ML043280683

Requests for Additional Information for Review of Nine Mile Point Nuclear Station, Units 1 and 2, License Renewal Application

ML043280683
Person / Time
Site:	Nine Mile Point
Issue date:	11/22/2004
From:	Le N NRC/NRR/DRIP/RLEP
To:	Spina J Nine Mile Point
Le N. B., NRR/DRIP/RLEP, 415-1458
References
TAC MC3272, TAC MC3273
Download: ML043280683 (13)

Text

November 22, 2004 Mr. James A. Spina Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TACS MC3272 AND MC3273)

Dear Mr. Spina:

By letter dated May 26, 2004, Constellation Energy Group Inc., submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew the operating licenses for the Nine Mile Point Nuclear Station (NMP), Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC). The NRC staff is reviewing the information contained in the license renewal application (LRA) and has identified, in the enclosure, areas where additional information is needed to complete the review.

Based on discussions with Mr. Peter Mazzaferro of your staff, a mutually agreeable date for your response is within 30 days from the date of this letter. If you have any questions regarding this letter or if circumstances result in your need to revise the response date, please contact me at 301-415-1458 or by e-mail at nbl@nrc.gov.

Sincerely,

/RA/

N. B. (Tommy) Le, Senior Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket Nos.: 50-220 and 50-410

Enclosure:

As stated cc w/encl: See next page

ML043280683 DISTRIBUTION: See next page Document Name: E:\\Filenet\\ML043280683.wpd OFFICE RLEP:PM RLEP:LA RLEP:SC-A NAME N. Le Via Template S. Lee DATE 11/22/04 11/22/04 11/22/04

DISTRIBUTION: Ltr. To J. Spina, Re: RAI for the Review of Nine Mile Nuclear Station, Units 1 and 2, License Renewal Application (TAC NOS. MC3272 and MC3273), Dated: November 22, 2004 Adams accession no.: ML043280683 HARD COPY RLEP RF N. Le (PM)

E-MAIL:

RidsNrrDrip RidsNrrDe G. Bagchi K. Manoly W. Bateman J. Calvo R. Jenkins P. Shemanski PY Cheng RidsNrrDssa RidsNrrDipm D. Thatcher R. Pettis G. Galletti C. Li M. Itzkowitz (RidsOgcMailCenter)

T. Smith M. Lemoncelli M. Mayfield A. Murphy S. Smith (srs3)

S. Duraiswamy Y. L. (Renee) Li RLEP Staff C. Holden R. Laufer P. Tam B. Fuller, RI E. Knutson, RI J. Trapp, RI T. Mensah OPA Attachment NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 (NMP1, NMP2)

LICENSE RENEWAL APPLICATION (LRA)

REQUEST FOR ADDITIONAL INFORMATION (RAI)

RELATED TO:

Scoping and Screening Audit September 27 - October 1, 2004 RA 2.1 - Review Methodology for Non-Accident Design Basis Events 10 CFR 54.4(a)(1), states, in part, that safety systems, structures, and components (SSCs) within the scope of license renewal include safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49(b)(1)). 10 CFR 50.49, states that design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed. In regard to identification of design basis events, NUREG-1800, Section 2.1.3, Review Procedures, states:

The set of design basis events as defined in the rule is not limited to Chapter 15 (or equivalent) of the UFSAR. Examples of design basis events that may not be described in this chapter include external events, such as floods, storms, earthquakes, tornadoes, or hurricanes, and internal events, such as a high-energy-line break. Information regarding design basis events as defined in 10 CFR 50.49(b)(1) may be found in any chapter of the facility UFSAR, the Commissions regulations, NRC orders, exemptions, or license conditions within the CLB. These sources should also be reviewed to identify systems, structures and components that are relied upon to remain functional during and following design basis events (as defined in 10 CFR 50.49(b)(1)) to ensure the functions described in 10 CFR 54.4(a)(1).

During the scoping and screening methodology audit, the Nuclear Regulatory Commission (NRC) audit team questioned how non-accident design basis events, particularly design basis events that may not be described in the UFSAR, were considered during scoping. The NRC audit team noted that limiting the review of design bases events to those described in the UFSAR accident analysis could result in omission of safety-related functions described in the current licensing basis.

