ML050040332

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License Renewal Application - Responses to NRC Request for Additional Information Regarding Scoping and Screening Methodology
ML050040332
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 12/22/2004
From: O'Connor T
Constellation Energy Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NMPIL 1904, TAC MC3272, TAC MC3273
Download: ML050040332 (25)
v • d • e


Text

Constellation Energy' Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, New York 13093 December 22, 2004 NMP1L 1904 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Nine Mile Point Units 1 and 2 Docket Nos. 50-220 and 50-410 Facility Operating License Nos. DPR-63 and NPF-69 License Renewal Application - Responses to NRC Requests for Additional Information Regarding Scoping and Screening Methodology (TAC Nos. MC3272 and MC3273)

Gentlemen:

By letter dated May 26, 2004, Nine Mile Point Nuclear Station, LLC (NMPNS) submitted an application to renew the operating licenses for Nine Mile Point Units 1 and 2.

In a letter dated November 22, 2004, the NRC requested additional information regarding the scoping and screening methodology utilized to prepare the License Renewal Application. The NMPNS responses to six (6) of the eight (8) requests for additional information (RAIs) contained in the NRC's November 22, 2004 letter are provided in Attachment 1. Responses to the two (2) remaining RAIls will be submitted by January 31, 2005. Attachment 2 provides a list of the regulatory commitments associated with this submittal.

If you have any questions about this submittal, please contact Peter License Renewal Project Manager, at (315) 349-1019.

V ruly Timo y

. O'C Plant eneral TJO/DEV/jm NMPNS

- /I

Page 2 NMP1L 1904 STATE OF NEW YORK

TO WIT:

COUNTY OF OSWEGO 1, Timothy J. O'Connor, being duly sworn, state that I am Nine Mile Point Plant General Manager, and that I am duly authorized to execute and file this supplemental information on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this submittal are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provide by other Nine Mile Point employees and/or consultants. Such information Vs been reviewd in accordance with company practice and I believe it to be relia e

Subscribed and sworn before me, Notary Public in and for the S of New York and County of Oswego, this me day of

>rh er 2004.

WITNESS my Hand and Notarial Seal: -A,4-n Notary Public My Commission Expires:

bCC TnheV c) O j USA M.CAKDate Notary Public in the State of New York OMsweg Cuy Rag. Ho. OjLOg2 Attachments:

1. Responses to NRC Requests for Additional Information (RAI) Regarding Scoping and Screening Methodology
2. List of Regulatory Commitments cc:

Mr. S. J. Collins, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR Mr. N. B. Le, License Renewal Project Manager, NRR Mr. J. P. Spath, NYSERDA

ATTACHMENT I Nine Mile Point Nuclear Station Responses to NRC Requests for Additional Information (RAI)

Regarding Scoping and Screening Methodologv This attachment provides Nine Mile Point Nuclear Station, LLC (NMPNS) responses to six,(6) of the eight (8) requests for additional information contained in the NRC letter dated November 22, 2004, regarding the scoping and screening methodology used to prepare the License Renewal Application (LRA). Each NRC RAI is repeated, followed by the NMPNS response.

Responses to the two (2) remaining RAls will be provided by January 31, 2005. These two RAIs are 2.1-2 and 2.1-7.

RAI 2.1 Review Methodology for Non-Accident Design Basis Events 10 CFR 54.4(a) (1) states, in part, that safety systems, stnrctures, and components (SSCs) within the scope of license renewal include safety-related systems, stnrctures, and components which are those relied upon to remain finctional during and following design-basis events (as defined in 10 CFR 50.49(b)(1)J.

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 nrle 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. Infornation regarding design basis events as defined in 10 CFR 50.49(b)(1) may befound in any chapter of the facility UFSAR, the Commission s 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 andfollowing design basis events (as defined in 10 CFR 50.49(b)(1)) to ensure thefiunctions described in 10 CFR 54.4 (a) (1).

During the scoping and screening methodology audit, the Nuclear Regulatory Commission (NRRQ 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 Page 1 of 22

accident analysis could result 1in omission of safety-relatedfunctions described in the current licensing basis.

The staff therefore, requests the applicant provide the following additional information: a list of the design basis events evaluated as part of the license renewal scoping process, a describe 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) vere addressed during license renewal scoping. For each response, please indicate the docutnmentation sources reviewed to ensure that all design basis events were identified.

For each of the above issues, if the response indicates that additional scoping evaltations 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 renewval efforts, and list those stnrctures and components for which aging nmanagenment reviews wvere conducted. For each additional SC, please describe the aging management programs, as applicable, to be credited for managing the identified aging effects.

