3F0909-03, Response to Requests for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application (TAC No. ME0274) and Amendment 2

Response to Requests for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application (TAC No. ME0274) and Amendment 2

ML092580095
Person / Time
Site:	Crystal River
Issue date:	09/11/2009
From:	Franke J Progress Energy Co, Progress Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0909-03, TAC ME0274
Download: ML092580095 (30)

Text

0Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 54 September 11, 2009 3F0909-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Requests for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application (TAC NO. ME0274) and Amendment #2

References:

(1)

CR-3 to NRC letter 3F1208-01, dated December 16, 2008, "Crystal River Unit 3 -

Application for Renewal of Operating License" (2)

NRC to CR-3 letter dated August 14, 2009, "Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application (TAC NO. ME0274)," Sections 2.2 and 2.3 (3)

NRC to CR-3 letter dated August 14, 2009, "Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application (TAC NO. ME0274)," Section 2.3

Dear Sir:

On December 16, 2008, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc. (PEF), requested renewal of the operating license for Crystal River Unit 3 (CR-3) to extend the term of its operating license an additional 20 years beyond the current expiration date (Reference 1).

Subsequently, the Nuclear Regulatory Commission (NRC), by two letters dated August 14, 2009, provided requests for additional information (RAI) concerning the CR-3 License Renewal Application (References 2 and 3). Enclosures I and 2 to this letter provide the response to References 2 and 3, respectively. Enclosure 3 provides an amendment to the License Renewal Application.

No new commitments are contained in this submittal. However, the RAI responses resulted in a change to License Renewal Commitment #19. Enclosure 3 describes revised Commitment #19.

If you have any questions regarding this submittal, please contact Mr. Mike Heath, Supervisor, License Renewal, at (910) 457-3487, e-mail at mike.heath@pgnmail.com.

SSi

rely, J

A. Franke ice President Crystal River Unit 3 JAF/dwh

Enclosures:

1.

Response to Request for Additional Information (Reference 2)

2.

Response to Request for Additional Information (Reference 3)

3.

Amendment #2 Changes to the License Renewal Application xc:

NRC CR-3 Project Manager NRC License Renewal Project Manager NRC Regional Administrator, Region II Senior Resident Inspector Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428

U. S. Nuclear Regulatory Commission 3F0909-03 Page 2 of 2 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, infor.atio, and belief.

on A. Franke Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this

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"*I"Lbj&,, 2009, by Jon A. Franke.

day of Signature of Notary Public State of Florida "of"".

CAROLYN E. PORTMAHN vp No" W*

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5 B N--o*Nal M-om A--m.

(Print, type, or stamp Commissioned Name of Notary Public)

Personally Known

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Produced

-OR-Identification

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (REFERENCE 2)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 1 of 12 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (REFERENCE 2)

RAI 2.2-01: General Scoping Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

The Hydrogen Monitoring System is mentioned in the Crystal River Unit 3 Nuclear Generating Plant (CR-3) Final Safety Analysis Report (FSAR) in Section 9.11.2.1.2 and in the CR-3 license renewal application (LRA) in Section 2.3.3.61 under the Post Accident Containment Atmospheric Sampling System (PASS) discussion.

In both references, the Hydrogen Monitoring System is noted to share two sampling points with the PASS. No separate scoping discussion or scoping result regarding the Hydrogen Monitoring System is presented in the LRA.

Request:

Explain the exclusion of the Hydrogen Monitoring System from scope of license renewal per 10 CFR 54.4.

CR-3 Response:

CR-3 does not identify the components performing the hydrogen monitoring function as a unique system; rather, the hydrogen monitoring flow paths and hydrogen analyzers discussed in LRA Section 2.3.3.61 are a subsystem of the Post-Accident Containment Atmospheric Sampling System.

Hydrogen analyzers and associated components performing the containment hydrogen monitoring function are included within the scope of license renewal, and are depicted on scoping drawing 302-693-LR, Sheet 1.

RAI 2.2-02: General Scoping Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

The Auxiliary Feedwater (AFW) System is discussed in the CR-3 FSAR in Section 10.6.

Scoping drawing 302-081-LR, Sheet 4 shows the entire AFW System in scope.

LRA Table 2.3.4-8 indicates that the AFW pump, AFW pump bearing cooler housing, and AFW pump bearing cooler tubes are all included in the aging management review (AMR), however no separate discussion for the AFW System is presented in the LRA.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 2 of 12 In the CR-3 FSAR, Section 10.6.1, the AFW pump is designed to provide an additional non-safety grade source of secondary cooling to the once-through steam generators (OTSGs),

should a loss of all main and emergency feedwater (EFW) occur. This "AFW source" was added in response to U.S. Nuclear Regulatory Commission (NRC) concerns on the issue of EFW'reliability (Generic Issue 124 and SRP Section 10.4.9).

Request:

Provide clarification if the AFW system should or should not be in scope for license renewal. If the AFW system is in scope, provide a discussion of the AFW system similar to those provided in LRA Section 2.3 for other systems including the specific license renewal intended functions in accordance with 10 CFR 54.4 that the AFW system is credited with performing.

CR-3 Response:

The Auxiliary Feedwater Pump and related components are not an independent system, but are part of the Main Feedwater System described in Section 2.3.4.10 of the LRA. As stated in Section 10.6.2 of the CR-3 FSAR:

The Auxiliary Feedwater source consists of one motor driven pump (FWP-7), two pneumatically operated flow control valves (FWV-216 and FWV-217), and associated piping, valves, equipment, instruments, and controls. This equipment is part of the Main Feedwater System.

These components are within the scope of license renewal as demonstrated by the highlighting on license renewal drawing 302-081-LR, Sheet 4. As indicated on this drawing, FWP-7 can take suction from the Condenser Hotwell, CDHE-4B (Drawing 302-082-LR, Sheet 1, coordinate F9) or the Condensate Storage Tank, CDT-1 (302-101-LR, Sheet 2, coordinate B2). Also, the Dedicated Emergency Feedwater Tank, EFT-2 (Drawing 302-082-LR, Sheet 2, coordinate C5) may be aligned as a backup source of feedwater.

Auxiliary Feedwater (AFW) provides an additional non-safety grade source of secondary cooling to the steam generators. It is designed for use on an "as available" basis, only if emergency feedwater is unavailable and is not required for plant startup or normal plant operation.

When used for secondary cooling, the AFW source serves as a supplement to the High Pressure Injection/Power-Operated Relief Valve (HPI/PORV) method of cooling, but does not supersede procedural criteria governing when HPI/PORV cooling must be initiated and when it can be secured. AFW will not be operated simultaneously with emergency feedwater, except if transferring from one to the other. This transfer will be controlled to prevent violating the steam generator upper nozzle flow limits. As part of a defense-in-depth strategy, AFW, if available, may be used instead of the cyclic mode of the Turbine-Driven Emergency Feedwater Pump, EFP-2.