The staff therefore, requests the applicant to provide the following additional information: a list of the design basis events evaluated as part of the license renewal scoping process, a description of the methodology used to ensure that all design bases events (including conditions of normal operation, anticipated operational occurrences, design basis accidents, external events, and natural phenomena) were addressed during license renewal scoping. For each response, please indicate the documentation sources reviewed to ensure that all design basis events were identified.

For each of the above issues, if the response indicates that additional scoping evaluations are required, please describe these additional scoping evaluations performed to address the 10 CFR 54.4(a)(1) criteria. As applicable, please list any additional SSCs to be included within scope of the license renewal efforts, and list those structures and components for which aging management reviews were conducted. For each additional SC, please describe the aging management programs, as applicable, to be credited for managing the identified aging effects.

RAI 2.1 10 CFR 54.4(a)(1) Scoping of Safety-Related SSCs 10 CFR 54.4(a)(1)(iii) requires, in part, that the applicant considers within the scope of license renewal those systems, structures, and components that ensure the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in §50.34(a)(1), §50.67(b)(2), or §100.11. In Section 2.1.1.1, Application of Safety-Related Scoping Criteria, of the LRA, the applicant stated that, because of plant-unique considerations or preferences, some components classified as safety-related in the facility database did not perform any of the safety-related intended functions of 10 CFR 54.4(a)(1). The applicant further stated that these components may have been considered outside the scope of 10 CFR 54.4(a)(1). During the audit, the applicant described the process used to evaluate components classified as safety-related that did not perform a safety-related intended function. As part of the process, the applicant stated that the safety-classification of many safety-related components was re-evaluated in order to reconcile differences between scoping determinations and facility database information. Based on the audit, the staff requests a description of the process used during license renewal scoping activities to disposition components classified as safety-related that do not perform a safety-related intended function. In particular, the staff requests the applicant provide the following information:

a.

A description of any components or structures classified as safety-related in the facility safety-classification database that were not included within the scope of license renewal under the 10 CFR 54.4(a)(1) criteria. This description should include the basis for determining that these components do not perform a safety-related intended function. The response should also indicate if these components were included within the scope of license renewal under a different scoping criteria (e.g. §54.4(a)(2) or (a)(3)).

b.

Describe the process used to reconcile the facility database safety classification information with scoping intended function determinations. The response should contain a description of the process including the scope of the review used to re-evaluate the safety-classification of SSCs to reconcile disparities with intended function determinations.

RAI 2.1 Scope of Current Licensing Basis (CLB) Document Review for Identification of System Descriptions and System Intended Functions 10 CFR 54.21(a)(3) requires, in part, that the integrated plant assessment contained in the license renewal application demonstrate that the effects of aging will be adequately managed so that the intended function(s) of systems, structures and components within the scope of 10 CFR 54 will be maintained consistent with the current licensing basis for the period of extended operation. 10 CFR 54.3(a) states that the current licensing basis is the set of NRC requirements applicable to a specific plant and a licensees written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis that are docketed and in effect.

10 CFR 54.3(a) further states that the CLB includes certain NRC regulations; orders; license conditions; exemptions; technical specifications; design basis information documented in the most recent final safety evaluation report; and licensee commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports.

As part of the LRA review, the NRC staff evaluates the scope and depth of the applicants document review to provide assurance that the scoping methodology considered all SSC intended functions. In reviewing the LRA and scoping and screening implementation procedures, the NRC staff was unable to determine the extent that the CLB was reviewed during the development of the system description and intended function evaluations performed during the scoping phase of the review. With regard to the development of system descriptions and identification of system intended function, the staff requests the applicant provide the following information:

a.

Describe the methodology used to develop system descriptions and identify the system intended functions. Please state in the response which CLB source documents were used for these activities.

b.

In discussions with the Nine Mile Point license renewal project team during the scoping and screening methodology audit, it was identified that an electronic document database was used to identify CLB documents pertinent to the development of system descriptions and identification of system intended functions. Please describe the controls and processes, including any proceduralized controls, that were used to ensure that the electronic current licensing basis document database was complete and accurate.

RAI 2.1-4 - 10 CFR 54.4(a)(2) Scoping Criteria for Non Safety-Related SSCs By letters dated December 3, 2001, and March 15, 2002, the NRC issued a staff position to the Nuclear Energy Institute (NEI) which described areas to be considered and options it expects licensees to use to determine what SSCs meet the 10 CFR 54.4(a)(2) criterion (i.e., All non safety-related SSCs whose failure could prevent satisfactory accomplishment of any safety-related functions identified in paragraphs (a)(1)(i),(ii),(iii) of this section.) The NRC December 3rd letter provided specific examples of operating experience which identified pipe failure events (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") and the approaches that the NRC considers acceptable to determine which piping systems should be included in scope based on the 54.4(a)(2) criterion.

