ML23159A118
| ML23159A118 | |
| Person / Time | |
|---|---|
| Issue date: | 10/17/1997 |
| From: | Hoyle J NRC/SECY |
| To: | |
| References | |
| PR-050, 62FR53932 | |
| Download: ML23159A118 (1) | |
Text
{{#Wiki_filter:ADAMS Template: SECY-067 DOCUMENT DATE: 10/17/1997 TITLE: PR-050 - 62FR53932 - CODES AND STANDARDS; IEEE NATIONAL CONSENSUS STANDARD CASE
REFERENCE:
PR-050 62FR53932 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete
DOCKET NO. PR-050 (62FR53932) In the Matter of CODES AND STANDARDS; IEEE NATIONAL CONSENSUS STANDARD DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 10/09/97 10/09/97 DIRECT FINAL RULE - 11/24/97 11/10/97 COMMENT OF MDM/LAMB., INC. (EDWARD (TED) L. QUINN, V.P.) ( 1) 11/26/97 11/25/97 COMMENT OF IES UTILITIES, INC. (KENNETH E. PEVELER) ( 2) 11/26/97 11/24/97 COMMENT OF FAROUK D. BAXTER ( 3) 11/26/97 11/26/97 COMMENT OF NUCLEAR ENERGY INSTITUTE (ALEXANDER MARION) ( 4) 11/28/97 11/25/97 COMMENT OF TENNESSEE VALLEY AUTHORITY (MARK J. BURZYNSKI) ( 5) 11/28/97 11/21/97 COMMENT OF WESTINGHOUSE ELECTRIC CORPORATION (S. A. SEPP) ( 6) 11/28/97 11/25/97 COMMENT OF ENTERGY OPERATIONS, INC. (JERROLD G. DEWEASE, VICE PRESIDENT) ( 7) 12/01/97 12/01/97 COMMENT OF JOHN L. KNOX ( 8) 12/01/97 11/25/97 COMMENT OF PACIFIC GAS AND ELECTRIC COMPANY (GREGORY M. RUEGER, SR. V.P.) ( 9) 12/01/97 11/26/97 COMMENT OF DUKE ENERGY CORPORATION (M. S. TUCKMAN, EXEC. V.P.) ( 10) 12/01/97 11/29/97 COMMENT OF IEEE NUCLEAR POWER ENGINEERING COMMITTEE (WESLEY W. BOWERS, CHAIR IEEE NPEC) ( 11) 12/01/97 11/26/97 COMMENT OF SOUTH CAROLINA ELECTRIC & GAS COMPANY (DAVE A. LAVIGNE) ( 12) 12/01/97 11/26/97 COMMENT OF SOUTHERN CALIFORNIA EDISON (E. S. MEDLING) ( 13)
DOCKET NO. PR-050 (62FR53932) DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 12/01/97 11/26/97 COMMENT OF ELECTRIC POWER RESEARCH INSTITUTE (RAYMOND C. TOROK) ( 14) 12/01/97 11/27/97 COMMENT OF GARY JOHNSON ( 15) 12/01/97 12/01/97 COMMENT OF NUCLEAR UTILITY BACKFITTING AND REFORM GROUP (DANIEL F. STENGER, ESQ.) ( 16) 12/01/97 12/01/97 COMMENT OF NUCLEAR UTILITY GROUP ON EQUIPMENT QUALIFICATION (WILLIAM A. HORIN, ESQ., ET AL.) ( 17) 12/02/97 11/24/97 COMMENT OF ROCHESTER GAS AND ELECTRIC COMPANY (ROBERT C. MECREDY, VICE PRESIDENT) ( 18) - 12/02/97 11/26/97 COMMENT OF PECO ENERGY COMPANY (G. A. HUNGER, JR.) ( 19) 12/02/97 12/01/97 COMMENT OF SOUTHERN NUCLEAR OPERATING COMPANY (D. N. MOREY) ( 20) 12/03/97 12/01/97 COMMENT OF STRATEGIC TECHNOLOGY AND RESOURCES, INC. (PHILIP M. HOLZMAN) ( 21) 12/03/97 12/01/97 COMMENT OF CAROLINA POWER & LIGHT COMPANY (D. B. ALEXANDER) ( 22) 12/04/97 12/01/97 COMMENT OF APS (JAMES M. LEVINE, SR. VICE PRESIDENT) ( 23) 12/04/97 11/28/97 COMMENT OF THE SKS GROUP, INC. (DALE R. WUOKKO, PE) ( 24) - 12/08/97 12/01/97 COMMENT OF NEW YORK POWER AUTHORITY (JAMES KNUBEL, SR. VICE PRESIDENT) ( 25) 12/08/97 12/01/97 COMMENT OF COMMONWEALTH EDISON COMPANY (JOHN B. HOSMER, VICE PRESIDENT) ( 26) 12/08/97 12/04/97 COMMENT OF TOLEDO EDISON (R. E. DONNELLON) ( 27) 12/08/97 12/01/97 COMMENT OF FLORIDA POWER & LIGHT COMPANY (HARRY N. PADUANO) ( 28) 12/15/97 12/10/97 COMMENT OF VIRGINIA POWER (MICHAEL R. KANSLER, VICE PRESIDENT) ( 29) 12/23/97 12/16/97 DIRECT FINAL RULE: WITHDRAWAL 01/06/98 01/06/98 DIRECT FINAL RULE: CORRECTION 01/23/98 01/16/98 COMMENT OF NUCLEAR UTILITY BACKFITING AND REFORM GROUP (JAMES R. FITZGERALD, ESQ., ET AL.) ( 30)
' 35 WEST WACKER DRIVE CHICAGO. ILLINOIS 60601-9703
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1400 L STREET. N.W WASHINGTON. O.C. 20005-3502 DOCKETED USNRC 6 . RUE OU CIRQU E 75008 PARIS . FRANCE 200 PARK AVENUE NEW YORK. NY 10166-4193 (202) 371-5700 '98 JAN 23 Al 1 ~~ laiiE DU RHONE 1204 clt~EVA. SWITZERLAND FACSIMILE (202) 371 -5950 January 16, 1998 Mr. John C. Hoyle, Acting Secretary So Office of the Secretary ( '7.:l F~ 53 '175) U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ( ~ :{ Fa 53 "J 1.2) Re: Codes and Standards: IEEE National Consensus Standard (62 Fed. Reg. 53,932 (October 17, 1997))
Dear Mr. Hoyle:
On December 1, 1997, we submitted comments in the captioned rulemaking proceeding on behalf of the Nuclear Utility Backfitting and Reform Group (NUBARG). Several members ofNUBARG own and operate plants having construction permits issued prior to January 1, 1971. This letter is being submitted on behalf ofNUBARG as supplemental comments to confirm what we believe to be the NRC's position in the proposed rule, i.e., that nuclear power plants with construction permits issued prior to January 1, 1971, would not be required to comply with the provisions of IEEE Standards 279 and 603-1991 for future modifications.!! While we believe that the proposed rulemaking never intended to add additional requirements for plants whose construction permits were issued prior to January 1, 1971, the rulemaking (62 Fed. Reg. 53,932, October 17, 1997) to incorporate~ reterence IEEE Std. 603-1991 eliminated the explicit provision in 10 C.F.R § 50.55a(h) regarding nJ8f~ar plants with construction permits issued prior to January 1, 1971. The current version of 10 C.F.R. § 50.55a(h) explicitly excludes such plants from complying with the requirements of IEEE-279, "Criteria for Protection Systems for Nuclear Power Generating Stations." In that respect, we understand that comments made by the NRC representative at the November 1997 IEEE Nuclear Power Engineering Committee Meeting in Albuquerque, New Mexico, indicated that the NRC would encourage, but not require, such older plants to comply with IEEE Std. 603-1991 for future major ( system-level) modifications. 1' In view of the withdrawal of the direct final rule (62 Fed. Reg. 66,977, December 23, 1997), it should be practicable for the NRC to consider these brief comments as part of its ongoing review of the proposed rule. Ackno ed
U.S NUCLEAR REGULATORY CC RULEMAKINGS & ADJUDICATI S OFACEOFTHESECRETA OF THE COMMISSI Postmark Data C R8ClillVOO I ~J9 8' I A 3 _J>D1'-,-1?.~ - - I
\VT~STO~ & STR....\.\~ The statement of considerations accompanying the final rule should clearly state that the new 10 CF .R. § 50.55a(h)(3) applies only to plants with construction permits issued after January 1, 1971 and clarify that the second sentence of proposed§ 50.55a(h)(3) applies to plants covered by the first sentence of that subsection.
DOCKETED 7590-~1~ RC Nuclear Regulatory Commission
*95 JAN -6 Al 1 :49 10 CFR Part 50 OFFICI- oc SFCH . ,-iv RIN 3150-AF73 Rljll Iv\/.** \ \,'.ir
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t .j ,_l ADJUUIC;.\ i""i-. *~~- 'Tt FF Codes and Standards; IEEE National Consensus Standard , Withdrawal ; Correction AGENCY: Nuclear Regulatory Commission. OCKET PA ED RULE So ( (,'J.t=/<.t5°3'17S) ACTION: Direct final rule; correction. ( (,~Ffl.53q3:z.)
SUMMARY
- This document corrects a notice appearing in the Federal Register on December 23, 1997 (62 FR 66977) . This action is necessary to correct an erroneous Federal Register citation.
FOR FURTHER INFORMATION CONTACT: Michael T. Lesar, Acting Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, Washington, D.C. 20555-0001, telephone (301) 415-7163. - SUPPLEMENTARY INFORMATION: On page 66977, in the first column, in the last paragraph, in the second line, Federal Register citation "(62 FR 53933)" is corrected to read "(62 FR 53932)". Dated at Rockville, Maryland, this 6th day of January 1998. For the Nuclear Regulatory Commission.
~~/-:4 Michael T. Lesar, Acting Chief Rules and Directives Branch Division of Administrative Services Office of Administration
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[7590-01-PJ DOCKET ED USNRC NUCLEAR REGULATORY COMMISSION "97 DEC 23 Al 1 :54 10 CFR Part 50 RIN 3150 - AF73 OFFICE. or RUL,:tA: f C'
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fa ,.:L I ADJLJDlCl,i IL !N*: r TAFF Codes and Standards; IEEE National Consensus Standard, Withdrawal * ' *- '- AGENCY: Nuclear Regulatory Commission. DOCKET N BER PROPOSED RULE........_ __ ( lP'-F-"A.S3'115) ACTION: Direct final rule: Withdrawal. ( Co 2 f"ll5'3"13:2.)
SUMMARY
- The Nuclear Regulatory Commission is withdrawing a direct final rule that would have amended Commission's regulations to incorporate by reference the most recent published version of IEEE Std. 603-1991, a national consensus standard for power, instrumentation, and control portions of safety systems in nuclear power plants.
The NRC is taking this action because it has received significant adverse comments in response to an identical proposed rule which was concurrently published in the Federal Register. FOR FURTHER INFORMATION CONTACT: Satish K. Aggarwal, Senior Program I Manager, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, I I Washington, DC 20555. Telephone (301) 415-6005, Fax (301) 415-5074 (e-mail: i I SKA@NRC.GOV). I I I SUPPLEMENTARY INFORMATION: I I I On October 17, 1997 (62 FR 53933), the Nuclear Regulatory Commission I I published in the Federal Register a direct final rule amending its regulations at I I 10 CFR50.55a (h) to incorporate by reference the most recently published version of a national I I I P.uh. ,n-, 1:i/:J1/,7 I a:r (p:, Flt. r,t,q 17 I
2 consensus standard. The direct final rule was to become effective on January 1, 1998. The NRC also concurrently published an identical proposed rule on October 17, 1997 (62 FR 53975). In these documents, the NRC indicated that if it received significant adverse comments in response to this action, the NRC would withdraw the direct final rule and would consider the comments received as in response to the proposed rule and address these comments in a subsequent final rule. The NRC has received significant adverse comments on the direct final rule. Therefore, the Commission is withdrawing the October 17, 1997, direct final rule. The public comments received will be addressed in a subsequent final rule issued in either a notice of final rulemaking or in a notice of withdrawal of the proposed rule. I ti-, Dated at Rockville, Maryland, this/ i - day of December, 1997. For the Nuclear Regulatory Commission. l____ ohn C. Hoyt , Secretary of the Commission.
MICHAEL R. KANSLER Innsbrook Technical Center Vice President - Nuclear Operations 5000 Dominion Boulevard Glen Allen, Virginia 23060 DOCKETED 804*273*3586 USNRC
'97 DEC 15 P12 :32 December 10, 1997 VIRGINIA POWER Secretary Serial No. GL97-114 Attention: Rulemakings and Adjudications Staff Office of Administration DOCKET N,_..._,._.
U. S. Nuclear Regulatory Commission PROPOSED RULE 50 Washington, D. C. 20555-0001 ( C,:l Fl< 53ct 7S) ( '1 ~F~ S3"!3:L) Gentlemen: COMMENTS ON THE NUCLEAR REGULATORY COMMISSION'S PROPOSED RULE CHANGE TO 10CFR50.55a CONCERNING STANDARD IEEE 603, "CRITERIA FOR SAFETY SYSTEMS FOR NUCLEAR GENERATING STATIONS This letter provides comments in response to the Proposed Rule and Direct Final Rule which revises 10CFR50.55a to invoke the 1991 revision of IEEE Standard 603, as published in the Federal Register on October 17, 1997 on pages 53932 to 53935 and pages 53975 to 53976. Virginia Power is appreciative of the opportunity to provide comments on the proposed changes and commends the NRC for involving the industry prior to implementing regulatory modifications. The following comments are offered in regard to the proposed rule change:
- 1. The Direct Final Rule, on page 53933, states that "The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same.
Therefore, for purposes of paragraph (h) of 10CFR50.55a, "protection systems," and "safety systems" are synonymous." However, IEEE 279 clearly states that the "protection system" encompasses all electric and mechanical devices and circuitry (from sensors to actuation device input terminal) involved in generating those signals associated with the protective function. This is clearly the "sense and command" portion of a system. IEEE 603 recognizes the "sense and command" elements of a safety system, but also defines the "actuate" and "power supply" features which make the complete "safety system." Close inspection of the scope of IEEE 603 shows that the protection systems identified in IEEE 279 are a subset of the safety system. Therefore, the "protection systems" and "safety systems" are not synonymous, and the use of IEEE 603 represents a large increase in scope over IEEE 279. OEC 1 9 1997 Acknowledged by card .. 1111 I II --
U.S. UClEAR REGULATORY OOMMISSI RULEMAKIN8S &ADJUDICATIONS STAFF OFFICE OF TtiE SECRETARY OF THE COMMISSION DocumnStatlatk:s Postmark Date lj lj
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Secretary, NRC 12/10/97 Page 2. It should be noted that the replacement of IEEE 279 with IEEE 603 is llil1 a "like for like" replacement. When IEEE 279 was written, it contained all the guidance available to design protection systems for nuclear power generating stations. However, since the withdrawal of IEEE 279 and the development ofIEEE 603, much of the information associated with the detailed design of a safety system has been placed in companion standards which are referenced in IEEE 603 (i.e., IEEE 379 for single failure criterion and IEEE 384 for physical independence criteria, as examples). The Direct Final Rule mandates the use of IEEE 603, but does not address the mandatory use of these companion standards even though they are referenced in IEEE 603. The Direct Final Rule states that if these referenced standards are endorsed by a regulatory guide, they represent methods acceptable to the commission. If not, then the information in the referenced standard may be used consistent with current regulatory practice. The Direct Final Rule should be clarified to define if the companion standards for IEEE 603 are mandatory requirements and how the companion standards are to be used.
- 3. The Direct Final Rule allows current licensees to continue to meet the requirements set forth in IEEE 279 in effect on the date of their application for a construction permit. However, the Rule requires that "changes to protection systems in operating nuclear power plants which are initiated on or after January 1, 1998 must meet the requirements in IEEE 603-1991." The Rule defines what "changes" are in general terms and states that in-kind (like-for-like) replacements of protection system components are not considered changes to the protection system. The definition of an in-kind (like-for-like) replacement component is subjective and should not be limited to an original manufacturer's exact replacement. Because original manufacturers are leaving the nuclear parts business, utilities must be allowed to use replacement parts which are supplied by other manufacturers and which meet the original design requirements. Engineering verification and dedication of in-kind (like-for-like) substitute replacement equivalents must be allowed (i.e., form-fit-function equivalents).
The issue of in-kind (like-for-like) replacement is not clearly addressed in the Backfit Analysis. The backfit analysis does not recognize that the replacement of a component with another component that may not be an exact duplicate (like-for-like equivalent) is different from "changes which are voluntarily initiated by the licensee, or separately imposed by the NRC after a separate backfit analysis." Virginia Power agrees that the Direct Final Rule should mandate IEEE 603-1991 for future nuclear plants, and that current licensees should be allowed to continue to meet the requirements set forth in the edition of IEEE 279 in effect on the formal date of their application for a construction permit. In addition, Virginia Power proposes that both
Secretary, NRC 12/10/97 Page whole system modifications and in-kind (like-for-like) component replacements should be per either the existing plant's licensing basis or IEEE 603 at the plant's option. We appreciate the opportunity to provide comments on the Proposed Rule Change to 10CFR50.55a. Should you have any additional questions, please feel free to contact me. Sincerely, cc: Mr. Satish K. Aggarwal, Sr. Program Manager Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Florida Power & Ligt J ~ l { o >
- Box 14000, Juno Beach, FL 33408-0420 USNRC December 1, 1997 "97 OEC -8 P1 :55 L-97-304 OFFICE OF SEC-:'E1ARY Mr. John C. Hoyle, Secretary RUU: M:\J<lf*!,.,S .i\l~O U.S. Nuclear Regulatory Commission ADJUDICATIOl* S STAFF Washington, DC 20555-0001 DOCKET BERDI 50 pROP()SEDRlJLE,..L[!J.!J,...;;..;;...-
Attn: Rulemakings and Adjudications Staff ( /p~Fte53976) ( I, 'J. Fte 5313 :J.)
SUBJECT:
Codes and Standards, IEEE 603-1991 Direct Final Rule 62 FR 53932, October 17, 1997 Florida Power & Light Company (FPL) endorses the comments on the subject rule that were submitted by the Nuclear Energy Institute (NEI) in a letter dated November 26, 1997 (A Marion to J. Hoyle). The pending requirement that IEEE 603-1991 become part of the licensing basis for other than "like-for-like" modifications after January 1, 1998, is of particular concern to FPL. The scope of IEEE 603 (" safety systems") is much broader than the scope of IEEE 279 ( protection systems"). Since IEEE 279 is typically the licensing basis for operating nuclear plants, the adoption of IEEE 603 could have a significant impact on licensee design modifications. A reasonable notice and comment period is warranted to permit a thorough evaluation of the proposed change. We also disagree with the conclusion that the direct final rule is non-controversial simply because there was no adverse reaction to DG-1042 (proposed Revision 1 to Regulatory Guide 1.153), which endorsed IEEE 603 as one option for complying with NRC regulations. In general, licensees are neither obligated to comment on draft regulatory guides nor obligated to incorporate revised regulatory guides into the licensing basis. On the other hand, to make IEEE 603 a regulatory requirement is a significant action that would draw significant industry comment. FPL recommends the direct final rule be replaced by a proposed rulemaking to allow comment by licensees and resolution of the comments by the NRC staff Very truly yours, ~~d~?n~i . Licensing and Special Programs HNP:dcr an FPL Group company OEC ' 9 \997 Acknowledged by card ........., 11111 SC
U.S. NUCLEAR REGUI.ATORY COMMlSSIOK RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OFTffE SECRETARY OF THE COMMISSION DoQlnel'4Statilllcl Postmark Dale I r;@/9'1 Coples Recalvad----'------ 3__
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- e ED1"soN A Centerior Energy Company DOCKETED US RC Davis-Besse Nuclear Power Station 5501 North State Route 2 "97 OEC -a A7 :sa Oak Harbor, OH 43449-9760 Serial Number 2503 December 4, 1997 DOCKET NlJABER PROPOSED RULE p 50 Secretary (r.,a i:~ 5311s) United States Nuclear Regulatory Commission ( IP~F(<53'F~;;.) Document Control Desk Washington, D. C. 20555-0001 Attn: Rulemakings and Adjudications Staff
Subject:
Comments Regarding the NRC Direct Final Rule to Incorporate by Reference IEEE Standard 603-1991 in 10 CFR 50.55a, "Codes and Standards" Ladies and Gentlemen: Toledo Edison, a subsidiary of FirstEnergy, is partial owner of and is responsible for operation of the Davis-Besse Nuclear Power Station (DBNPS). As a 10 CFR Part 50 licensee, Toledo Edison has a vested interest in any rules the U.S. Nuclear Regulatory Commission (NRC) may adopt which can affect the management and operation of a commercial nuclear power plant. Toledo Edison is responsible for not only ensuring that the DBNPS is operated safely, but for also ensuring the DBNPS is operated cost-effectively for its customers. Industry Comments Toledo Edison has reviewed both the NRC direct final rule to incorporate by reference IEEE Standard 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," in 10 CFR 50.55a, "Codes and Standards", as noticed in the Federal Register on October 17, 1997 (62FR53932) and the comments submitted on November 26, 1997 to the NRC on behalf of the nuclear power industry by the Nuclear Energy Institute (NEI). Based on this review, Toledo Edison endorses the comments provided by NEI in response to the 62FR5332 notice. Additional Toledo Edison Comments In addition to the NEI comments, Toledo Edison provides the following comments. With respect to the NRC's claim that the backfit rule, 10 CFR 50.109, does not apply because the rule change does not impose any backfits, the rule would require licensees to adopt OEC 1 9 \997 Acla1owledged 'by' cartl .................nllvbl I ,. II
U.S. NUCLEAR REGULATORY COMMISS!ON RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY Of THE COMMISSION Oocllnent Statistics Poslmm1< oate 1I
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Serial 2503 Page 2 IEEE Std. 603-1991 over IEEE Std. 279 for changes to protection systems. The NRC's basis for this claim is that changes made to the protection systems would be voluntary and at the licensee's option. This would not be the case when components fail and are obsolete. In this situation, the licensee has no alternative but to find a suitable replacement component for the failed or obsolete component. It is not good engineering practice and would be very difficult to maintain configuration control in attempting to apply different IEEE standards within the same protection system. In mandating that the new component satisfy IEEE Std. 603-1991, the proposed rule essentially requires the licensee to upgrade and/or modify the remainder of that protection system train to satisfy IEEE Std. 603-1991. Defining "changes to protection systems" for purposes of when IEEE Std. 603-1991 is to be applied should be defined as including modifications, augmentation or replacement of complete protection systems channels. The term "channel" is defined in both IEEE standards (IEEE 279 and 603), as well as the Improved Standardized Technical Specifications. Toledo Edison believes that the IEEE standards were written with the intent that they would apply to an entire system, not on a component-by-component basis. In conclusion, Toledo Edison believes this rule will have a significant impact on licensees with little added safety value and will not result in cost-justified safety improvements. Should you have any questions or require additional information, please contact Mr. James L. Freels, Manager - Regulatory Affairs, at (419) 321-8466. Very truly yours, ~ s_~{f}- R. E. Donnellon Director - Engineering & Services cc: A. B. Beach, Regional Administrator, NRC Region III S. J. Campbell, DB-1 NRC Senior Resident Inspector A. G. Hansen, DB-1 NRC/NRR Project Manager Utility Radiological Safety Board
Commom,T*llt h Fd i,on Cll mp.1m 1-~00 ( l pu, !'lace Downer, c ;ron' . IL (,O "i 1:;.:;-0 I DOCK -TED USNR December 1, 1997
'97 DEC -8 A7 :59 ComEd OFrr ll.;l"- 1~-,F 1 r* ( ,!
*[ \I Secretary RUL;:: J\.1/., :If f ._, . " L) DOCKET U.S. Nuclear Regulatory Commissio~ JUD!CJ1*1 1C,r\ 1', ~) 1/i..FF PROPOSED RULE 50 Washington, DC 20555-0001 (c, ~ F~ 53"15}
Attention: Rulemakings and Adjudications Staff ( r,i FflS3'13:i..)
Subject:
Proposed Rule Change to 10CFR50.55a Concerning IEEE Standard 603, "Criteria for Safety Systems for Nuclear Power Generating Stations"
Reference:
Federal Register Vol 62 No. 53932 This letter provides the Commonwealth Edison Company (ComEd)'s comments on the subject Nuclear Regulatory Commission (NRC) proposed rulemaking. As stated in the Federal Register Notice, the Nuclear Regulatory Commission (NRC) considers this rulemaking, which endorses IEEE Std. 603-1991, to be non-controversial because there was no adverse public comments on Regulatory Guide 1.153, Rev. 1, which endorses this standard. ComEd believes that the lack of public comment was due to the fact that the Regulatory Guide was considered to apply to new plant construction, not existing plant modifications. When the NRC withdrew the IEEE Std . 279, and issued IEEE Std.603-1991 as a replacement standard, the committee responsible for the withdrawn standard expected IEEE Std. 279 to be used on modifications to existing design. Use of IEEE Std. 603-1991 on existing designs would result in substantial rework of existing systems without a corresponding improvement in their design. IEEE Std. 603-1991 should be used on all new designs such as for any new systems or for major additions to existing systems. The Proposed Rule requires any changes made pursuant to 10CFR50.59 initiated on or after January 1, 1998, and plant-specific departures from a design certification rule under 10 CFR Part 52, must meet the requirements IEEE Std 603-1991 . In-kind (like-for-like) replacement of protection system components are not considered changes to the protection systems. Under the Proposed Rule, unnecessary work would result. For example, a minor change such as a mounting change resulting from replacing a failed or obsolete component would subject the component's system to meet the requirements of IEEE Std. 603-1991 since such a change is required to be made under 50.59. The component's system on which the modification is being made would require substantial additional rework. This rework could include the separation of circuits, installation of additional conduit or cable tray to meet the independence and associated circuit requirements. This unnecessary work would be required without a corresponding improvement in safety. Modifying a component under IEEE Std. 603-1991 would result in substantial cost as a result of having to modify the full system because of a single component. c , 9 \997 AcknoWfedged by cmd ..- - - -
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U.S. Hua.EAR REQUlA10RYCXIIMSSION RULEMAKINGS &ADJUDlr,ATIONS STAff OFFICE OF TIE SECRE1'ARY Of THE COtMSSION Oocllllai . . . . PostmaJtcoate Coples RecQlved _ __ 1~f~/q1_ _ , Add'I Cq>les Reproduced __._ _ _ _ Special Dlstrlbution 14 ;.,AV:4J 4.~ _ "v't>R _ ......:lJe!!:..,__ _ _ __ 1
Secretary U.S. NRC December 1, 1997 As stated in the Federal Register, "The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same. Therefore, for purposes of paragraph (h) of 10 CFR 50.55a, "protection systems," and "safety systems" are synonymous." This statement is inconsistent with the "Forward" of IEEE Std. 603-1977 (the trial use issue of the standard) which clearly states that IEEE Std. 603-1977 is an expanded scope verses IEEE Std. 279-1971. The section of the IEEE Std. 603-1977 forward titled "Evolution" clearly explains the expanded scope of 603-1977 verses 279. This discrepancy needs to be clarified. The Federal Register Notice also states, "Section 3 of IEEE Std. 603-1991 references several industry codes and standards. If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices." This section may be interpreted as mandating implementation of these additional standards by reference. Implementation of additional codes and standards should be accomplished by separate rulemaking. The following are additional items that need clarification by the NRC and should be addressed in the Federal Register Discussion:
- 1. Define the legal interpretation of "acceptable reliability" (Ref. IEEE Std. 603-1991 Section 8.3).
- 2. The Rule states, "... changes to protection systems initiated on or after January 1, 1998, must meet the requirements set forth in IEEE Std. 603-1991, and the correction sheet dated January 30, 1995." The Staff needs to define "initiated" (i.e., Are plant changes that have been designed and approved prior to January 1, 1998, subject to the requirements?).