Response

The main focus of this request for additional information is the potential omission of safety-related functions from being identified if the source of this information was limited to the Updated Final Safety Analysis Report (UFSAR) chapter on accident analyses. The scoping process used for the Nine Mile Point (NMP) license renewal application (LRA) utilized several categories of document types that are part of the current licensing basis, in addition to the UFSAR accident analyses chapter, to identify all the safety-related functions for both Nine Mile Point stations. As such, the systems, structures and components (SSC) identified in the NMP LRA as meeting the scoping criterion of 10 CFR 54.4(a)(1) is a complete list.

The list of design basis events applicable to NMP1 and NMP2 is derived from each station's current licensing basis (CLB). The CLB is different for each station since they were licensed under different regulations. NMPI obtained its operating license in 1969, which was prior to the issuance of the General Design Criteria (GDC) of 10 CFR 50, Appendix A. However, there were similar Principal Design Criteria to which NMP 1 was designed and licensed. NMP2 received its operating license in 1986 and is a GDC plant that was reviewed in accordance with the Standard Review Plan. For both stations, as new regulatory issues arose, their CLB was revised to address the resolution of that issue. Examples of new regulatory issues include the TMI-related upgrades to plant designs and operation, boiling water reactor (BWR) reactor core stability, and emergency core cooling system (ECCS) suction strainer debris concerns.

The design basis events that are included in the current licensing basis for NMPl are:

Anticipated Operational Occurrences o Turbine Trip Without Bypass o Turbine Trip With Bypass o Loss of Feedwater Heating Page 2 of 22

I o Feedwater Controiler Failure Without Bypass o Feedwater Controller Failure With Bypass o Control Rod Withdrawal Error o Closure of All Main Steam Line Isolation Valves o Closure of One Main Steam Line Isolation Valve o Inadvertent Startup of Cold Recirculation Loop o Trip of All Recirculation Pumps o Trip of One Recirculation Pump o Recirculation Flow Controller Failure - Increasing Flow o Recirculation Flow Controller Failure - Decreasing Flow o Inadvertent Actuation of One Solenoid Relief Valve o Feedwater Controller Failure - Zero Flow o Inadvertent Actuation of One Bypass Valve o One Feedwater Pump Trip and Restart o Loss of Condenser Vacuum o Generator Load Rejection With Bypass o Generator Load Rejection Without Bypass o Loss of Auxiliary Power o Pressure Regulator Malfunction - Open o Pressure Regulator Malfunction - Closed o Power Bus Loss of Voltage o Shutdown Cooling Malfunction - Temperature Decrease o Inadvertent Emergency Condenser Start o Inadvertent High Pressure Coolant Injection Start o Fuel Assembly Insertion o Inadvertent Actuation of One Safety'Valve o Loss of Shutdown Cooling

  • Accidents o Pipe Breaks Outside Primary Containment o Loss of Coolant Accident (Design Basis Equipment Performance) o Loss of Coolant Accident (Emergency Operating Procedures) o Fuel Handling Accident o Control Rod Drop Accident o Fuel Loading Error o Recirculation Pump Seizure
  • Other Events o Shutdown Without Control Rods o Overpressure Protection o Stability o Anticipated Transients Without Scram o Instrument Air Failure o D-C Power Interruptions o Station Blackout o Safe Shutdown for Fire Protection Page 3 of 22

o Metal Water Reaction Capability Analysis o Pipe Whip Capability Analysis The NMPI current licensing basis also indicates that NMPl was designed in accordance with the applicable codes to withstand the most severe earthquake, flooding condition, windstorm, ice condition, temperature and other deleterious natural phenomena which can be expected to occur at the site" (Reference FSAR Section I.A).

The design basis events that are included in the current licensing basis for NMP2 are:

  • Abnormal Operational Transients o Main Generator Trip (Load Rejection) With Bypass System Failure o Main Turbine Trip With Bypass System Failure o Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position o Recirculation Loop Pump Seizure for One Loop o Recirculation Loop Pump Shaft Break
  • Design Basis Accidents o Control Rod Drop Accident o Fuel Handling Accident o Loss of Coolant Accident Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Inside Containment Page 4 of 22

o Small, Large, Steam and Liquid Piping Breaks Outside Containment o Instrument Line Break Outside Drywell o Feedwater Line Break Outside Containment o Gaseous Radwaste System Leak or Failure o Augmented Offgas Treatment System Failure o Liquid Radwaste System Leak or Failure o Liquid Radwaste System Storage Tank Failure Special Events o Shipping Cask Drop o Reactor Shutdown from Anticipated Transients without Scram (ATWS) o Reactor Shutdown from Outside Main Control Room o Reactor Shutdown Without Control Rods The NMP2 current licensing basis also indicates that the SSC important to safety are designed for:

