ML090350419

Proposed Staff Guidance on the Position of the Gall Report Presenting One Acceptable Way to Manage Aging Effects for License Renewal

ML090350419
Person / Time
Site:	PROJ0690
Issue date:	11/23/2001
From:	Charemagne Grimes Office of Nuclear Reactor Regulation
To:	Lochbaum D, Alexis Nelson Nuclear Energy Institute, Union of Concerned Scientists
References
Download: ML090350419 (70)

Text

{{#Wiki_filter:-. November 23, 2001 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW Washington, DC 20006-3708 Suite 600 Washington, DC 20006-3919 SUB~IECT: PROPOSED STAFF GUIDANCE ON THE POSITION OF THE GALL REPORT PRESENTING ONE ACCEPTABLE WAY TO MANAGE AGING EFFECTS FOR LICENSE RENEWAL

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the enclosed guidance clarifying that NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," presents one acceptable way to manage aging effects for license renewal. This is consistent with our goal to more efficiently resolve license renewal issues identified by stakeholders as outlined in NRR Office Letter No. 805, "License Renewal Application Review Process." This letter reflects one of the lessons learned from the license renewal demonstration project that is Item No. 2.1 listed in a letter to Alan P. Nelson of Nuclear Energy Institute, from Christopher I. Grimes of NRC, dated October 3, 2001. A statement of the issue and background information is provided in Enclosure 1. We are requesting your comments on the proposed staff guidance (Enclosure 2), and we request that you submit comments within 30 days following the date of this letter to ensure a timely resolution of this issue. The staff plans this addition to NUREG-1800, "Standard Review Plan for License Renewal Applications for Nuclear Power Plants" (SRP-LR) in a future update. Also it is recommended that conforming changes be made to NEI 95-10, Revision 3, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule." If you have any questions regrading this matter, please contact Shoji Takeyama at 301-415-3873. Sincerely, IRAI Christopher I. Grimes, Chief License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosure:

As stated cc w/encl: See next page

Issue Heading: The GALL report presents one acceptable way to manage aging effects for license renewal

== Description:==

The following observation and its lesson learned is based on the license renewal demonstration project that demonstrated how an applicant would use the Generic Aging Lessons Learned (GALL) report in preparing its application and how the NRC staff would use the improved license renewal guidance documents to perform its review. This is Item No. 2.1 listed in a letter to Alan P. Nelson of NEI, from Christopher I. Grimes of NRC, dated October 3, 2001. ObseNation: The GALL report presents one acceptable way to manage aging. During the preparation of the request for additional information, the GALL report was sometimes treated as the only acceptable way. This is not consistent with the purpose of the GALL report. Lesson Learned: The GALL report indicates that it contains one acceptable way and not the only way to manage aging. However, this observation indicates that the GALL report or the other license renewal guidance documents should be revisited to see if further enhancement is necessary. Evaluation: In NUREG-1801, Vol. 1, "Generic Aging Lessons Learned (GALL) Report Summary," on page 3 under "APPLICATION OF THE GALL REPORT," third paragraph, the description of the application of the GALL report is as follows: The GALL report contains one acceptable way to manage aging effects for license renewal. An applicant may propose alternatives for staff review in its plant-specific license renewal application. Use of the GALL report is not required, but its use should facilitate both preparation of a license renewal application by an applicant and timely, uniform review by the NRC staff. Thus, it is clear that The GALL report indicates that it contains one acceptable way and not the only way to manage aging for license renewal. However, in NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (SRP-LR), there is not the clear description that the GALL report contains one acceptable way to manage aging effects for license renewal. The SRP-LR contains the following descriptions: SRP-LR. CHAPTER 3. SECTION 3.X.1. "AREAS OF REVIEW" The staff has issued a generic aging lessons learned (GALL) report addressing aging management for license renewal (Ref2). The GALL report documents the staffs basis for determining whether generic existing programs are adequate to Enclosure 1

-. manage aging without change or generic existing programs should be augmented for license renewal. The GALL report may be referenced in a license renewal application and should be treated in the same manner as an approved topical report. Because a license renewal applicant mayor may not be able to reference the GALL report as explained below, the following areas are reviewed. SRP-LR. CHAPTER 3, SECTION 3.X.1.3, "AGING MANAGEMENT EVALUATIONS THAT ARE DIFFERENT FROM OR NOT ADDRESSED IN THE GALL REPORT" The GALL report provides a generic staff evaluation of certain aging management programs (AMPs). If the applicant does not rely on a particular program for license renewal, or if the applicant indicates that the generic staff evaluation of the elements of a particular program does not apply to its plant, the staff should review each such AMP to which the GALL report does not apply. The GALL report provides a generic staff evaluation of programs for certain components and aging effects. If the applicant has identified particular components subject to aging management review (AMR) for its plant that are not addressed in the GALL report, or paticular aging effects for a component that are not addressed in the GALL report, the staff should review the applicant's AMPs applicable to these particular components and aging effects. SRP-LR. CHAPTER 4, SECTION 4.3.2.1.1.3, "10 CFR 54.21(c)(1)(iii)" In Chapter X of the GALL report (Ref 13), the staff has evaluated a program that monitors and tracks the number of critical thermal and pressure transients for the selected reactor coolant system components. The staff has determined that it is an acceptable aging management program to address metal fatigue of the reactor coolant system components according to 10 CFR 54.21(c)(1)(iii). The GALL report may be referenced in a license renewal application and should be treated in the same manner as an approved topical report. In referencing the GALL report, the applicant should indicate that the material referenced is applicable to the specific plant involved and should provide the information necessary to adopt the finding of program acceptability as described and evaluated in the report. The applicant should also verify that the approvals set forth in the GALL report for the generic program apply to the applicant's program. SRP-LR. CHAPTER 4, SECTION 4.3.3.1.1.3, "10 CFR 54.21 (c)(1)(iii)" The applicant may reference the GALL report in its license renewal application,as appropriate. The review should verify that the applicant has stated that the report is applicable to its plant with respect to its program that monitors and tracks the number of critical thermal and pressure transients for the selected reactor coolant system components. The reviewer verifies that the applicant has identified the appropriate program as described and evaluated in the GALL report. The reviewer

                                              -2

-. also ensures that the applicant has stated that its program contains the same program elements that the staff evaluated and relied upon in approving the corresponding generic program in the GALL report. No further staff evaluation is necessary. SRP-LR CHAPTER 4, SECTION 4.4.2.1.3, "10 CFR 54.21 (c)(1)(jii)" In Chapter X of the GALL report (Ref. 14), the staff has evaluated the environmental qualification program (10 CFR 50.49) and determined that it is an acceptable aging management program to address environmental qualification according to 10 CFR 54.21(c)(1)(iii). The GALL report may be referenced in a license renewal application, and should be treated in the same manner as an approved topical report. In referencing the GALL report, the applicant should indicate that the material referenced is applicable to the specific plant involved and should provide the information necessary to adopt the finding of program acceptability as described and evaluated in the report. The applicant should also verify that the approvals set forth in the GALL report for the generic program apply to the applicant's program. SRP-LR. CHAPTER 4, SECTION 4.4.3.1.3. "10 CFR 54.21(c)(1)(jii)" The applicant may reference the GALL report in its license renewal application, as appropriate. The review should verify that the applicant has stated that the report is applicable to its plant with respect to its environmental qualification program. The reviewer verifies that the applicant has identified the appropriate program as described and evaluated in the GALL report. The reviewer also ensures that the applicant has stated that its environmental qualification program contains the same program elements that the staff evaluated and relied upon in approving the corresponding generic program in the GALL report. No further staff evaluation is necessary. SRP-LR. CHAPTER 4, SECTION 4.5.2.1.3. "10 CFR 54.21(c)(1)(jii)" In Chapter X of the GALL report (Ref. 4), the staff has evaluated a program that assesses the concrete containment tendon prestressing forces, and has determined that it is an acceptable aging management program to address concrete containment tendon prestress according to 10 CFR 54. 21 (c)(1)(iii), except for operating experience. The GALL report recommends further evaluation of the applicant's operating experience related to the containment prestress force. The GALL report may be referenced in a license renewal application, and should be treated in the same manner as an approved topical report. In referencing the GALL report, the applicant should indicate that the material referenced is applicable to the specific plant involved and should provide the information necessary to adopt the finding of program acceptability as described and evaluated in the report. The applicant should also verify that the approvals set forth in the GALL report for the generic program apply to the applicant's program.

                                              -3

~ . SRP-LR. CHAPTER 4. SECTION 4.5.3.1.3. "10 CFR 54.21<c)(1)(jii)" An applicant may reference the GALL report in its license renewal application, as appropriate. The review should verify that the applicant has stated that the report is applicable to its plant with respect to its program that assesses the concrete containment tendon prestressing forces. The reviewer verifies that the applicant has identified the appropriate program as described and evaluated in the GALL report. The reviewer also ensures that the applicant has stated that its program contains the same program elements that the staff evaluated and relied upon in approving the corresponding generic program in the GALL report. The GALL report recommends further evaluation of the applicant's operating experience related to the containment prestress force. The applicant's program should incorporate the relevant operating experience that occurred at the applicant's plant as well as at other plants. The applicant should consider applicable portions of the experience with prestressing systems described in Information Notice 99-10 (Ref.3). Tendon operating experience could vary among plants with prestressed concrete containments. The difference could be due to the prestressing system design (for example, button-heads, wedge or swaged anchorages), environment, or type of reactor (PWR or BWR). The reviewer reviews the applicant's program to verify that the applicant has adequately considered plant-specific operating experience. The GALL report already states that it contains one acceptable way to manage aging effects. Therefore, no changes are needed in the GALL report. However, the SRP-LR should be clarified as attachment to this paper in 3.x.1 Areas of Review, 4.3.2.1.1.310CFR54.21(c)(1)(iii), 4.4.2.1.310CFR54.21(c)(1)(iii), and 4.5.2.1.3 10CFR54.21(c)(1)(iii) to explicitly indicate that the GALL report presents one acceptable way to manage aging effects for license renewal, and in 4.3.3.1.1.3 10CFR54.21 (c)(1)(iii), 4.4.3.1.3 10CFR54.21 (c)(1)(iii), and 4.5.3.1.3 10CFR54.21 (c)(1)(iii) to capture the thought that additional NRC staff evaluation will be required if a method other than the GALL report is relied on in the application for license renewal. Also, the staff recommends that NEI considers changing NEI 95-10, Revision 3, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule," to note that while the GALL report provides one acceptable way to manage aging effects, additional staff evaluation will be required if a method other than the GALL report is relied on in the application for license renewal.

                                               -4

Reference:

NUREG-1801, Vo\.1, "Generic Aging Lessons Learned (GALL) Report Summary," U.S. Nuclear Regulatory Commission, July 2001. NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, July 2001. NEI 95-10, Revision 3, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule," Nuclear Energy Institute, March 2001. Attachments: Markups of Chapters 3 and 4 of the SRP-LR are attached.

                                       -5

corresponding generic program in the GALL report. No further staff evaluation is necessary. If the applicant does not reference the GALL report in its renewal application. additional staff evaluation is necessary to determine whether the applicant's program is acceptable for this area of review. 4.4.2.1.3 10 CFR 54.21(c)(1)(iii) In Chapter X of the GALL report (Ref. 14), the staff has evaluated the environmental qualification program (10 CFR 50.49) and determined that it is an acceptable aging management program to address environmental qualification according to 10 CFR 54.21 (c)(1)(iii). The GALL report may be referenced in a license renewal application, and should be treated in the same manner as an approved topical report. However, the GALL report contains one acceptable way and not the only way to manage aging for license renewal. In referencing the GALL report, the applicant should indicate that the material referenced is applicable to the specific plant involved and should provide the information necessary to adopt the finding of program acceptability as described and evaluated in the report. The applicant should also verify that the approvals set forth in the GALL report for the generic program apply to the applicant's program. 4.4.3.1.3 10 CFR 54.21 (c)(1)(iii) The applicant may reference the GALL report in its license renewal application, as appropriate. The review should verify that the applicant has stated that the report is applicable to its plant with respect to its environmental qualification program. The reviewer verifies that the applicant has identified the appropriate program as described and evaluated in the GALL report. The reviewer also ensures that the applicant has stated that its environmental qualification program contains the same program elements that the staff evaluated and relied upon in approving the corresponding generic program in the GALL report. No further staff evaluation is necessary. If the applicant does not reference the GALL report in its renewal application. additional staff evaluation is necessary to determine whether the applicant's program is acceptable for this area of review. 4.5.2.1.3 10 CFR 54.21 (c)(1)(iii) In Chapter X of the GALL report (Ref.4), the staff has evaluated a program that assesses the concrete containment tendon prestressing forces, and has determined that it is an acceptable aging management program to address concrete containment tendon prestress according to 10 CFR 54.21 (c)(1)(iii), except for operating experience. The GALL report recommends further evaluation of the applicant's operating experience related to the containment prestress force. The GALL report may be referenced in a license renewal application, and should be treated in the same manner as an approved topical report. However, the GALL report contains one acceptable way and not the only way to manage aging for license renewal. In referencing the GALL report, the applicant should indicate that the material referenced is applicable to the specific plant involved and should provide the information necessary to adopt the finding of program acceptability as described and evaluated in the report. The applicant should also verify that the

                                           -2

approvals set forth in the GALL report for the generic program apply to the applicant's program. 4.5.3.1.3 10 CFR 54.21 (c){1)(iii) An applicant may reference the GALL report in its license renewal application, as appropriate. The review should verify that the applicant has stated that the report is applicable to its plant with respect to its program that assesses the concrete containment tendon prestressing forces. The reviewer verifies that the applicant has identified the appropriate program as described and evaluated in the GALL report. The reviewer also ensures that the applicant has stated that its program contains the same program elements that the staff evaluated and relied upon in approving the corresponding generic program in the GALL report. The GALL report recommends further evaluation of the applicant's operating experience related to the containment prestress force. The applicant's program should incorporate the relevant operating experience that occurred at the applicant's plant as well as at other plants. The applicant should consider applicable portions of the experience with prestressing systems described in Information Notice 99-10 (Ref.3). Tendon operating experience could vary among plants with prestressed concrete containments. The difference could be due to the prestressing system design (for example, button-heads, wedge or swaged anchorages), environment, or type of reactor (PWR or BWR). The reviewer reviews the applicant's program to verify that the applicant has adequately considered plant-specific operating experience. Also. if the applicant does not reference the GALL report in its renewal application. additional staff evaluation is necessary to determine whether the applicant's program is acceptable for this area of review.