Finally, ComEd recommends that the NRC consider incorporation of the latest revision of IEEE Std. 603 (which is currently in the final balloting process) in lieu of IEEE Std. 603-1991. ComEd believes this revision resolves many issues addressed in this correspondence. Sincerely, ~t:f!=-, Vice President cc: G. Dick, Generic Issues Project Manager - NRR A.B. Beach, Regional Administrator- RIii Office of Nuclear Safety - IONS
123 Main Street White Plains, New York 10601 DOCKE TED 914 681.6840 US RC 914 287.3309 (FAX)
"97 OEC -8 A7 :57 1997 0[C - tJ PM 3: 20 James Knubel
, . NewVorkPower . , Authority RULEs
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- 11'-IG'- .i' D US NRC ADJUD!Ct*. rJ;-; !* J,. c.* '"/,FF DOCKET NtlilBER PROPOSED RULE PR 5 o (If~ l='R 53~7-') December 1, 1997
( (I) .:2 P ~ S-3 '13 .l) JPN-97-037 IPN-97-164 Chief, Rules and Directives Branch Division of Administrative Services Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Indian Point 3 Nuclear Power Plant Docket No. 50-286 James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 COMMENTS ON FINAL DIRECT RULE CHANGES TO PARAGRAPH (h) OF 10 CFR 50.55a "CODES AND STANDARDS" REFERENCES : 1. October 17, 1997 FEDERAL REGISTER , Vol. 62, No. 20, pages 53932-53935, "Nuclear Regulatory Commission Final Direct Rule, Codes and Standards; IEEE National Consensus Standard"
- 2. IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Generating Stations"
- 3. IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations"
Dear Sir:
The Authority has reviewed the direct final rule (Reference 1) published October 17, 1997 amending 10 CFR 50.55a(h). This new rule incorporates a reference to IEEE Std. 603-1991 (Reference 2) to replace IEEE Std. 279 (Reference 3) which has been withdrawn by the IEEE. The Authority has several concerns regarding this new rule, which are detailed below. Imposes New Requirements on Existing Operating Plants The current version of 10 CFR 50. 55a(h) includes a provision that excludes plants with construction permits issued prior to January 1, 1971 . The final direct rule includes no such provision. The absence of this provision would have a significant effect on the Authority's Acknowledged by card . D~, .!..~?.!,.:._
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ability to change protection systems at the Authority's nuclear plants. Both Indian Point 3 and James A. FitzPatrick received their construction permit before January 1, 1971, and consequently, are exempt from the requirements of 10 CFR 50.55a(h). While both plants have committed to portions of IEEE Std. 279 (1968 or 1971 ), this change to 10 CFR 50.55a(h) would impose IEEE Std. 603-1991 as a new requirement on both plants. The imposition of IEEE Std. 603-1991 as a regulatory requirement may meet the definition of a backfit in 10 CFR 50.109(a)(1) and requires the preparation of a backfit analysis. Broadened Scope of IEEE Std. 603-1991 The scope of IEEE Std. 603-1991 (including the correction sheet dated January 30, 1995) goes beyond that of IEEE Std. 279. While a section-by-section comparison of the two standards shows them to be similar, IEEE Std. 603-1991 includes features not addressed in IEEE Std. 279. IEEE Std. 279 is limited to the sense and command features of safety systems. IEEE Std. 603-1991 includes not only those features, but expands the scope of the standard to address execute features, power sources and supporting systems, such as heating, ventilating and air conditioning (HVAC). IEEE Std. 603-1991 is also broader in scope than IEEE Std. 279 since it invokes several other industry standards as requirements. Many of these sub-tier standards are not part of the current licensing basis of either of the Authority's plants and were issued after they received their operating licenses. As currently proposed, the new rule could be interpreted as elevating these sub-tier standards to the status of regulatory requirements. Administrative Procedures Act The Administrative Procedures Act (APA) typically requires a rule to be issued as a proposed rule and allow for public comment. A final direct rule may only be used when the issue is entirely non-controversial. The October 17, 1 997 Federal Register Notice concludes that because the Commission did not receive any adverse comments on a draft regulatory guide (Reference 3) which also endorsed IEEE Std. 603-1991, that the issuance of this rule is non-controversial. The lack of public comments, adverse or otherwise, on a draft regulatory guide is not a reliable indicator of whether or not an NRC staff position is controversial. As stated at the bottom of most regulatory guides," ... Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and solutions different from those set out in the guides will be acceptable if they provide a basis for the findings ... (emphasis added)." In contrast, the rules and regulations contained in Title 10 are legally binding and compliance is required. While the Authority closely monitors developing regulatory issues and NRC concerns, it does not routinely develop and submit comments on new or revised NRC staff positions expressed in guidance documents such as Regulatory Guides.
Conclusion The effective date of the new rule should be delayed until these concerns about it can be resolved and appropriate changes incorporated. The Authority does not consider the adoption of IEEE Std. 603-1991, as a regulatory requirement, non-controversial. This letter does not contain any new commitments. If you have any questions regarding this matter, please contact the Director - Nuclear Licensing, Ms. C. D. Faison. Very truly yours,
~=:JD~f Senior Vice President and Chief Nuclear Officer cc: Regional Administrator U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Office of the Resident Inspector U.S. Nuclear Regulatory Commission Indian Point 3 P. 0. Box 337 Buchanan, NY 10511 Mr. George F. Wunder, Project Manager Project Directorate 1-1 Division of Reactor Projects 1/11 U. S. Nuclear Regulatory Commission Mail Stop 14B2 Washington, DC 20555 Ms. K. Cotton, Acting Project Manager Project Directorate 1-1 Division of Reactor Projects 1/11 U. S. Nuclear Regulatory Commission Mail Stop 14B2 Washington, DC 20555
DOCKETED THE SKS GROUP, INC. USNRC Dale R. Wuokko, PE President YI IEC -4 PS :11 686:1 Perivale Park Toledo, OH 43617 USA "'OCKET tlROPOSED fUE 5o OFFICE or SFC"-{1- y AULEtviAJ<!~ *r*:;--* -~- O { tp!l Fl? 5°3'115) ADJUDIC/1.!IOI ~S TAFF (~~~~ 5313~ Secretary ovember 28, 1997 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemakings and Adjudications Staff
Subject:
Comments Pursuant to Direct Final Rule to Amend 10 CFR 50.55a(h) On October 17, 1997, the Nuclear Regulatory Commission (NRC) published in Volume 62 of the Federal Register, pages 53932-53935, a direct final rule to amend 10 CFR 50.55a(h) by incorporating by reference IEEE Standard 603-1991 , "Criteria for Safety Systems for Nuclear Power Generating Stations.11 The SKS Group, Inc., as a small business, is a provider of services to the nuclear power industry and thus has a vested interest in new rulemaking that affect this industry. The SKS Group, Inc. offers the following comments for consideration by the NRC: I . According to the Federal Register notice, the NRC has proceeded with direct final rulemaking because there was no adverse public comment on draft regulatory guide, DG-1042, which was proposed Revision 1 to Regulatory Guide 1.153, "Criteria for Safety Systems. 11 This draft regulatory guide proposed to endorse IEEE Std. 603-1991 . Due to the significant regulatory differences between a draft regulatory guide made available for public comment and a rule change, it is inappropriate for the NRC to conclude that the rule change is noncontroversial simply because no adverse comments received on the draft regulatory guide. The future impact on nuclear power plant modifications due to a rule change can be significant, while a change to a regulatory guide, which is not a regulation, may have little effect. Accordingly, a ptoposed rule change is likely to receive a closer analysis than a proposed change to a regulatory guide. It does not appear that a good cause existed for the NRC to dispense with its normal rulemaking procedures in determining that such procedures were unnecessary and proceeding directly to a final rule, especially considering that nearly two full years had elapsed since the draft regulatory guide had been issued.
- 2. The rule will require that changes to protection systems in operating nuclear power plants initiated on or after January 1, 1998 meet the requirements in IEEE Std. 603-1991.
This has the potential to significantly impact the Spring 1998 refueling outages of nuclear Acknowledged by card ....~c 1 9 ..1~~. -
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power plants due to its fast track implementation. Typically, modifications to protection systems are performed during refueling outages. The rule is unclear as to what is meant by "changes to protection systems initiated". For example, is a change that is in the design concept form on January 1, 1998, considered already initiated? Or, does this rule wording mean that the change must be physically initiated in the plant? Without clearly defining this term, it will be up to NRC inspectors to subjectively determine what constitutes a change initiated before January 1. Furthermore, due to the fast track implementation of the rule, some licensees may need to obtain exemptions under 10 CFR 50.12 in order to implement their planned, designed, budgeted, and material-ready plant modifications in the Spring of 1998. A better implementation of this rule would be to clearly define in a statement of considerations what is meant by the term "changes to protection systems initiated" and allow in the effectiveness of the rule a reasonable amount of time for those plants with changes already budgeted, designed, and scheduled in the near-term to proceed without potential rule violation. This rule change should not be effective until these issues have been addressed.
- 3. The previous rule only applied to "protection systems". The revised rule defines the terms "protection systems", "safety systems", and "safety-related systems" as synonymous.
However, these terms are not synonymous in the licensing bases for all nuclear power plants, in particular to those plants which are older plants. The updated final safety analysis reports for plants define which systems are considered protection systems, and safety systems and safety-related systems are not always categorized as protection systems. This rule will, therefore, impact the licensing bases of some plants and require that IEEE Std. 603-1991 now be applied to these systems. The significance of this impact should be analyzed prior to the implementation of the rule. Yours very truly,
/JPS Commitment. Innovation. Energy.
DOCKETED USt-!RC James M. Levine Tf;k WJ.~3jl3-5300 Mail Station 7602 Palo Verde Nuclear Senior Vice Presr,?it 0£C -4 Flt( ($0~3-60n P.O. Box 52034 Generating Station Nuclear Phoenix, AZ 85072-2034 U. S. Nuclear Regulatory Commission J t\l:l NUMBER ATTN: Document Control Desk PROPOSED RULE.......=-~ Washington, DC 20555-0001 ( {p:J.F~ S3t:t75) Attention: Rulemakings and Adjudications Staff c"a ~,e S-3'i 3 :i.)
Subject:
Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 Docket Nos. STN 50-528/529/530 Comments on Proposed/ Final Rule (10 CFR 50.55a(h))
Dear Sirs:
In the October 17, 1997, Federal Register (62FR53932 and 62FR53975), the NRC published changes to 10 CFR 50.55a(h) that would require compliance with the Institute of Electrical and Electronics Engineers (IEEE) Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," instead of IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations" for protection system changes. The rule changes were published concurrently as both a proposed rule and a direct final rule, effective January 1, 1998, unless significant adverse comments are received by the NRC by December 1, 1997. APS has reviewed the proposed/final rule changes to 10 CFR 50.55a(h) and is providing comments in the enclosure. The comments are considered to be significant adverse. Compliance with the changes to the regulations would impose considerable costs on APS and other licensees that would not have a commensurate safety benefit above compliance with the current regulations. The costs do not appear to have been considered by the NRC. If the NRC plans to go forward with implementation of the proposed/final rule, APS requests that the effective date be delayed from January 1, 1998, to June 1, 1998, in order to provide enough time to request and obtain NRC review of an exemption from the rule. OEC 1 9 1997 Acknowledged by I 181
U.S. NUa.EAR REGULATORY COMMISSKJ( RULEMAKINGS &ADJUDICATIONS 8Wf OFFICE OF THE SECRETARr OFTHECOMMI~ OocllnentStallab Postmark Date _' :J 3 / 'i 1 Copies Recalvoo _/ _ , _ _ __ Add'I Cople8 ____3_ __ _ s rw""
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Comments on Proposed / Final Rule (10 CFR 50.55a(h)) Page 2 Please contact Mr. Scott Bauer at (602) 393-5978 if you have any questions or would like additional information regarding this matter. Sincerely, JMUSAB/RKR/mah
Enclosure:
Comments on Proposed / Final Rule - (10 CFR 50.55a(h)) cc: E. W. Merschoff K. E. Perkins J. W. Clifford J. H. Moorman
ENCLOSURE Comments on Proposed / Final Rule (10 CFR 50.55a(h))
Comments on Proposed/ Final Rule- (10 CFR 50.55a(h}) The direct final rule as written represents a significant impact on licensees' design and licensing basis. Incorporation of IEEE-603 into 10 CFR 50.55a does not recognize the substantial impact on existing licensed facilities. This change will significantly revise the design and licensing basis of installed plant protection systems, since there is an ongoing need for plant modifications that maintain the original system design requirements. The implementation schedule for the direct final rule does not take into account the level of effort required to review revisions to regulatory guides and the more significant impact of code of federal regulations revisions. The direct final rule does not include an implementation schedule and does not consider the impact on plant modifications that have already been designed, but have not been installed. This would require re-review of the plant modification to ensure that the requirements of the direct final rule will be met. Our comments included below are grouped into functional categories that provide specific examples of the impact of the direct final rule, including the costs associated with approval of the direct final rule without regard to the applicability to existing facilities. Increased Scope of the new standard and backfit concerns As stated in the direct final rule, the backfit rule was not intended to apply to regulatory actions which change expectations of prospective applicants, and therefore the backfit rule does not apply to the portion of the rule applicable to new construction permits, new operating licenses, new final design approvals, new design certifications and combined licenses. This is consistent with past NRC practice and the discussions on backfitting in the "Value-Impact Statement" prepared for Revision 1 to Regulatory Guide 1.153. This does not acknowledge that significant licensing and design basis impacts are incurred when the same standard is incorporated into the code of federal regulations without consideration to existing licensed facilities. The principle rationale for issuance of the direct final rule was the perceived lack of comments to the revision to Regulatory Guide 1.153. The "Value-Impact" statement from regulatory guide 1.153 (Draft DG-1042) in section 2.1.2 clearly acknowledges that the IEEE-603 standard expands the scope of IEEE-279 " ...to include power sources and actuation functions .... " While the statement does continue to state that the scope expansion is essentially covered by the guidance provided in existing regulatory guides, it fails to recognize that not all-existing facilities are licensed to those same regulatory guides. We feel that this direct final rulemaking action both directly and indirectly introduces substantial changes to the licensing basis of many existing facilities. This must be considered as a backfit action and transition provisions must be identified to allow existing facilities to continue to operate under the licensing basis currently in effect.
The "Value-Impact" statement continues in each succeeding paragraph to conclude that no new requirements are imposed. This is true in the context of a regulatory guide in which an existing licensee has the ability to maintain the licensing basis in accordance with the original license conditions, or voluntarily revise commitments to endorse the new regulatory guidance. Approval of this direct final rule would compel the licensee to involuntarily revise the licensing basis to conform to the IEEE-603 standard if a plant design change was required to maintain the protection system capabilities to meet the original licensing basis (i.e. IEEE-279). Specific sections of the IEEE Standard that represent significant departures from the PVNGS Design and licensing bases are as follows:
- The fourth paragraph of the "Discussion" section in the "Direct Final Rule" states, "The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same." This is clearly unfounded, since many of the "Auxiliary Supporting Features" and "Other Auxiliary Supporting Features," for which examples are given in Fig. 3 of IEEE 603-1991, were not previously identified as being within the scope of IEEE 279-1971.
- IEEE Std. 279-1971 classifies single-failure criterion in Section 4.2 as "Any single failure within the protection system shall not prevent proper protective action at the system level when required". IEEE Std. 603-1991 classifies single-failure criterion in Section 5.1 as "The safety systems shall perform all safety functions required for a design basis event in the presence of: (1) any single detectable failure within the safety systems concurrent with all identifiable but non-detectable failures; (2) all failures caused by the single failure; and (3) all failures and spurious system actions that cause or are caused by the design basis event requiring the safety functions."
Furthermore, Section 5.1 of IEEE-603 states that if a design does not meet the single-failure criterion then design features shall be provided or corrective modifications shall be made to ensure that the system meets the specified reliability requirements. The single-failure criterion specified in IEEE Std. 279-1971, was one of the major guidelines/criteria used in the design and development of the protection systems in current nuclear plants. The single-failure criterion specified in IEEE Std. 603-1991 constitutes a major change from the original criteria and scope. To evaluate the impact of the revised single-failure criteria would require a detailed review of existing systems just to determine the scope of impact. Since the new criteria was not a design guideline in current protection systems and constitutes a significantly expanded scope, revising existing systems to meet this new criteria would typically require major modifications or in some cases complete system replacements. This would be particularly true with the various electronic systems, such as the protection systems, since the existing electronic design and circuitry do not allow major changes in logic and function. 2
- IEEE Std. 603, Section 5.3, Quality, requires that all components be designed, manufactured, inspected, installed, tested, operated, and maintained in accordance with ANSI/ASME NQA1-1989. Palo Verde is not currently committed to ANSI/ASME NQA1-1989. This requires that as a minimum, that a quality control program be implemented to maintain items covered by IEEE Std. 603 requirements separately from the current quality control program requirements for the remainder of the components.
- Section 5.12 of IEEE 603 provides requirements for Auxiliary Features. Section 5.12.1 states that "Auxiliary support features shall meet all requirements of this standard." Section 5.12.2 (2) further defines other auxiliary features as " ... part of the safety system by association (that is, not isolated from the safety system) .... "
This section also refers to Appendix A as an illustration of this criterion. The definition on page 13 provides further definition of "auxiliary supporting features, Systems or components that provide services (such as cooling, lubrication, and energy systems) required for the safety systems to accomplish their safety functions." This would include systems such as instrument air, non-class AC power, and non-class DC power as well as many other systems that would be categorized as an auxiliary feature under the new requirements. Modification of any system/component and it associated auxiliary system would require an extensive, costly modification, including the redesign of all or part of non-class auxiliary systems.
- Section 6.2.3 of IEEE 603 states that override capability must be provided for all protective functions, such that manual actions can be taken to control safety equipment after automatic protective actions have occurred. This is a new requirement that was not previously specified in IEEE 279-1971. Therefore under the new requirements of the proposed change to 10CFR50.55a, any modification to safety equipment, its associated actuation or control logic, or its associated power supplies, would have to incorporate override capability, along with an analysis showing that the displays and controls necessary to perform override functions are, "located in areas that are accessible, located in an environment suitable for the operator, and suitable arranged for operator surveillance and action." This would significantly increase the costs associated with designing and implementing necessary modifications. No explicit design analyses were previously required or performed to demonstrate the adequacy of manual override capabilities for safety functions.
- Section 8 of IEEE 603 has no equivalent requirement in IEEE 279. The design and licensing basis for PVNGS electric power systems is based on IEEE 308-1971.
With the requirement that changes and modifications to the protection system conform to the standards cited within IEEE 603 and associated regulatory guides, this is clearly an increase in the scope of plant changes and will incur an associated increase in the cost of modifications. 3
- The PVNGS digital computer Core Protection Calculators (CPC's) were not designed to any standard for software-based protection systems, since none existed when the plant was licensed. The CPC's were licensed via analyses presented in CESSAR and in the FSAR. IEEE 603 requires that software-based protection systems adhere to IEEE/ANS 7.4.3.2. Since the CPC's were designed and licensed under a different standard, any changes to the CPC's would require major changes if not complete system replacement due to the fundamental differences in the original design and the new requirements.
- The identification of requirements in Section 5.11 of IEEE 603 is different than those in Section 4.22 of IEEE 279. IEEE 603 refers to the requirements of IEEE 384-1981, IEEE 420-1982, and IEEE 494-1974. The PVNGS UFSAR commits to the requirements of IEEE 384-1974 and IEEE 420-1973. This would require a revision to the PVNGS labeling procedure before we could perform any modifications that relabel plant components.
The NRC staffs backfit analysis states in part that this is not considered to be a backfit because any changes required by the NRC would need to have a backfit review and any other changes would be done voluntarily by licensees. Since the requirements of IEEE Std. 603-1991 affect more systems than IEEE Std. 279, "voluntary" changes to these added systems will be more costly than the current requirements. Based on the increased costs, modifications that could improve plant reliability may no longer be economically viable. Even though this is not considered to meet the backfit criteria, this rule change will increase modification costs such that licensees may be discouraged from making plant improvements unless mandated by the NRC. The rationale for direct final rule The background section of the direct final rule action uses as a justification for this action that "Because there were no adverse public comments to Revision 1 to Regulatory Guide 1.153, the Commission believes that there is general public consensus that IEEE Std. 603-1991 provides acceptable criteria for safety systems in nuclear power plants." This neglects the very different purpose of a review of a regulatory guide, which does not mandate licensee compliance for holders of operating licenses. The lack of comments received on the regulatory guidance revision is incorrectly used to conclude that the utility industry has no substantial comments to the mandatory imposition of the IEEE-603 standard to existing plants. The revision to R.G. 153 was not reviewed from the perspective of this rulemaking action. Indeed, section D of R.G. 1.153 states that the IEEE-603 standard will be used to evaluate applications for construction permits and operating licenses. Since most operating nuclear electric utilities are not applying for construction permits or operating licenses, there is little, if any, impact of the revision of R.G. 1.153 to operating facilities. 4
Palo Verde is not committed to Regulatory Guide 1.153. The Palo Verde UFSAR does not refer or commit to Regulatory Guide 1.153. Therefore, there would be no reason to review draft regulatory guide DG-1042. Since the NRC staff issued the draft regulatory guide for public comment in November 1995, it is likely that most currently licensed nuclear plants are committed to IEEE Std. 279 and if committed to Regulatory Guide 1.153, the commitment is to rev. 0 only. Since Palo Verde is not committed to Regulatory Guide 1.153, it is incorrect to assume that the draft regulatory guide was acceptable to Palo Verde. As issued for comment, there was no indication that the requirements of the draft regulatory guide would be imposed on licensees by rulemaking. Furthermore, the line of reasoning concluded that "Because there were no adverse public comments to Revision 1 to Regulatory Guide 1.153, the Commission believes that there is general public consensus that IEEE Std. 603-1991 provides acceptable criteria for safety systems in nuclear power plants." Again this is a valid conclusion as far as it goes, and we also believe that the IEEE-603 document provides acceptable criteria for safety system designs, but we do not conclude that this criteria should be retroactively applied to operating nuclear facilities that were licensed to a different standard. The backfit analysis section of the final rulemaking posting concludes incorrectly that approval of this action does not constitute a backfit situation because: "... This direct final rule does not change the licensing basis (i.e., IEEE Std. 279) for plants that do not intend to make any changes to their power and instrumentation and control systems. However, the direct final rule would require future changes to existing power and instrumentation and control portions of protection systems to comply with the new standard. This would not be considered a backfit, since the changes are voluntarily initiated by the licensee, or separately imposed by the NRC after a separate backfit analysis." This reasoning does not consider that some plant changes are necessitated by the discovery of an existing plant condition that does not conform to the original licensing and design basis of the facility. If the condition is not expected to be corrected, it is considered a "defacto change" that must receive a 10 CFR 50.59 review. If the defacto change would be acceptable under the plant licensing basis established pre-1998 using IEEE-279; it could be evaluated and implemented under the 50.59 process. However, post 1998, the 50.59 review must consider the IEEE-603 standard, and since the plant was not originally designed to this standard, it would most likely fail the 50.59 change criteria, forcing the plant to implement an unnecessary and costly modification. This does not conform to the staffs stated understanding that changes to the protection system are "voluntary" and as such are not "back fit" considerations. Plant modifications are often not voluntary. They may be forced by such factors as equipment obsolescence or wear out, increased regulatory interest in some issue, industry experience, discovery of failure modes, etc. Also, the definition of a modification's "scope of effect" as defined by IEEE 603 is not clear. It is common in the 5
nuclear industry for utilities to experience ever-widening scopes of effect whether related to modifications, new programs, or new regulations. Revisions to EQ Requirements contained within the IEEE 603 Standard as applied to existing Licensees Implementation of IEEE Std. 603-1991 invalidates licensee commitment to those older standards endorsed by regulatory guides. Therefore, the licensee will be required to reevaluate their current licensing basis in terms of the requirements of the later revision of the standards. Since, the standards are concerned primarily with equipment qualification, there is a very large impact on plant equipment and equipment installations. Paragraph 5.4, Equipment Qualification," requirements for safety system equipment conflicts with the requirements of 10CFR50.49(c)(3) that only require "harsh environment" equipment qualification. Programs based on good commercial quality that have been invoked by some licensees does not appear to meet the IEEE Std. 603-1991 requirements for a type test, previous operating experience, analysis, or combination of these that substantiates the performance requirements for safety systems that are not located in "harsh environments." The PVNGS Post-Accident Monitoring Instrumentation was not designed to IEEE 279, but instead meets specific requirements of equipment qualification, separation, etc. as laid out in the various categories of RG 1.97. If IEEE 279 type requirements are imposed on any of this equipment via a mod developed in accordance with IEEE 603, major redesign or equipment change-out may be required. Dual Licensing Basis A dual licensing basis will be created for those design modifications initiated after 1/1 /98. For example, if a component is replaced with a like component or with a slightly different component make or model number, or a minor modification to the system architecture is required, an evaluation will be required to determine which portions of the circuit and auxiliary systems must be upgraded to meet IEEE 603. It is not clear whether a partial implementation of the new design standard would be acceptable or if entire system would have to be replaced. Even if IEEE 603 applied only to the portion of the protective system that was changed, it will be difficult and confusing to keep track of the portion of the system that complies with IEEE-279 and the portion that complies with IEEE-603. 6
The direct final rule is ambiguous as to which versions of industry standards licensees are required to comply with for new work. IEEE 603 is significantly different from IEEE 279 in that it references many other standards. This leads to the following questions:
- If a certain version of a standard is referenced in IEEE 603, a second version in the Regulatory Guides, and a third version in the licensee's UFSAR, which version must be followed?
- If a Regulatory Guide is updated in the future to reference a new version of an IEEE standard referenced in IEEE 603, does such "endorsement" by the NRC constitute a de facto revision to 10CFR50.55a? If so, at what time does the change take effect?
- If an industry standard is referenced in IEEE 603 and references other standards not listed in IEEE 603; does the direct final rule require that the subsidiary references also be followed?
- Some of the standards referenced in IEEE 603 have been revised and reissued since the publication of IEEE 603. Do licensees have the flexibility to adopt the newer requirements, or must they adhere to the older ones?
- In cases where the UFSAR now takes exception to certain prov1s1ons of the standards, must licensees revise the UFSAR to eliminate the exception for future work?
Revisions to current Licensing Basis The level of effort and expense to incorporate IEEE 603 into Palo Verde's licensing basis documents is not trivial. It would be necessary to update the UFSAR and other licensing and design basis documents to reflect the new requirements, including the newer versions of industry standards referenced by IEEE 603. This would include the need for a demarcation between the portions of the plant that are still under the old standard vs. those under the new requirements. Considering the number of disciplines that would be affected, it is estimated that this work would cost at least $1 00K. IEEE 603 requirements will effectively change our design basis with regard to modification work, because the new requirements would need to be specified as base design inputs. These inputs are reflected in plant output documents such as design basis manuals, engineering drawings, engineering specifications, installation specifications, etc. Therefore, significant engineering manhours will need to be spent in a review to identify the differences between our current committed documents and the later revisions of the inferred documents, and determine the impact and complete any required updates. Additionally, personnel would have to be trained on the changes and new requirements. 7
The accelerated approval schedule for the implementation of the direct final rule to incorporate the IEEE-603 standard into 10 CFR Part 50.55a represents a substantial financial and logistical impact to the operation and maintenance of nuclear electric utilities. The impact is manifested in many areas of station engineering, design, and licensing activities. First the impact to the plant licensing basis is more than trivial. The incorporation of the IEEE 603 standard and by reference the many attendant and supporting industry standards represent a major change in the licensing basis of virtually every electric utility in commercial operation. In this one action a large percentage of the industry standards that PVNGS has committed to and was licensed to are changed to later revisions, by virtue of their endorsement in regulatory guides, without an appropriate period of time to assess the impact to the station of this indirect revision to its licensing basis. As identified in the direct final rule notification published on 10/17/97, "If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices." This paragraph implies that if the industry standard is endorsed by a regulatory guide, then the utility must now comply with the industry standard endorsed by the regulatory guidance, even if the utility is committed to a different industry standard in the currently approved licensing basis. This action in itself represents a backfit issue in that utilities will be required to comply with requirements that are currently outside of its their current licensing basis. Implementation considerations This accelerated approval cycle makes no allowance for in-process modifications. At a multi-unit facility such as PVNGS, design changes require 18 months following the issuance of the final engineering design change package to be fully implemented in all three units. Specifically, a design change implemented in Unit 2 in the Fall, 1997 refueling outage would comply with IEEE-279, and the same modification implemented in Unit 1 in the Spring, 1998 refueling outage must comply with IEEE-603. There is insufficient time to assess the impact of this direct final rule on the acceptability of the plant design change scheduled for implementation in Unit 1. If the plant design change must be reengineered to comply with the requirements of IEEE 603, it must be rescheduled to the next available outage window. This could represent a three year delay in implementation of the design change. This will also lead to additional documentation burden on station staff in maintaining the design and licensing basis for pre 1998 modifications and post 1998 modification implementation. Additionally, if a design modification is installed pre-1998, and required redesign to install post-1998, due to new design requirements imposed by the mandatory compliance to IEEE 603, must the pre-1998 installations be updated to the revised 8
design? If redesign is not required, this represents at a minimum an additional burden to configuration management activities, and brings into question the value added by mandatory compliance to a revised licensing basis condition. If redesign is required, then the final rulemaking as published does not require pre-1998 installations to be modified with the new IEEE 603 compliant design, apparently in an effort to address the lack of a backfit analysis for this proposed rulemaking activity. Within the context of compliance to IEEE 603 as incorporated into 10 CFR 50.55a, several areas appear to be open to interpretation. For example:
- Within the context of 10 CFR 50.55a, how is the phrase "to the extent feasible and practical (Reference IEEE 603 Section 6.4) to be interpreted?
- IEEE 603 Section 2 provides several definitions that have not been previously defined in law such as "Class 1E", "division" and "associated circuit." The definition of division varies between operating plants and significantly between the part 50 and part 52 plants.
- What is the legal interpretation of "acceptable reliability" (Reference IEEE 603 Section 8.3)?
- IEEE 603 requires that indication shall be automatically actuated if the bypass is expected to occur more than once a year. How is this frequency to be monitored and what regulatory consequences would result from more frequent bypasses?
- The term "acceptable reliability" is extremely subjective.
