg; Ouality Assurance Branch; Division of Quality Assurance, Safeguards, and Inspection Programs; Office of Inspection and Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. Page 1 of 11
FOR FUF.THER INFORMATION CONTACT: Stephen M. Goldberg; Quality Assurance Branch; Division of Quality Assurance, Safeguards, and Inspection Programs; Office of Inspectien and Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. Telephone: 301 492-4968. SUPPLEMENTARY INFORMATION: The following is the body of a p,,pcsed generic letter and an enclosure to the letter which is a table summart:ing the QA guidance for non-safety-related ATW5 equipment. This letter is to be addressed to all i i licensees of operating reactors, applicants for operating licenses, and holders i l of construction permits. This letter describes how this guidance was developed, its relationship to Appendix P to 10 CFR Part 50, and its use for meeting NRC QA requirements for non-safety-related ATWS equipment. LETTER On June 1, 1984, the Commission approved publication of a final rule, 10 CFR 1
- 50. 6 c' regarding the reduction of risk from anticipated transients without scram (ATWS) events for light-water cooled nuclear pcwer plants (49 FR 26036).
At the same time, the Commissicn directed the staff to complete and issue in the form of a generic letter explicit quality essurance (QA) guidance for non-safety-related equipment encompassed by the ATWS rule. l Section 50.6 'c! of the ATW5 rule requires that eacn licensee cevelop and submit (to the Director of the Office of Nuclear Reactor Regulation) a proposed schedule for meeting the requirements of the rule within 180 days after issuance of the OA guidance. i To develop QA guidance for ATWS ecuipment that is not safety related, the NRC l staff considered it necessary to survey QA practices applied to non-safety-related equipment at some operating nuclear power plants. During these olant visits. the NRC staff generally found that utility procedures were similarly Page 2 of 11
.._.c applied for safety-related and non-safety-related equipment or activities l (e.g., design modifications and procurement procedures). At some point in these utility procedures, CA practices for safety-related equipment or activities diverged from the practices employed for non-safety-re'ated equipment or activities. As e.n example, design mos'ificaticns to safety-related equipment l 1 required QA cre,anizational involvement and design verification; design I modifications to non-safety-related equipment required no design verification other than normal supervisory review and no QA organizational involvement. On j the basis of these plant visits, the NRC staff concluded that, as a general i matter, the quality practices now applied to non-safety-related equipment would be adequate for non-safety-related ecuipment encompassed by the ATW5 rule. Licensees, applicants, and the NRC staff have the desire to minimi:e the proli-feration of QA programs as opposed to the establishment of new and separate OA programs for non-saftty-related equipment. The practices that were observed during the plant visits were either to apply the 10 CFR Part 50, Appendix B program to ncn-safety-related ecuipment or to apply QA controls consistent with selected portions of their Appenc1x B program, although the utility procedures and practices did not specifically reference such controls as Appendix B requirements. Because of familiarity of the staff and industry with Appendix B requiremerts and because of demonstrated industry preference, the st.ff has chosen to develop explicit guidance for non-safety-related ATW! eculpment by fra-ing it in terms of Appendix B criteria. Accordingly, OA controls for non-safety-related Page 3 of 11
= ATWS equipment which meet Appendix B criteria except for those portions described below would fulfill NRC requirements: The QA organization is not required to participate ir developing and implementing QA practices for this equipment, provided that normal super-i visory controls exist to verify that the QA practices are being applied. { Audits are not required, provided that lire management periodically i reviews the adequacy of the QA practices as one of its internal control functions. A new and separate QA program is not required provided that the licensees or applicants are committed to establish QA controls for this equipment by utility policy statements, by procedures, instructions, or directives, or by other suitable means. In addition, new or separate programs or measures i are not required in the follcwing areas: - Inspection, provided that the line organization is responsible for f.etermining inspection requirements and for assuring that sufficient insr,ections are performed. - Test control, provided that the line organi:ation is respu 2 1ble for determining test requirements and for assuring tnat suf ficient testing is perf ormed. - Nonconformances and corrective action, provided that the line organization is responsible for controlling nonconformances, takes prompt action to correct conditions adverse to quality, and, as app;opriate, implements measures to preclude repetition. Page 4 of 11