                                           -3

• April 1, 2002 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW., Suite 600 Washington, DC 20006-3708 Washington, DC 20006-3919

SUBJECT:

STAFF GUIDANCE ON SCOPING OF EQUIPMENT RELIED ON TO MEET THE REQUIREMENTS OF THE STATION BLACKOUT (SBO) RULE (10 CFR 50.63) FOR LICENSE RENEWAL (10 CFR 54.4(a)(3>>

Dear Messrs. Nelson and Lochbaum:

The Nuclear Regulatory Commission (NRC) staff has reviewed the Nuclear Energy Institute's (NEI) comments, dated March 19, 2002, and the Union of Concerned Scientists' letter, dated February 19, 2002, on the proposed staff guidance for identifying equipment relied on to meet the requirements of the SBO rule 10 CFR 50.63, as it affects scoping for license renewal rule under 10 CFR 54.4(a)(3). The staff is enclosing a copy of the revised staff position on scoping of SBO equipment for license renewal. However, the staff would like to clarify the use of alternate ac power sources within the context of the SBO rule. Alternate ac power sources were accepted under the SBO rule as an alternate means of withstanding an SBO. The definition of an alternate ac power source is contained in 10 CFR 50.2. The definition addresses the capability of these power sources to cope with an SBO but not to recover from an SBO. While a very small number of alternate ac power sources may have capabilities beyond those required for coping, the staff nevertheless finds that they were only reviewed as a means of coping with an SBO for the plant specified coping duration. Reference to alternate ac power sources as a means of recovering from an SBO is therefore not intended within the context of the SBO rule. Within the context of the rule, only offsite power and onsite power are credited as means of recovering from an SBO event; and both must therefore be included within the scope of license renewal. An aging management program for SBO equipment that is within the scope of license renewal should address the 10 attributes described in the Standard Review Plan for License Renewal. For the attributes that address corrective action, confirmation process, and administrative controls, the staff has determined that 10 CFR Part 50, Appendix B is acceptable. However, Appendix A "Quality Assurance Guidance for Non-Safety Systems and Equipment" of Regulatory Guide 1.155, "Station Blackout" may be used subject to the staff review if and when a specific SBO aging management program is submitted by the applicant. The implementation of this staff position will start with the license renewal applications currently under review. Additional staff guidance for implementation of this staff position at Calvert Cliffs, Oconee, ANO-1, and Hatch will be issued separately.

                                                 -2 With the enclosed staff position, it is also possible that comparable changes might need to be made to NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54:- The License Renewal Rule." If you have any questions regarding this matter, please contact Peter Kang at 301-415-2779.

Sincerely, IRAI David B. Matthews, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project 690

Enclosure:

As stated cc w/encl: See next page

NRC Staff Position on the License Renewal Rule (10 CFR 54.4) as it relates to The Station Blackout Rule (10 CFR 50.63) Staff Position Consistent with the requirements specified in 10 CFR 54.4(a)(3) and 10 CFR 50.63(a)(1), the plant system portion of the offsite power system should be included within the scope of license renewal. The reasons for support of this position follow: Rationale The license renewal rule, 10 CFR 54.4(a)(3), requires that, "All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for. ....station blackout (10 CFR 50.63)" be included within the scope of license renewal. The station blackout (SBO) rule, 10 CFR 50.63(a)(1), requires that each light-water-cooled nuclear power plant licensed to operate be able to withstand and recover from a station blackout of a specified duration that is based upon factors that include: "(iii) The expected frequency of loss of offsite power; and (iv) The probable time needed to restore offsite power." The SBe rule in this regard is consistent with the staff findings identified in the statement of considerations and NUREG-1032, "Evaluation of Station Blackout Accidents at Nuclear Power Plants." In particular, with regard to factor (iv), the staff found that offsite power is more likely to be restored (0.6 hours median time to restore) than the emergency diesel generators (8 hours median time to repair) in terminating an SBO event. Station Blackout is the loss of offsite and onsite ac electric power to the essential and non essential switchgear buses in a nuclear power plant. It does not include the loss of ac power fed from inverters powered by station batteries nor loss of ac power from an SBe defined alternate ac power source. The SBO rule was added to the regulations in 10 CFR Part 50 because, as operating experience accumulated, concern arose that the reliability of both the offsite and onsite ac power systems might be less than originally anticipated, even for designs that met the requirements of General Design Criteria 17 and 18. As a result, the SBe rule required that nuclear power plants have the capability to withstand and recover from the loss of offsite and onsite ac power of a specified duration (the coping duration). Licensees' plant evaluations followed the guidance specified in NRC Regulatory Guide (RG) 1.155 and NUMARC 87-00 to determine their required plant-specific coping duration. The criteria specified in RG 1.155 to calculate a plant-specific coping duration were based upon the expected frequency of loss of offsite power and the probable time needed to restore offsite power, as well as the other two factors (onsite emergency ac power source redundancy and reliability) specified in 10 CFR 50.63(a)(1). In requiring that a plant's coping duration be based in part on the probable time needed to restore offsite power, 10 CFR 50.63(a)(1) is specifying that the offsite power system be an assumed method of recovering from an SBO. Disregarding the offsite power system as a means of recovering from an SBO would not meet the requirements of the rule and would result in a longer required coping duration. Enclosure

                                                 -2 The use of the offsite power system within 10 CFR 50.63(a)(1) as a means of recovering from an SBO should not be construed to be the only acceptable means of recovering from an SBO.

A licensee could for example recover offsite power or emergency (onsite) power. It is not possible to determine prior to an actual SBO event which source of power can be returned first. As a result, 10 CFR 50.63(c)(1)(ii) and its associated guidance in RG 1.155, Section 1.3 and Section 2, requires procedures to recover from an SBO that include restoration of offsite and onsite power. Based on the above, both the offsite and onsite power systems are relied upon to meet the requirements of the SBO rule. Elements of both offsite and onsite power are necessary to determine the required coping duration under 10 CFR 50.63(a)(1), and the procedures required by 10 CFR 50.63(c)(1)(ii) must address both offsite power and onsite power restoration. It follows, therefore, that both systems are used to demonstrate compliance with the SBO rule and must be included within the scope of license renewal consistent with the requirements of 10 CFR 54.4(a)(3). License renewal applicants are presently including the onsite power system within the scope of license renewal on the basis of the requirements under 10 CFR 54.4 (a)(1) (safety-related systems). They are also including equipment that is relied upon to cope with an SBO (e.g., alternate ac power sources) on the basis of the requirements under 10 CFR 54.4(a)(3). Only the addition of the offsite power system is therefore necessary to complete the required scope of the electrical power systems under license renewal. The offsite power systems of U.S. nuclear power plants consist of a transmission system (grid) component that provides a source of power and a plant system component that connects that power source to a plant's onsite electrical distribution system which powers safety equipment. The staff has historically relied upon the well-distributed, redundant, and interconnected nature of the grid to provide the necessary level of reliability to support nuclear power plant operations. For purposes of the license renewal rule, the staff has determined that the plant system portion of the offsite power system that is used to connect the plant to the offsite power source should be included within the scope of the rule. This path typically includes the switchyard circuit breakers that connect to the offsite system power transformers (startup transformers), the transformers themselves, the intervening overhead or underground circuits between circuit breaker and transformer and transformer and onsite electrical distribution system, and the associated control circuits and structures. Ensuring that the appropriate offsite power system long-lived passive structures and components that are part of this circuit path are subject to an aging management review will assure that the bases underlying the SSO requirements are maintained over the period of the extended license. This is consistent with the Commission's expectations in including the SSO regulated event under 10 CFR 54.4(a)(3) of the license renewal rule.

NUCLEAR ENERGY INSTITUTE Project No. 690 cc: Mr. Joe Bartell Mr. Robert Gill U.S. Department of Energy Duke Energy Corporation NE-42 Mail Stop EC-12R Washington, DC 20585 P.O. Box 1006 Charlotte, NC 28201-1006 Mr. Richard P. Sedano, Commissioner State Liaison Officer Mr. Joseph Gasper Department of Public Service Manager - Nuclear Licensing 112 State Street Omaha Public Power District Drawer 20 Fort Calhoun Station FC-2-4 Adm. Montipelier, Vermont 05620-2601 Post Office Box 399 Hwy. 75 - North of Fort Calhoun Mr. Stephen T. Hale Fort Calhoun, NE 68023-0399 Florida Power & Light Company 9760 S.W. 344 Street Mr. Paul Gunter Florida City, Florida 33035 Director of the Reactor Watchdog Project Nuclear Information & Resource Service Mr. William Corbin 1424 16th Street, NW, Suite 404 Virginia Electric & Power Company Washington, DC 20036 Innsbrook Technical Center 5000 Dominion Boulevard Mr. Hugh Jackson Glen Allen, Virginia 23060 Public Citizen's Critical Mass Energy & Environment Program Mr. Frederick W. Polaski 215 Pennsylvania Ave. SE Manager License Renewal Washington DC 20003 Exelon Corporation 200 Exelon Way Mary Olson Kennett Square, PA 19348 Nuclear Information & Resource Service, Southeast Office P.O. Box 7586 Asheville, North Carolina 28802

'. November 23, 2001 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 17761 Street, NW., Suite 400 1707 H Street, NW Washington, DC 20006-3708 Suite 600 Washington, DC 20006-3919 SUB~IECT: PROPOSED REVISION OF CHAPTERS II AND III OF GENERIC AGING LESSONS LEARNED (GALL) REPORT ON AGING MANAGEMENT OF CONCRETE ELEMENTS

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the proposed revision of Chapters II (Containment) and III (Structures) of the GALL report. The need for the proposed clarification on concrete in these GALL chapters was discussed with the Nuclear Energy Institute (NEI) during the lessons learned meeting on the license renewal demonstration project held on October 11, 2001. This issue is listed as item No. 3.9 in the letter to Alan Nelson of NEI, from Christopher Grimes of NRC, dated October 3, 2001. This process is consistent with our goal to more efficiently resolve license renewal issues identified by the staff or the industry as outlined in NRR Office Letter No. 805, "License Renewal Application Review Process." The staff has revised sections of Chapters II and III of the GALL report and a brief description of the basis for the clarification and its changes is provided in Enclosure 1. Since the proposed changes also affect other parts of the guidance documents, all pertinent changes in Chapters II and III of GALL Volume 2 and Table 5 of GALL Volume1 (NUREG-1801), and Table 3.5-1 of Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (NUREG-1800) have been included as Enclosure 2. We are requesting your comments on the proposed markup, and we request that you submit comments within 45 days following the date of this letter to ensure a timely resolution of this issue. The staff plans on incorporating this position into the improved renewal guidance documents in a future update. If you have any questions regarding this matter, please contact Peter Kang at 301-415-2779. Sincerely, IRAI Christopher I. Grimes, Chief License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

As stated cc w/encl: See next page Mr. Alan Nelson Washington, DC 20006-3708 Nuclear Energy Institute 17761 Street, NW., Suite 400

Description of the Basis for the Clarification and Changes for Concrete Elements (Item No. 3.9 in a letter to NEI dated October 3, 2001) The staff position is that aging of concrete elements of the containment and other Class 1 structures should be managed. Contrary to this position, the staff observed during review of the Nuclear Energy Institute (NEI) license renewal demonstration project that the participants misinterpreted the staffs license renewal guidance on this issue and determined that there were no aging effects requiring management for concrete elements for the above structures. The reasons for the misinterpretation were that: (1) applicable aging management programs (AMPs) for concrete elements in the current Generic Aging Lessons Learned (GALL) report were not clearly stated and (2) some inconsistencies were found between Chapters II and III of the GALL report for the concrete elements. During the review, all participants agreed that each AMP should be divided into accessible and inaccessible areas that are addressed separately. The staff agreed that applicable AMPs for degradation of concrete should be clarified in the GALL report and the Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR) of the improved renewal guidance documents. Changes: Under the "Aging Management Program (AMP)," column for concrete elements in the containment and Class 1 structures in Chapter II of the GALL report, the staff replaced the current AMP with a new AMP that addresses accessible and inaccessible areas separately based on: (1) the inspection for accessible areas would be required to be performed in accordance with ASME section XI, Subsection IWL under the requirements of 10 CFR 50.55a (existing mandated program) and (2) a plant-specific aging management program is required for inaccessible areas if the below grade environment is found to be aggressive or exposed to flowing water. Therefore, with the existing mandated program, the plant specific AMP for concrete elements in the containment and Class 1 structures is only required to address inaccessible areas. For Chapter III, the staff again replaced the current AMP with a new AMP that addresses accessible and inaccessible areas separately. The inspection for accessible areas would be required in accordance with a "Structural Monitoring Program" based on the requirement of 10 CFR 50.65 (Maintenance Rule), while a similar plant-specific program would be required for inaccessible areas by closely following the Chapter II example. The staff chose not to change the answer in the "Further Evaluation" column of the GALL report, but has clarified when further evaluation is required following the "Yes" or "No response. The staff has determined this was an efficient and effective method for updating the license renewal guidance documents. Enclosure 1

GALL REVISION Chapter II Containment Structures Enclosure 2

2 " Containment Structures

      - --  - --  ~ - -----  - - _._- ---------   - -_.- - - - -  ----_. - - ---- -----

Structure and/or Aging Effect! Further Item Component Material Environment Mechanism Aging Management Program (AMP) Evaluation A1.1-a Concrete elements: Concrete Outside Loss of material Chapter XI.S2, "ASME Section XI, No, if the containment (spalling, scaling) Subsection IWL" stated Dome; wall; basemat; ring and cracking I conditions girder; buttresses Freeze-thaw Accessible Areas: are satisfied Inspections performed in accordance for with IWL will indicate the presence of inaccessible loss of material (spalling, scaling) and areas cracking due to freeze-thaw. Inaccessible Areas: Evaluation is needed for plants that are located in moderate to severe weathering conditions (weathering index> 100 day-inch/yr) (NUREG-1557). Documented evidence to confirm that the in-place concrete had the air content between 3% to 6% and the subsequent inspections performed did not exhibit degradations related to freeze-thaw should be considered a part of the evaluation. The weathering index for the continental US is shown in ASTM C33 90, FiQ. 1. A1.1-b Concrete elements: Concrete Outside Increase in Chapter XI.S2, "ASME Section XI, Yes, a containment porosity, Subsection IWL" plant-Dome; wall; basemat; ring permeability, and specific girder; buttresses loss of strength/ Accessible Areas: aging Leaching of Inspections performed in accordance manageme calcium hydroxide with IWL will indicate the presence of nt program increase in porosity, and permeability is required due to leaching of calcium hydroxide. for inaccessible areas as