And finally, this rulemaking does not include consideration of license life extensions. By incorporating this standard in 10 CFR 50.55a without transition considerations established for plant license extensions, it would require reanalysis of the protection systems design bases originally approved to IEEE-279, against the criteria of IEEE 603. It is extremely unlikely that plants designed in the 1960's and 1970's would be capable of meeting licensing conditions established in the late 1990's. This effectively precludes the application for plant life extension without substantial protection system redesign and increased costs. 9
CP&L Carolina Power & Light Company PO Box 1551 411 Fayetteville Street Mall Raleigh NC 27602 CP&L Letter: PE&RAS-97-099 December 1, 1997 DOCKETNlMBEAM1
~ROPOSED RULE n 5 0 ,,
Secretary ( <P:!FR53C,15) U.S. Nuclear Regulatory Commission ( <P~F~5'3'13;t) Washington, DC 20555-0001 Attn: Rulemakings and Adjucations Staff
Subject:
Requested Comments on the NRC Proposed and Direct Final Rules on IEEE 603-1991 (62 FR 53975 and 62 FR 53932)
Dear Sir or Madam:
Attached are the comments of Carolina Power & Light Company (CP&L) on the NRC proposed and direct final rules amending 10CFR50.55a(h), "Protection and Safety Systems," to incorporate by reference IEEE standard 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations." This proposed amendment was published in the Federal Register (62 FR 53975 and 62 FR 53932) on October 17, 1997. CP&L recognizes that the proposed amendment to the regulations was prompted by the withdrawal of IEEE 279-1971, and CP&L supports efforts to improve nuclear regulations. However, the amendment appears to be self-contradictory and lacks sufficient coordination with respect to more recent NRC guidance. Therefore, CP&L recommends that the NRC withdraw the direct final rule and postpone the proposed amendment pending further revision and some additional clarifications, as described in the attached specific comments. Please contact me at (919) 546-6901 if you have questions. Sincerely,
~~:::f--D3A Performance Evaluation & Regulatory Affairs HAS Attachment DEC - 4 1997 AcknoWledged by card----
U.S. NUCLEAR REGUlATORY COll1881JN AULEMAKINGS &AOJUDICATIONBIWF OFFICE OF THE SECRETARr OF THE COMMISSION Docllnn St lllc:t Postmark Da1a if< hrniffed by {'ttrD ( Gt..lle, ~er btl_ 1 12/,/n wl,o r,driut'tl Ct>ntrnt11T l'r-ott1 ;,dtracf;c rukMJc/f'lj C, ipies Received I we.b.s, ft j 7'o..Jf°rM;riet/ on ,~J~/11 P~d'I Copies Reprocllced ___J____,..__ Snecial Distribution ,A§J ti r:: '4./a. / 1:, '7?I
Page 2 CP&L Letter PE&RAS-97-099 December 1, 1997 Requested Comments on the NRC Proposed and Direct Final Rules on IEEE 603-1991 (62 FR 53975 and 62 FR 53932) cc: Mr. L.J. Callan, Executive Director for Operations Mr. S.J. Collins, Director, USNRC Office of Nuclear Reactor Regulation Mr. L.A. Reyes, Regional Administrator, Region II Mr. J.B. Brady, USNRC Resident Inspector - HNP, Unit 1 Mr. B.B. Desai, USNRC Resident Inspector - HBRSEP, Unit 2 Mr. V.L. Rooney, USNRC Project Manager - HNP, Unit 1 Ms. B.L. Mozafari, USNRC Project Manager- HBRSEP, Unit 2 Mr. C.A. Patterson, USNRC Resident Inspector - BSEP, Units 1 and 2 Mr. D.C. Trimble, USNRC Project Manager - BSEP, Units 1 and 2 Chairman J.A. Sanford - North Carolina Utilities Commission USNRC Document Control Desk
Page 3 CP&L Letter PE&RAS-97-099 December 1, 1997 Attachment Requested Comments on the NRC Proposed and Direct Final Rules on IEEE 603-1991 (62 FR 53975 and 62 FR 53932) Comment 1: As written, the proposed amendment appears to be self-contradictory or at least confusing. Item 3 of the proposed amendment to 10CFR50.55a(h) states, in part:
" ... changes to protection systems in operating nuclear power plants, initiated on or after January 1, 1998 must meet the requirements in IEEE Std. 603-1991, and the correction sheet dated January 30, 1995."
Also, Item 4 of the proposed amendment to 10CFR50.55a(h) states, in part:
" ... For construction permits, operating licenses, final design approvals, design certifications and combined licenses issued after January 1, 1998, safety systems must meet the requirements set forth in IEEE Std. 603-1991, and the correction sheet dated January 30, 1995 ."
However, the first paragraph of 10CFR50.55a, which was unchanged by the proposed amendment, appears to be in conflict with the above changes. The first paragraph of 10CFR50.55a states, in part:
"Each operating license for a boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f) and (g) of this section and each construction permit for a utilization facility is subject to the following conditions .... "
The inclusions of operating plants in Items 3 and 4 of paragraph 10CFR50.55a(h) and the exclusion of paragraph 10CFR50.55a(h) from the applicability statement for operating plants creates a confusing contradiction in the regulations. Comment 2: If the intent of the proposed amendment was to expand the regulations to make paragraph 10CFR50.55a(h) applicable to operating plants, then considerably more guidance is needed in the following areas: a) What are the requirements on existing equipment that interfaces with plant modifications subjected to the new recommendations set forth in IEEE Std. 603-1991? b) What are to become of the commitments that licensees made to the recommendations contained in IEEE Std. 279 which are not continued in IEEE Std. 603-1991? This issue is especially important since the NRC did not provide an analysis of the differences between these two standards with the proposed amendment to expand paragraph 10CFR50.55a(h) to operating plants. c) How can IEEE Std. 603-1991 be reconciled with a Risk-Informed and Performance-Based regulatory philosophy? In the nearly seven years since IEEE Std. 603-1991 was issued, the NRC has promoted development of Risk-Informed and Performance-Based regulations. At the NRC Workshop on Risk-Informed Regulation in August 1997, the NRC included Defense-In-Depth, rather than Single Failure Proof, to be a part ofRisk-lnformed and Performance-Based regulation. In apparent conflict with this approach, IEEE Std. 603-1991 indicates that "The performance of a probable assessment of safety systems ... shall not be used in lieu of the single-failure criterion."
From: <PMHStar@aol.com> DOCKETED To: TWD2.TWP9(CAG) USNRC Date: 12/1/97 10:31pm
Subject:
Public Comments re: Codes and Standards; IEEE Standard From: Philip M. Holzman Strategic Technology And Resources, Inc. OFFIC ;:.-- OF SE0't.,J 1* *it1 ' ~ .\ )\' { 195 High Street RULEMAl<li',i*-.:s th ) ADJUDI(.,;t", ' nI .,j I\~-*
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If.FF Winchester, MA 01890 617 729-9212 PMHStar@aol.com DOCKET SD PROPOSEORULE..a.::i.......- - To : NRC Interactive Rulemaking (Ms. Carol Gallagher) ( tp~F~ Ss1,s) ( (,,,~PllS3 '132) Public Comments Re: Codes and Standards; IEEE National Consensus Standard
@ 62 Fed. Reg. 5373 (October 17, 1997)
The attached Word for Window 2.0 file (PMH603.doc) contains our comments on the subject rulemaking. As we discussed our browser (AOL) evidently does not upload to your system. Please submit the attached comments. I appreciate your efforts and a confirmation of receipt. If there are any problems or questions please call or e-mail. PhilH OEC . 4 i997 Acknowtedged by * ***.... 11111 .......... III llt
U.S. NUCLEAR REGUlATORY ISSION RULEMAKINGS &ADJUDICATlONS STAFF OFFICE OF THE SECRETARV OF THE COMMISSION DoQlnartS1allllca " ' - " DolB $1, 1.c,;ffe/ ht e-~-1f,6I1-stH ,,, I./, /11 {r-.fr;MI Ro,. j.. 1,,,..,,,-1;*o1L ,,. It,.,.,,.,,, ,,,,,;1,,) ~Received _ ___,,_ _ _ __ Add'I ~ Reproduced --=- 3 _ __ Speclal()isd)UtlOn,~ afp.*.:..,
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From: Philip M. Holzman Strategic Technology And Resources, Inc. 195 High Street Winchester, MA 01890 617 729-9212 PMHStar@aol.com To: NRC Interactive Rulemaking
Subject:
Public Comments re: Codes and Standards; IEEE National Consensus Standard 62 Fed. Reg. 5373 (October 17, 1997)) Strategic Technology And Resources, Inc. (STAR) hereby submits the following comments on the Nuclear Regulatory Commission's (NRC) proposed rule to revise the
- current provisions of 10 C.F .R. Part 50.55a(h). STAR is an engineering consulting firm providing technical services to individual licensees and industry groups.
We have reviewed the proposed revision and based on the comments provided below are both significant and adverse. Accordingly, we recommend that the direct final rule be withdrawn and that public comment resolution be achieved prior to issuing a final rule. Since significant changes to the rule may be necessary to resolve the comments provided here and by others and to ensure that all comments are appropriately accommodated, we further recommend that the revised rule be issued as a proposed final rule with an additional comment period. Detailed Comments Inappropriate Expansion of 10 CFR 50.55a(h) Scope
- 1. The proposed rule incorrectly expands the scope of "protections systems" by equating protection systems, safety systems, and safety-related systems. 50.55a(h) codifies IEEE 279 provisions as they apply to "protection systems". The proposed rule incorrectly concludes that protection systems, safety systems, and safety-related systems" are synonymous. This is inconsistent with past practice, existing NRC guidance, and common usage of these terms and the scope of protection systems stated in IEEE 603. There are numerous examples supporting the a more limited protection system scope definition, including:
- IEEE 603-1991 Figure 2 presents a 3X3 matrix of safety system elements and states "The protection system of IEEE 2 79-1971 (withdrawn) and of this standard is the sense and command features for the reactor trip system and the engineered safety features." This properly defined scope encompasses only one of the nine safety system elements presented in the figure. Power sources and execute features are elements of the safety system but are appropriately excluded for the definition of
protection system. Sense and command features associate with auxiliary supporting features and other auxiliary features are similarly excluded
- SRP Section 7.1. Instrumentation and Controls; II. Acceptance Criteria, support the view that "protection systems" are a limited set of safety systems. This section states in part:
Although ANSI/IEEE Std 279 contains acceptance criteria only for protection systems, the concepts ofANSI/IEEE Std 279 are applicable as guidance to other I&C safety systems and to non-safety I&C system for which high functional reliability is a goal."
"The scope of IEEE Std 603 includes all I&C safety systems, which are the systems covered in Sections 7.2 through 7.6 of the safety analysis report (SAR) . Therefore,
- while the guidance of IEEE Std 603 and the requirements ofANSI/IEEE Std 279 are equally applicable to protection systems, IEEE Std 603 is more directly applicable to I&C safety systems other than the protection systems (i.e., information systems, safe shutdown systems, and interlock systems)."
- 2. Although important design criteria, such as the single failure criterion and equipment qualification, have been applied to these other safety system elements (e.g., power sources, auxiliary supporting features), their use has not been based on the provisions of 50.55a(h) or the use ofIEEE 279. Rather, the General Design Criteria, Standard Review Plan, industry standards, and other NRC staff guidance have been used to determine the applicable requirements. This is particularly true for the electrical system. For example, IEEE 308, IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations, has been the design standard used by virtually all plants and AL WR applications for electrical systems. There are a plethora ofNRC documents including regulatory guides, standard review plan, and standard technical specification (to name just a few) and licensee FSARs which utilize this (uncodified) standard as the basis for system design and acceptability. In none of these instances has IEEE 279 been cited since it only deals with "protection systems". Now the NRC proposes to utilize IEEE 603 as the design requirement for these systems without any information or technical considerations as to its potential impact on the operating characteristics of these power systems and their associated equipment.
- 3. Historically, neither the NRC staff nor the industry has interpreted the existing 50.55a(h) scope as applying to these other safety system elements. For example, we are unaware of the use of IEEE 279 as the design basis document for power systems.
Individual plant licensing bases, including SERs, and the SRP all recognize IEEE 308 as the appropriate design basis document for safety system power sources.
- 4. Related to the above, we have questioned others regarding the extent of current IEEE 603 usage by the industry and the NRC during the past decade for safety system
modifications or reviews of new plant designs (e.g., AL WRs). It was assumed that the proposed codification of IEEE 603 would have been preceded by its usage, at least as a guidance document, for such safety system designs and modifications. Surprisingly, we learned that during this period IEEE 603 use has been limited to guidance (i.e., specific provisions were found to be inappropriate in certain cases) and even this use has been limited to traditional protection systems (i.e., RPS and ESF AS). In these instances the IEEE 603 provisions were used to amplify on the criteria contained in IEEE 279. Given this practice for the newest designs, including the AL WRs, requiring compliance with IEEE 603 for an expanded equipment scope (all safety systems as compared to traditional protection systems) is untried and should not be codified until its technical and safety implications are determined.
- 5. There are numerous IEEE and other industry standards that are referenced and utilize by the NRC and licensees to establish/review system designs. However, these
- documents, except for IEEE 279, serve as guidance with selective exceptions and interpretations made on case-by-case bases by licensees and the NRC staff. IEEE 603 is appropriately in this category and should remain so until there has been a detailed study of the technical, safety and cost implications of codifying its requirements for both modifications and new systems. At a minimum, it is inappropriate for the NRC to implement such a pervasive rule change without a reasonable showing of no adverse overall safety impact.
IEEE 279 is Not Obsolete it was Simply Withdrawn
- 6. The proposed rule implies that codification of IEEE 603 is needed because the provisions in IEEE 279 are obsolete. Nothing is further from the truth. IEEE 279 was withdrawn by the IEEE because its fundamental principles were carried forward and expanded in IEEE 603 . Importantly and from a regulatory perspective, IEEE 279 appears appropriate as a requirement since it articulates broad principles that must apply to protection systems. Codification of IEEE 603, which is a more detailed document, may be inappropriate since certain of its requirements may be inappropriate or extremely difficult to implement in all cases. This is particularly true for its use as a requirement for all system modifications and its expanded safety system scope. In summary, the general principles ofIEEE 279 and not the more detailed provisions of IEEE 603 are the appropriate basis for a rule.
- 7. There is little evidence that compliance with IEEE 603 for modifications is reasonably achievable without some exceptions. There are numerous industry document including FSARs, topical reports, and safety system analyses that establish the IEEE 279 compliance bases for existing protection systems. Similarly, there are numerous NRC documents that interpret and clarify the IEEE 279 provisions. Until there were few if any such documents addressing compliance methods for all the provisions of IEEE 603 for either an entire system or portions of a system. A significant new body of documents must be developed to provide NRC and industry personnel with guidance for modification compliance. Many of the
topics may be controversial and not easily resolved. In the interim, regulatory stability will be lost as licensees, NRC personnel, and manufacturers struggle to define modification limits and interpret how the new requirements can be integrated into modifications.
- 8. While IEEE 603 can serve as guidance for these protection system modifications, fully compliance is neither practical or achievable in all cases. Further, its imposition may effectively inhibit modification or upgrades to such systems. In our discussions with designers involved in upgrading existing protection systems with digital and computer-based devices, it is clear that full compliance with IEEE 603 is virtually impossible for all modified aspects of the system. Some exceptions are needed to safely and efficiently integrate the new equipment into the existing system. We believe that, after thoughtful consideration and reasonable efforts to achieve full compliance with IEEE 603 for a range of protection system
- modification, both NRC and industry technical personnel must conclude that fully compliance in all instances is either impractical or unachievable.
Inappropriate to Apply IEEE 603 to all Modifications
- 9. It is difficult or virtually impossible to define the boundaries of modifications when determining which portions of the modified system must comply with IEEE 603. Is compliance limited to only those items that are physically replaced or added or must all the circuits that interface with these devices meet the IEEE 603 criteria. Both cases present conceptual, practical, and safety implications and need thoughtful consideration.
- 10. For those unfamiliar with protection and electrical safety system designs, requiring full compliance with a new standard may be conceptually appealing but we strongly
- caution against such a superficial conclusion. Analogies to ASME piping system upgrades or to upgrading commercial/residential wiring during renovations are simplistic and inappropriate. Safety system electrical design concepts, such as separation/isolation criteria and testability, interrelate and are pervasive throughout a system design. Given these interrelationships, the complexity of system circuitry, and the physical proximity of various circuits within electrical panels and in cable raceways, it may be virtually impossible to implement new criteria for portions of a system.
If an entire system is replaced then compliance is possible. If major pieces of equipment are replaced (e.g., analog instrument racks or RPS logic cabinets), compliance may be achievable within the new devices but problems may arise when these new devices are interfaced with existing system equipment. Since both new and old equipment and wiring will exist within each modified circuit, full compliance might be virtually impossible for the entire circuit without significantly expanding the modification scope.
For smaller scope modifications (e.g., addition of a time delay relay to an ESFAS circuit or other modifications to existing protection schemes) which do not involve replacement of major devices, compliance, even if achievable, may significant expand the design and modification efforts. What are now relatively simple modifications, such as replacement of a electromechanical relay with a solid state relay, may become substantially more complex and intractable. One obvious implication is that licensees may be reluctant to voluntary implement some of these modifications. No Compelling Safety Need
- 11. We are unaware of any compelling safety issue prompting the adoption ofIEEE 603 in lieu of IEEE 279 for modifications to existing facilities. Absent this safety need we are concerned that adoption of IEEE 603 will promote instability in a technical
- area (i.e., protection system compliance with IEEE 279) currently containing a rich source of documents and practices defining NRC expectations and industry interpretations of the existing 50.55a(h) reference to IEEE 279.
Adverse Safety Impact Requiring forced implementation of the proposed rule's provisions may have significant adverse safety impact. Some examples are:
- 12. Currently, modifications to "protection systems" must meet the current licensing basis (IEEE 279). To the extent feasible equipment manufacturers and licensees may apply the provisions ofIEEE 603 with due recognition that all aspects of the standard cannot be incorporated without design inconsistencies. Analysis and judgment are applied to determine which provisions can be incorporated without producing unnecessary complexity or compromising the overall system design. We believe the NRC staff, when reviewing proposed designs and modifications, applies a similar logic. This is clearly preferable to forced compliance to all IEEE 603 criteria. Under the revised rule, forced compliance to all IEEE 603 provisions can result in instances where unnecessary complexity or a compromised overall system design must be implemented.
- 13. Forced IEEE 603 compliance will involve increased costs as manufacturers and licensees establish compliance review processes and expand the modification scope to implement unnecessary compliance-driven design changes. With due consideration to these unnecessary compliance-drive costs, licensees may conclude that system upgrades and modernization that were previously justified will not be implemented due to increasing costs, complexity, delayed licensing reviews/approvals, and similar considerations.
Pending Revision to IEEE 603
- 14. The IEEE is in the process of issuing a revised IEEE 603 in the very near future and, apparently, the NRC is aware of this. It appears inappropriate to formally codify a standard that will imminently be reissued. This is particularly true when this revision will more fully address the use of digital-based systems. What rational does the NRC have to rush to rule an IEEE standard that will be out of date within a few months?
Summary In summary, we recommend the following:
- Retain the existing provisions of 10 CFR 50.55a(h) and implement protection system compliance for major system modifications or for new facilities using the guidance of IEEE 603 and the Standard Review Plan.
The proposed definition for "protection systems" should be revised to reflect current guidance and practices. The term "protection system" is not synonymous with "safety system" or "safety-related system". If reference to a withdrawn standard (IEEE 279) for new plants is unacceptable from a policy perspective then the NRC should consider incorporating the general principles of IEEE 279 directly into the rule. Alternatively, those selective provisions ofIEEE 603 applicable to "protection systems" can be referenced in a proposed rule revision. However, adoption of a broader rule scope regarding equipment/system coverage or compliance criteria should not implemented without a detailed studies regarding overall safety implications and associated implementation costs.
December 2. 1997 NOTE TO: Emile Julian Chief . Docketing and Services Branch FROM: Carol Gallagher RES. ORA SUBJECT : DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING Attached for docketing is a comment letter related to the Direct Final/Proposed Rulemaking on Codes and Standards: IEEE National Consensus Standard. This letter was received via e-mail on December 1. 1997. The commenter's address is Philip M. Holzman. Strategic Technology and Resources . Inc .. 195 High Street. Winchester . MA 01890 . Please send a copy of the docketed comment to Satish Aggarwal (mail stop TlO-ElO) for his records.
Attachment:
As stated cc w/o attachment : S. Aggarwal
Davu Nioret Southern Nuclear Vice Pr~< id~nt Operating Company Fa rley P* oJect PO Box 1295 Birmingham, Al aba ma 35201 Tel 205.992 .5 131 OOCl<ETED DEC - 2 1997 December 1, 1997 ~.., FKUIWllaMD I. THERN A COMPANY nergy to Serve Your World"" Docket Nos. 50-321 50-348 50-424 HL-5529 50-366 50-364 50-425 LCV-1133 Mr. John C. Hoyle Secretary of the Commission D CKET BER SO U.S. Nuclear Regulatory Commission PROPOSED RUlE,...!l.,::!~----- Washington, DC 20555-0001 ((p~Fl<.53'17S) ,7;;\ (~~FR 53 '13:J.) ~ ATTENTION: Rulemakings and Adjudications Staff Comments on Proposed Rule "Codes anq Standards; IEEE National Consensus Standard" (62 Federal Register 53932 of October 17, 1997)
Dear Mr. Hoyle:
Southern Nuclear Operating Company (Southern Nuclear) has reviewed the proposed rule regarding the incorporation by reference of IEEE 603-1991 and the direct final rule which is scheduled for implementation on January 1, 1998. In accordance with the request for comments, Southern Nuclear is in total agreement with the NEI comments which are to be provided to the NRC. Southern Nuclear has a sincere concern that the basis for the direct final rulemaking is ungrounded and, as discussed in the NEI comments, it has circumvented the normal course of rulemaking promulgation. Should you have any questions, please advise. Respectfully submitted, DNM/JMG cc: See following page Et - 4 997 Ackn edged by - - -- -*
U.S. NlXlEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF Cff1CE Of THE SECRETARY OF THE COMMISSION Docll'nentS1atlsllcs ?ostmarkDate CopiesRacelvad fa.Kt.I on
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U.S. Nuclear Regulatory Commission Page2 cc: Southern Nuclear Operating Company Mr. C. K. McCoy, Vice President, Plant Vogtle Mr. H. L. Sumner, Jr., Vice President, Plant Hatch Mr. J. B. Beasley, General Manager - Plant Vogtle Mr. R. D. Hill, General Manager - Plant Farley Mr. P.H. Wells, General Manager - Plant Hatch U.S . Nuclear Regulatory Commission, Washington, DC Mr. J. I. Zimmerman, Licensing Project Manager - Farley Mr. N. B. Le, Licensing Project Manager - Hatch Mr. D . Jaffe, Senior Project Manager - Vogtle U. S. Nuclear Regulatory Commission, Region II Mr. L.A. Reyes, Regional Administrator Mr. T. M. Ross, Senior Resident Inspector - Farley Mr. B . L. Holbrook, Senior Resident Inspector - Hatch Senior Resident Inspector - Vogtle HL-5529 LCV-1133
DEC ~1 1 '97 04: 11PM SOUTHERN t~UCLEHR 205 992 6108 be: Mr. W. G. Hairston, ill Mr. K. W. McCracken Mr. J. W. McGowan Mr. G. Bockhold Commitment Tracking System (2) Document Control
Station Support Department DOCKETED PECO NUCLEAR US HRC PECO Energy Comp;my 965 Chesterbrook Boulevard A Unit of PECO Energy WaynP PA 19087-5691 "97 DEC -2 P4 :21 November 26, 1997 Mr. John C. Hoyle Secretary of the Commission Nuclear Regulatory Commission Attn: Rulemaking and Adjudications Staff Washington, DC 20555-0001
Subject:
Comments Concerning NRC Direct Final Rule and Proposed Rule "Codes and Standards; IEEE National Consensus Standard" (62FR53932 and 62FR53975, dated October 17, 1997)
Dear Mr. Hoyle:
This letter is being submitted in response to the NRC's request for comments concerning the 10 CFR 50 Direct Final Rule and Proposed Rule, "Codes and Standards; IEEE National Consensus Standard," which were published in the Federal Register on October 17, 1997 (i.e., 62FR53932 and 62FR53975, respectively). It is the NRC's intent to amend its regulations to incorporate by reference the Institute for Electrical and Electronic Engineers (IEEE) Standard 603 -1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," a national concensus standard for power, instrumentation, and control portions of safety systems in nuclear power plants. Specifically, 10CFR50.55a(h) would be revised to include an endorsement to IEEE Standard 603-1991, and replace an IEEE standard currently endorsed in the regulations which has been withdrawn by IEEE. The final rule, if promulgated, will become effective on January 1, 1998, unless significant adverse comments are received by December 1, 1997. PECO Energy appreciates the opportunity to comment on the subject direct final rule and proposed rule. We are concerned that this rulemaking effort may result in added costs in maintaining a plant in peak operating condition without a commensurate increase in safety. It is our understanding that this rulemaking will not mandate the use of the guidance specified in IEEE 603-1991 on plants with construction permits issued prior to January 1, 1971. Accordingly, PECO Energy offers the attached comments on the direct final rule and proposed rule for consideration by the NRC. If you have any questions or require additional information, please do not hesitate to contact us. Very truly yours, ,-.j a. I.L,,t" ,j,- G. A. Hunger, Jr. Director - Licensing Attachment DEC - 4 1997
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ATTACHMENT Attachment Page 1of 3 PECO Energy Comments NRC Direct Final Rule and Proposed Rule "Codes and Standards; IEEE National Consensus Standards" PECO Energy representatives recently attended a meeting of the Nuclear Power Engineering Committee (NPEC) of the Institute for Electrical and Electronic Engineers (IEEE)on November 12, 1997. At this meeting, there was significant discussion with regard to the impact that this rulemaking may have on existing plants. An NRC staff representative at this meeting provided an interpretation with regard to the intent and implications of this rulemaking effort. Therefore, if the NRC continues with promulgation as a final rule, PECO Energy recommends that the current language presented in this rule be revised to more accurately reflect the NRC staffs intent, as presented by its representative, in order to minimize the potential impact on operating plants. PECO Energy offers the following comments for consideration by the NRC.
- 1. Defining *protection system" as synonymous with "safety systems" and "safety-related systems" in new paragraph 10CFRS0.55a(h)(2)(i) definitions, is a significant increase in the scope of the rule.
The "protection system" as defined by both IEEE 279 and IEEE 603 is only a portion of the safety system. For example, the protection system does not include the auxiliary supporting features or the execute features. An impact of this rulemaking is to apply the requirements of IEEE 279 to these portions of the safety system. This change in definition expands the scope on existing designs even if plant changes are not made. PECO Energy judges that such an expansion of scope will lead to additional cost without a material increase in the availability or reliability of the systems. It is therefore recommended that "protection system" be deleted from paragraph 10CFRS0.55a(h)(2)(i) and that a new paragraph be added defining "protection system" consistent with IEEE 603. Note: For clarity, we have used the definitions contained in IEEE 603 in the comments to follow in lieu of equating "protection system" with "safety system."
- 2. It is PECO Energy's understanding that the intent of this rulemaking applies to system-level replacements. The Statements of Consideration as published in the Federal Register notice indicate that in-kind (i.e., like-for-like) replacements of protection system components are not considered changes to the protection system for the purposes of this rule. However, neither the rule nor the Statements of Consideration discuss the impact of modifications which fall between system-level replacements and in-kind replacements. Therefore, we recommend that the rule be modified to explicitly state that IEEE 603 only applies to system-level replacements of protection systems and additions of new protection systems, and suggest that paragraph 10CFRS0.55.a(h)(2)ii be revised to state:
"Changes to protection systems discussed in this paragraph encompass system-level replacements of protection systems and additions of new protection systems permitted by license amendments or 10CFRS0.59, and plant specific departures from a design certification rule under 10CFR part 52."
Attachment Page 2 of 3 The following are a few examples in order to assist in clearly defining a system-level replacement to which the rule is intended to apply.
- The replacement of a reactor protection system transmitter with a transmitter manufactured by a different vendor must meet IEEE 279 since it is a like-for-like replacement.
- The replacement of the High Pressure Coolant Injection (HPCI) system speed controls with new digital equipment must meet IEEE 279 since the modification only replaces a portion of the HPCI system.
- The upgrade of the Average Power Range Monitor (APRM) portion of the Neutron Monitoring System to add the ability to detect and suppress potential reactor instability must meet IEEE 279 since the modification only replaces the APRM signal processing components, output relays, recirculation flow transmitters, and operator displays. If this modification were to also replace the neutron detectors, Local Power Range Monitor (LPRM) cards, and associated power supplies, the modification would be considered a complete replacement and must meet IEEE 603.
- The replacement of the Source Range Monitors (SRMs) and Intermediate Range Monitors (IRMs) with Wide Range Neutron Monitors (WRNMs) must meet IEEE 603 since it involves the complete replacement of the system, including sensors, preamps, signal processors, output relays, and operator displays. Reuse of a few components (e.g., selected cables, raceway, and control room panels where the displays are mounted) would still place this type of modification in the category of a complete replacement.
- 3. It is PECO Energy's understanding that IEEE 603 was written by the IEEE for application to new designs; there was no intent to backfit any revised requirements onto existing plants. However, the subject rulemaking rule will impose IEEE 603 criteria on changes made at PECO Energy's Limerick Generating Station facility. The impact of invoking IEEE 603 guidance on existing plants must therefore be carefully assessed. We note that the Statements of Consideration published in the Federal Register states:
*section 3 of IEEE Std. 603-1991 references several industry codes and standards. If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices.*
We further understand that the discussion provided in the Statements of Consideration delineate that standards referenced in IEEE 603, but not endorsed by a regulatory guide, are not mandatory requirements even if IEEE 603 invokes the referenced standard by the use of "shall." It also indicates that an NRC regulatory guide which endorses a previous revision of a standard is the current NRC recommended practice even if IEEE 603 invokes the current revision of the standard by the use of *shall." However, this specific information is not included in the language of the rule itself. Omission of this information from the rule could lead to confusion in the industry as well as the NRC during implementation of the requirements. Therefore, we recommend that the above portion of the Statements of Consideration be included in the specific language to the rule.
Attachment Page 3 of 3 Additionally, Limerick Generating Station's Safety Analysis Report contains commitments to various design criteria and requirements which are identified as major sections of IEEE 603 (e.g., independence and electrical separation, capability for test and calibration, bypass and inoperative status indication, setpoints, electrical power sources, etc.). Some of these commitments were made prior to publication of some regulatory guides, and others are commitments to previous revisions of IEEE standards and NRC regulatory guides. Furthermore, some of the commitments in the Safety Analysis Report provide alternate approaches to meeting the guidance stipulated in the various regulatory guides. Although portions of these commitments may differ from current regulatory guides, and may not be all-inclusive when compared to IEEE 603 or regulatory guide criteria, they constitute an acceptable design and licensing basis for modifications to our facility. Therefore, we recommend that the following wording be added to the rule:
"For plants licensed prior to January 1, 1998, commitments in the Safety Analysis Report for a facility to the topics covered by IEEE 603 also constitute a method acceptable to the Commission for meeting these regulatory requirements."