During normal plant operation, the AFW pump can be operated in the recirculation mode without restriction. Periodic full flow testing of the pump is accomplished in this manner.

Except at interfaces with safety related equipment and structures, the AFW source is non-safety grade and is not Class 1 E powered or electrically connected to the emergency diesel generators. As such, it is not relied upon during design basis events and is intended for use on

U. S. Nuclear Regulatory Commission 3F0909-03 Page 3 of 12 an "as available" basis, only. AFW performs no safety function, and there is no impact on nuclear safety if it fails to operate.

10 CFR 54.4(a)(1) Functions This portion of the Main Feedwater System performs no safety related functions.

10 CFR 54.4(a)(2) Functions This portion of the Main Feedwater System is credited with functioning in support of mitigation of a Steam Generator Tube Rupture accident.

This portion of the Main Feedwater System provides a flow path for Auxiliary Feedwater from the Condensate Storage Tank to the Once-Through Steam Generators.

This portion of the Main Feedwater System is credited with functioning in support of safe shutdown in Unresolved Safety Issue (USI) A-46, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, evaluations.

10 CFR 54.4(a)(3) Functions This portion of the Main Feedwater System includes components relied upon for compliance wkith 10 CFR 50.48 for fire protection.

This portion of the Main Feedwater System includes components relied upon for compliance with 10 CFR 50.63 for Station Blackout (SBO).

RAI 2.2-03: General Scoping Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

On LRA Figure 2.2-1, CR-3 Plant Structures, the applicant shows structures in light lines, denoting the structure is not in scope of license renewal. Among the structures the applicant depicts as not in scope are the Reactor Building (RB), Maintenance Building and the Health Physics (HP) Office structures. In FSAR Chapter 5, Section 5.1.1.1, the applicant lists Class I structure, system, and components (SSCs). Among the list is the EFW Tank Enclosure, which corresponds to the Dedicated EFW Tank Enclosure Building on LRA Figure 2.2-1. Shown next to this Class I structure, are the RB Maintenance Building and the HP Office. However these structures are shown as not in scope of license renewal. Due to their proximity, these structures could have the potential to interact with the adjacent Class I structure; and if so would be included in scope of license renewal under 10 CFR 54.4(a)(2).

U. S. Nuclear Regulatory Commission 3F0909-03 Page 4 of 12 Request:

Explain the exclusion of the RB Maintenance Building and HP Office structures from scope of license renewal per 10 CFR 54.4.

CR-3 Response:

The Reactor Building Maintenance Support Building is a non-safety related sheet metal structure, supported by a structural steel frame on a concrete slab. The Health Physics (HP)

Office is a non-safety related concrete block structure on a concrete slab. For License Renewal scoping, the HP Office was included as part of the Reactor Building Maintenance Support Building and not as a separate structure. The purpose for the Reactor Building Maintenance Support Building (including the HP Office) is to provide support for maintenance activities during outages, a machine shop during non-outage periods, and office space for Health Physics personnel supporting maintenance activities.

The Reactor Building Maintenance Support Building (including the HP Office) was designed and constructed using the plant modification review process or Modification Approval Record (MAR).

The potential failure of the Reactor Building Maintenance Support Building (including the HP Office) was considered in the MAR.

Per the MAR, the Reactor Building Maintenance Support Building (including the HP Office) was designed to wind loads of 110 mph consistent with other Class III plant structures at CR-3. The MAR also states, "The failure of this building will not impact other safety related structures or components." Expansion devices between the Reactor Building Maintenance Support Building (including the HP Office) and the EFW Tank Enclosure Building were provided to allow for differential movement. Ample clearances or expansion joints were provided between concrete, masonry wall sections, structural steel framing and the EFW Tank Enclosure Building. Special architectural tie-in details for differential movement were used for sealing between the Reactor Building Maintenance Support Building (including the HP Office) and the EFW Tank Enclosure Building with flashing, sealant and use of minimal length anchorages. There is no structural connection between the Reactor Building Maintenance Support Building (including the HP Office) and the EFW Tank Enclosure Building.

Since the Reactor Building Maintenance Support Building (including the HP Office) is designed as a separate free standing structure and incorporates design details to structurally separate interaction with the EFW Tank Enclosure Building, the Reactor Building Maintenance Support Building (including the HP Office) was excluded from the scope of License Renewal per 10 CFR 54.4.

In addition, there were no components supported by the Reactor Building Maintenance Support Building (including the HP Office) which were in the scope of License Renewal.

RAI 2.2-04: General Scoping Items

Background:

10 CFR 54.4(a) provides three criteria for determining whether systems or components are in scope for license renewal. The applicant follows their stated methodology to ensure that this regulation is met.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 5 of 12 Issue:

In FSAR, Chapter 1, Figure CR3-G86-D, shows an Outage Support Building adjacent to the Borated Water Storage Tank. However, LRA Figure 2.2-1, CR-3 Plant Structures, does not show this structure. Since this structure is adjacent to the Borated Water Storage Tank, which is a Class I structure, it has the potential to interact with the adjacent Class I structure; and if so would be included in scope of license renewal under 10 CFR 54.4(a)(2).

Request:

Explain the exclusion of the Outage Support Building from scope of license renewal per 10 CFR 54.4.

CR-3 Response:

The Outage Support Building was removed prior to submittal of the License Renewal Application and was therefore not discussed in the application. Plant drawing CR3-G86-D has been revised to show the Outage Support Building removed as part of the Engineering Change process. An FSAR change request was approved to revise the Chapter 1 figures that are based on drawing CR3-G86-D during the next update of the FSAR. The Outage Support Building was excluded because it has been removed from the site.

RAI 2.2-05: General Scoping Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

On LRA Figure 2.2-1, CR-3 Plant Structures, the applicant shows structures in light lines, denoting the structure in not in scope of license renewal. Among the structures the applicant depicts as not in scope are the Traveling Screens. In FSAR Chapter 5, Section 5.1.1.1, Class I, the applicant list Class I SSCs. Among the list is the Nuclear Steam Supply Systems intake structure, which corresponds to the Circulating Water Intake structure on LRA Figure 2.2-1. The traveling screens are a part of this Class I structure; however, they are shown as not in scope of license renewal.

Due to their proximity, these traveling screens have the potential to interact with the adjacent Class I structure. Therefore, the traveling screens should be included in the scope of license renewal under 10 CFR 54.4(a)(2). In addition, the travel screens may have a filtering function, which may require them to be in the scope of license renewal.