3 II Containment Structures

        --  - ---   -- - -  --- ----_._--   - ---.- --    - ..  --- -----,

Structure and/or Aging Effect! Further Item Component Material Environment Mechanism Aaina Manaaement Proaram (AMP) Evaluation Inaccessible Areas: stated A plant-specific aging management program is required for below-grade inaccessible areas (basemat and concrete wall), if the concrete is exposed to flowing water (NUREG 1557). An aging management program is not required, even if reinforced concrete is exposed to flowing water, if there is documented evidence that confirms the in-place concrete was constructed in accordance with the recommendations in ACI 201.2R-77. A1.1-c Concrete elements: Concrete Inside or Increase in Chapter XI.S2, "ASME Section XI, Yes, a outside porosity and Subsection IWL" plant-Dome; wall; basemat; ring containment permeability, specific girder; buttresses cracking, loss of Accessible Areas: aging material (spalling, Inspections performed in accordance manageme scaling) / with IWL will indicate the presence of nt program Aggressive increase in porosity and permeability, is required chemical attack cracking, or loss of material (spalling, for scaling) due to aggressive chemical inaccessible attack. areas as stated Inaccessible Areas: A plant-specific aging management program is required for below-grade exterior reinforced concrete (basemat, embedded walls), if the below-grade environment is aggressive (pH < 5.5. chlorides> 500 ppm, or sulfates> 1,500 ppm). Examination of representative samples of below-grade concrete, when excavated for any reason, is to be included as Dart of a

4 " Containment Structures

      --- - --- -------- ---~------------- --_.-------- ---- - ---------

Structure and/or Aging Effect! Further Item Component Material Environment Mechanism Aging Management Program (AMP) Evaluation plant-specific program. Note: Periodic monitoring of below-grade water chemistry (inclUding consideration of potential seasonal variations) is an acceptable approach to demonstrate that the below-grade environment is aggressive or non-aggressive. A1.1-d Concrete elements: Concrete Inside or Cracking and Chapter XI.S2, "ASME Section XI, No, if the outside expansion I Subsection IWL" stated Dome; wall; basemat; ring containment Reaction with conditions girders; buttresses aggregates Accessible Areas: are satisfied Inspections performed in accordance for with IWL will indicate the presence of inaccessible cracking due to reaction with areas aggregates. Inaccessible Areas: Evaluation is needed if investigations, tests, and petrographic examinations of aggregates performed in accordance with ASTM C295-54, ASTM C227-50, or ACI201.2R-77 (NUREG-1557) demonstrate that the aggregates are reactive. A1.1-e Concrete elements: Concrete; Inside or Cracking, loss of Chapter XI.S2, "ASME Section XI, Yes, a carbon outside bond, and loss of Subsection IWL" plant-Dome; wall; basemat; ring steel containment material (spalling, specific girders; buttresses; reinforcing scaling) / Accessible Areas: aging steel Corrosion of Inspections performed in accordance manageme embedded steel with IWL will indicate the presence of nt program cracking, loss of bond, and loss of is required material (spalling, scaling) due to for

5 II Containment Structures

       .. . .... .. ...... _ --- --_
                             .. ---   .. __ ........ _.. .. . .. _...... --. --- .. - .... -_ ..... -----,

Structure and/or Aging Effect! Further Item Component Material Environment Mechanism Aging Management Program (AMP) Evaluation corrosion of embedded steel. inaccessible areas as Inaccessible Areas: stated A plant-specific aging management program is required for below-grade exterior reinforced concrete (basemat, embedded walls), if the below-grade environment is aggressive (pH < 5.5, chlorides> 500 ppm, or sulfates> 1,500 ppm). Examination of representative samples of below-grade concrete, when excavated for any reason, is to be included as part of a plant-specific program. Note: Periodic monitoring of below-grade water chemistry (including consideration of potential seasonal variations) is an acceptable approach to demonstrate that the below-grade environment is aggressive or non aQQressive.

6 GALL REVISION Chapter III Structures and Component Supports

7 III Structures and Component Supports Al. uroup 1 ::structures ItsVVK Keactor tslog ., t"VVK ::SOlelo tsIOg., l,;ontrol Koom/tslog.) Structure and/or Aging Effect! Further Item Component Material Environment Mechanism Aging Management Program (AMP) Evaluation A1.1-a Concrete: Reinforced Weather Loss of material Chapter XI.S6, "Structures Monitoring No, if within Exterior above and below concrete exposed (spalling, scaling) Program" the scope of grade; foundation and cracking / the Freeze-thaw Accessible Areas: applicant's Inspections performed in accordance structures with "Structures Monitoring Program" monitoring will indicate the presence of loss of program material (spalling, scaling) and cracking and the due to freeze-thaw stated conditions Inaccessible Areas: are satisfied Evaluation is needed for plants that are for located in moderate to severe inaccessible weathering conditions (weathering areas index>100 day-inch/yr) (NUREG-1557). Documented evidence to confirm that the in-place concrete had the air content between 3% to 6% and the subsequent inspections performed did not exhibit degradations related to freeze-thaw should be considered a part of the evaluation. The weathering index for the continental US is shown in ASTM C33 90, FiQ. 1.

8 A1.1-b Concrete: Reinforced Flowing water Increase in Chapter XI.S6, "Structures Monitoring No, if within Exterior above and below concrete porosity and Program" the scope of grade; foundation permeability, and the loss of strength I Accessible Areas: applicant's Leaching of Inspections performed in accordance structures calcium hydroxide with "Structures Monitoring Program" monitoring will indicate the presence of increase in program porosity and permeability due to and a plant-leaching of calcium hydroxide specific aging Inaccessible Areas: manageme A plant-specific aging management nt program program is required for below-grade is required inaccessible areas (basemat and for concrete wall) if the concrete is inaccessible exposed to flowing water (NUREG areas as 1557). An aging management program stated is not required, even if reinforced concrete is exposed to flowing water, if there is documented evidence that confirms the in-place concrete was constructed in accordance with the recommendations in ACI201.2R-77.

9 A1.1-c Concrete: Reinforced Any Expansion and Chapter XI.S6, "Structures Monitoring No, if within All concrete cracking/ Program" the scope of Reaction with the aggregates Accessible Areas: applicant's Inspections/evaluations performed in structures accordance with "Structures Monitoring monitoring Program" will indicate the presence of program expansion and cracking due to reaction and the with aggregates. stated conditions Inaccessible Areas: are satisfied Evaluation is needed if investigations, for tests, and petrographic examinations of inaccessible aggregates performed in accordance areas with ASTM C295-54, ASTM C227-50, orACI201.2R-77 (NUREG-1557) demonstrate that the aggregates are reactive. A1.1-d Concrete: Reinforced Exposure to Cracking, loss of Chapter XI.S6, "Structures Monitoring No, if within Interior and above-grade concrete aggressive bond, and loss of Program" the scope of exterior environment material (spalling, the scaling) / Accessible Areas: applicant's Corrosion of Inspections performed in accordance structures embedded steel with "Structures Monitoring Program" monitoring will indicate the presence of cracking, program loss of bond, and loss of material (spalling, scaling) due to corrosion of embedded steel.

10 A1.1-e Concrete: Reinforced Exposure to Cracking, loss of Inaccessible Areas: Yes, a Below-grade exterior; concrete aggressive bond, loss of A plant-specific aging management plant-foundation environment material (spalling, program is required (may be a part of specific scaling) / structures monitoring program) if the aging Corrosion of below-grade environment is aggressive manageme embedded steel (pH < 5.5, chlorides >500ppm, or nt program sulfates> 1500 ppm). Examination of is required representative samples of below-grade for concrete, when excavated for any inaccessible reason, is to be included as part of a areas as plant-specific program. stated Note: Periodic monitoring of below-grade water chemistry (including consideration of potential seasonal variations) is an acceptable approach to demonstrate that the below-grade environment is aggressive or non aaaressive. A1.1-f Concrete: Reinforced Exposure to Increase in Chapter XI.S6, "Structures Monitoring No, if within Interior and above-grade concrete aggressive porosity and Program" the scope of exterior environment permeability, the cracking, loss of Accessible Areas: applicant's material (spalling, Inspections performed in accordance structures scaling)/ with "Structures Monitoring Program" monitoring Aggressive will indicate the presence of increase in program chemical attack porosity and permeability, cracking, or loss of material (spalling, scaling) due to aggressive chemical attack.

11 A1.1-g Concrete: Reinforced Exposure to Increase in Inaccessible Areas: Yes, a Below-grade exterior; concrete aggressive porosity and A plant-specific aging management plant-foundation environment permeability, program is required (may be a part of specific cracking, loss of structures monitoring program) if the aging material (spalling, below-grade environment is aggressive manageme scaling) I (pH < 5.5, chlorides >500 ppm, or nt program Aggressive sulfates >1500 ppm). Examination of is required chemical attack representative samples of below-grade for concrete, when excavated for any inaccessible reason, is to be included as part of a areas as plant-specific program. stated Note: Periodic monitoring of below-grade water chemistry (including consideration of potential seasonal variations) is an acceptable approach to demonstrate that the below-grade environment is aggressive or non aqqressive.

12 Proposed changes to SRP NUREG-1800 (Table 3.5-1) and GALL NUREG-1801 Vol-1 (Table 5) Table 3.5-1. Summary of Aging Management Programs for Structures and Component Supports Evaluated in Chapters /I and 1/1 of the GALL Report (continued) Aging Aging Effect! Management Further Evaluation Type Component Mechanism Programs Recommended BWRlPWR Concrete elements: Aging of accessible Containment lSI Yes, a plant-Foundation, wall, and inaccessible specific aging dome concrete areas due management to leaching of program is required calcium hydroxide, for inaccessible aggressive chemical areas as stated attack, and corrosion of embedded steel (see Subsection 3.5.2.2.1.1) BWR/PWR Concrete elements: Scaling, cracking, Containment lSI No, if stated foundation, dome, and and spalling due to conditions are wall freeze-thaw; satisfied for expansion and inaccessible areas cracking due to reaction with aggregate Class I Structures BWRlPWR All Groups except All types of aging Structures No, if within the Group 6: accessible effects Monitoring scope of the interior/exterior applicant's structures concrete & steel monitoring program components and a plant-specific aging management program is required for inaccessible areas as stated (see Subsection 3.5.2.2.2.1) BWR/PWR Groups 1-3, 5,7-9: Aging of Plant-specific Yes, a plant-inaccessible concrete inaccessible specific aging components, such as concrete areas due management exterior walls below to aggressive program is required grade and foundation chemical attack, and for inaccessible corrosion of areas as stated embedded steel (see Subsection 3.5.2.2.2.2)

13 TABLE 5.

SUMMARY

OF AGING MANAGEMENT PROGRAMS FOR THE STRUCTURES AND COMPONENT SUPPORTS EVALUATED IN CHAP"rERS II AND III OF THE GALL REPORT (CONTINUED) Aging Further Aging Effect! Management Evaluation GALL Type Component Mechanism Programs Recommended Item Number PWR Concrete (Reinforced and Prestressed) and Steel Containment BWR Concrete (Mark II and III) and Steel (Mark I, II, and III) Containment BWRI Concrete Aging of Containment Yes, a plant- II.A1.1-b, PWR elements: accessible and 151 specific aging II.AU-c, foundation, inaccessible management II.AU-e, dome, and wall concrete areas program is 11.A2.2-b, due to leaching required for 11.A2.2-c, of calcium inaccessible 11.A2.2-e, hydroxide, aggressive areas as stated II.B2.2.1-a. chemical attack, 1I.B2.2.1-b, and corrosion of II.B2.2.1-d, embedded steel II.B3.1.2-a, II.B3.1.2-b, II.B3.1.2-d, II.B3.2.1-b, II.B3.2.1-c, II.B3.2.1-e. BWRI Concrete Scaling, Containment No. if stated II.A1.1-a, PWR elements: cracking, and lSI conditions are II.AU-d, foundation, spalling due to satisfied for 11.A2.2-a, dome, and wall freeze-thaw; inaccessible II.A2.2-d, expansion and areas cracking due to II.B2.2.1-c, reaction with II.B3.1.2-c, aggregate 11.83.2.1-a, II.B3.2.1-d. Class I Structures BWRI All Groups All types of aging Structures No, if within the liLA1.1-a. PWR except Group 6: effects monitoring scope of the liLA1.1-b, accessible applicant's III.A1.1-c. interior/exterior structures liLA1.1-d, concrete and monitoring steel program and a liLA1.1-f, components III.A1.2-a. plant-specific aging 11I.A2.1-a. management III.A2.1-b, program is III.A2.1-c. required for III.A2.1-d. inaccessible III.A2.1-f, areas as stated III.A2.2-a, III.A3.1-a.