1_ _ _ _ Ind:viduals outside the responsible design organization are not required to perform design verification (i.e., measures provided to verify or check the adequacy of the design by competent individuals or groups other than those who performed the original design), instead, a design review by i 3 the designer's supervisor would be adequate. l l 1 Contractors and subcontractors are not recu1 red to establish QA programs as a condition of the contract. The licensee's or applicant's QA controls for non-safety-related ATWS equipment beccme effective at the time the material or equipment is received at the plant. Contractors and subcontractors who perform services at the plant would be subject to the licensee's or applicant's QA controls for non-safety-related ATWS equipment. Documentation is not required to be available at the licensee's or applicant's i l plant to verify procedures were followed for the purposes of satisfying i internal control recuirements (i.e., documentation that verifies that receipt inspections were conducted need not be retained). However, documentation is required to verify that this e uipment is designed, installed, tested, operated, and maintair.ed so as to assure that the design specificatiers listed in the table published with the ATWS rule (49 FR 26036, pp. 260a2-26043) have been met. Enclosed with this letter is a summary of the QA guidance for non-safety-related ATWS equipment framed in the format of Appendix B to assist licensees and applicants. Page 5 of 11
i 1 l i in summary, the staf f concludes that either the rpplication of QA controls i based on the guidance in this letter or the application of Appendix B require-ments in their entirety is an acceptable method for satisfying NRC requirements. The staf f anticipates that licensees and applicants will select an approach which does not result in the establishment of a new and separate OA program for I i such equipment. l Issuance of this OA guidance shall be considered the reference date initiating the schedule in 10 CFR 50.62(d). The establishment of requirements under the ATWS rule was approved by the Office of Management and Budget under clearance number 3150-0111 which expires April 30, 1985 (49 FR 26036, p. 26044). Comments on burden and duplication may be directed ~ j to the Office of Management and Budget, Reports Management, Room 3208, New l j Executive Office Building, Washington, DC 20503. i I i i i i Page 6 of 11
ENCLOSURE i
SUMMARY
OF THE OA GUIDANCE FOR NON-SAFETY-RELATED ATW5 EQUIPMENT REQUIREMENT GUIDANCE f. Organization None (QA organization nct involved) II. Program None (no new or separate program requirec) III. Design Control Establish measures 1 to assure design specifications are included or correctly translated into design documents. Safety evaluations and reviews by designer's supervisor are required. !V. Procurement Document Establish measures to assure specifications Control and QA requirements are.ncluded in procurement documentation. II Except far design control, where the utility is responsible for ensuring that design control measures are applied at contractor or subcontractor organizations, the term " establish measures" applies to activities within the licensee's or applicant's organization, only. Also, the term " measures" is used synonymously with the term " controls" that appe?rs in the letter itself. Page 7 of 11
~ . REQUIREMENT GUIDANCE V. Instructions, Procedures Establish measures for documenting the ar.d Drawings controls appiied to activities that affect quality. VI. Document Control Establish measures to control issuance and changes to documents. VII. Control of Purchased Establish measures at plant to assure that items and Services all purchases conform to procurement documents. Stores or warehouse personnel or engineers may perform this verification. Vill. Identification and Control. Establish measures to identify and control of Purchased Items purchased items (i.e., traceability from receipt at the plant). !X. Control of Special Establish measures to control special Processes processes on the basis of codes, standards, and other requirements. X. Inspection Establish measures to inspect activities affecting quality. Verify ccnformance to documentation. Accomplish inspection by trainea personnel who did not perform the work. Page 8 of 11