  • Protection Against Natural Phenomena (GDC 2) o Earthquakes o Tornadoes o Hurricanes o Floods o Tsunami o Seiches
  • Fire Protection (GDC 3) o Fire o Explosions
  • Environmental and Missile Design (GDC 4) o Missiles o Pipe Whip o Discharging Fluids The NMP design and configuration control processes ensure that those SSCs required to perform a safety-related function are properly evaluated and identified. The process for evaluating the safety classification for a SSC requires that the function of the SSC be identified via a review of the following documents.
  • Licensing and Design Basis Documentation for the Component
  • Failure Modes and Effects Analysis (Unit 2) - Components
  • Nuclear Safety Operational Analysis (Unit 2 FSAR App 15A) - Systems Level
  • Color Coded Piping and Instrumentation Diagrams (P&IDs) and Q-List Drawings (Unit 1)

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Once the function(s) of the SSd are determined, they are compared against the criteria of 10 CFR 50.49(b)(1). The evaluation and resulting safety classification is documented in controlled design documents, such as Safety Classification (Appendix B) Determinations and Safety Class Boundary Drawings, and the Master Equipment List (MEL) database. Therefore, the NMP design and configuration control process has ensured that the design basis events applicable to each station have been reviewed when determining the safety classification for each SSC.

The identification of safety-related functions was also determined during the scoping process use for compliance with 10 CFR 50.65, the Maintenance Rule.

The methodology used to ensure that all design basis events were addressed during license renewal scoping was to utilize the NMP controlled documents and databases that identified those SSCs and functions classified as safety-related. These documents and databases consist of the NMP1 UFSAR and NMP2 Updated Safety Analysis Report (USAR), Safety Class Boundary Drawings, Appendix B Determinations, Maintenance Rule Scoping Documents, and Master Equipment Lists. Evidence of different documents being used to identify safety-related functions for licensing renewal scoping is documented in the Scoping and Screening Reports developed for each NMP I and NMP2 system and structure. These documents are available for review at the NMP site.

Based upon the above description of how the safety classification of the NMP SSC and functions are determined and documented, NMP is confident that all design basis events were considered in the scoping process for license renewal.:

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 consider 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 ofplant-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 IO CFR 54.4(a)(1). The applicant 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 andfacility database information. Based on the audit, the staff requests a description of the process used during license reneival scoping activities to disposition components classified as safety-related that do not perfonn 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 ihe basis for Page 6 of 22

determining that these comiponents do not perform a safety-related intendedfiunction.

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 thefacility database safety classification information with scoping intended function determinations. In particular, the staff requests a description of the process including the scope of the review used to re-evaluate the safety-classification of SSCs to reconcile disparities with intendedfunction determinations.

Response

The response to this RAI will be provided in a future submittal.

RAI 2.1 Scope of Current Licensing Basis 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 intendedjfunction(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 ofNRC requirements applicable to a specificplant and a licensee s 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 recentfinal 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 applicant's 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 intendedffunction 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 thefollowing information:

a.

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

b.

In discussions with the Nine Mile Point license renewvalproject team during the scoping and screening methodology audit, it was identified that an electronic document database Page 7 of 22

was used to identify CLB documents pertinent to the development of system descriptions and identification of systenm intendedfunctions. 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.

Response

a. As part of the License Renewal (LR) scoping process employed for the NMP systems and structures, system descriptions and system intended functions were identified. The methodology used for researching and documenting these was controlled in LR project procedure LRG-02, "Licensing Renewal Scoping and Screening." The direction provided for the system descriptions was to document "A concise but complete description of the system or structure. The system description should describe the entire system, at a summary level, generally emphasizing the parts that are in scope for license renewal." Similarly for the system intended functions, the direction provided was to document "A complete list of functions performed by the system or structure. (Note that these functions may or may not be intended functions.)"

The LRG-02 procedure also provided direction on the sources of information to be researched to develop the system descriptions and intended functions. The primary sources for system descriptions and functions were:

  • UFSARIUSAR
  • Maintenance Rule Scoping Documents
  • Design Basis Documents (DBDs)
  • Nuclear Engineering Reports (NERs) for the regulated events.

Other sources of information included:

  • Appendix B Determinations (safety class information)
  • Design documents (including specifications and calculations)
  • Master Equipment List (MEL)
  • Operating procedures and instructions
  • Plant drawings (including P&IDs and safety class boundary drawings)
  • Licensing documents (including docketed correspondence)
  • Training materials (for system descriptions)

The results of the research were documented in system and structure scoping and screening reports that undervent a review and approval process. This process involved an independent LR project engineer review, discipline lead review, supervisor review, system engineer review, and, finally, project manager review and approval. Thus there is high confidence that system descriptions are accurate and all functions have been properly identified. The specific documents used for the generation of the system descriptions and intended functions are referenced in the individual system and structure scoping and screening report. These reports are available for review at the NMP site.

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b. During the initial phase of the NMP LR project, the correspondence between NMP and NRC was identified as a required source of information. However, since these documents were only available in hard copies (paper form), a task was completed to create a digital image (i.e., electronic copy) which was text searchable. It is these electronic copies that are referred to in the request for additional information as the "electronic current licensing basis document database." However, this set of electronic copies is not the complete set of documents constituting the current licensing basis nor is it a database.