14 TABLE 5.

SUMMARY

OF AGING MANAGEMENT PROGRAMS FOR THE STRUCTURES AND COMPONENT SUPPORTS EVALUATED IN CHAPTERS II AND 1\1 OF THE GALL REPORT (CONTINUED) Aging Further Aging Effect! Management Evaluation GALL Type Component Mechanism Programs Recommended Item Number III.A3.1-b, III.A3.1-c, III.A3.1-d, III.A3.1-f, III.A3.2-a, III.A4.1-a, III.A4.1-b, III.A4.1-d. III.A4.2-a, III.A4.2-b, III.A5.1-a, III.A5.1-b, III.A5.1-c, III.A5.1-d, III.A5.1-f, III.A5.2-a, III.A7.1-a, III.A7.1-b. III.A7.1-c, III.A7.1-d. III.A7.1-f, III.A7.2-a, III.A8.1-a, III.A8.1-b, III.A8.1-c, III.A8.2-a, III.A9.1-a. III.A9.1-b, III.A9.1-c, III.A9.1-d, III.A9.1-f. BWRI Groups 1-3, 5, Aging of Plant Specific Yes, a plant III.A1.1-e, PWR 7-9: inaccessible specific aging III.A1.1-g, inaccessible concrete areas management III.A2.1-e, concrete due to program is III.A2.1-g, components, aggressive required for III.A3.1-e, such as chem ical attack inaccessible III.A3.1-g, and corrosion of exterior walls areas as stated III.A5.1-e, embedded steel below grade III.A5.1-g, and foundation III.A7.1-e, III.A7.1-g, III.A8.1-e,

15 TABLE 5.

SUMMARY

OF AGING MANAGEMENT PROGRAMS FOR THE STRUCTURES AND COMPONENT SUPPORTS EVALUATED IN CHAPTERS II AND III OF THE GALL REPORT (CONTINUED) Aging Further Aging Effect! Management Evaluation GALL Type Component Mechanism Programs Recommended Item Number III.A8.1-g, IlI.A9.1-e, III.A9.1-g.

December 3,2002 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW., Suite 600 Washington, DC 20006-3708 Washington, DC 20006-3919

SUBJECT:

INTERIM STAFF GUIDANCE (ISG)-04: AGING MANAGEMENT OF FIRE PROTECTION SYSTEMS FOR LICENSE RENEWAL

Dear Messrs. Nelson and Lochbaum:

By letter dated January 22, 2002, the Nuclear Regulatory Commission (NRC) staff provided the Nuclear Energy Institute (NEI) and the Union of Concerned Scientists the opportunity to comment on proposed guidance to clarify the fire protection systems program described in NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," dated July 2001. The staff received NEl's written comments on June 17,2002. The staff discussed these comments with NEI during a public meeting on July 25, 2002. The staff finalized the Interim Staff Guidance (ISG) to address these comments. The enclosed ISG consists of: (1) staff position (Enclosure 1); (2) aging management programs (Enclosure 2) for Chapter XI.M26, "Fire Protection" and Chapter XI.M27, "Fire Water Systems" of NUREG-1801, GALL Report; and (3) FSAR Supplement Table 3.3-2 (Enclosure 3) of NUREG-1800, "Standard Review Plan for Review of License Renewal." The staff plans to incorporate these comments into the improved license renewal guidance documents in a future update. The staff considers this ISG as clarifications with no additional requirements and, therefore, did not perform a backfit evaluation. As requested by NEI, the staff has assigned the following ISG designations for previously approved ISGs, to date, in order to assist the applicants with their license renewal applications before a future update. The corresponding ADAMS accession numbers are shown in the parentheses:

• ISG-01 Staff Guidance on the Position of the GALL Report Presenting One Acceptable Way to Manage Aging Effects for License Renewal (ML013300531 )
• ISG-02 Staff Guidance on Scoping of Equipment Relied on to Meet the Requirements of the Station Blackout Rule for License Renewal (ML020920464)
• ISG-03 Revision of Chapters II and III of GALL Report on Aging Management of Concrete Elements (ML013300426)

A. Nelson and D. Lochbaum -2 For the above approved ISGs, the staff requests that the NEI consider corresponding changes in NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule," if any. Should you have any questions regarding this matter, please contact Peter Kang at 301-415-2779. Sincerely, IRA! David B. Matthews, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

As stated cc w/encls: See next page

NRC Staff Position on Aging Management of Fire Protection Systems Introduction The staff plans to revise the Fire Protection (FP) system aging management program in NUREG-1801.

1. Staff Position for Wall Thinning of FP Piping Due to Internal Corrosion Fire Protection (FP) piping is typically designed for a 50-year life in industrial applications. The limiting aging mechanism is general corrosion. Because the general corrosion of FP piping is typically very uniform, loss of intended function as a result of catastrophic failure caused by wall thinning throughout the system is possible and needs to be managed. However, internal inspections performed during each refueling cycle by disassembling portions of the FP piping, as stated in NUREG-1801, Chapter XI.M27, "Fire Water Systems," may not be most effective means to detect this aging effect. Each time the system is opened, oxygen is introduced into the system and this accelerates the potential for general corrosion. Therefore, the staff recommends that the applicant perform a baseline pipe wall thickness evaluation of the fire protection piping using a non-intrusive means of evaluating wall thickness, such as volumetric inspection, to detect this aging effect before the current license term expires. The staff also recommends that the applicant perform pipe wall thickness evaluations at plant-specific intervals during the period of extended operation. The plant-specific inspection intervals are determined by engineering evaluation performed after each inspection of the fire protection piping to detect degradation prior to the loss of intended function. As an alternative to pipe wall thickness evaluations, an applicant may use the existing Chapter XI.M27.

As part of the review of this issue and the above stated approach, a concern was raised as to the inspection specifications of the internal surface of below grade FP piping. The staff acknowledges that some applicants may be able to demonstrate that the environmental and material conditions that exist on the interior surface of below grade FP piping are similar to the conditions that exist within the interior surface of the above grade FP piping. If an applicant makes such a demonstration, the staff agrees that the results of the interior inspections of the above grade FP piping can be extrapolated to evaluate the interior condition of the below grade FP piping. If not, additional inspection activities are needed to provide the reasonable assurance that the intended function of below grade FP piping will be maintained consistent with an applicant's current licensing basis for the period of extended operation.

2. Staff Position for Testing of Sprinkler Heads NFPA 25, 1999 Edition, Section 2.3.3.1, "Sprinklers," states, where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." NFPA 25 also contains guidance to perform this sampling every 10 years after the initial field service testing.

Enclosure 1

                                                  -2 The 50-year service life of sprinkler heads does not necessarily occur at the 50th year of operation in terms of licensing. The service life is defined from the time the sprinkler system is installed and functional. In most cases, sprinkler systems are in place several years before the operating license is issued. However, sprinkler systems in some plants may have been installed after the plant was placed in operation. The staff recommends, in accordance with NFPA 25, that sprinkler head testing should be performed at year 50 of sprinkler system service life, not at year 50 of plant operation, with subsequent sprinkler head testing every 10 years thereafter.

3. Staff Position for Valve Line-up Inspections of Halon/Carbon Dioxide Fire Suppression Systems NUREG-1801, Chapter XI.M26, "Fire Protection," currently identifies the need to perform a functional test of the halon/carbon dioxide fire suppression systems to determine the suppression agent charge pressure and verify that the extinguishing agent supply valves are open and the system is in automatic mode. Section 54.21 of Title 10 of the Code of Federal Regulations (CFR) specifies that an aging management review is to be performed for those structures and components that perform an intended function without moving parts, or without a change in configuration or properties, and that are not subject to replacement based on a qualified life or specified time period. The staff reviewed these items and determined that a valve lineup inspection, charging pressure inspection, and an automatic mode of operation verification are operational activities pertaining to system or component configurations or properties that may change, and are not related to aging management. Therefore, the staff position is to revise NUREG-1801 to eliminate the halon/carbon dioxide system inspections for charging pressure, valve lineups, and automatic mode of operation.

Backfit consideration The staff has determined that this ISG clarifies the staff position on aging management of FP systems and does not affect adequate protection or compliance with 10 CFR Part 54, the License Renewal Rule. Therefore, the staff did not evaluate this ISG for backfit.

XI.M26 FIRE PROTECTION Program Description For operating plants, the fire protection aging management program (AMP) includes a fire barrier inspection program and a diesel-driven fire pump inspection program. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection of fire rated doors to ensure that their operability is maintained. The diesel-driven fire pump inspection program requires that the pump be periodically tested to ensure that the fuel supply line can perform the intended function. The AMP also includes periodic inspection and test of halon/carbon dioxide fire suppression system. Evaluation and Technical Basis

1. Scope of Program: For operating plants, the AMP manages the aging effects on the intended function of the penetration seals, fire barrier walls, ceilings, and floors, and all fire rated doors (automatic or manual) that perform a fire barrier function. It also manages the aging effects on the intended function of the fuel supply line. The AMP also includes management of the aging effects on the intended function of the halon/carbon dioxide fire suppression system.

2. Preventive Actions: For operating plants, the fire hazard analysis assesses the fire potential and fire hazard in all plant areas. It also specifies measures for fire prevention, fire detection, fire suppression, and fire containment and alternative shutdown capability for each fire area containing structures, systems, and components important to safety.

3. Parameters Monitoredllnspected: Visual inspection of approximately 10% of each type of penetration seal is performed during walkdowns carried out at least once every refueling outage. These inspections examine any sign of degradation such as cracking, seal separation from walls and components, separation of layers of material, rupture and puncture of seals which are directly caused by increased hardness and shrinkage of seal material due to weathering. Visual inspection of the fire barrier walls, ceilings, and floors examines any sign of degradation such as cracking, spalling, and loss of material caused by freeze-thaw, chemical attack, and reaction with aggregates. Hollow metal fire doors are visually inspected on a plant specific interval to verify the integrity of door surfaces and for clearances. The plant specific inspection intervals are to be determined by engineering evaluation to detect degradation of the fire doors prior to the loss of intended function.

The diesel-driven fire pump is under observation during performance tests such as flow and discharge tests, sequential starting capability tests, and controller function tests for detecting any degradation of the fuel supply line. Periodic visual inspection and function test at least once every six months examines the signs of degradation of the halon/carbon dioxide fire suppression system. Material conditions that may affect the performance of the system, such as corrosion, mechanical damage, or damage to dampers, are observed during these tests. Enclosure 2

                                             -2

4. Detection of Aging Effects: Visual inspection of penetration seals detects cracking, seal separation from walls and components, and rupture and puncture of seals.

Visual inspection (VT-1 or equivalent) of approximately 10% of each type of seal in walkdowns is performed at least once every refueling outage. If any sign of degradation is detected within that sample, the scope of the inspection is expanded to include additional seals. Visual inspection (VT-1 or equivalent) of the fire barrier walls, ceilings, and floors performed in walkdown at least once every refueling outage ensures timely detection for concrete cracking, spalling, and loss of material. Visual inspection (VT-3 or equivalent) detects any sign of degradation of the fire door such as wear and missing parts. Periodic visual inspection detect degradation of the fire doors before there is a loss of intended function. Periodic tests performed at least once every refueling outage, such as flow and discharge tests, sequential starting capability tests, and controller function tests performed on diesel-driven fire pump ensure fuel supply line performance. The performance tests detect degradation of the fuel supply lines before the loss of the component intended function. Visual inspections of the halon/carbon dioxide fire suppression system detect any sign of degradation, such as corrosion, mechanical damage, or damage to dampers. The periodic function test and inspection performed at least once every six months detects degradation of the halon/carbon dioxide fire suppression system before the loss of the component intended function.

5. Monitoring and Trending: The aging effects of weathering on fire barrier penetration seals are detectable by visual inspection and, based on operating experience, visual inspections performed at least once every refueling outage detect any sign of degradation of fire barrier penetration seals prior to loss of the intended function.

Concrete cracking, spalling, and loss of material are detectable by visual inspection and, based on operating experience, visual inspection performed at least once every refueling outage detects any sign of degradation of the fire barrier walls, ceilings, and floors before there is a loss of the intended function. Based on operating experience, degraded integrity or clearances in the fire door are detectable by visual inspection performed on a plant specific frequency. The visual inspections detect degradation of the fire doors prior to loss of the intended function. The performance of the fire pump is monitored during the periodic test to detect any degradation in the fuel supply lines. Periodic testing provides data (e.g., pressure) necessary for trending. The performance of the halon/carbon dioxide fire suppression system is monitored during the periodic test to detect any degradation in the system. These periodic tests provide data necessary for trending.

6. Acceptance Criteria: Inspection results are acceptable if there are no visual indications of cracking, separation of seals from walls and components, separation of layers of material, or ruptures or punctures of seals; no visual indications of concrete cracking, spalling and loss of material of fire barrier walls, ceilings, and floors; no visual indications of degraded integrity or clearances in the fire doors. No corrosion is

                                             -3 acceptable in the fuel supply line for the diesel-driven fire pump. Also, any signs of corrosion and mechanical damage of the halon/carbon dioxide fire suppression system are not acceptable.

7. Corrective Actions: For fire protection structures and components identified within scope that are subject to an aging management review for license renewal, the applicant's 10 CFR Part 50, Appendix B, program is used for corrective actions, confirmation process, and administrative controls for aging management during the period of extended operation. This commitment is documented in the final safety analysis report (FSAR) supplement in accordance with 10 CFR 54.21(d). As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address corrective actions, confirmation process, and administrative controls.

8. Confirmation Process: See Item 7, above.

9. Administrative Controls: See Item 7, above.

10. Operating Experience: Silicone foam fire barrier penetration seals have experienced splits, shrinkage, voids, lack of fill, and other failure modes (IN 88-56, IN 94-28, and IN 97-70). Degradation of electrical raceway fire barrier such as small holes, cracking, and unfilled seals are found on routine walkdown (IN 91-47 and GL 92-08). Fire doors have experienced wear of the hinges and handles. Operating experience with the use of this AMP has shown that no corrosion-related problem has been reported for the fuel supply line, pump casing of the diesel-driven fire pump, and the halon/carbon dioxide suppression system. No significant aging related problems have been reported of fire protection systems, emergency breathing and auxiliary equipment, and communication equipment.

References NRC Generic Letter 92-08, Thermo-Lag 330-1 Fire Barrier, December 17, 1992. NRC Information Notice 88-56, Potential Problems with Silicone Foam Fire Barrier Penetration Seals, August 14, 1988. NRC Information Notice 91-47, Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test, August 6, 1991. NRC Information Notice 94-28, Potential problems with Fire-Barrier Penetration Seals, April 5, 1994. NRC Information Notice 97-70, Potential problems with Fire Barrier Penetration Seals, September 19, 1997.

                                              -4 XJ.M27 FIRE WATER SYSTEM Program Description This aging management program applies to water-based fire protection systems that consist of sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes, water storage tanks, and aboveground and underground piping and components that are tested in accordance with the applicable National Fire Protection Association (NFPA) codes and standards. Such testing assures the minimum functionality of the systems.