Request:

Explain the exclusion of the traveling screens from scope of license renewal per 10 CFR 54.4.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 6 of 12 CR-3 Response:

The traveling screens in question are identified in the PassPort Equipment Database (EDB) as CWTS-1A through -1G and CWTS-2.

Circulating water traveling screen CWTS-2 is a single, dedicated traveling screen system that filters all of the seawater flow conveyed by the B train of the Nuclear Service and Decay Heat Sea Water System. Train A seawater is drawn from a separate forebay area of the intake structure. The forebay area is common to all circulating water pumps, and water entering this area is filtered by seven large traveling screens (CWTS-1A through -1G).

The CR-3 scoping process is described in Section 2.1.1 of the LRA. Section 2.1.1.1 of the LRA states:

CR-3 components identified in PassPort EDB as safety related meet the criteria of 10 CFR 54.4(a)(1) and are in the scope of License Renewal unless specific evaluation and justification is provided to exclude them.

PassPort EDB does not identify these components as safety related. Therefore, the traveling screens are not within the scope of license renewal pursuant to 10 CFR 54.4(a)(1).

The scoping process pursuant to 10 CFR 54.4(a)(2) is described in Section 2.1.1.2 of the LRA.

It states in part:

CR-3 has made use of the CLB-based information regarding quality classification, functional data, and regulatory requirement data contained in PassPort EDB to identify SSCs that have functional or physical interactions with safety related SSCs.

The NRC previously questioned the safety classification of traveling screen CWTS-2 in Section 4.2.1.3 of the letter from S.A. Varga (NRC) to W.S. Wilgus (CR3), Inspection Report No. 50-302/87-22, dated December 30, 1987. It states:

The team questioned the licensee's non-safety classification of CWTS-2 in that a safety-related method may be required to prevent accumulated debris from reaching the suction of the B train raw water pumps. The FSAR stated that one intake conduit shares a common intake structure, bar racks, and traveling screens with the circulating water system and that the other intake conduit is supplied with a bar rack and separate traveling screen located in a separate intake structure. The team was concerned with the ability of the traveling screens to withstand debris loading during design bases events and remain structurally intact.

As debris accumulates at the surface of the screen, the pump suction continuously removes water from the pit causing the level in the pit to drop. As a result, debris may accumulate on the upstream side of the screen and start to form a dam, first at the surface then eventually drawn down the screens. As a consequence, an increasing differential pressure could be experienced across the screen.

A sufficiently large accumulation of debris will cause the screen to fail, permitting the accumulated debris to flow to the pump suction.

For train A of the emergency seawater pumps, debris may take a longer time to accumulate because the suction for this train is from a large forebay and through seven traveling screens. The team was informed that when pumps RWP-2B and RWP-3B operate simultaneously, the level in the pit is a little over 3 feet lower than the intake canal level.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 7 of 12 A specific safety classification review for CWTS-2 was provided to the NRC in a letter dated June 30, 1988 (R.C. Widell (CR-3) to S.A Varga (NRC),

Subject:

Crystal River Unit 3, Docket No. 50-302, Operating License DPR-72, Inspection Report 87-22). It states:

The Traveling Screen CWTS-2 is not classified as safety related because this component is not required to function to support safe shutdown of the plant using the Alternate Nuclear Service Seawater Cooling System (RW).

The letter of June 30, 1988 also provided the following discussion concerning failure of the screen:

Even if failure of the screen were to occur, the massed debris would enter the intake structure pre-chamber and float on the surface. At the minimum water level of 79 ft.

elevation, at least 7-1/2 ft. of submergence is maintained above the top of the 48-inch intake pipe to the pump suction chamber. With a suction pipe entrance velocity of -4.2 ft./sec. and 7-1/2 ft. of submergence, vortexing, will not occur. The debris will remain on the surface in the pre-chamber until shutdown of the alternate loop and cleaning of the chamber is feasible.

A review of the quality classification, functional data, and regulatory requirement data contained in PassPort EDB did not identify that the traveling screens were required to be within the scope of license renewal pursuant to 10 CFR 54.4(a)(2).

Additional reviews were performed to determine if there was either a functional dependency between the traveling screens and the Nuclear Service and Decay Heat Sea Water System (described in Section 2.3.3.49 of the LRA) or credible physical interactions between the traveling screens and the 48-inch intake conduits.

A review of industry operating experience was performed for the hypothetical failure of a traveling screen.

None was identified where the safety related function of a service water system was compromised. Per Section 3.1.2 of NEI 95-10:

Consideration of hypothetical failures that could result from system interdependencies that are not part of the CLB and that have not been previously experienced is not required.

Therefore, the traveling screens are not within the scope of license renewal pursuant to 10 CFR 54.4(a)(2).

The scoping process pursuant to 10 CFR 54.4(a)(3) is described in Section 2.1.1.3 of the LRA.

Scoping pursuant to 10 CFR 54.4(a)(3) is performed with a combination of reviews of PassPort EDB and topical reviews. These reviews did not identify the traveling screens as being within the scope of license renewal.

Therefore, the traveling screens are not within the scope of license renewal pursuant to 10 CFR 54.4(a)(3).

In conclusion, since the traveling screens do not meet any of the scoping criteria in 10 CFR 54.4, they are not within the scope of license renewal.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 8 of 12 RAI 2.2-06: General ScoDing Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

On LRA Figure 2.2-1, CR-3 Plant Structures, the applicant shows structures in dark lines, denoting the structure is in scope. of license renewal.

Among the structures the applicant depicts as in scope is the Machine Shop. LRA Section 2.4.1.14, Machine Shop, states that the Machine Shop structure intended function is to "Provide structural support and/or functional support to non-safety related components." Also, the Machine Shop structure is shown next to Class I structures, Control Building, Auxiliary Building, and Diesel Generator Building. Due to the proximity, the Machine Shop structure has the potential to interact with the adjacent Class I structure; and if so would be included in the scope of license renewal under 10 CFR 54.4(a)(2).

During the CR-3 plant audit on June 23, 2009, the applicant discussed a portion of the machine shop being in scope for license renewal to support the Appendix R equipment on the roof. The applicant's reasoning for excluding the remaining support structure of the machine shop from the scope of license renewal is because the failure of the supporting steel would be hypothetical in nature. However, this explanation conflicts with industry operating experience.

In addition, in the FSAR, Section 2.4.2.4, "Facilities Required for Flood Protection," the applicant describes equipment required to remain functional during a postulated hurricane to assure maintenance of the reactor in a safe condition. The applicant describes five large doors that have an inflatable-type seal that serves as a back-up in the unlikely event of a compression-type seal failure, one of which describes a water-tight door into the Hot Machine Shop. CR-3 FSAR Figure 2.30 shows water-tight doors, but not the machine shop. It is not clear if this component is physically located in the machine shop structure or other structure.

