ML23151A481

PR-050 - 56FR03796 - Codes and Standards for Nuclear Power Plants

ML23151A481
Person / Time
Issue date:	01/31/1991
From:	Taylor J NRC/EDO
To:
References
PR-050, 56FR03796
Download: ML23151A481 (1)

Text

DOCUMENT DATE:

TITLE:

CASE

REFERENCE:

KEYWORD:

ADAMS Template: SECY-067 01/31/1991 PR-050 - 56FR03796 - CODES AND STANDARDS FOR NUCLEAR POWER PLANTS PR-050 56FR03796 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE:

PR-050 OPEN ITEM (Y/N) N RULE NAME:

CODES AND STANDARDS FOR NUCLEAR POWER PLANTS PROPOSED RULE FED REG CITE:

56FR03796 PROPOSED RULE PUBLICATION DATE:

01/31/91 ORIGINAL DATE FOR COMMENTS: 05/27/91 NUMBER OF COMMENTS:

EXTENSION DATE:

I I

30 FINAL RULE FED. REG. CITE: 57FR34666 FINAL RULE PUBLICATION DATE: 08/06/92 NOTES ON FINAL RULE SIGNED BY EDO.

FILE LOCATED ON Pl.

STATUS F RU!JE TO FIND THE STAFF CONTACT OR VIEW THE RULEMAKING HISTORY PRESS PAGE DOWN KEY HISTORY OF THE RULE PART AFFECTED: PR-050 RULE TITLE:

CODES AND STANDARDS FOR NUCLEAR POWER PLANTS

-PROPOSED RULE SECY PAPER:

FINAL RULE SECY PAPER:

PROPOSED RULE SRM DATE:

FINAL RULE SRM DATE:

DATE PROPOSED RULE I

I SIGNED BY SECRETARY:

01/16/91 DATE FINAL RULE I

I SIGNED BY SECRETARY:

10/20/92 STAFF CONTACTS ON THE RULE CONTACTl: GILBERT C. MILLMAN CONTACT2:

MAIL STOP: NS-217-B PHONE: 492-3848 MAIL STOP:

PHONE:

DOCKET NO. PR-050 (56FR03796)

DATE DOCKETED 01/29/91 04/11/91 04/15/91 04/15/91 04/15/91 04/16/91 04/16/91 04/16/91 04/16/91 04/16/91 04/16/91 04/17/91 04/18/91 04/18/91 DATE OF DOCUMENT 01/16/91 04/05/91 04/11/91 04/09/91 04/15/91 04/11/91 04/12/91 04/16/91 04/15/91 04/15/91 04/16/91 04/16/91 04/15/91 04/16/91 In the Matter of CODES AND STANDARDS FOR NUCLEAR POWER PLANTS TITLE OR DESCRIPTION OF DOCUMENT FEDERAL REGISTER NOTICE - PROPOSED RULE COMMENT OF NORTHERN STATES POWER COMPANY (GERALD H. NEILS, EXECUTIVE ENGR.) (

1)

COMMENT OF PUBLIC SERVICE ELECTRIC AND GAS COMPANY (B. A. PRESTON, MANAGER) (

2)

COMMENT OF OHIO CITIZENS FOR RESPONSIBLE ENERGY,INC (SUSAN L. HIATT) (

3)

COMMENT OF NUMARC (JOE F. COLVIN) (

4)

COMMENT OF WASHINGTON PUBLIC POWER SUPPLY SYSTEM (G.C. SORENSEN, MANAGER) (

5)

COMMENT OF FLORIDA POWER & LIGHT (FPL)

(W.H. BOHLKE, VICE PRESIDENT) (

6)

COMMENT OF BWR OWNERS' GROUP (GEORGE J. BECK, CHAIRMAN) (

7)

COMMENT OF ENTERGY OPERATIONS (ARKANSAS/WATERFORD)

(GERALD W. MUENCH, VICE PRESIDENT) (

8)

COMMENT OF ALABAMA POWER COMPANY (J.D. WOODWARD) (

COMMENT OF GEORGIA POWER COMPANY (W.G. HAIRSTON, III) (

10)

COMMENT OF GPU NUCLEAR CORPORATION (J. KNUBEL) (

11)

COMMENT OF GULF STATES UTILITY COMPANY (GSU)

(W.H. ODELL, MANAGER-OVERSIGHT) (

12)

COMMENT OF TENNESSEE VALLEY AUTHORITY (TVA)

(E.G. WALLACE, MANAGER) (

13)

9)

DOCKET NO. PR-050 (56FR03796)

DATE DOCKETED 04/18/91 04/18/91 04/19/91 04/19/91 04/22/91 04/22/91 04/22/91 04/22/91 04/22/91 04/22/91 04/22/91 04/22/91 04/22/91 04/23/91 04/26/91 04/ 26/91 05/03/91 08/17/92 DATE OF TITLE OR DOCUMENT DESCRIPTION OF DOCUMENT 04/12/91 04/11/91 04/15/91 04/12/91 04/15/91 04/17/91 04/15/91 04/12/91 04/15/91 04/15/91 04/17/91 04/15/91 04/18/91 04/22/91 04/18/91 04/22/91 04/29/91 07/15/92 COMMENT OF WOLF CREEK NUCLEAR OPERATING CORPORATION (FORREST T. RHODES, VICE PRESIDENT) (

14)

COMMENT OF ILLINOIS DEPARTMENT OF NUCLEAR SAFETY (THOMAS W. ORTCIGER, DIRECTOR) (

15)

COMMENT OF SOUTH CAROLINA ELECTRIC & GAS CO.

(JOHN L. SKOLDS, VICE PRESIDENT) (

16)

COMMENT OF NIAGARA MOHAWK (CARL D. TERRY, VICE PRESIDENT) (

17)

COMMENT OF WISCONSIN PUBLIC SERVICE CORP.

(K. H. EVERS, MANAGER) (

18)

COMMENT OF CONSUMERS POWER CO.

(DENNIS R. HUGHES, DIRECTOR) (

19)

COMMENT OF VERMONT YANKEE NUCLEAR POWER CORP.

(J. P. PELLETIER, VICE PRESIDENT) (

20)

COMMENT OF ROWLEY CONSULTANTS (C. WESLEY ROWLEY, P.E.) (

21)

COMMENT OF DUKE POWER CO. (M. S. TUCKMAN) (

22)

COMMENT OF (W. G. GATES, DIVISION MANAGER) (

23)

COMMENT OF FLORIDA POWER CORP.

(P.M. BEARD, JR., SR. VICE PRESIDENT) (

24)

COMMENT OF NEBRASKA PUBLIC POWER DISTRICT (G.R. HORN, NUCLEAR POWER GROUP MGR. ) (

25)

COMMENT OF PENNSYLVANIA POWER & LIGHT CO.

(H. W. KEISER) (

26)

COMMENT OF YANKEE ATOMIC ELECTRIC CO (D.W. EDWARDS, DIRECTOR) (

27)

COMMENT OF DUKE POWER COMPANY (M.S. TUCKMAN, VICE PRESIDENT) (

28)

COMMENT OF BOSTON EDISON (G.W. DAVIS) (

29)

COMMENT OF TOLEDO EDISON (DONALD C. SHELTON, VICE PRESIDENT) (

30)

FEDERAL REGISTER NOTICE - FINAL RULE

COCK[i ED USNRC

• 92 AUG 17 P 6 :34 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AD05

[7590-01]

Codes and Standards for Nuclear Power Plants AGENCY:

Nuclear Regulatory Commission.

ACTION:

Final rule.

SUMMARY

The Commission is amending its regulations to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section XI, Division 1, of the ASME Code.

The final rule imposes an augmented examination of reactor vessel shell welds and separates the requirements for inservice testing 1

inservice inspection by placing the requirements for inservice testing in a separate paragraph.

The ASME Code addenda and edition incorporated by reference provide updated rules for the construction of components of light-water-cooled nuclear power plants, and for the inservice inspection and inservice testing of those components.

This final rule permits the use of improved methods for construction, inservice inspection, and inservice testing of nuclear power plant components; requires expedited implementation of the expanded reactor vessel shell weld examinations specified in the 1989 Edition of Section XI; and more clearly distinguishes in the regulations the requirements for inservice testing from those for inservice inspection.

EFFECTIVE DATE: (30 days after publication in the Federal Register).

The incorporation by reference of certain publications listed in the regulations is approved by the Office of the Director of the Office of the Federal Register as of (30 days after publication in the Federal Register).

FURTHER INFORMATION CONTACT:

Mr. G. C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: {301) 492-3848.

SUPPLEMENTARY INFORMATION:

Background

On January 31, 1991 {56 FR 3796), the Nuclear Regulatory Cormnission published in the Federal Register a proposed amendment to its regulation, 10 CFR Part 50, RDomestic Licensing of Production and Utilization Facilities,"

to update the reference to editions and addenda of the American Society of 2

Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

This proposed amendment would revise§ 50.55a to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section Ill, Division 1, of the ASME Code, and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section XI, Division 1, of the ASME Code, with a specified modification.

The modification would require implementation of certain requirements for containment isolation valve (CIV) testing that appear in Section XI Subsection IWV prior to the 1988 Addenda, but which do not appear in the later addenda.

The amendment would impose an augmented examination of reactor vessel shell welds, and separate in the regulations the requirements for inservice testing from those for inservice inspection by placing the requirements for inservice testing in a separate paragraph.

Su11111ary of Comments Interested parties were invited to submit written comments for consideration in connection with the proposed amendment by April 16, 1991.

Co11111ents were received from 29 separate sources. These sources consisted of 23 utilities, one service organization representing four nuclear power plants,*

the Nuclear Management and Resources Council (NUMARC), one owners group (BWR Owners Group (BWROG)), one state entity (Illinois Department of Nuclear Safety (IONS)), one public citizens group (Ohio Citizens for Responsible Energy (OCRE)), and one independent consultant.

The submitted connents generally addressed one -of the following subject areas: (1) the incorporation by reference of the specified later addenda and edition of Section III, Division 1, and Section XI, Division 1, of the ASME 3

Code into§ 50.55a; (2) the endorsement of comments submitted by NUMARC; (3) the proposed modification to Section XI Subsection IWV rules for CIV testing; (4) the proposed augmented reactor vessel examination; (5) the separation of the rules for inservice inspection and inservice testing; (6) the existing scope of§ 50.55a for pump and valve testing; and (7) the potential endorsement in§ 50.55a of ASME/ANSI OM Part 4 on snubbers.

Those who co1T111ented on the updating of existing references to Section III and Section XI of the ASME Code in§ 50.55a generally noted their approval.

One commentor, however, expressed significant concern with the new provision initially specified in the Section XI 1988 Addenda which expands the existing requirement to examine one circumferential and one longitudinal reactor vessel shell weld during the 2nd and subsequent inspection intervals to essentially 100 percent of all reactor vessel shell weld during those intervals.

Volumetric examination of all reactor vessel shell welds during the first inspection interval has been a requirement in Section XI since the 1975 Addenda.

The commentor believes that the expanded examination is unnecessary and that examination efforts should focus on the beltline welds or welds that exceed a specified fluence level. The NRC agrees with the ASME action to expand the reactor vessel examination on the basis that the importance of the reactor vessel, and previous unexpected cracking of primary coolant pressure boundary components, requires that the expanded examinations be performed to ensure the integrity of the reactor vessel. The importance of reactor vessel integrity in protecting the public health and safety demands that periodic, comprehensive inservice examinations of the reactor vessel be made to ensure that structural degradation, if it occurs, does not go undetected. Although 4

the beltline welds do receive the highest radiation, there is simply no assurance that service induced cracking would be limited to those welds.

An examination once every ten years of essentially 100 percent of all reactor vessel shell welds is both reasonable and necessary.

The comments submitted by NUMARC relate to: (1) the proposed endorsement of a later edition and addenda of the ASME Code, which NUMARC considers to be a positive step; (2) the proposed modification to Section XI Subsection IWV (i.e., the reference to Part 10 of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987 (OM Part 10)), which NUMARC considers to be inappropriate and unnecessary on the basis that 10 CFR Part 50, Appendix J testing is adequate; *(3) the proposed augmented reactor vessel examination, which NUMARC recognizes to be important, but suggests that more flexibility be incorporated into the implementation provisions; and (4) the scope of§ 50.55a which NUMARC believes should not be influenced by Generic Letter 89-04. Approximately one-half of the utility commentors specifically endorsed the corrments by NUMARC.

In general, comments from the other utilities were consistent with one or more of the convnents from NUMARC.

The co11111ents from NUMARC are discussed below, along with co1T111ents from others on the same subject.

Most of the convnents addressed, in part, the proposed modification to Section XI Subsection IWV rules for containment isolation valve testing.

Utility comments supported the NUMARC convnent, which expressed the belief that the current Appendix J containment leakage testing program already provides an adequate basis for assessing and controlling containment leakage and that the modification could result in a valve having to be declared inoperable 5

immediately, in spite of the fact that the total containment leakage may be substantially less than allowable.

NUMARC suggested t~at, in lieu of reinstating requirements for specific valves, NRC recommend to the ASME Operations & Maintenance (O&M) Committee that it perform a comprehensive review of the testing requirements for containment isolation valves and acceptance standards for those tests.

IONS agreed with the NRC position that the requirements for leakage rate analysis and provisions for corrective action should be maintained, but believed that it would be less confusing for licensees if those requirements were incorporated into the existing requirements for Type C testing in Appendix J.

OCRE strongly supported the action by NRC to modify the Section XI rules for CIV testing.

The NRC concern that resulted in the proposed modification to Section XI Subsection IWV stems from the findings of two reviews and a follow-on study of Appendix J leak test results. The overall findings show that valve leakage is the primary contributor to occurrences of containment unavailability and that such occurrences generally involve small, rather than large, leaks.

Risk to the public from small leakage events is very low, but the NRC is concerned that eliminating the e~isting Section XI requirement to analyze leakage rates and to take corrective action in the event of abnonnally high leakage rates for those CIVs that do not provide a reactor coolant system pressure isolation function could reduce the ability to detect degrading valves and, thereby, could permit an unacceptable reduction in the present safety margin associated with the leak tight integrity of those CIVs and, thereby, the containment.

6

It was specifically noted in the proposed rule that the NRC was interested in receiving comments on the discussed basis for and content of the proposed modification, and was particularly interested in receiving comments that would provide insight and justification, based upon plant experiences, relative to the need for revising or possibly eliminating the proposed modification.

Many comments were received that express concern with the proposed modification.

However, these corrments, which generally state the opinion that Appendix J requirements are adequate and sufficient with regard to ensuring containment integrity, are of a qualitative nature and no specific plant data or operational experiences were provided or referenced that updated the results of the earlier studies.

No additional substantive information was provided for the NRC to consider relative to the need for revising or possibly eliminating the proposed modification.

It has not been demonstrated, by analysis of more recent and comprehensive containment leakage test data, that containment leakage integrity can be maintained at an acceptable level without continued implementation of the existing Appendix J valve leak rate test program in conjunction with the Section XI requirement for analysis of leak rates.

Consistent with the comment by NUMARC, the NRC staff discussed the basis for OM Part 10 CIV testing requirements with representatives from the ASME O&M Committee.

Based on these discussions and in concert with the O&M Convnittee organization, the O&M Committee has initiated action to (1) perform a comprehensive review of OM Part 10 CIV testing requirements and acceptance standards and (2) develop a basis document that would provide, as a minimum, a documented basis for not including the requirements for analysis of leakage 7

rates and corrective actions in OM Part 10 for those CIVs that do not provide a reactor coolant system pressure isolation function.

The NRC will reevaluate the need for the modification to Section XI Subsection IWV, following review of this basis document.

It is anticipated that this will occur as part of a future rulemaking proceeding that will address the incorporation by reference of the ASME O&M Code into§ 50.55a.

In the meantime, this final rule incorporates by reference the 1988 Addenda and 1989 Edition of Section XI, Division 1, with a specified modification for CIV testing that is provided in a new§ 50.55a(b)(2)(vii).

The modification substantially preserves the existing requirements for analysis of leakage rates and corrective actions that exist in Subsection IWV prior to the 1988 Addenda.

Specifically, the modification requires that licensees implement the requirements of Paragraph 4.2.2.3(e), "Analysis Qf Leakage Rates," of Part 10 and Paragraph 4.2.2.3(f), "Corrective Action," of Part 10, in addition to the requirements of Paragraph 4.2.2.2 of Part 10, for all Category A valves that are CIVs, regardless of whether or not they provide a reactor coolant system pressure isolation function.

Because paragraph 4.2.2.3(e) of Part 10 is specified in the modification rather than the existing IWV-3426, the existing Section XI requirement is somewhat relaxed by permitting valve combinations rather than specific valves to be analyzed.

This recognizes that, in the past, requests for relief have been granted where design constraints necessitate testing combinations of valves with permissible leak rate limits applied to valve groups.

The specified modification does not require the present practice of trending NPS 6 and larger valves because that requirement has not been carried from IWV-3427(b) to OM Part 10.

8

Section XI Subsection IWV (1988 Addenda and 1989 Edition), Subsection IWP (1988 Addenda and 1989 Edition), and Subsection IWF (1987 Addenda, 1988 Addenda, and 1989 Edition) reference ASME/ANSI OM Part 10, ASME/ANSI OM Part 6, and ASME/ANSI OM Part 4, respectively. During preparation of this final rule, it was recognized that Table IWA-1600-1 in the applicable Section XI addenda and edition specifies a nonexistent revision for OM Part 10 and Part 6, and does not specifically identify the applicable revision for OM Part 4.

The Section XI Subco111T1ittee on Inservice Inspection has taken action to correct the revision reference, which, for these standards, should be the ANSI/ASME OMa-1988 Addenda to the ASME/ANSI OM-1987 Edition.

To ensure that licensees are aware of the correct revision reference to the OM standards, an additional modification,§ 50.55a(b)(2)(viii), has been added to specify that the OMa-1988 Addenda is the applicable revision to the OM-1987 Edition for OM Part 4, Part 6, and Part 10 when using the noted Section XI addenda and edition.

The NUMARC comment relative to the proposed augmented examination of the reactor vessel indicates an understanding of the NRC position on the need for this examination, but notes concern with the specifics of the proposed implementation.

Specifically, NUMARC expresses concern that: (1) better utilization of available inspection resources could be accomplished by limiting application of the augmented inspection program to the reactor vessel beltline shell welds, or by limiting implementation of the augmented examination to reactor vessel shell welds that exceed a specific neutron flux exposure (this comment differs from the one utility comment noted above relative to updating later edition and addenda of Section XI in that it only 9

refers to the augmented examination); (2) tooling for the older Boiling Water Reactors (BWRs) may generally not be available in the time-frame needed; (3) only those reliefs which address the scope and extent of shell weld examinations should be revoked, and they should be revoked on a plant specific basis; and (4) the NRC should state its willingness to accept requests for specific new exemptions, based on the availability of suitable equipment and technology at the time of the scheduled inspection and the appropriate technical justification.

Other corrunents on the augmented examination include those from:

BWROG, which noted concern for those plants close to the end of the current intervai that could not practically incorporate the augmented examination into the current interval and would have to perform that examination during the first period of the next interval (Note: The deferred augmented examination may be used as a substitute for the reactor vessel shell weld examination normally scheduled for the interval in which the deferred examination was performed

(§ 50.55a(g)(6)(ii)(A)ill, therefore, the impact of deferring the augmented examination will be reduced); IONS, which strongly supports the NRC position regarding the augmented examination of the reactor vessel; and OCRE, which also strongly supports the augmented examination and notes that the examination will not only provide an additional assurance of safety, but will aid in understanding aging degradation phenomena which will assist licensees that wish to pursue license renewal.

The NRC position with regard to the augmented examination of the reactor vessel, as previously stated in the Supple~entary Information to the proposed 10

rule, is that degradation of reactor vessel materials has become more of a concern recently, because (1) results from irradiation surveillance material tests show that certain reactor vessel materials undergo greater radiation damage than previously expected, (2) indications from operational data show that stress corrosion cracking of BWR reactor vessels is more probable than was thought several years ago, and_ (3) significant service induced cracking has occurred in large vessels (i.e., pressurizer, steam generators) designed and fabricated to the ASHE Code.

It is the judgment of the NRC that, because of new information and previous limited examinations of reactor vessels, there may exist a substantially greater potential for reactor vessel degradation, in all areas of the reactor vessel, than previously considered and that maintenance of the level of protection presumed by the regulations requires more than compliance to existing regulatory requirements.

The NRC has determined that the augmented examination of reactor vessels will result in a substantial increase in the overall protection of the public health and safety, and that the costs of implementation are justified in view of the increased protection. The backfit analysis required by§ 50.109, "Backfitting," is provided as part of the regulatory analysis that supports this final rule.

However, the NRC agrees with comments that additional flexibility and specificity will improve implementation of the augmented examination of reactor vessel examination.

To this end, the augmented examination of reactor vessel shell welds specified in this final rule includes the following new provisions and clarifications: (1) the revocation of previously granted reliefs is limited to those reliefs that deal with the extent of volumetric 11

examination of reactor vessel shell welds; (2) the augmented examination will be performed in accordance with the Section XI edition and addenda applicable to the inspection interval in which the examination is actually performed; (3)

"essentially 100%" as used in§ 50.55a(g)(6)(ii)(A) means "more than 90 percent of the examination volume of each weld, where the reduction in examination volume is due to interference from another component, or part geometry;" (4) licensees that defer the augmented examination to the next interval are permitted to retain all existing approved reliefs for the current interval; (5) licensees with fewer than 40 months remaining in the inspection interval in effect when the rule becomes effective are permitted to extend the interval in accordance with the provisions of Section XI (1989 Edition)

IWA-2430(d); (6) licensees that are unable to satisfy completely the requirements for the augmented examination may request to perform alternate examinations in accordance with§ 50.55a(g)(6)(ii)(A)fil. These items are addressed individually in the discussion below regarding provisions of the augmented reactor vessel shell weld examination.

Section 50.55a{g){6){ii) addresses augmented inservice inspection programs for those systems and components for which the Commission determines that added assurance of structural reliability is necessary.

For that purpose, and consistent with the discussion in this final rule,

§ 50.55a(g){6){ii)(A) has been added to require expedited implementation of the reactor vessel shell weld examinations specified in the 1989 Edition of Section XI, Division 1, in Item Bl.IO, 11Shell Welds,* of Examination Category 8-A, 11 Pressure Retaining Welds in Reactor Vessel, 11 in Table 2500-1 of Subsection IWB, 11Requirements for Class 1 Components of Light-Water Cooled 12

Power Plants."

In order to ensure the applicability of the new augmented examination to all licensees,§ 50.55a(g)(6)(ii)(A)ill revokes all previously granted reliefs relating to the extent of volumetric examination of the reactor vessel shell welds that apply to examinations for the inservice inspection interval that is in effect when the rule becomes effective subject to a specified modification.

Limiting the revocation of previously granted reliefs to those that deal with the extent of the volumetric examination pennits the retention of those approved reliefs that deal with issues such as specification of calibration blocks.

Licensees that choose to defer the augmented examination to the next interval in accordance with§ 50.55a(g)(6)(ii)(A)ill should note that paragraph.LiY.l of that section modifies the revocation of approved reliefs to permit retention of previously approved reliefs for the current interval when the augmented examination is deferred. This provision recognizes that plants that previously received relief from the Section XI reactor vessel shell weld examination and satisfy the condition to defer the augmented examination may find it impractical to implement the Section XI examination during the current inspection interval.

Section 50.55a(g)(6)(ii)(A).ill requires all licensees to implement the specified augmented examination of reactor vessels during the inspection interval in effect when this rule becomes effective, subject to conditions specified in§ 50.55a(g)(6)(1i)(A)ill and ill-Section 50.55a(g)(6)(ii)(A)ill specifically permits the use of the augmented examination, when not deferred, as a substitute for the reactor vessel shell weld examinations scheduled for 13

the inspection interval in effect when the rule becomes effective, and specifies that, for the purpose of this rule, "essentially 100 percent" as used in Table IWB-2500-1 means "more than 90 percent of the examination volume of each weld, where the reduced examination volume is due to interference from another component, or part geometry." This is consistent with Section XI Code Case N-460, which previously has been approved for use in Regulatory Guide 1.147.

It is recognized that it may be necessary to implement a combination of internal and external diameter examinations to achieve nessentially 100%"

examination volume coverage for each weld. A clarification has been included in this section to note that the augmented examination may be used as a substitute for the reactor vessel shell weld examination in the interval in effect when the rule becomes effective when the augmented examination is not deferred. This is a reinforcement of§ 50.55a(g)(6)(ii)(A)ill, as it appears in both the proposed and final rule, which specifies that the deferred examination may not be used as a substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection interval in effect when the rule becomes effective.

The NRC recognizes that plants with fewer than 40 months remaining in the inspection interval when this rule becomes effective may find it impractical to implement the augmented examination of the reactor vessel during that inspection interval. Therefore,§ 50.55a(g)(6)(ii)(A)(l) permits plants with fewer than 40 months remaining in the inspection interval when this rule becomes effective to defer the augmented examination until the first period of the next inspection interval. However, this same paragraph specifically prohibits the use of the deferred augmented examination as a substitute for 14

reactor vessel shell weld examinations scheduled for the inspection interval in effect when the rule becomes effective. The intent is to ensure that the examinations are deferred only when necessary and not to have the rule encourage a 40-month delay in reactor vessel shell weld examinations.

Further, § 50.55a(g)(6)(ii)(A)(J) permits using the deferred examination, with a condition, as a substitute for reactor vessel shell weld examinations scheduled for the inspection interval in which the deferred examinations are performed.

The condition is that subsequent reactor vessel shell weld examinations for successive inspection intervals be performed in the first period of the inspection interval. This condition is necessary to prevent a potential 160-month gap between reactor vessel shell weld examinations. This gap would occur if a plant used the deferred examination perfonned in the first period as a substitute for the scheduled examination and then deferred the examination for the next inspection interval to the end of that interval as permitted by Section XI.

In addition, this section specifies that licensees with fewer than 40 months remaining in the inservice inspection interval in effect when the rule becomes effective may extend that interval in accordance with the provisions of Section XI (1989 Edition) IWA-2430(d) to pennit implementation of the augmented examination during the current interval. It is not the intent of the NRC to permit licensees in the second period of an inspection interval to reduce the interval length for the purpose of "being within 40 months of the end of the interval" and, thereby, deferring the augmented examination to the first period of the subsequent interval.

Section 50.55a(g)(6)(ii)(A)ill specifies that a licensee that has either completed or has scheduled an inspection of essentially 100 percent of the 15

length of all Examination Category 8-A shell welds during the inservice inspection interval in effect when the rule becomes effective does not have to implement the required augmented examination of the reactor vessel shell welds.

Primarily, this paragraph is intended to permit licensees who are in the first inspection interval to use the essentially 100 percent reactor vessel shell weld examination required for that interval by Section XI to satisfy the requirement for the augmented examination of the reactor vessel.

The technical objective of the augmented examination will be accomplished under these conditions. These licensees will continue to apply the current requirements of§ 50.55a(g)(4) until the next inspection interval when future examinations will be performed based on ASHE Section XI, 1989 Edition, or later Code edition and addenda specified in§ 50.55a(b).

The augmented examination specified in§ 50.55a(g)(6)(ii)(A) is not an ASME Code requirement.

It is a requirement specifically developed and additionally imposed by the Commission. Therefore, except for the specific provisions in§ 50.55a(g)(6)(ii)(A)ill and ill that permit using the augmented examination as a substitute for Section XI required reactor vessel shell weld examinations, the closing out of an inservice inspection interval is not dependent on completion of the augmented examination.

In the specific instance where the augmented examination is deferred to the first period of the next inspection interval, the current inspection interval could be closed out relative to reactor vessel shell weld examinations by implementing the regularly scheduled reactor vessel shell weld examinations as modified by previously approved applicable relief requests for that interval.

16

The NRC recognizes that, as noted by commentors, there may exist conditions that prevent licensees from completely satisfying the requirements for the augmented reactor vessel shell weld examination as specified in

§ 50.55a(g)(6)(ii)(A).

For this reason, § 50.55a(g)(6)(ii)(A)ill has been added to permit licensees that make a determination that they are unable to completely satisfy the specified augmented examination to propose and use alternatives that have been authorized by the NRC's Director of the Office of Nuclear Reactor Regulation.

This final rule amends§ 50.55a to separate the requirements for inservice testing from those for inservice inspection by moving the requirements for inservice testing to a separate paragraph.

Previously,

§ 50.55a(g), "Inservice inspection requirements,R specified the requirements for (1) preservice and inservice examinations for Class 1, Class 2, and Class 3 components and their supports, (2) system pressure tests for Class 1, Class 2, and Class 3 components, and (3) inservice testing of Class 1, Class 2, and Class 3 pumps and valves.

In order to emphasize the importance of inservice testing and to distinguish more clearly its requirements from those of inservice inspection, this final rule moves the requirement for inservice testing from§ 50.55a(g), "Inservice inspection requirements,R to a separate (previously reserved)§ 50.55a(f), which is titled "Inservice testing requirements.* All existing requirements for inservice examination and system pressure testing are retained in§ 50.55a(g).