REQUIREMENT GUIDANCE x!. Test Control Establish measures to test non-safety-related ATWS equipment prior to installation and operation and periodically. Document l and evaluate results. XII. Control of Measuring Establish measures to control, calibrate, and Test Equipment and adjust measuring and test equipment { at specisfic intervals. XIII. Handling, Storage, an( Establish measures to control handling, Shipping storage, shipping, cleaning, and preservation of purchases in accordance with utility documentation and manufacturer's recomendations. i ? XIV. Inspection, Test, and Establish measures to indicate status of Operating Status inspection, test and operability of installed non-safety-related ATWS equipment. XV. Nonconformances Establish measures to identify nonconformances. i XVI. Corrective Action Establish measures for prompt correction of j System conditions which are adverse to quality l l (i.e., nonconformances). Establish rneasures, if appropriate, to preclude repetition. Page 9 of 11
l ( REQUIREMENT Glt!0ANCE I i I i XVll. Records Establish measures to mairtain and control records which turnish evidence tnat system specifications described in the table of the ATWS rule have been met. I A XVIII. Audits None (audits not required if line management I reviews adequacy of CA controls). 4 t a l l l G =
m Dated at Bethesda, Maryland this 31st day of October 1984 1 FOR THE NUCLEAR REGULATORY COMMISSION / s e5 M. T lor, Deputy Director 0 fice of nspection and Enforcement i l i i l Page 11 of 11
j t DESCRIPTION OF DIFFERENT POINTS OF VIEW AND HISTORY OF THE IMPORTANT-TO-SAFETY ISSUE Different Points of View The Commission has had separate meetings on this subject with the NRC staff on May 11, 1984, and representatives of the nuclear industry 2 on July 31, 1984. The staff position, as presented at the May 11 meeting, has been that (1) the Appendix A to 10 CFR 50,2 term "important to safety" encompasses the broad sco and that (2) the term " safety-related" applies to a narrower subset of this equipment as defined in Appendix A to 10 CFR 100 (fII)(c) and in 10 CFR 50.49(b)(1)3 Although GDC-1 of Appendix A requires 1 that a QA program be developed for the broad scope of equipment "important to safety," no specific guidance regarding the characteristics of such a program has been provided. On the other hand, specific QA program requirements for the subset of safety related equipment are established by Appendix B to 10 CFR 50. The industry position as presented at the July 31 meeting, is that (1) the two terms are synonymou,s and limited to the set of eccipment tlready defined to be safety-related by the staf f; (2) the QA requirements for this of equipment (e.g., fire protection, radwaste, ATW5, and secur in adopted regulatory guidance; (3) HRC has existing regulatory authority to take appropriate action if a safety concern involving equipment which is not safety-related is apparent; and (4) Without further rulemaking, there is no regulatory basis for requiring licensecs to have QA controls for any type of equipment other than the narrow set called " safety related." The industry 2 The industry representatives were M. Edelman, representing the Atomic Inoustrial Forum (AIF); W. Counsil, representing the Utility Safety Classification Group (USCG); and R. Mcdonald, representing the Nuclear Utility Management and Human Resources Committee (NUMARC). 2 Equipment (i.e., systems, structures, and components) which provides reasonable assurance that the facility can be operated without undue risk to the health and safety of the public (10 CFR 50, Appendix A). 3 "This equipment is that relied upon to remain functional dur'.ng and t following design basis events to ensure: i (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; and j (3) the capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the ) 10 CFR Part 100 guidelines." t
',s. , to the Commission a decision on the correct use of terminology and the appropriate-QA requirements. Generic Letter 84-01 In October 1953, NRR and ELD discussed with the Committee to Review Generic Requirements (CRGR) various staff approaches to inform utilities about the staff's current definition and interpretation of the terms "important to safety" and " safety-related." Frcm that meeting and resultant EDO decision, NRR issued a generic letter (GL 84-01) on January 5,1984, to all power reactor permittees and licensees setting forth the staff's position on this issue (see j ). This letter described the staff's position concerning require-ments for the two classes of equipment. The letter stated that the staff viewed normal industry practice as generally acceptable for equipment considered important to safety but not covered by the Appendix 8 QA program. On April 6, 1984, IE issued a Federal Register Notice (49 FR 13775) requesting public and industry views on whether further guidance is needed on this issue. Comments received were along the lines of the position presented by industry representatives to the Commission in the July 31, 1984 meeting. NRC held a public meeting on April 30, 1984, with interested