Also, these systems are normally maintained at required operating pressure and monitored such that loss of system pressure is immediately detected and corrective actions initiated. A sample of sprinkler heads is to be inspected by using the guidance of NFPA 25, Section 2.3.3.1. This NFPA section states that "where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." It also contains guidance to perform this sampling every 10 years after the initial field service testing. The fire protection sprinkler system piping is to be subjected to full flow tests at the maximum design flow and pressure or evaluated for wall thickness (e.g., non-intrusive volumetric testing or plant maintenance visual inspections) to ensure that corrosion aging effects are managed and that wall thickness is within acceptable limits. These inspections are performed before the end of the current operating term and at plant specific intervals thereafter during the period of extended operation. The plant specific inspection intervals are to be determined by engineering evaluation of the fire protection piping to detect degradation prior to the loss of intended function. The purpose of the full flow testing and wall thickness evaluations is to ensure that corrosion, microbiological influenced corrosion (MIG), or biofouling is managed such that the system function is maintained. Evaluation and Technical Basis

1. Scope of Program: The aging management program focuses on managing loss of material due to corrosion, MIC, or biofouling of carbon steel and cast-iron components in fire protection systems exposed to water. Hose stations and standpipes are considered as piping in the AMP.

2. Preventive Actions: To ensure no significant corrosion, MIC, or biofouling has occurred in water-based fire protection systems, periodic flushing, system performance testing, and inspections may be conducted.

3. Parameters Monitored/Inspected: Loss of material due to corrosion and biofouling could reduce wall thickness of the fire protection piping system and result in system failure. Therefore, the parameters monitored are the system's ability to maintain pressure and internal system corrosion conditions. Perform periodic flow testing of the fire water system using the guidelines of NFPA 25, Chapter 13, Annexes A & D at the maximum design flow or perform wall thickness evaluations to ensure that the system maintains its intended function.

                                             -5

4. Detection of Aging Effects: Fire protection system testing is performed to assure that the system functions by maintaining required operating pressures. Wall thickness evaluations of fire protection piping are performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections are performed before the end of the current operating term and at plant specific intervals thereafter during the period of extended operation. As an alternative to non-intrusive testing, the plant maintenance process may include a visual inspection of the internal surface of the fire protection piping upon each entry to the system for routine or corrective maintenance, as long as it can be demonstrated that inspections are performed (based on past maintenance history) on a representative number of locations on a reasonable basis.

These inspections must be capable of evaluating (1) wall thickness to ensure against catastrophic failure and (2) the inner diameter of the piping as it applies to the flow requirements of the fire protection system. If the environmental and material conditions that exist on the interior surface of the below grade fire protection piping are similar to the conditions that exist within the above grade fire protection piping, the results of the inspections of the above grade fire protection piping can be extrapolated to evaluate the condition of below grade fire protection piping. If not, additional inspection activities are needed to ensure that the intended function of below grade fire protection piping will be maintained consistent with the current licensing basis for the period of extended operation. Repair and replacement actions are initiated as necessary. Continuous system pressure monitoring, system flow testing, and wall thickness evaluations of piping are effective means to ensure that corrosion and biofouling are not occurring and the system's intended function is maintained. General requirements of existing fire protection programs include testing and maintenance of fire detection and protection systems and surveillance procedures to ensure that fire detectors, as well as fire protection systems and components, are operable. Visual inspection of yard fire hydrants performed annually per NFPA 25 ensures timely detection of signs of degradation, such as corrosion. Fire hydrant hose hydrostatic tests, gasket inspections, and fire hydrant flow tests, performed annually, ensure that fire hydrants can perform their intended function and provide opportunities for degradation to be detected before a loss of intended function can occur. Sprinkler heads are inspected before the end of the 50-year sprinkler head service life and at 10 year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner.

5. Monitoring and Trending: System discharge pressure is monitored continuously.

Results of system performance testing are monitored and trended as specified by the NFPA codes and standards. Degradation identified by non-intrusive or internal inspection is evaluated.

6. Acceptance Criteria: The acceptance criteria are (a) the ability of a fire protection system to maintain required pressure, (b) no unacceptable signs of degradation observed during non-intrusive or visual assessment of internal system conditions,

                                             -6 and (c) that no biofouling exists in the sprinkler systems that could cause corrosion in the sprinkler heads.

7. Corrective Actions: For fire water systems and components identified within scope that are subject to an aging management review for license renewal, the applicant's 10 CFR Part 50, Appendix B, program is used for corrective actions, confirmation process, and administrative controls for aging management during the period of extended operation. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable to address corrective actions, confirmation process, and administrative controls.

8. Confirmation Process: See Item 7, above.

9. Administrative Controls: See Item 7, above.

10. Operating Experience: Water-based fire protection systems designed, inspected, tested and maintained in accordance with the NFPA minimum standards have demonstrated reliable performance.

References NFPA 25: Inspection, Testing and Maintenance of Water-Based Fire Protection Systems, 1998 Edition. NFPA 25: Inspection, Testing and Maintenance of Water-Based Fire Protection Systems, 2002 Edition.

Table 3.3-2. FSAR Supplement for Aging Management of Auxiliary Systems (continued) Implementation Program Description of Program Schedule* Compressed air The program consists of inspection, monitoring, and testing Existing program monitoring of the entire system, including (1) frequent leak testing (BWRlPWR) valves, piping, and other system components, especially those made of carbon steel; and (2) preventive monitoring that checks air quality at various locations in the system to ensure that oil, water, rust, dirt, and other contaminants are kept within the specified limits. This program is in response to NRC GL 88-14 and INFO's Significant Operating Experience Report (SOER) 88-01. It also relies on the ASME OM Guide Part 17, and ISA-S7.0.l-1996 as guidance for testing and monitoring air quality and moisture. Fire protection The program includes a fire barrier inspection program and a Existing program (BWRlPWR) diesel-driven fire pump inspection program. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained. The diesel-driven fire pump inspection program requires that the pump be periodically tested to ensure that the fuel supply line can perform the intended function. The AMP also includes periodic inspection and test of halon/carbon dioxide fire suppression system. Fire water system To ensure no wall thinning or biofouling has occurred in the Program should be (BWRlPWR) fire protection system, periodic full flow flush test and modified before the system performance test are conducted to prevent corrosion period of extended from biofouling of components. Also, the system is normally operation maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions initiated. The AMP relies on testing of water based fire protection system piping and components in accordance with applicable NFPA commitments. In addition, this program will be modified to ensure fire protection piping is full flow tested or the wall thickness is evaluated prior to the end of the period of extended operation and at a plant specific intervals thereafter. Fuel oil chemistry The AMP relies on a combination of surveillance and Existing program (BWRJPWR) maintenance procedures. Monitoring and controlling fuel oil contamination in accordance with the guidelines of ASTM Standards 01796, D2276, D2709, and D4057, maintains the fuel oil quality. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by periodic cleaning/draining of tanks and by verifying the quality of new oil before its introduction into the storage tanks. Enclosure 3

Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW., Suite 600 Washington, DC 20006-3708 Washington, DC 20006-3919 SUB..IECT: INTERIM STAFF GUIDANCE (ISG) - 5 ON THE IDENTIFICATION AND TREATMENT OF ELECTRICAL FUSE HOLDERS FOR LICENSE RENEWAL

Dear Messrs. Nelson and Lochbaum:

The Nuclear Regulatory Commission (NRC) staff (Staff) has finalized the proposed ISG on the identification and treatment of electrical fuse holders for license renewal that was issued on May 16, 2002. The Staff considered comments from a NuClear Energy Institute (NEI) letter, dated June 19, 2002, and a Union of Concerned Scientists letter, dated May 23, 2002. Based on inslghts gained during the Staff's review of license renewal applications, the Staff finds that the previous ISG is sufficient to address the aging effects on insulation material for fuse blocks, but not sufficient to detect the aging effects on metallic clamps for the fuse clips of the fuse holder. Thus, the revised ISG concludes that both the insulation material and the metallic clamps of fuse holders are subject to aging management for license renewal. is a copy of the revised ISG for fuse holders. Enclosure 2 includes pertinent changes to (1) Chapter VI of "Generic Aging Lessons Learned (GALL) Report" (NUREG-1801) and (2) Table 2.1-5 of "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (NUREG-1800). The Staff is currently developing an appropriate aging management program for metallic metal clips which will be incorporated into NUREG-1801. The implementation of this Staff position will start with the license renewal applications currently under review. Additional Staff guidance for implementation of the Staff position at plants with a renewed license will be issued separately.

A. Nelson and D. Lochbaum -2 For the resolved ISGs, it is also possible that comparable changes might need to be made to NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule." If you have any questions regarding this matter, please contact Peter Kang at 301-415-2779. Sincerely, David B. Matthews, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

As stated cc w/encls: See next page

INTERIM STAFF GUIDANCE (ISG)-5 ON THE IDENTIFICATION AND TREATMENT OF ELECTRICAL FUSE HOLDERS FOR LICENSE RENEWAL Staff Position Consistent with the requirements specified in 10 CFR 54.4(a), fuse holders (including fuse clips and fuse blocks) are considered to be passive electrical components. Fuse holders would be scoped, screened, and included in the aging management review (AMR) in the same manner as terminal blocks and other types of electrical connections that are currently being treated in the process. This staff position only applies to fuse holders that are not part of a larger assembly, but support safety-related and non safety-related functions in which the failure of a fuse precludes a safety function from being accomplished [10 CFR Part 54.4(a)(1) and (a)(2)]. Examples are fuses that are used as protective devices to ensure the integrity of containment electrical penetrations when they are challenged by electrical faults, or as isolation devices between Class 1E and non-Class 1E electrical circuits to ensure that the safety function is not compromised as a result of faults in the non-Class 1E circuits. An appropriate aging management program (AMP) should be adopted to manage the effects of aging where necessary. Rationale The intended functions of a fuse holder are to provide mechanical support for the fuse and to maintain electrical contact with the fuse blades or metal end caps to prevent the disruption of the current path during normal operating conditions when the circuit current is at or below the current rating of the fuse. Fuse holders perform the same primary function as connections; they provide electrical connections to specified sections of an electrical circuit to deliver rated voltage, current, or signals. The intended functions of fuse holders meet the criteria of 10 CFR 54.4(a) and are performed without moving parts or without a change in configuration or properties as described in 10 CFR 54.21 (a)(1 )(i). The staff concludes that fuse holders are passive, long-lived electrical components within the scope of license renewal and subject to an AMR. However, fuse holders inside the enclosure of an active component, such as switchgear, power supplies, power inverters, battery chargers, and circuit boards, are considered to be piece parts of the larger assembly. Therefore, under 10 CFR 54.21, fuse holders that are parts of a larger assembly are considered outside the scope for license renewal. For license renewal purposes, fuse holders/blocks are classified as a specialized type of terminal block because of the similarity in design and construction. Terminal blocks are passive components subject to an AMR for license renewal. However, like fuses, terminal blocks located inside the enclosure of an active component are considered to be piece parts of the larger assembly and, thus, are outside the scope of license renewal. The fuse holders are typically constructed of blocks of rigid insulating material, such as phenolic resins. Metallic clamps are attached to the blocks to hold each end of the fuse. The clamps can be spring-loaded clips that allow the fuse ferrules or blades to slip in, or they can be bolt lugs, to which the fuse ends are bolted. The clamps are typically made of copper. Enclosure 1

                                               -2 Operational experience, as discussed in NUREG-1760 (Aging Assessment of Safety-Related Fuses Used in Low- and Medium-Voltage Applications in Nuclear Power Plants), identified fuse holders as experiencing a number of age-related failures. Aging stressors such as vibration, thermal cycling, electrical transients, mechanical stress, fatigue, corrosion, chemical contamination, or oxidation of the connecting surfaces can result in fuse holder failure. On this basis, fuse holders (including both the insulation material and the metallic clamps) are subject to both an AMR and AMP for license renewal. Typical plant effects observed from fuse holder failures due to aging have resulted in: challenges to safety systems, cable insulation failure due to over-temperature, failure of a containment spray pump to start, a reactor trip, etc. Therefore, managing age-related failures of fuse holders would have a positive effect on the safety performance of a plant. Information Notices 91-78, 87-42, and 86-87 provide examples that underscore the safety significance of fuse holders and the potential problems that can arise from age-related fuse holder failures.

GALL AMP for Fuse Holders Fuse holders, are considered as electrical connections and, thus, are subject to GALL XI.E1 "Electrical Cables and Connections Not SUbject to 10 CFR 50.49 Environmental Qualification Requirements." However, the AMP for fuse holders needs to include the following aging stressors, if applicable: fatigue, mechanical stress, vibration, chemical contamination, and corrosion. Where environments or operating conditions preclude such aging effects (e.g., fuse holders not SUbject to vibration from rotating machinery), they need not be addressed by the AMP. GALL XI.E1 is based on only a visual inspection of accessible cables and connections. Visual inspection, alone, may not be sufficient to detect the aging effects from fatigue, mechanical stress, vibration, or corrosion on the metallic clamps of the fuse holder. Other methods of aging detection may be necessary. Alternatively, plant modifications or administrative controls that have been made, which preclude these types of aging effects from occurring, would eliminate the need for an additional AMP (Le., the GALL XI.E1 program will be adequate).

A. ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 Environmental Qualification Requirements Systems, Structures and Components This section addresses electrical cables and connections that are not subject to the environmental qualification requirements of 10 CFR 50.49, and that are installed in power and instrumentation and control (I&C) applications. The power cables and connections addressed are low-voltage <<1000V) and medium-voltage (2kV to 15kV). High voltage (> 15kV) power cables and connections have unique, specialized constructions and must be evaluated on an application specific basis. Electrical cables and their required terminations (Le., connections) are typically reviewed as a single commodity. The types of connections included in this review are splices, mechanical connectors, fuse holders, and terminal blocks. This common review is translated into program actions, which treat cables and connections in the same manner. Electrical cables and connections that are in the plant's environmental qualification (EO) program are addressed in VI.B. System Interfaces Electrical cables and connections functionally interface with all plant systems that rely on electric power or instrumentation and control. Electrical cables and connections also interface with and are supported by structural commodities (e.g., cable trays, conduit, cable trenches, cable troughs, duct banks, cable vaults and manholes) that are reviewed, as appropriate, in the Structures and Components Supports section. NUREG-1801 VI A-2 April 2001 Enclosure 2

77 Electrical and I&C Cables and Connections, Bus, electrical Yes portions of Electrical and I&C Penetration Assemblies (e.g., electrical penetration assembly cables and connections, connectors, electrical splices, fuse holders, terminal blocks, power cables, control cables, instrument cables, insulated cables, communication cables, uninsulated ground conductors, transmission conductors, isolated-phase bus, nonsegregated-phase bus, segregated-phase bus, switchyard bus) 78 Electrical and I&C Chargers, Converters, Inverters No (e.g., converters-voltage/current, converters voltage/pneumatic, battery chargers/inverters, motor-generator sets) 79 Electrical and I&C Circuit Breakers No (e.g., air circuit breakers, molded case circuit breakers, oil-filled circuit breakers) 80 Electrical and I&C Communication Equipment No (e.g., telephones, video or audio recording or playback equipment, intercoms, computer terminals, electronic messaging, radios, transmission line traps and other power-line carrier equipment) 81 Electrical and I&C Electric Heaters No Yes for a Pressure Boundary if applicable 82 Electrical and I&C Heat Tracino No 83 Electrical and I&C Electrical Controls and Panel Internal No Component Assemblies (may include internal devices such as, but not limited to, switches, breakers, indicating lights, etc.) (e.g., main control board, HVAC control board) 84 Electrical and I&C Elements, RTDs, Sensors, Thermocouples, No Transducers Yes for a Pressure (e.g., conductivity elements, flow elements, Boundary if applicable temperature sensors, radiation sensors,watt transducers, thermocouples, RTDs, vibration probes, amp transducers, frequency transducers, power factor transducers, speed transducers, var. transducers, vibration transducers, voltaae transducers) 85 Electrical and I&C Fuses No 86 Electrical and I&C Generators, Motors No (e.g., emergency diesel generators, ECCS and emergency service water pump motors, small motors, motor-generator sets, steam turbine generators, combustion turbine generators, fan motors, pump motors, valve motors, air compressor motors) NUREG-1800 2.1-2 April 2001

A. ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 Environmental Qualification Requirements Systems, Structures and Components This section addresses electrical cables and connections that are not subject to the environmental qualification requirements of 10 CFR 50.49, and that are installed in power and instrumentation and control (I&C) applications. The power cables and connections addressed are low-voltage <<1 OOOV) and medium-voltage (2kV to 15kV). High voltage (>15kV) power cables and connections have unique, specialized constructions and must be evaluated on an application specific basis. Electrical cables and their required terminations (Le., connections) are typically reviewed as a single commodity. The types of connections included in this review are splices, mechanical connectors" fuse holders, El!'9 J~rI!lLn_al plo_c_k~~ TbLs_ ~~I'DIJ1_o_n 1- --I~Fo_rm_att_ed --J review is translated into program actions, which treat cables and connections in the same manner. Electrical cables and connections that are in the plant's environmental qualification (EO) program are addressed in VI.B. System Interfaces Electrical cables and connections functionally interface with all plant systems that rely on electric power or instrumentation and control. Electrical cables and connections also interface with and are supported by structural commodities (e.g., cable trays, conduit, cable trenches, cable troughs, duct banks, cable vaults and manholes) that are reviewed, as appropriate, in the Structures and Components Supports section. NUREG-1801 VI A-2 April 2001 Enclosure 2

77 Electrical and I&C Cables and Connections, Bus, electrical Yes portions of Electrical and I&C Penetration Assemblies (e.g., electrical penetration assembly cables and connections, connectors, electrical splices,

                       .fuse holders, t~~~ilJ,!1 PI9c;:~s" p~~e!p~~I~s-, __ -----------------

___ - -{ Formatted control cables, instrument cables, insulated cables, communication cables, uninsulated ground conductors, transmission conductors, isolated-phase bus, nonsegregated-phase bus, seareaated-ohase bus, switchyard bUS) 78 Electrical and I&C Chargers, Converters, Inverters No (e.g., converters-voltage/current, converters voltage/pneumatic, battery chargerslinverters, motor-aenerator sets\ 79 Electrical and I&C Circuit Breakers No (e.g., air circuit breakers, molded case circuit breakers, oil-filled circuit breakers) 80 Electrical and I&C Communication Equipment No (e.g., telephones, video or audio recording or playback equipment, intercoms, computer terminals, electronic messaging, radios, transmission line traps and other power-line carrier eauioment\ 81 Electrical and I&C Electric Heaters No Yes for a Pressure Boundarv if aoolicable 82 Electrical and I&C Heat Tracina No 83 Electrical and I&C Electrical Controls and Panel Internal No Component Assemblies (may include internal devices such as, but not limited to, switches, breakers, indicating lights, etc.) (e.g., main control board, HVAC control board) 84 Electrical and I&C Elements, RTDs, Sensors, Thermocouples, No Transducers Yes for a Pressure (e.g., conductivity elements, flow elements, Boundary if applicable temperature sensors, radiation sensors,watt transducers, thermocouples, RTDs, vibration probes, amp transducers, frequency transducers, power factor transducers, speed transducers, var. transducers, vibration transducers voltage transducers\ 85 Electrical and I&C Fuses No 86 Electrical and I&C Generators, Motors No (e.g., emergency diesel generators, ECCS and emergency service water pump motors, small motors, motor-generator sets, steam turbine generators, combustion turbine generators, fan motors, pump motors, valve motors, air comoressor motors) NUREG-1800 2.1-2 April 2001

A. ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 Environmental Qualification Requirements Systems, Structures and Components This section addresses electrical cables and connections that are not subject to the environmental qualification requirements of 10 CFR 50.49, and that are installed in power and instrumentation and control (I&C) applications. The power cables and connections addressed are low-voltage <<1000V) and medium-voltage (2kV to 15kV). High voltage (>15kV) power cables and connections have unique, specialized constructions and must be evaluated on an application specific basis. Electrical cables and their required terminations (Le., connections) are typically reviewed as a single commodity. The types of connections included in this review are splices, mechanical connectors.. fuse holders, ElM J~rI"DLnEll !>!o_c.!<§~ T~Ls_~~I"DIJl_O.!1 1- -1'-Fo_nn_a_tted ____ review is translated into program actions, which treat cables and connections in the same manner. Electrical cables and connections that are in the plant's environmental qualification (EO) program are addressed in VI.B. System Interfaces Electrical cables and connections functionally interface with all plant systems that rely on electric power or instrumentation and control. Electrical cables and connections also interface with and are supported by structural commodities (e.g., cable trays, conduit, cable trenches, cable troughs, duct banks, cable vaults and manholes) that are reviewed, as appropriate, in the Structures and Components Supports section. NUREG-1801 VI A-2 April 2001 Enclosure 2

77 Electrical and I&C Cables and Connections, Bus, electrical Yes portions of Electrical and I&C Penetration Assemblies (e.g., electrical penetration assembly cables and connections, connectors, electrical splices, luse holders, t!l~rT!iQC!1 !>~9'5.S1. p<!V!e! .9~~I~sJ __ - - - - - - - - - - - - - - - - - - _ ** - { Fonnatted control cables, instrument cables, insulated cables, communication cables, uninsulated ground conductors, transmission conductors, isolated-phase bus, nonsegregated-phase bus, seareaated-ohase bus switchyard bus) 78 Electrical and I&C Chargers, Converters, Inverters No (e.g., converters-voltage/current, converters voltage/pneumatic, battery chargerslinverters, motor-aenerator sets) 79 Electrical and I&C Circuit Breakers No (e.g., air circuit breakers, molded case circuit breakers, oil-filled circuit breakers) 80 Electrical and I&C Communication Equipment No (e.g., telephones, video or audio recording or playback equipment, intercoms, computer terminals, electronic messaging, radios, transmission line traps and other power-line carrier equipment) 81 Electrical and I&C Electric Heaters No Yes for a Pressure Boundary if applicable 82 Electrical and I&C Heat Tracing No 83 Electrical and I&C Electrical Controls and Panel Internal No Component Assemblies (may include internal devices such as, but not limited to, switches, breakers, indicating lights, etc.) I (e.g., main control board, HVAC control board) 84 Electrical and I&C Elements, RTDs, Sensors, Thermocouples, No Transducers Yes for a Pressure (e.g., conductivity elements, flow elements, Boundary if applicable temperature sensors, radiation sensors,watt transducers, thermocouples, RTDs, vibration probes, amp transducers, frequency transducers, power factor transducers, speed transducers, var. transducers, vibration transducers, voltaae transducers) 85 Electrical and I&C Fuses No 86 Electrical and I&C Generators, Motors No (e.g., emergency diesel generators, ECCS and emergency service water pump motors, small motors, motor-generator sets, steam turbine generators, combustion turbine generators, fan motors, pump motors, valve motors, air comoressor motors) NUREG-1800 2.1-2 April 2001

Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW., Suite 600 Washington, DC 20006-3708 Washington, DC 20006-3919

SUBJECT:

PROPOSED INTERIM STAFF GUIDANCE (ISG) ON IDENTIFICATION AND TREATMENT OF HOUSING FOR ACTIVE COMPONENTS FOR LICENSE RENEWAL

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the proposed ISG on identification and treatment of housing for active components for license renewal. The NRC staff (the staff) has developed the subject ISG to clarify that screening of housings for active mechanical components is conducted in accordance with the requirements of Section 54.21 (a){1) of the license renewal rule (10 CFR Part 54). The staff has made further revisions to this ISG since it issued a May 1, 2002, letter that solicited comments on the issue. The proposed ISG incorporates lessons learned from recent license renewal reviews to indicate that housings for fans, dampers, and heating and cooling coils may be subject to an aging management review for license renewal. is a copy of the proposed ISG on identification and treatment of housing for active components for license renewal. Enclosure 2 includes pertinent changes to Table 2.1-5 of the Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR). The staff plans to incorporate this ISG position into the improved renewal guidance documents (NUREG-1800 and 1801) in a future update. We request that you comment on the proposed ISG and submit a schedule for resolution, in order to ensure a timely closure of this issue.

A. Nelson and D. Lochbaum -2 For the resolved ISGs, it is also possible that comparable changes might be needed to NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule." If you have any questions regarding this matter, please contact Peter Kang at 301-415-2779. Sincerely, David B. Matthews, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project 690

Enclosures:

As stated cc w/encls: See next page

PROPOSED INTERIM STAFF GUIDANCE (ISG) ON SCREENING OF HOUSINGS FOR ACTIVE COMPONENTS FOR LICENSE RENEWAL Staff Position Pump casings and valve bodies, as described in both the license renewal rule (10 CFR Part 54 or the Rule) and its statement of consideration (SOC), are examples of how an applicant should evaluate housings for active mechanical components. The proper implementation of 10 CFR Part 54 requires screening evaluations to consider not just the active mechanical component, but the intended function of its associated housing. Specifically, based on experience with recent license renewal reviews, the staff believes that the housings for fans, dampers, and heating and cooling coils may perform a critical pressure retention and/or structural integrity function which, should that function not be maintained, could prevent the associated active mechanical component from performing its intended function. Further, if such housings perform these functions and meet the long-lived and passive criteria, then the housings should be subject to an aging management review (AMR). The staff expects applicants for license renewal to identify active mechanical component housings that require an AMR. This determination should consider whether failure of the housing would result in a failure of the associated active mechanical component to perform its intended function, and whether the housing meets the long-lived and passive criteria as defined in the Rule. The reasons for support of this position follow: Rationale Section 54.29 of the Rule states that a renewed license may be issued by the Commission if the Commission finds that actions have been or will be taken with respect to the matters identified in paragraphs (a)(1) and (a)(2) of this section, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis (CLB), and that any changes made to the CLB, in order to comply with this paragraph, are in accord with the Atomic Energy Act and the Commission's regulations. These matters include managing the effects of aging during the period of extended operation in order to assure the functionality of structures and components that have been identified as requiring review under Section 54.21 (a)(1). The SOC for the Rule states that the objective of a license renewal review is to determine whether the detrimental effects of aging, which could adversely affect the functionality of systems, structures, and components (SSCs) that the Commission determines require review for the period of extended operation, are adequately managed. The SOC articulates the underlying philosophy of the Rule; that during the period of extended operation, safety-related functions should be maintained in the same manner and to the same extent as they are during the current licensing term. Aging effects, which could adversely impact the ability of SSCs to maintain these safety-related functions during the period of extended operation, should be evaluated. Section 54.21 (a)(1) of the Rule states that components that perform their intended functions without moving parts and without a change in configuration or properties (Section 54.21 (a)(1 )(i>> and that are not subject to replacement based on qualified life or a specified time period ENCLOSURE 1

                                                 -2 (Section 54.21 (a)(1 )(ii)), are subject to an AMR. Such components are commonly considered to be "long-lived" and to perform a passive function. Section 54.21 (a)(1 )(i) states, These structures and components include, but are not limited to ... pump casings, valve bodies ..."

and lists other components that perform passive functions. The examples cited in the Rule illustrate components with passive functions. Section 1I1.f.i(a) of the SOC explains that major components may have active functions, passive functions, or both, and cites pumps and valves as examples. Pumps and valves have moving parts, but the Commission concluded that the pressure-retaining function performed by the pump casing and the valve body should be subject to an AMR. The SOC further explains that the Commission does not limit the consideration of pressure boundaries to the reactor coolant pressure boundary. The exclusion regarding components is focused on active functions, rather than on the exclusion of the entire component, while the AMR applies to the passive function of the component. On this basis, the staff concludes that the discussion of pump casings and valve bodies in both the Rule and its SOC are provided as examples of how an applicant should evaluate housings for active mechanical components, and that proper implementation of the Rule requires that screening evaluations consider not just the active mechanical component, but the intended function of its associated housing. Specifically, the staff believes that the housings for fans, dampers, and heating and cooling coils may perform a critical pressure retention and/or structural integrity function which, should that function not be maintained, could prevent the associated active mechanical component from performing its intended function. Further, if such housings perform these functions and meet the long-lived and passive screening criteria, then they should be subject to an AMR.

PROPOSED REVISIONS TO SRP-LR TABLE 2.1-5 Current Table Entries Item Category Structure, Component, or Structure, Component, or Commodity Group Commodity Group Meets 10 CFR 54.21 (a}(1 )(i) (Yes/No) 58 Heat Exchangers HVAC Coolers Yes 116 Valves Dampers No 124 Fans Ventilation Fans No ProposedR" eVlslons t 0 the T abl e Ent nes

                                        .