Since the digital imaging of the NMP/NRC correspondence was completed using standard commercial products, there was no need for specific proceduralized controls. A member of the NMP LR project staff was assigned oversight responsibility to ensure that the correspondence was imaged and filed properly. The NMP/NRC correspondence up to February 2003 was copied into this electronic file. Correspondence after this date was available for review only as hard copies. However, since the number was relatively small and the topics generally did not impact the LR project, it was deemed acceptable to not transfer them to the electronic file.

These documents, which constitute part of the current licensing basis, were used to support the development of various reports for the project. They were extensively used in researching potential time-limited aging analyses but were also used in the scoping and aging management review phases as well. In support of the scoping phase, these documents were researched to ensure that all functions were properly identified for the fire protection, anticipated transients without scram (ATWS) and station blackout (SBO) regulated events.

Specific documents reviewed included NMP responses to the issuance of new regulations (i.e., ATWS and SBO), NRC Safety Evaluations, NMP responses to the safety evaluations, as applicable, and NMP responses to generic letters. The electronic files were also researched when specific questions arose during scoping and aging management program reviews.

RAI 2.1 10 CFR 54.4(a)(2) Scoping Criteria for Nonsafetv-related SSCs By letters dated December 3, 2001, and March 15, 2002, the Nuclear Regulatory Commission (NRC) issued a staffposition 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-relatedfunctions identified in paragraphs (a)(1) (i), (ii), (iii) of this section.) Tle December 3rd letter provided specific examples of operating experience whlich identifiedpipefailure events (summarized in Information Notice (IN) 2001-09, "Main Feedwvater 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 15 1h 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 hypotheticalfailures, but rather should base their evaluation on the plant s CLB, engineeringfiudgement and analyses, and relevant operating experience. The March 1 5th letterfiurther describes operating experience as all documented Page 9 of 22

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 applicant 's 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 applicant's evaluation of the 10 CFR 54.4(a) (2) criteria.

The staff requests the applicant provide the following information:

a.

Non Safetv-Related OVSR) Pipin Attached to Safetv-Related (SR) Pipinz Section 2.1.4.2.3, "SR/NSR Piping Interface, " of theLRA states thatfor 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 firthest.

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 inon 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-relatedpiping up to its anchors and (2) the associatedpiping 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 usedfor the purposes of the 10 CFR 54.4(a) (2) evaluation;
2. A description of how the first seismic anchor was identifiedfor 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./;

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3. Confirmation that for the NSR piping, associated plant equipment, and their supports, tip to and including the first seismic anchor, were within the scope of license renewal and subject to aging management review.
4. A discussion ofhowv plant equipment identified as the tennination point for non safety-relatedpiping was evaluated during the scopinlgprocess. Aspart ofthat discussion, if the plant equipment provides a structural supportfillnction 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 fitnction is maintained) is not in scope.

7he 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 afail-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, ifthe response indicates that use of the scoping methodology screened 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 ofyour efforts, and list those SCsfor which aging management reviews were conducted. For each SC also describe the aging management programs, as applicable, to be creditedfor managing the identified aging effects.

Response

a. Responses to each of the four itemized requests are provided as follows.
1. The definition of an equivalent anchor, which was used for the purposes of the 10 CFR 54.4(a)(2) evaluation, is different between NMP1 and NMP2. For NMP1, which was licensed in 1969, the definition of an equivalent anchor is a combination of structural supports and changes in piping geometry that provides restraint to the piping in six degrees of freedom. Mechanical components; i.e., pumps and heat exchangers, can also provide restraints to one or more of the degrees of freedom.

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For NMP2, which was licensed in 1986, the definition of the equivalent anchor is an actual structural anchor that is designed to provide restraint to the piping in six degrees of freedom.

LRA Section 2.1.4.2.3, "SR/NSR Piping Interface," discusses the term "seismic anchor" and includes a footnote with the above information.

2. The NMP LR scoping process did not identify the physical end point in the field of the NSR piping, fittings, and equipment within scope to support submittal of the license renewal application. This level of detail was determined to be unnecessary at this point in time to demonstrate adequate aging management. The aging management programs credited for the NSR piping, fittings, and equipment are not dependent upon the physical location of the end point. The NSR portions are constructed of the same material and exposed to the same environments. Therefore, the aging management reviews are complete to determine the appropriate aging management programs.

The determination of the actual "seismic anchor" is important during the implementation phase of the license renewal project. At this future point in time, NMP will use the definition of equivalent anchor stated in LRA Section 2.1.4.2.3 to define the end points of the applicable NSR piping, fittings and equipment. This information will be part of the implementing procedures for the applicable aging management programs; i.e., Systems Walkdown Program, Flow Accelerated Corrosion Program, etc.

3. For NSR piping attached to SR piping, all NSR piping, fittings, and equipment (associated plant equipment), and their supports, up to and including the first seismic anchor, are within the scope of license renewal and subject to aging management review.