Request:

Explain the exclusion of portions of the Machine Shop, as discussed during the June 23, 2009 audit, and the Hot Machine Shop structures from scope of license renewal per 10 CFR 54.4.

CR-3 ResDonse:

After further review, CR-3 has decided to include the remainder of the Machine Shop support structure in the scope of License Renewal and make Criterion 10 CFR 54.4(a)(2) applicable to the Machine Shop.

These changes will revise LRA Subsection 2.4.2.14, Table 2.4.2-14, Subsection 3.5.2.1.15, Table 3.5.1, Table 3.5.2-15, Subsection A.1.1.29, and Subsection B.2.29. The specific revisions to the LRA are described in Enclosure 3 to this submittal.

On Table 2.4.2-14, commodities were added for the supporting structure. A Fire Door which is actually located in the Ready Warehouse was deleted from the Machine Shop.

The water-tight door shown on FSAR Figure 2-30 is attached to the exterior walls of the Auxiliary Building. The water-tight door is closed to seal flood water from entering the Auxiliary Building. The water-tight door is included in the Component/Commodity Door (Non-Fire) in LRA

U. S. Nuclear Regulatory Commission 3F0909-03 Page 9 of 12 Table 2.4.2-1, Component Commodity Groups Requiring Aging Management Review and Their Intended Functions: Auxiliary Building, with a Flood Barrier intended function.

RAI 2.2-07: General ScoDinc Items

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

In LRA Section 2.2, Table 2.2-1, the applicant lists systems in the scope of license renewal.

The Reactor Building Pressure Sensing and Testing System was listed as not in the scope of license renewal. The applicant does not provide an explicit explanation of what components comprise this system. The applicant did include the Leak Rate Test System in the scope of license renewal, which seems to have the same function as the Reactor Building Pressure Sensing and Testing System. Therefore, the Reactor Building Pressure Sensing and Testing System should be included in scope of license renewal under 10 CFR 54.4.

Request:

Explain the exclusion of the Reactor Building Pressure Sensing and Testing System from scope of license renewal per 10 CFR 54.4.

CR-3 Response:

The Reactor Building Pressure Sensing and Testing System, as listed in Table 2.2-1, is identified in the CR-3 Equipment Database (EDB) as a separate system, but this system is not populated with components in the EDB. The components associated with the Reactor Building pressure sensing function are included within the Reactor Building Spray System at CR-3, and are in the scope of License Renewal as depicted on 302-712-LR, Sheet 1.

RAI 2.3.3.33-01: Emergency Diesel Generator System

Background:

10 CFR 54.4(a) provides criteria for determining whether systems or components are in scope for license renewal.

Issue:

The applicant identifies the AFW Pump as being in scope of license renewal and is highlighted as such on LRA drawing 302-081-SH-004. However, the power sources for the AFW pump, the Alternate AC (AAC) diesel and building, are not identified as in scope per LRA Section 2.3.3.33 and structures Table 2.2-2. In the CR-3 FSAR, Section 10.6.1, the AFW pump is designed to provide an additional non-safety grade source of secondary cooling to the OTSGs, should a loss

U. S. Nuclear Regulatory Commission 3F0909-03 Page 10 of 12 of all main and EFW occur. This "AFW source" was added in response to NRC concerns on the issue of EFW reliability (Generic Issue 124 and SRP Section 10.4.9).

In addition, the AAC may be credited as backup power supply to the emergency diesel generator (EDG) in the event of a loss of all AC. License Amendment 207, regarding technical specification change request for EDG allowed outage time extension (from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days), issued June 13, 2003, indicates that "AAC Diesel is intended to provide defense in depth during EDG online maintenance and other times when it is available. The AAC Diesel will be capable of carrying the loads required for safe shutdown, including maintaining adequate voltage and frequency such that the performance of safety systems is not degraded."

The technical evaluation for this licensing amendment included a probabilistic safety assessment evaluation which incorporated the availability of the AAC Diesel.

License Amendment 228, issued on December 26, 2007, by the NRC, involved the measurement uncertainty recapture power uprate, which referenced the AAC Diesel. The applicant noted that the AAC Diesel can be aligned to either safety-related AC distribution bus.

Though it is noted in the CR-3 FSAR that the AAC Diesel does not have a Station Blackout function, it is evidently relied upon in subsequent license amendment requests to provide defense in depth for the Emergency Diesel Generator System. Also, in LRA Figure 2.2-1, CR-3 Plant Structures, the applicant shows structures in light lines, denoting the structure in not in scope of license renewal. Among the structures the applicant depicts as not in scope is the AAC Diesel Generator Building.

Request:

Explain the exclusion of the AAC Diesel System and AAC Diesel Generator Building from scope of license renewal per 10 CFR 54.4.

CR-3 Response:

Consistent with the CR-3 scoping process, the scoping result of the Alternate AC (AAC or Aac)

Diesel Generator was based on a comparison of design and license basis requirements to the License Renewal scoping criteria of 10 CFR 54.4:

The AAC Diesel Generator is not safety related, and is not in scope under 10 CFR 54.4(a)(1).

Regarding interactions between non-safety related and safety related SSCs:

The AAC Diesel Generator is capable of providing power to auxiliary feedwater pump FWP-7, but FWP-7 itself is non-safety related. FWP-7 was installed to resolve Generic Safety Issue (GSI) 124 at CR-3, but the design as accepted to resolve the GSI utilized power from non-safety Reactor Auxiliary Switchgear Bus 3, without reliance on diesel generator backup.

Neither the AAC Diesel Generator nor its predecessor (a trailer-mounted diesel generator) was installed at that time, nor were they the result of CR-3 commitments associated with the resolution of GSI 124. Since the resolution of GSI 124, CR-3 has installed an additional safety related diesel-driven Emergency Feedwater System pump. Thus CR-3 now has a total of three safety related pumps to supply the Once-Through Steam Generators (OTSGs) as described in LRA Subsection 2.3.4.8.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 11 of 12 License Amendment 207 states "The Aac source is intended to provide defense in depth during EDG online maintenance and other times when it is available and is not intended to be used to change the CR-3 licensing basis for compliance with SBO." As approved in License Amendment 207, the use of an Aac source to extend EDG outage duration is fully elective on the part of CR-3, and License Amendment 207 notes that the criteria for an Aac source could be met with temporary or permanent options. There are no Limiting Conditions for Operation for the AAC Diesel Generator in the CR-3 Technical Specifications.

In addressing the adequacy of the EDGs for measurement uncertainty recapture power uprate, License Amendment 228 based its acceptance on existing diesel generator margin and the conclusion that the power uprate would not significantly change diesel generator loading. The AAC Diesel Generator, while referenced, was not credited with any specific functional requirements in this regard.