There is overall favorable acceptance of the separation of the requirements in the regulation for inservice testing and for inservice 17

inspection.

It is generally believed by the commentors, as it is believed by the NRC, that the separation serves to clarify and emphasize the requirements for inservice testing.

Two administrative changes were made in the development of§ 50.55a(f) relative to existing§ 50.55a(g). First,

§ 50.55a(f)(6)(ii) has been added to indicate the Commission's intent to impose an augmented inservice testing program if added assurance of operational readiness is deemed necessary. This paragraph only indicates intent and does not impose a specific requirement.

It does parallel the existing§ 50.55a(g)(6)(ii) which specifies that the Convnission may require an augmented inservice inspection program for systems and components for which it deems that added assurance of structural reliability is necessary.

One utility co11111entor expressed concern that the addition of§ 50.55a(f)(6)(ii) would permit the Commission to impose an augmented inservice testing program without further justification. This is not the case.

Any program for augmented inservice testing will be fully justified with a documented regulatory analysis that includes the appropriate backfit analysis.

The intent of the NRC to perform the necessary backfit analysis is clearly demonstrated by the backfit analysis that was performed to require the augmented examination of the reactor vessel that is specified in

§ 50.55a(g)(6)(ii)(A) of this final rule.

Second, this final rule includes the addition of introductory text to

§ 50.55a(g) which states that the requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in§ 50.55a(f). This change is necessary because the placement of inservice testing requirements into a separate§ 50.55a(f), as included in the proposed rule, would have 18

caused administrative inconsistencies with regard to existing references to

§ 50.55a(g) for inservice testing in documents such as technical specifications, safety analysis reports, procedures, and records.

With this change, existing references to§ 50.55a(g) for inservice testing will refer the user to§ 50.55a(f), where the specific requirements for inservice testing are located.

The NRC recommends that as the governing documents are updated, the direct reference to§ 50.55a(f) be incorporated, as appropriate.

Two editorial revisions, relative to the previous§ 50.55a(g), are included in the new§ 50.SSa(f). These editorial revisions: (1) reserve

§ 50.55a(f)(3)(i) and (ii) so that the structure of§ 50.SSa(f) will parallel that of§ 50.SSa(g) for the purpose of promoting easier cross-referencing between the two paragraphs; and (2) modify the reference to 120-month inspection interval in§ 50.SSa(g) to 120-month interval in§ 50.SSa(f),

because the term "inspection interval," as used in Section XI, is used only in the context of inservice inspection.

(The term *test interval 11 was not used because, unlike inspection interval, the 120-month time frame does not designate a period of required actions for the testing program.

The 120-month interval used in§ 50.SSa{f) and the 120-month inspection interval used in§ 50.SSa(g) are considered by the staff to be coincident for the purpose of 120-month updating requirements.)

A number of co11111ents were received regarding the scope of§ 50.55a as applied to pump and valve testing. These comments ranged from reco11111ending that the scope of§ 50.55a be expanded* to be consistent with the scopes of OM Part 6 and Part 10, which go beyond Class 1, Class 2, and Class 3 components, 19

to recommending that the scope of§ 50.55a be limited to ASME Code classified components.

One corrnnentor expressed concern that the Supplementary Information in the proposed rulemaking addressed Generic Letter 89-04 in a way that seemed to include the letter in the rulemaking.

That was not intended.

To the contrary, the intent of this rulemaking is to maintain the existing scope of§ 50.55a for pump and valve testing.

For plants whose construction permits were issued on or after January 1, 1971, that scope constitutes Code classified components as specified in existing§ 50.55a(g)(2) and (3)

[i.e., § 50.55a(f)(2) and (3) by this rulemaking].

For those plants whose e

construction permits were issued prior to January 1, 1971, that scope

~

constitutes components of the reactor coolant pressure boundary wnich must meet the requirements applicable to components that are classified as ASME Code Class 1, and other safety-related pumps and valves which must meet the requirements applicable to components that are classified as ASME Code Class 2 or Class 3, as specified in existing§ 50.SSa(g)(l) [i.e., § 50.SSa(f)(l) by this rulemaking].

The reference to the generic letter has not been included in the final rule.

A number of comments were received with regard to snubber testing which is outside the scope of this rulemaking.

Commentors generally suggested that ASME OM Part 4, RExamination and Perfonnance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers),R which is referenced in Subsection IWF in the 1987 Addenda, 1988 Addenda and 1989 Edition of Section XI, be incorporated by reference into§ 50.55a. Subsection IWF, "Component Supports,R provides rules for the examination of component supports, and the testing of snubbers.

Prior to the 1987 Addenda, Subsection IWF provided self-contained rules for the 20

testing of snubbers.

Section 50.55a does not specify requirements for the testing of snubbers. This was clarified by the separation of requirements for inservice testing and inservice inspection.

Inservice testing requirements specified in§ 50.55a(f) apply only to pumps and valves.

The testing requirements specified in OM Part 4 and referenced in Section XI Subsection IWF article IWF-5000 are not incorporated by reference into§ 50.55a.

Requirements for the testing of snubbers are generally governed by plant technical specifications.

NRC is in the process of initiating a proposed rul-emaking that would, among other things, address the incorporation by reference of the ASME OM Code, which contains rules for pump, valve, and snubber testing, into§ 50.55a(f).

The NRC will as a part of this future rulemaking determine the need for and acceptability of endorsing the ASME OM Code rules for snubber testing. However, in accordance with requirements for examination of component supports specified in§ 50.55a(g), licensees are required to implement the rules for examination of snubbers that are provided in OM Part 4 as referenced in Subsection IWF Article IWF-5000 in the applicable Section XI addenda and edition of this final rule.

Section 50.55a(g) provides requirements for selecting the ASME Code edition and addenda of Section XI to be complied with during the preservice inspection(§ 50.55a(g)(3), for plants whose construction permit was issued on or after July 1, 1974); the initial IO-year inspection interval

(§ 50.55a(g)(4)(i)); and successive IO-year inspection intervals

(§ 50.55a(g)(4)(ii)). As noted in the final rule codifying the most recent amendment to§ 50.55a (May 5, 1988; 53 FR 16051), paragraph IWA-2400 of Section XI (as revised by the Winter 1983 Addenda) incorporated rules for 21

selecting the applicable edition and addenda of Section XI during the preservice inspection (IWA-2411), the initial 10-year inspection interval (IWA-2412), and successive 10-year inspection intervals (IWA-2413).

The criteria provided in the regulations and Section XI are effectively the same for the preservice inspection and the successive 10-year inspection intervals, but differ for the initial 10-year inspection interval.

In general, use of the Co1T111ission requirements will result in the selection of a more recent edition and addenda than will use of the Section XI rules. Satisfying the requirements of§ 50.55a{g){4)(i) for the initial 10-year inspection interval will, in general, also satisfy the rules of Section XI.

Although the Section XI requirements for selecting editions and addenda remain unchanged in the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition, the Commission is reaffirming its intent that in all cases the existing requirements in

§ 50.55a{g) be the basis for selecting the edition and addenda of Section XI to be complied with during the preservice inspection, the initial 10-year inspection interval, and the successive 10-year inspection intervals.

This final rule makes a number of editorial changes to§ 50.55a for the purpose of adopting a standard convention for imposing an obligation or expressing a prohibition.

In this convention "shall" is used to impose an obligation on an individual or legal entity capable of performing the required action, "mustn is used as the mandatory form when the subject of the sentence is an inanimate object, and "may not" is used to impose a prohibition. The following paragraphs are amended solely to be consistent with this convention:

the introductory paragraph to the section; paragraphs {a){l), {a){3),

-(b)(2)(iii), (b)(2)(iv), (g)(l), (g)(3)(ii), (g)(3)(iii), (g)(3)(iv),

22

introductory paragraph to (g)(4), (g)(4)(i), (g)(4)(ii), (g)(S)(i),

(g)(S)(iv), (g)(6)(i), (h), and footnote 8. Other paragraphs are amended for the same editorial reason, but they also contain technical revisions relevant to other parts of this final rule. Section 50.55a(f) has been developed consistent with the noted convention.

Subsection IWE, "Requirements for Class MC Components of Light-Water-Cooled Power Plants,* was added to Section XI, Division 1, in the Winter 1981 Addenda.

Since§ 50.55a does not currently address the inservice inspection of containments and the scope of§ 50.55a is not affected by this final rule, the requirements of Subsection IWE are not imposed upon Commission licensees by this amendment.

The incorporation by reference of Subsection IWE into

§ 50.55a is presently the subject of a separate rulemaking action. Section 50.55a(b)(2)(vi) is reserved for that action.

The NRC previously alerted all holders of operating licenses or construction permits for nuclear power reactors, through NRC Information

,e Notice No. 88-95 (IN 88-95), "Inadequate Procurement Requirements Imposed by

/

Licensees on Vendors,* to the potential that inadequate licensee procurement requirements or implementation by vendors in supplying components under the ASME Code could result in failure by these vendors to fully implement 10 CFR Part -so, Appendix B (Quality Assurance Criteria). The problem, which was revealed during routine NRC inspections of vendors, resulted from the belief by some vendors that if an item was exempted by the ASME Code from Code requirements, the item was exempt from all other regulatory requirements.

The apparent belief of some vendors was that since NRC endorses the ASME Code in 23

its regulations and has accepted the various exemptions, there are, therefore, no other applicable regulatory requirements.

This belief is not consistent with the NRC position.

The NRC reaffirms its position which, as previously put forth in IN 88-95, states that all safety-related items, even those exempted from ASME Code requirements, are required to be manufactured under a quality assurance program that meets the requirements of 10 CFR Part 50, Appendix B.

Finding of No Significant Environmental Impact: Availability The Convnission has determined under the National Environmental Policy Act of 1969, as amended, and the Co11111ission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action that significantly affects the quality of the human environment and therefore an environmental impact statement is not required.

This final rule is one part of a regulatory framework directed to ensuring pressure vessel integrity, and the operational readiness of pumps and valves. Therefore, in the general sense, this rule will have a positive impact on the environment.

This rule incorporates by reference into the NRC regulations improved rules contained in the ASHE Code for the construction, inservice inspection, and inservice testing of components used in nuclear power plants.

In addition, this rule requires an augmented examination of reactor vessel shell welds to further ensure the structural integrity of the reactor vessel. The occupational exposures attributable to the expanded reactor vessel examinations contained in the ASHE Code and the augmented 24

examination are not expected to be significant because exposures will be limited by the use of remote examination equipment.

Occupational exposures associated with the augmented reactor vessel examination will be further limited by provisions in the final rule that permit, under certain conditions, the licensee to satisfy the requirement for the augmented examination by previously scheduled or implemented reactor vessel examinations, or by deferring the examination to the next interval and using the deferred examination as a replacement for the previously scheduled examination for that interval. The actions required by applicants and licensees to implement the final rule are of an established nature that should not increase the potential for a negative environmental impact.

The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

Single copies of the environmental assessment and the finding of no significant impact are available from Gilbert C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-3848.

Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

These requirements were approved by the Office of Management and Budget 25

e approval number 3150-0011.

The public reporting burden for this collection of information is estimated to average 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.

Send co1T111ents regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Convnission, Washington, DC 20555, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management and Budget, Washington, DC 20503.

Regulatory Analysis The Co1T111ission has prepared a regulatory analysis for this amendment to the regulations.

The analysis examines the costs and benefits of the alternatives considered by the Commission.

Interested persons may examine a copy of the regulatory analysis at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

Single copies of the analysis may be obtained from Mr. G. C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Conrnission, Washington, DC 20555, Telephone (301) 492-3848.

26

Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.

605(b), the Co1m1ission hereby certifies that this rule will not have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants.

The companies that own these plants do not fall within the scope of the definition of "small entitiesR set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. Since these companies are dominant in

.:_ ----their service areas, this rule does not fall within the purview of the Act.

Backfit Analysis The final rule incorporates by reference a later edition and addenda to Section III, Division 1, and, with both a technical and nontechnical

,e modification,Section XI, Division 1, of the ASME Code; imposes an augmented examination on reactor vessels; and separates the requirements for inservice inspection from those for inservice testing.

The incorporation by reference into the regulations of later editions and addenda of Section III and Section XI of the ASME Code is not a backfit because Section III requirements apply only to new construction, except as voluntarily implemented by licensees, and because updated Section XI requirements are an integral part of the longstanding§ 50.55a(g)(4)(ii) 27

requirement to update inservice inspection and inservice testing programs to reflect the requirements of the latest edition and addenda of Section XI incorporated by reference in§ 50.55a(b) 12 months prior to the start of the 120-month inspection interval, subject to specified limitations and modifications.

The technical modification to Part 10 of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987 specified in§ 50.55a(b)(2)(vii) is not a backfit because it simply retains an existing Section XI requirement for containment isolation valve testing that licensees now are required to implement in accordance with§ 50.55a(g).

The nontechnical modification specified in

§ 50.55a(b)(2)(viii) is not a backfit because it only serves to properly identify an incorrectly referenced standard in Section XI.

The NRC has concluded, based on the analysis required by§ 50.109(a)(3) which is provided in the regulatory analysis, that the backfit that will be imposed by the augmented reactor vessel examination specified in

§ 50.55a(g)(6)(ii)(A) will result in a substantial increase in the overall protection of the public health and safety, and that the direct and indirect costs of implementation are justified in view of the increased protection.

The separation in the regulation of the inservice inspection and inservice testing requirements is an administrative reorganization of§ 50.55a that has no impact on existing technical requirements and, therefore, has no effect on backfitting.

28

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, 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 Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments 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. 5851).

Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 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, 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, 29

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).

For the* purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273);

§§ 50.5, 50.46(a) and (b), and 50.54(c) are issued under sec. 161~, 68 Stat.

948 as amended (42 U.S.C. 220l(b); §§ 50.5, 50.7(a), 50.lO(a)-(c), 50.34(a) and (e), 50.44(a)-(c), 50.46(a) and (o), 50.47(b), 50.48(a), (c), (d), and (e), 50.49(a), 50.54(a), (i), (i)(l), (1)-(n), (p), (q), (t), (v), and (y),

50.55(f), 50.55a(a), (c)-(e), (g), and (h), 50.59(c), 50.60(a), 50.62(b),

50.64(b), 50.65, and 50.80(a) and (b) are issued under sec. 16li, 68 Stat.

949, as amended (42 U.S.C. 2201(1)); and§§ 50.49(d), (h), and (j), 50.54(w),

(z), (bb), (cc), and (dd), 50.55(e), 50.59(b), 50.6l(b), 50.62(b), 50.70(a),

50.71(a)-(c) and (e), 50.72(a), 50.73(a) and (b), 50.74, 50.78, and 50.90 are issued under sec. 1610, -68 Stat. 950, as amended (42 U.S.C. 2201(0)).

2.

In§ 50.55a, the introductory text, paragraphs (a), (b)(l), the introductory text of (b)(2), (b)(2)(iii), (b)(2)(iv)(A), (g)(l), (g)(2),

(g)(3)(i), (g)(3)(ii), (g)(4), (g)(5)(i), (g)(5)(iv), (h), and footnote 8 are revised; paragraphs (g)(3)(iii) and (g)(3)(iv) are removed and reserved; paragraph (b)(2)(vi) is added and reserved; and paragraphs (b)(2)(vii),

30

(b)(2)(viii), (f), introductory text to (g), and (g)(6)(ii)(A) are added to read as follows:

§ 50.55a Codes and standards.

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 in addition to those specified in§ 50.55.

(a)(l) Structures, systems, and components must be designed, fabricated, erected, constructed, tested, and inspected to quality standards co11111ensurate with the importance of the safety function to be performed.

(2)

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code specified in paragraphs (b), (c),

(d), (e), (f): and (g) of this section.

Protection systems of nuclear power reactors of all types must meet the requirements specified in paragraph (h) of this section.

(3)

Proposed alternatives to the requirements of paragraphs (c), (d), (e}, (f}, (g}, and (h} of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear 31

Reactor Regulation.

The applicant shall demonstrate that:

(1)

The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

(b) * * *

(1)

As used in this section, references to Section III of the ASME Boiler and Pressure Vessel Code refer to Section III, Division l, and include addenda through the 1988 Addenda and editions through the 1989 Edition.

(2)

As used in this section, references to Section XI of the ASME Boiler and Pressure Vessel Code refer to Section XI, Division 1, and include addenda through the 1988 Addenda and editions through the 1989 Edition, subject to the following limitations and modifications:

(iii) Steam generator tubing (modifies Article IWB-2000).

If the technical specifications of a nuclear power plant include 32

surveillance requirements for steam generators different than those in Article IWB-2000, the inservice inspection program for steam generator tubing is governed by the requirements in the technical specifications.

(iv)

Pressure-retaining welds in ASME Code Class 2 piping (app]jes to Tables IWC-2520 or IWC-2520-1. Category C-F),

(A) Appropriate Code Class 2 pipe welds in Residual Heat Removal Systems, Emergency Core Cooling Systems, and Containment Heat Removal Systems, must be examined.

When applying editions and addenda up to the 1983 Edition through the Summer 1983 Addenda of Section XI of the ASME Code, the extent of examination for these systems must be determined by the requirements of paragraph IWC-1220, Table IWC-2520 Category C-F and C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the Su111T1er 1975 Addenda.

(vi)

[Reserved]

(vii) Inservice testjng of containment isolation valves.

When using Subsection IWV in the 1988 Addenda or the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, leakage rates for Category A containment isolation valves that do not provide a reactor coolant system pressure isolation function must be analyzed in accordance with paragraph 4.2.2.3(e) of Part 10, 33

and corrective actions for these valves must be made in accordance with paragraph 4.2.2.3(f) of Part 10 of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987.

(viii)Section XI References to OM Part 4, OM Part 6 and OM Part 10 (Table IWA-1600-1}.

When using Table IWA-1600-1, "Referenced Standards and Specifications" in the Section XI, Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified "Revision Date or Indicator" for ASME/ANSI OM Part 4, ASME/ANSI Part 6, and ASME/ANSI Part 10 shall be the OMa-1988 Addenda to the OM-1987 Edition.

(f) Inservice testing requirements.

(1)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, pumps and valves must meet the test requirement of paragraphs (f)(4) and (5) of this section to the extent practical.

Pumps and valves which are part of the reactor coolant pressure boundary must meet the requirements applicable to components which are classified as ASHE Code Class 1. Other safety-related pumps and valves must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

34

(2)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, pumps and valves which are classified as ASME Code Class 1 and Class 2 must be designed and be provided with access to enable the performance of inservice tests for operational readiness set forth in editions of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda 6 in effect 6 months prior to the date of issuance of the construction permit.

The pumps and valves may meet the inservice test requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed therein.

(3)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after July 1, 1974:

(i)

[Reserved]

(ii) [Reserved]

{iii)

Pumps and valves which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6 applied to 35

the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.

(iv)

Pumps and valves which are classified as ASME Code Class 2 and Class 3 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 5

applied to the construction of the particular pump or valve or the Surrmer 1973 Addenda, whichever is later.

(v)

All pumps and valves may meet the test requirements set forth in subsequent editions of codes and addenda or portions thereof which are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section.

(4)

Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves which are classified as-ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements, except design and access provisions, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda that become effective subsequent to editions specified in paragraphs (f)(2) and (f)(3) of this section and that are incorporated by reference in paragraph (b) of this section, to the extent practical within the limitations of design, geometry and 36

materials of construction of such components.

(i)

Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during the initial 120-month interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section on the date 12 months prior to the date of issuance of the operating license, subject to the limitations and modifications listed in paragraph (b) of this section.

(ii) Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed in paragraph (b) of this section.

(iii) [Reserved]

(iv)

Inservice tests of pumps and valves may meet the requtrements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section, and subject to Convnission approval.

Portions of editions 37

or addenda may be used provided that all related requirements of the respective editions or addenda are met.

(5)(i) The inservice test program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (f)(4) of this section.

{ii) If a revised inservice test program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Co11111ission for amendment of the technical specifications to conform the technical specification to the revised program.

The licensee shall submit this application, as specified in§ 50.4, at least 6 months before the start of the period during which the provisions become applicable, as determined by paragraph (f){4) of this section.

{iii) If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shal1 notify the Convnission and submit, as specified in

§ 50.4, information to support the determination.

(iv) Where a pump or valve test requirement by the code or addenda is determined to be impractical by the licensee and is not included in the revised inservice test program as permitted by paragraph {f)(4) of this section, the basis for this determination 38

must be demonstrated to the satisfaction of the Colllllission not later than 12 months after the expiration of the initial 120-month period of operation from start of facility commercial operation and each subsequent 120-month period of operation during which the test is determined to be impractical.

(6)(i) The Commission will evaluate determinations under paragraph (f)(S) of this section that code requirements are impractical.

The Commission may grant relief and may impose such alternative requirements as it detennines is authorized by law and will not endanger life or property or the co1T111on defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

(ii) The Commission may require the licensee to follow an augmented inservice test program for pumps and valves for which the Colllllission deems that added assurance of operational readiness is necessary.

(g)

Insery1ce inspection requirements.

Requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in§ 50.55a(f).

(1)

For a boiling or pressurized water-cooled nuclear power facility whose construction pennit was issued prior to January 1, 39

1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (5) of this section to the extent practical.

Components which are part of the reactor coolant pressure boundary and their supports must meet the req4irements applicable to components which are classified as ASME Code Class 1. Other safety-related pressure vessels, piping, pumps and valves must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

(2)

For a boiling or pressurized water-cooled nuclear power facility whose construction pennit was issued on or after January 1, 1971, but before July 1, 1974, components (including supports) which are classified as ASME Code Class 1 and Class 2 must be designed and be provided with access to enable the perfonnance of inservice examination of such components (including supports) and must meet the preservice examination requirements set forth in editions of Section XI of the ASHE Boiler and Pressure Vessel Code and Addenda~

in effect six months prior to the date of issuance of the construction permit.

The components (including supports) may meet the requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section.

(3)

For a boiling or pressurized water-cooled nuclear power facility whose construction pennit was issued on or after 40

July 1, 1974:

(1)

Components which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of such components and must meet the preservice examination requirements set forth in Section XI of editions of the ASHE Boiler and Pressure Vessel Code and Addenda 6

applied to the construction of the particular component.

(ii} Components which are classified as ASME Code Class 2 and Class 3 and supports for components which are classified as ASME Code Class 1, Class 2, and Class 3 must be designed and be provided with access to enable the performance of inservice examination of such components and must meet the preservice examination requirements set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6 applied to the construction of the particular component.

(iii) [Reserved]

(iv)

[Reserved]

(4)

Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) 41

which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda that become effective subsequent to editions specified in paragraphs (g)(2) and (g)(3) of this section and that are incorporated by reference in paragraph (b) of this section, to the extent practical within the limitations of design, geometry and materials of construction of the components.

(i) Inservice examinations of components and system pressure tests conducted during the initial 120-month inspection interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section on the date 12 months prior to the date of issuance of the operating license, subject to the limitations and modifications listed in paragraph (b) of this section.

(ii) lnservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed in paragraph (b) of this section.

42

(iii) [Reserved]

(iv)

Inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section, and subject to Corrmission approval.

Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.

(5)(1) The inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section.

(iv) Where an examination requirement by the code or addenda is determined to be impractical by the licensee and is not included in the revised inservice inspection program as permitted by paragraph {g){4) of this section, the basis for this determination must be demonstrated to the satisfaction of the Commission not later than 12 months after the expiration of the initial 120-month period of operation from start of facility corrmercial operation and each subsequent 120-month period of operation during which the examination is determined to be impractical.

43

(6) * * *

(ii) * * *

(A)

Augmented examination of reactor vessel.

ill All previously granted reliefs under§ 5O.55a to licensees for the extent of volumetric examination of reactor vessel shell welds specified in Item Bl.IO of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," in Table IWB-25OO-1 of Subsection IWB in applicable edition and addenda of Section XI, Division I, of the ASME Boiler and Pressure Vessel Code, during the inservice inspection interval in effect on ____ (effective date of rule will be inserted) are hereby revoked, subject to the specific modification in§ 5O.55a(g)(6)(ii)(A){3}{iv} for licensees that defer the augmented examination in accordance with

§ 5O.55a(g)(6)(ii)(A)fil.

ill All licensees shall augment their reactor vessel examination by implementing once, as part of the inservice inspection interval in effect on ___ _

(effective date of rule will be inserted), the examination requirements for reactor vessel shell welds specified in Item Bl.IO of Examination Category B-A, nPressure Retaining Welds in Reactor Vessel," in Table IWB-25OO-1 of Subsection IWB of the 1989 Edition of Section XI, Division I, of the ASME Boiler and Pressure Vessel Code, subject to the conditions 44

specified in§ 50.55a(g)(6)(ii)(A)ill and ill-The augmented examination, when not deferred in accordance with the provisions of

§ 50.55a(g)(6)(ii)Aill, shall be performed in accordance with the related procedures specified in the Section XI edition and addenda applicable to the inservice inspection interval in effect on ____ (effective dat_e of rule will be inserted), and may be used as a substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection interval in effect on ____ (effective date of rule will be inserted). For the purpose of this augmented examination, "essentially 100%" as used in Table IWB-2500-1 means more than 90 percent of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

ill Licensees with fewer than 40 months remaining in the inservice inspection interval in effect on ____ (effective date of rule will be inserted) may defer the augmented reactor vessel examination specified in § 50.55a{g){6){ii){A)ill to the first period of the next inspection interval under the following conditions:

ill The deferred augmented examination may not be used as a substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection interval in effect on ____ {effective date of rule will be inserted).

45

il.il The deferred augmented examination may be used as a substitute for the reactor vessel shell weld examination normally scheduled for the inspection interval in which the deferred examination is performed.

illil If the deferred augmented examination is used as a substitute for the normally scheduled reactor vessel shell weld examination, subsequent reactor vessel shell weld examinations must be performed during the first period of successive inspection intervals.

.L1Yl Licensees that defer the augmented examination, as permitted herein, may retain all previously granted reliefs that otherwise would be revoked by§ 50.55a(g)(6)(ii)(A)ill for the inservice inspection interval in effect on ____ (effective date of rule will be inserted).

.!.Yl Licensees with fewer than 40 months remaining in the inservice inspection interval in effect on ___ _ (effective date of rule will be inserted) may extend that interval in accordance with the provisions of Section XI (1989 Edition) IWA-2430(d) for the purpose of implementing the augmented examination during that interval.

Wl The deferred augmented examination shall be performed in accordance with the related procedures specified in the Section XI 46

edition and addenda applicable to the inspection interval in which the augmented examination is performed.

ill The requirement for augmented examination of the reactor vessel may be satisfied by an examination of essentially 100 percent of the reactor vessel shell welds specified in

§ 50.55a(g)(6)(ii)(A)ill that has been completed, or is scheduled for implementation with a written commitment, or is required by

§ 50.55a(g)(4)(i), during the inservice inspection interval in effect on ---- (effective date of rule will be inserted).

ill Licensees that make a determination that they are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in§ 50.55a(g)(6)(ii)(A) shall submit infonnation to the Co1TU11ission to support the determination and shall propose an alternative to the examination requirements that would provide an acceptable level of quality and safety.

The licensee may use the proposed alternative when authorized by the Director of the Office of Nuclear Reactor Regulation.

(h)

Protection systems.

For construction permits issued after January 1, 1971, protection systems must meet the requirements set forth in editions or revisions of the Institute of Electrical and Electronics Engineers Standard: "Criteria for Protection Systems for Nuclear Power Generating Stations," (IEEE-279) in effect 7 on the 47

6 formal docket date 0 of the application for a construction permit.

Protection systems may meet the requirements set forth in subsequent editions or revisions of IEEE-279 which become effective.

ASME Code cases that have been determined suitable for use by the Commission staff are listed in NRC Regulatory Guide 1.84, 11Design and Code Case Acceptability -- ASME Section III Division l," NRC Regulatory Guide 1.85, "Materials Code Case Acceptability -- ASME Section III Division l, 11 and NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability -- ASME Section XI Division l.

11 The use of other Code cases may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to§ S0.55a(a)(3).

7 For purposes of this regulation the proposed IEEE 279 became "in effect* on August 30, 1968, and the revised issue IEEE 279--1971 became "in effect" on June 3, 1971.

Copies may be obtained from the Institute of Electrical and Electronics Engineers, United Engineering Center, 345 East 47th St., New-York, NY 10017.

Copies are available for inspection at the CoDJDission's Technical Library, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland.

48

\\.

8 Where an application for a construction permit is submitted in four parts pursuant to the provisions of§ 2.lOl(a-1) and Subpart F of Part 2 of this chapter, "the formal docket date of the application for a construction permit" for purposes of this section is the date of docketing of the information required by§ 2.lOl(a-1), (2), or (3),

whichever is later.

Dated at Rockville, Maryland this d~ay c}f---1992_

For the Nuclear Regulatory Commission.

M. T~~

Executive Director for Operations.

49

tF~ CENTERIOR

~ ENERGY Donald C. Shelton Vice President

• Nuclear Davis-Besse Docket Number 50-346 License Number NPF-3 Serial Number 1933 April 29, 1991

• Secretary

• 91 MAY - 3 A10 :28 United States Nuclear Regulatory Commission Attention:

Docketing and Service Branch Washington, D. C.