- 4. PECO Energy recommends that additional guidance be provided in Section 7 of the Standard Review Plan (SRP), as well as the NRC Inspection Manual procedures, to ensure consistent interpretation of the revised rule. We suggest that this guidance include the following:
- Examples of the types of modifications which must meet IEEE 603 vs. IEEE 279,
- Guidance on proper application of standards referenced in IEEE 603, and
- Guidance on identifying previous licensee commitments which constitute an acceptable method of meeting the requirements of IEEE 603.
This guidance will assist in meeting the requirements of the revised rule. It will also help to facilitate a consistent interpretation of the requirements by NRC inspectors. Finally, it is PECO Energy's understanding that this rulemaking does not mandate the use of IEEE 603 criteria on plants with construction permits issued prior to January 1, 1971. Therefore, the guidance stipulated in this IEEE standard does not apply to changes at our Peach Bottom Atomic Power Station facility, since its construction permit was issued prior to January 1, 1971. Commitments to IEEE 279 contained in the Peach Bottom Atomic Power Station Safety Analysis Report will continue to apply without being upgraded to IEEE 603.
DOCKETED US. C !!¥~i~KC~~OON *sm~A=E, romE~R, Nr 14~ AREA CODE 716 546-2700 "91 OEC -2 p 2 :40 ROBERT C. MECREDY Vice President November 24, 1997 Nuclear Operations Secretary U.S. Nuclear Regulatory Commission Attn: Rulemaking and Adjudications Staff DOCKET PROPOSED RULE.,.___S ____o__ Washington, D.C. 20555
Subject:
Proposed Change to 10 CFR 50.55a(h) R. E. Ginna Nuclear Power Plant ( ~~F~S'3'i?S] ( (,:J..~SJ"l:J~ g I Docket No. 50-244
Dear Secretary:
This letter is in response to the proposed rule to incorporate the requirements of IEEE Std. 603-1991 into 10 CFR 50.55a(h), effective January 1, 1998. In the procedural background section of the Federal Register notice, it is stated that the NRC considers this rulemaking noncontroversial, and therefore is publishing this proposed rule as a direct final rule. However, if the NRC receives significant adverse comments by December 1, 1997, the NRC will publish a document that withdraws the direct final rule. Our purpose is to provide such significant adverse comments, as follows:
- 1) The apparent basis for judging that this rulemaking is noncontroversial is that no significant comments were received on the proposed Revision 1 to Regulatory Guide 1.153, which would endorse IEEE Std. 603-1991. Since a regulatory guide is not binding, and merely suggests a method for meeting regulatory criteria, RG&E considers that issuing a final direct rule based on such feedback is not an appropriate action.
In subparagraph (3) of this proposed rule, it is stated that changes to protection systems initiated on or after January 1, 1998 must meet the requirements set forth in IEEE Std. 603-1991, and the correction sheet dated January 30, 1995. This is considered a significant backfit for Ginna Station, which was issued a Construction Permit in 1966 and whose design is generally in conformance with IEEE Std. 279-1971. No backfit analyses was performed by the NRC for this proposed Rule, as required by 10 CFR 50.109. It cannot be stated that all future changes to the Ginna Station protection systems would be voluntary, since the definition of "changes" in subparagraph (2) (ii) of the proposed rule includes modifications permitted by license amendments, OEC - 4 1997 Acknowtedged by card----
U.S. NUCLEAR REGULATORY G"* ***,: ,-1 RUl.EMAKINGS&ADJUn 1r. l' .* ,*..* OFACE OF r1-*- 0F THE Docu Postmark Date _ II/ :J ". / tJ 7 Copies Received /. . Add'I Coples Rer 3 Special Distribut, A5yar-wa.t
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which could be imposed by the NRC. Furthermore, changes made on account of equipment obsolescence are not entirely voluntary, but would have to meet the provisions of IEEE Std. 603-1991 under this proposed rule. We do not believe that the NRC has legitimately shown that this proposed rule is noncontroversial, and it must therefore be withdrawn, to be replaced by a rule (if any) which would have gone through the 10 CFR 50.109 process, as required by NRC regulations. Very truly yours, Robert c. GJW\482 xc: Mr. Guy S. Vissing (Mail Stop 14B2} Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 US NRC Ginna Senior Resident Inspector
12-01-1997 20:30 P.02 NUCLEAR UTILITY GROUP ON E:0UIPM£NT QUALIFICATION UITIC aoo 1400 I. TIIICET, N. W. WAeklNOTON, D, c. eooos-:seoa Tl:1.ll!~HONIC (102) 371*15700 December 1, 1997 DOCKET pfa>OSEDRIUE PR ~ -::;O V ( IP'(;.F/l5.3'J75) /j'i\ Mr. John C. Hoyle Acting Secretary, U.S. Nuclear Regulatory Commi11ion ( G,~FR.S3'13'J.) ~ 1 Wuhiqton, D.C. 20555 Re: Code1 and Standard,; IEEE National Comen1u1 Stapdu,I ,,21e4. Ba, 53.932 lOctobfr 17, 1997))
Dear Mr. Hoyle:
The Nuclear Utility Group on Equipment Qualification ( 11NUOEQ")1 hereby 1ubmit1 the following comments on the Nuclear Regulatory Commiuion'a ("NRC") proposed rule to revise the current provisions of 10 C.F.R. Part 50.SSa(h). 2 The NUGEQ baa reviewed the comments developed by the Nuclear Utility Backfitting and Reform Group ("NUBARG") and the Nuclear Energy Institute ("NEI") and agrees with the comments, observations, and recommendations presented by these two organizations. The NUGEQ agrees that the propolcd revision is controversial and believes that the adverse comments developed by NEI, NUBARO and the NUOEQ are significant. Accordingly, we recommend that the direct final rule be withdrawn and that public comment resolution be achieved prior to issuina a final rule. Since significant changes to the rule may be neceaaary to resolve these adverse comments and to enaure that all comments arc appropriately aeeommodatcd, we further recommend that the revised rule be iasued as a proposed rule with an additional comment period. The NUGBQ is comprised of 36 electric utilities in the United States and Canada, including NRC licensees authorized to operate over 100 nuclear power reactors: The NUOEQ wu Conned in 1981 to addre11 and monitor topics and i1aue1 related to equipment qualification, particularly with respect to the environmental ** qualification of electrical equipment pursuant to IO C.F.R. S0.49. z 62 Fed. Reg. 53,932 (October 17. 1997). EC - 4 1997 Ackr,owledgecf by card HIIIIIIIIIIIIIIIIIHlilliUI i 1$221
U.S. tlDf.AR REGUlATORV COIIIISSIOI( RUlEMAl<INGS &AllJUDK:ATIONS STAR= OFFICE OFllfE SECRETARY OF THE COMMISSION Oocunai Slallallcl ~-0.. Fa.fe/ l:J. f/'1 7 I
12-01-1997 20:30 P.03 Mr. John C. Hoyle December 1, 1997 Page2 Additional observations reprdin1 changes to the current provisions of 10 C.F.R. Part SO.SSl(h) include: No Safety Need: We are unaware of any compelling safety issue prompting the adoption oflEEB 603 in lieu of IEEE 279 for modifications to existing facilities. Absent this 1afety need, we are concerned that adoption of IEEE 603 will promote instability in a technical area (i.e., protection system compliance with IEEE 279) currently containing an extanaive history of satisfactory implementation. reflected in licensina buea documentation and practices for NllC power reactor licensees, which defines NRC expectations and industry application of the existing 50.SSa(h) reference to IEEE 279. Advene Con1equenca: We arc concerned that the adoption of IEEE 603 for plant modifications may have adver1e aafety consequence,. Since the propo1ed revilion will provide little if any safety enhancemcnta, it ii reasonable to conclude that nesanve considerations may dominate the overall effect on safety of the propo1ed rule change. Suc:h negative comiderations include:
- Inappropriate implementation of all criteria applicable to "protection 1ystems11 to other safety ayatem features (e.g., instrumentation and control (I&C) portions of auxiliary 1Upporting features and auxiliary features) not originally designed to meet all theu criteria.
- Inappropriate implementation of all criteria applicable to "protection 1yatem1" to other aafety system elements (I.e., execute and power aourcca) not originally designed to meet theae criteria.
- Elimination of prudent ul8 of engineerins judgment by licensees and the NR.C technical staff when determinins the extent to which plant/system/
equipment modiftcationa can reasonably accommodate all the provilion1 of IEEE 603.
- Existence of a confusing two-tiered compliance scheme for protection systems.
- Uncertainties regarding "modification boundaries" when establishina compliance.
12-01-1997 20:31 P.04 Mr. John C. Hoyle December l, 1997 Pagel
- Expenditure and diversion of resources to interpret and revise licensee, industry, and NR.C guidance documents regarding compliance with 10 C.F.R. SS.SSa(h) rather than addreaaina matters of obvious safety benefit.
- Added modifieation complexities and co1t1 which may discourage implementation of modifications that would otherwiae be cost/benefit justified.
E:s.panded Definition or "protecdon 1y1tem1": The NEI and NUBARO comments make clear that the proposed definition of 11 protoction systems" significantly expands the scope of "protection 1ystem1" beyond current uaage. We agree that both current practice and the language of IEEE 279 limit the scope of protection systems to the sense and command features of the Reactor Protection Syatem (RPS) and Engineered Safety Features Actuation Systems (ESFAS). We alao note that Standard Review Plan (SRP), Section 7.1. Inatnunentation and Controls; II. Acceptance Criteria, supports the view that "protection systems" are a limited set of aafety systems. This SRP section states in part:
Although ANSI/IEEE Std. 2 79 contain, acceptance criteria only for protection systems, the concepts ofA.NSIIIEEE Std. 279 are applicable as guidance to other I&C safety systems and to non-sqfoty I&:C system for which high functional r,liability is a goal. "
"The scopB of IEEE Std. 603 Includes all I&C safety systemsi which are the systems covered in Sections 7.2 through 7.6 of the SQjety analysis report (SAR). Therefore, while the guidance of IEEE Std 603 and the r,quirwm,nts ofANSI/IEEE Std. 279 are equally applicable to protection synem.s, IEEE Std 603 is more directly applicable to /d:C safety systems other than th, prot,ctton systems (I.e., Information systems, sqfe shutdown systems, and interlock systems). "
12-01-1997 20:31 P.05 Mr. John C. Hoyle December 1, 1997 Page4 In summary, the NUGEQ recommends that compliance with the provisions of IEEE 603 for modifications to existing facilities be voluntary and that the current licensing basia (i.e., IEEE 2.79) and definitions for "protection sy1tem" remain unchanged. We appreciate the opportunity to comment.
- colm H. Philip11 Jr.
William A Horin Counsel to the Nuclear Utility Group on Equipment Qualification
12-01-1997 20:30 P.01 WINSTON & STRAWN Telephone 1400 L Street. N.W. Facsimile (202) 371-5700 Wahington, D.C. 20005-3S02 (202) 371-S9S0 FACSIMILE DATE: December t, 1997
- PLEASE DELIVER THE FOLLOWING PAGE(S) TO:
4t NAME:_ _ ___.M~r.., ..,J.oh..,n_.c..,_H_o...y,. .I~e______________ FIRM/COMPANY: Office of the Secretaa - U.S. NRC M-.1..00~"" ~. ~~*~,,.s/ \4'"-L',-..""' A.. \.\ otL,w FROM: TELECOPY NUMBER:_~(3......0.....,1)_..4._1_5-_t_to_t_ _ _ _ _ _ _ __ TELECOPY VERIFICATION NUMBER: _ _ _ _ _ _ _ __ CLIENT/MATTER NUMBER: NUMBER OF PAGES (INCLUDING COVER SHEET):_... f ___s _ __ NOTES OR COMMENTS: _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Th11 lntbnnallon oonlalned In thl1 flo1lmlle muup II IIIDmey prlvll11ed end aonlldentllll lnfonnarlon Intended only for the 1111 or die lndlvldllll or entll.y n1med llbave. ltrhe ruderofdll1 m11nae 11 not the intended recipient, or tho employco or 111nt rupon1lltl11 u, deliver le to th11 ln111iuled rKlpient, )'OIi ue llllreby nadfled Iha! any dluemllllllon. dllcrlbutlon or CXIPYUII oClhll c:ommunlcallon II saictly prolllbltlld. 1ryo11 have 1eGDlved thl1 aommunic:llllon In 11rror, ple111 immedlllely notify 111 by t.el11phou, 111d l'lllllm die orl1lnll m11111111 to III It the above lddrea via the U.S. .Pmtal Servloe. 1111111k you OPERATOR INITIALS: _ _ __ N~: _ _ _ _ _ _ _ __ NO_ __ CONFIRMATION: Yes_ _ If you do not roGOive all die pqes, pleue call Mary Wood at (202) 371-5881 as soon u poulble. Thank you.
WINSTON & STRAvVN 35 WEST WACKER DRIVE 1400 L STREET, N.W. 6 . RUE OU CIRQUE CHICAGO, ILLINOIS 60601-9703 75008 PARIS. FRAN~ WASHINGTON , D.C. 20005-3502 200 PARK AVENUE 43. RUE OU RHONE (202) 371-570Q NEW YORK, NY 10166-4193 1204 GENEVA. SWITZERLAND FACSIMILE (202) 371-5950 December 1, 1997 DOCKET N BER SQ 1\11r. John C. Hoyle PROPOSED RULE,_!..!!...,,;;::;..:- , : _ (tt,a,~~63'115) Office of the Secretary U.S . Nuclear Regulatory Commission ( fo')..F~S391:J.) Washington, D.C. 20555-0001
- Attention: Rulemakings and Adjudications Staff
Dear 1\11r. Hoyle:
On behalf of the Nuclear Utility Backfitting and Reform Group, we are submitting these comments to address the direct final rule regarding IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations." The direct final rule would require licensees to comply with IEEE Std. 603-1991 for future modifications to safety-related protection systems in lieu of IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," which is part of the licensing bases for many plants. The scope of IEEE Std_ 603-1991 is substantially broader than IEEE Std. 279 and addresses additional functional and design requirements that are not addressed in IEEE Std. 279. For example, IEEE Std_ 603-1991 establishes requirements for the 11 sense-and-command," "execute," and "power supply" functions
- for all safety-related systems, while IEEE Std. 279 only addresses the "sense-and-command" function for reactor protection systems and engineered safety feature actuation systems.
Under 10 C.F.R. § 50.109, the direct final rule would appear to constitute a backfit for the reasons discussed below, and therefore a backfitting analysis should be performed prior to the rule becoming effective. As also discussed below, the use of a direct final rule to impose new standards on licensees appears inconsistent with the Administrative Procedure Act (APA). The use of a direct final rule is typically limited to issues which have no substantial impact. 10 C.F.R. § 50.55a(h) currently requires licensees with construction permits issued after January 1, 1971 to comply with the revision of IEEE Std. 279 in effect on th~ date of the application for a construction permit. The proposed revision to 10 C.F.R. § 50.55a(h) would require changes to protection systems initiated on or after January 1, 1998 to comply with the additional requirements of IEEE Std. 603-1991. IEEE Std. 603-1991 also requires adherence to additional IEEE standards referenced in IEEE Std. 603-1991. However, the revisions of the IEEE standards referenced in IEEE 603 Std. 1991 are not necessarily the revision that licensees are currently committed to. The additional requirements of IEEE Std. 603-1991 would thus c - 4 1997 AcknoWledged by
U.S. UCl.EAR REGUlATORY COMMISSION RU Kl GS & ADJUDICATIONS STAFF OFFICE OF THESECREl'ARV OF THECOMMISSION Ooc'8n8ntStatilllcl Postmark Date fated on 1/Jll /q'&' ~ f,n*rkJ on 1a/z./'J'1 Copies Recelv8d _ _ _ , ----- Add'I Coples Reproduced --=3:;,.._ _ __ Special Distl1blltion~ 1........__.......__
'Rl't>S
WINSTON & STRAWN Mr. John C. Hoyle December 1, 1997 Page 2 impose a change to the current licensing basis of plants, and for that reason, would represent a backfit. The rule would also result in changes to plant procedures, and therefore would constitute a backfit as defined in Section 50.109(a)(l). Due to the broader scope of IEEE Std. 603-1991 in comparison with IEEE Std. 279, licensees would be required to make changes to their existing procedures that administratively control the power source and instrumentation and control functions of protection systems. 1 The NRC concluded that a backfit analysis for the implementation of IEEE Std. 603-1991 was not necessary because it apparently considered future changes to protection systems to be "voluntarily initiated by the licensee... ." 62 Fed. Reg. at 53,934. We recognize that it may be appropriate, in some circumstances, for licensees to use new criteria or standards for plant changes initiated by the licensee, such as changes that involve advances in design or operations. However, to mandate the use of new criteria for all future plant changes would constitute a backfit. For example, the NRC has long recognized that using criteria more stringent than the Standard Review Plan (SRP) would be a backfit except in limited circumstances. See NRC Manual Chapter 0514 (Appendix at pp. 1-2). In addition, the NRC's assumption that all changes are voluntarily initiated is not necessarily accurate. Changes made on account of equipment obsolescence, planned maintenance, or scheduled replacement would not be voluntary. Nevertheless, under the proposed rule, these changes would have to comply with IEEE Std. 603-1991 . Historically, most licensees have relied on IEEE Std. 279 as their current licensing basis for changes to protection systems. The proposed rule could result in dual licensing bases. For example, if a licensee replaced a single component in a protection system, the replaced component would comply with IEEE Std. 603-1991, but the remainder of the system would comply with IEEE Std. 279. In addition to the backfitting issues identified above, we have other concerns regarding whether or not the use of a direct final rule in this manner is consistent with the AP A Typically, a direct final rule can only be used when there is no significant impact. The additional requirements imposed by IEEE Std. 603-1991 could have significant impacts as noted above. Moreover, we note that the "good cause" exemption of Section 553(b)(3)(B) of the APA (which An additional impact would be the increase in systems, structures, and components that would be brought within the scope of the Maintenance Rule as discussed in I 0 C.F.R. § 50.65. In addition, it is our understanding that IEEE Std. 603-1991 is currently being revised to address the use of digital and computer-based systems. A more efficient use of both the NRC's and licensees' resources would be to wait until IEEE Std. 603-1991 has been revised before initiating rulemaking.
WINSTON & STRAWN Mr. John C. Hoyte December 1, 1997 Page 3 is the basis for a direct final rule) only allows an agency to forego the notice-and-comment rulemaking "when the agency for good cause finds that notice and public procedure thereon are impracticable, unnecessary, or contrary to the public interest." None of these factors apply here. Therefore, ordinary rulemaking with the notice-and-comment format is the appropriate rulemaking process for this issue. Indeed, the very purpose of notice-and-comment rulemaking is to give the agency the benefit of the public's views. Just because an IEEE standard is characterized as a national consensus standard does not mean that the public has had an opportunity to express its views. Nor would such an opportunity relieve the NRC from its legal responsibilities under the AP A In bypassing the normal comment process, the NRC relied upon the lack of adverse comments on proposed Regulatory Guide 1.153, "Criteria for Safety Systems." Licensees are not required to comply with the guidance in a regulatory guide and may do so voluntarily. Thus, the lack of adverse comments on a regulatory guide should not be taken as a reason for dispensing with public involvement in the rulemaking process here. In lieu of rulemaking, the NRC should consider less burdensome alternatives. One alternative would be to allow licensees the option of complying with either IEEE Std. 279 or IEEE Std. 603 on a voluntary basis. However, if the NRC wishes to adopt IEEE Std. 603-1991 as a binding requirement, it must go through the ordinary notice-and-comment rulemaking process, including a backfitting analysis, to determine if the imposition of IEEE Std. 603-1991 is justified. The NRC should resolve the public comments received on this rulemaking before making the direct final rule effective. If necessary, the NRC should allow for a second comment period to ensure that all comments are adequately resolved.
~:2J~~u DanielF. Stenger James R. Fitzgerald Counsel to the Nuclear Utility Backfitting and Reform Group
pfaOSEDRll.E 0 (<,~FA. s1,15J /ic\ DOCKE TED ( '12t:,e 5J'i1;.) \!,b USNRC Comments on Proposed Rulemaking Implementing IEEE 603 into 10 CF 50.55a(h)
"97 OEC -1 P3 :02
- 1. Because IEEE 279 is out of date, and no is longer supported by IEEE, I agree that it is no longer appropriate to cite this standard in 10 CFR 50.
OFFlCl. OF S[Cf; I , ~,RY RUL=1\1L ,,1r-.~r ,__. !..)
- 2. I do not agree, however, that the proposed change is an appropriat&ibtt(?jfutfd 'file'S ,'-
- r,~FF problem for reasons outlined below.
2a. Incorporating IEEE 603-1991 into the rules is unnecessary and unnecessarily prescriptive. The General Design Criteria (GDC) of 10 CFR 50, Appendix A contain sufficient basis for protection systems. IEEE 603, IEEE 279, and a variety of other documents contain acceptable design approaches to fulfill the requirements of these GDC. Incorporating IEEE 603-1991 (or IEEE 279, for that matter) into regulation unnecessarily restricts design approaches. 2b. Requiring that all changes to protection systems meet the requirements of IEEE 603 unnecessarily requires changing the licensing basis of many existing plants when modifications are made. This will severely inhibit introducing modern technology into safety systems. 2c. Since changes to the licensing basis will be required when a modification is made, it appears that all safety system changes involving a change from IEEE 279 to IEEE 603 will also require a license amendment. If this is the correct interpretation, the proposed rule will halt almost all upgrades. 2d. It will be impossible to rigidly implement the requirement that changes comply with the requirements of IEEE 603. IEEE 603 and 279 are systems standards. Most changes to safety systems only change a part of the system. For example, some BWRs are replacing the nuclear instrumentation signal processing and command equipment with modern digital systems. These changes do not necessarily involve replacing field wiring, sensors, auxiliary supporting features, or other auxiliary features. It is not clear how the proposed rule would apply in this case. One interpretation would be that the entire system must be upgraded to 603 whenever a change is made. This interpretation would essentially halt safety system upgrades. The complementary interpretation is that only the changed equipment must meet the requirements of IEEE 603. In this case, it may be very difficult to design the interface between the new and old portions of the system such that the all new equipment meets IEEE 603 . Requiring compliance with IEEE 603 by law severely limits the flexibility of licensees, applicants, and regulators to arrive at practical resolutions to this interface problem. 2e. Codifying IEEE 603-1991 freezes regulatory guidance on a standard that is already nearly out of date. Instrumentation and control is an area of rapid technological development. As transition from analog to digital instrumentation and control systems progresses, IEEE 603 and the supporting standards will undergo continual update to incorporate new technology and lessons learned. In fact, IEEE 603-1991 will almost
U.S. NUaJ:AR REGUlATORV COMMISSION RUt.EMAl<INGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Oocunent Statistics Postrnarlc Date f&iv~I hf C4.n;( ~(~,,.- o~ 12-/, h7 { tdr1~11eJ f+ol'h wei.si-k) Copies Received _ _;..- , ---- Add'I Cq:)i8s Reproduced __.3_____ Special Distribution ~arh,)4-l 2 I r
certainly be out of date within the next few months as balloting of IEEE 603-1996 is in progress. Placing any specific version of a standard into the regulation inappropriately discourages the use of more recent guidance as it is developed. IEEE 279 has remained in the regulation for more than 10 years since IEEE 603 has been available. This demonstrates the difficulty of incorporating new guidance into regulations. Freezing the guidance on safety system design on IEEE 603-1991 for another 10 years will do a disservice to the industry and the public. 2f. Incorporating IEEE 603 into the rule precludes the use of alternatives. When NRC originally developed 10 CFR 50.55a(h), IEEE 279 was the only available guidance on the design of protection systems. Today alternative international guidance, such as IAEA SG-50-D3 and SG-50-D8 exist. A suite of IEC nuclear standards support implementation of the principles outlined in the IAEA guidance. Furthermore, safety systems standards are beginning to emerge in the industrial sector. ISA S84.01, IEC 1508, and UL 1998 are examples of industrial standards that might reasonably be adapted for use in the design of nuclear power plant safety systems or elements of these systems. These international nuclear and industrial standards contain requirements that are different in some ways from the requirements of IEEE 603. Nevertheless, these alternative standards could be intelligently applied to accomplish the same intent. Indeed, the IAEA and IEC standards have been applied in that way in certain European countries. In the environment of a stagnant U.S. nuclear industry it is imprudent to close the possibility of using such alternative standards by referencing a single standard in regulation. Opening the possibility would of using other standards would offer an opportunity for the U.S. industry to take advantage of equipment and systems developed for the larger industrial or foreign nuclear markets.
- 3. Defining the terms protection systems, safety systems, and safety-related systems as synonymous is in direct conflict with the use of the terms in industry, and the use of the terms in IEEE 603. As noted in Figure 2 of IEEE 603, OThe protection system of IEEE Std 279-1991 (withdrawn) and this standard is the sense and command features of the reactor trip system and the engineered safety features. 6 Figure 2 shows that the protection system is only one of seven elements of scope of IEEE 603. The IEEE definition is consistent with industry usage.
3a. Using the definition section to set these terms equal can be only be interpreted as either significantly increasing the scope of 10 CFR 50.55a(h), or reinterpreting the scope of IEEE 603 . Neither of these interpretations is acceptable. Direct rule making appears to be an inappropriate procedure for making a significant increase in scope of the rules. I agree that 10 CFR 50 is rather sparse in its discussion of safety systems other than protection systems. Improvements in this area may be beneficial. Such improvements, however, should be made via a more considered process, following industry and public discussion.
Changing the scope of IEEE 603 such that it is limited to protection systems is contrary to good judgment, and is not in the spirit of Public Law 104-113. This interpretation would mean that that NRC has, by changing the definition of safety system, actually adopted a standard that is significantly different from the industry consensus standard. In fact, IEEE 603 almost certainly would not receive consensus with this revised definition. 3b. Defining protection systems and safety systems as equivalent conflicts with the use of the term protection system in 10 CFR 50, Appendix A. Clearly the requirements of GDC 20 through 25 and 29 apply to protection systems as defined by IEEE 279 and 603. It is also widely recognized that the NRC extends these requirements to the execute features of protection systems as well. Setting the terms safety system and protection system as equal extends the scope of these GDC to the auxiliary supporting features, other auxiliary features, and power sources that are within the scope of IEEE 603. Perhaps it is appropriate to extend these criteria to other portions of safety systems, but again such an extensive change in the scope of regulations should be more carefully considered. 3c. The implications of defining the terms safety related systems, safety systems, and protection systems as equivalent appear to have not been completely considered by the staff. Given the pervasive use of these terms in regulatory and licensing documentation, there are likely to be many other implications beyond those identified above. Such a far reaching change should not be made without very careful consideration.
- 4. A better approach to the problem would be to remove the direct reference to IEEE standards in 10 CFR 50 and to develop a Reg. Guide discussing the use of IEEE 279 in existing applications. For new systems, Reg. Guide 1.153 and NUREG 0800 already identify IEEE 603 as the preferred guidance. This proposed approach allows the flexibility needed to arrive at intelligent approaches to upgrades of existing facilities and hold open the opportunity to take advantage of appropriate international and industrial standards.
Gary Johnson 1255 Higuera Ct. Livermore, CA 94550 510-423-8834 kg6un@ricochet.net
December 1, 1997 NOTE TO: Emile Julian Chief, Docketing and Services Branch
~~
FROM: Carol Gallagher RES, ORA
SUBJECT:
DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING Attached for docketing is a comment letter related to the Direct Final/Proposed Rulemaking on Codes and Standards: IEEE National Consensus
- Standard. This letter was received on our interactive rulemaking website on November 27, 1997. The commenter's address is Gary Johnson, 1255 Higuera Ct.,
Livermore. CA 94550. Please send a copy of the docketed comment to Satish Aggarwal (mail stop TlO-ElO) for his records.