parties, including industry, to discuss the technical issues involved. Shoreham Case On June 5.,1984, the Commission stated they will initiate a rulemaking proceeding regarding the important to safety issue which had been certified to the Commission (on April 23, 1984) by the Atomic Safety and Licensing Board In the Matter of Long Island Lightino Company (Shoreham Nuclear Power Station, Unit 1), CLI-84-9, 19 NRC 1323 (June 5, 1984). In addition, the Ccmmission stated: "The Commission understands current precedent to hold that the term 'important to safety' applies to a larger class of equipment than the term ' safety-related. ' However, this does not mean that there is a pre-defined class of equipment at every plant whose functions have been determined by rule to be important to safety although the equipment is not safety related. Rather, whether any piece of equipment has a function important to safety is to be determined on the basis of a particularized showing of clearly identified safety concerns for the specific equipment, and the requirements of General Design Criterion 1 (GDC-1) must be tailored to the identified safety concerns." QA Guidance for Non-Safety-Related ATWS Equipment In conjunction with the approval for publication of the final ATWS rule on June 1, 1984, the Commission directed the staff to issue QA guidance in the form of a generic letter for non-safety related equipment that is associated with 10 CFR 5B.62, " Requirements for Reduction of Risk from Anticipated Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants." l
i ,e, 1 The staff had discussions with interested industry groups, met with the CRGR, and conducted visits to a limited number of plants to develop its approach on this subject. On November 6, 1984, the staff issued a proposed generic letter for comment in the Federal Register (49 FR 44337). Comments were due from all parties on this guidance by December 10, 1984. Attachments: 1. Denton Memerandum to Staff Concerning Safety Classification 2. Generic Letter 84-01 t 7 4 i i l i l l l
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i UNITED sT,:.Tas l /l'{ .,g i NUCLEAR REGULATORY COMMISSION l
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i t i TO ALL HOLDERS OF 00ERATING LICENSES, APPLICANTS FOR OPERAT!!iG LICENSES !.ND HOLDERS OF CONSTRUCTION PERMITS FOR POWER REACTORS Gentlemen:
Subject:
NRC Use cf the Terms, "Importan: to Safety" anc " Safe:y R211 red" (Generic Letter 84-01) As you muy knew, there has been cencern expressed recently by the Utility Classification Group over NRC.use of the terms "important to safe:y" :nd " safety-related." The concern appears to be principally derived fr;n recent licensing cases in which the meaning.of the terms in regard to.RC quality assurance recuirements has been at issue, and frcm a memorandun frca the Director, Office of Nuclear Reactor Regulatien, to NRR persennel ~ 6ted November 20, 1981. i Er.cic:cd fur ycur information are tvec letters to the NRC f rca this Gr u-i, a..d the L'.C 'res:onse dated December 19,19S3. In particular, you should note that the NRC reply makes it very clear that NRC regula cry,ierisciction invol.ing a safety matter is not controlled by the use of tems sucn as !' safe:y..related" and "important to safety," and our ccnclusien tha*. ;ur-suant to cur regulations, nuclear pcwer plant permittees c5 licensees are responsible for develeping and implementing quality assurance programs fer plant design and construction or for pi'nt operaticn which meet the... re a ceneral requirements of General Design Criterion 1 for plan: ecuipinen: "i..sortant to safety," and the rore crescriptive requirements of Appendix 3 to 10 CFR Par: 50 for " safety-related" piant equi; ment. Qile previcus staff licensing reviews were net scocifically ci ect:d tcwards deten.1ining whether,'in fact, permittees or licensees have develc;tt cur.ii y assurance prcgrams which adecuately addres.- all structures, syste.ms ar.d com- ,mnen:s i:;.per ant to safe'y, this was rtot because of a.:y cencern cycr 9e lad of regula cry recuirercents for this class of equipr.ent. Rather, cur l practice ;as bascd upon the s:aff view that normal industry practice is I generally acceptable for mest e;uipment not covered by Appendix 3 within this class. Never:heless, in specific situatiens in the past wtere we have found tnat cuality assurance rt *irements bayond no m.al incustry pr:c;ics l l were needed fer ocuipment "im;;r:3n: to safety," we have nct hesita.e>: :n imc0 sing acditional recuireme:.:s commensurate with the im;0rtance :e safety cf the equip:ent involved. k'e intend to centinue that prac-ice. GC DM W ) 6,l / W h W G , sg n < ,t n i L (' l0 1
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i. 1 i i i ! The NRC staff ir, interested in your comments and views on whether further
- guidance is needed related to this issue.