20555 300 Madison Avenue Toledo, OH 43652-0001 (419) 249-2300

Subject:

Comments on Proposed Rule:

Codes and Standards for Nuclear Power Plants (56 FR 3796)

Gentlemen:

Federal Register, Volume 56, Number 21, dated January 31, 1991, solicited comments on a proposed revision of 10 CFR 50.55a, "Codes and Standards."

Toledo Edison has reviewed the proposed revision of 10 CFR 50.55a and comments prepared on behalf of the nuclear power industry by the Nuclear Management and Resources Council (NUMARC).

Toledo Edison's comments on the proposed rule are fully reflected in the NUMARC comments.

Accordingly, Toledo Edison endorses the NUMARC comments.

If you have any questions regarding Toledo Edison endorsement of the NUMARC comments, please contact Mr. R. W. Schrauder, Manager - Nuclear Licensing, at (419) 249-2366.

cc:

P. M. Byron, DB-1 NRC Senior Resident Inspector A. B. Davis, Regional Administrator, NRC Region III J. R. Hall, NRC Senior Project Manager M. D. Lynch, NRC Senior Project Manager Nuclear Management and Re~ources Council Utility Radiological Safety Board

~~,

tl~'f 2 Operating Companies:

Cleveland Electric Illuminating Toledo Edison c1mowledged by card............................,m

U.S. NUCLEAR REGUL TORY COMM! SION OOCKETING & SER'/iCF. SECTION OFFICE OFT riE S~CRl?T.:,n*r OF THE en*.** <

DOCKET NUMBER PROPOSED RULE PR 5 O f' 1 ff$

. (_56 FR. 03 7'f(y /2qi BOSTON EDISON L "uct,1t' 0

Pilgrim Nuclear Power Station Rocky Hill Road Plymouth, Massachusett~ 123~

R 26 p J :42 George W. Davis Senior Vice President -

Nuclear Mr. Samuel J. Chilk Secretary of the Commission Washington, DC 20555 Attention:

Docketing and Service Branch AQril 22, 1991 BtCo 91-057 License DPR-35 Doer.et 50-293 Codes and Standards for Nuclear Power Plants 56 Federal Register 3796 (January 31, 1991)

Comments on Proposed Rule

Dear Mr. Chilk:

These comments are submitted by Boston Edison Company in response to the proposed rule of the U.S. Nuclear Regulatory Commission (NRC) incorporating the 1986 Addenda, 1987 Addenda, 1988 Addenda and the 1989 Editions of the ASME Boiler and Pressure Vessel Code,Section III, Division l and Section XI, Division I, by reference into 10 CFR Part 50.55a.

He have reviewed the proposed rule and the associated draft regulatory analysis and we offer the following comments for your consideration.

The rule reflects a review of the potential impacts on the industry as a who l e, but does not address in sufficient detail the impact on an individual power plant.

Inspection of the Pilgrim reactor vessel will be limited by the reactor vessel internal design. Similarly, it may not be possible to conduct the inspections on the schedule reflected in the proposed rule because the required technology does not exist and may not be available in the time frame stated in the rule, as proposed.

He believe a balance needs to be struck in the final rule between the proposed inspections and the overall impact on an individual plant.

He, therefore, request the NRC consider the potential hardship inherent in the proposed rule and add the words "to the extent practical" in Section 50.55a(g)(6)(ii)(A)(2) and provide greater latitude in the implementation schedule in the final rule.

He also encourage the NRC to review previously-granted reliefs under 10CFR50.55a for the examination of reactor vessel shell welds on a case-by-cas e basis, or in the alternative, as part of an Owner's Group effort prior to revoking any reliefs granted under 50.55a.

This will allow the industry and the NRC to develop a consistent basis for requesting and granting reliefs and reduce the cost to the industry and the NRC to refile and revi ew i ndividual relief requests.

MA't 2 o 199\\

Acknowledged b\\i..... ti:il............................... "":

1"""'

... ******4*......

l.J.S. N~~Lt:AR R[GdLATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Copies Received __ "'--------

Add'I Copies Reproduced __

if. ___ _

Special Distribution Pot., r'< I j) S',

fl,,

. IJ r

7

.ulfYYlobtli mdlm(i.¥1 r

I

BOSTON EDISON COMPANY S. J. Chil k Page 2 We agree with the concerns raised by the Nuclear Management and Resources Council and endorse their comments on the proposed amendment to 10CFR5O.55a.

We appreciate the opportunity to comment on the proposed rule and would welcome an opportunity to discuss our comments further.

RAH/clc/5711 cc:

Sr. NRC Resident Inspector-Pilgrim Station

II Duke Power Company Nuclear Production Dept.

PO Box 1007 Charlotte.NC. 28201-/00, DOCKET NUMBER PR PROPOSED RULE 6IJ DUKE POWER.

April 18, 1991 The Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.

20555

..,L~~i~~ 0.31t;~)

U::i NiiC

• 91 APR 26 P 3 :37 Attention: Docketing and Service Branch

Subject:

Codes and Standards for Nuclear Power Plants Proposed Rule Duke Power Company Comments

Dear Sir:

I M.S. TUCKMAN Vice President Nuclear Operations (704)373-385 /

On April 15, 1991, Duke Power Company submitted comments on a proposed amendment to 10 CFR 50.55a. We would like to amend our comments as follows:

Duke Power Company has reviewed the proposed rule and concurs with the comments submitted by NUMARC with the following exception.

The proposed 10 CFR 50.55a revision would further separate testing requirements for containment isolation valves (CIVs). Prior to Part 10, CIV testing requirements were contained in both Subsection IWV and Appendix J. Part 10 consoiidated CIV testing requirements by specifying that CIVs shall be tested in accordance with Appendix J.

The proposed change would distribute CIV testing requirements between Appendix J, 10 CFR 50.55a, and Part 10. This would be confusing and would not promote consistent testing. CIV testing requirements should be consolidated.

Duke appreciates the opportunity to comment on this proposed rule.

Very truly yours, M.S. Tuckman JAR MAY 2 o 1991 Acknowledged by card..................................

. (,JLJCL'-/,R nrnuLATORY COMMISSION' DOCKETING e, SERVICE SECTION OFFICE OF THI: SECRETARY OF THE COMMISSION Document StaUstics Postmark Data '-I /2 / If I CooiP.s Receivad,_1__,_/ _

1

-.rJd'I Cr,!

::Eo: Reproduced __,_f ___ _

SpeC!2i Disb bu~ion f Dr?,. f? / or/

n1: 11mm,. faldJ.IJ b.t1f II

DOCKET NUMBER PR 5f)

PROPOSED RULE G

TIT YANKEE ATOMIC ELECTRIC COMPANYl.s~FP.tJ!4..~¥e!epnone(S08) 779-6711 0

'y TWX 710-3~~-~~1~

• ~

ld~!\\;,c;/27/

~

.~

1 h1 580 Main Street, Bolton, Massachusetts 01740-1398

,HDi*

U'.:J Nr~C G-

• JANK.EE

• 91 APR 23 Al 1 :QQ Secretary April 22, 1991 SPS 91-42 FYC 4-91 United States Nuclear Regulatory Commission Washington, DC 20555 Attention:

Subject:

Docketing and Service Branch Proposed Rule on Codes and Standards for Nuclear Power Plants to Amend 10CFR50.55a (56FR3796)

Dear Mr. Chilk:

Yankee Atomic Electric Company (Yankee) appreciates the opportunity to comment on the proposed rule to revise 10CFR50.55a.

Yankee owns and operates a nuclear power plant in Rowe, Massachusetts.

Our Nuclear Services Division also provides services to other nuclear power plants in the Northeast, including Vermont Yankee, Maine Yankee, and Seabrook.

The Nuclear Management and Resources Council (NUMARC) is also submitting comments on this proposed rule.

Yankee is a member of NUMARC and fully endorses their comments.

Of particular importance is the NUMARC comment on Containment Isolation Valve (CIV) testing.

All Yankee plants have tested CIVs in accordance with 10CFR50 Appendix J since its inception.

The proposed rule change substitutes individual valve leakage rates for the Appendix J required limitation on the summation of all containment penetration and isolation valve leakages.

It is not always physically possible to isolate all individual valves for testing due to the construction characteristics of certain valve groupings.

Redundancy should also be taken into consideration.

The inability to test all valves individually does not indicate an inability to meet Technical Specifications or design requirements.

Implementation of the present Appendix J limit at Yankee facilities involves a vigorous valve maintenance program, thus the intent of the proposed approach is already implemented without the negative effect of limitation of operational flexibility which could cause unnecessary shutdowns.

Consequently, we oppose this change.

The NRC is also proposing to adopt, by reference, more recent editions of ASME Section XI.

The effect of this change is to require 100% volwnetric inspection of all reactor pressure vessel welds in every 10-year inspect i on ffi~'{ 2 0 \\~'l\\

I Acknowledged by card..................................

U.S. NUCLEAR Rc:GULI\\ i UH V ~UMMl6Si0N DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date £E t/lr Copies Received_---'--~ ---

Add'I Copies Reproduced _

Special Distribution,PJ),.q, 1? -Z: Z> s;,

/IJ:/ I m1~ C4y7.6e1/

United States Nuclear Regulatory Commission Attention:

Docketing and Service Branch April 19, 1991 Page 2 interval.

We can understand the NRC staff's need for absolute assurance of the integrity of reactor pressure vessels and agree that it is vitally important.

However, we also feel that inservice inspections can be more specifically concentrated in the area of concern (i.e., the vessel beltline region) and still adequately supplement the information obtained during vessel fabrication regarding integrity of the welds.

This viewpoint is particularly true in the case of Boiling Water Reactors (BWR), where vessel dimensions assure that fluence at the vessel wall and, thus, embrittlement are substantially lower.

Consideration of the 100% inspection requirement, in light of the fact that only particular regions of the vessel are susceptible to service-induced degradation, makes it seem inappropriate.

Under this regimen, weld zones least likely to incur inservice degradation receive inspections with equal rigor during each 10-year interval as those welds most likely to experience degradation.

By NRC's own assessment not only will these inspections cost millions of dollars at each facility but will result in an increase in worker dose.

It seems imperative that NRC take steps to optimize the requirement to the safety benefit obtained.

First, the rule should recognize interferences and accessibility problems by establishing a more realistic objective than the 100% weld inspection.

Second, the rule should preferentially focus the inspection on the vessel beltline region as contrasted to all welds.

Third, the rule should draw a clear distinction between PWR and BWR vessels.

In addition, it should acknowledge the drastically reduced susceptibility of BWR vessels to irradiation embrittlement by easing the inspection objective for percent of weld to perhaps 100% of beltline welds only.

In the backfit analysis accompanying this rule, experiences with vessel nozzle cracks or cracks in pressurizers and steam generators are identified as the reason for this rule.

The particulars of each of these instances are such that the welds involved are clearly not similar to reactor vessel shell welds.

Consequently, the apparent reason for this blanket requirement does not appear to be adequate underpinning for such a sweeping regulatory backfit.

We ask the NRC to reconsider this rulemaking.

requirement should not be used as a replacement for requirements focused on specific areas where safety assurance of safety can be achieved.

D. W. Edwards A blanket inspection technically sound enhancements or stronger Director, Industry Affairs DWE/dhm/WPP76/320

DOCKET NUMBER PR 5o PROPOSED RU(i°l, ;::,e t'.3 fq,:.)

I $@I Pennsylvania Power & Light Company Two North Ninth Street*Allentown, PA 18101-1179*215/774-5151 Harold W. Keiser Senior Vice President-Nuclear 215/774-4194 APR, 8 1991 Mr. Samuel J. Chi l k Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Docketing and Service Branch SUSQUEHANNA STEAM ELECTRIC STATION COtl4ENTS ON PROPOSED RULE - REVISION TO 10 CFR 50.55A PLA-3568 FILES R41-2. A17-11

Dear Mr. Chilk:

• 91 APR 22 P 3 :39 Docket Nos. 50-387 and 50-388 Pennsylvania Power & Light Company (PP&L) is submitting comments in response to the proposed rule incorporating the 1986 Addenda, 1987 Addenda, 1988 Addenda and 1989 Editions of ASME Boiler and Pressure Vessel Code,Section III, Division 1 and Section XI, Division 1, by reference into 10CFR50.55a (55 Federal Register 53220 - December 27, 1990).

PP&L supports the addition of later addenda and editions of the ASME Boiler and Pressure Vessel Code, Sections III and XI to 10CFR50.55a.

However, we would request that the Commission consider the following:

A clearer definition of "essentially 100%" would help licensees assess RPV shell weld examination l imitations so inspection alternatives could be evaluated.

If based on the total length of RPV shell welds approximately 90%

examination coverage for all the RPV shell welds would be accomplished, it is not clear that this 90% examination coverage would be acceptable in accordance with the proposed revision.

For example, two longitudinal shell welds having only 22% examination coverage from the outside of the vessel, but could be 100% exami ned from the i nside of the vessel may not be acceptable under the APR.?. 4 199.l Acknowledged by card..................................

U.S. UCLEAR REGULAiORY COMMI ION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMiSS!ON Document Statistics Postmmt Date _ 1_-_1_? __ _

Co~es Received

~

Add'I Copies Reprod1Jted

~

Special Distribution !fD:,; J?I::I)s, m, LlmA~ v-;;,ri/Jbel/

I I

r FILES R41-2, A17-11 PLA-3568 Mr. Samuel J. Chilk proposed rule change even though the overall examination coverage is 90% based on the total length of all RPV shell welds, and would be "essentially 100%".

A clearer understanding of what is meant by "essentially 100%" is needed.

Very truly yours, w~

H. W. Keiser cc:

NRC Document Control Desk (original)

NRC Region I Mr. G. S. Barber, NRC Sr. Resident Inspector Mr. J. J. Raleigh, NRC Project Manager

DOCKET NUMBER PR

~2 5

PROPOSED RULE SO

~

C5(p F/2. e:>"3 19~

• l3i:NERALOFFICE N

~.*o

.*.11.-

=================================P.O=.=BO=t=~=t\\,=;~=op=L~o~MN~B~==;~~* ~~f!=~=~i=s~=~=68=60=2--0=4=99=

=Nebraska Public Power District FAX(402)563-sss1

• 91

~PR 22 P 3 :23 NLS9100241 April 15, 1991 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 A TIN: Docketing and Service Branch

Subject:

Notice of Proposed Rulemaking:

Codes and Standards for Nuclear Power Plants 56 Federal Register 3796 (January 31, 1991)

Dear Mr. Chilk:

In accordance with the above Notice of Proposed Rulemaking, the Nebraska Public Power District (the District) submits the following comment on the NRC's proposed amendment to 1 0CFR50.55a. The District appreciates this opportunity to express its views on the promulgation of this proposed rule.

The proposed text of 1 0CFR50.55a(g)(6)(ii)(A)(1) would revoke all previously granted reliefs for reactor vessel shell weld examinations for the lnservice Inspection Interval in effect on the effective date of the rule. This includes reliefs for those licensees with fewer than 40 months remaining in the Inspection Interval that is in effect on the effective date of the rule. The District requests that consideration be given to revising the proposed text of 1 0CFR50.55a(g)(6) (ii)(A)(1) to specify that previously granted relief requests for examination of reactor vessel welds will not be revoked for those licensees who elect to defer the augmented examination as allowed in 10CFR50.55a(g)(6)(ii)(A)(3). This will allow those licensees who are able to defer the augmented reactor vessel examination to the first period of their next 1 0 year Inservice Inspection Interval to do so, while still meeting their current interval examination requirements without submitting new relief requests.

The District believes that incorporating this comment into the proposed rule will not weaken the purpose of the proposed rule, but rather would clarify it and aid licensees in their implementation of it. Your consideration of this comment is appreciated.

sell]~

~

Horn Nuclear Power Group Manager

~PR 2, 4 1991 Acknowledged by card........................ --..

Powerful Pride in Nebraska

U.S. NUCLEAR Rt.GL:v ;, J 11 t v~i.fo,11 :::.ION OOCKETING & SEf-lVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics J)ommar1( Date ___.;,..,___._......_-'-+--.........,--

Copies Received

/

Add'I Copies Reproduced -~---

Spec;2*

~!*t'on PPl2) l?ID

-'}J, Ll.m4"".J.<2A-01t26e II

Florida Power CO RP ORA T ION April 17, 1991 NL91-0017 Mr. Samuel J. Chilk Secretary of the Commission DOCKET NUMBER Pft ~

PROPOSED R~

F,e 0"3?q,£}

• 91 APR 22 P 3 :32 U.S. Nuclear Regulatory Commission Attn: Docketing and Service Branch

Subject:

NRC Proposed Rule, "Codes and Standards For Nuclear Power Plants"

Dear Sir:

Florida Power Corporation endorses the comments provided by NUMARC (Nuclear Utility Management and Resources Council) on the subject proposed amendment to 10 CFR 50.55a.

Sincerely, l.!.~f.

setior Vice President Nuclear Operations PMB:LVC:wla A Florida Progress Company

U.S. NUCLEAR REGt.JLATOiiY COM ~ISSION OOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Dat1 __,___-~ / -~-----

Copies Recmlf'.r; ______ / _--,,-,. __

~~'.~~i~

0

~:t~i~~~;~~-.tb£,

L J) ~

mr It m,4 N;

<!Ao/fa' IL

45-5124 April 15, 1991 LIC-91-0063 Secretary DOCKET NUMBER PROPOSED RULE PR 5 0

{§Co FR-O~~~)

~:*~.~-: ~Nj)~,iX!t-~;f-:*' Pt~~~.,*<:-.\\*:~*~

., / _,,~1tt;,,.-~t:~8"~'.

~A

(

~

~

S Omaha Public Power District 444 South 16th Street Mall Omaha. Nebraska 68102-2247 402/636-2000

• 91 APR 22 P 3 :32 U. S. Nuclear Regulatory Commission Attn: Docketing and Service Branch Washington, DC 20555

References:

Gentlemen:

SUBJECT:

1.

2.

Docket No. 50-285 Codes and Standards for Nuclear Power Plants 56 Fed. Reg.

3796 (January 31, 1991) Comments on Proposed Rule Comment on Proposed Amendment to 10 CFR 50.55a These comments are submitted by Omaha Public Power District (OPPD) in response to the proposed rule of the U.S. Nuclear Regulatory Commission incorporating the 1986 Addenda, 1987 Addenda, 1988 Addenda and the 1989 Editions of ASME Boiler and Pressure Vessel Code,Section III, Division 1 and Section XI, Division 1, by reference into 10 CFR Part 50.55a (55 Fed. Reg. 52330 - December 27, 1990).

Regarding the proposed amendment concerning Containment Isolation Valves (CIV),

OPPD has the following comments:

Omaha Public Power District opposes the proposed amendment to modify the provisions of OM Part 10 to eliminate the exemption of Containment Isolation Valves not providing a reactor coolant system pressure isolation function from mandatory leak rate analysis and corrective action if the valve, or combination of valves, fail to meet the leak rate acceptance criteria.

OPPD ' s position is based on the concern that the amendment would require and result in unneeded shutdowns or outage extensions when the 0.6 overall containment leak rate acceptance criteria is not in jeopardy.

Regarding the increased scope of reactor pressure vessel ISi from 100% of one circumferential and one longitudinal shell weld to 100% of the length of all reactor pressure vessel shell welds, OPPD has the following comment:

OPPD is opposed to the proposed amendment.

The bas is for selection of welds for examinati on under the ISi Program is statistical in nature.

The statistical nature of examination selection is based on the assumption that components (e.g. welds) that are carefully controlled during manufacture will behave (X)

Employment with Equal Opportunity Male/Female APR 2 4 19 Acknowledged by card............... "....... __..

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFF!CE OF THE SECRETARY OF THE C0),,,1(*!.1$SION l)_,q.1menl S~t1stlc6 Pomtark Da,? :{~-...:../....c.~--a-----

Copies Recs,.'):, ~-~L----.----

Add'I Copi=*;; f.ep;t:, 1~ed ~-

~'-'::=""'"'1---

i~! ol tt:rutic,n~-

LIC-91-0063 Page 2 similarly under similar operating conditions. Therefore, a 100%

examination sample is not felt to be necessary to detect component degradation.

The proposed amendment appears to be contrary to this philosophy and the cost of the additional examinations and extended outages is not felt to be justified by a commensurate improvement in confidence in RPV integrity.

Regarding the proposed expedited examination requirement for the !SI of 100% of the length of all reactor pressure vessel shell welds, OPPD has the following comment:

OPPD is opposed to the proposed amendment.

Since Fort Calhoun Station (FCS) is planning their RPV examination during the Winter 1992 Refueling Outage, and since FCS is in the 3rd period of our Second Ten Year inspection interval, the proposed expedited schedule would require OPPD to either react rapidly to expand the scheduled RPV examination scope to 100% in less than one year, or face the potential requirement of removing the core support barrel again during the 1st period of the 3rd ten year interval (prior to 1996). Thus, the proposed expedited schedule that is intended to prevent certain plants from going 20 years before performing a 100% examination of their RPV shell welds could require FCS to perform two RPV examinations in less that five years.

While expansion of a presently scheduled examination from 22% to 100% of the shell welds would cost a substantial amount of money and increase the 1992 outage by an estimated 3.5 days, performing an additional examination during the 1993, 1995 or 1996 outages is expected to cost over ten times as much and require a 17 day outage extension.

OPPD feels that neither of these costs justify the benefit to be gained by expediting the 100% RPV shell weld examination.

Should you have any questions regarding OPPD's comments, please contact Mr.

Chuck Bloyd directly at (402) 533-6921.

Sincerely, W. G. Gates Division Manager Nuclear Operations WGG/pjc c: LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator, Region IV W. C. Walker, NRC Project Manager R. P. Mullikin, NRC Senior Resident Inspector Document Control Desk

II Duke Power Company Nuclear Production Dept PO Box 1007

~~2~!~UM.~~~ PR 50 Charlotte. NC 28201-100'7

{f_6Fi!.0379~ __ -1_,r-.i_ -;u-:, -

UJNHC DUKEPOWER April 15, 1991 The Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.

20555

• 91 APR 22 A11 :05 "r,, L:

:,: r\\ t-. l;... 1: ~,

t)UCKt i 1N[_) ', ::; t* Vlf:f

[,rn,\\NC 1i Attention: Docketing and Service Branch

Subject:

Codes and Standards for Nuclear Power Plants Proposed Rule Duke Power Company Comments

Dear Sir:

I MS TUCKMAN Vice President Nuclear Operations (704)373-38.51 In the Federal Register (56FR3796) dated January 31, 1991, the Nuclear Regulatory Commission published for comment a proposed amendment to 10 CFR 50.55a.

Duke Power Company has reviewed the proposed rule and concurrs with the comments submitted by NUMARC.

Very truly yours, t-\\_~-~~

M.S. Tuckman Acknowledged by card....... ~f.~... t~.. ~:.!_

L1.::i. *0cLc:AR R::GuJ:1 :)~1

• coMrJ:°ss:or~

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postma~ Date _._//_--"/_h ____ _

Coples Received _

Add'! C i

~~g~61~i~~~; PR t>

RC ROWLEY CONSULTANTS

~

r£ CJ3 7'1 i,)

Holly Forest Lt.lChL; UJ P. 0. Box 980 U~NRC Cashiers, N.C. 28717 (704) 743-2925

• 91 APR 22 A11 :07 April 12, 1991 Mr Samuel J. Chil k Secretary

• JF'!CE :*i StC~Fii\\r7 "

OOCK['i ING *, ~, il Vl'T US NRC bt L*1 Washi ngton, DC 20555 Subject : Proposed Amendment to 10 CFR 50.55a Endorsing ASME Codes Reference : Federal Regi ster, Vol 56, No 21, dated Jan 31, 1991

Dear Mr Chilk :

My comments are as follows:

(1) The proposed exception to OM Part 10, Inservice Testing of Valves in LWR Power Plants of ASME/ANSI OM-1987 (including the ASME/ANSI OMa-1988 Addenda), is totally unnecessary.

(2) The previous requirements for IST of valves (BPVCode, Sec-tion XI, IWV, prior to 1988 Addenda) contributed to over-lapping requirements for the same containment isolation valves from these documents :

• 10 CFR 50, Append i x J

• BPVCode,Section XI, Subsection IWV

( 3) When OM Part 10 was developed these overlappi ng requirements were spec i fically addressed by the ASME O&M Committee and via the technical consensus process (which the NRC was part of) the following paragraph was inserted into OM Part 10:

( 5) 4.2.2.2 Contai nment Isolation Valves Catagory A valves, whi ch are containment isolation valves, shall be tested in accordance with Federal Regulation 10 CFR 50, Appendix J.

Containment iso-lation valves which also provide a reactor coolant system pressure i solation function shall additionally be tested i n accordance with para 4. 2. 2.3.

Note that para 4.2. 2. 3 is Leakage Rate for Other than Contain-ment Isolation Valves, which covers freaquency, differential test pr essure, seat leakage measurement, test medium, analysis of leakage rates, and corrective action requirements.

Then the ASME BPVCode Committee deleted its previous IWV re-quirements and adopted OM Part 10 by reference.

So two ASME consensus committees looked at this issue and both agreed to this change.

Acknowledged by card............... ".. "".. --**

IJ :. **'..K,... i..:,'1h nc"'vUl i RY COMM,SSIO DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date _1_-___,;_t_7..__ __ _

Copies Received __

• dd'I Copies Reproduc~ _---,,--:.-,,,....--

';pecial Distribution r.Dtf?. I? IJJ 5' D1 1 /lfhA:e1, ~);et/ 7 I

--r

Mr Samuel J. Chilk page 2 April 12, 1991 (6) Although there are related aspects of these two sets of re-quirements, they really have different objectives:

• Appendix J--barrier to radioactive gaseous (from the core source term) leakage through the contain-ment boundary in an accident scenerio with result-ing dose to the public.

• IWV--part of the operational readiness program for engineered safeguard systems.

Related to this proposed amendment to 10 CFR 50.55a is another nuclear industry problem--that is Appendix J has been in need of modification for almost twenty years, yet the only modifications issued have been edito~ial (NRC vice AEC, etc).

I propose that the entire set of requirements for CLRT in Appen-dix J be transferred to a new section of the ASME OM Code (Sec-tion CLT).

This would be in accordance with the Federal Policy (promulgated by 0MB) that encourages the use of nuclear industry programmatic documents in lieu of regulations.

Then an ASME con-sensus committee could develop, approve, and publish the needed changes.

This proposal is already being discussed within the NRC and the ASME O&M Committee--so far it appears quite attractive.

Since I am a senior independent consultant in the nuclear indus-try with substantial experience in both CLRT and IST, I would be happy to provide additional comments as appropiate.

Sincerel

~

C.Wesley Rowle Consultant

DOCKET NUMBER PR VERMONT y ANKEE PROPOSED RULE 56> 50 03 ',,9 NUCLEAR POWER CORPORA TIO~JJt 0

April 15, 1991 Ferry Road, Brattleboro, VT 05301-7002

~PR 22 1\\11 :'\\ 2 James P. Pelletier Vice President, Engineering (802) 257-5271 Mr. Samuel J. Chilk Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Docketing And Service Branch

Dear Mr. Chilk:

1. Proposed Rule -

10 CFR Part 50: RIN 3150-AD05 Codes and Standards for Nuclear Power Plants, January 31, 1991

2. Standard Review Plan for the Review of License Renewal Application for Nuclear Power Plants, Draft Report for Comment, NUREG 1299 (November 90)

3. Draft Regulatory Analysis -

"Amendment to 10 CFR

& 50.55a Codes and Standards," January 4,1991 (Backfit analysis for Reference l)

4. BWROG-PLEX/EPRI Joint Meeting, March 25, 1991 Vermont Yankee's preliminary review of Reference 1 yielded the following comments:

1. Reactor Vessel Shell welds:

Reference 1 proposes to adopt the 1989 Section XI criteria for reactor vessel shell welds. These criteria require 100% volumetric examination of these welds.

Reference 1 further states that required actions "are of a routine nature that should not increase the potential for a negative environmental impact." Vermont Yankee supports prudent testing of reactor vessel shell welds to help ensure reactor vessel integrity both now and into the future. We are aware of (and have been investigating) the various internal and external techniques being used to examine these welds in a boiling water reactor. We are concerned, however, that some of these techniques are relatively new and have not received industry-wide BWR use. There are only a select few vendors that are developing these techniques, and the current proposed rule may strain these vendor resources and ultimately the licensee's ability to implement the rule.

Additionally, to achieve 100% volumetric examination, Vermont Yankee and other early BWRs would need to perform extraordinary measures (such as disassembling the biological shield) which would significantly impact our ALARA program.

APR 2 4 1991 Acknowledged by card....... __..,.,Mfs.u...... >>

U.S. NUCLEAR REGULATORY CO, \\MiSS10N DOCKETING & SERVICE SECTION OFFICE OF THE SECRET ARY OF THE COMMISSION Document Statistics Postmartc Date --..L..---1--~----

Copies R8ceived !