The AAC Diesel Generator is not required to supply any accident loads, or safe shutdown loads in the event of a fire or seismic event.

Based on these considerations, the AAC Diesel Generator and associated support systems are not included in the scope of License Renewal under 10 CFR 54.4(a)(2).

The AAC Diesel Generator is not credited in CR-3 CLB evaluations with satisfying the requirements of any regulated events identified in 10 CFR 54.4(a)(3).

Summarizing, the AAC Diesel Generator does not satisfy any of the criteria of 10 CFR 54.4.,

and so has not been included in scope of license renewal. Similarly, the AAC Diesel Generator Building is not included in the scope of License Renewal on the basis there is no intended function associated with protecting / supporting the AAC Diesel Generator.

RAI 2.3.3.35-01: Fuel Handling System Backqround:

10 CFR 54.21 (a)(1) requires the applicant to provide a list of structures and components subject to an AMR. The staff reviews the LRA, FSAR, and license renewal boundary drawings to verity that list of components provided for each system is complete.

Issue:

LRA Section 2.3.3.35, Fuel Handling System, states: "There are no License Renewal scoping drawings that depict the Fuel Handling System." LRA drawing 302-621 depicts the fuel transfer canal (B-2). The CR-3 LRA Table 2.3.3-35 lists piping and components for the Fuel Handling system.

Request:

Verify that LRA drawing 302-621 depicts all the components for the fuel handling system that are in scope and excluded from scope.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 12 of 12 CR-3 Response:

Subsection 2.3.3.35 of the LRA correctly states that:

The Fuel Transfer Tubes are categorized as mechanical components, the remaining cranes, gates, and racks are civil/structural components and, for License Renewal, are addressed with the structure in which they are located.

The fuel transfer tubes are the mechanical components included in the Containment isolation piping and components identified in LRA Subsection 2.3.3.35.

The mechanical portion of the Fuel Handling System, consisting of the two fuel transfer tubes, is depicted on drawing 302-621-LR, Sheet 1, at coordinate B-2. Note that the piping, valves and fittings associated with the Fuel Transfer Tubes are included in the Spent Fuel Cooling System.

As depicted on 302-621-LR, Sheet 1, the two Fuel Transfer Tubes and associated piping, valves, and fittings are within the scope of license renewal.

In summary, LRA Subsection 2.3.3.35 should have stated:

The License Renewal scoping boundaries for the Fuel Handling System are shown on the following scoping drawing. (Scoping drawings have been submitted separately for information only.) 302-621-LR, Sheet 1 instead of:

There are no License Renewal scoping drawings that depict the Fuel Handling System.

Therefore, this RAI response requires an amendment to the LRA. Refer to Enclosure 3 of this submittal.

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (REFERENCE 3)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 1 of 5 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (REFERENCE 3)

RAI 2.3.3.1-1:

The Air Handling Ventilation and Cooling System description states air reservoirs/accumulators are included in the system. Neither Table 2.3.3-1 nor in Table 3.3.2-1 include air reservoirs/accumulators. Provide a discussion, or indicate where in the application there is a discussion, justifying that the air reservoirs/accumulators are not in scope as components subject to aging management review as a pressure boundary.

CR-3 Response:

Air reservoirs/accumulators are small air-filled tanks utilized in the Air Handling Ventilation and Cooling System. These components are in scope and subject to aging management review with the pressure boundary intended function.

The Air Handling Ventilation and Cooling System air reservoirs/accumulators are included in the component/commodity identified as Piping, piping components, piping elements, and tanks identified in LRA Table 2.3.3-1 and LRA Table 3.3.2-1, page 3.3-101.

RAI 2.3.3.3-1:

Drawing 302-751-LR Sheet 1 shows the Reactor Building Reactor Cavity Cooling System heat exchanger in scope for licensing renewal. Table 2.3.3.3-3 indicates that only the housing of the heat exchanger is in scope subject to an aging management review as a pressure boundary. Provide a discussion, or indicate where in the application is the discussion, justifying the tubing of the heat exchanger is not in scope for aging management review as a pressure boundary.

CR-3 Response:

The Reactor Building Cavity Cooling subsystem non-safety related cooling coil heat exchanger housings are in the scope of license renewal as indicated in LRA Table 2.3.3.3-3.

The subject heat exchanger cooling coils (tubing) form part of the Reactor Building pressure boundary, and the water is supplied by the Industrial Cooling System as identified in LRA Section 2.3.3.3 (paragraph three page 2.3-33) and LRA Section 2.3.3.24.

The cooling coils (tubing) of the Reactor Building Cavity Cooling heat exchangers are in scope with a pressure boundary intended function.

The cooling coils (tubing) are included in the Industrial Cooling System component/commodity identified as Containment isolation piping and components and as shown in LRA Table 2.3.3-24.

RAI 2.3.3.7-1:

The Decay Heat Closed Cycle Pump Cooling System description states air accumulators are provided to ensure adequate air volume is available to operate required pneumatic fan dampers. Provide a discussion, or indicate where in the application there is a discussion, justifying that the air accumulators are not in scope for aging management review as a pressure boundary.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 2 of 5 CR-3 Response:

Air accumulators are utilized in the functioning of the Decay Heat Closed Cycle Pump Cooling System pneumatically operated fan dampers. These air accumulators are in scope and subject to aging management review with the pressure boundary intended function.

In the plant database, these air accumulators are included as components in the Air Handling Ventilation and Cooling System. Based on their location in the plant database, the subject air accumulators are included in the component/commodity identified as

Piping, piping components, piping elements, and tanks identified in LRA Table 2.3.3-1 and LRA Table 3.3.2-1, page 3.3-101.

RAI 2.3.3.8-1:

The Spent Fuel Coolant Pump Cooling System description states that air accumulators are provided to ensure adequate air volume is available to operate required pneumatic fan dampers. Provide a discussion, or indicate where in the application there is a discussion, justifying that the air accumulators are not in scope for aging management review as a pressure boundary.

CR-3 Response:

Air accumulators are utilized in the functioning of the Spent Fuel Coolant Pump Cooling System pneumatically operated fan dampers. These air accumulators are in scope and subject to aging management review with the pressure boundary intended function.

In the plant database, these air accumulators are included as components in the Air Handling Ventilation and Cooling System. Based on their location in the plant database, the subject air accumulators are included in the component/commodity identified as Piping, piping components, piping elements, and tanks identified in LRA Table 2.3.3-1 and LRA Table 3.3.2-1, page 3.3-101.

RAI 2.3.3.10-1:

The Auxiliary Building Exhaust System functions to limit the release of radioactivity to the environment. The redundant fans are provided. Automatic control dampers isolate the inactive fans. The auxiliary building exhaust fans are provided with inlet and outlet pneumatic dampers. The dampers have a required function to open to permit airflow when the fan is operating. The dampers also have a required function to close when the fan is stopped to prevent air recirculation to the operating fan.