If you are interested in partici-l Dating in a meeting with NRC to discuss this subject, please contact ! Mr. James M. Taylor, Deputy Director, Office of.Inspecti:n and Enforcement. Sincerely, i \\ f F. &M4 l Jarrell G. Ebenhut, Director Division of L9 censing l
Enclosure:
l 1. Two Letters from Utility Safety l Classification Group 2. NRC Response dated December 19, 1983 i i l I l l i 1 4 4 l i l o
N.( A.ta!.ent 1 ~ ', '( f-G 6 8 November 20, 1981
- E 'C?.A N L'M FCP.: A'l f. R Personnel FROM:
Harold R. Denton, Director i Of. ice of Nuclear Reactor Regulation i SU?, JECT: STANDARD DEFINITIO::S FOR CO."MONLY-USED SAFETY CLASSIFI TERMS Litigatien of one of the principal issues in the THI-l Restart Hearing broust to light the fact that there is r.ot complete consistency among all cica.er..s of in the conduct of !;RR's safety review and licensing activit.as..he I;R More specifi-related" have been used at times interchangeably, cr in way consistent with the definitions and usage of such terms in the regulatior.s, and which do not fully reflect the intent of the regulation's or current licar. sing practice. Efforts have been undcruay kr some m:nths now to develop guidance for the consistent usage of these terms. These efforts have included: (a) review of c 1r.rge number of Reg Guides and SRP's, in conjunction with parts of the regula-tior.: uron which they are based, for consistency in the application of nicty cla..sification terminology, (2) extensive discussions among cognizant t3.2, t.Es (Stds. Devel.) ar.d ELD representatives regarding pmper interpretation t.nd tpplication of such terms, including consideratten of alternative "staadrrd" definitiens and (3) consultation with the cognizant ACRS Subco the.'s natters, and consideration by the full ACRS as well. rrittee regarding i As a result of these efforts, I am endersing and~ prescribing for use by all Ima perscnr.el the standard definitions set forth 'in the enclosure to this letter. It should be noted that in connection with icng-term efforts to devtlop r. cans for renhing reactor plant syster.3 with re;;c:t to ia ccm.a: tion with related efforts to develop a graded Q.A. approach in r Ectoregree of licensinc, the general question of safety classifications and safety classiMr. tion
- t..af nologics will be reexamined; and this could result in chansts to t".e defini-ti?ns set forth in the enclosure or perhaps in develop.at of a completely na schene in this regard.