Add'I Copies Reprod~

J{

Special DistritxJtion r ~D

~ R ::~ Sy

{YI, If ov+"'. c.;.;;4 ai._-6 e IL

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission Page 2 The following quote from the draft NRC Standard Review Plan on License Renewal (Reference 2, page 37) summarizes this concern:

"BWR [Reactor Pressure Vessel] RPVs The methods and requirements of ASME Code,Section XI (Ref.

3), BWR RPVs are the same as those for PWR vessels. However, many older BWRs have very limited accessibility for external ISI of the vessel. Typically 75 to 90 percent of the vessel weld lengths are exempted because they are inaccessible. The only alternative is ISI methods of examination from the inside surface. This is of particular importance at the bel tline welds. Minor and major repairs were made to shell plates during construction, but some of these cannot be examined because of limited accessibility. This is also true for some of the pipe-to-nozzle welds that were not confi<:1ured to facilitate ISI. Their configuration was changed in later reactors after the requirements of ASME Code,Section XI were published."

Vermont Yankee proposes that the proposed rule be modified to permit BWRs to develop an equivalent inspection plan, that inspects (and appropriately analyzes) those accessible beltline welds that have been exposed to relatively high fluence or have been known to be reworked. This is technically acceptable in that:

1. Per the draft NRC Standard Review Plan on License Renewal (Reference 2, Page 35),

BWRs receive 80 times less fluence than PWRs and "although irradiation embrittlement should not be ignored, it is of lesser concern in BWR RPVs than in PWR RPVs."

2. BWR beltline welds in the lower jet pump/shelf region are below the core region, and well below the maximum core fluence. These welds will be impacted less than those located higher in the BWR vessel. By assuring maximum beltline weld integrity higher in

• the vessel

annulus, extraordinary insl?ection measures, lower in the vessel annulus can be avoided.

Based on the above, Vermont Yankee also suggests that the following statement be inserted into Reference 1, Page 3798:

For BWRs, the NRC intends that only accessible beltwelds (both internal and external) be volumetrically examined, and that extraordinary examinations (i.e., vessel/biological shield disassembly) are not intended by this proposed rule.

The NRC further recognizes that at this time, that internal BWR beltweld insl?ection techniques are new and may take time to implement. This proposed rule intends that these internal inspection techniques be sought and implemented on a prudent basis.

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission Page 3

2. Backfit Analysis for Reactor Vessel Beltline Welds-Vermont Yankee has reviewed the backfit analysis (Reference 3) for this proposed rule and has the following comments:

A) Crack Propagation-The backfit analysis on Pg A-3 and A8, noted that stress corrosion cracks have propagated from the stainless steel cladding into the carbon steel base metal based on Quad Cities 2

inspection.

VY understands per Reference 4, that the observed stainless and carbon steel cracks were independent, and that there was no crack propagation from clad to base metal.

Additionally,

we understand that the root cause failure mechanism is not yet totally known. Therefore this information can not be directly applied to beltline welds.

Similarly, comparisons were also made with feedwater nozzles (page A-9) to beltline welds. This is not appropriate in that the primary wear mechanism for feedwater nozzles is fatigue and not stress corrosion cracking or embrittlement. VY recommends that these issues be further investigated to ensure that the backfit analysis is properly supported.

B) Available Technology-On Pg 7, the backfit analysis states that technology is being developed commensurate with the proposed schedule. As of today:

• Per Reference 4, there are no code approved interim inspection methods for internal BWR beltline welds.

• There are only 2 vendors that have made significant progress in BWR in-vessel UT measurement; however, the "behind jet pump" delivery system is still under development and has not been functionally tested for lower annulus inspections.

• Without the development of the above two items, there are significant implications, both ALARA and economic.

Using a system that is not fully tested could yield improper

results, and have unnecessary actions implemented in the field. Performing the inspections, even under perfect conditions will cost up to

$6,000, 000/BWR based on the NRC' s estimate of

$2,000,000/inspection and downtime cost estimate of

$400,000/day (pages A21-22).

Based on the above, VY recommends that the cost/benefit analysis and implementation schedule be reexamined for BWRs.

VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission Page 4 C) Reg Guide 1.99 Embrittlement Results-Per page A-9, the NRC stated that "recent surveillance capsule reports from U.S.

BWRs **...* are not as conservative as anticipated". VY has noted that the Reg Guide is very conservative in making assessments based on limited data which is totally a(>propriate for reassessing vessel pressure/temperature operating curves but not totally appropriate for beltline embrittlement evaluations. For example, VY determined that its worst case material to have an NDT of 40 degrees F via the Reg Guide, and this provided conservative pressure/temperature heat up curves. However independent inspection of the material by a national lab, indicated these materials tc have an NDT of 10 degrees F. Further, Reg Guide 1.99, Rev. 2, becomes a better predictor as more sample results are gathered. The net result is that (as stated in Reference 2) BWR embrittlement is not considered significant at this time.

Additionally, per Reference 4, accessible beltline weld inspections performed to date show no unanticipated beltline weld embrittlement.

3.

Using Generic Letters for Rulemaking-Vermont Yankee is concerned about the way the NRC has referenced Generic Letter 89-04 in this proposed rule. In general, Vermont Yankee sees generic letters as a way for the NRC (who has an integrated view of plant operation) to bring to licensee attention potential industry generic weaknesses that the licensees can then use to improve their plants. As presented in this rule, it appears to be used as a form of rulemaking. This can be counterproductive. For Vermont Yankee and our Inservice Test program, Generic Letter 89-04 is addressed with Region I. However, as presented in the rule, it appears that non-safety pumps and valves will be added to this scope and undo work already achieved. This appears to VY to be well beyond the intended scope of the generic letter and Section XI. VY recommends that the NRC delete the first three paragraphs on Pg 3798, column 1, of Reference 1, in lieu of Generic Letter 89-04, which is already issued and implemented at many plants.

Thank you for this opportunity to comment on the proposed references. Please acknowledge receipt of this letter.

Sincerely, gfP~--

J. P. Pelletier Vice President,Engineering cc: J.Gary Weigand, President/CEO, VY W.H. Raisin, Division Director, NUMARC APR 2 4 1991 Acknowledged by card..................................

consumers Power POWERINli MICHlliAN'S PRDliRESS DOCKET NUMBER PR J::'/J PROPOSED RULE...:...:.;.~v'V

--~

~

@J rs, F£ tJa7rdJ. __ l, t\\l *,i 1~ /9

\\:

u<:iN\\*,C I\\PR 22 ~1 i *.i 4 General Offices :

1945 West Parnall Road, Jackson, Ml 49201 * (517) 788-0550 April 1 7, 1991 Mr Samuel J Chilk Secretary of the Commission US Nuclear Regulatory Commission Washington, DC 20555 ATTN:

Docketing and Service Branch

SUBJECT:

CODES AND STANDARDS FOR NUCLEAR POWER PLANTS 56 FR 3796 (1/13/91)

COMMENTS ON PROPOSED RULE

Dear Mr. Chilk:

These comments are submitted by Consumers Power Company in response to the proposed rule of the NRC incorporating the 1986 Addenda, 1987 Addenda, 1988 Addenda and the 1989 Editions of the ASME Boiler and Pressure Vessel Code,Section III, Division 1 and Section XI, Division 1, by reference into 10 CFR Part 50.55a (55 FR 53220 - December 27, 1990).

Augmented Reactor Vessel Inspection Since the proposed rule and associated regulatory analysis states that all previously granted reliefs for reactor vessel shell weld examinations will be revoked, we request that the NRC state their willingness to accept requests for specific exemptions.

Consumers Power Company believes that is is not the intent of the NRC to require extraordinary measures to complete the examinations.

Some early nuclear plants like Big Rock Point were designed and constructed prior to the development of ASME Section XI and its access and examination requirements.

Thus some of the required inspections are not feasible due to access limitations, necessitating the need for specific exemptions.

On November 1, 1985, Consumers Power Company was granted relief from volumetric exams on portions of the reactor vessel shell welds found to be impractical due to inaccessibility.

Consumers Power Company believes that these relief requests are still valid and that revoking all previously granted reliefs for reactor vessel exams is not prudent or required to insure the safety of the facility.

Dennis R Hughes Director, Nuclear Services CC:

NUMARC WLBeckman, Big Rock OC0491-0004-NU02

~~K 2, 4: 1SS Acknowledged by card.......... """'""~.:.~,;..n A CMS ENERGY COMPANY

U.S. UCLEAR REGULATORY COMM1S::,10 DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date _ lf-4-----"-l 7__.__ ___ _

Co~esRerei~-- -~---

Add'I Copies Reprodoced _ _.__=---

Special Distribution f?.JJlfJ J? ;£.7) 5 M,'/(m1t-tv7 ~.6ell,:

WPSC!414l433-1598 TELECOPIER !4141433-5544

~,~~~~ULEPR 5fJ

• 7)

/,;;;,

(5'1pr£ tf)379':J NRC-91-051 ~

EASYLINK 62891993 COCK[1CO WISCONSIN PUBLIC SERVICE CORPORATION U'.:> NRC 600 North Adams

• P 0. Box 19002

• Green Bay, WI 54307-9002 April 15, 1991 Secretary of the Commission Attention Docketing and Service Branch U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Comments on Proposed Amendment to 10 CFR 50.55a

• 91 APR 22 Al 1 : 1 5

,_ ~ r-, :

~ ~. 1..., j \\ t r /\\ :

t_.,J(_;f'\\:* f !th: ', < i I ',,'F:f

~ h

1

• t-h,...

Reference:

1) Federal Register, Vol. 56, No. 21, Thursday, January 31, 1991, pp. 3796-3804 In the January 31, 1991 Federal Register (Reference 1), the Nuclear Regulatory Commission (NRC) proposed an amendment to 10 CFR 50.55a. This amendment would incorporate by reference the 1986-1988 Addenda and 1989 Edition of Section III and Section XI (with a specified modification) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code into regulation.

The resulting regulation would impose augmented examination of reactor vessel shell welds. In addition, it would also separate in the regulations the inservice testing requirements from the inservice inspection requirements by placing the inservice testing requirements in a separate paragraph. Wisconsin Public Service Corporation (WPSC), the licensee for the Kewaunee Nuclear Power Plant (KNPP), reviewed comments submitted to the NRC by the Nuclear Management and Resources Council (NUMARC) regarding the proposed amendment to 10 CFR 50.55a. Although WPSC is in general support of NUMARC's comments, we offer the attached additional comments to stress our most significant areas of concern. WPSC requests that the NRC consider these comments prior to final issuance of the regulation.

APR 2 4 1991 Acknowledged by card......... ffHUNmuu..........

Doalnent Statistics f'ostmark Date ---'.1/ __

I _;;5"

opies Received

/

Add'I Copies Rep,,

Secretary of the Commission April 15, 1991 Page 2 Sincerely,

-~~

~

f. ~

1

• Evers Manager - Nuclear Power SLC/jms Attach.

cc - US NRC - Region III Mr. Patrick Castleman, US NRC US NRC Document Control Desk LIC\\NRC\\N473

ATTACHMENT To Letter from K. H. Evers (WPSC) to Docketing and Service Branch (NRC)

Dated April 15, 1991

Secretary of the Commission April 15, 1991 Attachment Page 1 Endorsing ASME Codes NUMARC has supported the proposed endorsement of later addenda and editions of the ASME Boiler and Pressure Vessel Code, Sections III and XI. Wisconsin Public Service Corporation (WPSC) generally concurs that this is a positive step on the part of the NRC and that endorsing new code editions and addenda is necessary as new technologies and testing methods are developed. However, implementing new requirements carries with it a significant burden.

Therefore, WPSC does not agree with the NRC' s conclusion that endorsing later editions and addenda of section XI of the ASME Code is not a backfit simply because updated section XI requirements are an integral part of the longstanding 10 CFR 50.55a(g)(4)(ii) requirements.

While the regulations require that licensees update their program to the latest edition endorsed by the NRC, the regulations do not require the Commission to update their endorsed code.

Clearly, specifying new code editions imposes new requirements on licensees and therefore is subject to the provisions of 10 CFR 50.109. Therefore, prior to endorsing a new edition of the code, a backfit analysis should be performed to determine if the direct and indirect costs of implementation would be justified in view of the potential increased protection of the public health and safety.

p Secretary of the Commission April 15, 1991 Attachment Page 2 Inservice Testing Requirements WPSC concurs with the proposed amendment to separate the requirements for inservice inspection (ISi) and inservice testing (1ST). Separating the regulations would more clearly distinguish the requirements for ISi and 1ST. In addition, since the requirements for 1ST are currently a subset of the ISi Requirements, the 1ST program could be viewed as less important.

Placing ISi and 1ST requirements in separate sections places equal emphasis on the importance of both programs. WPSC recognized that it was prudent to separate the ISi Plan from the 1ST Plan prior to entering the KNPP second inspection interval. As a result, each program is its own entity and responsibility for implementation of inservice testing and inservice inspections are assigned independently.

WPSC does not agree with the proposed regulation in 10 CFR 50.55a(f)(6)(ii). This allows the Commission to require the licensee to follow an augmented inservice test program for pumps and valves for which the Commission deems that added assurance of operational readiness is necessary. WPSC contends that this rule allows the Commission too much latitude in imposing additional testing requirements and bypasses the controls provided in 10 CFR 50.109. To justify the proposed augmented reactor vessel exams, a backfit analysis was performed although the current regulation does not require it. To ensure consistency, the regulations should state that the provisions of 10 CFR 50.109 will be implemented prior to imposing any augmented tests or inspections.

t Secretary of the Commission April 15, 1991 Attachment Page 3 Containment Isolation Valve Tests WPSC agrees with the NUMARC position that 10 CFR 50, Appendix J provides an adequate basis for testing containment isolation valves (CIVs).

The Kewaunee Appendix J testing program has establishe,d advisory action levels and mandatory action levels, based on effective penetration size, for each type C tested penetration. The ability of containment isolation valves or groups of containment isolation valves to perform their intended function is demonstrated by this testing method. If the Commission determines that it is necessary to retain the requirements of leakage rate analysis and corrective action for CIVs, it is more appropriate to incorporate these requirements into Appendix J of 10 CFR 50.

Therefore, it is WPSC's opinion that retaining the requirements of Subsection IWV for leakage rate analysis and corrective action for specific valves is not necessary.

LIC\\NRC\\N473

N T NIAGARA UMOHAWK DOCKET NUMBER PROPOSED RULE PR 5 ()

($~£ 0379(p NINE MILE POINT NUCLEAR STATION /P.O. BOX 32 LYCOMING, NEW YORK 13093 /TELEPHONE (315) 343-2 11 O

u. s. Regulatory Commission April 12, 1991 NMP lL 0579

• 91 APR 19 P 4 :26 Attn:

Docketing and Service Branch Washington, D.C.

20555 Nine Mile Point Unit 1 Docket No. 50-220 DPR-63 Gentlemen:

Nine Mile Point Unit 2 Docket No. 50-410 NPF-69 On January 31, 1991, the Nuclear Regulatory Commission published for comment proposed revisions to 10CFR50.55a regarding Inservice Testing and Inservice Inspection (50 Fed. Reg. 3796, January 31, 1991).

Attached please find Niagara Mohawk's comments regarding the proposed rule.

xc:

Ron Simard, NUMARC Very truly yours, WZ>.c -

Carl D. Terr~

Vice President Nuclear Engineering Acknowledged by card................... '"......,.....

u.s NUCLt.AH r1cuvu"I I Viii i..,\\.)i~iMt".3ii.)N DOCKETING & SERVICE SECTION OFFICE OF TYE SECRETARY OF THE COMMISSION Document Statistics Postmark Date __ 4...._-_..;..(_0 ___ _

Copies Received_~ /--~---

Add'I Copies Reproduced _ _,3/--=---

Special Distribution P22~) 1,<.I]J ~

/Y\\ i If <<114/'11 Z!4mf /;;e //

7

NIAGARA MOHAWK POWER CORPORATION COMMENTS ON PROPOSED CHANGES TO 10CFR50.55A The proposed changes to 10CFR50.55a will separate Inservice Inspection (ISI) and Inservice Testing (IST) into two separate sections, 10CFR50.55a(f) and (g).

This change will help clarify which requirements are applicable to ISI and which are applicable to IST.

The proposed change anticipates the deletion of IST requirements from ASME XI and prepares the way for the imposition of the O&M Code.

Niagara Mohawk believes additional clarification is required for component supports.

Supports are visually inspected under ISI rules, but snubbers are functionally tested to both technical specification requirements and OM-4.

Referencing OM-4 in 10CFR50.55a(f) would clarify functional testing requirements and would complement Technical Specification improvement efforts.

The proposed changes to 10CFR50.55a(f) endorse the use of OM-6&10 for testing pumps and valves.

OM-6&10 expand the scope of testing programs to include pumps and valves that were not originally designed, purchased, or installed to ASME rules (e.g., Diesel auxiliary systems).

This expanded scope creates a conflict with the Technical Specifications which requires ASME Class 1, 2, or 3 components to be tested.

Furthermore, this expansion in scope does not take into consideration the impact of additional testing on the reliability of these systems nor the fact that these systems were not designed to be tested to OM standards.

Consequently, this rule change will result in more testing on auxiliary components and more requests for relief from testing requirements.

We also object to the more restrictive valve leakage limits being proposed.

OM-10 valve leakage limits are based on the expected leakage of a valve in good operating condition.

These limits are not based on containment leakage limits.

10CFR50 Appendix J contains the overall leakage limits for containment isolation valves.

The Appendix J limit is based on 10CFR100.

It is inappropriate to apply a specific containment leakage limit on a valve-by-valve basis.

If changes to local leak rate test requirements are necessary, they should be addressed in Appendix J, not in 10CFR50.55a.

Niagara Mohawk has no problem with the proposed change to 10CFR50.55a(g).

We have been pursuing alternate techniques for examining the reactor vessel beltline region for several years.

We note, however, that even with the improved techniques, there will still be some access restrictions and that full coverage will not be obtained.

Thus, relief requests for the reactor welds will still be required.

It might, therefore, be more prudent to require existing relief requests to be revised and resubmitted rather than cancel all the relief requests for category B-A weldments.

~

SCE&G ASC/lt.lnCompany DOCKET NUMBER soJ?iHQPo0£EiOtlitUl<bsfB,pany 5 0 P.O. Box 88 Q

o*v Jenkinsville, SC 29065 5bF/::..C> 3 77CP (803) 345-4040 APR 1 5 1991 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

Subject:

VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12

• 91 APR 19 P 4 :26

, t t ! I f

. ; t

' 1, "*

bu(it, i1,[, --: ~*(,,,,,*r

!*l\\,\\N\\ t":

COMMENTS ON THE PROPOSED RULE TO 10CFR50.55a (PR 910003)

South Carolina Electric & Gas Company (SCE&G) has reviewed the proposed rule to 10CFR50.55a (reference 56 Federal Register 3796) endorsing later addenda and editions of ASME Code Sections III and XI.

SCE&G supports the endorsement with the exceptions noted.

SCE&G also endorses the comments submitted by the Nuclear Management and Resources Council (NUMARC).

SCE&G considers that 10CFR50 Appendix J provides an adequate basis for testing containment isolation valves and limiting leakage.

SCE&G implements Appendix J by establishing conservative leakage limits by valve type to ensure total containment leakage is significantly below the allowable limit.

Valves with leakage approaching/exceeding the conservative limit are repaired as soon as practical.

By this means, total leakage it maintained significantly below the Appendix J allowable value.

The proposed rule would require immediate repair or replacement of a valve exceeding the licensee defined limit. This is considered too restrictive, based on the conservative limits already in place.

In order to provide the needed flexibility, SCE&G 1s very conservative limits would be raised to a value closer to the ASME OMa-1988 permissible leakage rate.

We believe this would result in actual containment leakage rates greater than what is allowed in our current program.

The reference to Position 11 of Generic Letter 89-04 (testing of non-code components) in the Supplementary Information section of the rule is confusing since 10CFR50.55a deals with ASME Class 1, 2, and 3 pumps and valves.

Non-code pumps and valves and the associated piping are not usually designed to meet the requirements of ASME Section XI.

SCE&G supports the NUMARC position that ASME Section XI test requirements should not be linked to non-code pumps and valves.

APR 2. 4 \\99\\

Acknowledged by card...........................,......

j

,\\('. r.

JI..,\\L l,i,\\ i :c:t V,(,f:: ((.,,:, i 1vN OF-FICE OF TYE SECRETARY CF THE COMMISSION Document Statistics Postm rk Date i -1 0 Copies Received __ ~~-~--

Add'I Copies Reproduced ____ _

~pecial Distribution,P))R 7? 1]) S (0_, /lMA"' J CA-alphe/1 1

Mr. Samuel J. Chilk PR 910003 Page 2 of 2 SCE&G considers the present snubber testing and inspection requirements of Technical Specifications (with the provisions of Generic Letter 90-09) and ASME Section XI 1977 Edition to be more than adequate.

The requirements of the revision of 0M-4 (0M-1987, Part 4; now ISTD) endorsed under the proposed rule would result in a significant manpower expenditure and radiation exposure with no identifiable increase in any margin of safety.

SCE&G does support ISTD Revision 7 as a replacement for the applicable Technical Specification and current ASME Section XI requirements for snubber testing and inspection.

If you have any questions concerning these comments, please call.

ARR:JLS:lcd c:

0. W. Dixon Jr.

R. R. Mahan R. J. White G. F. Wunder NUMARC General Managers NSRC RTS (PR 910003)

File (811.02, 50.002F)

~di;;&

John L. Skolds NUCLEAR EXCELLENCE - A SUMMER TRADITION!

DOCKET NUMBER AOPOSED RULE PR so, ".. i [,

( 6{.p F", f?. 037ttp}

STATE OF ILLIN0IS DEPARTMENT OF NUCLEAR SAFETY 1035 OUTER PARK DRIVE SPRINGFIELD, IL 62704 THOMAS W. 0RTCIGER DIRECTOR Secretary of the Commission (217) 785-9900 April 11, 1991 U.S. Nuclear Regulatory Commission Docketing and Service Branch One White Flint North 11555 Rockville Pike Rockville, MD 20852

• 91 APR 1 8 p 4 :Q 7 JIM EDGAR GovERNOR The Illinois Department of Nuclear Safety (IONS) appreciates this opportunity to comment on the proposed rule change to 10 CFR 50.55a.

IONS recognizes that adherence to the American Society of Mechanical Engineers Code (ASME) is important to reactor safety, and IONS staff members participate as members of various ASME and Operations and Maintenance (O&M) consensus committees in the Code development processes.

In regard to the proposed changes to 10 CFR 50.55a, IONS has the following comments.

The proposed modification of the 1988 Addenda and 1989 Edition of Section XI, Division 1, Subsection IWV, Inservice Testing of Valves, found in 10 CFR 50.55a(b)(2)(vii), should not be incorporated.

IONS believes that similar requirements should be commonly located in the regulations to avoid licensee confusion.

Paragraph 4.2.2.2 of Part 10 of ASME/ANSI OMa-1988, which is incorporated by reference in the 1988 Addenda and the 1989 Edition of Section XI, was written so that requirements for containment isolation valve (CIV) leak testing would be found in documents other than O&M Part 10.

It was expected that all requirements for CIV leak testing would be placed in 10 CFR 50, Appendix J, as Type C testing.

IONS agrees with the NRC position that the leakage rate analysis requirements and corrective action provisions from IWV in the earlier editions of Section XI for CIVs should be maintained.

However, we believe the proposed modification could perpetuate confusion among licensees about where requirements for CIV leak testing are found.

This could prevent the NRC from achieving its objective of clearly distinguishing in the regulations the requirements for CIV testing.

If the NRC believes that leak analysis and corrective action requirements are necessary, in addition to those currently found in Appendix J, Type C testing, the IONS staff believe that it would be less confusing for the licensees if those requirements were incorporated into the existing requirements for Type C testing in Appendix J.

To keep requirements for CIV leak testini in both Appendix J and the ASME Code could be detrimental to the industry and to plant safety to the extent that doing so may unnecessarily perpetuate existing confusion.

APR 2 4 l99J Acknowledged by card.......................... "......

U.S. NUCLl:AR Rfov~-*....,1 ** *

• J i-.1lv\\'.SS10N OOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark r 41 / Jt5/'J I CopieE,.**

i --,-1-~

Add'I Ct

'f Special L,1 PDR.., Rt D5 _ __

rn, tL man I C-a,mpb...d!f

U.S. Nuclear Regulatory Commission Page 2 April 11, 1991 The scopes of ASME/ANSI OMa-1988, Parts 6 and 10, were modified to delete reference to Class 1, 2, and 3 at the urging of the NRC.

As the supplementary information to the Federal Register Notice states, Generic Letter 89-04 already requires the In Service Testing (1ST) program to be extended to all valves necessary for safe operation of the plant. This has also been the position of the NRC staff for many years prior to the Generic Letter. This rulemaking presents the NRC with an opportunity to end some confusion on the part of the licensees.

IONS thinks related requirements should, to the extent practicable, appear in the same document or reference.

If the NRC believes that inservice testing of pumps and valves beyond Class 1, 2, and 3 is required, this is an opportunity to place the requirement in the appropriate regulations.

IONS strongly supports the NRC position proposed to be codified in 50.55a(g)(6)(ii)(A). Because of Code changes, older reactor vessels may have had only small samples of reactor vessel shell welds examined as part of the inservice inspection program in the first and second intervals. Although the probability of a failure of the reactor vessel is small, the consequences of failure are very large.

IONS believes that this change will make a positive contribution to plant safety.

TWO:rkw Since~

Thomas Director

DOCKET NUMBER PR PROPOSED RULE SO (51PrRC'XYJqf,)

~fOP~B§§l{

Forrest T. Rhodes Vice President Engineering & Technical Services U.S. Nuclear Regulatory Carmission Attention:

Mr. Samuel J. Chilk, Secretary Mail Station Pl-137 W~shL,gton, D.C.

20555 April 12, 1991 ET 91-0071 SUBJECI':

Propose:::i Arrendment to 10 CFR 50.SSa, Endorsing ASME Ccxies

Dear Mr. Chilk:

- ~

1., J LI\\LIL~

USNilC

• 91 APR 18 P 3 :28 Plea.se find enclose:::i the camenLs fran myself and several key nanbers of the ASME Operation and.Maintenance (O&M) Main Camri.ttee on the propose:::i amandrrent to 10 CFR 50.SSa as publishe:::i in the Fe:::ieral Register, Vol. 56, No. 21, date:::i January 31, 1991.

I am the present Chai.nran of the ASME O&M Main Ccmnittee.

We are conceme:::i that the NRC is taking exception to Part 10, "Inservice esting of Valves in Light-Water Reactor Power Plants" of ASME/ANSI CMa-1988 Addenda to ASME/ANSI CM-1987, Operation and Maintenance of Nuclear Power Plants".

CM Part 10 was develope:::i as an industry standard using the consensus process with full NRC invol varent.

we also recamend that the new 10 CFR 50.SSa(f) "Inservice Testing Requirarents" be expande:::i to address testing of snubbers. As currently written, 10 CFR 55a ( g) "Inservice Inspection Requirarents" covers inservice inspection of carp::>nent supports, but it is not clear that paragraph (f) cove.....-rs inse.rvice testing of snubbers.

we appreciate the opportunity to be able to cament on this propose:::i regulation anendrrent. Additional specific caments are attache:::i. If ~ can be of further assistance, please contact ne.

FIR/aew Attachnent Forrest T. Rhodes Vice President Engineering & Technical Services Acknowledged by card--~~~ ** :.. ~-~!.....

P.O. Box 411 / Burlington, KS 66839 / Phone: (316) 364-8831 An Equal Opportunity Employer M/F/HCNET

CXHml1'S Of 10 CFR 50.55a The NRC is proposing to separate inservice inspection ( ISI) and inservice testing (IST), which are currently containe::J. in 10 CFR 50.SSa(g), into separate sections, 10 CFR 50.SSa(f) and (g).

'lhi.s propose::J. change will be better for the industry by clarifying which requirarents are ISI and which are IST.

'lhere does appear to be a oversight which should be corrected and include::J.

as part of the propose::J. change.

Section XI, Subsection IWF makes reference to CM Part 4 Examination and Perfonnance Testing of Nuclear Power Plant Dynamic Restraints ( Snubbers).

Part 4 contains inservice testing requirarents for snubbers and should be referenced. in the new 10 CFR 50.SSa(f).

The carmission has aske::J. for caments relative to the nee::J. for revising, or possibly eliminating the propose::J. nodification to Part 10. 'lhis propose::J.

nocli.fication is not nee::J.e::J., or prudent for the follow.ing reasons:

'lhe ASME Operation & Maintenance 'v<<>rking Group on Pumps and Valves detennine::J. that having two sets of testing requirarents for contairment isolation valves was an elerent of confusion in the cur.rent testing programs.

'lherefore, when Part 10 (CM-10) was developed fran its predecessor IWV, the requirarents for testing Catego:ry A contaimrent isolation valves, (valves in which leakage is.inportant) ~

change::J. to require testing as required by 10 CFR 50,.Appendix J. 'lhe -working group felt that the.Appendix J Program, as cur.rentl y implerente::J. by the industry, is ve:ry credible program, which is specific to containnent isolation valves.

'lhus

.Appendix J Program treats containnent leakage as a system, and acceptance criteria is base::J. on a system approach.