Item Category Structure, Component, or Structure, Component, or Commodity Group Commodity Group Meets 10 CFR 54.21 (a}(1 }(i) (Yes/No) 58 Heat HVAC Coolers Yes Exchangers (including housings) 116 Valves Housings for Dampers Yes (including louvers and gravity dampers) 124 Fans Housings for Ventilation Fans Yes (Including exhaust fans)

November 13, 2002 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 17761 Street, NW., Suite 400 1707 H Street, NW. Washington, DC 20006-3708 Suite 600 Washington, DC 20006-3919 SUB~'ECT: PROPOSED STAFF GUIDANCE ON THE SCOPING OF FIRE PROTECTION EQUIPMENT FOR LICENSE RENEWAL

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the enclosed guidance on the scoping of fire protection equipment for license renewal. This is consistent with our goal to more efficiently resolve license renewal issues identified by the staff or the industry, as outlined in NRR Office Letter No. 805, "License Renewal Application Review Process." Your response to this letter will assist the staff in deciding how to finalize and implement the guidance. During previous NRC scoping and screening inspections for license renewal, issues regarding the scoping and screening of fire protection equipment have arisen, indicating that additional guidance would be useful. Enclosure 1 provides guidance that was developed to clarify the requirements of 10 CFR 54.4(a)(3) as it pertains to 10 CFR 50.48 (including General Design Criterion 3, Appendix R, and associated license conditions). Proposed revisions to NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," are provided as Enclosure 2. A copy of Generic Letter 84-01 is provided for convenient reference (Enclosure 3). Although this proposed guidance does not convey a change in the NRC's regulations or how they are being interpreted, it is being provided to facilitate complete preparation of future applications for license renewal. As such, we are interested in receiving comments on the proposed guidance as well as an indication of when comments can be provided to ensure its timely release. The staff plans to incorporate this guidance into the improved renewal guidance documents (NUREG-1800 and/or NUREG-1801) in a future update. Additionally, comparable augmentation of NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule," might be warranted. Because this guidance provides a clarification of existing guidance with no additional requirements, the staff did not perform a backfit evaluation.

A. Nelson and D. Lochbaum -2 If you have any questions regarding this matter, please contact Rani Franovich at 301-415-1868. Sincerely, IRAJ Pao-Tsin Kuo, Program Director License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

As stated cc w/encls: See next page

PROPOSED STAFF POSITION ON THE LICENSE RENEWAL RULE (10 CFR 54.4) AS IT RELATES TO THE FIRE PROTECTION (FP) RULE (10 CFR 50.48) Staff Position Consistent with the requirements specified in 10 CFR 54.4(a)(3) and 10 CFR 50.48, all systems, structures, and components (SSCs) relied upon to perform a function that demonstrates compliance with the Commission's regulations for FP (10 CFR 50.48) are within the scope of license renewal. Consistent with General Design Criterion (GDC) 3, the scope of SSC's included in 10 CFR 50.48 goes beyond the protection of safety-related equipment. According to NUREG-0800, Section 9.5.1, "Fire Protection Program," the scope of equipment required for compliance with 10 CFR 50.48 also includes FP SSCs relied on to minimize the effects of a fire and to prevent the release of radiation to the environment. Components required to comply with 10 CFR 50, Appendix R, and with commitments to Appendix A to Branch Technical Position (BlP) APCSB 9.5-1, "Fire Protection For Nuclear Power Plants," or BTP CMEB 9.5-1, as documented in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," are within the scope of license renewal. Each nuclear station has a unique FP program, and the licensing basis for meeting FP requirements is plant-specific. In short, plant-specific licensing basis documents establish the basis for making FP scoping determinations. Rationale The License Renewal Rule, 10 CFR Part 54.4(a)(3), states that all plant systems, structures and components (SSCs) relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48) are within the scope of license renewal. The Statement of Considerations (SOC) for the license renewal rule, published in the May 8, 1995, edition of the Federal Register (60 FR 22461), states that 10 CFR 50.48(a) requires each nuclear power plant licensee to have in place a fire protection plan (FPP) that satisfies 10 CFR Part 50, Appendix A, GDC 3. The SOC further states, "the FPP establishes the fire protection policy for the protection of systems, structures, and components important to safety at each plant and the procedures, equipment, and personnel requirements necessary to implement the program at the plant site" (60 FR @ 22472). Regulatory Guide 1.189, "Fire Protection for Operating Nuclear Power Plants," provides the following definition of important to safety in its glossary: "nuclear power plant structures, systems, and components 'important to safety' are those required to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public." The scope and meaning of SSCs important to safety also are discussed in Generic Letter (GL) 84*01, "NRC Use of the Terms, 'Important to Safety' and 'Safety Related.'" The effects of fires on SSCs "important to safety" are addressed by 10 CFR 50.48 to provide a general level of protection that is afforded to all systems, not only those required for safe shutdown. The scope of SSCs required for compliance to GDC 3 and 10 CFR 50.48 goes beyond preserving the ability to achieve and maintain the plant in a safe shutdown condition in the event of a fire. In fact, NUREG-0800 states that the purpose of the FP program is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary safe shutdown functions and will not significantly increase the risk of radioactive releases to the environment, in accordance with GDC 3 and 5. Commitments to Enclosure 1

meet Appendix A to BTP APCSB 9.5-1 or BTP CMEB 9.5-1, as documented in Safety Evaluation Reports (SERs), which are directly referenced in the fire protection license condition, illustrate how a licensee complies with the regulations in 10 CFR 50.48. Each nuclear station has a unique FP program, and the licensing basis for meeting FP requirements is plant-specific. To determine the current licensing basis (CLB) for a nuclear power facility and perform an effective, complete scoping review for license renewal, an applicant should review applicable license renewal guidance and licensing basis documents. Documents that either specify FP requirements or define the CLB for FP include. but are not limited to, the following:

• The facility operating license and associated FP license conditions
• NRC SERs referenced in the FP license condition
• Applicable National Fire Protection Association (NFPA) codes (if commitments are made by the applicant to adopt NFPA code recommendations)
• Exemptions that may contain licensee commitments as they pertain to 10 CFR 50.48
• The most up-to-date fire hazards analysis (FHA)
• Design basis documents and specifications governing fire protection plans, systems and structures
• Technical Specifications (TS) and related operating commitments (e.g., those relocated from TS to the Updates Final Safety Analysis Report [UFSAR])
• UFSAR descriptions and drawings depicting systems and structures required for compliance with 10 CFR 50.48
• Code of Federal Regulations (Part 50 and Part 54) and associated sacs
• Appendix A to BTP APCSB 9.5-1, "Fire Protection For Nuclear Power Plants" or NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 9.5.1 [as referenced in 10 CFR 50.48 (b)(1)]
• Docketed correspondence [e.g., applicant commitments to Appendix A to BTP 9.5-1, NUREG-0800 exemption requests, etc.] pertaining to compliance with 10 CFR 50.48.

The staff should review the SERs or other licensing documents identified in the applicant's license condition that contain licensee commitments to 10 CFR 50.48. An applicant may sometimes exclude a particular component from the scope of license renewal on the basis that, although the component was discussed in the SER or FSAR (such as a fire protection jockey pump or a portion of an automatic sprinkler system), this does not constitute a "commitment" or imply that the component is required for compliance to 10 CFR 50.48. To determine if the exclusion of a component is valid, the applicant should review its response(s) to Appendix A to BTP 9.5-1 or to Section 9.5.1 of NUREG-0800 and other similar docketed correspondence that forms the basis of the SEA. If a particular component is provided for compliance with the approved FP program, as required by 10 CFR 50.48, then that particular component is relied upon to meet the requirements of 10 CFR 50.48 and should be included within the scope of license renewal. The exception to this involves changes to the FP program through a number of regulatory processes (e.g., 10 CFR 50.59, "Changes, Tests and Experiments"). Changes to the FP program can also occur through GL 86-10, "Implementation of Fire Protection Requirements," and GL 88-12, "Removal of Fire Protection Requirements From Technical Specifications." These changes typically involve license amendment requests to relocate TS governing FP operability and performance testing requirements to their UFSAA. Additionally, an applicant 2

may have relocated their FP program for meeting Appendix A to BTP 9.5-1 into their FHA or into some other licensing or design basis document. Applicants sometimes assume that commitments were documented in the UFSAR at the time a GL 86-10 license amendment was approved by the NRC and, for this reason, rely upon the UFSAR as their primary scoping document. However, information in the SERs that document an applicant's response to Appendix A to BTP 9.5-1 is not always documented in the UFSARs. Therefore, applicants for license renewal should review all documents that define their licensing basis for meeting fire protection requirements in performing scoping reviews for their license renewal applications. Backfit consideration The staff has determined that this guidance clarifies the staff's guidance on scoping of FP structures and components. Therefore, the staff did not evaluate this ISG for backfit. 3

For under 10 CFR 54.4(a)(3), are required to be included within the scope of the rule. For example, if a nonsafety-related diesel generator is required for safe shutdown under the fire protection plan, the diesel generator and all SSCs specifically required for that generator to comply with and NRC regulations shall be included within the scope of license renewal under 10 CFR 54.4(a)(3). Such SSCs may include, but should not be limited to, the cooling water system or systems reqUired for operability, the diesel support pedestal, and any applicable power supply cable specifically required for safe shutdown in the event of a fire. In addition, the last sentence of the second paragraph in Section III.c(iii) of the SOC provides the following guidance for limiting the application of the scoping criteria under 10 CFR 54.4(a}(3} as it applies to the use of hypothetical failures: Consideration of hypothetical failures that could result from system Interdependencies, that are not part of the current licensing bases and that have not been previously experienced is not required. (60 FR 22467) The SOC does not provide any additional guidance relating to the use of hypothetical failures or the need to consider second-, third-, or fourth-level support systems for scoping under 10 CFR 54.4(a}(3}. Therefore, in the absence of any guidance, an applicant need not consider hypothetical failures or second-, third-, or fourth-level support systems in determining the SSCs within the scope of the rule under 10 CFR 54.4(a}(3}. For example, if a nonsafety-related diesel generator is relied upon only to remain functional to demonstrate compliance with the NRC SSO regulations, the applicant need not consider the following SSCs: (1) an altemate/backup cooling water system, (2) non-seismically-qualified building walls, or (3) an overhead segment of nonseismically-quaJified piping (in a Seismic 11/1 configuration). This guidance is not intended to exclude any support system (whether identified by an applicant'S CLB, or as indicated from actual plant-specific experience, industrywide experience [as applicable]. safety analyses. or plant evaluations) that is specifically required for compliance with, the applicable NRC regulation. For example. if a nonsafety-related diesel generator (required to demonstrate compliance with an applicable NRC regulation) specifically requires a second cooling system to cool the diesel generator jacket water cooling system for the generator to be operable, then both cooling systems must be included within the scope of the rule under 10 CFR 54.4(a)(3). The applicant is required to identify the SSCs whose functions are relied on to demonstrate compliance with the regulations identified in 10 CFR 54.4(a}(3} (that Is, whose functions were credited in the analysis or evaluation). Mere mention of an SSC in the analysis or evaluation does not necessarily constitute support of an intended function as required by the regulation. For environmental qualification, the reviewer verifies that the applicant has indicated that the environmental qualification equipment is that eqUipment already identified by the licensee under 10 CFR 50.49(b), that is, equipment relied upon in safety analyses or plant evaluations to demonstrate compliance with NRC regUlations for environmental qualification (10 CFR 50.49). The PTS regulation is applicable only to PWRs. If the renewal application is for a PWR and the applicant relies on a RegUlatory Guide 1.154 (Ref. 5) analysis to satisfy 10 CFR 50.61, as described in the plant's CLB, the reviewer verifies that the applicant's methodology would include SSCs relied on in that analysis that are within the scope of license renewal. For SBO, the reviewer verifies that the applicant's methodology would include those SSCs relied upon during the "coping duration" phase of an SBO event (Ref. 6). NUREG-1800 2.1-9 April 2001 Enclosure 2

--~-~"'tf   - ....-..,.--..:-""'..--.-.......r- ..... - r ..- - ., p, ... _,.....:--~.~' .... --*I""': ~'.....*fl'_:-~.        i ' i. . . ~-f_*~~*~_'_"'_--'  -"-'*-*

For fire protection, the reviewer verifies ;that the applicant's"methodo'6~v would include those SSCs relied upon to meetthe reQuirements of 10 CFR 50.48 (Reference t6 ISG)J 'Potential information - '. sources

                     ,-

that should, be'reviewed to

                                        , . _ . ...                                      , -

determins'an'*appJica*iit's

                                                                                                              ... -~"'_ .... -           ." Iicensin~'

_.. basis .for ~ ~eetin~ th~ re.Quir~m~nts,~J 10 ~F:R 50AB.are provide.d inJabJe,2.1~2;,Specif!c.StafCGuidance Ol"!.Scoping (Issue: fire. protection). 2.1.3.2 Screening Once the 5SCs within the scope of license renewal have been identified, the next step is determining which structures and components are subject to an AMR (i.e., "screening") (Ref. 1). 2.1.3.2.1 "Passive" The reviewer reviews the applicant's methodology to ensure that "passive" structures and components are identified as those that perform their intended functions without moving parts or a change in configuration or properties in accordance with 10 CFR 54.21 (a)(1 )(i). The description of "passive" may also be interpreted to include structures and components that do not display "a change in state." 10 CFR 54.21 (a)(1 )(i) provides specific examples of structures and components that do or do not meet the criterion. The reviewer verifies that the applicant's screening methodology includes consideration of the intended functions of structures and components consistent with plant CLB, as typified in Table 2.1-4 (Ref. 1). The license renewal rule focuses on "passive" structures and components because structures and components that have passive functions generally do not have performance and condition characteristics that are as readily observable as those that perform active functions. "Passive" structures and components, for the purpose of the license renewal rule, are those that perform an intended function, as described in 10 FR 54.4, without moving parts or without a change in configuration or properties (Ref. 2). The description of "passive" may also be interpreted to include structures and components that do not display "a change of state." Table 2.1-5 provides a list of typical structures and components identifying whether they meet 10 CFR 54.21 (a)(1)(i). 10 CFR 54.21 (a)(1 )(i) explicitly excludes instrumentation, such as pressure transmitters, pressure indicators, and water level indicators, from an AMR. The applicant does not have to identify pressure-retaining boundaries of this instrumentation because 10 CFR 54.21 (a)(1 )(i) excludes this instrumentation without exception, unlike pumps and valves. Further, instrumentation is sensitive equipment and degradation of its pressure retaining boundary would be readily determinable by surveillance and testing (Ref.6). If an applicant determines that certain structures and components listed in Table 2.1-5 as meeting 10 CFR 54.21 (a)(1 lei) do not meet that requirement for its plant, the reviewer reviews the applicant's basis for that determination. 2.1.3.2.2 "Long-Lived" The applicant's methodology is reviewed to ensure that "long-lived" structures and components are identified as those that are not SUbject to periodic replacement based on a qualified life or specified time period. Passive structures and components that are not replaced on the basis of a qualified life or specified time period require an AMR. NUREG-1800 2.1-10 April 2001