Attachment:
As stated cc w/o attachment: S. Aggarwal
KET PROPOSED RUlE~~..1!5-- DOCKETED (~:zr:tJ.53 CJ15) ( (p~,:'~ ~~,1~)
@tf US RC
SUBJECT:
Concerns on the Proposed Rule Change to Incorporate the illEECJi'b3- l P3 :03 Standard OFF,( ,f OF "'E*' 1*!1 . R\' The proposed rule in its current form will cause significant uncertainty in r~~ 1911tlie1: !'"~ , :~6 ' licensing requirements for digital upgrades to safety systems, and may the~UDI( ,;-,. ,- \ .. ) E'l ,,FF discourage utilities from making needed improvements. This has happened in the past 1992 draft generic letter on analog-to-digital replacements (the Oautomatic USQ6 generic letter) had such an effect. This would not be in the best interest of the NRC, the utility industry, or the public. As currently written, the applicability of the Standard is subject to interpretation. It could have a large impact on the industry. IEEE 603 is a system oriented standard, but the rule will apply it to "changes" to systems. Will this mean changeout of one component will invoke IEEE 603 for the entire system? There are significant differences in requirements of IEEE 603 vs. IEEE 279. IEEE 603 requires compliance with several other IEEE standards as indicated above. There is some question as to whether the NRC intends to require compliance with the secondary standards (even though IEEE 603 is clear on this point). Although NRC points out this rule change doesn't qualify as a mandated backfit because changes to protection systems are "voluntary," changes to these systems will in fact be necessary to maintain their reliability, because of increasing obsolescence and decreasing vendor support of existing systems. Imposing new, more stringent requirements on the changes (and possibly on the whole system as a result) could have a significant impact on cost and schedule, and may cause utilities to delay making changes that would enhance plant safety. A dual licensing basis could be created within a system; this could introduce significant confusion, since IEEE 603 is written as a system standard. The proposed rule references IEEE 603-1991. There is a new version due out soon, so the regulatory requirement will be out of date again almost as soon as it is issued. There has been insufficient dialogue between the NRC and the industry on this proposed rule change. The NRC concluded that, because they did not receive significant industry comments on the draft Reg. Guide DG-1042 endorsing IEEE 603-1991, this rule change does not require a significant comment period. However, a rule change that requires compliance with a standard has much more impact than a Reg. Guide that simply endorses the standard as an acceptable method. Raymond C. Torok Electric Power Research Institute 3412 Hillview Ave. Palo Alto, CA 94303 cknowfed ed by
U.S. NUCLEAR REGUlATORY OOMMISSION flJLEMAIONGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Docl.lnent Statlsllcs Postmark Date f'l,prni-Hd hy ~ar~I C.t.!~kr on n/, Copies Recelved ___/________
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Recommendations EPRI recommends that utilities voice their opinions in letters to NEI and the NRC, requesting that the decision on the new rule be delayed until the concerns noted above are satisfactorily resolved. Because Osignificant6 adverse comments will carry more weight than unsupported opinions, it will be helpful to include an explanation of the significance of your concern, including the potential impact on plant operation, finance or safety. Comments to NRC Mail comments to: Secretary, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Attention: Rulemakings and Adjudications Staff or hand deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4:15 p.m. on Federal workdays. You may also provide comments via the NRC's interactive rulemaking website through the NRC home page (http://www.nrc.gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905 (e-mail CAG@nrc.gov). For further information on NRC's proposed rule, contact: Satish K. Aggarwal, Senior Program Manager, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 415-6005, Fax (301) 415-5074 (e-mail: SKA@NRC.GOV). Comments to NEI Forward comments to: Alex Marion, (202) 739-8080 or e-mail am@nei.org. 3
December 1. 1997 NOTE TO: Emile Juli an Chief. Docketing and Services Branch FROM : Carol Ga 11 agher /I /) A. 1 o _L ~ RES . DRA ~ / ~ ' -~
SUBJECT:
DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING Attached for docketing is a comment letter related to the Direct Final / Proposed Rulemaking on Codes and Standards; IEEE National Consensus Standard . This letter was received on our interacti ve rulemaking website on November 26 . 1997. The commenter ' s address is Raymond C. Torok . Electric Power Research Institute. 3412 Hillview Avenue. Palo Alto . CA 94303 . Please send a copy of the docketed comment to Satish Aggarwal (mail stop TlO-ElO) for his records. Attachment : As stated cc w/o attachment : S. Aggarwal
DOCKET ... E. S. Medling An EDISON INTERNATIONAL'" Company PRaQElRWll 5o Manager, Regulatory Projects ( {,:J FIJ 63 '?75) ( ~ :J.t:1(5.3 '13-2) November 26, 197 Secretary, U.S. Nuclear Regulatory Commission E:o~ c~o Ornr, Attention: Rulemakings and Adjudications Staff Cl~O Washington, D. C. 20555-0001 -!::5:"T1 T-z I
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Subject:
Southern California Edison (SCE) Comments on Dirif'II' "TT :< lSinal t!: Rule and Proposed Rule - Codes and Standards; IEEE National Consensus Standard (62 Federal Register 53932-53935 and 53975-53976, October 17, 1997) San Onofre Nuclear Generating Station, Units 2 and 3 In the subject Federal Register Notice, the NRC issued a direct final rule which would incorporate by reference IEEE Std. 603-1991 into 10 CFR 50.55(a). The rule will become effective on January 1, 1998 unless significant adverse comments are received. This letter provides SCE's comments to the direct final rule. SCE strongly opposes the direct final rule and believes it should be withdrawn and industry comments addressed. In general, we bel i eve the rule would 1) substantially impact existing design and not meet the provisions of the backfit rule, 2) be very burdensome and costly to implement, without a commensurate increase in safety, 3) potentially reduce safety and, 4) conflict - with existing regulations. Although the NRC did not receive adverse public comments on the proposed revision 1 to Regulatory Guide 1.153, which endorsed IEEE-603, this does not indicate that mandatory compl i ance with the IEEE-603 standard is acceptable to licensees. Since compliance with Regulatory Guides i s dependent on each individual plant's licensing basis, one would not expect signi f icant public comment on regulatory guides as compared to proposed rul emaking. In add i tion, this standard was apparently directed at future applications using advanced reactors. We believe public comment on the rule is essential in this case. Additional specific comments are provided in the Enclosure. San Onofre Nuclear Generating Station P. 0. Box 128 San Clemente, CA 92674-0128 714-368-7492 Fax 714-368-7575 IE - 4 \9ST bycam, I I I
U.S. NUCLEAR REGULATORY C0MMISSKlN IU.EMAKINGS&ACWDICA110NS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION DocllnnStatBIICI Postmark Date 1-ka,d '/Je/,'vtr'"~d on ,~/, Coples Received ____, _ _ _ __
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Secretary, U. S. Nuclear Regulatory Commission If you have any questions please contact me at 714-368-7492. Sincerely, Enclosure
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cc: E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 U. S. Nuclear Regulatory Commission, Document Control Desk
Enclosure Specific Comments on Direct Final Rule Regarding IEEE-603
- 1. The rule would not meet the provisions of the backfit rule, 10 CFR 50.109.
We provide four bases for this position.
- a. Compliance with the IEEE-603 requirements for separation would require that the new standard be applied to existing plant design. This is contrary to the NRC s assumption that IEEE-603 can be separately imposed 1
upon new changes only. The Notice states: 11 The Commission has prepared a regulatory analysis which shows that the proposed amendment does not impose any new requirements or costs on current licensees who do not make changes to safety systems. However, licensees planning or proposing changes to power and instrumentation and control systems will be impacted because they will be required to meet the requirements of IEEE Std. 603-1991 for the changes even though the remainder of the plant power and I&C systems are only required to meet their current licensing basis. 11 The IEEE-603 Standard, and its incorporated-by-reference Standards, set criteria for the separation of safety and non-safety related circuits and equipment. It is a fundamental characteristic of separation that any change to the existing design, such as the addition of a new cable, requires evaluation of the impact on all other circuits and equipment in proximity to the new circuit. This concept was specifically recognized in the Value section of Draft Regulatory Guide 1042 which states: 11 IEEE Std. 603-1991 has no major application difficulties for advanced reactor types. It should be noted, however, that safety systems common to current reactors may not exist in passive advanced reactors or they may not be safety systems. Many of the principles outlined in this standard may have applicability to non-safety systems. To determine the systems subject to these criteria, an analysis of the overall plant response to postulated design basis events should be performed. 11 Adoption of the new standard as regulation would require this evaluation be conducted under the new standard's criteria for the existing design as well as the new design. This would clearly extend the application and impact of the new standard to the existing plant design.
Enclosure
- b. Compliance with the new rule would require significant procedure revisions to describe points in time during which manual control is allowed.
The Federal Register Notice states that it "considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same. 11 This conclusion is specifically contradicted by IEEE Standard 603-1991. Note 1 to Figure 2 states: "The protection system of IEEE Std. 279-1971 (withdrawn) and of this standard is the sense and command features for the reactor trip system and the engineered safety features." Figure 2 identifies a three by three matrix of safety system elements, only three of which are related to sense and command functions. Figure 2 is also a subset of safety systems as noted in Figure 1 of the Standard. We believe that substituting IEEE Standard 603-1991 for IEEE 279-1971 will result in substantial broadening of the scope of applicability for 10 CFR 50.55a(h). IEEE Standard 603-1991 places new criteria upon the use of manual action. For example, Section 4.5.1 requires that design basis documentation identify 11 (t)he points in time and the plant conditions during which manual control is allowed." Our Emergency Operating Instructions provide safety function status check guidelines for manual actions, consistent with the ABB-CE Emergency Procedure Guidelines, CENPD-152. This guidance is not time-based. In order to comply with the proposed rule, our design basis documents and procedures would require revision to describe points in time during which manual control is allowed. This would be a fundamental change to the Emergency Procedure Guidelines and our operating strategies.
- c. Full compliance with the provisions of IEEE-603 will require the development of design basis information for existing safety systems that is irrelevant to safe operation and not presently available.
As an example, Section 4.4 of the standard adds a requirement to identify acceptable rates of change for variables that control protective action. Our protection system actuation is based upon variables exceeding selected setpoints and is generally insensitive to the rate of change of the variable. While specific compliance exists for startup rate-related events, the rate of change of other variables, such as pressure, temperature, flow, or power, is unrelated to safety system actuation. A retrospective reevaluation of the design bases for Enclosure safety systems would be necessary to comply with the standard, even if only minor changes to the systems were contemplated.
- d. The IEEE Standard would require the installation of new design features not now required by the licensing or design bases. In addition, safety margins could be reduced.
Section 6.2.1 of the IEEE Standard states that: "Means shall be provided in the control room to implement manual protective initiation at the division level of the automatically initiated protective actions." San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 do not have the capability to manually initiate a Recirculation Actuation Signal (RAS) or a Loss of Voltage Signal (LOV) at the division level. It appears that a backfit regarding RAS, for example, would be required if SCE were to implement any design change related to: i) any of the following systems: Chemical and Volume Control System Containment Isolation System Safety Injection System Emergency Diesel Generators Class IE 125 V Power Class IE 480 V Power Reactor Protection System ii) any of the following areas: Control Room near Main Control Board Panels Control Room near Panels Refueling Water Tanks Containment Sumps Safety Equipment Building (in proximity to recirculation fluids) iii) or any system or component in proximity to piping and circuitry related to these systems or areas. The SONGS design and operating procedures allow component level control to accomplish the desired safety function, if needed. The NRC has not demonstrated a need for manual, division-level initiation. Indeed, installation of manual, division level actuation for RAS could actually be adverse to safety. Inadvertent initiation of RAS could Enclosure prevent adequate core and containment cooling following a design basis accident. As noted in Section 7.3.2.1 of the Safety Evaluation Report for SONGS Units 2 and 3, the Staff confirmed that there is no single failure in the current SONGS design that could inadvertently initiate RAS. Clearly, the new rule would create new requirements for existing designs, contrary to the backfit rule and the NRC s intentions. 1 Implementation of the direct final rule or the proposed rule as written, will require modifications to the procedures or organization required to design, construct, or operate a facility. In addition, compliance with the new rule would obviously result in substantial expense for implementation.
- 2. Implementation of the rule would be extremely burdensome. Section 6.4 of the standard requires that: "To the extent feasible and practical, sense and command feature inputs shall be derived from signals that are direct measures of the desired variables as specified in the design basis. 11 SCE has already obtained application-specific regulatory concurrence for the use of indirect process monitoring, such as reactor coolant pump speed sensors in lieu of RCS flow instruments or the use of pressure instruments in lieu of level instruments. The need to repeatedly justify these selections during partial or complete changes in specific instruments is unwarranted.
- 3. We believe the rule is unnecessary. Although IEEE-279 has been withdrawn, continued compliance with the historical version meets the safety and licensing bases of operating plants.
- 4. The substitution of IEEE-6O3 for IEEE-279 will force licensees to incur significant expense without a commensurate increase in safety.
SCE provides three bases for this argument:
- a. Section 4 of the standard provides prescriptive standards for the breadth and depth of detail to be included in design basis information that varies from current SONGS commitments. Section 3 of the standard also invokes more recent revisions of many IEEE Standards than the revisions to which SONGS is committed. Minor system level changes, such as the substitution of a fuse, would trigger a comprehensive revision process as the entire system would thereafter need to meet the IEEE-6O3 criteria and the criteria of the more recent versions of the referenced standards
[criteria].
- b. Existing, approved deviations from Regulatory Guide 1.75 would be Enclosure superseded and discarded. This could require significant licensee expense to obtain new approval for existing deviations.
- c. The existence of two standards would require licensees to develop and maintain additional documentation and records of changes that demonstrate the correct standard was applied to any particular design .
E*ncl osure
- d. The items discussed in Section 1 of this attachment are also examples of substantial cost incurrence.
- 5. The IEEE Standard includes design criteria that conflicts with other Commission regulations. The following examples are noted:
- a. Section 5.6.3.2 provides criteria for equipment in proximity to safety systems that conflicts with Appendix R. It is not necessary to separate both trains of safety-related equipment from a fire.
- b. Section 5.6.3.3 provides criteria for single random failure that conflicts with General Design Criterion 21. The Standard fails to
- account for safety-related portions of a non-safety-related system
- c. Section 5.13 provides criteria for multi-unit stations, such as SONGS, that conflicts with General Design Criterion 5. Simultaneous accidents are not assumed at multi-unit sites.
6 South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station . Dave A. Lavigne P.O. Box 88 General Manager Jenkinsville, SC 29065 DOC"'ETEQ n Nuclear Suppa1 Services SCE&G (803) 345.5209 (803) 634-2011 USNRC RC-97-0243 "fJ7 lIC -1 P l :5 2 November 26, 1997 Secretary of The Commission OFF/r.~ OF SECF?t i//. /f ATTN: Rulemakings and Adjudications sw/jJ~~t;*Ky,U;;;) ~-t-q_ U.S. Nuclear Regulatory Commission ~ t, i iv. 011:V: F Washington, DC 20555-0001 OCKET Nlll':R Gentlemen: PfO'OSED11J.Ell 5o ( /p:JF~5"~'1"75)
Subject:
VIRGIL C. SUMMER NUCLEAR STATION ( ~:JF/<53,?:J.) DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 COMMENTS ON PROPOSED RULE CHANGE TO INCORPORATE IEEE 603 STANDARD South Carolina Electric & Gas (SCE&G) Company submits the following comments pursuant to the subject proposed rule change. COMMENTS
- 1. The proposed rule change regarding the incorporation of IEEE 603 Standard should be considered as a mandated backfit. The voluntary" changes (i.e.
modifications) which are necessary for protective systems to maintain the required reliability will impose new, more stringent requirements on these future changes (and possibly the entire system) which will have a significant impact on the cost and schedule, while delaying changes which would enhance plant safety.
- 2. The proposed rule change could result in a dual licensing basis within a system, introducing significant confusion, since IEEE 603 is written as a system standard.
- 3. There has been insufficient dialogue between the NRC and the industry on this proposed rule change. (Draft Guide 1042 was an endorsement of IEEE 603 Standard and did not carry the same potential impact as a rule change.)
- 4. As currently written, the applicability of the Standard is subject to interpretation. (i.e. Will the change out of one component invoke IEEE 603 for the entire system? )
- 5. The proposed rule change for endorsement of IEEE 603 should not be a rule change. The regulatory guide process is a sufficient method for the endorsement of new Standards.
OEC
- 4 l997 AcknoWfedged by card'"' 1 UH PS2
U.S. NUCLEAR REGULATORY COMMISSION IU.EMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Oocll1lent Statlstlcl Postmark 0ate Copies Reoelved _ _ _ 11 1~(, I q1 Add'I ~ Reproduced __,;3;;;;..__ __ Special Distribution A-59qrw a. I I
Secretary of The Commission PR 970008 RC-97-0243 Page 2 of 2 In the view of SCE&G, the NRC should present a sufficient comment period to allow the industry to assess the proposed rule change and its potential impact on operating plants. Should you have any questions, please contact Mr. Jeffrey Pease of my staff, at (803) 345-4124, at your convenience. Very truly yours,
/
/
L;b;£ 11~~k_D Dave A. Lavigne
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DALJjwp c: J.L.Skolds W.F.Conway R>R>Mahan R.J.White G.D.Moffatt A. R.Johnson NSRC RTS (PR 970008) File 811.02 (50.002H) OMS 2
IEEE POWER ENGINEERING SOCIETY NUCLEAR POWER ENGINEERING COMMITTEE CHAIR VICE CHAIR SECRETARY W. W. BOVl.ers R. E. Hall B. P. Grim PECO Energy Company Brookhaven National Laboratory General Electric Company Peach Bottom Sta. SMB4~ Dept of Advanced Technology 175 Curtner Ave. MIC 187 1848 Lay Road Bldg. 130 San Jose, CA 95125 Pa5tChalr Delta , PA 17314 Upton, NY 11973 Vox: 408-925~130 G. K. Henry Vox: 717-456-3581 Vox: 516-344-2144 Fax: 408-925-13n ComEd Quad CHfes station Fax: 717-456-3099 Fax: 516-344-3957 grimb@ajcpol.ne.ge.com 530 Aspen HIiis Circle
\1\001M1ra@peco-energy.com r.e.hall@bnl.gov BettendOlf, IA 52722 Vox: 30~5 -22 1 x 3228 Fax: 309-65-2178 )-. A.n*u... T qacghQccmall.ceco.com SUBCOMMITTEE.CHAIRS PROPOSED November 29, 1997 C (,.~ F~ 63'115 SC-2, Qualification G.J. Toman Nu1henn International, Inc.
Secretary ( "a Fi 53~3.2.) 501 S. 11th. S1reet U.S. Nuclear Regulatory Commission Ml Vernon, IL 6286 Vox: 618-2-6000 / Fax: 618-2-66 1 Washington, DC 20555-0001 SC-3, Operations, Surveillance & Testing S. Kasturi Attention: Rulemakings and Adjudications Staff MOS 25 Piermont Drive Mell/Ille, NY 117 7
Subject:
Proposed Ru le Change to 10CFR50.55a Cone Vox: 516--423-782 / Fax: 516-385-783 skasturiQmc12000.com IEEE Standard 603, "Criteria for Safety Syste SC-4, Auxiliary Power Nuclear Power Generating Stations" P.B. stevens Stone & Webster Englnaering Corp. 25 Summer st
Dear Sir or Madam:
Boston, MA 02210 Vox: 617-689-27 15 / Fax: 61Hi89-2969 peter.stevensQstonaweb.com The Nuclear Power Engineering Committee (NPEC) of the Institute of SC-5, Rellablllty Electrical and Electronic Engineers (IEEE) heartily supports the NRC J . R. Fragola SAIC effort of endorsing industry standards for use in the commercial nuclear 7 West 36th. Street, 10th. Floor New York, NY 10018 power industry. IEEE representatives have worked with NRC personnel Vox: 212-239-8510 / Fax: 212-239-851 2 joseph.r.fi'agolaQcpmx.salc.com on numerous occasions to facilitate this effort. As the sponsoring committee within the IEEE, NPEC affirms NRC endorsement of IEEE sc-e, Safety Related Sy5tem M. S. Zar Standard 603 , "Criteria for Safety Systems for Nuclear Power Generating Sargent & Lundy, 23rd. Floor 55 Eest Monroe Street Stations," for new nuclear power generating facilities . Chicago, IL 60603 Vox: 312-269-2222 / Fax: 312-269-2198 mark.zarQslchlcago.lnfonelcom The proposed rule and direct final rule, as published on October 17, SC-7, Human Factors & Control Facllltles 1997, on pages 53975 and 53976 and pages 53932 through 53935 of the
- s. Malcolm Federal Register, change 10CFR50.55a by invoking the 1991 revision of AECL 2251 Speakman Drive IEEE 603. The IEEE has worked for several years to update this l\llsslssauga, Ontario L5K 1B2 Canada standard. The revision incorporates user comments especially in the Vox: 905-823-90 0x3157 / Fax: 905-823-8006 malcolmsQaecl.ca area of electromagnetic interference (EMI) and radio-frequency SC-8, Quality Maintenance & Improvement interference (RFI). The revision incorporates input from NRC staff LP. Gradln members. The revision has been balloted within the IEEE and QUAL-TEC, Inc.
127 Cabot Street consensus has been reached on the revised wording. The revision has West Bab)'lon, NY 1 1 70 Vox: 516--420-1060 / Fax: 5 1 6 19 been sent for final approval to the IEEE Standards Board . The IEEE l.gradinQleee.org NPEC understands that the NRC staff will consider endorsing this new Standards Coordinator revision when it becomes available. M. S. Zar (see above) Technical Sessions Coordinator At the November 12, 1997, meeting of the IEEE NPEC there was much J. T. Keiper The Foxboro Company discussion of the impact that the proposed rule will have on existing MIS B51-1C plants. NPEC members participating in th is meeting represented a cross 33 Commercial Street Foxboro, MA 02035 section of the industry, including licensees, NRC staff, reactor vendors, Vox: 508~9-6332 / Fax: 508~ 9-6580 JkeiperQ~xboro.com architect engineers, and consultants. As a result of these discussions it Awards Chair was clear to NPEC members that the wording of the rule should be D. F. Brosnan revised to more accurately reflect the intent of the NRC staff. The Paclftc Gas & Bectric Company 25 Market st, Room 725-N9B following comments place the concerns of NPEC on the record and San Francisco, CA 9 105 Vox: 1 5-973-8657 / Fax: 15-973-9061 suggest changes in the proposed rule to alleviate these concerns. dlbQpge.com - 4 \997 Acknowledged by Clld---....... IEEE THE INSTITUT E OF ELECTRICAL AND ELECTRONICS ENGINEERS, Inc.
U.S. NUCLEAR REGUlATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSJON OocumentSlatllllcl Postmark Date t/ruJdelivereJ 6y 5a.f(5 h 4j6(r1Ai~I tJn 1r;./,/q~. pf).Jfmrcrt~d 1~/ /'11 Copies Received _ _I______........._ Add'I Copies Reproduced _3--.____ Special Distribution A- a. n,.u...l D tD.S
November 29, 1997 Page 2
- 1. Defining "protection system" as synonymous with "safety systems" and "safety-related systems" in new paragraph 10CFR50.55a(h)(2)(i), definitions, is a significant increase in scope of the rule for existing plants. Both IEEE 279 and IEEE 603 clearly define protection system as a subset of the safety system . Figure 2 of IEEE 603 shows this relationship and includes a note which states,
'The protection system of IEEE 279-1971 (withdrawn) and of this standard is the sense and command features for the reactor trip system and the engineered safety features ."
Execute features, power supplies, auxiliary supporting features, and other auxiliary features become part of the protection system under the proposed rule . Therefore, when 10CFR50.55a(h)(3) states that protection systems must meet the requirements set forth in either IEEE 279 or 603, the requirements of either IEEE 279 or 603 are imposed on the execute features , power supplies, auxiliary supporting features, and other auxiliary features in existing plants even if design changes are not made. The IEEE NPEC judges that this expansion in scope will lead to additional cost without a material increase in the availability or reliability of the safety systems. Therefore, it is recommended that "protection system" be deleted from paragraph 10CFR50.55a(h)(2)(i) and that a new paragraph be added defining "protection system" consistent with IEEE 603. Note: For clarity, we have used the definitions contained in IEEE 603 in the comments which follow in lieu of equating protection system with safety system.
- 2. The IEEE NPEC understands that the intent of the proposed rule is for system-level replacements. The statement of considerations published in the Federal Register states that in-kind (like-for-like) replacements of safety system components are not considered changes to the safety system for the purposes of this rule . However, neither the rule nor the Federal Register discuss the modifications made in existing plants which fall between system-level replacements and in-kind replacements. The rule should be modified to explicitly state that IEEE 603 only applies to system-level replacements and additions of new safety systems.
One possible way to both clarify the definition of safety system and clarify the type of modifications to which the rule applies is to delete the last sentence of paragraph 10CFR50.55a(h)3, "protection systems, " and add the following to paragraph 10CFR50.55a(h)4, "safety systems":
"For nuclear power plants with construction permits issued after January 1, 1971, but prior to January 1, 1998, system-level replacements of safety systems and additions of new safety systems initiated after January 1, 1998, must meet the requirements set forth in IEEE Std . 603-1991 , and the correction sheet dated January 30, 1995."
Based on the discussions at the NPEC meeting, a few examples assist in clearly defining a system-level replacement to which the rule is intended to apply:
- The replacement of a reactor protection system transmitter with a transmitter manufactured by a different vendor must meet IEEE 279 since it is a like-for-like replacement.
- The replacement of the high pressure coolant injection (HPCI) system speed controls with new digital equipment must meet IEEE 279 since the modification only replaces a portion of the HPCI system .
- The upgrade of the average power range monitor (APRM) portion of the neutron monitoring system to add the ability to detect and suppress potential BWR reactor
November 29, 1997 Page 3 instability must meet IEEE 279 since the modification only replaces the APRM signal processing components, output relays, recirculation flow transmitters, and operator displays. If this modification were to also replace the neutron detectors, local power range monitor (LPRM) cards, and associated power supplies, the modification would be considered a complete replacement and must meet IEEE 603.
- The replacement of the source range monitors (SRM) and intermediate range monitors (IRM) in a BWR with wide range neutron monitors (WRNM) must meet IEEE 603 since it involves the complete replacement of the system, including sensors, preamps, signal processors, output relays , and operator displays. Reuse of a few components (e.g.,
selected cables, raceway, and control room panels where the displays are mounted) would still place this type of modification in the category of a complete replacement.
- 3. The complete set of requirements in IEEE 603 includes the criteria contained in referenced standards. This set of requirements was approved by the IEEE by a consensus process.
However, the statement of considerations published in the Federal Register change this set of requirements by stating :
"Section 3 of IEEE Std . 603-1991 references several industry codes and standards. If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices. "
Based on the discussion at the NPEC meeting, we understand that this statement in the Federal Register means that a standard referenced in IEEE 603 but not endorsed by a regulatory guide is not a mandatory requirement even if IEEE 603 invokes the referenced standard by the use of "shall. " It also means that an NRC regulatory guide which endorses a previous revision of a standard is the current NRC recommended practice even if IEEE 603 invokes the current revision of the standard by the use of "shall." The IEEE NPEC believes that the lack of endorsement of referenced standards creates an incomplete set of requirements. Therefore, it is suggested that endorsement of IEEE 603 by the proposed rule be interpreted as endorsement of all standards referenced in IEEE 603.
- 4. IEEE 603 was written for application to new designs; there was no intent by the IEEE to backfit any revised requirements onto existing plants. However, the proposed rule makes IEEE 603 a requirement for changes in existing nuclear power generating stations. Some licensees have commitments in their safety analysis report to various design criteria and requirements which are identified as major sections of IEEE 603 (e.g., independence and electrical separation, capability for test and calibration, bypass and inoperative status indication, setpoints, electrical power sources, etc.). Some of these commitments were made prior to publication of some regulatory guides. Some licensees committed to previous revisions of IEEE standards and NRC regulatory guides. Furthermore, some of the commitments in the safety analysis reports provide alternate approaches to meeting regulatory requirements.
Although portions of these commitments may differ from current regulatory guides and may not be all-inclusive when compared to IEEE 603 and current regulatory guides, they constitute an acceptable design and licensing basis for modifications to the facility . Therefore, it is recommended that the following words be added to the rule :
"For plants licensed prior to January 1, 1998, commitments in the safety analysis report for a facility to the topics covered by IEEE 603 also constitute a method acceptable to the Commission for meeting these regulatory requirements."
November 29, 1997 Page4
- 5. The IEEE NPEC understands that the draft rule does not apply to plants with construction permits issued prior to January 1, 1971. Therefore, the requirements of IEEE 603 do not apply to changes in plants with a construction permit issued prior to January 1, 1971, even if the licensee previously committed to applying IEEE 279 to portions of the safety systems.
- 6. Based on the discussion at the NPEC meeting, the members of NPEC recommend adding guidance in section 7 of the standard review plan (SRP) and in appropriate NRC inspection guides to ensure consistent interpretation of the revised rule. It is suggested that this guidance include:
- Examples of the types of modifications which must meet IEEE 603 vs. IEEE 279.
- Guidance on application of standards referenced in IEEE 603.
- Guidance on identifying previous licensee commitments which constitute an acceptable method of meeting the requirements of IEEE 603.
Please contact me at 717-456-3581 if there are any questions concerning these comments. Sincerely,
- ~~~
Chairman, IEEE NPEC
~ Duke r,Power. . DOCKETED Duke Power Company A Duk, Energy Company EC07H A Dult Entrgy Company 526 South Church Street USNRC P.O. Box 1006 Charlotte, NC 28201-1006 M. S. Tuckman Executive Vice President "97 0£C -1 AlO :04 (704) 382-2200 OFFICE Nuclear Generation (704) 382-4360 FAX November 2 6, 1997 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Attention: Rulemakings and Adjudications Staff
Subject:
Proposed Rule To Amend 10 CFR 50.55 a(h) To Incorporate IEEE Standard 603-1991 In response to the proposed rulemaking to amend 10 CFR 50.55 a(h) to incorporate IEEE Standard 603-1991 which appeared in the October 17, 1997 Federal Register, Duke Energy Corporation offers the following comments. Comment 1 The Commission states at Federal Register (FR) 53933, "The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same. Therefore, for purposes of paragraph (h) of 10 CFR 50.55a, "protection systems" and "safety systems" are synonymous." We disagree that "protection systems" and "safety systems" are synonymous. Both IEEE Std. 279 and 603 define protection system as a subset of the safety system. Figure 2 of IEEE Std. 603 shows this relationship and includes a note which states:
"The protection system of IEEE Std. 279-1971 (withdrawn) and of this standard is the sense and command features for the reactor trip system and the engineered safety features."
Additionally, IEEE Std. 603-1991 expands the scope to include "auxiliary supporting features" and "other auxiliary features." For example, IEEE Std. 603-1991, Figure 3, identifies HVAC as an "auxiliary supporting feature." HVAC, and other auxiliary support systems, at some older plants are non-safety. Therefore, defining "protection system" as synonymous with "safety system" and including auxiliary supporting features represent a significant increase in scope of the rule for existing plants. OEC * \S97 AcknoWledged by card---- 4
-I U.S. NUCLEAR REGULATORY COMM!