By.inposing the propose::J. nocli.fication, each valve -would have a specific leak rate associated with it and corrective action mandated when the leakage is excee::J.e::J..

'lhi.s approach raroves flexibility fran operations of a unit and could force unnecessa:ry shutdowns.

'lhe leakage limits containe::J. in CM-10 are based on expecte::J. leakage of a valve in gcx::x:i operating cor.di tion.

'lhese limi. ts have not basis when CClTlfxlred to the function of the contai.nrrent.

.Appendix J does contain an overall limit, which was specifically developed for leakage for a contaimrent in order to naintain off-site releases below that establishe::J. by 10 CFR 100. 'lherefore the.Appendix J limit is rrore appropriate.

Page 2 It appears that the reason the NRC is taJdng exception to Part 10 is a perceived. inadequacy in the containnent isolation valves testing p:rograms as contained in 10 CFR 50,.Appendix J. '!he Operation & Maintenanoo Ccmnittee feels that the testing program referenced by Part 10, that is, 10 CFR.Appendix J, is adequate and no augrrented requi:raients need to be imposed by the NRC.

If the

.Appendix J limit is not appropriate, then this limit should be changed instead of imposing additional Code limits.

'!he NRC contends that approxinately 30% of the tine containnent leakage ~uld exC'eed plant technical specifications. It should be noted that these results are based on the very conservative approach that the lowest leakage rate valve of two valves in series has failed to isolate. '!his analysis technique is referred. to as "Maximum Pathway leakage", which goes beyond any plant design basis where the plant is designed to protect the health and safety of the public £:ran the consequenoos of a design basis accident, assuming a single failure. A true neasure of the containnent structure to limit the release of radioactivity within the limits of 10 CFR 100

~d be an analysis of the minimum pathway leakage of each containnent penetration, including the irost limiting single failure.

10 CFR 50.Appendix J, as implarented by irost nuclear utilities, and as enforced by the NRC, requires that the surmation of all containnent penetration leakages, calculated using Maximum Pathway analysis, be less than 0.6L (where L = Allowable Leakage Limit for Containnent) prior to d&laring p~imary containnent operable and subsequently restarting after each shutdown for refueling (where L = Allowable leakage Limit for Containnent). '!his plant technicaf specification requi:raient precludes allowing a containnent isolation valve to be declared operable with gross leakage. '!here exists no basis for applying a specific leakage limit on a valve by valve basis, nor will such a limit appreciably affect the total.Appendix J leakage value calculated via maxi.mum pathway leakage techniques. 'lhe ClL"Te.."'lt.Appendix J/Technical Specification limit is based on the 10 CFR 100 analysis for the function of a valve as it relates to containnent as a whole.

'!he ASME limit (Part 10 paragraph 4.2.2.3(e)) has no basis for application to total contairnnent leakage.

Page 3 The NRC, by this proposed regulatory action, implied that containrrent isolation valves need to be viewed fran both an overall containrrent leakage and canponent specific leakage standpoint. It is true that CXIrp)Ilent specific leakage rate values are inportant.

Every utility has to make a detennination of when to repair a valve. '!his detennination will oonsider many factors, including target values, current leakage rate, leakage rate history, canbined leakage rate of other valves tested, difficulty of repair, availability of spare parts, outage schedule, system availability for testing, etc. 'lhl.s approach to repair adds flexibility, but still ensures that the Appendix J limit is net.

Many utilities establish administrative limits, or target values for each valve.

One nethod used is to divide the allowable leakage

(. 6 L ) by the total valves tines valve size. 'lhl.s gives a limit, based.aon valve size, which will ensure the Appendix J limit is not exceeded. '!his approach is a good guide, but depending upon other oonsiderations as described above, exceeding a target value for any particular valve may or may not lead to repair.

However, when the target value is exceeded., a penalty must be paid by reducing limits on other valve ( s).

A running total of the leakage of all valves is maintained. to ensure the Appendix J limit is not exceeded..

It does not appear that additional regulations, i.e. in addition to Appendix J, is an appropriate way to regulate a testing program. If changes to Appendix J are in order, the NRC can certainly make then. If a new.ANSI standard is appropriate, that can and will be developed. by the oonsensus standard developed. programs of which the NRC is a vital and appropriate participant.

[W DOCKET NUMBER PRoPosEo RULE PR 5o (5CP FR 03r"fqu)

LuLKi.. H.:O USNFC Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 APR 16 1991 Mr. Samuel J. Chilk, Secretary ATTN:

Docketing and Service Branch U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Sir:

• 91 APR 18 P 3 :18 NUCLEAR REGULATORY COMMISSION (NRC) - PROPOSED RULE 10 CFR PART 50.55a, CODES AND STANDARDS FOR NUCLEAR POWER PLANTS TVA has reviewed and is pleased to provide comments on the subject proposed rulemaking recently published in the January 31, 1991, Federal Register (56 FR3796-3804).

TVA supports the rulemaking comments made by the Nuclear Management and Resources Council.

In addition, we offer the enclosed specific comments for the NRC's consideration.

TVA appreciates this opportunity to comment on the proposed rule.

Very truly yours, TENNESSEE VALLEY AUTHORITY Rd~

~E.G. Wallace, Manager Nuclear Licensing and Regulatory Affairs Enclosure cc:

See page 2 Acknowledged by card,mtHII... HH&essantttttstWM

U.S. NUCLcl"lh ht.l-vLATORY CC*'

DOCKETING & SERVICE sr,

OFFICE OF THE SECRE i *'.:'.'

OF THE COMMISSIC~.

2 U.S. Nuclear Regulatory Commission cc (Enclosure):

Ms. s. c. Black, Deputy Director Project Directorate II-4 U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. G. C. Millman Division of Engineering Offi ce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C.

20555 APR 16 1991

ENCLOSURE CONTAINMENT ISOLATION VALVE (CIV) TESTING 10 CFR 50.55a, Appendix J testing of containment isolation valves (CIV), when properly and conservatively implemented, provides an adequate basis for maintaining the function of the components.

The addition of specific individual leak rate limits and imposed corrective actions based on trends places an undue burden on the licensee that is not warranted in light of experience gained from the testing currently being done.

Test data performed by the current Appendix J method has been shown not to be trendable.

American Society of Mechanical Engineers (ASME)Section XI paragraphs IWV-3426 and IWV-3427 and the associated acceptance criteria tables are structured so as to impose percent limits on the change of the measured leak rates.

The individual leak rates measured in accordance with Appendix J are in such small quantities that the data varies back and forth across the reference values and the limits that precipitate required actions.

This is due to the nature of an active component.

A single cycle of a valve open and back closed might change the measured leak rate enough to place the valve under increased test frequency or drop the leak rate below the reference leak rate by a significant percentage.

In addition, valves that have not been operated since the previous test might exhibit an increase in leak rate sufficient to cause additional testing at an increased frequency only to exhibit a decrease in leak rate to values below the reference values at the next test performed.

The result is an unwarranted testing burden on the licensee, a reduction in the operational flexibility to defer maintenance until the next outage period while still remaining within the safe limits of the containment system, which could force unnecessary shutdowns.

Appendix J of 10 CFR 50 is structured to treat the containment as an integrated system while the ASME Code is component oriented.

Appendix J testing, when properly implemented, provides a fully adequate means of maintaining the integrity of the system and protecting the safety of the public.

Though TVA does not believe it is necessary, to the extent NRC wishes to incorporate component objectives and obtain specific CIV component condition information, we believe that a better approach would be to amend Appendix J requirements.

It should also be noted that the proposed change will be in direct conflict with many current utility programs which have requested and obtained relief to perform the Appendix J type of CIV testing in lieu of using the ASME Code paragraphs requiring analysis of leakage rates and associated corrective actions.

In addition, increasing the frequency of tests is in direct conflict with the published position No. 10 of attachment 1 to Generic Letter 89- 04, "Guidance on Developing Acceptable Inservice Testing Programs."

2 APPLICABILITY TO ASME CODE CLASS 1 1 2 1 AND 3 PUMPS AND VALVES Supplementary information in the Federal Register makes it clear that the objective of the proposed change was to move the current requirements into the new paragraphs for the purpose of emphasizing the importance of testing.

In doing this, the words that limit the applicability of the paragraphs to ASME Code Class 1, 2, and 3 components were included.

This is in direct conflict with the stated position No. 11 of Generic Letter 89- 04 and the stated premise for the rulemaking to put more emphasis on testing.

In addition, this maintains a burden on the NRC to impose augmented program requirements for tests on pumps and valves that fall outside these classifications; instead of placing the responsibility more appropriately with the licensee who knows the physical configuration and importance of the component to the integrated plant safety response and is in a better position to decide which components should be included in the program and how they must be tested.

Moreover, the proposed language is contrary to the standards recently published by ASME in cooperation with the NRC.

Rather than adopting past standards in this area, we believe that NRC should use this proposed rulemaking as an opportunity to align 10 CFR 50.SSa(f) with the new standards agreed upon by NRC.

APPLICATION OF THE NEW 10 CFR 40.44a(f) TO DYNAMIC RESTRAINTS The proposed change should be amended to reference dynamic restraint (snubber) testing through ASME Operation and Maintenance Standards, Part 4, "Examination and Testing of Nuclear Power Plant Dynamic Restraints (Snubbers). "

The language of 10 CFR 50.SSa(f) should be broadened to allow inclusion of other components and systems codes and standards that may be appropriate for reference in the future.

GULF DOCKET NUMBER PROPOSED RULE PR 50 (SIPFR o31q(p)

STATES UTILITIES CO.IW.PAN.

RIVER BEND STATION POST OFFICE BOX 220 ST. FRANCISVILLE, LOUISIANA 70775 AREA CODE 504 635-6094 346-8651 Secretary of the Commission U. s. Nuclear Regulatory Commission Washington, D.C.

20555 Gentlemen:

i...; CL*

C APR 18 p 3

07 Gulf States Utilities Company (GSU) is pleased to comment on the Commission's proposed rule regarding codes and standards for nuclear power plants (56FR3796 dated January 31, 1991).

GSU supports the NRC's position in expediting the implementation of the expanded reactor vessel shell weld (Category B-A) examinations specified in the 1989 Edition of section XI, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

However, licensees that are in the first inspection interval when the rule becomes effective and have already completed or scheduled essentially 100 percent of the length of all Examination Category B-A welds as discussed in the proposed 10CFR50.55a(g) (6) (ii) (A) (4) should not have all previously granted reliefs revoked as provided in the proposed 10CFR50.55a(g) (6) (ii) (A) (1).

Many plants, similar to River Bend Station, have developed their Inservice Inspection Programs in strict accordance with the ASME Code and have requested relief for only those instances where it is physically impossible, due to the design of the vessel, to examine welds according to the ASME Code.

Examples of these types of relief include the reactor pressure vessel bottom head welds, the reactor pressure vessel-to-bottom head circumfrential weld and the reactor pressure vessel top head-to-vessel flange weld.

GSU believes that revoking all reliefs without regard to the reason for the relief will cause undue hardship on those utilities who would, in turn, have to resubmit relief requests identical to those which previously have been reviewed and found to be acceptable by the staff.

Furthermore, GSU endorses NUMARC's supporting position on this proposed rule and appreciates the opportunity to provide comments to the NRC.

Manager-Oversight River Bend Nuclear Group APR 2 4. 1991 Acknowledged by card....... "'""'"""'""'~.....

U.S. NUCLEAR AEOOlATORY COMMf8SION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmm Date

'-+ /; 5 J 71 Coples Received ____ /_

Add'I Copies Reprodt.ad --~---

Special Di6'1butm Ppg 'RIP$

m,LlfY¥111, ~bd(

Nuclear Secretary of the Commission

• 91 A.PR 17 P12 :O 1

,:,F~!C:, F ~.f:Crfl !M' DOCKfT *~h, ', [ I VICf

!~f? ~1.li_H 16, 1991 C300-91-1082 U. S. Nuclear Regulatory Commission Washington, DC 20555 Attn:

Docketing and Service Branch

Dear Sir:

GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054 201-31 6-7000 TELEX 136-482 Writer's Direct Dial Number:

Subject:

Request for Comments on the Proposed Rule Regarding Codes and Standards for Nuclear Power Plants GPU Nuclear Corporation herewith submits its comments on the subject rule.

The proposed rule was published in the Federal Register on January 31, 1991 and comments were requested by April 16, 1991.

We agree with the comments being submitted separately by the Nuclear Management and Resources Council (NUMARC).

In addition, four (4) specific comments are offered for your consideration and are contained in the enclosure.

Sincerely, J. Knubel Director, Licensing & Reg. Affairs JK/MRK Enclosure APR 2 4 1991 Acknowledged by card....... :~.:.-.:....................

GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation

U.S. NUCLt.Ah : *~-

_,...)N DOCKETING & SEli vi, *.,,>,:.CflON OFFICE OF THE SECRET ARY OF THE COMMJSSION Document Statistics Postmark Date _ P_t _____ _

Copies Reef',

11. _ 1 __ ____

Add'I Copie:.-.

• 4 ____ _

Speciat Di~:i,t;...

KIDS, PDR I

~

• t-D) C (u(YJ..,C\\

C300-91-1082 April 16, 1991 ENCLOSURE GPU Nuclear Comments on the Proposed Rule Regarding Codes and Standards for Nuclear Power Plants l)

Depending on the timing of the rule and a particular plant's schedule with respect to its current interval, the wording of the rule could require additional reactor vessel examinations within a time frame varying from 40 to 120 months.

Such a requirement will place an undue burden on certain plants while allowing other plants considerably more time.

If it is acceptable for some plants to take up to 120-months to complete the examinations, it would only be equitable for the final rule to prescribe that the additional examinations be performed within 120 months of the effective date of the final rule for all plants.

2)

As written, the rule could create a situation where older BWRs will fail to satisfy the regulatory requirements through no fault of their own.

Currently, there are only two remote ultrasonic manipulator systems available.

Qualification of these instruments is nearing completion; however, these tools have not been proven to provide adequate coverage and ultrasonic sensitivity in the field.

This situation could lead to scheduling conflicts among the utilities and there may be plants that are unable to comply with the new requirements.

Unless this problem is addressed adequately by the rule, plants may be required to request exemptions that will consume the administrative resources of both the utility and the NRC unnecessarily.

3)

The proposed rule does not incorporate Appendix VIII to the 1989 ASME Code Addenda.

This Appendix requires verification of the inspection process by performance demonstration using similar materials and thicknesses on mockups containing both simulated and real flaws.

The nuclear industry has, for the most part, endorsed the concept of Appendix VIII and is vigorously pursuing implementation through the Performance Demonstration Initiative (PDI).

The NRC is aware of this effort and is being kept aware of its progress.

Given the industry experience with intergranular stress corrosion cracking (IGSCC) in BWRs, we are already aware that a Code examination alone may not be adequate to detect expected flaws.

Therefore, we strongly recommend that the rule be changed to reflect that exams be performed using processes that meet Appendix VIII.

The rule would need to consider completion of the PDI which is approximately 2-3 years away.

4)

Relative to revocation of relief requests as proposed, for plants that a r e within the last year of their current interval and have no scheduled refueling outage until the beginning of the next interval, those plants should be permitted to operate until the next scheduled refueling or ISI inspection as appropriate; they should not be obligated to shutdown only to perform the examinations required by this rule.

Georgia Power Company 333 Piedmont Avenue Atlanta, Georgia 30308 Telephone 404 526-3195 0i)

Malling Address 40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Telephone 205 868-5581 OOCKET NUMBER p1t PFtOPOSED RULE n 50 -

(_5~ F~ D37q6) t11e soutt1em el1?ct11c sv'>tem W. G. Hairston, Ill Senior Vice President Nuclear Operations Docket Nos.

50-321 50-366 April 16, 1991 50-424 50-425 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docketing and Service Branch Comments on Proposed Rule "Codes and Standards for Nuclear Power Plants" (56 Federal Register 3796 of January 31, 1991)

Dear Mr. Chilk:

HL-1567 ELV-02704 Georgia Power Company has reviewed the proposed rule, 10 CFR Part 50, "Codes and Standards for Nuclear Power Plants," published in the Federal Register on January 31, 1991.

In accordance with the request for comments, Georgia Power Company is in total agreement with the NUMARC comments which are to be provided to the NRC.

Should you have any questions, please advise.

Respectfully submitted, WGH,111/JMG r :J. ~~~

APR 2 4 1991

__........... *-*-*mmu pJeo,(q pa6pa1MOU>IOV

L*.

DOCKETl1 *G & SERVICE SECTIO OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date 1/- I 7 (FA x)

Co~esReceived. __ ~/'---r----

Add'I Copies Reproduced~

.-'-/

Special Distribution P ~&;~ U s:;

Mi tf roA Al, C.4-ao./Jhet!

7

~

Mr. Samuel J. Chilk Page 2 cc: Georgia Power Company Mr. J. T. Beckham, Jr., Vice President - Nuclear, Plant Hatch Mr. C. K. McCoy, Vice President - Nuclear, Plant Vogtle Mr. W. B. Shipman, General Manager - Plant Vogtle Mr. H. L. Sumner, Jr., General Manager - Plant Hatch NORMS U. S. Nuclear Regulatory Commission, Washington, DC Mr. K. N. Jabbour, Licensing Project Manager - Hatch Mr. D. S. Hood, Licensing Project Manager - Vogtle U. S. Nuclear Regulatory Commission, Region II Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch Mr. B. R. Bonser, Senior Resident Inspector - Vogtle

Alabama Power Company 40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Telephone 205 868-5086 DOCKET NUMBER PROPOSED RULE PR s D I\\

• {5~ F~ 0 3 79t:,/

J. D. Woodard Vice President-Nuclear Farley Project Docket Nos.

50-348 50-364 Mr. Samuel J. Chilk Secretary of the Commission April 15, 1991 U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docketing and Service Branch APR 1 6 1991 oocKETINGl SERVICE BRANCH aECV-tfflC Comments on Proposed Rule "Codes and Standards for Nuclear Power Plants" (56 Federal Register 3796 of January 31, 1991)

Dear Mr. Chilk:

.\\

Alabama Power the southern electric system Alabama Power Company has reviewed the proposed rule, 10 CFR Part 50, "Codes and Standards for Nuclear Power Plants," published in the Federal Register on January 31, 1991.

In accordance with the request for comments, Alabama Power Company is in total agreement with the NUMARC comments which are to be provided to the NRC.

Should you have any questions, please advise.

JDW/JMG cc: Mr. S. D. Ebneter Mr. S. T. Hoffman Mr. G. F. Maxwell Respectfully submitted, w~~

. Woodard (j)

~-***,.. m~ 1,,VMMISSION L-

• C j il~G & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics

~ostmark Date

- 9 f'"A 0 So~esReceived. __

'\\dd'I Copies Reproduced __ J/_,__ __ _

,pecial Distribution P]) I( 1 I? :L. j) S IY\\ i < ( MI1/4: ;.,J CA-ayzhe ({

7

~"-

-===- Entergy Operations Entergy Operations, Inc.

PO Box 31995 Jackson. MS 39286-1995 Tel 601-984-9740

• 91 APR 16 P 1 :50 O,E 7' ns Sup1 r* f Gerald W. Muencw Vice Pre

J, nt April 15, 1991 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:

Docketing and Service Branch

Subject:

Comments on Proposed Rule Change to 10 CFR 50.55a CNRO-91/00007

Dear Mr. Chilk:

Entergy Operations, Inc. has reviewed the proposed rule change to 10 CFR 50.55a, "Codes and Standards for Nuclear Power Plants," posted in Volume 56, Number 21 of the Federal Register.

In accordance with the invitation to comment, we wish to submit the following on the behalf of Arkansas Nuclear One, Units 1 &

2, Grand Gulf Nuclear Station, and Waterford III Steam Electric Station.

1)

The timing of proposed rule implementation in its current form is critical due to the potential to impact future outage work with respect to the 40 month exemption for ISI.

The anticipated implementation date would be very benefi-cial for outage planning.

2)

The term "essentially 100%" inspection of shell welds is vague and should be specific to avoid inconsistent inter-pretation.

3)

BWR-6 plants should be exempted from the 40 month require-ment where committed Code requirements specify 100% shell weld coverage.

The existing program is adequate since the biological shield wall causes mi n i mal inspection limitations, and 100% shield inspections are already required.

4)

Testing requirements of IWV-3426 and IWV-3427 of Section XI of the ASME Code were deleted in Subsection IWV of the 1988 Addenda and the 1989 Edition of the ASME Code for contain-ment isolation valves (CIVs) that do not perform a reactor coolant system pressure isolation function.

Position 10 of Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs," issued April 3, 1989, delin-eated that CIVs are subject to the requirements APR 2 4 J99l Acknowledged by card,... ;;.;.-.;..,.....,..............

OO~!(i:f:NG SERVICE :.~.'.:,rioN OFFICE OF 1HE SECRt: ! s..RY OF THE COM SlO Document Statistics Pbstrnd.Date __

~-----

Copes Received __ / _____ _

Mrl Coples Reproduced __._ ___

Qsdllllon fDB, 6105.J.,

Pl,' I (C] fill

Comments on Proposed Rule Change to 10 CFR 50.55a April 15, 1991 CNRO-91/00007 Page 2

5) of IWV-3426 and IWV-3427(a) only.

Position 10 also stated that the usefulness of IWV-3427(b) does not justify the burden of complying with this requirement based on input from many utilities and staff review of testing data at some plants.

Therefore, we believe the proposed rule change should not reinstate the requirements of IWV-3427(b).

If the requirements of IWV-3427(b) are reinstat-ed, the licensee should be allowed to document noncompli-ance with these requirements of the Code in the inservice testing program only and not be required to submit a relief request to the NRC as is currently allowed by Position 10 of Generic Letter 89-04.

Appendix J of 10 CFR 50, as currently implemented by the industry, provides an adequate basis for testing CIVs.

Therefore, a requirement to perform individual valve leak-age rate analyses is unnecessary and will not provide a substantial increase in the overall protection of the public health and safety. The reasoning is as follows:

a)

An overall containment isolation valve leakage limit already exists which in conjunction with the Technical Specifications, precludes allowing one of the valves in a series to be declared operable with gross leak-age.

b)

The existing leakage limit is extremely conservative because all leakage rates are calculated at peak containment accident pressure; the design basis acci-dent that requires containment isolation is conserva-tive; and leakage rates are calculated by maximum pathway leakage.

c)

The existing leakage limit adequately protects against valves with gross leakage since any valve with a leakage rate contributing a significant portion of the 0.60 La permissible leakage would be analyzed careful-ly, and in most cases, repaired or replaced.

d)

The proposed limits for individual valve leakage are not related to the limits or accidents for which containment isolation valve limits are required as they do not consider the specific purpose of each valve.

Comments on Proposed Rule Change to 10 CFR 50.55a April 15, 1991 CNRO-91/00007 Page 3 We appreciate this opportunity to express our views on the proposed revision and the Commission"s consideration of the above comments.

Sincerely,~~~

(__)

~~

GWM/sep cc:

Mr. T. w. Alexion Mr. s. D. Ebneter Mr. L. L. Kintner Mr. R. D. Martin Mr. R. B. McGehee Ms. Sheri Peterson Mr. N. s. Reynolds Mr. R. Simard Mr. H. L. Thomas Mr. D. L. Wigginton NRC Resident Inspectors Office:

Arkansas Nuclear One Grand Gulf Nuclear Station Waterford Steam Electric Station, Unit 3 DCC (ANO)

Central File (GGNS)

Records Center (Waterford-3)

Corporate File ( 4)

Document Control Desk U. s. Nuclear Regulatory Commission Mail Station Pl-137 Washington, D. c.

20555

DOCKET NUMBER PR 50 BWR PROPOSED RULE (S& FR 037qv;)

OWNERS' GROUP George J. Beck, Chairman (215) 640-6450 BWROG-91043 c/o Philadelphia Electric Company

• 955-65 Chesterbrook Blvd., MIC 638-5

• Wayne, PA 19087-5691 April 16, 1991 Secretary U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTN:

Docketing and Service Branch

Subject:

COMMENTS ON PROPOSED AMENDMENT TO 10CFR50.55a The BWR Owners' Group (BWROG) appreciates the opportunity to provide comments on the NRC's proposed amendment to 10CFR50.55a.

We have sol ici ted comments from our member utilities, and their comments are summarized in the attachment to this letter.

The comments/positions provided in this letter have been endorsed by a substantial number of the members of the BWROG; however, it should not be interpreted as a commitment of any individual member to a specific course of action. Each member must formally endorse the BWROG position in order for that position to become that member's position.

Regards, George J. Beck, Chairman BWR Owners' Group WAZ7/GJB/waz Attachment cc:

R. D. Binz IV, BWROG Vice Chairman S. D. Floyd, RRG Chairman BWROG Primary Representatives BWROG IST/ISI Committee BWROG Containment Testing Committee D. J. Walters, NUMARC R.

Torok, EPRI G.

Oakley, INPO S. J. Stark, GE APR 2 4 1991 Acknowledged by card........................... ".....

U.S. NUCLEA9 REGULA i v1=iY G0MM,SSION DOCKETING & SERViCc: SECTION OFFICE OF T>-'E Sf :"l1c f "RY OF THE v:,.'.vi~ ~-:~)N Do<:t.ltrtf;llt. ta!tSIICS Postmark D, !E _-:_ _ _ (b)__

Copies Rece1,.,..!**.. _L _________ _

Adcfl Copies Hep10CUCbJ. 4. -*- - --

Special Distribution. _,_,:;;_,.=..._...:,..,._al.,__,_ __ _

rhj lf.'O<ln,~eR

ATTACHMENT BWR OWNERS' GROUP COMMENTS ON PROPOSED AMENDMENT TO 10CFR50.55a Comment The NRC is proposing to separate inservice inspection (ISI) and inservice testing (IST), which are currently contained in 10CFR50.55a(g), into separate sections, 10CFR50.55a(f) and (g). This proposed change will clarify which requirements are ISI and which are 1ST.

Additionally, an avenue would be open to reference the new O&M Code directly when appropriate.

Comment The new 10CFR50.55a(f) should be changed to reference OM Part 4, _

Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers).

Section XI Subsection IWF makes reference to OM Part 4, which contains IST requirements for snubbers.

Comment The Commission has asked for comments relative to the need for rev1s1 ng or possibly eliminating the proposed modifications to Part 10.

Th is proposed revision is not needed or prudent for the following reasons:

0 The leakage limits contained in Section XI are based on expected leakage of a valve in good operating condition. These limits have no basis when compared to the function of the containment.

Appendix J does contain an overall limit which was specifically developed for leakage for a containment in order to maintain leakage below that established by 10CFRlOO.

Therefore the Appendix J limit is more appropri ate.

If the Appendix J limit is not appropriate, then instead of t ryi ng to impose Code limits in addition to the Appendix J limit, a change to the Appendix J limit is recommended.

o The NRC contends that containment leakage during operation would exceed plant technical specification limits approximately 30 percent of the time. It should be noted that these results are based on the

• very conservative approach that the best valve of two valves in series has failed to isolatef. This analysis technique is referred to as

Maximum Pathway Leakage". A true measure of a containment structure to limit the release of radioactivity within the limits established by 10CFRlOO would be an analysis of the minimum pathway leakage of each containment penetration, including the most l imiting, single failu re.

o 10CFR50 Appendix J, as implemented by most nucl ear utilities and as enforced by the NRC, requires that the summation of a17' containment penetration leakages, calculated using Maximum Pathway analysis, be less than 0.6 La (where La is the allowable leakage limit for

containment) prior to declaring primary containment operable and restarting after each shutdown for refueling. This plant technical specification requirement absolutely precludes allowing one of the two valves in series which provide a containment isolation function, to be declared operable with gross leakage. There exists no basis for applying a specific leakage limit on a valve-by-valve basis, nor will such a limit appreciably affect the total Appendix J leakage value calculated using Maximum Pathway Leakage techniques.

The current Appendix J/technical specification limit is based on 10CFRl00 analysis and on the function of a valve as it relates to containment as a whole.

The ASME limit [Part 10 paragraph 4.2.2.3(e)] has no basis for application to total containment leakage.

It is recommended that any revisions required for containment isolation valve leak rate testing be made in conjunction with Appendix J. Appendix J programs have proven to be effective in identifying containment isolation valve leak rate problems.

Comment Section 50.55a(g)(6)(ii)(A)(2) would require all licensees to implement the specific augmented reactor vessel examination during the inspection interval in force when the proposed rule becomes effective, with an exception for plants with fewer than 40 months remaining in the inspection interval when the proposed rule becomes effective.

With longer operating cycles, some plants with more than 40 months remaining in their inspection interval may find they have only one refueling outage in which to perform the augmented examination. These plants will find it impractical to implement the augmented reactor vessel examination during their current inspection interval.

For example, the equipment needed to perform the inspections takes time to develop, and at this time the number of vendors capable of performing the inspections is small.

Thus utilities may not be able to find an available vendor when needed.

It is recommended that the proposed rule provide greater flexibility for scheduling this examination.

Comment Paragraph 50.55a(g)(6)(ii)(A)(3) allows plants with fewer than 40 months rema1n1ng in their inspection interval to defer the augmented examination of the reactor vessel shell welds to the first period of the next interval.