The NRC March 15th letter further described the staffs expectations for the evaluation of non-piping SSCs to determine which additional non safety-related SSCs are within scope. The staff position states that applicants should not consider hypothetical failures, but rather should base their evaluation on the plants CLB, engineering judgement and analyses, and relevant operating experience. The March 15th letter further describes operating experience as all documented plant-specific and industry-wide experience which can be used to determine the plausibility of a failure.

Operating experience documentation sources would include NRC generic communications and event reports, plant-specific condition reports, industry reports such as SOERs, and engineering evaluations.

Based on a review of the LRA, the applicants scoping and screening implementation procedures, and discussions with the applicant, the staff determined that additional information is required with respect to certain aspects of the applicants evaluation of the 10 CFR 54.4(a)(2) criteria. The staff requests the applicant provide the following information:

a.

Non Safety-Related (NSR) Piping Attached to Safety-Related (SR) Piping Section 2.1.4.2.3, "SR/NSR Piping Interface," of the LRA states that the for NSR piping containing water or steam [attached to SR piping], the NSR portion within the scope of license renewal extends beyond the depicted class change [SR to NSR] until no longer in the vicinity of SR equipment [defined in the LRA as within the same building, corridor or floor] or until the first seismic anchor is reached, whichever is the furthest.

During the audit, the NRC audit team noted that in some cases where non safety-related plant equipment provided a termination point for non safety-related piping that was attached to safety-related piping. In these cases, the non safety-related piping was placed within the scope of license renewal, but the plant equipment (such as a heat exchanger) was not considered to be within scope.

For cases where an entire pipe run including both safety-related and non safety-related piping was analyzed as part of the current licensing basis to establish that it could withstand design basis event loads, NUREG-1800, Section 2.1.3.1.2 indicates that the scoping methodology includes: (1) the non safety-related piping up to its anchors and (2) the associated piping anchors as being within the scope of license renewal under 10 CFR 54.4(a)(2). Because in some instances plant equipment was used as a termination point for the non safety-related piping within the scope of license renewal, this plant equipment appears to be equivalent to an associated piping anchor as described in NUREG-1800.

The staff therefore, requests that the applicant provide additional information regarding their 10 CR 54.4(a)(2) evaluation to include:

1. The definition of equivalent anchor consistent with the plant CLB which was used for the purposes of the 10 CFR 54.4(a)(2) evaluation;

2. A description of how the first seismic anchor was identified for NSR pipe attached to SR pipe, within the scope of license renewal. Specifically include a discussion of how the scoping process identified the first seismic anchor, and established the license renewal boundaries which extended beyond the vicinity of SR equipment [beyond the building, corridor or room];

3. Confirmation that for the NSR piping, associated plant equipment, and their supports, up to and including the first seismic anchor, were within the scope of license renewal and subject to aging management review.

4. A discussion of how plant equipment identified as the termination point for non safety-related piping was evaluated during the scoping process. As part of that discussion, if the plant equipment provides a structural support function to the non safety-related piping, please identify whether that plant equipment is within scope of renewal and subject to an AMR. If it is not provide a justification for not including this plant equipment within the scope of license renewal.

b.

NSR SSCs Which Functionally Interact With SR SSCs LRG-02, "License Renewal Scoping and Screening," paragraph 3.4.3.2, states that malfunctions of NSR equipment that result in a challenge to SR equipment

([where] the SR function is maintained) is not in scope. The staff requests that the applicant provide the basis for this position and all applications of this position during the scoping process.

c.

Fail-Safe Components LRG-02, "License Renewal Scoping and Screening," paragraph 4.1.2, states that fail-safe components are components whose failure (through interaction with the failed NSR SSC) cannot prevent the accomplishment of a safety-related function since the NSR SSC causes the SR SSC to attain a fail-safe state. The staff requests that the applicant provide the basis for this position and all applications of this position during the scoping process.

In addressing each of the above issues, if the response indicates the use of the scoping methodology to screen out any non safety-related SSCs that could spatially interact with safety-related SSCs, please describe any additional scoping evaluations performed to address the 10 CFR 54.4(a)(2) criteria. As part of the response, please list any additional SSCs included within scope as a result of your efforts, and list those SCs for which aging management reviews were conducted. For each SC also describe the aging management programs, as applicable, to be credited for managing the identified aging effects.