RULEMAKINGS & ADJUDICATIO c OFFICE OF THE SFCP OFTHECOl11 1 Document Postmarl< Date ~and 1).e./,*,;qrqd ~Y S-i.f;~J, Cjjfir ~4./ bn 1:i/, / 17 Coples Received _/ _ Add'I Copies ReprodL* * .3 Special Distnbution_ '4-15~,:w a.l
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Comment 2 IEEE Std. 603-1991 requires compliance with many other IEEE standards that are not required by IEEE Std. 279-1971. Meeting these additional requirements can cause significant impact. For example, IEEE Std. 603-1991 requires separation of Class lE equipment to meet the requirements of IEEE Std. 384-1981. These requirements differ significantly with the existing licensing basis of many plants. To require cabling systems and internal wiring of cabinets affected by a modification to be different from all circuits not affected by the modification is impractical, could cause significant increases in modification costs, would not necessarily provide any commensurate increase in safety, and would significantly complicate plant work processes since separation criteria for modifications would differ from those for maintenance. Comment 3 The existing rule does not backfit IEEE Std. 279 on plants with construction permits issued prior to January 1, 1971. The Commission states at FR 53933, "However, changes to protection systems in operating nuclear power plants initiated on or after January 1, 1998 must meet the requirements in IEEE Std. 603-1991. II We consider the proposed rule to represent a backfit issue for plants with construction permits issued before January 1, 1971, in that any change/enhancement, other than like-for-like component replacements, to applicable systems could be interpreted to require upgrading affected portions of the existing system to completely comply with the design criteria (single failure, separation/isolation, etc.) of IEEE 603-1991. This could have the unintended effect of discouraging or prohibiting licensees from upgrading their systems. Comment 4 The Commission states at FR 53975, "Because NRC considers this rulemaking noncontroversial, we are publishing this proposed rule concurrently as a direct final rule." The Commission states at FR 53933 that, "Because there were no adverse public comments to Revision 1 to Regulatory Guide 1.153, the Commission believes that there is general public consensus that IEEE Std. 603-1991 2
provides acceptable criteria for safety systems in nuclear power plants. These statements indicate that the NRC equated the lack of adverse comments to Regulatory Guide 1.153, a guidance document, with tacit agreement on the part of the industry, and justification to proceed with direct rulemaking. The lack of industry response to a guidance document that does not constitute a regulatory requirement should not be interpreted as endorsement of a new rule, that if implemented as written, would constitute a significant potential backfit for some utilities. Concl.usion The proposed rule and direct final rule would result in the backfit of significant new requirements on some facilities; particularly if they elect to make changes to upgrade systems/equipment. Therefore, the impact of implementing the direct final rule on existing plants should be fully evaluated and thoroughly assessed before the rule is adopted. We do not agree with the proposed rule for existing plants, and strongly recommend existing plants be grandfathered if this rule is adopted. Very truly yours, M. S. Tuckman xc: Alex Marion Nuclear Energy Institute 1776 I Street, NW - Suite 400 Washington, D. C. 20006-3708 3
Pacific Gas and Electric Company 245 Market Street. Room 937-N9B Gregory M. Rueger San Francisco. CA 94105 Senior Vice President and l/11ili11.~. ld<ir<'s., General Manager Mail Code N9B Nuclear Power Generation P 0 Box 770000 San Francisco. CA 94177 415/973-4684 Fax 415/973-2313 November 25, 1997 OOCKtT PROPOSED so PG&E Letter DCL-97-197 ( ~:l.F~ 53'115) ( (,:2 F~ S3'/ 3:J-)
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c_c :!J ~ Mr. John C. Hoyle cO ,r-,; o C) Office of the Secretary ;:::;;7 _:~:-) ~ co ('") n U') ::x: U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 '
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Docket No. 50-275, OL-DPR-80 Docket No. 50-373, OL-DPR-82 Diablo Canyon Units 1 and 2 Comments on Codes and Standards; IEEE National Consensus Standard 62 Federal Register 53932, October 17, 1997
Dear Mr. Hoyle:
On October 17, 1997, the NRC noticed issue of a direct final rule in the Federal Register (62 FR 53932) amending 10 CFR 50.55a(h) , "Protection and Safety Systems," to incorporate by reference IEEE Standard (Std .) 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations." Pacific Gas and Electric Company's (PG&E) comments on the direct final rule are provided below. Overall, PG&E believes that the revised rule will have a significant adverse effect on licensees. The revised rule mandates that changes to protection systems in operating nuclear power plants initiated on or after January 1, 1998, must meet the requirements in IEEE Std. 603-1991 . Incorporation of this standard for all future modifications to the plant protection system at Diablo Canyon Power Plant (DCPP) will have a major impact on PG&E for the following reasons.
- 1. As with various other operating power plants, DCPP is designed and licensed to IEEE Std. 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations." To date, no safety-significant issues have been identified with existing plants licensed to this standard that would justify the added regulatory burden that will be incurred by licensees in implementing the revised regulation.
DEC - 4 1997 Acknowledged by card_ .. ,_,_ _,_,_
,o u,vi..lJ'l n REGULATORY COMMISSION M KINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Document Statlstlcl Postmark Data JJ/~'-/q7 Copies Received ____/ _ _ _ __
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John C. Hoyle November 25, 1997 Page2
- 2. The Federal Register describes the revised rule as a noncontroversial issue because the IEEE standard was endorsed by a regulatory guide with no adverse public comment. The publication of regulatory guides by the NRC has been endorsed and supported by the industry. In this regard, PG&E did not comment on the endorsement of the standard in a regulatory guide.
However, neither the standard nor the regulatory guide require mandatory application, whereas the revised rule would implement IEEE Std. 603-1991 as a requirement on licensees for applicable plant changes initiated on or after January 1, 1998. In addition, IEEE Std. 603-1991 includes reference to various other standards that must also be implemented to achieve full compliance with its provisions. Thus, the revised rule is a significantly different regulatory action that may warrant backfit considerations, particularly since implementation of the revised rule may redirect resources from more safety-significant activities. Further, during a plant's operating lifetime, a number of plant changes may be made due to equipment obsolescence issues. Such changes are obviously not voluntary. For example, when a vendor no longer supports a particular piece of equipment, licensees often have no reasonable alternative other than to find a suitable replacement, which may require replacing an entire system. Under the revised rule mandating use of IEEE Std. 603-1991, the cost of such plant changes could significantly increase with no commensurate enhancement to safety. As a consequence, rather than replacing such equipment or system when warranted, a licensee may seek to avoid incurring these excessive costs by extending the life of the equipment involved, which can be contrary to the intent of the revised rule.
- 3. IEEE Std. 603-1991 refers to numerous revisions of other IEEE standards, as well as revisions to other industry standards that are more recent than those applicable to DCPP. The effort required to fully assess the impact of the revised rule, including the referenced standards, cannot be completed before the stated effective date of the rule. One example of such a referenced standard is IEEE Std. 308-1980.
DCPP is designed and licensed to IEEE Std. 308-1971. According to the revised rule, a change after 1997 to a component within the plant protection system may require a full redesign. More specifically, to comply with IEEE Std. 603-1991, a change in a protection system component at DCPP might necessitate replacement or upgrading of the entire 12-kV and offsite power system. This modification would involve significantly greater resources than simply changing the component itself and would be prohibitively expensive.
- 4. A dual licensing basis may be created for a plant system or component once the revised rule is implemented, thereby leading to possible confusion.
John C. Hoyle November 25, 1997 Page 3
- 5. As a regulation instead of a regulatory guide, it is not clear how the phrase "to the extent feasible and practical" (Section 6.4) should be interpreted.
6 The term "acceptable reliability" (Section 8.3) lacks definition.
- 7. IEEE Std. 603-1991 states that indication shall be automatically actuated if the bypass is expected to occur more than once a year. The basis for establishment of this frequency is not clear.
- 8. The proposed rule states that safety-related systems covered by IEEE Std. 279 and IEEE Std. 603-1991 are the same. It has not been demonstrated that auxiliary supporting features in operating plants are classified as safety-grade in all cases pursuant to IEEE Std. 279. Therefore, it is not clear that the scope of these two standards are identical.
Note that the scope of IEEE Std. 603 includes all safety-related systems as interpreted by the NRC in the Federal Register notice, while the scope of IEEE Std. 279 clearly includes only the plant protection system and its supporting features (such as its power supply configuration). Codifying IEEE Std. 603 will cause the licensing basis for a significant and broad range of systems to change to IEEE Std. 603. For example, if the DCPP containment reactor cavity sump level transmitter (a Technical Specification designated post-accident monitoring instrument) were to be changed from a Barton to a Rosemount transmitter, the design calculation for the instrument uncertainty would have to be changed. Since the design calculation would be revised for the new Rosemount transmitter (the revision could also be necessitated by just changing the measured span of the parameter), the design of the level instrument has changed, and an evaluation would be required to determine whether other changes were necessary to comply with IEEE Std. 603. IEEE Std. 279 and Std. 603 can be interpreted to apply not only to changes to system hardware, but also to non-hardware design changes, of which there are many. Seemingly insignificant changes (such as changing an instrument span or changing a setpoint in the Eagle 21 plant protection system, which requires no hardware change) could be interpreted under Std. 603 to require full reevaluation to the new design basis. The resulting reevaluation would primarily be procedural, and merely increases overhead costs without contributing to plant safety. PG&E has already prepared a series of Replacement Part Evaluations to support the replacement of like-for-like parts when it becomes necessary. This body of substantial work would have to be reevaluated and modified, reviewed, and reapproved for use in the event that IEEE Std. 603 were to become law. In fact, it is probable that several like-for-like changes would no longer be possible under the Replacement Part
John C. Hoyle November 25, 1997 Page4 Evaluation Program because of the substantial supporting system design changes necessary to comply with IEEE Std. 603. The effect of the new design standard would be to make small changes become very substantial and burdensome. Given the limited time available to respond to the Federal Register notice, PG&E did not perform a comprehensive review of the revised rule. Nevertheless, PG&E believes that licensees should be able to retain their original design and licensing bases, and maintain their commitments to existing standards such as IEEE Std. 279-1971. PG&E agrees that, where practical, licensees should adopt standards that are more current. However, licensees should not be required by rulemaking to do so. IEEE Std. 603-1991 is a well-written, technically feasible standard that should be retained for new plants where it can be effectively incorporated into the design. Based on the above comments, PG&E requests that the NRC rescind the direct final rule. Alternatively, should the NRC proceed with the rule, PG&E requests that the rule be revised to delete any requirement for implementation of IEEE Std. 603-1991 by operating plants. Sincerely, c: Steven D. Bloom Alex Marion, NEI Ellis W. Merschoff Kenneth E. Perkins David L. Proulx Diablo Distribution dwo/220
From: Adria Byrdsong <atbl@nrc.gov> To: WND2.WNP6(JLK) Date: 12/5/97 8:45am
Subject:
Note: Comment forwarded to Rulemakings and Adjudications Staff HyperNews notification. Reply via: http://itm25.ore.gov: 88/HyperNews/get/EELB/Distribution/6/ l / l .html From: Adria Byrdsong To: JLK Date: 12/4/97 1:32pm
Subject:
Proposed and Direct Final Rules re IEEE Std. 603-1991 John: On December 1, we received your correspondence dated November 25 regarding the Proposed and Direct Final Rules entitled, "Codes and Standards; IEEE National Consensus Standard." We forwarded your correspondence to the appropriate staff for consideration. Thank you for your interest. Adria Byrdsong Rulemakings and Adjudications Staff Office of the Secretat*y
J KET PROPOSED fU.E~~--o-mo ( &, 'JJ~ '53'115) From: Admin <jlk@nrc.gov> ( "1 ~ F~ 53'13 :J.) DOCKETED To: WND2.WNP6(JLK) US RC Date: 11/25/97 9:56am M:DEe. I
Subject:
Note: Request for feedback for making electri497PollJ¥ 3~st~~~58 synonymous with protection systems by Regulatory Rule HyperNews notification. Reply via: OFFIC[ ( F Ecq, f\RY http:/ /irm25. nrc. gov: 88/HyperNews/get/EELB/Distribution ,R;IJh. ~html!t,JG, /., 'O ADJUDICA i 10NS S AFF Comments/questions/concerns are requested relating to the following proposed comment being developed by EELB. The proposed comment relates to the Federal Register notification of NRC's plan to amend paragraph (h) of 10 CFR 50.55a to replace IEEE 279 with IEEE 603. The letter from the EDO to the Commission dated September 8, 1997 states the following:
"The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same. Therefore, for purposes of paragraph (h) of 10 CFR 50.55a, "protection systems," and "safety systems" are synonymous. The Commission notes that these two terms are also synonymous with the term "safety-related systems," used elsewhere in Commission's regulations. Therefore, licensees are expected to apply IEEE Std. 279-1971 and IEEE Std. 603-1991, as appropriate, to "safety-related systems."
It is agreed that systems covered by 603 and 279 are the same; but, what is not clear is how the Commission, for purposes of paragraph (h) of 10 CFR 50.55a, therefore concludes that "protection systems" and "safety systems" are synonymous. Protection systems, as defined by 603, are consider to be one of many safety systems. One of many safety systems, by definition, cannot be synonymous with all (the many) safety systems; thus, protection systems are not synonymous with safety systems. - The onsite electric power system, the component cooling water system, the residual heat removal system, and many other mechanical and electrical systems, in addition to the protection system, are defined as safety systems by IEEE 603. These mechanical and electrical systems, as specified by 603, are not required to meet the guidelines of IEEE 603. However, by inference, if it is stated as part of NRC regulations that protection systems are synonymous with safety systems, then these many other mechanical and electrical safety systems can also be considered synonymous with protection systems and be required to meet 603 requirements; but, since these other mechanical and electrical systems are outside the scope of 603 and have no stated requirements in the standard, the commission (by defining the existence of a synonymous relationship between safety systems and protection systems) has added an unnecessary level of confusion to our regulations and perpetuates a dichotomy between the NRC staff and IEEE. It is recommended that the proposed rule be revised to remove the synonymous relationship between protection systems and safety systems so that NRC's definitions of safety systems will be the same as IEEE's. E*Mittl DEC - 4 1997 Acknowledged by
U.S. NUCLEAR REGULATORY COMMISSION flJLEMAKINGS & ADJUDICATIONS STAFf OFFICEOFTHES 1 OF THE CO ISS
- J DocumP ~
Postmark Date Hl4~1 ])el,*v,red 6y_5t;.f,'si A95qrJ.Jt./ on/:,./, /c,1 Coples Received ---=-'----- Add'! Copies Rep1 ~u 3 _ ~ I Oistributi )1 ~, r w ~ f, Pi '1:fl.PS.
-===- ENTERGY DOCKETED USHRC Ente,gy Ope,at;ons, Inc. ( } ) PO Box 31995 Jackso11. MS 39286-1995 1el 601 368 5760 Fax 6013685768 "97 NOV 28 P2 :Q9 Jerrold G. Dewease Vice Pres1de111 Opcratio11s Sr 1ppnrt November 25, 1997 Mr. John C. Hoyle Secretary, U.S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 ATTN.: Rulemakings and Adjudications Staff
Subject:
Comments on Direct Final Rule amending 10CFR50.55a(h), Protection and Safety Systems, to incorporate by reference IEEE std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations"
Reference:
Federal Register Volume 62, Page 53932, dated October 17, 1997 CNRO-97/00022
Dear Mr. Hoyle:
Attached please find Entergy Operations, Inc. (Entergy) comments on the above captioned matter. We agree in principal that the NRC should consid er adopting industry standards vice de nova agency development of standards affecting the industry. The direct final rule mandates the use of IEEE std. 603-1991 for changes to protective systems in operating nuclear power plants initiated after January 1, 1998 except for like-for-like replacements. Entergy is concerned that the direct manner in which the NRC is promulgating the final rule short circuits the comment process and introduces issues not clearly explained in the Statement of Considerations. The NRC's basis for the direct final rule is the disingenuous conclusion that lack of comments on the Draft Regulatory Guide 1.153 connotes public agreement for the promulgation of the final rule. Regulatory Guides establish "an acceptable way" (e.g., optional way ) of meeting the Commission's regulations, while a rulemaking rigidly sets the standards. In issuing this direct final rule, the NRC has inappropriately shifted the burden to the public (i.e., the licensees) to prove the rule is not justified. The numerous changes incorporated by the new standard constitute a significant backfit which has not been properly subjected to the requirements of 10CFR50.109. DEC - 4 \991_ Acknowledged by card II IE I $b$ 0 ;o&,: 6W
.I U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Docwnent Statlsllcs Postmark Date // / :J t, / CJ 1 Copies Received _ __:... I _ _ __
Add'I Copies Reproduced - =--.--- S11ecial Oistribution~ a ~ :..L.!-,.;:::.:..;.=, , __ IDS - -(- -*
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 2 of 3 A plant's licensing basis establishes a set of legally binding requirements the plant must meet to satisfy the commission's regulations. It also establishes reasonable assurance that the facility operation will not endanger the health and safety of the public. The referenced standard is more detailed than the standards to which most plants were originally licensed (IEEE-279) and references other documents that have not been endorsed by the Commission in a Regulatory Guide. This introduces confusion about how to interpret the new standard. Evaluation and tracking of these additional requirements for a modification clearly imposes a new backfit burden on licensees. The required §50.59 review for planned modifications determines whether or not an unreviewed safety question (USQ) is being introduced by the planned modification. Any modification involving a USQ requires prior NRC approval. Imposing the direct final rule circumvents the §50.59 process by removing the continued acceptability for changes that fall within the original licensing basis. There has been nothing to suggest that compliance with IEEE-279 is technically unacceptable. For this reason, the NRC should permit licensees the option to continue to meet their current licensing basis for modifications after January 1 , 1998, or meet the later standards. Entergy has worked with the Nuclear Energy Institute (NEI) to develop industry comments. Entergy endorses NEl's letter on this subject. In addition, we have reviewed and concur with Winston & Strawn comments on this subject prepared in behalf of the Nuclear Utility Backfitting and Reform Group (NU BARG), of which Entergy is a member. Our additional technical comments are attached. If there are any questions regarding these comments, please contact Les England at (601 )-368-5766. Very truly yours, ~!rfJ&o attachments cc: (see next page) _ __J
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 3 of 3 cc: Mr. C. M. Dugger (W-GSB-300) Mr. J. J. Hagan (G-ESC3-VPO) Mr. C. R. Hutchinson (N-GSB) Mr. J. R. McGaha (R-GSB-40) Mr. J. W. Yelverton (M-ECH-65) Mr. Jack N. Donohew Project Manager (GGNS) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13-H-3 Washington, DC 20555 Mr. George Kalman Project Manager (ANO) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13-H-3 Washington, DC 20555 Mr. Chandu P. Patel Project Manager (W-3) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13-H-3 Washington, DC 20555 Mr. David L. Wigginton Project Manager (RBS) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13-H-.3 Washington, DC 20555
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 1 of 4 COMMENTS ON DIRECT FINAL RULE AMENDING 10CFR50.55a(h), Protection and Safety Systems
- 1) None of Entergy's nuclear units are presently committed to IEEE 603-1991.
Each site would be affected by this change since it is applicable to all future modifications. This constitutes a significant backfit which has not been subjected to the requirements of 10CFR50.109.
- 2) The direct rulemaking has been proposed due to the perceived lack of comment on the issuance of IEEE 603-1991 and Regulatory Guide 1.153 Rev. 1. It is reasonable to assume that comments were not made because no plants are committed to nor were required to meet these publications. In addition, the author of the Federal Register posting, Mr. Satish Agerwal, was informally provided immediate negative comments at the IEEE NPEC 97-1 meeting when he announced the NRC's intent to publish this item.
- 3) The discussion in the Federal Register posting states that Section 3 of IEEE Std.
603-1991 references several industry codes and standards. This posting further states that "If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If the referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with regulatory practices." Regulatory Guide 1.153, Rev. 1 currently endorses IEEE 603-1991 as" ... a method acceptable to the NRC staff for satisfying the Commission's regulations with respect to the design, reliability, qualification, and testability of the power, instrumentation, and control portions of the safety systems of nuclear power plants." The actual text of IEEE 603-1991 Section 3 states, "This standard shall be used in conjunction with the following publications." In addition, several sections throughout IEEE 603-1991 invoke these latest standards, individually (see Attachment 2 for examples). This list of standards and publications is beyond the current license basis of Entergy's nuclear units. One example of impact is the invocation of IEEE 323-1983 and IEEE 627-1980 in the section related to Equipment Qualification. These standards require environmental qualification of mechanical components. The programs at Entergy's sites are presently limited to electrical components.
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 2 of 4 In addition, Regulatory Guide 1.1.53 Rev. 0 includes statements supporting the relief implied in the Federal Register posting. However, no words supporting the relief are included in Revision 1 of this Regulatory Guide.
- 4) The Discussion section of the Federal Register listing includes a position that the NRC views the systems covered by IEEE 603-1991 and IEEE 279-1971 to be the same. It goes on to state that "... for the purposes of paragraph (h) of 10CFR50.55a, 'protection systems,' and 'safety systems' are synonymous."
Entergy disagrees with this position. IEEE 279-1971 defines protection systems carefully to encompass all electric and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating those signals associated with the protective function. These signals are those that actuate reactor trip and that, in the event of a serious reactor accident, actuate engineered safeguards such as containment isolation, core spray, safety injection, pressure reduction, and air cleaning. IEEE 279-1971, Section 4.7 clarifies the control and protection system interaction and states, "Any equipment that is used for both protective and control functions shall be classified as part of the protection system and shall meet all the requirements of this document." This is clearly distinctive from the definition found in IEEE 603-1991 for "safety system" which includes all safety related systems regardless of whether they generate signals associated with protective functions or actuate engineered safeguards. The Forward of IEEE 603-1991, under the heading of "Evolution", states in part, "The series began with IEEE Std. 279-1968, a trial-use protection systems standard, followed by IEEE Std. 279-1971, a full protection system standard; IEEE Std. 603-1977, a trial-use standard for safety systems; and IEEE Std. 603-1980, a full safety system standard [emphasis added)." It appears from this discussion the IEEE committee recognizes the difference in scope between the two standards. The NRC's correlation between "protection system" and "safety system" is unsupported by the words of the IEEE committee.
- 5) Appendices to IEEE 603-1991 are specifically listed as not being part of the standard but are included for information only. IEEE standards are also preceded with the statement, "Use of an IEEE Standard is wholly voluntary." Due to inclusion of the entire standard in 10CFR50.55a, it is unclear if the rulemaking also includes the appendices in its scope.
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 3 of 4
- 6) Various sections of Entergy's UFSARs discuss compliance of systems with specific criteria contained in IEEE 279-1971. Revision and maintenance of these UFSAR sections to delineate the portion of systems that comply with the earlier criteria versus the portion that comply with the later criteria would be overly burdensome. Also, no clear guidance is given to assist in determining the bounds of the modified system; i.e., if a transmitter is replaced, what portions of the instrument loop are required to meet the new standards?
- 7) The posting in the Federal Register is unclear about the definition of the word "initiated" in regard to the statement "... changes to protection systems in operating nuclear power plants initiated on or after January 1, 1998 must meet the requirements in IEEE 603-1991." Does the term "initiated" refer to the design phase or implementation phase? Also, would this be applicable to changes implemented prior to January 1, 1998 but not completely closed in regard to paperwork?
- 8) The Background section states, in part; "However, IEEE 279 is obsolete, has been withdrawn by IEEE and has now been superseded by IEEE Std. 603-1991,
'Criteria for Safety Systems for Nuclear Power Generating Stations."' No information has been found in either IEEE Std. 279-1971 or IEEE Std. 603-1991 which states that IEEE Std. 279 is considered either obsolete or superseded by IEEE Std. 603-1991.
- 9) For existing licensees, only "protection systems" are currently regulated under 10CFR50.55a(h). The proposed rule expands this scope to include all "safety systems", which is well beyond the scope of the existing rule. This expansion clearly represents a backfit. The proposed rule changes the licensing basis of existing plants from application to "protection systems" to application to "safety systems". The Commission backfit analysis should reflect this expanded scope.
- 10) The Backfit Analysis section of the proposed rulemaking states, "... this would not be considered a backfit, since the changes are voluntarily initiated by the licensee ... " In the case of obsolescence of components, the modification that a plant is required to make to a protection system is involuntary. As such, this rulemaking applies a change to the current licensing basis for involuntary changes and is, therefore, a backfit.
- 11) The existing 10CFR50.55a(h) states in part, "For construction permits issued after January 1, 1971, ... " indicating that the rule is only applicable to plants with a construction permit issued after January 1, 1971. The proposed rulemaking recognizes this, but still requires plants with construction permits issued before
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 4 of 4 January 1, 1971 to meet the requirements of the new rule for changes to safety systems initiated after January 1, 1998. This constitutes a change to the licensing basis for those plants not currently required to satisfy the requirements of IEEE Std. 279-1971 and, therefore, a backfit.
- 12) Many of the existing Regulatory Guides reference IEEE Std. 279-1971 and its codification in 10CFR50.55a(h). With removal of IEEE Std. 279-1971 via this rulemaking, how are existing Regulatory Guides affected? Will each of the Regulatory Guides that reference IEEE Std. 279-1971 be revised to reference IEEE 603-1991?
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 1 of 2 EXAMPLES OF NEW STANDARDS REFERENCED IN IEEE 603-1991
- 1. IEEE 603-1991 Section 4 states, "The design basis shall be consistent with the requirements of ANSI/ANS 51.1-1983 [14] or ANSI/ANS 52.1-1983 [15] and shall document as a minimum: ... " These documents were written after most plants were licensed.
- 2. IEEE 603-1991 Section 5.3 states, "Safety system equipment shall be designed, manufactured, inspected, installed, tested, operated, and maintained in accordance with a prescribed quality assurance program (ANSI/ASME NQA1-1989 [16])." This document was written after most plants were licensed.
- 3. IEEE 603-1991 Section 5.4 states, "Qualification of Class 1E equipment shall be in accordance with the requirements of IEEE Std 323-1983 and IEEE Std 627-1980[11 ]." This specifically is in direct contrast with Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants." This Regulatory Guide specifically calls out IEEE 323-1974 as the acceptable method of qualifying Class 1E equipment. Also, 10CFR50.49 references IEEE 323-1974 in paragraph (b)(1)footnote #3.
- 4. IEEE 603-1991 Section 5.6.3.2(1) states, "The separation of Class 1E equipment shall be in accordance with the requirements of IEEE Std 384-1981 [6] [B3]."
This document was written after most plants were licensed.
- 5. IEEE 603-1991 Section 5.7 states, "Testing of Class 1E systems shall be in accordance with the requirements of IEEE Std 338-1987 [3]." This document was written after most plants were licensed.
- 6. IEEE 603-1991 Section 5.8.1 states, "The display instrumentation provided for manually controlled actions for which no automatic control is provided and that are required for the safety systems to accomplish their safety functions shall be part of the safety systems and shall meet the requirements of IEEE Std 497-1981[9]."
This document was written after most plants were licensed.
Comments on Direct Final Rule amending 10CFR50.55a(h) November 25, 1997 CNRO-97/00022 Page 2 of 2
- 7. IEEE 603-1991 Sections 5.11, 5.14, 5.15 state similar requirements to meet IEEE 384-1981, IEEE 420-1982, and IEEE 1023-1988. These documents were written after most plants were licensed.
- 8. IEEE 603-1991 Section 6.8.1 states, "The allowance for uncertainties between the process analytical limit documented in Section 4.4 and the device setpoint shall be determined using a documented methodology. Refer to ISA S67.040-1987[18 This document was written after most plants were licensed.
- 9. IEEE 603-1991 Section 8.1 states, "Specific criteria unique to the Class 1E power systems are given in IEEE Std 308-1980[1]." This document was written after most plants were licensed.
Westinghouse Energy Systems DOCKETED USHRC
- NSD-NRC-97-5453 Nuclear Services Division Electric Corporation "97 NOV 28 Pl2 :36 PO Box 355 Pittsburgh Pennsylvania 15230-0355 OFFICE o..: F,* + 'Afft RULEM,\l,lf'l ::; ,,: J November 21, 1997 ADJUOiCP,TI( ~._,,S 3TAFF
~ecretary, U.S. Nuclear Regulatory Commission DOCKET NlllEfill ATTN: Rulemakings and Adjudications Staff PROPOSED Ill f _ 50 Washington, DC 20555 ( ~~F~ ~3915)
( 1,2~/l 53 93:L)
Subject:
Comments on NRC Proposed Rule Change to Incorporate IEEE 603-1991, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations" into 10CFR 50.55a. These comments are submitted by Westinghouse Electric Corporation ("Westinghouse") in response to the United States Nuclear Regulatory Commission request for public comments on the proposed rule change to incorporate IEEE 603-1991 into 10CFR 50.55a. Westinghouse supports the effort of the NRC to review and endorse the latest revisions of IEEE standards into the rules and regulations. Realizing that it is the intent of the NRC to encourage plant modifications that will enhance safety, the following comments and suggested rule modifications are oriented toward reducing the confusion that may result from applying more recent standards to operating plant modifications.
- 1. Since it appears that one of the primary applications of IEEE 603 will be with regard to digital replacement systems, it would be advisable to wait for the proposed revision to IEEE 603-1991. The new version includes stronger and more complete references to IEEE 7-4.3 .2 to cover diversity, defense-in-depth, EMI/RFI, etc.
- 2. The new rule should clarify the application of IEEE 603, i.e., which plants, which modifications. It is suggested that the following changes be incorporated into the new rule:
a) Section 50.55a (h) (2) (ii) - Changes to protection systems include major system level modification§., augmentation§. or replacement of protection systems permitted by license amendments or made by licensees pursuant ..... b) Section 50.55a (h) (3) - ....... However, for nuclear ower lants with construction ermits issued after Januar 1 1971 ma*or s stem level ch-anges to protection systems initiated on or after ....... . ..