This paragraph also prohibits the use of the deferred augmented examination as a substitute for the reactor vessel shell examination scheduled for implementation during the current inspection interval.

The proposed rule should be revised to allow plants within the last 12 months of their current interval to substitute the deferred augmented reactor vessel shell examinations for the remaining reactor vessel shell examinations scheduled in the current interval.

Prohibiting the use of deferred augmented examinations as a substitute for the reactor shell examinations scheduled within the next 12 months would cause additional expenditure of radiation exposure, time and money within a short period of time.

It would be too late to adequately plan to perform the augmented examinations during the next scheduled outage, even though the reactor vessel core barrel would be removed to complete the examination

requirements of the current interval. Since the core barrel must be removed to perform the augmented examinations required by this rule change, a plant would therefore be forced to remove the core barrel again within the subsequent 40 month period.

It has always been recognized within the Code that removing the core barrel is a massive undertaking and performed only once in ten years.

It requires removing all fuel from the vessel, lifting the core barrel partly out of the reactor cavity water (resulting in high exposures on the refueling floor), and adds days to the duration of the outage, significantly increasing the cost of the outage.

DOCKET NUMBER PROPOSED RULE PR 5?

Mr. Samuel J. Chilk Secretary U.S. Nuclear Regulatory Washington, D.C. 20555

{s&FR03 1~)

Commission Attention: Docketing and Service Branch Re:

Proposed Rule P.O. Box 14000, Juno Beach, FL 33408-0420 DOCKETED AP~ 16 ;gg1 DOCKETING.!i, SERVICE BRANCH SECY-NRC

' l9/

APRIL: 1 2 1991 L-91-112 Codes and Standards for Nuclear Power Plants 56 FR 3796 (January 31, 1991)

Request for Comments

Dear Mr. Chilk:

At 56 FR 3796, the NRC requested comments on a proposed rule that would incorporate the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of ASME Boiler and Pressure Vessel Code,Section III, Division 1 and Section XI, Division 1, by reference into 10 CFR Part 50.55a.

These comments are submitted on behalf of the Florida Power & Light Company (FPL}, a licensed operator of two nuclear power plant units in Dade County, Florida and two units in St.

Lucie County, Florida.

The Nuclear Management and Resources Council, Inc.

(NUMARC) is offering comments on the subject proposed rule. In addition to endorsing the incorporation of later editions and addenda of the ASME Boiler and Pressure Vessel Code,Section III and XI into 10 CFR Part 50. 55a, NUMARC provides additional comments that the Commission is requested to consider prior to the issuance of the final rule.

FPL endorses both the additional comments submitted by NUMARC and endorses the incorporation of later editions and addenda into 10 CFR Part 50.55a.

Thank you for the opportunity to comment.

Sincerely, o 1 e Vice President Nuclear Engineering and Licensing cc: R.L. Simard, NUMARC APR Acknowledged by card 2 4 1ss1 an FPL Group company

U.S. UCU:.*, *

-"'r_:r Mrlli~':J!')~J OOCKETINu * :SEflVk~f SECTION OFFICE OF THE SECRETA8 OF THE COMMISSION Doarnent Statistics Postmark Date J.,t / 1 $ I Cj t {£Ax)

Copies 9E're1ved_-'-/ _____ _

Addi U*r,\\. * **-i~ured 4

Spoc1ai ;~,:-

,*

• _./J. I Q'S I P/Je

_{lh.J. /..fil(J,9'),.J}aynf /:xd!t

DOCKET NUMBER PR

~~PROPOSED RULE

!50 i;.

(SbF£0379j}

WASHINGTON PUBLIC POWER SUPPL y SYSTs,rN~-llD P.O. Box 968

• 3000 George Washington Way

• Richland, Washington 99352

• 91 APR 16 A 8 :1 6 Apri 1 11, 1991 Samuel J. Chilk, Secretary U. S Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

PROPOSED AMENDMENT TO 10 CFR 50.55a, ENDORSING ASME CODES Attached please find the Supply System's comments on the proposed amendment to 10 CFR 50.55a as published in the Federal Register, Vol. 56, No. 21, dated January 31, 1991.

/

The Supply System ) ppreciates the opportunity to be able to comment on this proposed regulation amendment.

If we can be of further assistance, please contact Mr. T. F. Hoyle (509) 372-5236.

Very truly yours,

~

~

(280)

~

~~g~i 1tory Programs TFH/bw

Attachment:

Comments on 10 CFR 50 cc:

GC Millman, NRC APR 2 4 10n.

Acknowledged by card...........

Iv.If

• • • • • • • • • • • • • • • .. *****n

.. -.* *"" t:i[GULATOAY COMMI tt----

uUCKETING & SERvrce SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Doamenl Statistics

'-'ostmark Date _ '1+-/1_2.:.../q_1 _ ___ _

~o~ s ReceiVed___j_ ____ _

\\dd't Copies ReprOducoo.;L... 4 ___ _

~~

OislributiOn PPR, 1., ID i 1 ~

m,ucnao CMnPbeli

}

COl'-l'ENTS ON 10 CFR 50 The NRC is proposing to separate inservice inspection (ISI) and inservice testing (IST), which are currently contained in 10 CFR 50.55a(g), into separate sections, 10 CFR 50.55a(f) and (g).

This proposed change will be better for the industry by clarifying which requirements are ISI and which are IST; however, we feel additional work is necessary on the proposed sections.

Please consider the toll owing comments in your final review of 10 CFR 50. 55a ( f) and

( g)

• SNUBBERS There appears to be an omission which should be corrected as part of the proposed change.

Section XI, Subsection IWF, makes reference to OM Part 4 Examination and Performance Testing of Nuclear Power Pl ant Dynamic Retrai nts

( Snubbers).

Part 4 contains i nservi ce testing requirements for snubbers and should be referenced in the new 10 CFR 50.55a(f).

If this change is not made, it will be very difficult for the industry to make needed changes to Part 4 and to get them implemented via Section XI in a timely manner.

INSERVICE TESTING OF CONTAINMENT ISOLATION VALVES The commission has asked for comments relative to the need for revising, or possibly eliminating the proposed modification to Part 10.

This proposed revision is not needed or prudent.

The leakage limits contained in OM-10 are based on expected leakage of a valve in good operating condition.

These limits have no basis when compared to the function of the containment.

Appendix J does contain an overall limit which was specifically developed for leakage for a containment in order to maintain off-site releases below that established by 10 CFR 100.

Therefore, the Appendix J limit is appropriate.

It appears that the reason the NRC is taking exception to Part 10 is a perceived inadequacy in the containment i sol ati on valve testing program as contained in 10 CFR 50, Appendix J.

The Supply System feels that the testing program referenced by Part 10, that is 10 CFR 50, Appendix J, provides an acceptable level of protection of public health and safety, and that no augmented requirements need to be imposed through 10 CFR 50.55a.

If the Appendix J limit (.6La, where L = Allowable Leakage Limit for Containment) is not appropriate then this li~it should be changed instead of imposing additional Code limits.

The NRC contends that approximately 30% of the time containment leakage would exceed plant technical specifications.

It should be noted that these results are based on the very conservative approach that the lowest leakage rate valve of two valves in series has failed to isolate.

This analysis technique is referred to as "Maximum Pathway Leakage," which goes beyond any pl ant design basis where the plant is designed to protect the health and safety of the public from the consequences of a design basis accident assuming a single failure.

A true measure of a containment structure to limit the release of radioactivity within the limits of 10 CFR 100 would be an analysis of the minimum pathway leakage of each containment penetration, including the most limiting single failure.

10 CFR 50, Appendix J, as implemented by most nuclear utilities, and as enforced by the NRC, requires that the summation of all containment penetration leakages, calculated using Maximum Pathway analysis, be less than 0.6 L prior to declaring primary containment operable and subsequently restartin? after each shutdown for refueling.

This plant technical specification requirement precludes allowing a containment isolation valve to be declared operable with gross leakage.

There exists no basis for applying a specific leakage limit on a valve by valve basis, nor will such a limit appreciably affect the total Appendix J leakage value calculated via maximum pathway leakage techniques.

The current Appendix J/technical specification limit is based on the 10 CFR 100 analysis for the function of containment as a whole.

The ASME limit (Part 10 paragraph 4.2.2.3(e))

has no basis for application to total containment leakage.

The NRC, by this proposed regulatory action, implies that containment isolation valves need to be viewed from both an overall containment leakage and component specific leakage standpoint.

It is true that component specific values are important.

At WNP-2 target leak rate values were established for each valve in the Appendix J program.

As experience was gained these values have been modified.

The decision as to when a valve is to be repaired is dependent on many factors, including:

current leakage rate for the valve in question, leakage rate history, combined leakage rate of other valves tested, difficulty of repair, availability of spare parts, outage schedule, system availability for testing, allowable leakage limits, etc.

These decisions, however, do not impact' the ability to meet Appendix J.

The Supply System's Appendix J leakage rate program uses allowable leakage limits.

These limits, if exceeded, require a local leakage determination at the next outage (one year) instead of the normal time of two years required by Appendix J.

This concept is similar to an alert limit of Section XI or Operations & Maintenance.

Generic Letter 89-04 requires leakage limits for each individual valve.

The Supply System has established limits based on IWV-3426.

If the limits of IWV-3426 cannot be met, a larger limit is specified and used as allowed by IWV-3426.

IWV-3426 is a required action limit and some type of rework is required if the limit is exceeded.

For the allowable leakage limits previously mentioned,.6 times the IWV-3426 limit was used for valves greater than 10 inches.

This has now been changed to 1.0 times the IWV-3426 limit to ensure premature corrective action is not mandated.

This example shows that the program will change with the new imposed limits, but it is not a better program, or a program which will provide greater safety to the public.

If Code limits are not mandated, then smaller limits (.6 x IWV-3426 limits) could be used.

The Supply Sys tern maintains a "running checkbook" on CI V leakage and maintains total leakage at approximately 0.25 L versus an acceptance criteria of 0.6 L.

Our philosophy and practice is exempfary, and does not substantiate the NRCis position that "[c]ontainment leakage during operation would exceed plant technical specification limits approximately 30 percent of the time."

We do not feel we need to change our program, and believe the new proposed regulation should provide allowances for currently sound programs like ours.

It does not appear that a Federal Regulation is the appropriate location for standards for a testing program.

It would be more appropriate for the industry to develop a consensus standard (or Code) which would replace or supplement Appendix J.

The reason for this proposal is that it would, in the long run, be easier to change and have acceptance by the NRC, industry and the genera 1 public.

The Operations & Maintenance standards planning committee is currently reviewing the need for an Operations & Maintenance standard on Containment Isolation Valve testing.

SCOPE, INSERVICE TESTING OM-6 and OM-10 require all pumps and valves that perform a safety function to be tested, not just ASME Class 1, 2, and 3 pumps and valves as required by Section XI.

Many valves and some pumps associated with non-ASME safety related systems or components will require testing under OM-6 or OM-10.

The proposed regulation should exempt pumps and valves which are associated with equipment that is tested as a system.

For example, it is not practical to individually test an emergency diesel jacket water cooling pump or temperature regulating valve.

INSERVICE INSPECTION The new paragraph (A)(l), which revokes all previously granted relief requests, should be modified to only revoke relief requests for the second and subsequent i nterva 1 s.

This proposed change to 10 CFR 50 is based on a change in the 1988 Addenda to Section XI, which modifies the 1986 edition to require, in the second, third and fourth inspection intervals, examination of essentially 100% of the length of the reactor vessel shell welds.

The requirement to examine essentially 100% of the length of all reactor vessel shell welds during the first inspection interval has been in Section XI since the 1975 winter Addenda to the 1974 edition.

Therefore, relief requests granted for the first interval should remain valid.

NUCLEAR MANAGEMENT AND RESOURCES COUNCIL 1776 Eye Street. N.W.

• Suite 300

• Washington.DC 20006-2496

[202) 872-1280 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.

20555 ATTENTION:

Docketing and Service Branch April 15, 1991 RE:

Codes and Standards for Nuclear Power Plants 56 Fed. Reg. 3796 (January 31, 1991)

Comments on Proposed Rule

Dear Mr. Chil k:

oocKcteO APR 1 5 1991 These comments are submitted by the Nuclear Management and Resources Council, Inc. ("NUMARC") in response to the proposed rule of the U.S. Nuclear Regulatory Commission ("NRC") incorporating the 1986 Addenda, 1987 Addenda, 1988 Addenda and the 1989 Editions of ASME Boiler and Pressure Vessel Code,Section III, Division 1 and Section XI, Division 1, by reference into 10 CFR Part 50.55a (55 Fed. Reg. 53220 - December 27, 1990).

NUMARC is the organization of the nuclear power industry that is responsible for coordinating the combined efforts of all utilities licensed by the NRC to construct or operate nuclear power plants, and of other nuclear industry organizations, in all matters involving generic regulatory policy issues and on the regulatory aspects of generic operational and technical issues affecting the nuclear power industry.

Every utility responsible for constructing or operating a commercial nuclear power plant in the United States is a member of NUMARC.

In addition, NUMARC's members include major architect/engineering firms and all of the major nuclear steam supply system vendors.

We believe that the proposed endorsement of later addenda and editions of the ASME Boiler and Pressure Vessel Code, Sections III and XI is a positive step on the part of the NRC, and we support this endorsement.

Before the final rule is issued, however, we request that the Commission consider the following comments:

Containment Isolation Valve Tests We believe that 10 CFR Part 50, Appendix J, as currently implemented by the industry, provides an adequate basis for testing containment isolation valves. This was also recognized by the ASME APR 2, 4 199\\

Acknowledged by card....... H..... &M_.auwwu,

S. NUCLEAR 'RtGvU\\ T 0RY L;OMMr5SIO~

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY CF THE COMMISSION Document Statistics ostmarit Date -----1~+:./1~~~/ f,__I ___

o~esReceived.__..___......,_ __

dd'I Copies Reproduced --"'-'-/ ___ _

pecial Distribution PDf<, R !OS,-J W,'llman. ~du£R '

Mr. Samuel J. Chilk April 15, 1991 Page 2 Operations and Maintenance Committee in developing OM Part 10, "Inservice Testing of Valves in Light Water Power Reactors," (which has now replaced Section XI, Subsection IWV by reference).

OM Part 10 changed the requirements for testing Category A containment isolation valves, requiring them to be tested in accordance with Appendix J.

The proposed rule, however, would preserve pre-1988 requirements, which had been removed from Subsection IWV in the 1988 addenda, for leakage rate analysis and corrective action for specific valves.

We believe that retaining these requirements is not necessary since adequate system acceptance criteria are contained in Appendix J.

The Appendix J program treats containment leakage as an integrated system total, while the NRC-proposed modification to Subsection IWV would impose specific leak rates for each valve.

Depending on the results of the leakage rate analysis, additional data analysis and corrective action for those containment isolation valves, which do not perform a reactor coolant system pressure isolation function, may be required.

The NRC proposed modification, which adds the requirements of OM Part 10 Para. 4.2.2.3 to these valves, also applies the requirement of sub-paragraph (f). Thus, a valve could be declared inoperable immediately in spite of the fact that total containment leakage may be substantially less than allowable. Therefore, unnecessary extensions of outages may be incurred.

We understand that the O&M Standards Planning Committee is currently reviewing the need for a consensus standard on containment isolation valve testing.

In lieu of reinstating requirements on specific valves that were removed from Subsection IWV, the NRC could recommend that the O&M committee perform a comprehensive review of containment isolation valve testing requirements and acceptance standards.

Augmented Reactor Vessel Inspection As noted in the Regulatory Analysis prepared for the subject changes, it would be possible, using earlier editions of the code and exemptions granted by the NRC, to meet the requirements of 10 CFR Part 50 without inspecting some vessel welds for 30 years. Technically, it is difficult to justify this lengthy interval for a component as critical as the reactor vessel, and the Section XI Subcommittee has acknowledged this by revising the Code to require a 100% inspection for each inspection interval.

We concur with the NRC staff that these inspections are important; however we have some concerns over the timing of implementation of the 100% inspection requirements.

As noted in the regulatory analysis, the time frame in the proposed rule for completion of the reactor pressure vessel inspections varies considerably from plant to plant, depending on the time remaining in the inspection interval.

Some older BWR plants nearing the end of

Mr. Samuel J. Chilk April 15, 1991 Page 3 the second or third period of a 10 year inspection interval at the time the rule becomes effective can face particular hardship, since the tooling for performing inspections on these plants is still under development and may not be generally available at the time needed.

In addition, because of the need for field experience and assessment of the impact of the examination on the outage schedule, it may be prudent to schedule examinations over several outages rather than a single outage.

Specifically, in order to allow the time for plants to plan and schedule equipment with limited availability, we believe that the proposed rule should be revised to provide additional flexibility, for example, using the following schedule. A plant within the second period of an interval at the time the rule becomes effective should be allowed to defer the completion of the examination until the first period of the next inspection interval. A plant with less than 40 months in the present inspection interval should be permitted to defer the completion of the examination until the second period of the next inspection interval.

The concern over potential susceptibility of reactor vessel shell weld materials to degradation would be greatest in the beltline region, where the welds are exposed to higher neutron flux levels than other locations.

To permit better utilization of available inspection resources, the NRC should consider limiting the application of the augmented inspection program to the beltline reactor vessel shell welds.

Alternatively, a bounding value of neutron flux exposure could be established, limiting the application of the augmented inspection program to reactor vessel shell welds which have exceeded this value.

The proposed rule states that all previously granted reliefs for the examination of reactor vessel shell welds will be revoked.

Only those reliefs which address the scope and extent of shell weld examinations should be revoked.

For example, we would expect that a relief covering a calibration block design for reactor vessel shell weld examination would not be revoked.

Further, we believe the NRC should revoke the reliefs on a plant specific basis, notifying each licensee of those specific reliefs which will be revoked.

Finally, we believe that it is not the intent of the NRC to require extraordinary measures to complete the examinations.

We would expect that internal inspection techniques be developed and utilized on a reasonable and prudent basis.

We request that the NRC state their willingness to accept requests for specific new exemptions, based on the availability of suitable equipment and technology at the time of the scheduled inspection and the appropriate technical justification.

Mr. Samuel J. Chilk April 15, 1991 Page 4 Applicability to Code Pumps and Valves The Supplementary Information is very clear that the proposed rule specifies requirements only for pumps and valves that are classified as Cl ass I, Class 2 or Class 3.

However, the inclusion of Position II of Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs", makes the applicability less clear, since the generic letter (which is not a vehicle for imposing requirements) applies to non-code, safety-related pumps and valves as well.

The testing program for non-code, safety-related pumps and valves may be different from that required for code items.

Some pumps and valves that perform a safety function but are not classified as Code Class I, Class 2, or Class 3 are often very difficult to test individually using the quantitative acceptance criteria required by ASME Section XI.

An example would be some of the pumps and valves associated with the diesel generator whose proper operation can best be verified on a system basis. A system test of the generator can verify proper operation of the equipment and components, while individual equipment or component tests may be impractical.

In general, we believe that ASME Section XI test requirements should not be required for pumps and valves that are not ASME code components, such as pumps and valves that are skid mounted or do not have separate control circuits which allow testing of the individual pump or valve.

We appreciate the opportunity to comment on the proposed rule and would welcome the opportunity to discuss our comments further with the appropriate NRC personnel.

Sincerely, JFC/ALM c<,](fl*f-~

"Je F. Colvin

April 9, 1991 DOCKET NUMBER

~

PROPOSED RULE PR so fl

~

( 5" FR O 3 7 92:) 00CKOEO USNRC COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC. ~'QCR~'

ON PROPOSED RULE, "CODES AND STANDARDS FOR NUCLEAR '~bw~IR( 15 t-'

27 PLANTS," 56 FED. REG. 3796 (JANUARY 31, 1991)

.!~ -~*-.: ! ~ t~ ?_f' 'I if; t("f.

OCRE supports this proposed rule.

OCRE particularly supports the proposed modifications which will require that leak rates be analyzed and corrective actions taken in the event of abnormally high leakage rates for all containment isolation valves, not just those which isolate the reactor coolant pressure boundary.

The revelation that operating experience shows that "containment leakage during operation would exceed plant technical specification limits approximately 30% of the time" indicates that containment leakage is a serious problem.

The NRC is absolutely correct that this situation "indicates a need to improve, rather than relax, the present requirements concerning containment test, leak monitoring and maintenance programs, including the ASME section XI requirement for valve leak rate analysis."

Containment performance is the most important factor in determining the offsite consequences of accidents.

Improving containment performance is essential, and that iyiprovement will be impossible if the licensees do not analy~e valve leak rates to determine the root cause of the problem and then take the appropriate corrective actions.

OCRE also supports the proposed augmented reactor vessel examination and expedited implementation of same, including the revocation of previously granted reliefs.

The NRC has again presented compelling evidence that certain RPV materials have undergone greater radiation damage than previously thought, that stress corrosion cracking of BWR RPVs is more probable than previously thought, and that cracking has indeed occurred in large vessels designed and fabricated to the ASME code.

These concerns are corroborated by the enclosed excerpt from a report prepared by the Ohio State University Nuclear Engineering Program for the Public Utilities Commission of Ohio regarding the General Electric "Reed Report."

This excerpt, Comments by Dr. Michael P. Manahan, Senior Research Scientist at the Battelle Memorial Institute in Columbus, Ohio, addresses BWR reactor pressure vessel embrittlement.

Dr. Manahan states that "more severe embrittlement has been observed in several BWR pressure vessels than origially anticipated and in many vessels the nil-ductility transition temperature shift exceeds the Regu latory Guide 1.99 (Rev. 2) model prediction by a statistically significant amount."

Dr. Manahan provides data to substantiate this statement.

Furthermore, he indicates that

• there appears to be a neutron dose rate effect on RPV embrittlement, that predictive models are based largely upon 1

APR 2 4 1991 Acknowledged by card............................. -*

• u.s.

GLEAA AE LATORY COMM!

IOM DOCKETING & SERVICE SECTJON OFFICE OF THE SECRETARY CF THE COMMISSION Document Statiab l'oslmm Date - ~..;..+-/ '-~,;..;,_1 ___ _

Coples Received_r/ _____ _

Iden Copies Reproduced __,,_ __ 4 __ _

SpeclaDis~ PDR I KIDS Cl-r(J,11cnao, f/;aipi;el/..... * -

~

aoca.

PWR data, and that variables other than copper and nickel content of the RPV base metal may be involved.

These factors suggest that, for BWRs, "any agreement with the RG 1.99 (Rev.

2) trend curve is only fortuitous.

This would, of course, mean that current pressure/temperature operating limits are arbitrary."

Given this situat ion, augmented inspections of RPV welds are necessary and appropriate.

These requirements should be implemented on an e xpedited basis.

These provisions of the proposed rule will not only provide an additional assurance of safety, but will expand our data base which will aid in understanding aging degradation phenomena.

This understanding is essential for any consideration of operating license renewal.

Thus, this proposed rule will assist licensees who wish to pursue license renewal.

OCRE urges the NRC to adopt these proposed rules as soon as possible.

Respectfully submitted, Susan L. Hiatt OCRE Representative 8275 Munson Road Mentor, OH 44060 (216) 255-3158 2

Comments on Issues 5 and 24 by Michael P. Manahan The MHB report mixes quotes pertaining to the pressure (ferric steel) and the core internals (many of which are fabricated from stainless steel).

The technical issues should be separated into two groups by material types (austenitic steel issues and ferric steel issues) since the in-service damage and fracture mechanisms of concern are different.

I believe that the current state of knowledge concerning embrittlement and cracking of these materials is not, at present, adequate to justify closure of Issues 5 and 24.

I do believe that there are sound, conservative, and justifiable engineering solutions to these problems which would enable safe operation during the interim period over which the industry could gather more data and knowledge and thus improve predictive capabilities.

It is my opinion that many of the technical arguments presented in favor of issue closure are not adequate based on current knowledge and do not constitute a consistent, logical framework which justifies issue closure. The specific reasons for my concerns are described below.

Austenitic steel Embrittlement and cracking While ductility loss, corrosion, erosion, and wear are valid concerns, which must be guarded against, stress corrosion cracking 30

(SCC) and irradiation assisted cracking (IAC) are is likely to be the dominant failure mechanisms for austenitic steel in the foreseeable future.

Knowledge concerning sec has grown over the past 20 years.

Early cracking problems were attributable to the use of sensitized stainless steel.

Later, the mechanism of IASCC was identified in relation to cracking in non-sensitized stainless steel.

Recent, work at Battelle [1,2 ) and Westinghouse [3) on PWR control rods (type 304 stainless steel) has shown that intergranular fracture can occur at high fluences, without a corrosive medium, and at stresses at or below the yield stress.

Additional work is needed to fully characterize this newly discovered fracture mechanism.

In addition to these concerns, the fluence threshold for cracking in stainless is uncertain.

The GE update report states that the expected threshold is 2.0 x 10 21 n/cm2 for core internals.

The technical justification for this fluence threshold is not provided.

A recent paper [3] by GE indicates that the incubation fluence is 5 x 10 20 n/cm 2 for control blades fabricated from type 304 stainless steel.

The threshold fluence for the new cracking mechanism is not known at present.

Ferric Pressure Vessel steel Embrittlement More severe embrittlement has been opserved in several BWR pressure vessels than originally anticipated and in many vessels the nil-ductility transition temperature shift exceeds the 31

Regulatory Guide 1.99 (Rev. 2) model prediction by a statistically significant amount.

Currently available models which predict the trend in ferric steel embrittlement are inadequate when used outside a narrow flux, fluence, temperature, and chemistry window.

Since accurate trend curve models for BWRs are not available '

overly conservative assumptions may be necessary to ensure safe operation.

In the near term, this will most likely result in hydro-test problems.

In the long run, these conservative assumptions could adversely affect life extension plans.

With regards to safety, it is difficult to determine adequate margins in P-T curve calculations since it is difficult to identify the most severely embrittled region of the vessel (i.e. the worst combination of flu x, fluence, chemistry, temperature etc.).

Some specific technical considerations are summarized below:

1 A number of surveillance capsule specimens removed from operating BWR's show more embrittlement than would be expected from current theoretical models for neutron damage or from current damage correlations specified for use by the USNRC to predict embrittlement (RG 1.99, Rev.

2).

As shown in Table 1, 11 out of 32 (34%) vessels have measured nil-ductility transition temperature shifts

( tiT30S) which exceed RG 1. 99 (Rev. 2) by more than two standard deviations, and 19 out of 32 (59%) fall outside of the 95 percent confidence interval (+/- 2 standard deviations).

2 There appears to be an effect of neutron dose~ or flux on the embrittlement of PV steels [4,5,6].

Current models and correlations for damage depend on a data base largely developed from PWR experience where neutron flux is considerably higher than for BWR's.

In the case of Big Rock, for example, the same material was irradiated to 2.3 x 1019 n/cm 2 at two different fluxes (0.83 x 10 11 n/cm 2/sec and 9. 6 x 1011 n/cm 2/sec).

The lower flux irradiation yielded a much higher 6T30 (107 F for low flux, vs. 66 F for high flux).

These data along the reference [5] and

[6] data and the fact that damage rate effects have been demonstrated in the national breeder reactor and fusion 32

materials program strongl y suggest a damage rate effect in BWR pressure vessels.

3 Current RG 1.99 correlations between total neutron dose or fluence and resulting damage (6T30 ) consider only the chemistry of the base material (Cu and Ni) and product form (plate vs. weld).

Variables other than Cu, Ni, and fluence have been observed in test reactor irradiations (where the flux is substantially different from the PWR mean flux which dominates the surveillance data base) to contribute to pressure vessel steel embrittlement.

Since many BWR fluxes are one or more orders of magnitude below the flux range of the data used to develop RG 1.99 (Rev.

2) ( 3. o x 10 10

3. o x 10 11 n/cm 21sec), it is probable that subtle microstructural effects similar to those observed in test reactor irradiations are contributing to BWR vessel embrittlement.

4 The total fluence expected *tor a typical BWR is considerably lower than for PWRs (a factor of 10 lower).

However, correlations stipulated by the NRC for calculation temperature shifts for determining PWR and BWR operating limits are based largely on PWR data.

The BWR surveillance data base is sparse and was not well represented in RG 1.99 (Rev. 2).

There were only 19 data points out of a total of 216 with fluences typical for BWRs included in the RG 1.99 (Rev. 2) development.

5 If there are damage rate effects and/or other variables which should be included in a trend curve for BWRs, then any agreement with the RG 1.99 (Rev. 2) trend curve is only fortuitous.

This would, of course, mean that current pressure/temperature operating limits are arbitrary.

Summary e

In light of these concerns and observations of operating experience, it is possible that some of the Perry vessel internals may need to be replaced prior to the current design life.

Radiat~on damage to the pressure vessel may cause significant hydro-test problems early in life, and may necessitate annealing during life extension.

References 33

[l]

Manahan, M.P., Kohli, R., Santucci,J., and Sipush. P., "A Phenomenological Investigation of In-Reactor Cracking of Type 304 Stainless Steel Control Rod Cladding", Invited Paper, Journal of Nuclear Engineering and Design, Accepted for Publication January 1988.