RAI 2.1 Development and Use of Technical Position Papers NUREG-1800, Section 2.1.3.1.3, Regulated Events, states that all SSCs that are relied upon in the plants CLB (as defined in 10 CFR 54.3), plant-specific experience, industry-wide experience (as appropriate), and safety analyses or plant evaluations to perform a function that demonstrates compliance with NRC regulations identified under 10 CFR 54.4(a)(3), are required to be included within the scope of the rule. As part of the LRA review, the NRC staff evaluates the scope and depth of the applicants document review to provide assurance that the scoping methodology considered all SSC intended functions.

During the NRCs scoping and screening methodology audit, the applicant identified several Technical Position Papers as a documentation source for license renewal scoping under 10 CFR 54.4(a)(3). In reviewing the LRA, scoping and screening implementation procedures, and evaluation of the Feedwater System during the audit, the NRC audit team was informed by the applicant that two technical position papers (anticipated transients without scram (ATWS) and station blackout (SBO)) had not been adequately reviewed and incorporated into the LRA during the LRA verification activities.

This discrepancy was identified by the applicant during the audit and documented in a Deviation Engineering Report DER-NM-2044, dated September 30, 2004. DER-NM-2044 states that Section 2.3.4.B.3 of the LRA is incomplete because it does not reference an SBO event in the description of why components in the Unit 2 Feedwater system are in the scope of LR. DER-NM-2044 also states that an extent of condition review is necessary to determine if there are similar instances effecting other system descriptions in the LRA.

Based on the above discrepancy, the staff requests the applicant provide the following information:

a.

Describe the methodology used to develop technical position papers. In the response, please state which CLB source documents were used to develop the position papers.

b.

Describe the actions taken to ensure that both the NMP Unit 1 and the Unit 2 LR Scoping and Screening Reports adequately address the new ATWS and SBO design basis documents, as well as any potentially affected LRA sections.

c.

DER-NM-2044 states that Section 2.3.4.B.3 of the LRA is incomplete because it does not reference an SBO event in the description of why components in the Unit 2 feedwater system are in the scope of LR. Describe the actions taken to determine if there are similar instances for other system descriptions in the LRA.

In addressing each of the above issues, if the response indicates that failure to incorporate the information in the Technical Position Papers has resulted in the omission of relevant information in the LRA or omission of SSCs from within scope, please describe any additional scoping evaluations performed to address is concern. As part of the response, list any additional SSCs included within scope as a result of the review efforts, and list those SCs for which aging management reviews were conducted.

For each SC describe the aging management programs, as applicable, to be credited for managing the identified aging effects.

RAI 2.1 Implementation of the Electrical Spaces Scoping Approach NUREG-1800, Section 2.5.3.1, Components Within the Scope of License Renewal, states that an applicant may use the plant spaces approach in scoping electrical and instrumentation and control (I&C) components. In the plant spaces approach, an applicant may indicate that all electrical and I&C components located within a particular area are either within or not within the scope of license renewal. NUREG-1800, Table 2.5-1, Examples of Plant Spaces Approach for Electrical and I&C Scoping and Corresponding Review Procedures, provides guidance for the review of scoping performed in accordance with the plant spaces approach. If the applicant limits the scope of electrical and I&C components considered within the scope of license renewal by excluding components in certain plant spaces, Table 2.5-1 indicates that this approach should not result in failing to place electrical and I&C components that perform intended functions within the scope of license renewal.

During the audit, the staff reviewed the applicants methodology for scoping and screening of electrical and I&C components. The staff found that the procedures related to electrical and I&C scoping and screening lacked sufficient detail to determine if the applicants methodology was adequate for scoping and screening of electrical and I&C components. Specifically, Procedure LRG-02, Scoping and Screening, Section 3.8.2.1 requires that system components be identified and incorporated into the LRA Electronic Database (CONRAD). LRG-02 Section 3.9.2, further states that all electrical components that are long-lived and passive are addressed as commodities and therefore no electrical systems encompassed by the LRG-02 guidelines are considered to contain electrical components subject to an AMR. LRG-04, Aging Management Review of Electrical Commodities, Section 3.1.1.1.1, states that the reviewer should utilize available documented and electronic resources to generate a preliminary list of items in the commodity group. In practice, once the system intended functions were identified the applicant developed a commodity list of electrical commodities which were then screened and evaluated in the aging management review phase. The staff was unable to determine the specific activities which were performed by the applicants staff to identify the applicable intended functions, plant electrical equipment required to perform those functions, and subsequent development of the electrical commodity list from which the aging management reviews were conducted.