If air accumulators are provided to assure the required air supply to the damper actuators, please discuss if they are in scope for license renewal for the function as a pressure boundary.

CR-3 Response:

The Auxiliary Building Exhaust System uses instrument air for automatic pneumatic control of system dampers to accomplish proper system functioning.

The Auxiliary Building Exhaust System does not require air reservoirs/accumulators to ensure functioning of associated system dampers in the event of loss of instrument air. The system dampers do not provide a safety function, only an operational function.

Air accumulators are not included in this system nor required for Auxiliary Building Exhaust System operation. Therefore, they are not in scope for license renewal for the function as a pressure boundary.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 3 of 5 RAI 2.3.3.11-1: Drawing 302-753-LR Sheet 1, coordinates F4, shows damper AHFD-25 as not in scope for license renewal.

Drawing 302-753-LR Sheet 2 coordinates El, shows damper AHFD-25 in scope for license renewal.

Please clarify if this damper is in scope for license renewal.

CR-3 Response:

Damper AHFD-25 is included in scope for license renewal as identified on drawing 302-753-LR, Sheet 2, coordinates El.

RAI 2.3.3.11-2:

Drawing 302-753-LR Sheet 1 shows AHU-33 to be in scope. Sheet 2 of the same drawing shows the same item, AHU-33, to be not in scope. Drawing 302-001-LR Sheet 1, Symbols, does not define the symbol. U.S. Nuclear Regulatory Commission (NRC) staff cannot determine the function of AHU-33 to identify if it should be in scope for license renewal. Provide clarification if AHU-33 is in scope for license renewal.

If it is determined to be not in scope, please include the justification.

CR-3 Response:

Bathroom exhaust fan AHU-33 is included in scope for license renewal as identified on drawing 302-753-LR, Sheet 1, coordinates C7, with a pressure boundary intended function.

RAI 2.3.3.11-3: For the Control Complex Ventilation System the NRC staff cannot determine from the drawings or from the descriptions in the application or the Final Safety Analysis Report (FSAR) if the pneumatic operated dampers shift to their safety position on the loss of air or on the application of air. If air accumulators are provided to assure the required air supply to the damper actuators, please discuss if they are in scope for license renewal for the function as a pressure boundary.

CR-3 Response:

Control Complex Ventilation System air accumulators provide air (in the event of loss of compressed air) to ensure required fan dampers fully re-open following a design basis event, as noted in LRA Section 2.3.3.11.

The air accumulators are in scope and subject to aging management review with the pressure boundary intended function.

The Control Complex Ventilation System air accumulators are included in the component/commodity identified as Piping, piping components, piping elements, and tanks identified in LRA Table 2.3.3-11 and LRA Table 3.3.2-11, pages 3.3-150 and 3.3-151.

RAI 2.3.3.12-1:

For the Emergency Diesel Generator Air Handling System the NRC staff cannot determine from the drawings or from the descriptions in the application or the FSAR if the pneumatic operated dampers shift to their safety position on the loss of air or on the application of air. If air accumulators are provided to assure the required air supply to the damper actuators, please discuss if they are in scope for license renewal for the function as a pressure boundary.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 4 of 5 CR-3 Response:

No air accumulators are required for operation of the Emergency Diesel Generator Air Handling System. The system's safety related dampers are required to open on an Emergency Diesel Generator Room Fan start signal. On loss of air supply, the system's safety related dampers fail open to their required safety position. Therefore, they are not in scope for license renewal for the function as a pressure boundary.

RAI 2.3.3.12-2: The Emergency Diesel Generator Air Handling System description indicates end baffles are installed to help assure the exhaust air from below the generator does not recirculate back to the cooling air inlets.

These help minimize generator heat rejection to adjacent electrical equipment/components and enhance the capability of the ventilation system to maintain the room temperatures within acceptable limits. Please discuss, or indicate where in the application it is discussed, why the baffles are not in scope for license renewal.

CR-3 Response:

In 1997, a concern was identified for recirculation of exhausted cooling air from the emergency diesel generator (EDG) to the generator air inlets. In addition, the generator exhausted air was blowing on nearby electrical cabinets.

A design was installed which incorporated steel baffle plates and exhaust ductwork to vent the exhausted cooling air above and away from the generator air inlets and the nearby electrical cabinets.

The EDGs are in the scope of license renewal. The end baffles are steel plates mounted to the generator housing skid.

The generator housing and the baffle plates are integral to the generator and are scoped as part of the generator.

The exhaust ductwork installed as part of the new design and associated with the generator end baffles is included in the Emergency Diesel Generator Air Handling System and in the scope of license renewal.

This ductwork is subject to aging management review with the pressure boundary intended function and included in the Emergency Diesel Generator Air Handling System component/commodity Ductwork and components as shown in LRA Table 2.3.3-12 and LRA Table 3.3.2-12, page 3.3-153.

RAI 2.3.3.14-1:

FSAR Section 9.7.2 states that the switchgear room smoke detectors and temperature switches in the return duct close fire dampers.

Drawing 302-754-LR Sheet 1, coordinates A8, shows the fire damper on the discharge of the switchgear room (AHFD-40) is in scope and that inlet fire damper (no equipment number shown) is not in scope. Provide a discussion, or indicate where in the application the discussion is located, justifying inlet fire damper (no equipment number shown) housing is not in scope for license renewal.

CR-3 Response:

On drawing 302-754-LR, Sheet 1, coordinates A8, the fire damper tagged as Auto Fire Damper (4 Req'd) should have been highlighted. This fire damper symbol represents four fire dampers.

To indicate the four fire dampers are in scope, the highlighting was 5laced around the box

U. S. Nuclear Regulatory Commission 3F0909-03 Page 5 of 5 identifying AHFD-38, AHFD-48, AHFD-49, and AHFD-50.

The four fire dampers are in the scope of license renewal.

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ENCLOSURE 3 AMENDMENT #2 CHANGES TO THE LICENSE RENEWAL APPLICATION

U. S. Nuclear Regulatory Commission 3F0909-03 Page 1 of 8 Amendment #2 Changes to the License Renewal Application Source of Change License Renewal Application Amendment #2 Changes RAI 2.3.3.35-01 In the FSAR and Drawing References subsection for the Fuel Handling System on page 2.3-93, replace the statement There are no License Renewal scoping drawings that depict the Fuel Handling System.

with The License Renewal scoping boundaries for the Fuel Handling System are shown on the following scoping drawing. (Scoping drawings have been submitted separately for information only.)