[2]

Manahan, M.P., Kohli, R., Santucci, J., Sipush, o., and Harris, R.L., "Irradiation Assisted Cracking of -Control Rod Cladding", Invited Paper, Structural Mechanics in Reactor Technology, Conference Proceedings, Vol. C, pp. 75 -

85, SMIRT-9 Conference, Lausanne, Switzerland, August 17-21, 1987.

[3]

Sipush, P.J., Woodcock, J., Chickering, R.W., "Lifetime of PWR Silver-Indium-Cadmium Control Rods", EPRI NP-4512, March 1986.

[4]

Manahan, M.P., "Neutron Damage Rate Effects and Trenel Curve Development for BWR Pressure Vessel Stub", Proposal to the BWR Owners Group, July 1986.

[5]

Corwin, W.R., "Heavy Section Steel Technology Program Semiannual Progress Report for October 1987 -

March 1988",

NUREG/CR-42.

[6]

Hawthorne, J.R., Hiser, A.L., "Experimental Assessments of Gundremmingen RPV Archive Material for Fluence Rate Effects Studies", NUREG-CR-5201, September 1988.

34

Tab l e 1 Co rlson Between Measured and Predicled At Capsule Capsule Rog Guide 1.99 iHmeas - i\\Tcalc Vessel Cu NI Fl11ence 2 Fh11< 2 Measurod Rovlslon 2 Reactor Maloriat Wl.(.,.o)

WI.(%)

(1E19 n/cm )(1E11 111cm *S)

~ T :,o 6T:xi 6T:,o I Mar9ln Big Rock A302B 0.100 0.18u 0.1500 06100 00 20 3 56 6

-200 819 Rock A3028 0.100 0.180 2.3000 9 fiOOO 65.7 69 0 10'.l 0

- 0.19 Oig Rock A3028 0.100 0.180 10.7000 16 0000 152. 1 85.7 119.7 3.91 Big Rock AJ020 0.100 0.180 2.2700 0.8330 106.7 60.B 102.B 2.23 Big Rock AJ02B 0.100 0.180 0.7100 0.7500 49.1 509 84 9

- 011 Dresden-2 A302B 1.9000 1.6500 50.0 1705 204.5

-709 Oresrlen-2 AJ02B 0.0052 0 0027 23 0 100 200 2.60 Dresrlon-3 AJ02B 0.0029 0 0032 50 3.3 66 1.0J Dres<.lon-3 A302B 0.105 0.508 0.0071 0.0074 12 0 6.5 13 0 1.69 Halch-1 00240 47.0 10 0

m 0 3.22 Lacrosse A3020

• 0.070 0.210 0.4580 1.0200 87.5 33.7 67.4

3. 16 LaCrosse A3028 0.070 0.210 1.0200 1.0500 77.6 43.3 77.3 2.02 Millslone A5338

0. 140 0.610 0.3780 0.3990 67.J 73.3 107.J

-0.35 Millstone A5338 0.140 0.610 0.37B0 0.3990 94.0 73.3 107.3 1.22 Millstone A533B 0.0330 58.0 31.0 62.0 1.59 w

Monticello 0.170 42 o*

27.B 55.6 0.B4 Ul 0.654 00293 0 0121 Nine Mile Pofnl A302BM 0.230 0.460 0.0478 0 Ot90 114.0 41.9 75.9 4.24 Nine Mile Point A3020M 0.240 0.540 0.0360 00200 69 0 39.8 73 B 2 B!J Oyster Creek A302B 0.173 0.107 0 07'16 0 0290 720 28.6 57.2 3 03 Pilgrim 0.0230 25.0 19.0 38.0 0.63 Ouad Cilles-1 A302BM 0.210 0.560 0.8200 0 0024 125 0 14 t.5 175.5

-0.97 Quad Cilies-1 A3028M 0.210 0.560 4.0400 3.2500 166.9 203.S 237.5

- 2.16 Quad Cilies-1 A302BM 0.210 0.560 0.0010 00030 0.0 2.7 5.4

-2.00 Quad Cilies* 1 A302BM 0.210 0.560

1. 1900 3 0700 107.B 157.1 191.1

-2.90 Quad Cities-1 A302BM 0.182 0.480 0.0055 0.0026 40 8.8 17.6

-1.09 Quad Cilies-2 A302BM 0.100 0.540 0 0016 5.0 1.9 3.8 3.26 Quad Cities-2 A302BM

. 0.100 0.540 1.2700 40 0 69.3 103.3

- 1.72 Quad Cilies-2 AJ02BM 0.075 0.510 0.0066 0.0037 0.0 4.2 8.4

-2.00 Quad Cities-2 A302BM 0.100 0.540 4.1400 2.9100 51.8 88 6 122 6

- 2.1 fi Quad Cilies-2 A302BM 0.100 0.540 0.0020 0 0030 00

2. 1 4.2

- 2.00 Quad Cilies-2 A302BM 0.100 0.540 1.2700 2.4700 36.5 69.3 103.3

- 1.93 Vermonl Yankee A533B 0.110 0.6,80 0.0043 00018 19.0 4.5 9.0 6.44

• eased on an estimated unlrradlated T :,o *

• • All 6T tn Fahrenheit degrees.

O PS~G DOCKET NUMBER PROPOSED RULE PR 5 !)

(5~ Fe tt?3'/9&,J Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department

• 91 P.PR l5 P 4 :26

~ -

/~

Secretary, US Nuclear Docketing and Service Washington, DC 20555 Gentlemen:

l

~l NLR-N91060 Regulatory-Commission Branch COMMENTS ON PROPOSED CHANGE TO 10CFR50.55A INSERVICE TESTING OF CONTAINMENT ISOLATION VALVES SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Public Service Electric & Gas Company (PSE&G) hereby submits comments for the proposed rule change to 10CFR ~p.55a.

• ,. contains comments on paragraph b(2)vii -

Inservice Testing of Containment Isolation Valves and paragraph (6) (ii) (A) (3) Augmented Examination of Reactor Vessel.

Should there be any questions with regard to this submittal, please do not hesitate to contact us.

Attachment Sinc:;1ly, 0 fr1~;;v B. A. Preston Manager -

Licensing and Regulation APR 2 4 1991 Acknowledged by card.......... "............... "'-

The Energy People 95-2168 (87.SM) 12-90

U.S. NUCLEAR RE8ULATOAY COMM! 'iON OOCKETING & SERVICE SECTION OFFICE OF me SECRETARY OF THE COMMISSION Docunent Statfsticl Postmartt Date l/ I I 2. / q I Co~esRecelved_t _____ _

Adcfl Copies Reproduced...-::;_4..:,_.. __

Special Distribution W 1 g ID~. 4 rn; llm/Lf'l. ' fJampbdi

. ~~ ;~

NLR-N91060 ATTACHMENT 1 COMMENTS TO 10CFR50.55A PROPOSED RULE CHANGE Paragraph b(2) (vii) - Inservice Testing of Containment Isolation Valves The proposed rule, through the adoption of the 1989 Edition of ASME Section XI, adopts ASME/OM-10 for valve testing.

This specific paragraph takes exception to certain requirements of OM-10.

Specifically, containment isolation valves must be analyzed in accordance with paragraph 4.2.2.3(e) and corrective action taken in accordance with paragraph 4.2.2.3(f) of ASME/OM-10.

However, the ASME/OM-10 standard refers all requirements for containment isolation valves to 10CFR50 Appendix J.

Containment isolation valve leak rate testing is required by 10CFR50 Appendix J.

These requirements, implemented through technical specifications, require valve leakage to be evaluated against acceptance criteria.

Unacceptable leakage rates require valves to be repaired, leakage paths to be isolated, or operating units to be shutdown.

In addition, if unacceptable leakage rates are identified during outages, plants are prevented restart.

If the proposed rule is adopted as written, it would require utilities to consider Technical Specification actions, in addition to those identified in the preceding paragraph, which may may force plants to shut down.

If they evaluated leakage in accordance with Appendix J, they would not have to.

There have been reports published on excessive leakage through containment isolation valves and the effect on containment leakage integrity.

One example is a paper titled, "Primary Containment Leakage Integrity: Availability and Review Experience"; published in Nuclear Safety, Vol. 21-5, dated September-October 1980.

This report identifies many of the valve leakage failures to be repetitive in nature with these problems associated with large diameter MSIV's, purge or ventilation valves.

As a result of reports such as these, the generic problems with these valves are being addressed and individual valve leakage acceptance criteria for these problem valves have been identified in plant Technical Specifications.

The remaining problems with valve leakage are more than likely plant specific and are identified during Appendix J, Type C Testing.

The Appendix J program has proven to be effective in identifying containment isolation valve leak rate problems.

NLR-N91060 It is recommended that any revisions the NRC feels is required to containment isolation valve leak rate testing requirements be made in conjunction with Appendix J.

By duplicating valve leak rate acceptance criteria in the plants IST Program, plant operability will be challenged more often without a corresponding increase in safety and it will add confusion in Technical Specification implementation.

Paragraph (6) (ii) (A) (3) - Augmented Inspection of Reactor Vessel This paragraph allows plants with fewer than 40 months remaining in the Inservice Inspection interval to defer the augmented examination of the Reactor Vessel shell welds to the first period of the next interval.

The term augmented examination refers to the examination of 100% of the shell welds, whereas,Section XI requires only a partial examination of these welds.

This paragraph also prohibits the use of the deferred augmented examination as a substitute for the reactor vessel shell examination scheduled for implementation during the current inspection interval.

Prohibiting the use of the deferred augmented examinations as a substitute for the reactor shell examinations scheduled within the next 12 months would cause additional expenditure of radiation exposure, time, and money within a short period of time.

It would be too late to adequately plan to perform the augmented examinations during the next scheduled outage, even though the reactor vessel core barrel would be removed to complete the examination requirements of the current interval.

Since the core barrel must be removed to perform the augmented examinations required by this rule change, PSE&G would thereafter be forced to remove the core barrel again within the subsequent 40 month period.

It has always been recognized within the code that removing the core barrel is a massive undertaking and performed only once in 10 years.

It requires removing all fuel from the vessel, lifting the core barrel partly out of the reactor cavity water resulting in high exposure levels on the refueling floor, adds days to the duration of the outage, and significantly increases the cost of the outage.

Based on the above information, we recommend the proposed rule be revised to allow plants within the last 12 months of their current interval be allowed to substitute the deferred augmented reactor vessel shell examinations for the remaining reactor vessel shell examinations scheduled in the current interval.