The associated aging management programs are/will be applied to all of the components in this component type.

4. As stated in Item 2 above, the scoping process did not identify the physical end point in the field of the NSR piping, fittings, and equipment within scope. This will be determined during the implementation phase of the license renewal project. However, plant equipment that is determined to be the termination point for NSR piping will be considered within scope of license renewal and subject to an AMR.
b. In developing the NMP LR project procedure for scoping and screening, clarification was provided for the NSR criterion; i.e., 10 CFR 54.4(a)(2). One of the clarifications was that malfunctions of NSR equipment that result in a challenge to SR equipment ([where] the SR function is maintained) are not part of the scope of 54.4(a)(2). The basis for this position is that this situation does not meet the criterion of 54.4(a)(2), which states that "All nonsafety-related SSCs whose failure could prevent satisfactory accomplishment of any safety-related functions identified in" 54.4(a)(1). The clarification states that even though a SR component may be impacted, the SR function that the component provides is not impacted. Criterion 54.4(a)(2) states that the SR function is prevented. Since the SR function is not prevented under the clarification provided, the clarification is considered acceptable.

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With respect to all applications where this clarification was used during the scoping process for NMP1 and NMP2, it was used for those SSCs that fall under one of the criteria from the Maintenance Rule (10 CFR 50.65). This criterion is identified as "SC5 - NSR SSC Failure Causes a Scram or actuates a SR system." For each NMPl and NMP2 SSC function that was determined to meet this criterion for the Maintenance Rule scoping evaluation, the corresponding function was classified as not within scope of LR. The details of the specific functions that met this clarification for LR are available in the NMP1 and NMP2 Scoping and Screening Reports, which are available at NMPNS for review.

c. In developing the NMP LR project procedure for scoping and screening, clarification was provided for the NSR criterion; i.e., 10 CFR 54.4(a)(2). Another of the clarifications was 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 basis for this position is that this situation does not meet the criterion of 10 CFR 54.4(a)(2), which states that "All nonsafety-related SSCs whose failure could prevent satisfactory accomplishment of any safety-related functions identified in..." 10 CFR 54.4(a)(1). The clarification states that even though a SR component may be impacted, the SR function that the component provides is not impacted.

Criterion 54.4(a)(2) states that the SR function is prevented. Since the SR function is not prevented under the clarification provided, the clarification is considered acceptable.

An example of this clarification is building blow out panels which are held in place by clips designed to fail and relieve an overpressure condition in the building. Premature failure of the clips would still permit the panels to perform their intended functions; thus, they would not be required to be in scope. However, there were no instances noted during the NMP1 and NMP2 scoping process where this logic was used to exclude any SSCs.

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 plant'S CLB (as defined in 10 CFR 54.3), plant-specific experience, industry-wvide experience (as appropriate), and safety analyses or plant evaluations to perform a fiunction that demonstrates compliance with NRC regulations identified under 10 CFR 54.4(a) (3), are required to be included within the scope of the nrle. As part of the LRA review, the NRC staff evaluates the scope and depth of the applicant 'S document review to provide assurance that the scoping methodology considered all SSC intended functions.

During the NRC's 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 implemeintation 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 (.4TWS) 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 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 Page 13 of 22

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 systenm 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 whlich 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 ATWMS 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 2feedwater 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 thatfailure to incorporate the infonnation in the Technical Position Papers has resulted in the onmission of relevant infornation in the LRA or omission of SSCs from within scope, please describe any additional scoping evaluations performed to address this concern. As part ofyour 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 creditedfor managing the identified aging effects.

Response

a. "Technical position papers" used at NMP are controlled in accordance with engineering administrative procedure NEP-DES-02, "Engineering Evaluations." The general methodology employed involves the preparation of the engineering evaluation, a technical review or design verification, and approval by the responsible supervisor. The preparer is directed to "Perform the evaluation and document in sufficient detail to allow a technically qualified reviewer/design verifier to understand the purpose, inputs, evaluation criteria, assumptions, method, references, and conclusions of the evaluation, and to conclude adequacy without recourse to the originator." Design verification is required when the evaluation involves safety-related systems, structures, or components. The evaluation is documented as a Nuclear Engineering Report (NER). (Note that a NER documents an investigation of technical information relating to the design or licensing basis but is not, by itself, a design or licensing basis document.)

This administrative procedure also applies to the review and acceptance of vendor-supplied documents. These documents also require a review and approval by NMP prior to use.

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(Note that the two technical position papers (ATWS and SBO) referred to in the request for additional information above were not reviewed and approved for use. It is this error that led to the apparent discrepancy'iith the Feedwater System described above.)

There were five NERs used to support the LR scoping process for NMP. These documents, and their CLB source documents, are listed below.