December 3, 2001 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW Washington, DC 20006-3708 Suite 600 Washington, DC 20006-3919

SUBJECT:

LICENSE RENEWAL ISSUE: SCOPING OF SEISMIC 11/1 PIPING SYSTEMS

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the enclosed guidance clarifying the scoping of seismic 11/1 piping systems. This is consistent with our goal to more efficiently resolve license renewal issues identified by the staff or the industry, as outlined in NRR Office Letter No. 805, "License Renewal Application Review Process." Based on your response to this letter, the staff will decide how to finalize and implement this guidance. The staff developed this guidance to ensure that scoping of seismic II1I piping systems, including piping and supports, is conducted in accordance with the requirements of 10 CFR 54.4(a)(2). We are requesting comments on the proposed guidance and request that you submit comments within 30 days following the date of this letter to ensure a timely resolution of this issue. The staff plans to incorporate this position into the improved renewal guidance documents (NUREGs 1800, andlor 1801) in a future update. It is also possible that comparable changes might be needed to NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule." If you have any questions regarding this matter, please contact Hai-Boh Wang at 301-415-2958, Christian Araguas at 301-415-3936, or William Burton at 301-415-2853. Sincerely, IRAI Christopher I. Grimes, Chief License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project 690

Enclosure:

As stated cc w/encl: See next page Mr. Alan Nelson Washington, DC 20006-3708 Nuclear Energy Institute 1776 I Street, NW., Suite 400

STAFF POSITION ON SCOPING OF SEISMIC II/I PIPING SYSTEMS

1. BACKGROUND Section 54.29 of 10 CFR Part 54 (the Rule) states that a renewed license may be issued by the Commission if the Commission finds that actions have been or will be taken with respect to the matters identified in paragraphs (a)(1) and (a)(2) of this section such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the Current Licensing Basis (CLB), and that any changes made to the CLB in order to comply with this paragraph are in accord with the Act and the Commission's regulations. These matters include managing the effects of aging during the period of extended operation to assure the functionality of structures and components that have been identified to require review under Section 54.21(a)(1).

The Statements of Consideration (SOC) for the Rule states that the objective of a license renewal review is to determine whether the detrimental effects of aging, which could adversely affect the functionality of systems. structures, and components (SSCs) that the Commission determines require review for the period of extended operation. are adequately managed. The SOC articulates the underlying philosophy of the Rule that during the extended period of operation, safety-related functions should be maintained in the same manner and to the same extent as during the current licensing term. Aging effects that could adversely impact on the ability of SSCs to maintain these safety-related functions during the extended period of operation should be evaluated. Section 54.4(a)(2) of the Rule states that all non-safety related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in Section 54.4(a)(1) should be included within the scope of the Rule. The SOC provides additional guidance related to this scoping criterion. Specifically, the SOC states that "To limit this possibility for the scoping category relating to nonsafety-related systems, structures. and components... An applicant for license renewal should rely on the plant's CLB, actual plant-specific experience, industry-wide operating experience, as appropriate, and existing engineering evaluations to determine those nonsafety-related systems, structures, and components that are the initial focus of the license renewal review. Consideration of hypothetical failures that could result from system interdependencies that are not part of to CLB and that have not been previously experienced is not required." (Federal Register, Volume 60, No. 88,22467).

2. DISCUSSION A subset of non-safety-related piping systems that meet the 54.4(a)(2) criterion is seismic II over I (seismic II/I) piping. Seismic II/I denotes non-seismic Category I SSCs interacting with seismic Category I SSCs as described in Position C.2 of Regulatory Guide 1.29, "Seismic Design Classification." The SOC specifically includes seismic 11/1 components as a subset of the 54.4(a)(2) scoping requirement. In addition, Section 2.1.III.B of the Standard Review Plan for License Renewal (September, 1997) states that "The reviewer verifies that the so-called

'seismic II over I' systems, structures, and components consistent with the plant's CLB are identified by the applicant's proposed screening methodology." Enclosure

                                                 -2

During the review of several license renewal applications, the staff has found that applicants have sometimes misinterpreted the requirements for scoping criterion 54.4(a)(2), particularly as it related to seismic 1111 piping. Specifically, some applicants have only considered seismicity in determining whether seismic 1111 piping should be included within the scope of license renewal. Several applicants have concluded that only the seismically-designed pipe supports for seismic 11/1 piping need to be included within the scope of license renewal. This conclusion is based on having the seismic 1111 piping seismically supported, and operating experience that has shown that seismically supported piping at nuclear plants, whether new or old, has not fallen during a seismic event. Applicants therefore conclude that consideration of seismic 11/1 pipe failures is hypothetical. The staffs concern is that seismic 1111 piping, though seismically supported, would be subjected to the same plausible aging effects as safety-related piping. For example, depending on piping material, geometrical configuration, operating condition such as water chemistry, temperature, flow velocity, and external environment, erosion and corrosion may be plausible aging effects for some seismic 11/1 piping. Those effects, if not properly managed, could result in age-related failures and adversely impact the safety functions of safety-related SSCs. As mentioned previously, applicants consider failure of pipe segments during a seismic event to be hypothetical for the reasons stated above. The staff agrees that operating experience shows that seismically-supported piping has not fallen during a seismic event, and agree that, on this basis, seismic supports should be included within the scope of license renewal. However, the staff maintains that industry operating experience has also shown that piping has failed for reasons other than a seismic event. For example, numerous erosion and corrosion related pipe wall thinning issues, and pipe failure events are documented in a number of generic communications, and are summarized in Information Notice (IN) 2001-09, "Main Feedwater System Degradation in Safety-Related ASME Code Class 2 Piping Inside the Containment of a Pressurized Water Reactor." The operating experience referenced in this IN shows that piping has failed for reasons other than seismic events. On this basis, the staff concludes that pipe failures due to age-related degradation are not hypothetical, and therefore, both seismic 11/1 piping segments and their supports should be included within the scope of license renewal.

3. CONCLUSION On the basis of the above discussion, the staff concludes that seismic 1111 piping systems, including both the piping segments and supports, should be included within the scope of license renewal. By including these components within scope, age-related degradation of these components can be evaluated and, if appropriate, adequately managed to ensure that intended functions can be maintained during the extended period of operation.

This staff position applies both to applicants who have committed in their CLB to follow the guidance of RG 1.29, as well as to applicants whose plants include piping segments whose failure could impact safety-related SSCs, whether due to a seismic event, or due to other reasons, as discussed above.

• March 15, 2002 Mr. Alan Nelson Mr. David Lochbaum Nuclear Energy Institute Union of Concerned Scientists 1776 I Street, NW., Suite 400 1707 H Street, NW Washington, DC 20006-3708 Suite 600 Washington, DC 20006-3919 SUB..IECT: LICENSE RENEWAL ISSUE: GUIDANCE ON THE IDENTIFICATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS WHICH MEET 10 CFR 54.4(a)(2)

Dear Messrs. Nelson and Lochbaum:

The purpose of this letter is to provide you with the opportunity to comment on the enclosed guidance on the identification and treatment of structures, systems, and components which meet the scoping criterion specified in 10 CFR 54.4(a)(2). This is consistent with our goal to more efficiently resolve license renewal issues identified by the staff or the industry, as outlined in NRR Office Letter No. 805, "License Renewal Application Review Process." Based on your response to this letter, the staff will decide how to finalize and implement this guidance. The staff developed this guidance to ensure that scoping of non-safety-related structures, systems, and components, is conducted in accordance with the requirements of 10 CFR 54.4(a)(2), and is intended to supplement the position on the identification and treatment of seismic 11/1 components we provided to you by letter dated December 3,2001. We are requesting comments on the proposed guidance and request that you provide us with a schedule for a timely resolution of this issue. The staff plans to incorporate this position into the improved renewal guidance documents (NUREGs 1800, andlor 1801) in a future update. It is also possible that comparable changes might be needed to NEI 95-10, Revision 3, "Industry Guidance for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule." If you have any questions regarding this matter, please contact Hai-Boh Wang at 301-415 2958, or William Burton at 301-415-2853. Sincerely, IRAI Christopher I. Grimes, Program Director License Renewal and Environmental Impacts Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project 690

Enclosure:

As stated cc w/encl: See next page

STAFF POSITION ON 54.4(8)(2) SCOPING CRITERION

1. BACKGROUND Section 54.29 of 10 CFR Part 54 (the Rule) states that a renewed license may be issued by the Commission if the Commission finds that actions have been or will be taken with respect to the matters identified in 54.29(a)(1) and (a)(2) such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the CLB, and that any changes made to the CLB in order to comply with this paragraph are in accord with the Atomic Energy Act and the Commission's regulations. These matters include managing the effects of aging during the period of extended operation to assure the functionality of structures and components that have been identified to require review under Section 54.21(a)(1).

The Statements of Consideration (SOC) for the Rule state that the objective of a license renewal review is to determine whether the detrimental effects of aging, which could adversely affect the functionality of systems, structures, and components (SSCs) that the Commission determines require review for the period of extended operation, are adequately managed. Section 54.4(a)(2) of the Rule states that all non-safety related SSCs whose failure could prevent satisfactory accomplishment of any of the functions identified in Section 54.4(a)(1) should be included within the scope of the Rule. The SOC provides additional guidance related to this scoping criterion. Specifically, the SOC states that "To limit this possibility for the scoping category relating to non safety-related systems, structures, and components... An applicant for license renewal should rely on the plant's CLB, actual plant-specific experience, industry-wide operating experience, as appropriate, and existing engineering evaluations to determine those non safety-related systems, structures, and components that are the initial focus of the license renewal review. Consideration of hypothetical failures that could result from system interdependencies that are not part of to CLB and that have not been previously experienced is not required." (Federal Register, Volume 60, No. 88, 22467).

2. DISCUSSION The SOC articulates the underlying philosophy of the Rule; that during the period of extended operation, safety-related functions should be maintained in the same manner and to the same extent as during the current license term.

The staff must have reasonable assurance that the applicant has identified all non safety related SSCs that meet the 54.4(a)(2) scoping criterion. To accomplish this, the applicant should clearly describe the methodology used to determine those non safety-related SSCs that meet this criterion. This description should include how plant-specific failures of non safety related SSCs and industry failures of such SSCs were considered in this determination, and should identify whether consideration was given to non safety-related SSCs which may not have failed during the current term, but may have a reasonable expectation of failure during the extended term. Such consideration should be based on sound engineering judgement that assures the failure of those non safety-related SSCs would not occur during the extended period of operation. Information which formed the basis for the applicant's conclusions need

• .1

                                                      -2 not be included in the application, but should be documented, auditable, and retrievable, in accordance with 10 CFR 54.37.

When demonstrating that failures of non safety-related SSCs would not adversely impact on the ability to maintain intended functions, a distinction must be made between non safety-related SSCs that are connected to safety-related SSCs and those that are not connected to safety related SSCs. For a non safety-related SSC that is connected to a safety-related SSC, the non safety-related SSC should be included within the scope of license renewal up to the first seismic anchor past the safety/non-safety interface. Further, if the in-scope non safety-related structure or component is of the same commodity group (i.e., the same material/environment combination) as the safety-related structure or component to which it's connected, the same aging management programs should be applied to both the safety-related and non safety related structures and components. If the in-scope non safety-related structure or component is not of the same commodity group, then aging management programs appropriate for the commodity should be applied. For non safety-related SSCs which are not connected to safety-related piping or components or are beyond the first seismic anchor past the safety/non-safety interface, but have a spatial relationship such that their failure could adversely impact on the performance of a safety related SSC's intended function, the applicant has two options when performing its scoping evaluation; a mitigative option or a preventive option. With the mitigative option, the applicant should demonstrate that plant mitigative features (e.g., pipe whip restraints, jet impingement shields, spray and drip shields, seismic supports, flood barriers) are provided which protect safety-related SSCs from failures of non safety-related SSCs. This demonstration should show that the mitigating devices are adequate to protect safety-related SSCs from failures of non safety-related SSCs regardless of failure location (consideration can be given to the likelihood of failure at a particular location based on sound engineering judgement). If this level of protection can be demonstrated, then only the mitigative features need to be included within the scope of license renewal. However, if an applicant cannot demonstrate that the mitigative features are adequate to protect safety-related SSCs from the consequences of failures of non safety-related SSC's, then the applicant should utilize the preventive option, which requires that the entire non safety-related SSC be brought into the scope of license renewal. An applicant may determine that, in order to ensure adequate protection of the safety-related SSC, a combination of mitigative features and non safety-related SSCs must be brought within scope. Again, it is incumbent upon the applicant to provide adequate justification for the approach taken with respect to scoping of non safety-related SSCs in accordance with the Rule. To ensure that all relevant non safety-related SSCs are captured within the scope of the Rule, an applicant should consider not only its CLB, but also plant and industry operating experience. Operating experience includes all documented plant-specific and industry-wide experience which can be used to determine the plausibility of a failure. Documentation would include NRC generic communications and event reports, plant-specific condition reports, industry reports such as SOERs, and engineering evaluations.

3. CONCLUSION On the basis of the guidance provided in the SOC, the staff expects applicants for license renewal to identify non safety-related SSCs whose failure could adversely impact intended

1 ' functions. Such SSCs are to be included within the scope of license renewal. The evaluation to determine which non safety-related SSCs are within scope should not consider hypothetical failures, but should, based on engineering judgement and operating experience, consider the likelihood of system failure during the extended period of operation. The information used to support the scoping determination should be documented and available for staff review.}}