For the time beir.g, however, the definitions in the ca-clos ra sh:uld be cor.sidered " standard" cud should bc applied consistently by all IER perseanel in all aspects of our saft:y review and Itcensing activities and 5 :.oul:1 be appropriately r.flected in our rc g nce docur.cnts. MM64 W [ n 5 NPP f9>a. k...18]J"3eumo: e v, n n ~ vi:hC L IC Ce~ V 'J V y n r^ y ^ y ^v ~ 3 m ..l. l 1 i ....p. o. .................l... -------a a
\\ er 4-g- All HRR Personnel It is expected that minor editorial revisions will have to be made to some existing Reg Guides and SRP's in order to make their wording consistert with these definitions. your purview in this regard and recommend the necessary change i expected that this will involve extensive revision efforts. clear that my interest here is only in establishing consistency in the language I want to make used by all cogni: ant groups within NRR in expressing our technical requirements. It is not my intention by this action to dictate new technical requirements, to modify existing technical requirements, or to broaden the existing scope of HRR licensing review. W 3
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LlLR.Dartas Harold R. Denton, Director Office of Nuclear Reactor Regulation Enciasure: Definition of Terms Distribution: i l F. Conran Conran Chron H. S. Hanauer M. L. Ernst Ernst Rdg A. Thadani F. Cofiman H. Denten E. Case i M B.:.D s,T3....R M B, ;,05 T,7f[.,R,R A B,: p,S T, g(,, M DST M A S T,,,,,,,,,,,H$Ma,,naWr [,,t3R t;R 'Cocc%w........ FCo f fman AThadani st h-tL a itt/ . HDen ...........................-{h..........................I-s N .1 N /81 11//$/81 II/ 8I II/h/0I II/[Y[I. RdUULL
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(. B DEFINITION OF TEPJ4,5_ imoortant to Safety i i e Definition - From 10 CFR 50, Appendix A (General Design Criteria) - see first paragraph of " Introduction." i "Those structures, systems, and components that provide reasonable assurance i that the facility can be operated without undue risk to the health and safety of the public~." W Encompasses the broad class of plant features, covered (not necessarily e explicitly) in the General Design Criteria, that contribute in important way to safe operation and protection of the pi.blic in all phases and aspects ~ of facility operation-(i.e., normal pperation and IFansient control as well as accident mitigation). Includes safety-Grade (or Safety-Related) as a subset. e Sa fe ty-Rela ted Definition - From 10 CFR 100, Appendix A - see sections III.(c), VI.a.(l), and a VI.b.(3). These structure, systems, or components designed to remain functional for the SSE (also termed ' safety features') necessary to assure recuired safety functions, i.e. :' (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which c:uld result in potential off-site exposures comparable to the guideline exposucas of this part. e Subset of "Important to Safety" e Regulatory Guide 1.29 providesan LWR-ceneric, function-oriented listing of " safety-related" structures, systems, anc components neeceo to provide or perform ~ required safety functions. Additional information (e.g., NSSS type, j BOP design A-E, etc.) is needed to generate the com'plete listing of safety-I related SSC's for any specific facility. Note: The term " safety-related" also appear:; in 10 CFR 50, Appendix B (Q.A. Program Requirements); however, in that context it is framed in somewhat different language' than its definition in 10 CFR 100, Appendix A. That difference in language between the two appendices has contributed to, confusion and misunderstanding regarding the exact meaning of " safety-related" and its relationship to "important to safety" and " safety-grade." A revision to the language of Appendix B has been proposed'to clarify this situation and remove any arbiquity in the meaning of these terms. Enclosure
i 1 s, .c, 2 Sa fe ty-Gra de e Term not used explicitly in regulations but widely used/ applied by staff and indus.try in safety review process.
- Equivalent to " Safety-Relate 6" 1.e., both terms apply to the same subset of the broad class "Important to Safety."
f l I I O e l 4 l e 9 9 e 1 l e [ l 9 d me
n> .f \\ k 2 Sa fe ty-Gra de e Term not used explicitly in regulations but_widely used/ applied by staff and industry in safety review process
- Equivalent to " Safety-RelateA" i.e., both terms apply to the same subset of the broad class "Important to Safety."
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