1. NER-IE-024, Identification ofNMP1 Offsite Power Station Blackout (SBO) Scope Addition Components for License Renewal
a. UFSAR Section IX, Electrical Systems
2. NER-lM-025, Station Blackout Evaluation
a. NRC Station Blackout Rule Safety Evaluation - Nine Mile Point Nuclear Station Unit 1, July 1, 1991.
b. NRC Station Blackout Rule Supplemental Safety Evaluation -Nine Mile Point Nuclear Station Unit 1, November 6, 1991
3. NER-2E-027, Identification of Systems and Components Required for Station Blackout Event (SBO) - Coping
a. NRC Station Blackout Rule Safety Evaluation - Nine Mile Point Nuclear Station Unit 2, dated May 29, 1991
4. NER-2E-028, Identification of Equipment and Components Required for Anticipated Transients Without Scram
a. NRC Safety Evaluation Report on Nine Mile Point Unit 2 Compliance with ATWS Rule 10 CFR 50.62 Relating to Alternate Rod Injection (ARI) and Reactor Protection Trip (RPT) System, dated March 24, 1988
b. NMP2 Technical Specifications B.3.3.4.2 Bases
5. NER-2E-029, Identification of NMP2 Offsite Power Station Blackout (SBO) Scope Addition Components for License Renewal
a. USAR Section 8.2.1.1, Offsite Power Grid System
b. As stated in the RAI, a discrepancy was identified by the NMP staff during the NRC scoping and screening methodology audit in which the NMP2 Feedwater System was not identified as also being in scope for the SBO regulated event. NMP initiated corrective action report DER-NM-2004-4466 to resolve this issue. This discrepancy was caused by the use of an unapproved document to determine the NMP2 systems and components (SCs) in scope for the ATWS and SBO regulated events. The actions taken to ensure that all the required NMP2 SCs were properly identified within the scope for LR were to compare the systems listed in the LRA to those identified in the approved engineering reports; i.e., NER-2E-027, NER-2E-028 and NER-2E-029. The results of this comparison identified that LRA Section 2.3.4.B.3, "NMP2 Feedwater System," did not need to be identified as in scope for the SBO regulated event since the Feedwater components credited for SBO (i.e., reactor coolant/containment isolation valves) were already properly included in LRA Section 2.3.2.B.5, "NMP2 Primary Containment Isolation System." This system includes the reactor Page 15 of 22

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coolant/containment isolation valves for all systems and is properly credited for being in scope for the SBO regulated event. Therefore, the apparent discrepancy identified during the audit was determined to be incorrectly characterized.

However, during the review to ensure all systems were properly identified for the ATWS and SBO regulated events, it was discovered that the NMP2 Common Electrical System should have been identified as in scope for the SBO regulated event. NER-2E-027 identifies this system as containing computers and printers as components credited for coping with a SBO event. However, LRA Section 2.5.B.4, "NMP2 Common Electrical System," indicates that this system is in scope only for the safety-related criterion and the fire protection, EQ and ATWS regulated events. Therefore, LRA Section 2.5.B.4 will be revised, as indicated below, to also include the SBO regulated event as a criterion for this system. Note that since the in-scope components in this system are classified as "Active," they do not require an aging management review and no changes to LRA Section 3.6, "Aging Management of Electrical and Instrument and Controls Systems," are required. The NMP LR Scoping and Screening Report will also require a change to properly indicate that this system is credited for SBO. This action is captured in the NMP corrective action program under DER NM-2004-4466. Another action captured in this DER is to revise the applicable NMP2 Scoping and Screening reports to delete reference to the unapproved engineering report and identify the appropriate approved engineering report.

No additional reviews were conducted for the NMPl 'systems to ensure they were properly identified for the ATWS and SBO regulated events since the original scoping and screening reports were based upon existing approved engineering reports. For the ATWS criterion, NMP1 installed a new system to meet the regulations. Therefore, this new system, also titled ATWS, was properly identified as being in scope. For the SBO criterion, NER-IM-025, "Station Blackout Evaluation," dated November 22, 1995 and NMP calculation S0-SBO-M016, "NMP1 Station Blackout Equipment List," dated August 23, 1995, were used to identify the systems and components credited for SBO coping. Those systems and components credited for SBO recovery were identified by using NER-IE-024, "Identification of NMP1 Offsite Power Station Blackout (SBO) Scope Addition Components for License Renewal." Since the scoping for the NMP1 systems originally utilized approved engineering documents, it was determined that a re-evaluation was not necessary.

c. Response provided in Item b above.

LRA Revisions In LRA Section 2.5.B.4, the second bulleted item is revised as follows (changes are highlighted in italics):

"It contains SCs 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), environmental qualification (10 CFR 50.49), anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63)."