302-621-LR, Sheet 1 Telecon with Delete the following sentence from the summary paragraph of subsection 4.3.4, "RCS NRC staff dated Loop Piping Leak-Before-Break" on page 4.3-15 and from the corresponding July 23, 2009 paragraph in Subsection A. 1.2.2.11 on page A-37:

In addition, recent NRC concerns related to Alloy 82/182 and LBB analyses are addressed in the industry's submittal MRP-140, "Materials Reliability Program: Leak-Before-Break Evaluation for PWR Alloy 82/182 Welds," EPRI, Palo Alto, CA: 2005, 1011808.

RAI 2.2-06 Revise the Description portion of LRA Subsection 2.4.2.14 on page 2.4-37 to read:

Description The Machine Shop is a two story structural steel and sheet metal building adjacent to the Turbine Building, Control Complex, and Auxiliary Building on the west side, the Diesel Generator Building on the south side, and the Ready Warehouse on the east side. The Machine Shop includes the "Clean Machine Shop" on the North end and a "Hot Machine Shop" on the South end separated with a masonry wall.

The Machine Shop is a Class III structure. Calculations have indicated that for Class III structures the wind imposes greater load than does an earthquake loading such that seismic loads need not be considered. As such, there are no seismic interaction issues. In addition, the Machine Shop construction drawings show the Machine Shop as a free standing structure only attached to the surrounding Class I structures with architectural details ensuring a water tight structure. However, 10 CFR 54.4(a)(2) functions will be assigned to the Machine Shop due to its proximity to Class I structures. The Machine Shop contains components required to support regulated events associated with Fire Protection. An Appendix R chiller is supported on the roof of the Machine Shop. The chiller is credited for Control Complex cooling if a fire should disable the normal HVAC cooling in the Control Complex. The support steel for the chiller and associated electrical conduit, panels, and enclosures are included in the scope of License Renewal. The supports for the fire protection piping inside the Machine Shop are also included in the scope of License Renewal. The structure commodities provide support or protection to the Fire Protection system and Appendix R Chilled Water components.

(continued)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 2 of 8 Source of License Renewal Application Amendment #2 Changes Change RAI 2.2-06 Based on the results of the CR-3 scoping and screening review, the Machine Shop (continued) performs the following intended functions:

C-7 Structural Support for Criterion (a)(2) and (a)(3) components-The Machine Shop is in the scope of License Renewal because it contains:

1. SSCs which are non-safety related whose failure could prevent satisfactory accomplishment of the safety related functions, and

2.

SSCs that are relied on during postulated fires.

Replace existing Table 2.4.2-14 with the revised Table 2.4.2-14 that follows this table.

Revise the Materials, Environments, Aging Effects, and Aging Management Programs for the Machine Shop in LRA Subsection 3.5.2.1.15 as follows:

Materials The materials of construction for the Machine Shop components are:

• Aluminum Carbon Steel Elastomers Reinforced Concrete Concrete Block Environment The Machine Shop components are exposed to the following:

Air-Indoor Air-Outdoor Reinforced Concrete Soil Aging Effects Requiring Management The following Machine Shop aging effects require management:

Loss of Material Reduction or Loss of Isolation Function Cracking Change in Material Properties Reduction in concrete anchor capacity due to local concrete degradation Aging Management Programs The following AMPs manage the aging effects for the Machine Shop components:

Structures Monitoring Program Masonry Wall Program (continued)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 3 of 8 Source of Chanae License Renewal Application Amendment #2 Changes RAI 2.2-06 (continued)

Add the Machine Shop to the list of structures managed by the Masonry Wall Program in the remarks column of line item 3.5.1-43 on LRA Table 3.5.1, page 3.5-47.

Add the Machine Shop to the list of structures managed by the Structures Monitoring Program in the remarks column of line item 3.5.1-44 on LRA Table 3.5.1, page 3.5-47.

Replace existing Table 3.5.2-15 with the revised Table 3.5.2-15 that follows this table.

Revise LRA Subsection B.2.29 by adding the Machine Shop to the list of structures in the Program Description, and revise Subsection B.2.29, Enhancements, Scope of Program to read:

Scope of Program Revise Program administrative controls to:

1) Identify the structures that have masonry walls in the scope of License Renewal, and

2)

Include inspection of the masonry walls in the Machine Shop in a periodic engineering activity.

This enhancement impacts License Renewal Commitment #19. Therefore, revise Commitment 19 on the list of CR-3 License Renewal Commitments, Enclosure 1 to CR-3 to NRC letter 3F1208-01, dated December 16, 2008, "Crystal River Unit 3 -

Application for Renewal of Operating License," as follows:

19 Program administrative controls will A. 1.1.29 Prior to the Masonry Wall be enhanced to (1) identify the period of Program structures that have masonry walls in extended the scope of License Renewal, and operation LRA Section.

(2) include inspection of the masonry B.2.29 walls in the Machine Shop in a periodic engineering activity.

RAI 2.2-06 Revise LRA Subsection A. 1.1.29, Masonry Wall Program, on LRA page A-16, to include the Machine Shop in the list of structures in the first paragraph. Also, add an enhancement by revising the second paragraph to read:

Prior to the period of extended operation, Program administrative controls will be enhanced to (1) identify the structures that have masonry walls in the scope of License Renewal, and (2) include inspection of the masonry walls in the Machine Shop in a periodic engineering activity.

U. S. Nuclear Regulatory Commission 3F0909-03 Page 4 of 8 TABLE 2.4.2-14 COMPONENT COMMODITY GROUPS REQUIRING AGING MANAGEMENT REVIEW AND THEIR INTENDED FUNCTIONS:

MACHINE SHOP Component/Commodity Intended Function(s)

(See Table 2.1-1 for function definitions)

Anchorage / Embedment C-7, Structural Support for Criterion (a)(2) and (a)(3) components Cable Tray, Conduit, HVAC Ducts, C-7, Structural Support for Criterion (a)(2) and (a)(3) components Tube Track (associated with the Appendix R Chilled Water System)

Concrete: Above Grade C-7, Structural Support for Criterion (a)(2) and (a)(3) components Concrete: Below Grade C-7, Structural Support for Criterion (a)(2) and (a)(3) components Concrete: Foundation C-7, Structural Support for Criterion (a)(2) and (a)(3) components Fire Hose Stations (associated with C-7, Structural Support for Criterion (a)(2) and (a)(3) components Fire Service piping)

Masonry Walls: All C-7, Structural Support for Criterion (a)(2) and (a)(3) components Racks, Panels, Cabinets, and C-7, Structural Support for Criterion (a)(2) and (a)(3) components Enclosures for Electrical Equipment and Instrumentation (associated with the Appendix R Chilled Water System)