- 3. IEEE 603-1991 references many other standards and states that they shall be used in conjunction with IEEE 603. To avoid future confusion it is recommended that the new rule include the intent of the NRC which as expressed in the discussion section is: If the standard referenced by IEEE 603 has been endorsed in a regulatory guide, the standard constitutes an acceptable method of meeting a regulatory requirement as described in the regulatory guide. If OEC -4 l99L.
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U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUOICATIONB STAR= OFFICE OF THE SECRETARY OF THE COMMISSION Docllnant StatiBtlcl Postmark Data II &'Il9 1 Coples Received _ _,_ _ __ Add'I C(¥Jie8 Reproduced ~ ---- Special Dis1ribution 1JJq('wal
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NSD-NRC-97-5453 2 November 21, 1997 a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices.
- 4. Equating protection systems and safety systems in 50.55a (h) (2) may tend to be confusing for operating plants. The intent of the rule is not to require additional review unless major system level modifications are implemented. The separation of the two definitions as exhibited in 50.55a (h) (3) and (4) is more appropriate.
Westinghouse appreciates the opportunity to provide these comments. Should you wish to discuss our comments in greater detail, please contact me at (412) 374-5282. Very truly yours, ~112.g& Regulatory and Licensing Engineering cc: Satish K. Aggarwal, Senior Program Manager Office of NRR 2583C-RBM-2:l 12197
DOCKETED [IE "97 NOV 28 P2 :Q8 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402-2801 Gr' ,-*;:.' .. *-*
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November 25, 1997 DOCKET PROPOSED RULE-......ii...:=;..o_ _, Mr. John c. Hoyle, Secretary ( &,!J. f P..£397E) U.S. Nuclear Regulatory Commission ( 1Pa t=~S3'j3~) ATTN: Rulemaking and Adjudications Staff Washington, D.C. 20555-0001
Dear Mr. Hoyle:
NRC - REQUEST FOR ADVERSE COMMENTS ON DIRECT FINAL RULEMAKING AND PROPOSED RULEMAKING - CODES AND STANDARDS; INSTITUTE OF ELECTRICAL AND ELECTRONIC ENGINEERS (IEEE) NATIONAL CONSENSUS STANDARD TVA is pleased to provide comments on the direct final rulemaking published at 62 Federal Register 53932 (October 17, 1997), and the associated proposed rulemaking published at 62 Federal Register 53975 (October 17, 1997). As stated in the summary section of the direct final rule, NRC is amending its regulations to incorporate, by reference, IEEE Standard 603-1991, a national consensus standard for power, instrumentation, and control portions of safety systems in nuclear power plants. NRC identified this action as necessary to endorse the latest version of this national consensus standard in NRC's regulations, and replace an IEEE standard currently endorsed in NRC's regulations which has been withdrawn by the IEEE. TVA does not believe that the direct final rule should be adopted for the following reasons:
- The language in the rule which invokes IEEE 603-1991 is ambiguous. The scope of the rule can be interpreted to apply to any modification (including piece part replacement) to a safety system made after January 1, 1998.
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U.S. NUCLEAR REGULATORY COUMISSlON RULEMAK1NGS &ADJUDICATIONS STAff OFFICE OF THE SECRETARY OF THE COMMISSION DocumentStatilllcl Postmartd>ate las/ JII I q1 Coples Received _ _.:... , ---- Add'I Coples Reproduced ~ 3 _ ___ _ Special DlstribUtion t,J a.
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Mr. John c. Hoyle Page 2 November 25, 1997
- The language in the rule which invokes 603-1991 is confusing. The referenced IEEE standard invokes other standards by reference. It is not clear whether the standards addressed by reference in IEEE 603-1991 are also requirements of the rule.
- Adoption of this rule would create confusion in the design and licensing bases for equipment as it is modified. The short implementation period is not sufficient to properly modify design procedures and train employees on the new rules. The lack of corresponding support information, such as Regulatory Guides to clarify the ambiguous and confusing language in the rule, compounds the problem.
TVA recommends that the proposed rule be modified as follows:
- The modified rule should be applicable only to those plants with a construction permit issued after the initial issue of IEEE 603 in 1977.
- The effective date of the modified rule should be changed to allow at least one year for the development of the revised design procedures and training of employees.
- The modified rule should apply only to design changes initiated after the effective date to avoid confusion over design requirements for design work in progress.
- The modified rule should clearly state that it applies only to system level replacements of the protection system {i.e., reactor trip system and engineered safety feature actuation system) or the addition of new safety systems.
Mr. John c. Hoyle Page 3 November 25, 1997
- The modified rule should clearly state that additional standards referenced in IEEE 603-1991 are not requirements of the rule. Instead, the rule should clearly state that commitments in the safety analysis report for the facility to the topics covered by IEEE 603 also constitute acceptable methods of meeting the requirements.
- Corresponding changes to the Standard Review Plan and/or Regulatory Guides should be made and issued concurrent with any rule change to provide additional guidance on the types of modifications that fall within the scope of the rule. Guidance should also be provided on the alternatives for addressing the topics covered by referenced standards in IEEE 603-1991.
TVA believes that the changes recommended above are consistent with the purpose of the proposed rule as stated by NRC's representative at the November 12, 1997, Nuclear Power Engineering Committee meeting. TVA also believes that the direct final rule should be replaced with a modified rule that resolves the ambiguity and confusion that exists within the language of the direct final rule. Without these changes, TVA is concerned that alternate interpretations could be made in the areas that are ambiguous and confusing. Questions about the interpretation of the rule would require additional resource expenditures to achieve resolution. Interpretations that extend the scope of the rule to any modification of safety-related systems would result in significant cost impacts to TVA. Similarly, any interpretations of the referenced standards in IEEE 603-1991 would result in significant cost impacts to TVA. Additionally, changes to existing analysis resulting from this new rule would also lead to significant cost impacts to TVA. As such, TVA would view these interpretations as backfits to the current licensing basis. TVA believes the best way to avoid these conflicts is to revise the direct final rule to clarify these points.
Mr. John c. Hoyle Page 4 November 25, 1997 We appreciate the opportunity to respond to the subject direct final rule. If you have any questions regarding TVA's comments, please contact me at (423) 751-2508. Sincerely, "/IUl_/~- MarK'iJ. Burzynski Manager Nuclear Licensing cc: Mr. R. W. Hernan, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. R. E. Martin, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381
Mr. John C. Hoyle Page 5 November 25, 1997 cc: U.S. Nuclear Regulatory commission Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission On White Flint, North 111555 Rockville Pike Rockville, Maryland 20852-2739 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
O~CKETEO USHRC NUCLEAR ENERGY INSTITUTE
*97 NOV 26 P5 :04 Alexander Marion DIRECTOR, PROGRAMS NUCLEAR GENERATION DIVISION November 26, 1997 DOCKET NlMBER- S 0 PROPOSED RULE.a..I!!~----
( (of). f (<, 53 Cf 13) Mr. John C. Hoyle, Secretary U.S. Nuclear Regulatory Commission ( fo '2. ~I< 53 '/3:J..) Washington, DC 20555-0001 Attn. Rulemakings and Adjudications Staff
SUBJECT:
Codes and Standards; IEEE National Consensus Standard 62 Fed. Reg. 53932, October 17, 1997 The Nuclear Energy Institute (NEI)1, on behalf of the nuclear energy industry, submits the enclosed comments on the NRC action to amend its regulations to incorporate by reference IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations." The Federal Register notice referenced above is a "direct final rule" that will become effective on January 1, 1998, unless significant adverse comments are received by December 1, 1997. NEI recommends the Commission withdraw the direct final rule pending resolution of the following comments. We believe it is important and beneficial for NRC to endorse voluntary industry standards. Regulatory endorsement enables a licensee to proceed in confidence when applying appropriate standards in the design and operation of a nuclear power plant. The NRC typically uses regulatory guides to endorse a standard by stating that it provides a methodology acceptable to the NRC staff. For example, in July 1996 the NRC published Revision 1 to Regulatory Guide 1.153 endorsing IEEE 603-1991. NEI believes that regulatory guides are an appropriate mechanism for NRC endorsement of voluntary industry standards. The direct final rule, however, imposes IEEE 603-1991 as a requirement for future changes to existing power and instrumentation and control portions of protection systems. Furthermore, the NRC 1 NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including regulatory aspects of generic operational and technical issues. NEI members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architecUengineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry. OEC - 4 1997 Acknowledged by card ....... ,.. ,........,,2ss,r,. I //I, I ',TRr"ET, t'-JW ',IJITE 400 1./VASHINGT()f'J [) 1 20006-37(1H PHONE '..:'ll/ 7J9 8000 f-AX 20278"i.1nl9
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Mr. John C. Hoyle, Secretary November 26, 1997 Page 2 concludes the direct final rule is not a backfit and does not impose new requirements or costs on current licensees. On the contrary, NEI believes the direct final rule is a backfit under 10 CFR 50.109. The rule is a new NRC position that would require changes to the design approval processes for licensees planning other than "like-for-like" protection system modifications after January 1, 1998. The majority of licensees are currently committed to IEEE 279. Therefore, the mandatory imposition of IEEE 603 represents a change that will result in a dual licensing basis for these licensees. This would especially be the case when component level modifications are under consideration. The fact that the scope and applicability of IEEE 603 is significantly different than IEEE 279 is of particular concern. Furthermore, the NRC staffs conclusion that the terms "protection system," "safety system" and "safety-related system" are synonymous is questionable because both standards define "protection system" as a subset of "safety system." In conclusion, NEI recommends the direct final rule be withdrawn until a more thorough backfit analysis can be performed and industry comments can be evaluated. Specific comments regarding the impact of the direct final rule on operating plants are enclosed. We appreciate the opportunity to provide these comments and would be pleased to discuss our views with the NRC staff. Sincerely, Alex Marion Enclosure
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997 Supplementary Information The federal register notice states that NRC considers the direct final rule to be non-controversial because there was no adverse public comment on the regulatory guide that endorsed IEEE 603-1991 (Regulatory Guide 1.153, "Criteria for Safety Systems"). This type of rationale should not be used as the basis for dispensing with public comment on a rule change. A regulatory guide by itself does not impose requirements, thus, the lack of licensee comments on a proposed guide does not indicate agreement with the content of the guide; it simply means that licensees decided to apply limited resources to other matters. To invoke an industry standard by direct final rule sidesteps public comment and improperly shifts the burden to - licensees to "prove" the rule is not necessary. The federal register notice that requested public comment on draft regulatory guide DG-1042 (proposed Revision 1 to Regulatory Guide 1.153) was published on November 15, 1995. It stated, in part, "The draft guide has not received complete staff review and does not represent an official staff position" (62 Fed. Reg. 57461). In a listing of applicable General Design Criteria, the draft guide contained the following statement, "In addition, 10 CFR 50.55a, 'Codes and Standards,' requires in paragraph (h) that protection systems meet the requirements set forth in IEEE Std 279-1971 ... " Thus, the draft guide did not suggest any change to the licensing basis for current operating plants and should not be used as the foundation for the direct final rule. Past NRC staff practice has been to use regulatory guides to endorse IEEE standards as an acceptable method for satisfying NRC requirements. In general, alternatives to regulatory positions published in regulatory guides are acceptable to NRC when documented in a plant's licensing basis. However, the direct final rule would apply IEEE 603 to new plant construction and major system upgrades at existing plants even if IEEE 603 is not currently part of the existing plant's licensing basis. This is contrary to the principle that the use of consensus standards should be voluntary. Under the terms of the direct final rule, IEEE 603 becomes a mandatory requirement effective January 1, 1998. It will apply to modifications to existing power and instrumentation and control portions of protection systems. While it is recognized that IEEE 603 provides acceptable criteria for safety systems, NEI does not agree with the mandatory application of this criteria to operating plants licensed to a significantly different standard, i.e., IEEE 279. The rule goes well beyond the application intended by IEEE when the older standard was developed. 1
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997
Background
The federal register notice states that, "Because there were no adverse comments to revision 1 to Regulatory Guide 1.153, the Commission believes there is a general public consensus that IEEE 603 provides acceptable criteria for safety systems in nuclear power plants." The NEI position on this statement appears above in the comments on the "Supplementary Information" section of the direct final rule. Discussion NEI believes that the NRC practice of issuing regulatory guides to endorse industry consensus standards is consistent with the National Technology Transfer and Advancement Act of 1995. However, when NRC regulatory action proceeds in the form of rulemaking, the NRC must determine that the imposition of the new regulatory requirement is justified in accordance with 10 CFR 50.109. Although that determination has been made in the case of the direct final rule, we believe the impact of this rule on utility configuration management processes, and the associated licensing and design bases, has not been adequately evaluated. The definitions of "protection system" as used in IEEE 279 and IEEE 603 are consistent in that they consider the protection system to include the sense and command features for the reactor trip system and the engineered safety features. However, this is significantly different from the IEEE 603 definition of a safety system as, "A system that is 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 down the reactor and maintain it in a safe shutdown condition, or (iii) the capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the 10CFR Part 100 guidelines." These terms are not synonymous based on the content and intended use of the cited IEEE standards. Figure 2 in IEEE 603-1991 represents a matrix of safety system elements. It states, "The protection system of IEEE 279-1971 (withdrawn) and of this standard is the sense and command features for the reactor trip system and the engineered safety features." This scope, which is the proper scope, encompasses only one of the nine safety system elements presented in Figure 2. It is consistent with past practice, existing guidance and common usage of these terms. The direct final rule, however, expands the scope to include all safety systems. The direct final rule would require that any plant change other than a "like-for-like" component change satisfy IEEE 603-1991. Like-for-like replacements are often difficult, especially for older plants, because in many cases older model components are no longer manufactured. Such cases of "equipment obsolescence" are a primary reason for many plant upgrades to new design and technology. These upgrades are 2
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997 often the only feasible "fix" to maintain the integrity and operational compliance of plant safety systems, thus, they cannot be viewed as "voluntary." In addition, work is often performed in accordance with Technical Specification Limiting Conditions for Operation (LCOs), where time to repair equipment or seek exemption under 10 CFR 50.12 is limited. Therefore, before NRC imposes new requirements that (1) will make many component replacements more complex and costly, and (2) may impede resolution of Technical Specification issues within the times permitted by LCOs. Regulatory Analysis The regulatory analysis attached to SECY-97-201, "Changes to Paragraph (h) of 10 CFR Part 50.55a, 'Codes and Standards,"' should be revised to consider the impact of this rule on modifications to current plant protection systems and safety systems. Based on NEI comments in the "Discussion" section above, the application of IEEE 603 to changes in existing plant protection systems coupled with the decision to declare certain terms as synonymous, should be considered in the NRC regulatory analysis. Backfit Analysis According to this rule, plants currently licensed to IEEE 279 are required to review future changes to protection systems and safety systems in accordance with IEEE 603 -1991. This review will result in significant changes to the plant licensing and design bases. A dual licensing basis will be created for those design modifications initiated after January 1, 1998. For example, if a utility initiates a design modification to replace a component with slightly different make/model number, or if the change requires a minor modification to the system architecture, which portions of the circuit and auxiliary systems must be upgraded to become IEEE-603 compliant? It is unclear whether this condition is acceptable as a partial implementation of the new design standard or whether the new standard is only applicable to the replacement of an entire system. Even if IEEE-603 only applied to the portion of the protective system that was changed, it would be difficult and confusing to track compliance with IEEE-279 for existing portions of a system as well as compliance with IEEE-603 for the modified portion. Further confusion would be introduced by the applicability and impact of the other standards referenced in IEEE 603 for only the portion of the system that was changed. At this point, the question of how compliance with this direct rule will be determined and how compliance will be maintained presents a significant challenge. An additional concern remains with regard to the applicable Technical Specifications, and if revisions are or will be required to reflect the dual licensing 3
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997 basis. This could prohibit modifications under 10 CFR 50.59. Also, a number utilities that are revising plant surveillance procedures as part of a technical specification improvement program will be impacted by IEEE 603 and the other referenced standards identified in this rule. IEEE 603 references 13 IEEE standards, three ANS/ANSI standards and one ISA standard. This is another significant difference with IEEE 279-1971, which did not contain such references. These differences lead to several questions:
- If a certain version of a standard is referenced in IEEE 603, a second version in the Regulatory Guides, and a third version in the licensee's UFSAR, which version must be followed?
- If a Regulatory Guide is updated in the future to reference a new version of an IEEE standard referenced in IEEE 603, does such "endorsement" by the NRC constitute a de facto revision to 10CFR50.55a? If so, at what time does the change take effect?
- If an industry standard, which is both referenced in IEEE 603 and endorsed by a Regulatory Guide, references other standards not listed in IEEE 603, does this rule require that the subsidiary references also be followed?
- Some of the standards referenced in IEEE 603 have been revised and reissued since the publication of IEEE 603. Do licensees have the flexibility to adopt the newer requirements, or must they continue with commitments to the older ones?
- In cases where the UFSAR now takes exception to certain provisions of the standards, must licensees revise the UFSAR to eliminate the exception for future work?
Plants are licensed to different versions of the referenced standards and in a number of cases are not committed to all of the existing regulatory guides endorsing IEEE standards. The direct final rule will require all such commitments to be revisited when modifications to protection systems and safety systems are planned. For example, Regulatory Guide 1.75 endorses IEEE 384, and this rule may impact existing control panels, annunciators, recorders, and computer monitoring even though current designs have been determined by NRC to be acceptable. The fourth paragraph of the "Discussion" section in the direct final rule states, "The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same." This is clearly unfounded, since many of the "Auxiliary Supporting Features" and "Other Auxiliary Supporting Features," for which examples are given in Fig. 3 of IEEE 603-1991, were not previously identified as being within the scope of IEEE 279-1971. 4
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997 IEEE Std. 279-1971 classifies Single Failure Criterion in Section 4.2 as "Any single failure within the protection system shall not prevent proper protective action at the system level when required." IEEE Std. 603-1991 classifies Single Failure Criterion in Section 5.1 as "The safety systems shall perform all safety functions required for a design basis event in the presence of: (1) any single detectable failure within the safety systems concurrent with all identifiable but non-detectable failures; (2) all failures caused by the single failure; and (3) all failures and spurious system actions that cause or are caused by the design basis event requiring the safety functions." Furthermore, Section 5.1 of IEEE-603 states that if a design does not meet the single failure criterion then design features shall be provided or corrective modifications shall be made to ensure that the system meets the specified reliability requirements. The single failure criterion specified in IEEE Std. 279-1971, was one of the major guidelines/criteria utilized in the design and development of the protection systems in current nuclear plants. The single failure criterion specified in IEEE Std. 603-1991 constitutes a major change from the original criteria and scope. To evaluate the impact of this revised criteria would require a detailed review of existing systems just to determine the scope of impact. Since the new criteria were not design guidelines applicable to current protection systems and constitute a significantly expanded scope, revising existing systems to meet the new criteria would typically require major modifications or in some cases complete system replacements. This would be particularly true with the various electronic systems, such as the protection systems, since existing electronic design and circuitry do not allow major changes in logic and function. IEEE 603, Section 5.3, Quality, requires that all components be designed, manufactured, inspected, installed, tested, operated, and maintained in accordance with ANSI/ASME NQAl-1989. Not all licensees are currently committed to NQAl. This will require the licensee to maintain dual quality programs; one for items that are included in IEEE 603 due to changes in the protection system, and second quality program for the remainder of the components. Another approach would be to change the quality program for all components to meet the requirements of NQAl. Either approach could require considerable expense to licensees. Section 5.12 of IEEE 603 sets requirements for Auxiliary Features. Section 5.12.1 states "Auxiliary support features shall meet all requirements of this standard." Also, Section 5.12.2 (2) attempts to define other auxiliary features as "... part of the safety system by association (that is, not isolated from the safety system) ... " Also, this section refers to Appendix A as an illustration of these criteria. This section and the definition on page 13 many systems that would be subject to the rule. The cost to modify any system/component and it associated auxiliary system would be high, and may even require the redesign of all or part of the non-class auxiliary systems anytime a modification is needed. 5
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997 Section 8 of IEEE 603 has no equivalent requirement in IEEE 279. For many licensees, the design and licensing basis for electric power systems is IEEE 308-1971. The requirement that changes and modifications to the protection system conform to the standards cited within the IEEE standard and associated regulatory guides, this is clearly an increase in the scope of plant changes and will incur an associated increase in the cost of the modification. The level of effort and expense to incorporate IEEE 603 into a plant's licensing documents is not trivial. It would be necessary to update the UFSAR and other licensing and design basis documents to reflect the new requirements, including the newer versions of industry standards referenced by IEEE 603. This would include the need for a clear delineation of plant systems and portions thereof that fall within the current licensing basis and the new licensing basis invoked by this rule. Furthermore, any ongoing or planned design work could not be completed until the UFSAR revision was issued, since the current IEEE references in the UFSAR would not be consistent with this rule. The impact on 10 CFR 50.59 reviews may lead to unreviewed safety questions requiring prior NRC staff review. This rule will impact a plant's design basis with regard to protection system modifications because base design inputs are changed by this rule. These inputs are reflected in plant design basis manuals, engineering drawings, engineering specifications, installation specifications, etc. Significant engineering man-hours will be needed to identify and review differences between current design documents and the later revisions of the referenced standards. This will also impact the review on the plant and design basis updates. In addition, training of plant personnel may be significant, especially for those personnel involved in other programs, e.g., equipment qualification, testing, operations, maintenance, human factors, and system reliability/availability assessment. This rule makes no allowance for in-process modifications. For example, at a two-unit facility with an outage scheduled for one unit in early 1998 and the other unit in the fall or early 1999, the design change package is essentially complete for the modification to be made on the first unit. According to this rule and depending upon how changes to protection systems and safety systems "initiated after January 1, 1998 ... " is interpreted, the impact on existing outage plans and schedules could be significant due to re-engineering the design packages, procurements of materials to new standards, exemption requests, etc. Within the context of compliance to the IEEE-603 standard as incorporated into 10 CFR 50.55a, several specific areas will require clarification as follows:
- Within the context of 10 CFR 50.55a, how is the phrase "to the extent feasible and practical" (Section 6.4) to be interpreted?
6
ENCLOSURE Specific Comments to 62 Fed. Reg. 53932, October 17, 1997
- Section 2 provides several definitions that have not been previously defined in the regulations such as "Class IE", "division" and "associated circuit." The definition of "division" varies between Technical Specification of operating plants licensed under the Part 50 and Part 52 plants.
- What is the legal interpretation of "acceptable reliability" (Section 8.3)? This term is subjective and needs to be clarified.
- IEEE 603 requires that indication shall be automatically actuated if the bypass is expected to occur more than once a year. How is this frequency to be monitored and what regulatory consequences would result from more frequent bypasses?
This rule will have an impact on license renewal. The inclusion of IEEE 603 into 10 CFR 50.55a would require significant re-analysis of the protection systems and safety systems for plants whose licensing and design bases includes IEEE-279 as well as older versions of the standards referenced in IEEE 603. It is highly unlikely that plants designed in the 1960's and 1970's would be able to meet such new license conditions without significant expenditure. Conclusion We believe that the NRC should withdraw this direct final rule pending resolution of the above comments as well as other comments received during this comments period. Recommendation The action taken by IEEE to withdraw IEEE 279 and incorporate its provisions into IEEE 603 cannot be prevented. Standard development organizations will continue to proceed with similar action in the future as design practices evolve. We continue to believe that regulatory guides are an appropriate mechanism for NRC endorsement of voluntary industry standards. For this specific case, NRC may wish to consider a revision to 50.55a(h) that directly cites those provisions of IEEE 603 that carry forward the fundamental design principles applicable to protection systems as described in IEEE 279. Such action would allow utility licensees to retain the current commitments to IEEE 279 and provide the NRC with regulatory authority via 50.55a. And, the endorsement of IEEE 603 by Regulatory Guide 1.153 will continue to identify this standard as an acceptable method for meeting NRC regulatory requirements. File: x:\603encl 7
DOCKETED Farouk D. Baxter USNRC 23 Pilgrims Path Sudbury, MA 01776 "97 NOV 26 P2 :48 (978)443 - 2914 FAX : (978)443 - 8556 OFFICE OF S *:, 1 HY RULEM;,r(IN,::,s *..,, 0 November 24, 1997 ADJUDICA110 I:' STAFF Secretary U.S. Nuclear Regulatory Commission DOCKET NI NBER Washington , DC 20555-0001 A11.o: Rulemakings and Adjudications Staff PROPOSED RULE " S 0 ( (l)~FR5397S) ( ~ :J.Ffe 5'"3 '13..1-)
Dear Sir:
I would like to comment on the proposed final rule amending NRC regulations to incorporate IEEE Standard 603-1991 as outlined in the Federal Register, October 17, 1997 (Volume 62, Number 201). I offer my comments as an independent consultant and industry recognized expert and specialist in nuclear power plant electrical systems. I am intimately familiar with both IEEE Standards 279 and 603, and was involved with the creation and development of Standard 603, and as such I am knowledgeable of its intended interface with Standard 279 and other IEEE standards. Comment 1. Under the heading of DISCUSSION, it is stated that "IEEE 603-1971 uses the term 'safety systems' rather than 'protection systems'.* Further on it is stated that "The Commission considers that the systems covered by IEEE Standard 603-1991 and IEEE Standard 279- 1971 are the same. Therefore for the purposes of paragraph (h) of 10 CFR 50.55a, 'protection systems' and 'safety systems' are synonymous." These two statements which are the very core of the rutemaking cause grave concern. The NRC has apparently misunderstood and misinterpreted the purpose and intent of the two referenced standards. The facts of the issue are presented below: (a) The use of the words 'protection system' in IEEE Standard 279 was intentional, as was the use of the words 'safety system' in IEEE Standard 603. The standards intended that each term be used as defined. Any suggestion that they mean one and the same thing, or that they are synonymous reflects a misunderstanding and misinterpretation of what was intended. (b) Both terms, 'protection system' and 'safety system' are presently defined in IEEE Standard 603-1991 as separate entities. It should be noted that 'protection system' has a very limited application, while 'safety system' is broad-based and all encompassing, and
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thereby embraces the 'protection system' in addition to other electrical systems. It is therefore not correct for the NRC to summarily conclude that the terms are synonymous, and to suggest that the terms were misused in the IEEE standards. Clearly 'protection system' and 'safety system' cannot be construed as having the same meaning. (c) The scope of IEEE Standard 279-1971 was very specific in the use of the term
'protection system', and this is best captured by a paragraph in the Foreword of the document which states *------the system presently described herein can be more appropriately called a 'protection signal system' which by definition includes both instrument and logic channels. The protection system will then consist of the protective signal system and the actuator system. The user should recognize that neither the present nor the expanded scope includes all of the structures and equipment required for complete plant protection.*
The message here is that IEEE Standard 279-1971 addressed only one aspect of the multiple systems and equipment needed for plant safety, and further acknowledged that other areas existed which were not addressed by IEEE Standard 279-1971. In the time frame when IEEE Standard 279-1971 was issued, IEEE Standards activities took the approach of developing specific purpose standards which provided unique detailed criteria for specific elements required for safe operation of the plant. Typical of these standards was IEEE Standard 308, 323, 336, 338, 379, 384, 387 and sn etc. For example, IEEE Standard 308-1971 , *criteria for Class IE Electric Systems for Nuclear Power Generating Stations" provided criteria, requirements, and minimum conditions, for the electric power system. These requirements were independent of IEEE Standard 279-1971 because Standard 279-1971 had no relationship with Standard 308-1971 and was never intended to be applicable to the Class 1E Power System. If it had been intended that Standard 279-1971 envelope the requirements for the power system, then Standard 279-1971 would have been referenced therein, or the term 'protection system' would have been used in Standard 308-1971 to indicate that the electric power system was considered part and parcel of the protection system, but no such link was ever intended or implied. Clearly IEEE Standard 279-1971 has a very limited scope which is to exclusively address the 'protection system'. (Also see Item (e) below relating to confusion of terms between engineering disciplines). Other specific purpose standards such as those mentioned above, were similar in their approach to IEEE Standard 308. Though some may have been applicable to the 'protection system' in addition to other systems, there was never an implication or insinuation that these standards were part of the 'protection system' addressed by IEEE Standard 279-1971. (d) IEEE Standard 603 evolved in 1976/77 as the nuclear industry matured and the need for a 'mother' safety document was envisioned. Standard 603 did not supersede Standard 279-1971 as is implied by the rulemaking, instead IEEE Standard 603 came up with as set of safety criteria and requirements that would dictate safety philosophy to all the specific purpose standards such as IEEE Standards 279, 308, 323, 338, 379, 384, 387, and 577. IEEE Standard 603 would be the safety system criteria document or umbrella 2
document for power, instrumentation, and control portions in nuclear power safety systems. The 'safety system' created and enveloped by IEEE Standard 603 established the the minimum functional design requirements for the power, instrumentation, and control portions in nuclear power safety systems, a concept that went far beyond the very limited scope of IEEE Standard 279-1971 . Standard 603 became the 'mother' document of all documents addressing power, control, and instrumentation, portions of the nuclear plant that were required for safety. Such a 'mother' document did not exist prior to the introduction of Standard 603. The NRC's attempt to mandate through rulemaking that IEEE Standard 279-1971 was such a 'mother' document is therefore without basis. (e) In 1978, in the process of revising Trial-Use IEEE Standard 603-1977, it was recognized that users of the document from different engineering disciplines were developing different meanings and interpretations of the terms 'protective' and
'protection'. This led to considerable confusion and misunderstanding, particularly when
'protection' was applied to the Class 1E Power System. To avoid future misapplication of the safety system criteria, it was decided to intrcx:tuce two new terms 'sense and command' and 'execute features' to essentially replace 'protection system' and 'protective action' respectively.