Page 16 of 22

RAT 2.1 Implementation of the Electrical Spaces Scoping Approach NUREG-1800, Section 2.5.3. 1, "Components Within the Scope ofLicense Renewal, " states that an applicant may use the plant spaces approach in scoping electrical and instrumentation and control (QI&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 ofscopingperfornmed 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 applicant 's methodology for scoping and screening of electrical and I&C components. T7e staiffound that the procedures related to electrical and I&C scoping and screening lacked sufficient detail to determine if the applicant 's 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, firther 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 intendedfunctions 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 wereperfonned by the applicant 's staff to identify the applicable intendedfunctions, plant electrical equipment required to perforni thosefunctions, 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 usedfor 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 detennine that an electrical or I&C component did not support a license renewal intendedfunction. In your response, please address 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 ofLRG-04 that were used for the evaluation.

Page 17 of 22

Response

The LR scoping and screening process used to evaluate electrical and instrumentation and controls (I&C) components at NMP addressed all components and divided them into three categories. The categories were electrical and I&C components associated with mechanical systems, electrical and I&C components associated with electrical systems, and electrical and I&C components associated with commodities. The first two categories were treated the same as mechanical components. That is, each component was listed in the scoping and screening report for the system with which it was associated. The components' system was as defined in each plants' Master Equipment List (MEL). Each electrical and I&C component was evaluated as to whether it supported a system intended function. If it did, the component was considered within scope of LR. Conversely, if the component did not support a system intended function, it was considered not within scope of LR. Each of the in-scope components was then evaluated for whether it was classified as an active or passive component and for whether it was considered short or long lived. For the electrical and I&C components in mechanical and electrical systems, all were determined to be active components and, therefore, not subject to aging management review. The above-described methodology is contained in project procedures LRG-01, "License Renewal Project General Guidance," and LRG-02, "License Renewal Scoping and Screening."

With respect to the third category of electrical and I&C components, the methodology was slightly different. Based upon industry experience and guidance, those electrical and I&C components that are typically passive and long-lived were assigned directly to commodity groups without being evaluated with the systems in which they were installed. The electrical commodity groups, and the component types assigned to each group, are as follows.

  • Cables and Connectors - cables (power and instrument), connectors attached to the cable, splices, fuse blocks and terminal blocks
  • Non-Segregated/Switchyard Bus - conductors and insulators installed in electrical buses
  • Containment Electrical Penetrations - internal cables of building penetrations, including the connectors at the end of the "pig tail"
  • Switchyard Components - high voltage transmission conductors and insulators All components within a commodity group were considered to be within the scope of LR.

Therefore, no effort was made to identify each component by its component ID (i.e., unique MEL identifier). However, each component type was considered subject to aging management review.

Using the above general description of the scoping and screening methodology for electrical and I&C components, the following provides more details based upon the issues raised in this RAI.

a. Electrical systems were identified based upon those defined in the MEL, UFSAR, and Maintenance Rule (MR) Scoping Reports. The boundaries of each electrical system are based upon the components assigned to the system as well as any descriptions in the UFSAR and/or other design basis document. The MEL was used as the design document/database that assigned components to a particular system. The electrical systems and components Page 18 of 22

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defined in MEL were imported into the NMP LR database, ConRAD. The information contained in ConRAD for each electrical system included a system description, list of system functions, identification of which functions met any of the LR scoping criteria, a list of NMP documents from which this information was derived, and any corresponding comments. For each electrical component in the system, ConRAD contained information on the component unique identifier, title/description, and whether the component was in scope, active or passive, short or long-lived and subject to aging management review or not.

A scoping and screening report was prepared for all electrical systems in accordance with the direction provided in project procedure LRG-02,"License Renewal Scoping and Screening."

A complete listing of the electrical systems evaluated for NMP1 and NMP2 is presented in LRA Tables 2.2-1 and 2.2-2, respectively Those electrical systems that have intended functions, and, therefore, are within scope of LR, are described in LRA Section 2.5.

b. The methodology used to determine whether an electrical or I&C component supported an intended function is described in project procedure LRG-02. (Note that this discussion applies only to those electrical and I&C components associated with the first two categories identified above. Those components assigned to a commodity group were categorically considered to be in scope and subject to AMR.) Section 3.7 of LRG-02 requires that the electrical or functional boundary be described for the intended functions of electrical systems. This activity identifies a group of components that support a specific intended function. For example, all electrical and I&C components that are identified on the environmental qualification (EQ) list for a system are the group of components that support the EQ intended function. Another example is the group of components that supports the ATWS criterion. These are the components in the circuits that trip the reactor recirculation pumps. Section 3.8 of LRG-02 then requires that a comprehensive list of components be generated and input into ConRAD. Only those components that support a system intended function are classified as in-scope. In Section 3.9 of LRG-02, each in-scope component is evaluated for whether it is active or passive and short or long-lived. However, in recognition of the categorization of electrical and I&C components described above, Step 3.9.2 states "All electrical components that are long-lived and passive are addressed as electrical commodities by LRG-04, Aging Management Review for Electrical Commodities Guideline.