Roof-Membrane / Built-up C-7, Structural Support for Criterion (a)(2) and (a)(3) components Seals and Gaskets C-7, Structural Support for Criterion (a)(2) and (a)(3) components Siding C-7, Structural Support for Criterion (a)(2) and (a)(3) components Steel Components: All Structural C-7, Structural Support for Criterion (a)(2) and (a)(3) components Steel Supports for EDG, HVAC System C-7, Structural Support for Criterion (a)(2) and (a)(3) components Components, and Other Miscellaneous Equipment (associated with the Appendix R Chilled Water System)

Supports for Non-ASME Piping &

C-7, Structural Support for Criterion (a)(2) and (a)(3) components Components (associated with Fire Service piping)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 5 of 8 TABLE 3.5.2-15 CONTAINMENTS, STRUCTURES, AND COMPONENT SUPPORT -

SUMMARY

OF AGING MANAGEMENT EVALUATION - MACHINE SHOP Component/

Intended Aging Effect Requiring Aging Management Vume Table 1 Commodity Function Management Program Volume2 Item Item Anchorage/

C-7 Carbon Steel Reinforced None None J, 501 Embedment Concrete Cable Tray, C-7 Carbon Steel Air - Outdoor Loss of Material Structures Monitoring II1.B2-10 3.5.1-39 A

Conduit, HVAC (T-30)

Ducts, Tube Track Aluminum Air - Outdoor Loss of Material Structures Monitoring III.B2-7 3.5.1-50 A

(TP-6)

Concrete:

C-7 Reinforced Air - Outdoor Loss of Material Structures Monitoring III.A3-9 3.5.1-23 A

Above Grade Concrete Cracking (T-04)

Change in Material Properties Loss of Material Structures Monitoring III.A3-10 3.5.1-24 A

Cracking (T-06)

Change in Material Properties Air - Indoor Loss of Material Structures Monitoring III.A3-9 3.5.1-23 A

Cracking IIlI.A5-9 Change in Material (T-04)

Properties Loss of Material Structures Monitoring II1.A3-10 3.5.1-24 A

Cracking II1.A5-10 Change in Material (T-06)

Properties Reduction in concrete Structures Monitoring II1.B2-1 3.5.1-40 A

anchor capacity due to local I1.B3-1 concrete degradation I11.B5.1 (T-29)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 6 of 8 TABLE 3.5.2-15 (continued) CONTAINMENTS, STRUCTURES, AND COMPONENT SUPPORT -

SUMMARY

OF AGING MANAGEMENT EVALUATION - MACHINE SHOP Component/

Intended MnAging Effect Requiring Aging Management NUREG-1 801 Table 1 Commodity Function Material Environment Management Program Volume2 Item Notes I

I jItem Concrete: Below C-7 Reinforced Soil Cracking Structures Monitoring IlI.A3-3 3.5.1-28 A

Grade Concrete (T-08)

Loss of Material Structures Monitoring IlI.A3-4 3.5.1-31 A

Cracking (T-05)

Change in Material Properties Loss of Material Structures Monitoring III.A3-5 3.5.1-31 A

Cracking (T-07)

Change in Material Properties Change in Material Structures Monitoring Ill.A3-7 3.5.1-32 A

Properties I

(T-02)

Concrete:

C-7 Reinforced Soil Cracking Structures Monitoring II1.A3-3 3.5.1-28 A

Foundation Concrete (T-08)

Loss of Material Structures Monitoring II1.A3-4 3.5.1-31 A

Cracking (T-05)

Change in Material Properties Loss of Material Structures Monitoring II1.A3-5 3.5.1-31 A

Cracking (T-07)

Change in Material Properties Change in Material Structures Monitoring II1.A3-7 3.5.1-32 A

Properties (T-02)

U. S. Nuclear Regulatory Commission 3F0909-03 Page 7 of 8 TABLE 3.5.2-15 (continued) CONTAINMENTS, STRUCTURES, AND COMPONENT SUPPORT -

SUMMARY

OF AGING MANAGEMENT EVALUATION - MACHINE SHOP AgingNUREG-1801 Component/

Intended Material Environment Aging Effect Requiring Aging Management Volume 2 Table 1 Notes Commodity Function M

Management Program Item Item Fire Hose Stations C-7 Carbon Steel Air - Indoor Loss of Material Structures Monitoring I1.B5-7 3.5.1-39 C, 522 (T-30)

Masonry Walls C-7 Concrete Air - Indoor Cracking Masonry Wall II1.A3-11 3.5.1-43 A

Block 1I1,A5-11 (T-12)

Racks, Panels, C-7 Carbon Steel Air - Outdoor Loss of Material Structures Monitoring I1l.B3-7 3.5.1-39 A

Cabinets, and (T-30)

Enclosures for Electrical Equipment and Instrumentation Roof-Membrane C-7 Elastomer Air - Outdoor Cracking Structures Monitoring II1.A6-12 3.5.1-44 C, 529 Built-Up Change in Material (TP-7)

Properties Seals and Gaskets C-7 Elastomer Air - Indoor Cracking Structures Monitoring III.A6-12 3.5.1-44 C, 509 Change in Material (TP-7)

Properties Air - Outdoor Cracking Structures Monitoring II1.A6-12 3.5.1-44 C, 509 Change in Material (TP-7)

Properties Siding C-7 Aluminum Air - Indoor None None II1.B5-2 3.5.1-58 C, 522 (TP-8)

Air - Outdoor Loss of Material Structures Monitoring I1l.B4-7 3.5.1-50 C, 532 (TP-6)

Steel Components:

C-7 Carbon Steel Air - Indoor Loss of Material Structures Monitoring III.A3-12 3.5.1-25 A

All Structural Steel (T-11) 1

U. S. Nuclear Regulatory Commission 3F0909-03 Page 8 of 8 TABLE 3.5.2-15 (continued) CONTAINMENTS, STRUCTURES, AND COMPONENT SUPPORT -

SUMMARY

OF AGING MANAGEMENT EVALUATION - MACHINE SHOP Component/

Intended Aging Effect Requiring Aging Management NUREG-1801 Table 1 Commodity Function Material Environment Management Program Volumem2 Item Notes j

~ItemI Supports for EDG, C-7 Elastomer Air - Outdoor Reduction or Loss of Structures Monitoring I11.B4-12 3.5.1-41 A, 538 HVAC System Isolation Function (T-31)

Components, and Other Carbon Steel Air - Outdoor Loss of Material Structures Monitoring III.B4-10 3.5.1-39 A

Miscellaneous (T-30)

Mechanical Equipment Supports for Non-C-7 Carbon Steel Air - Indoor Loss of Material Structures Monitoring 1II.B2-10 3.5.1-39 A

ASME Piping &

(T-30)

Components Air - Outdoor Loss of Material Structures Monitoring I1.B2-10 3.5.1-39 A

I_

I (T-30)