As a result of the above review, the staff requests the applicant provide a detailed description of the methodology used for the scoping and screening of electrical and I&C components to include:

a.

How electrical system boundaries and individual electrical components within those boundaries were identified and incorporated into CONRAD.

b.

Describe the methodology used to determine that an electrical or I&C component did not support a license renewal intended function. In the response, please address the how the procedural guidance contained in the scoping and screening procedure LRG-02 and AMR procedure LRG-04 was implemented.

c.

Describe the available documented and electronic resources described in Section 3.1.1.1.1 of LRG-04 that were used for the evaluation.

RAI 2.1 Evaluation of Insulation During the audit, the applicant was unable to adequately describe the evaluation that was performed to determine if any insulation installed in the plant was required to support any system intended functions identified during the scoping process. As a result, the staff requests that the applicant describe any intended functions performed by insulation or the basis for determining that insulation (e.g. piping insulation) did not meet the scoping criteria described in 10 CFR 54.4(a)(1), (a)(2) or (a)(3).

RAI 2.1-8 - Quality Assurance Program Attributes in Appendix A, Safety Analysis Report Supplement, and Appendix B, Aging Management Programs and Activities The NRC audit team reviewed the applicants aging management programs described in Appendix A, Safety Analysis Report Supplement, and Appendix B, Aging Management Programs and Activities, of the Nine Mile Point license renewal application. The purpose of this review was to assure that the aging management activities were consistent with the staffs guidance described in NUREG-1800, Section A.2, Quality Assurance for Aging Management Programs (Branch Technical Position IQMB-1), regarding quality assurance attributes of aging management programs.

Based on the staffs evaluation, the descriptions and applicability of the plant-specific aging management programs and their associated quality attributes provided Appendix B.1.3 of the LRA is consistent with the staffs position regarding quality assurance for aging management. However, the applicant has not sufficiently described the use of the quality assurance program and its associated attributes (corrective action, confirmation process, and administrative controls) in the discussions provided for aging management programs described in Appendix A1, NMP1 Updated Final Safety Analysis Report (UFSAR) Supplement, and Appendix A.2, NMP2 Updated Safety Analysis Report (USAR) Supplement.

The staff requests that the applicant supplement the descriptions in the Appendix A.1 and A.2 to include a description of the quality assurance program attributes, including references to pertinent implementing guidance as necessary, which are credited for the programs described in Appendix B.1.3 of the LRA. The descriptions in Appendix A.1 and A.2 should provide sufficient information for the staff to determine if the quality attributes for the Appendix A.1 and A.2 aging management programs are consistent with the review acceptance criteria contained in NUREG-1800, Section A.2, Quality Assurance for Aging Management Programs (Branch Technical Position IQMB-1).

Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 cc:

Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 126 Lycoming, NY 13093 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Mr. James R. Evans LIPA P.O. Box 129 Lycoming, NY 10393 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy Electric Division NYS Department of Public Service Agency Building 3 Empire State Plaza Albany, NY 12223 C. Adrienne Rhodes Chairman and Executive Director State Consumer Protection Board 5 Empire State Plaza, Suite 2101 Albany, NY 12223-1556 Kathryn M. Sutton, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Mr. Michael J. Wallace President Nine Mile Point Nuclear Station, LLC c/o Constellation Energy Group, Inc.

750 East Pratt Street Baltimore, MD 21202 Mr. James M. Petro, Jr., Esquire Counsel Constellation Energy Group 750 East Pratt Street, 5th Floor Baltimore, MD 21202 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Fred Emerson Nuclear Energy Institute 1776 I St., NW, Suite 400 Washington, DC 20006-3708 Mr. Mark Flaherty Manager - Fleet Licensing R.E. Ginna Nuclear Power Plant 1503 Lake Rd.

Ontario, NY 14519 Mr. M. Steven Leonard General Supervisor - Nuclear Regulatory Matters Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 cc:

Mr. Peter Mazzaferro Site Project Manager - License Renewal Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 Mr. Mike Heffley Senior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401