Thus, no systems encompassed by this guideline are considered to contain electrical components subject to AMR."

With respect to LRG-04, it does not contain any methodology to determine whether an electrical or I&C component supports an intended function since all components addressed by this procedure are categorically classified as in-scope. Therefore, there is no need for a methodology to determine if a component is in scope or not.

c. LRG-04 was not used for scoping and screening of electrical and I&C components (see above). Thus, no resources described in Section 3.1.1.1.1 of LRG-04 were used for scoping and screening purposes. The "list of items" referred to in this section is in regard to the different types or brands of components within the commodity being evaluated in the AMR.

This is a necessary step in determining the materials of construction. For example, all the Page 19 of 22

different types/brands of cables must be identified to perform the cable AMR. This "list of items" comprises all the cables to be reviewed.

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 intendedfiunctions identified during the scoping process. As a result, the staifrequests that the applicant describe any intendedfunctions performed by insulation or the basis/or 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).

Response

The response to this request will be provided in a future submittal.

RAI 2.1 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 applicant v aging management programs described in Appendix A, "Safety Analysis Report Supplementt " and Appendix B, "Aging Management Programs and Activities, "ofthe 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 staff's evaluation, the descriptions and applicability of the plant-specific aging management programs and their associated quality attributes provided in Appendix B. 1.3 of the LRA is consistent with the staff s 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 providedfor aging management programs described in Appendix Al, "NMPI 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. Th7e descriptions in Appendix A.I and A.2 should provide szfficient informnation for the staff to determine ifthe quality attributes for theAppendixA.1 and A.2 aging management programs are consistent with the review acceptance criteria contained in NUREG-1800,SectionA.2. "QualityAssuranceforAgingManagementiProgranms(BranchlTechnical Position IQMB-1). "

Page 20 of 22

Response

The Nine Mile Point LRA, Appendix A.1 and A.2, will be revised as follows to describe the use of the NMP quality assurance program attributes of corrective action, confirmation process, and administrative controls as applied to the aging management programs credited for license renewal.

Revise LRA Appendix A. 1 by adding the following:

A1.0 GENERIC QUALITY ASSURANCE PROGRAM REQUIREMENTS FOR LICENSE RENEWAL The NMP Quality Assurance Program implements the requirements of 10 CFR 50, Appendix B, and is consistent with the summary in Appendix A.2 of NUREG-1800, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants," published July 2001. The elements of corrective action, confirmation process, and administrative controls in the Quality Assurance Program are applicable to both safety-related and non-safety related systems, structures, and components that are subject to an aging management review. Generically, these three elements are applicable as follows:

Corrective Actions Corrective actions are implemented in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMPI UFSAR, Appendix B. The NMP corrective action program provides for the identification, evaluation, and resolution of non-conforming conditions.

Confirmation Process The confirmation process is part of the corrective action program, which is implemented in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMP1 UFSAR, Appendix B. The focus of the confirmation process is on the verification that corrective actions are effective. The measure of effectiveness is in terms of correcting the adverse condition and precluding repetition of significant conditions adverse to quality.

Administrative Controls Aging management programs are implemented through various plant documents. These implementing documents are subject to administrative controls, including a formal review and approval process, in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMP1 UFSAR, Appendix B.

Page 21 of 22

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Revise LRA Appendix A.2 by adding the following:

A2.0 GENERIC QUALITY ASSURANCE PROGRAM REQUIREMENTS FOR LICENSE RENEWAL The NMP Quality Assurance Program implements the requirements of 10 CFR 50, Appendix B, and is consistent with the summary in Appendix A.2 of NUREG-1800, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants," published July 2001. The elements of corrective action, confirmation process, and administrative controls in the Quality Assurance Program are applicable to both safety-related and non-safety related systems, structures, and components that are subject to an aging management review. Generically, these three elements are applicable as follows:

Corrective Actions Corrective actions are implemented in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMP2 USAR, Appendix B. The NMP corrective action program provides for the identification, evaluation, and resolution of non-conforming conditions.

Confirmation Process The confirmation process is part of the corrective action program, which is implemented in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMP2 USAR, Appendix B. The focus of the confirmation process is on the verification that corrective actions are effective. The measure of effectiveness is in terms of correcting the adverse condition and precluding repetition of significant conditions adverse to quality.

Administrative Controls Aging management programs are implemented through various plant documents. These implementing documents are subject to administrative controls, including a formal review and approval process, in accordance with the requirements of 10 CFR 50, Appendix B, as committed in the NMP2 USAR, Appendix B.

Page 22 of 22

ATTACHMENT 2 List of Regulatorv Commitments The following table identifies those actions committed to by Nine Mile Point Nuclear Station, LLC (NMPNS) in this submittal. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT l

DUE DATE Submit responses to the following two requests for additional January 31, 2005 information (RAI) contained in the NRC letter dated November 22, 2004: 2.1-2 and 2.1-7.

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