It iS believed that in deyelo.ping the present rulematdng the NBC may have succumbed to the same confusion and misunderstanding which which was prevalent in 1978. Comment 2. Under the heading of SUPPLEMENTARY INFORMATION it is stated that "NRC considers this rulemaking to be noncontroversial because as noted in the background discussion, there was no adverse public comment on the regulatory guide endorsing this standard." Under the heading of BACKGROUND it is stated that "Because there were no adverse public comments to Revision 1 to Regulatory Guide 1. 153, the Commission believes that there is general public consensus that IEEE Standard 603-1971 provides acceptable criteria for safety systems in nuclear power plants." (a) While the latter statement is true, that is, there is no disagreement that IEEE Standard 603 provides acceptable criteria for safety systems in nuclear power plants; and in fact NRC endorsement of this standard was encouraged by the nuclear industry. However, the former statement regarding the rulemaking is totally unrelated to the public's acceptance at the Regulatory Guide. The assumption made by the NRC is that because the regulatory guide was non-controversial the rulemaking would also be non-controversial. This logic is difficult to comprehend because these appear to be two unrelated issues. (b) While it is also true that IEEE Standard 279-1971 has been withdrawn by IEEE because it was no longer being updated or reaffirmed, it is not 'obsolete' as stated by the NRC. It is my understanding that IEEE's position is that the standard is withdrawn, and that IEEE takes no position on its continued use or disuse by users. It should however be of interest 3
to note that IEEE Standard 279 is currently referenced by almost every operating nuclear plant in the U.S, and remains an active design and licensing basis document at these plants. One must therefore question the basis on which the NRC has now declared that the standard is 'obsolete'. Recommendation
- 1. A very clear understanding of the differences between the 'protection system' as promulgated by IEEE Standard 279-1971 , and the 'safety system' as promulgated by IEEE Standard 603-1991 , must established by the NRC before even considering any proposed rulemaking on this subject. To do otherwise would amount to a technically irresponsible and flawed decision.
- 2. The endorsement of IEEE Standard 603-1991 by Regulatory Guide 1.153 would appear to be an adequate vehicle to fulfill the NRC's need to maintain their regulations current. Thus the purpose behind the rulemaking would appear to be met through the regulatory guide.
- 3. IEEE Standard 279 is referenced by almost every operating nuclear plant in the U.S., and furthermore, remains an active design and licensing basis document at these plants; therefore, for the NRC to declare the standard 'obsolete' and remove it from regulations is believed imprudent.
- 4. IEEE should be requested to provide further clarification in the next revision of Standard 603 to ensure that misunderstanding of critical terms is minimized.
I will be glad to meet with any NRC staff members, the ACRS, or the Commissioners to discuss any of the above comments, and to explain why I believe the proposed rulemaking incorporating IEEE Standard 603-1991 is technically flawed and is based on a misunderstanding of the issues involved. Sincerely, Farouk D. Baxter, PE Specialist Nuclear Power Plant Electrical Systems 4
Duane Arnold Energy Center 3277 DAEC Road 0 Palo, IA 52324 DOCKETED Telephone 319 851 7611 USNRC Fax 319 851 7611 UTILITIES "97 NOV 26 P1 :47 November 25, 1997 NG-97-2036 DOCKET NUMBEA Secretary, PROPOSED RULE 5 o U.S. Nuclear Regulatory Commission ( IP~Ft< 53~7.S Attn: Rulemakings and Adjudications Staff ( f,Q FlfS3'1J.2.) Washington, DC 20555-0001
Subject:
Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Comment on Direct Final Rule and Proposed Rule 10 CFR 50.55a(h)-Protection and Safety Systems File: A-106 On October 17, 1997, the Nuclear Regulatory Commission published for public comment a draft Regulatory Guide DG-1042, "Criteria for Safety Systems." Attached are IES Utilities Inc. ' s comments on the proposed rulemaking. Should you have any questions regarding the attached information, please contact this office. Sincerely, Kenneth E. Peveler Manager, Regulatory Performance
Attachment:
Comment on Direct Final Rule and Proposed Rule 10CFR50.55a(h) Protection and Safety Systems cc: E. Protsch J. Franz D. Wilson G. Kelly (NRC-NRR) A. B. Beach (Region III) NRC Resident Office DOCU ,,, 4 OEC - ~ 1997 Acknowledged An /ES Industries Company
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Attachment to NG-97-2036 Page l of 4 Comment on Direct Final Rule and Proposed Rule 10CFR50.55a(h) Protection and Safety Systems IES Utilities would like to document Significant Adverse Comments concerning the direct final rule (62 Fed. Reg. 53932) and the proposed rule (62 Fed. Reg. 53975) amending 10CFR50.55a(h), Protection systems, to incorporate by reference Institute of Electrical and Electronics Engineers Standard (IEEE) std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," and changing the section title to "Protection and Safety Systems." The amended rule is not clear regarding application to plants that received their construction permits prior to 1971. This is of particular concern to IES Utilities because the construction permit for the Duane Arnold Energy Center (DAEC) was issued on June 17, 1970 prior to IEEE 279(71) being issued. Plant protection systems were designed to meet General Electric (GE) Design Safety Standards. GE NEDO 10139 describes how the DAEC design meets the intent of IEEE-279 with exceptions. 10CFR50.55a(h) presently requires protection systems of plants that received a Construction Permit after 1/1/1971 to comply with IEEE-279. The amended rule would require compliance with either IEEE-279 or IEEE-603(91) by those plants. However, it adds a separate sentence stating that changes to protection systems initiated after January 1, 1998 must meet IEEE-603(91 ). The language of the amendment can be interpreted as requiring that after January 1, 1998, changes to plants will be required to meet IEEE-603(91), regardless of the original Design Safety Standards. We understand that the intent is that plants which have not been previously required to meet IEEE-279 will continue to be permitted to make changes to their plant consistent with Licensing Basis and commitments made to the NRC, without regard to either IEEE-279 or IEEE-603(91). However, this intent is not clearly stated. The proposed rule should be revised to make its intent in this regard clear. Our concern about this lack of clarity is heightened by the discussion of the rule change in the Federal Register, which states: "the rule would require future changes to existing power and instrumentation and control portions of protection systems to comply with the new standard. This would not be considered a backfit, since the changes are voluntarily initiated by the licensee, or separately imposed by the NRC after a separate backfit analysis." The quoted language does not distinguish between plants that are required to meet IEEE-279 and those that are not. The distinction is important because it would be particularly burdensome to apply IEEE-603(91) to plants that were not designed to IEEE-279. To make this intent clear, we suggest that proposed paragraph 50.55a(h)(3) be revised by inserting "in such plants" in the second sentence after the words "changes to protection systems."
Attachment to NG-97-2036 Page 2 of 4 If this is not the intent, then the application of this rule to plants with Construction Permits before 1/1/71 would constitute a "backfit" under 10CFR50.109. The language quoted above from the Federal Register notice argues that requiring changes to comply with IEEE-603(91) should not be viewed as a backfit because changes are "voluntary." This is not correct; most plant changes are not voluntary, but are imposed by new standards or lack of equipment availability of "like-for-like" replacement. Equipment obsolescence is a primary reason for many plant upgrades to new design and technology. These plant changes are often the only feasible "fix" to maintain the integrity and operational compliance of plant safety systems, and cannot be viewed as voluntary. Some of this work is done under Technical Specifications Limiting Condition for Operations (LCOs), where time is limited to repair equipment, or seek IOCFRS0.12 exemption requests. Before NRC imposes new requirements that may impede resolution of such issues within the times permitted by LCOs, it should conduct a careful analysis of the impacts and safety benefits in accordance with 10CFRS0.109. We also believe imposing IEEE std. 603(91) on future plant upgrades would significantly impact IES without a commensurate safety benefit. An exhaustive comparison of IEEE std. 603 against DAEC plant design would be a large effort. Completion of this comparison would be needed before the full impact of the new rule on future modifications can be known. However, a cursory review indicates that the impact would be significant. For example, in planning replacement of Reactor Water Cleanup GEMAC, "like-for-like" replacement of the obsolete GEMAC control components is not feasible. If IEEE std. 603 is applied, it would require the use of other standards by use of the word "shall", none of which were in effect at the time of the DAEC Construction License, and to none of which the DAEC is committed. As a result, it would impose IEEE-384 redundancy and separation requirements, which were not required in the original design. Since this system is not required to be redundant, and there would be no redundancy for system components that were not replaced, the imposition of a redundancy requirement to the replacement components would not significantly enhance the safety of the overall system. The replacement components would also have to meet the requirements of IEEE-323(83) for meeting mild environment qualification testing requirements, while other components in the same system are not designed to such requirements. Another concern is having a mix of plant equipment and systems that have different requirements for maintenance and testing. This would compound the number of test procedures for equipment (IEEE-338(87)), and increase the documentation in the Design Bases Documents and Updated Final Safety Analysis Report (UFSAR) for describing the interfaces between the different code applications. If a component in an instrument channel is changed, the extent of application of the new standards to other channel
Attachment to NG-97-2036 Page 3 of 4 components, or the whole channel itself would need to be decided. Then separation between channels might not meet the requirement of the new standard. Another example is the limited conformance to IEEE-384. The DAEC UFSAR states that " ... input and output power and instrumentation cables are routed independently and in separate conduit or cable trays to meet the divisional requirements of IEEE-384." DAEC was not built to other separation requirements of IEEE-384, for instance in the Control Room where divisions come together in some control panels, which is permitted per our original licensing bases. Maintaining the integrity of the plant's design and licensing bases is the foundation of nuclear safety. Arbitrarily requiring future changes to comply with a new standard is clearly a backfit. To conclude that it is not a backfit based on changes being voluntary is not generally well founded. The Federal Register Notice argues that a direct final rule is appropriate because there were not significant comments on draft Reg. Guide DG-1042, issued in November 1995, which proposed Rev. 1 to Reg. Guide 1.153, Criteria for Safety Systems, endorsing IEEE 603(91). This argument is misleading. Regulatory Guides are acceptable, but not the only means of meeting NRC requirements. Regulatory Guides merely describe methods acceptable to the NRC staff for complying with NRC regulations. Commitment to a Regulatory Guide is voluntary on the part of a licensee and there is no requirement to comment on Regulatory Guides (draft or final). IES has not committed to RG 1.153 (rev.
- 1) and has no plans to commit to it in the future. IES did not respond to Draft Reg. Guide DG-1042 because it did not apply to us. The burdens that would result if changes to DAEC must comply with IEEE-603(91), were not imposed by the approval of Rev. 1 to Reg. Guide 1.153.
The proposed amendments would apply to changes "initiated" after January 1, 1998. The word "initiated" is not defined in the regulation, but we understand that the modifications are considered to be "initiated" once design responsibility is assigned to a designer or design team. An interpretation that a change is not "initiated" until the start of fabrication of components or of installation in the plant, would work a particular hardship if it were applied to IES. The designs of many changes are already completed and are ready for installation in support of our April 1998 Refuel Outage (RPO), and in conformance to the existing design and licensing bases. Requiring such modifications to comply with this new standard would create significant impact on the RPO schedule or require IES to request exemption from the new regulation. With the limited amount of notice of the final rule, it is not practical for us to review all of these changes for compliance to this new requirement and make any necessary changes to comply, nor to do the research to support individual exemption requests per 10CFR50. l 2 This could result in the Startup
Attachment to NG-97-2036 Page 4 of 4 from the refueling outage being delayed at significant cost and inconvenience to IES. It appears that this additional burden on licensees was not considered in the staffs evaluation of this rule. The discussion above points out the significant hardship that would be imposed on IES if it is required to apply IEEE 603(91) to future plant changes and shows that there would not be a commensurate safety benefit. If the Commission nevertheless proceeds to require compliance with IEEE-603(91), it should revise the definition of changes in the direct/proposed rule. As written, it includes modifications, augmentations or replacements pursuant to license amendments and changes pursuant to 10 CFR 50.59. This definition should be revised to exclude application to 10 CFR 50.59 changes. No safety benefit would be achieved by requiring IEEE-603(91) to apply to changes made pursuant to 10 CFR 50.59, since by definition, such changes do not involve an unreviewed safety question. The majority of changes made pursuant to 10 CFR 50.59 are minor, and a requirement to analyze each against a new standard, particularly one not previously applied to the plant, would impose a significant burden on the engineering process that has no apparent safety benefit.
DOCKET PROPOSED fU.E II 5 6 rJ> (. ( fo!J.FR.5'3175) ( (,:JF/'<53'1'32) l.:;../ November 10, 1997 Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Attention: Rulemaking and Adjudications Staff
SUBJECT:
Nuclear Regulatory Commission, 10CFR Part 50, RIN 3150-AF73 This letter provides my sianificant adverse comments on the issue of the direct final rule to incorporate by reference, national standard IEEE 603-1991 in 10 CFR 50.55a, "Codes and Standards," as described in SECY-97-201. This incorporation of IEEE 603-1991, to replace IEEE 279 for all modifications is a major change in regulatory commitment for all cuLrent U.S. nuclear plants and will result in significant adverse effect on our current fleet of plants. The reasons are as fol lo~\-:;; l.) There iD ri.:::, IEEE 603 designed t:,lanL in the U.S. today that is t~ilt - only on paper. 1.) The supplementary information describes this as a non-controversial issue because the IEEE standard was endorsed by a Regulatcry Guide with little comment by the industry. This is a significantly different regulatory action and does invoke the backfit rule. The addition of guidance and further clarification in the form of Regulatory Guides has long been endorsed by the industry and supported as this one was. The issue of direct rulemaking with no exceptions for
~11 m~difi~~~ions is no lcn~Pr a guide but a law and cannot be implemented in our cLrrert olants for the reasons listed below.
3.) A cual licensing basis may be created. For example, if a
- t. '(,t,.,'
plant replaces a design with like component with slightly . . ~ ; * ~- differennt make/model number, is this now an IEEE 603 *- device? Or is it an IEEE 603 system? This makes it all the harder when you consider that we don't have any current IEEE 603 plants. Acknowledged~ cad ID 2 I 1987 MDM/Lamb, Inc.
- 28202 Cabot Road, Suite 205
- Laguna Niguel, California 92677 (714) 365-1350 Fax (714) 365-1360
- www.mdmcorp.com
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November 10, 1997 Page 2 of 3 4.) The discussion sections mentions the IEEE referenced standards. This is no discussions of standards which are included in the body of the standard as "shall." There are approximately 100 "sh:3.ll's in the standard including IEEE 323, 627, 384 and NQA-1. The example of IEEE 384 in itself is a very critical financial and design exposure, because almost all of today's olants do not comply with IEEE 384. In rulemaking implementation, IEEE 384 compliance within the modification boundary and not outside creates a design and safety hazard and in some cases will be impossible to implement, without redesigning all the way back to switchgear. An example $10,000 modification could result in may millions of dollars in cost to comply taking a power supply all the way back to Lhe source in implementing electrical separation. 5.) As a law instead of a Regulatory Guide, how is the phrase "to the extent feasible and practical (Reference IEEE 603 Section 6.4) to be interpreted? 6.) Will the applicable Technical Specifications require modification to reflect the dual licensing basis? Would that prohibit modifications under 50.59? 7.) The proposed rule seems to state that compliance to IEEE E:03 only applies to plants with permits issue after 1/1/71. However, the discussion section mentions no limitations. Are plants with construction permits prior to 1/1/71 included? 8.) IEEE 603 Section 2 provides several definitions that have not been previously defined in law such as "Class lE",
"division" and "associated circuit." The definition of division various between Technical Specification of operating plants and significantly between the part 50 and part 52 plants.
9.) What J_s the legal ir..t.e:rpre:taticr-. cf "~~ce::,t-.,,hl"" reli:::d'.)ili ty" (Reference IEEE 603 Section 8.3)? 10.) IEEE 603 requires that indication shall be automatically actuated if the bypass is expected to occur more than once a year .... How is this frequency determined on a legal basis? 11.) The proposed rule states that systems covered by IEEE 279 and IEEE 603 are the same. Are all of the auxiliary supporting features in the 279 plants classified as safety grade in all cases? Are the scopes completely identical?
November 10, 1997 Page 3 of 3 12.) The implementation with the backfit analysis is unrealistic. The is 3. MAJOR scope and licensing basis change to the utilitv owner and will result in massive dollar costs for minor safety related uogrades to improve safety. It will discourage licensees from making improvements. These are my initial comments. I would like the opportunity to meet in person and go over every comment with you and the resultant impact on plants. Please advise at your earliest convenience. I can be reached at (714) 365-1350. Sincerely yours;
~~Hnn Vice President - MDM and President-Elect American Nuclear Society cc: D. Miller - ACRS A. Marion - NEI R. Torok - EPRI Edward (Ted) L. Quinn Vice Pres id ent AMERICAN NUCLEAR SOCIETY 555 North Kensington Avenue La Grange Park, Illinois 60526 USA Tel: 714/ 365-1350 Fax: 714/ 365-1360 Web Site: www.ans.org E-Mail: equinn@mdmcorp.com
DOCKET NUMBER PR PROPOSED RULE...!..!!.....;;... - - DOCKETED (62 Ff<£" J Cf 5'J.J USNRC [7599-01-P]
"97 OCT -9 P8 :04 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AF73 Codes and Standards; IEEE National Consensus Standard AGENCY: Nuclear Regulatory Commission.
ACTION: Direct final rule.
SUMMARY
- The Nuclear Regulatory Commission is amending its regulations to incorporate by reference IEEE Std. 603-1991, a national consensus standard for power, instrumentation, and control portions of safety systems in nuclear power plants. This action is necessary to endorse the latest version of this national consensus standard in NRC's regulations, and replace an IEEE standard currently endorsed in the NRC's regulations which has been withdrawn by the IEEE.
EFFECTIVE DATE: The final rule is effective on January 1, 1998, unless significant adverse comments are received by December 1, 1997. If the effective date is delayed, timely notice will be published in the Federal Register. The incorporation by reference of IEEE Std. 603-1991 is approved by the Director of the Federal Register as of January 1, 1998.
2 ADDRESSES : Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Attention: Rulemakings and Adjudications Staff. Hand deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4: 15 p.m. on Federal workdays. FOR FURTHER INFORMATION CONTACT: Satish K. Aggarwal, Senior Program Manager, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC - 20555, Telephone (301) 415-6005, Fax (301) 415-5074 (e-mail: SKA@NRC.GOV). SUPPLEMENTARY INFORMATION: NRC considers this rulemaking, which endorses IEEE Std. 603-1991, to be noncontroversial because, as noted in the background discussion, there was no adverse public comment on the regulatory guide endorsing this standard. Accordingly, the Commission finds that public notice and opportunity for comment are unnecessary pursuant to 5 U.S.C. 553(b)(B). Thus, the Commission is publishing this rule in final form without seeking public comments on the amendment in a proposed rule. This action will become effective on January 1, 1998. However, if the NRC receives significant adverse comments by December 1, 1997, then the NRC will publish a document that withdraws this action, and will address the comments received in response to an identical proposed rule which is being concurrently published in the proposed rules section of this Federal Register. Any significant adverse comments will be addressed in a subsequent final rule . The NRC will not initiate a second comment period on this action in the event the direct final rule is withdrawn.
3
Background
In 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,"
§ 50.55a requires that the protection systems in nuclear power plants meet the requirements set forth in IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," in effect on the formal docket date of the application. However, IEEE Std. 279 is obsolete, has been withdrawn by IEEE and has now been superseded by IEEE Std. 603-1991, e "Criteria for Safety Systems for Nuclear Power Generating Stations."
In November 1995, the NRC staff issued for public comment a draft regulatory guide, DG-1042, which was proposed Revision 1 to Regulatory Guide 1.153, "Criteria for Safety Systems." This draft regulatory guide proposed to endorse IEEE Std. 603-1991 (including the correction sheet dated January 30, 1995). Because there were no adverse public comments to Revision 1 to Regulatory Guide 1.153, the Commission believes that there is general public consensus that IEEE Std. 603-1991 provides acceptable criteria for safety systems in nuclear power plants. Discussion The direct final rule incorporates a national consensus standard, IEEE Std. 603-1991, for establishing minimal functional and design requirements for power, instrumentation, and control portions of safety systems for nuclear power plants into NRC regulations. This action is consistent with the provisions of the National Technology Transfer and Advancement Act of 1995, Pub. L. 104-113, which encourages Federal regulatory agencies to consider adopting industry consensus standards as an alternative to de novo agency development of standards
4 affecting an industry. This action is also consistent with the NRG policy of evaluating the latest versions of national consensus standards in terms of their suitability for endorsement by regulations or regulatory guides. Currently, 10 CFR 50.55 a (h) specifies that "protection systems" for plants with construction permits issued after January 1, 1971, must meet the requirements in IEEE Std. 279 in effect on the formal docket date of the application for a construction permit. IEEE Std. 279, states that a "protection system" encompasses all electric and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating those signals associated with the protective function. These signals include those that actuate reactor trip and that, in the event of a serious reactor accident, actuate engineered safeguards such as containment isolation, core spray, safety injection, pressure reduction, and air cleaning. "Protective Function" is defined by IEEE Std. 279, as" the sensing of one or more variables associated with a particular generating station condition, signal processing, and the initiation and completion of the protective action at values of the variables established in the design bases." IEEE Std. 603-1991, uses the term "safety systems" rather than "protection systems." A "safety system" is defined by IEEE Std. 603-1991, as "a system that is 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 down the reactor and maintain it in a safe shut down condition, or (iii) 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. A "safety function" is defined by IEEE Std. 603-1991, as "one of the processes or conditions
5 (for example, emergency negative reactivity insertion, post accident heat removal, emergency core cooling, post-accident radioactivity removal, and containment isolation) essential to maintain plant parameters within acceptable limits established for a design basis event." The Commission considers that the systems covered by IEEE Std. 603-1991 and IEEE Std. 279-1971 are the same. Therefore, for purposes of paragraph (h) of 10 CFR 50.55a, "protection systems," and *"safety systems" are synonymous. The Commission notes that these two terms are also synonymous with the term "safety-related systems," used elsewhere in e Commission's regulations. Therefore, licensees are expected to apply IEEE Std. 279-1971 and IEEE Std. 603-1991, as appropriate, to "safety-related systems." This rule mandates the use of IEEE Std. 603-1991 (including the correction sheet dated January 30, 1995) for future nuclear power plants, including final design approvals, design certifications and combined licenses under 10 CFR Part 52. Current licensees may continue to meet the requirements set forth in the edition or revision of IEEE Std. 279 in effect on the formal date of their application for a construction permit or may, at their option, use IEEE Std. 603-1991, provided they comply with all applicable requirement& for making changes to their licensing basis. However, changes to protection systems in operating nuclear power plants initiated on or after January 1, 1998 must meet the requirements in IEEE Std. 603-1991 . For purposes of this rule, "changes" to protection systems include (i) modifications, augmentation or replacement of protection systems permitted by license amendments, (ii) changes made by the licensees pursuant to procedures in 10 CFR 50.59, and (iii) plant-specific departures from a design certification rule under 10 CFR Part 52. In-kind (like-for-like) replacement of protection system components are not considered changes to the protection systems. Section 3 of IEEE Std. 603-1991 references several industry codes and standards. If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a
6 method acceptable to the Commission of meeting a regulatory requirement as described in the regulatory guide. If a referenced standard has not been endorsed in a regulatory guide, the licensees and applicants may consider and use the information in the referenced standard consistent with current regulatory practices. Electronic Access You may also provide comments via the NRC's interactive rulemaking website through the NRC home page (http;//www.nrc.gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905 (e-mail: CAG@nrc.gov). Finding of No Environmental Impact: Availability of Environmental Assessment The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule would not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environment impact statement is not required. The Commission has prepared an Environmental Assessment supporting this finding of no significant environmental impact. The NRC has sent a copy of the environmental assessment and a copy of the Federal Register Notice to every State liaison officer and requested their comments on the
7 environmental assessment. The environmental assessment is available for inspection at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Also, the NRC has committed itself to complying in all its actions with the Presidential Executive Order # 12898-Federal Actions to Address Environmental Justice in Minority Populations and Low-Income Populations, dated February 11, 1994. Therefore, the NRC also has determined that there are no disproportionate, high, and adverse impacts on minority and low-income populations. The NRC uses the following working definition of environmental justice: environmental justice means the fair treatment and meaningful involvement of all people, regardless of race, ethnicity, culture, income, or educational level with respect to the development, implementation, and enforcement of environmental laws, regulations and policies. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et seq.). Existing requirements were approved by the Office of Management and Budget, approval No. 3150-0011. Public Protection Notification If a document used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, an information collection .
8 Regulatory Analysis The Commission has prepared a regulatory analysis which shows that the proposed amendment does not impose any new requirements or costs on current licensees who do not make changes to safety systems. However, licensees planning or proposing changes to power and instrumentation & control systems will be impacted because they will be required to meet the requirements of IEEE Std. 603-1991 for the changes even though the remainder of the plant power and l&C systems are only required to meet their current licensing basis. The draft regulatory analysis is available for inspection in the NRC Public Document Room, 2120 L Street, NW., Washington, D.C. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b}}, the Commission certifies that this rule will not have a significant economic impact on small entities. This rule affects only the operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the small business size standards adopted by the NRC (10 CFR 2.810). Since these companies are dominant in their service areas, this rule does not fall within the purview of the Act. Backfit Analysis The rule requires applicants and holders of new construction permits, new operating
9 licenses, new final design approvals, new design certifications and combined licenses to comply with IEEE Std. 603-1991 (including the correction sheet dated January 30, 1995). Changes to protection systems in existing operating plants initiated on or after January 1, 1998 must meet the requirements of IEEE Std. 603-1991. IEEE Std. 279 will continue to apply to existing nuclear power plants that do not make any changes to their protection systems, but the rule
- permits the licensee the option of meeting IEEE Std. 603-1991.
The backfit rule was not intended to apply to regulatory actions which change expectations of prospective applicants, and therefore the backfit rule does not apply to the portion of the rule applicable to new construction permits, new operating licenses, new final design approvals, new design certifications and combined licenses. This rule does not change the licensing basis (i.e., IEEE Std. 279) for plants that do not intend to make any changes to their power and instrumentation and control systems. However, the rule would require future changes to existing power and instrumentation and control portions of protection systems to comply with the new standard. This would not be considered a backfit, since the changes are voluntarily initiated by the licensee, or separately imposed by the NRC after a separate backfit analysis. This is consistent with past NRC practice and the discussions on backfitting in "Value-Impact Statement" prepared for Revision 1 to Regulatory Guide 1.153. A copy of the Value-Impact Statement is available for inspection or copying for a fee in the Commission's Public Document Room at 2120 L Street NW., Washington, DC, under Task DG-1042. In summary, the NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this direct final rule because it does not impose any backfits as defined in 10 CFR 50.109(a)(1) and, therefore, a backfit analysis has not been prepared for this direct final rule.
10 Small Business Regulatory Enforcement Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996, the NRC has determined that this action is not a major rule and has verified this determination with the Office of Information and Regulatory Affairs of 0MB. List of Subjects in 10 CFR Part 50 Antitrust, Classified Information, Criminal penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, and Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganizations Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR Part 50. Part 50 - Domestic Licensing of Production and Utilization Facilities
- 1. The authority citation for Part 50 continues to read as follows:
AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936,937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846). Section 50.7 also issued under Pub. L. 95-601, sec, 10, 92 Stat. 2951 (42 U.S.C.
11 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, and 50.54 (dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138), Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235), Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80 - 50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
- 2. In § 50.55a, paragraph (h) is revised to read as follows:
§ 50.55a Codes and standards. (h) Protection and Safety Systems. (1) IEEE Std. 603-1991 and the correction sheet dated January 30, 1995, which are referenced in paragraph (h)(3) and (h)(4), are approved for incorporation by reference by the Director of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR Part 51. A notice of any changes made to the material incorporated by reference will be published in the Federal Register. Copies of IEEE Std. 603-1991 may be purchased from the Institute of Electrical and Electronics Engineers Service Center, 445 Hoes Lane, Piscataway, NJ 08855. It is also available for inspection at the NRC Library, 11545
12 Rockville Pike, Rockville, MD 20852-2738, and at the Office of the Federal Register, 800 North Capital Street, NW, Suite 700, Washington, DC. IEEE Std. 279, which is referenced in paragraph (h)(2) of this section was approved for incorporation by reference by the Director of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR Part 51. Copies of this standard are also available as indicated for IEEE Std. 603-1991. (2) Definitions. (I) For purposes of this paragraph the terms "protection systems," "safety systems," and "safety-related systems" are synonymous. (ii) Changes to protection systems include modification, augmentation or replacement of protection systems permitted by license amendments, changes to protection systems made by licensees pursuant to 10 CFR 50.59, and plant specific departures from a design certification rule under 10 CFR Part 52. (3) Protection systems. For nuclear power plants with construction permits issued after January 1, 1971, but prior to January 1, 1998, protection systems must meet the requirements set forth either in the Institute of Electrical and Electronics Engineers (IEEE) Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," or in IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," and the correction sheet dated January 30, 1995. However, changes to protection systems initiated on or after January 1, 1998 must meet the requirements set forth in IEEE Std. 603-1991, and the correction sheet dated January 30, 1995.
13 (4) Safety systems. For construction permits, operating licenses, final design approvals, design certifications and combined licenses issued on or after January 1, 1998, safety systems must meet the requirements set forth in IEEE Std. 603-1991, and the correction sheet, dated January 30, 1995. qv:! Dated at Rockville, Maryland, this _ _ _ _ day of October, 1997. For the Nuclear Regulatory Commission. Jo S , etary of the Commission.}}