p

~~~1:~~~ PR 0 0 ':'Northern sm*~~L~J:oQ

( 5{;. Fe 03 794>

USNRC 414 Nicollet Mall April 5, 1991 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention:

Docketing & Service Branch Minneapolis, Minnesota 55401 Telephon~,12) f\\1lR5~f p 5 :11 RE:

PROPOSED AMENDMENT TO 10CFR 50.55A, ENDORSING ASHE CODES The purpose of this letter is to submit comments on the Proposed Amendment to 10CFR 50.55a, Endorsing ASME Codes, as noted in Federal Register Volume 56, No. 21/Thursday, J anuary 31, 1991.

These comments address only the ten year weld inspection programs required by existing rule and proposed to b e enlarged by the proposed amendment to the rule.

These comments take exception to the information provided and conclusion reached in the statement of considerations concerning this aspect of the proposed amendment and suggest an alternative be considered which might enhance the benefit to assurance of safety and reduce the costs incurred to achieve that benefit.

The proposed rule would adopt by reference more recent editions of ASME,Section XI, which addresses requirements for volumetric inspection of pressure boundary welds and requirements for equipment testing.

Insofar as the proposed rule concerns weld inspections, the most significant result of the rule is that it would compel volumetric inspection of essentially 100% of the welds of all reactor vessels in every ten year inspection interval.

That portion of ASME Section XI Code was recently revised to codify this requirement at the specific request, if not demand, of the NRG Director o f NRR.

Such a requirement was not reasonable nor necessary when codified in ASME Section XI, and it is not reasonable nor necessary at this time to adopt i t by Rule.

The consideration for this aspect of the rule cites certain new information on irradiation damage t o reactor vessel materials, operational indications of stress corrosion cracking in BWRs and service induced cracks in pressurizers and steam generators.

Without any reasonable showing of how this information and experience has any relevance to the structural integrity of reactor shell plate welds, it makes the giant stride to the conclusion that all s uch welds in all reactor vessels should the r e fore b e Acknowledged by card

• • • • • • • • • ff.............. """

APR 2 4 1991

.. 1.,0Mi~ii$.31ON

.*.....,.r,iJ & SERVICE SECTION CFF JCE OF THE SECRET ARY

(,F THE COMMISSION Document Statistics Postmark Date _t/_,_/J~~-----

Coples Received. _ _.____-__ _

Add'! Copies Rej)rodl~ _.,l ::;__.c,__'I __

Special Distri ticn7 JJ 12. J "K-I D.5 ~ _

~}~

inspected in every ten year inspection interval.

In recognition of the industry resources which must be expended to comply with the additional requirements, a more rigorous and compelling case should be made by the NRG to justify this aspect of the proposed rule.

Industry experience with the weld inspection programs required by Section XI indicates that rarely, if ever, have these ten year ISI program inspections detected a service induced flaw until two events have occurred.

First, the service induced flaw is detected by visual inspection or leakage in one or more facilities.

Then inspection methods are refined to focus on that particular type of flaw in that particular zone of similar plants.

Thereafter, such focussed inspections have been reasonably effective, as in BWR pipe weld inspections for IGSCC flaws.

The weld inspection programs requil:ed by Section XI have become more effective with the evolution of improvements in ultrasonic inspection technology.

However, absent the refinement to focus on a particular type flaw in a particular zone, experience indicates these inspections merely become more adept at detecting and characterizing original construction defects which may or may not have been previously identified.

If the purpose of the inspection program is to identify service-induced flaws, then such inspections appear not to have been a very effective use of resources.

The operational experience and new information cited in the consideration of proposed rule makes neither a compelling nor a reasonable case for mandating essentially 100% of the shell welds of all reactor vessels to be inspected every ten year interval.

For example, the results of irradiation surveillance material tests do indeed show certain reactor vessels have been fabricated with materials that undergo greater irradiation damage than previously expected. Those results also show that other reactor vessels have been fabricated with materials that undergo lesser irradiation damage than previously expected.

There appears to be rather illusory logic in the conclusion that increased inspection of vessels undergoing lesser irradiation damage can mysteriously improve assurance of safety of vessels undergoing greater irradiation damage.

Even in those reactor vessels fabricated of materials which do undergo greater than expected irradiation damage, only some of the shell plate welds are exposed to significant radiation.

Similarly, there appears to be illusory logic in the reasoning that increased inspection of shell plate welds not exposed to significant irradiation will enhance assurance of safety of those welds in these vessels that are exposed to significant irradiation.

If assurance of safety in light of this new information is the concern, then it would be a more appropriate use of resources to focus inspection on reactor vessel shell welds in the "belt-line" zone where significant irradiation is anticipated to occur, for those reactor vessels fabricated of materials which undergo greater irradiation damage than previously expected.

The consideration of the rule also cites indications from operational data that show that stress corrosion cracking of BWR vessel welds is more probable than was thought several years ago.

There have indeed been operational indications of stress corrosion cracking in BWR reactor vessels, but no such indications in vessel shell plate welds. These indications have appeared only in safe end to vessel nozzle welds, and only those specific welds in which a particular inconel weld metal, welding process, and weld design was used.

The geometry of this zone involves a safe end attachment to a "pipe-like" extension of the reactor vessel nozzle.

From a pragmatic perspective, it is an artifact of codification that this zone is classified as a "vessel weld".

That is, the geometry, materials, welding process and weld residual stress environment of this zone is virtually unrelated to that of vessel shell plate welds.

Also the failure of these particular weld zones would have safety consequences within the envelope of design basis pipe break accidents and are not comparable to the failure of a reactor vessel.

Again, it escapes logic as to how more inspection of vessel shell plate welds in which there has been no indications of stress corrosion is going to enhance assurance of safety of these nozzle welds.

The Section XI Code already addresses these nozzle weld zones by requiring special inspection attention to bimetallic welds and piping terminal end welds.

It requires a giant leap to conclude that indications in these nozzle welds are suggestive that similar cracking in virtually unrelated shell plate welds "might be more likely than was thought several years ago".

The consideration of the proposed rule also cites service induced cracking in PWR pressurizer vessels and steam generators to support the rule requirement for greater inspection of reactor vessel shell plate welds.

This citation of service induced cracks in pressurizer vessels is again misappropriately extrapolated to reactor vessel shell plate welds.

These serviced induced cracks in pressurizers are associated with bimetallic seal welds of small bore heater penetration sleeves of some pressurizers, and are unrelated to vessel s*hell plate welds.

The service induced cracks in several PWR steam generators are on specific steam generator weld designs and service exposure not related to reactor shell plate welds.

The nuclear utility industry is now expending significant resources on ten year ISI programs required by ASME Section XI and adopted by Rule.

A substantial portion of these resources are expended performing inspections of weld zones in which service degradation is highly unlikely, and hence that expenditure of resources adds little value to assurance of safety.

The proposed rule would command further inappropriate expenditure of resources without commensurate benefit to assurance of safety. The NRC, having unduly influenced the ASME Section XI Code revision, can not fairly contend that the proposed rule merely adopts what the industry has codified.

This experience to date strongly suggest that ten year In Service Inspection Programs commanded by Code and Rule have taken a "scatter gun" approach that does not achieve a safety assurance benefit commensurate with the costs

incurred.

Weld zones least likely to incur in-service degradation are required to be inspected with equal rigor and cost as those weld zones deemed most likely to experience degradation and most likely to experience that degradation earlier in service life.

Enactment of the Rule as proposed would exacerbate this inappropriate expenditure of resources without commensurate benefit of safety assurance.

Experience to date suggests a greater benefit could be achieved in relation to cost if the required inspection programs were based on a reasonable technical judgment of risk and focused on weld zones deemed most likely to incur earliest in-service degradation, if it were to occur at all.

Sincerely, Gerald H Neils Executive Engineer Nuclear Generation File: GHN1441

• .:uµ'J io 01;i;y-Uriginal sent to 'll'tt1 tlffice of the

~.-.~~-

• DOCKET NUMBER PROPOSED RULE PR StJ (sG re 03 ?9,)-

NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-ADOS Codes and Standards for Nuclear Power Plants AGENCY:

Nuclear Regulatory Commission.

ACTION:

Proposed rule.

'91 JAN 29 A10 :28 c,FF11 *..:r. ;:-t,1,t,,,..,

[,OCKl. ! !N[,,*, '*[ "\\/l[;f 15R.I\\NCI:

SUMMARY

The Commission proposes to amend its regulations to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III, Division l, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section XI, Division 1, of the ASME Code, with a specified modification.

The proposed amendment would impose augmented examination of reactor vessel shell welds, and would separate in the regulations the requirements for inservice testing from those for inservice inspection by placing the requirements for inservice testing in a separate paragraph.

The ASME Code addenda and edition being incorporated by reference 1

provide updated rules for the construction of light-water-cooled nuclear power plant components, and for the inservice inspection and inservice testing of those components.

Adoption of this proposed amendment would permit the use of improved methods for construction, inservice inspection, and inservice testing of nuclear power plant components; would require expedited implementation of the expanded reactor vessel shell weld examinations specified in the 1989 Edition of Section XI; and would more clearly distinguish in the regulations the requirements for inservice testing from those for inservice inspection.

DATES:

Comment period expires (75 days after publication in the Federal Register).

Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before this date.

ADDRESSES:

Send comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

ATTN:

Docketing and Service Branch.

Deliver comments to:

11555 Rockville Pike, Rockville, Maryland, between 7:45 a.m. and 4:15 p.m. Monday through Friday.

Examine comments received at the NRC Public 9

Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT:

Mr. G. C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory*Commission, Washington, DC 20555, Telephone: (301) 492-3848.

SUPPLEMENTARY INFORMATION:

On May 5, 1988, the Nuclear Regulatory Commission published in the Federal Register (53 FR 16051) an amendment to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,

which incorporated by reference new addenda and a new edition to the ASME 2

Code.

This amendment revised§ 50.55a to incorporate by reference the Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 Addenda, and 1986 Edition for Division 1 rules of Section Ill, "Rules for the Construction of Nuclear Pciwer Plant Components," and the Wjnter 1983 Addenda, Summer 1984 Addenda, Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 Addenda, and 1986 Edition for Division 1 rules of Section XI, "Rules for the Inservice Inspection of Nuclear Power Plant Components," of the ASME Code.

The Commission proposes to amend§ 50.55a to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and i989 Edition of Sectio~ III, Division 1, of the ASME Code, and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section XI, Division 1, of the ASME Code, with a specified modificaton.

(In 1986, the ASME Code initiated a once-a-year addenda system and dropped the Summer/Winter designator). Also, the proposed amendment would impose augmented examination of reactor vessel shell welds, and would separate in the regulations the requirements for inservice testing from those for inservice inspection by placing the requirements for inservice testing in a separate paragraph.

Subsection IWP, Inservice Testing of Pumps," and Subsection IWV, "Inservice Testing of Valves," as contained in the 1988 Addenda and 1989 Edition of Section XI, incorporate by reference, respectively, Part 6, "Inservice Testing of Pumps in Light-Water Reactor Power Plants," and Part 10, "Inservice Testing of Valves in Light-Water Reactor Power Plants," of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987, "Operation and Maintenance of Nuclear Powef Plants." The contents of Subsections IWP and IWV in the 1988 Addenda and 1989 Edition are replaced in their entirety by the referenced*

rules of Part 6 and Part 10, respectively.

The NRC believes that certain 3

requirements in Part 10 represent unacceptable changes from present requirements in Subsection IWV of Section XI editions and addenda that have been incorporated by reference into§ 50.55a.

Therefore, the proposed amendment would incorporate by reference the 1988 Addenda and 1989 Edition of Section XI, Division 1, with a specified modification to Subsection IWV.

The NRC is particularly interested in receiving comments on the following discussed basis for and content of the proposed modification to Subsection IWV of the 1988 Addenda and 1989 Edition of Section XI, Division 1.

Paragraph IWV-3420 of Subsection IWV of Section XI editions and addenda presently incorporated by reference in§ 50.55a require all Category A valves, except those that function in the course of plant operation in a manner that demonstrates functionally adequate leak tightness, ~o undergo a valve leakage rate test. Subsection IWV paragraphs IWV-3426 and IWV-3427, respectively, require analysis of leakage rates and implementation of corrective actions dependent upon results of the leakage rate analysis.

Subsection IWV in the 1988 Addenda and 1989 Edition of Section XI, which reference Part 10 for the inservice testing of valves, provide rules for testing containment *isolation valves (CIVs) (i.e., paragraph 4.2.2.2 of Part 10 of the ASME/ANSI 0Ma~l988 Addenda). *These rules specify that Category A CIVs be tested in accordance with 10 CFR Part 50, Appendix J, and that CIVs which also provide a reactor coolant system pressure isolation function additionally be tested in accordance with Part 10, paragraph 4.2.2.3, "Leakage Rate for Other Than Containment Isolation Valves."

Paragraph 4.2.2.3(e) of Part 10 requires analysis of leakage rates and paragraph 4.2.2.3(f) of Part 10 specifies requirements for corrective action for Category A CIVs that also provide a reactor coolant system pressure isolation function.

4

Subsection IWV in the 1988 Addenda and 1989 Edition eliminate the present requirement to analyze leakage rates and to take corrective action in the event of abnormally high leakage_rates for those CIVs that do not provide a reactor coolant system pressure isolation function.

The NRC is concerned that this could significantly reduce the ability to detect degraded valves and, thereby, could permit an unacceptable reduction 1n the safety margin associated with the leak tight integrity of those CIVs that_ do not provide a reactor coolant system pressure isolation function.

The NRC's concern regarding the revision stems from the findings of two independent reviews of 9_

containment leakage rate failure experiences.

Both reviews conclude from analysis of Appendfx J leak test results, which included analysis of va1v~

leakage, that containment leakage during operation would exceed plant technical specification limits approximately 30 percent of the time.

This indicates a -need to improve, rather than relax, the present requirements concerning containment test, leak monitoring, and maintenance programs, including the ASME Section XI requirement for valve leak rate analysis.

It has yet to be demonstrated by analysis of more recent and comprehensive containment leakage test experiences that-containment leakage integrity can be improved to an acceptable level without implementation of a rigorous valve leak rate test program in conjunction with the present Section XI requirement for leak rate analysis. -

In proposing the following modification, the NRC specifically requests comments that would provide insight and justification, based upon plant experiences, relative to the need for revising or possibly eliminating the proposed modification.

The NRC proposes to incorporate by reference the 1988 Addenda and 1989 Edition of Section XI with a modification that would be specified in a new§ 50.55a(b)(2)(vii). The proposed modification would 5

substantially preserve the existing requirements for analysis of leakage rates and corrective actions that exist in Subsection IWV prior to the 1988 Addenda.

Specifically, the modification would require licensees to implement the requirements of paragraph 4.2.2.3(e), "Analysis of Leakage Rates," of Part 10 and paragraph 4.2.2.3(f), "Corrective Action," of Part 10, in addition to the requirements of paragraph 4.2.2.2 of Part 10, for all Category A valves that are CIVs, regardless of whether or not they provide a reactor coolant system pressure isolation function.

Section XI Subsection IWP and Subsection IWV editions and addenda, published up through the 1987 addenda, address Class 1, Class 2, and Class 3 pumps and valves, respectively, that perform specific safety functions.

The reference to Part 6 in Subsection IWP and to Part 10 in Subsection IWV in the 1988 Addenda and 1989 Edition expands the scope of these subsections to potentially include certain pumps and valves that are not classified as Class 1, Class 2, or Class 3.

Because§ 50.55a, at this time, only specifies requirements for pumps and valves that are designated Class 1, Class 2, or Class 3, this proposed amendment does not impose requirements on those pumps and valves that are not Class 1, Class 2, or Class 3, but would be included in the expanded scope of Subsection IWP and Subsection IWV in the 1988 Addenda and 1989 Edition.

However, Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Program," addresses this issue and notes in Position 11 that "The intent of 10 CFR 50 Appendix A, GDC-1, and Appendix B, Criterion XI, is that all components, such as pumps and valves, necessary for safe operation are to be tested to demonstrate that they will perform satisfactorily in service. Therefore, while 10 CFR 50.55a delineates the testing requirements for ASME Code Class 1, 2, and 3 pumps and valves, the testing of pumps and valves is not to be limited to only those covered by 6

10 CFR 50.55a."

The 1988 Addenda to Section XI modifies the 1986 Edition to require in the 2nd, 3rd, and 4th inspection intervals examination of essentially 100 percent of the length of all reactor vessel shell welds (i.e., Item Bl.IO, "Shell Welds," of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB, "Requirements for Class I components of Light-Water Cooled Power Plants"). Since the 1989 Edition is identical to the 1986 Edition as modified by the 1986 Addenda, 1987 Addenda, and 1988 Addenda, this revision also appears in the 1989 Edition of Section XI.

The 1986 Edition of Section XI (the most current Section XI rules presently incorporated by reference into§ 50.55a) requires examination of only one longitudinal weld and one circumferential weld from the beltline region during the 2nd, 3rd, and 4th inspection intervals. The requirement to examine essentially 100 percent of the length of all reactor vessel shell welds during the 1st inspection interval has been in Section XI since the 1975 Winter Addenda to the 1974 Edition.

Recent information from reactor vessel material surveillance programs, and observed flaws in certain operating reactor and steam generator vessels, reveal the potential susceptibility of reactor vessel materials to degradation.

Because of these experiences and the limited examinations performed to date on some reactor vessels, the NRC is concerned with the length of time that might elapse before a licensee would be required to implement the reactor vessel shell weld examinations specified in the 1988 Addenda and the 1989 Edition of Section XI through routine updating of its inservice inspection program.

Section 50.55a(g)(4)(ii) requires that inservice inspection programs be updated to reflect the latest edition and 7

addenda of Section XI identified in§ 50.55a(b)(2) 12 months prior to the start of the next 120-month inspection interval. Routine updating in accordance with this requirement could result in the 1989 Edition not being implemented for as long as 240 months (20 years).

For example, a plant just entering the first period in the 2nd, 3rd, or 4th inspection interval when this rule becomes effective would not have to implement the reactor vessel examinations specified in the 1989 Edition for 20 years, because that inspection interval would be covered by a previousSection XI edition/addenda and because under existing Section XI rules, the reactor vessel examinations in the succeeding interval, which would implement the 1989 Edition or later, could be deferred another 10 years until the end of that interval. Similarly, a plant just entering the second or third period in the 2nd, 3rd or 4th inspection interval would not be required to implement the 1989 Edition, or subsequent addenda, for 200 months (16 years, 8 months) or 160 months (13 years, 4 months), respectively.

Consistent with the existing updating requirements of§ 50.55a(g)(4)(ii) and the changing requirements of Section XI, some inservice inspection programs based on certain editions and addenda of Section XI may have resulted in very limited reactor vessel examinations.

For example, if examinations of the beltline welds during the 1st inspection interval were performed to comply with the 1974 edition of Section XI, 5 percent of the circumferential welds and 10 percent of the longitudinal welds would have been examined.

If, for the same plant, examinations during the 2nd inspection interval were performed to comply with the 1980 Edition, including subsequent*addenda, one circumferential weld and one longitudinal weld would have been required to be examined.

[The 1974 Edition of Section XI (with addenda through the 1975 Winter Addenda) through the 1986 Edition (with addenda through the 1987 8

Addenda) require that all reactor vessel shell welds be examined volu-metrically during the 1st inspection interval, and that one circumferential and one longitudinal beltline weld be examined volumetrically in succeeding inspection intervals; whereas the 1971 Edition through the 1974 Edition* (with addenda through the 1975 Summer Addenda) require that 10 percent of the length of each longitudinal weld and 5 percent of the length of each circumferential weld be examined volumetrically each inspection interval.]

Degradation of reactor vessel materials.has become more of a concern recently, because (1) results from irradiation surveillance material tests show that certain reactor vessel materials undergo greater radiation damage than previously expected, (2) indications from operational data show that stress corrosion cracking of BWR reactor vessels is more probable than was thought several years ago, and (3) significant service induced cracking has occurred in large vessels (i.e., pressurizer, steam generators) designed and fabricated to the ASME Code.

The NRC is concerned that the inherent delay in implementing the expanded reactor vessel examinations is inconsistent with the importance of the reactor vessel, with recent new information regarding degradation of reactor vessel materials, with the limited examination of shell welds previously performed on many reactor vessels, and with the need to ensure that the failure probability of the reactor vessel remains extremely low.

It is the judgment of the NRC that, because of new information and limited previous reactor vessel examinations, there may exist a substantially greater potential for reactor vessel degradation than previously considered and that maintenance of the level of protection presumed by the regulations requires more than compliance to existing regulatory requirements.

9

The NRC has determined that the proposed augmented reactor vessel examination would result in a substantial increase in the overall protection of the public health and safety, and that the costs of implementation would be justified in view of the increased protection.

The backfit analysis required by§ 50.109, "Backfitting," is provided as part of the regulatory analysis that supports this proposed rule.

Section 50.55a(g)(6)(ii) addresses augmented inservice inspection programs for those systems and components for which the Commission deems that added assurance of structural reliability is necessary.

For that purpose, and consistent with the above discussion, it is proposed that

§ 50.55a(g)(6)(ii)(A) be added to require expedited implementation of the reactor vessel shell weld examinations specified in the 1989 Edition of Section XI, Division 1, in Item Bl.IO, 11Shell Welds," of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," in Table 2500-1 of Subsection IWB, "Requirements for Class 1 Components of Light-Water Cooled Power Plants."

Proposed§ 50.55a(g)(6)(ii)(A) was developed with two primary considerations in mind.

First, the proposed rule must require implementation of the provisions for reactor vessel shell weld examinations provided in the 1989 Edition as quickly as practicable. Second, to minimize unnecessary impact on licensees, the implementation requirements for the augmented examination should be integrated as closely as possible with existing examination requirements and practices.

In order to ensure the applicability of the proposed augmented examination to all licensees, § 50.55a(g)(6)(ii)(A)ill would revoke all previously granted reliefs to licensees for reactor vessel shell weld examinations for the inservice inspection interval that would be in effect 10

when the rule becomes effective.

This is consistent with the ongoing development schedule for equipment and techniques that would permit those licensees with limited accessibility to implement the proposed augmented examination.

The NRC has structured the proposed requirement for augmented examination of reactor vessel shell welds recognizing that plants will be on different schedules for their 120-month inservice inspection interval.

Section 50.55a(g)(6)(ii)(A)ill would require all licensees to implement the specified augmented reactor vessel examination during the inspection interval in force when this proposed rule becomes effective, subject to conditions specified in proposed§ 50.55a(g)(6)(ii)(A)ill and ill-Section 50.55a(g)(6)(ii)(A)ill would specifically permit the use of the augmented examination as a substitute for the reactor vessel shell weld examinations scheduled for the inspection interval in effect when this proposed rule becomes effective.

The NRC recognizes that plants with fewer than 40 months remaining in the inspection interval when this proposed rule becomes effective may find it impractical to implement the augmented reactor vessel examination during that inspection interval. Therefore, proposed§ 50.55a(g)(6)(ii)(A)(J) would permit plants with fewer than 40 months remaining in the inspection interval when this rule becomes effective to defer the augmented examination until the first period of the next inspection interval. However, this same paragraph would specifically prohibit the use of the deferred augmented examination as a substitute for reactor vessel shell weld examinations scheduled for the inspection interval in effect when the rule becomes effective.

The intent is to ensure that the examinations are deferred only when necessary and not to have the proposed rule encourage a 40-month delay in reactor vessel shell weld examinations.

11

Section 50.55a(g)(6)(ii)(A)(J) would permit using the deferred examination, with a condition, as a substitute for reactor vessel shell weld examinations scheduled for the inspection interval in which the deferred examinations are performed.

The condition is that subsequent reactor vessel shell weld examinations for successive inspection intervals be performed in the first period of the inspection interval. This condition is necessary to prevent a potential 160-month gap between reactor vessel shell weld examinations. This gap would occur if a pl~nt used the deferred examination performed in the first period as a substitute for the scheduled examination and then deferred the examination for the next inspection interval to the end of that interval as permitted by Section XI.

Proposed§ 50.55a(g)(6)(ii)(A)ill specifies that a licensee that has either completed or has scheduled an inspection of essentially 100 percent of the length of all Examination Category B-A shell welds during the inservice inspection interval in effect when the proposed rule becomes effective does not have to implement the proposed requirement for augmented examination of the reactor vessel shell welds.

Primarily, this proposed paragraph is intended to permit licensees who would be in the 1st inspection interval to use the essentially 100 percent reactor vessel shell weld examination required for that interval by Section XI to satisfy the requirement for the proposed atigmented reactor vessel examination.

The technical objective of the augmented examination would have been accomplished under such conditions.

These licensees would continue to apply the current requirements of

§ 50.55a(g)(4) until the next inspection interval when future examinations would be performed based on ASME Section XI, 1989 Edition, or later Code edition and addenda specified in§ 50.SSa(b).

12

The proposed amendment to§ 50.55a would separate the requirements for inservice testing from those for inservice inspection by moving the requirements for inservice testing to a separate paragraph.

Presently,

§ 50.55a(g), "Inservice inspection requirements," specifies the requirements for (1) preservice and inservice examinations for Class 1, Class 2, and Class 3 components and their supports, (2) system pressure tests for Class 1, Class 2, and Class 3 components, and (3) inservice testing of Class 1, Class 2, and Class 3 pumps and valves.

In order to emphasize the importance of inservice testing and to more clearly distinguish its requirements from, those of inservice inspection, the proposed rule would move the present requirement for inservice testing from existing§ 50.55a(g), "lnservice inspection requirements," to a separate.(presently reserved) § 50.55a(f), which would be titled "lnservice testing requirements." All existing requirements for inservice examination and system pressure testing would be retained in

§ 50.55a(g).

Two editorial revisions, relative to existing§ 50.55a(g), are included in the proposed new§ 50.55a(f). These editorial revisions (1) reserve

§ 50.55a(f)(3)(i) and (ii) so that the structure of§ 50.55a(f) would parallel that of§ 50.55a(g) for the purpos*e of promoting easier cross-referencing between the two paragraphs, (2) modify reference to 120-month inspection interval in§ 50.55a(g) to 120-month interval in proposed§ 50.55a(f}, because inspection interval, as used in Section XI, is used only in the context of inservice inspection.

(The term "test interval" was not used because, unlike inspection interval, the 120-month time frame does not designate a period of required actions for the testing program.

The 120-month interval used in

§ 50.55a(f) and the 120-month inspection interval used in§ 50.55a(g) are considered by the staff to be coincident for the purpose of 120-month updating 13

requirements.)

In addition, two administrative changes have been made in the development of proposed§ 50.55a(f) relative to existing§ 50.55a(g).

First,

§ 50.55a(f)(6)(ii) has been added to indicate intent by the Commission to impose an augmented inservice testing program if added assurance of operational readiness is deemed necessary.

This proposed paragraph only indicates intent and does not impose a specific requirement.

It does parallel the existing§ 50.55a(g)(6)(ii) which specifies that the CoITH11ission may require an augmented inservice inspection program for systems and components for which it deems that added assurance of structural reliability is necessary.

Second, the proposed amendment includes the addition of introductory text to§ 50.55a(g) which states that the requirements for inservice testing of Class 1, Class 2~ and Class 3 pumps and valves are located in§ 50.55a(f).

This change is necessary because the proposed placement of inservice testing requirements into a separate§ 50.55a(-f) would cause administrative inconsistencies with regard to existing references to§ 50.55a(g) for inservice testing in documents such as technical specifications, safety analysis reports, procedures, and records.

With the proposed change, existing references to§ 50.55a(g) for inservice testing would refer the user to

§ 50.55a(f) where the specific requirements for inservice testing would be located.

The NRC recommen~s that as the governing documents are updated, the direct reference to§ 50.55a(f) be incorporated, as appropriate.

Section 50.55a(g) provides requirements for selecting the ASME Code edition and addenda of Section XI to be complied with during the preservice 14

inspection (§ 50.55a(g)(3), for plants whose construction permit was issued on or after July 1, 1974); the initial 10-year inspection interval

(§ 50.55a(g)(4)(i)); and successive 10-year inspection intervals

(§ 50.55a(g)(4)(ii)).

As noted in the Supplementary Information to the final rule of the most recent amendment to§ 50.55a (May 5, 1988; 53 FR 16051),

paragraph IWA-2400 of Section XI (as revised by the Winter 1983 Addenda) incorporated rules for selecting the applicable edition and addenda of Section XI during the preservice inspection (IWA-2411), the initial 10-year inspection interval (IWA-2412), and successive 10-year inspection intervals (IWA-2413).

The criteria provided in the regulations and Section XI are effectively the same for the preservice inspection and the successive 10-year inspection intervals, but differ for the initial 10-year inspection interval.

In general, use of the Commission requirements will result in the selection of a more recent edition and addenda than will use of the Section XI rules.

Satis-fying the requirements of§ 50.55a(g)(4)(i) for the inital 10-year inspection interval will, in general, also satisfy the rules of Section XI.

Although the Section XI requirements for selecting editions and addenda remain unchanged in the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition, the Commission is reaffirming its intent that in all cases the existing requirements in

§ 50.55a(g) be the basis for selecting the edition and addenda of Section XI to be complied with during the preservice inspection, the initial 10-year inspection interval, and the successive 10-year inspection intervals.

The proposed amendment would make a number of editorial changes to

§ 50.55a for the purpose of adopting a standard convention for imposing an obligation or expressing a prohibition.

In this convention "shall" is used to impose an obligation on an individual or legal entity capable of performing the required action, "must" is used as the mandatory form when the subject of 15

the sentence is an inanimate object, and "may not" is used to impose a prohibition.

The following paragraphs were amended solely to be consistent with this convention: the introductory paragraph to the section; paragraphs (a)(l), (a)(3), (b)(2)(iii), (b)(2)(iv), (g)(l); (g)(3)(ii), (g)(3)(iii),

(g)(3)(iv), introductory paragraph to (g)(4), (g)(4)(i), (g)(4)(ii),

(g)(5)(i), (g)(5)(iv), (g)(6)(i), (h), and footnote 8.

Other paragraphs were revised for the same editorial reason, but they also contain technical revisions relevant to other parts of this proposed amendment.

Section 50.55a(f) has been developed consistent with the noted convention.

Subsection IWE, "Requirements for Class MC Components of Light-Water-Cooled Power Plants," was added to Section XI, Division 1, in the Winter 1981 Addenda.

However, 10 CFR 50.55a presently incorporatesSection XI inservice inspection requirements for only Class 1, Class 2, and Class 3 components and their supports.

The regulation does not currently address the inservice inspection of containments.

Because this amendment is only intended to update current regulatory requirements to include the latest ASME Code edition and addenda, the requirements of Subsection IWE would not be imposed upon Commission licensees by this amendment.

The incorporation by reference of Subsection IWE into§ 50.55a is presently the subject of a separate rulemaking action. Section 50.55a(b)(2)(vi) is reserved for that action.

The NRC previously alerted all holders of operating licenses or construction permits for nuclear power reactors, through NRC Information Notice No. 88-95 (IN 88-95), "Inadequate Procurement Requirements Imposed by Licensees on Vendors," to the potential that inadequate licensee procurement requirements or implementation by vendors in supplying components under the ASME Code could result in failure by these vendors to fully implement 16

10 CFR Part 50, Appendix B (Quality Assurance Criteria). The problem, which was revealed during routine NRC inspections of vendors, resulted from the belief by some vendors that if an item was exempted by the ASME Code from Code requirements, the item was exempt from all other regulatory requirements.

The apparent belief of some vendors was that *since NRC endorses the ASME Code in its regulations and has accepted the various exemptions, there are, therefore, no other applicable regulatory requirements.

This belief is not consistent with the NRC position.

The NRC reaffirms its position which, as previously put forth in IN 88-95, states that-all safety-related items, even those exempted from ASM~ Code requirements, are required to be manufactured under a quality assurance program that meets 10 CFR Part 50, Appendix B requirements.

Finding of No Significant Environmental lmpact: Availability 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, if adopted would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.

The proposed rule is one part of a regulatory framework directed to ensuring pressure vessel integrity, and the operational readiness of pumps and valves. Therefore, in the general sense, the proposed rule would have a positive impact on the environment.

The proposed rule would incorporate by reference into the NRC regulations improved rules contained in the ASME Code for the construction, inservtce inspection and inservice testing of components used in nuclear power plants.

In addition, the proposed rule would require an 17

augmented examination of reactor vessel shell welds to further ensure the structural integrity of the reactor vessel. Actions required of applicants and licensees to implement the proposed rule are of a routine nature that should not increase the potential for a negative environmental impact.

The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW (Lower Level), Washington, DC.

Single copies of the environmental assessment and the finding of no significant impact are available from Gilbert C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-3848.

Paperwork Reduction Act Statement This proposed rule would amend information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

This proposed rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

Public reporting burden for this collection of information is estimated to average 135 hours0.00156 days <br />0.0375 hours <br />2.232143e-4 weeks <br />5.13675e-5 months <br /> per response, including the time for reviewing instructi-ons, searching existing data sources, gathering and maintaining the

~

data needed, and completing and reviewing the collection of information.

Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory 18

Commission, Washington, DC 20555; and to the Paperwork Reduction Project (3150-0011), Office of Management and Budget, Washington, DC 20503.

Regulatory Analysis The Commission has prepared a regulatory analysis for this proposed amendment to the regulations.

The analysis examines the costs and benefits of the alternatives considered by the Commission.

Interested persons may examine a copy of the regulatory analysis at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

Single copies of the analysis may be obtained from Mr. G. C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 492-3848.

Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.

605(b), the Commission hereby certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and 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 set out in regulations issued by the Small Business Administration at 13 CFR Part 121.

Since these companies are dominant in their service areas, this proposed rule does not fall within the purview of the Act.

19

Backfit Analysis The NRC has concluded, based on the analysis required by§ 50.109(a}(3}

which is provided in the regulatory analysis, that the backfit that would be imposed by the proposed augmented reactor vessel examination would result in a substantial increase in the overall protection of the public health and safety, and that the direct and indirect costs of implementation would be justified in view of the increased protection.

The incorporation by reference into the regulations of later editions and addenda of Section III and Section XI of the ASME Code is not a backfit because Section III requirements apply only to new construction, except as voluntarily implemented by licensees, and-because updated Section XI requirements are an integral part of the longstanding§ 50.55a(g}(4}(ii}

requirement to update inservice inspection and inservice testing programs to the requirements of the latest edition and addenda of Section XI incorporated by reference in§ 50.55a(b} 12 months prior to the start of the 120-month inspection interval, subject to specified limitations and modifications.. The proposed modification to Part 10 of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987 is not a backfit because it simply retains a requirement that licensees now are required to implement in accordance with§ 50.55a(g).

List of Subjects In 10 CFR Part 50 Antitrust, Classified information, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and 20

recordkeeping requirements.

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments 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.~.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. 5851).

Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 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, 50.54(dd) and 50.103 also issued under Sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

Sections 50.~3, 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).

21

For the purposes of sec. 223, 68 Stat. 958, as amemded (42 U.S.C. 2273);

§§ 50.46(a) and (b), and 5O.54(c) are issued under sec. 161b, 16li and 1610, 68 Stat. 948 as amended (42 U.S.C. 220l(b); §§ 50.7(a), 50.l0(a)-(c), 50.34(a) and (e), 50.44(a)-(c), 50.46(a) and (b), 50.47(b), 50.48(a), (c), (d), and (e), 50.49(a), 50.54(a), (i), (i)(l), (l)-(n), (p), (q), (t), (v), and (y),

50.55(f), 50.55a(a), (c)-(e), (g), and (h), 50.59(c), 50.60(a), 50.62~c),

50.64(b), and 50.80(a) and (b) are issued under sec. 16li, 68 Stat. 949, as amended (42 U.S.C. 220l(i)); and§§ 50.49(d), (h), and (j), 50.54(w), (z),

(bb), {cc), and {dd), 50.55(e), 50.59(b), 50.6l(b), 50.62{b), 50.70{a),

50.7l(a)-(c) and (e), 50.72(a), 50.73(a) and (b), 50.74, 50.78, and 50.90 are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(0)).

2.

In§ 50.55a, the introductory text, paragraphs (a), {b)(l), the introductory text of (b)(2), (b)(2)(iii), (b)(2)(iv), (g)(l),

(g){2), (g)(3)(i), (g)(3)(ii), (g)(4), (g)(5)(i), (g)(5)(iv), (h),

and footnote 8 are revised; paragraphs (g)(3)(iii) and (g)(3)(iv) are removed and reserved; paragraph (b)(2)(vi) is added and reserved; and paragraphs (b)(2)(vii), (f), introductory text to (~),

and (g)(6)(ii)(A) are added to read as follows:

§ 50.55a Codes and standards.

Each operating license for a boiling or pressurized water-cooled nuclear power facility must be subject to the conditions in paragraphs (f) and (g) of this section and each construction permit for a ~tilization facility must be subject to the following conditions in addition to those specified in§ 50.55.

22

(a)(l) Structures, systems, and components must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.

(2)

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code specified in paragraphs (~), (c), (d), (e),

(f), and (g) of this section. Protection systems of nuclear power reactors of all types must meet the requirements specified in paragraph (h) of this section.

(3)

Proposed alternatives to the requirements of paragraphs (c), (d),

(e), (f}, (g), and (h) of thii section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation.

The applicant shall demonstrate that (i) the proposed alternativ~s would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

(b) * * *

(1)

As used in this section, references to Section III of the ASME Boiler and Pressure Vessel Code refer to Section III, Division 1, and include addenda through the 1988 Addenda and editions through the 1989 Edition.

23

(2)

As used in this section, references to Section XI of the ASME Boiler and Pressure Vessel Code refer to Section XI, Division 1, and include addenda through the 1988 Addenda and editions through the 1989 Edition, subject to the following 1imitations and modifications:

(iii) Steam generator tubing (modifies Article IWB-2000).

If the technical specifications of a nuclear power plant include surveillance requirements for steam generators different than those in Article IWB-2000, the inservice inspection program for steam generator tubing must be governed by the requirements in the technical specifications.

(iv) Pressure-retaining welds in ASME Code Class 2 piping (applies to Tables IWC-2520 or IWC-2520-1, Category C-F). (A) Appropriate Code Class 2 pipe welds in Residual Heat Removal Systems, Emergency Core Cooling Systems, and Containment Heat Removal Systems, must be examined.

When applying editions and addenda up to the 1983 Edition through the Summer 1983 Addenda of Section XI of the ASME Code, the extent of examination for these systems must be determined by the requirements of paragraph IWC-1220, Table IWC-2520 Category C-F and C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the Summer 1975 Addenda.

24

(vi)

[Reserved]

(vii)

Inseryjce testing of containment isolation valves.

When using Subsection IWV in the 1988 Addenda or the 1989 Edition of Section XI, Division 1, of the ASME Boiler and'Pressure Vessel Code, leakage rates for Category A containment isolation valves that do not provide a reactor coolant system pressure isolation function must be analyzed in accordance with paragraph 4.2.2.3(e) of Part 10, and corrective actions for these valves must be made in accordance with paragraph 4.2.2.3(f) of Part 10 of ASME/ANSI OMa-1988 Addenda to ASME/ANSI OM-1987.

(f)

Inservice testing requirements.

(1)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, pumps and valves must meet the test requirement of paragraphs (f)(4) and (5) of this section to the extent practical.

Pumps and valves which are part of the reactor coolant pressure boundary must meet the requirements applicable to components which are classified as ASME -

Code Class 1. Other safety-related pumps and valves must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

(2)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, 25

but before July 1, 1974, pumps and valves which are classified as ASME Code Class 1 and Class 2 must be designed and be provided with access to enable the performance of inservice tests for operational readiness set forth in editions of Section XI of the ASME Boiler and Pressure Velsel Code and Addenda 6 in effect 6 months prior to the date of issuance of the construction permit.

The pumps and valves may meet the inservice test requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed therein.

(3)

For~ boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or*after July 1, 1974:

(i) [Reserved]

(ii) [Reserved]

(iii) Pumps and valves which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6 applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.

26

(iv)

Pumps and valves which are classified as ASME Code Class 2 and Class 3 must be designed and be provided*with access to enable the performance of inservice testing of the pumps and I

valves for assessing operational readiness set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6 applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.

(v)

All pumps and valves may meet the test requirements set forth in subsequent editions of codes and addenda or portions thereof which are incorporated by_ reference in paragraph (b) of this section, subject to the limitations and modifications listed therein.

(4)

Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements, except design and access provisions, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda that become effective subsequent to editions specified i'n paragraphs (f)(2) and (f)(3) of this section and that are incorporated by reference in paragraph (b) of this section, to the extent practical within the limitations of design, geometry.and materials of construction of such components.

(i)

Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during the initial 120-month interval must comply with the 27

requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section on the date 12 months prior to the date of issuance of the operating license, subject to the limitations and modifications listed in paragraph (b) of this section.

(ii) Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed in paragraph (b) of this section.

(iii) [Reserved]

(iv) Inservice tests of pumps and valves may meet the requirements set forth in subsequent editions and addenda that are incor-porated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section, and subject to Commission approval.

Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.

(5) (i)

The inservice test program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph 28

(f)(4) of this section.

(ii) If a revised inservice test program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program.

The licensee shall submit this application, as specified in§ 50.4, at least 6 months before the start of the period during which the provisions become applicable, as determined by paragraph (f)(4) of thi~ section.

(iii) If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Comm.ission and submit, as specified in-§ 50.4, information to* s_upport the detenni nation.

(iv)

Where a pump or valve test requirement by the code or addenda

)

is determined to be impractical by the licensee and is not included in the revised inservice test program as permitted by paragraph (f)(4) of this section, the basis for this determination must be demonstrated to the satisfaction of the Commission not later than 12 months after the expiration of the initial 120-month period of operation from start of facility commercial operation and each s~bsequent 120-month period of operation during which the test is determined to be impractical.

29

(6) (i)

The Commission will evaluate determinations under paragraph (f)(5) of this section that code requirements are impractical.

The Commission may grant relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

(ii) The Commission may require the licensee to follow an augmented inservice test program for pumps and valves for which the Commission deems that added assurance of operational readiness is necessary.

(g)

Inservice inspection requirements.

Requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in

§ 50.SSa(f).

(1) For a boiling or pressuriJed water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4} and (5) of this section to the extent practical.

Components which are part of the reactor coolant pressure boundary and their supports must meet the requirements applicable to components which are classified as ASME Code Class 1.

Other safety-related pressure vessels, piping, pumps and valves must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

30

(2)

For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, components (including supports) which are classified as ASME Code Class 1 and Class 2 must be desig~ed and be provided with access to enable the performance of inservice examination of such components (including supports) and must meet the preservice examination requirements set forth in editions of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda 6

in effect six months prior to the date of issuance of the construction permit.

The components (including supports) may meet the requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to the limitation and modifications listed therein.

(3)

For a boiling or pressurized water-cooled nuclear power facility whose construction pe~mit was issued on or after July 1, 1974:

(i) Components which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of such components and must meet the preservice examination requirements set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6

applied to the construction of the particular component.

(ii) Components which are classified as ASME Code Class 2 and Class 3 and supports for components which are classified as ASME Code Class 1, Class 2, and Class 3 must be designed and be provided with access to enable the performance of inservice examination of such components 31

and must meet the preservice examination requirements set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda 6 applied to the construction of the particular component.

(iii) [Reserved]

(iv) [Reserved]

(4)

Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the requirements, except design and access provisions and preservice examination*requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Cod~ and Addenda that become effective subsequent to editions specified in paragraph~ (g)(2) and (g)(3) of this section and that are incorporated by reference in paragraph (b) of this section, to the extent practical within the limitations of design, geometry and materials of construction of the components.

(i)

Inservice examinations of components and system pressure tests conducted during the initial 120-month inspection interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section on the date 12 months prior to the date of issuance of the 32

operating license, subject to the limitations and modifications listed in paragraph (b) of this section.

(ii)

Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed in paragraph (b) of this section.

(iii) [Reserved]

(iv)

Inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section, and subject to Commission approval.

Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.

(5)

(i) The inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section.

33

(iv)

Where an examination requirement by the code or addenda is determined to be impractical by the licensee and is not included in the revised inservice inspection program as permitted by paragraph (g)(4) of this section, the basis for this determination must be demonstrated to the satisfaction of the Convnission not later than 12 months after the expiration of the initial 120-month period of operation from start of facility commercial operation and each subsequent 120-month period of operation during which the examination is determined to be impractical.

(6) * * *

(ii) * * *

(A)

Augmented examination of reactor vessel ill All previously granted reliefs under§ 50.SSa to licensees for the examination of reactor vessel shell welds specified in Item Bl.IO of Examination Category B-A, "Pressure Retaining Welds in Reactor

Vessel, 11 in Table.IWB-2500-1 of Subsection IWB in applicable edition and addenda of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, during the inservice inspection interval in effect on ____ (effective date.of rule will be ins*erted) are hereby revoked.

ill All licensees shall augment their reactor vessel examination by implementing once, as part of the inservice inspection interval in effect on ____ (effective date of rule will be inserted), the examination requirements for reactor vessel 34

shell welds specified in Item B1.10 of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in§ 50.55a(g)(6)(ii)(A)ill and ill, The augmented examination may be used as a substitute for the reactor vessel shell weld ~xamination scheduled for implementation during the inservice inspection interval in effect on

---

(effective date of rule will be inserted).

ill Licensees with fewer than 40 months remaining in the inservice inspection interval in effect on ____ (effective date of rule will be inserted) may defer the augmented reactor vessel examination specified in§ 50.55a(g)(6)(ii)(A)ill to the first period of the next inspection interval.

The deferred augmented examination may not be used as a substitute for the reactor vessel shell weld examination scheduled for implementation during the i nservi ce *inspection interval in effect on ____ (effective date of rule will be inserted).

The deferred augmented examination may be used as a substitute for the reactor vessel shell weld examination normally scheduled for the inspection interval in which the deferred examination is performed.

If the deferred augmented examination is used as a substitute for the normally scheduled reactor vessel shell weld examination, subsequent reactor vessel shell weld examinations must be performed during the first period of successive inspection intervals.

35

ill The requirement for augmented examination of the reactor vessel may be satisfied by an examination of essentially 100 percent of the reactor vessel shell welds specified in§ 50.55a(g)(6)(ii)(A)ill that has been completed, or is scheduled for implementation with a written commitment, or is required by§ 50.55a(g)(4)(i), during the inservice inspection interval in effect on ---- (effective date of rule will be inserted).

(h)

Protection systems.

For construction permits issued after 6

January 1, 1971, protection systems must meet the requirements set forth in editions or revisions of the Institute of Electrical and Electronics Engineers Standard: "Criteria for Protection Systems for Nuclear Power Generating Stations," (IEEE-279) in effect 7 on the formal docket date 8 of the application for a construction permit.

Protection systems may meet the requirements set forth in subsequent editions or revisions of IEEE-279 which become effective.

ASME Code cases that have been determined suitable for use by the Commission staff are listed in NRC Regulatory Guide 1.84, "Design and Code Case Acceptability ASME Section III Division l," NRC Regulatory Guide 1.85, "Materials Code Case Acceptability -- ASME Section III Division l," and NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptabilty -- ASME Section XI Division l."

The use of other Code cases may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to

§ 50.55a(a)(3).

36

B For purposes of this regulation the proposed IEEE 279 became "in effect" on August 30, 1968, and the revised issue IEEE 279--1971 became "in effect" on June 3, 1971.

Copies may be obtained from the Institute of Electrical and Electronics Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017.

Copies are available for inspection at the Commission's Technical Library, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland.

Where an application for a construction permit is submitted in four parts pursuant to the provisions of§ 2.lOl(a-1) and Subpart F of Part 2 of this chapter, "the formal docket date of the application for a construction permit" for purposes of this section must be the date of docketing of the information required by§ 2.lOl(a-1) (2) or (3), whichever is later.

Dated at Rockville, Maryland this /~day ~7'° 1991.

For the Nuclear Regulatory Commission.

Executive Director for Operations.

37