ML20197A490

Forwards Integrated Plant Assessment Commodity & Sys Repts for Review & Approval IAW 10CFR50.54,license Renewal Rule

ML20197A490
Person / Time
Site:	Calvert Cliffs
Issue date:	03/03/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9803090264
Download: ML20197A490 (140)

Text

b Csitut.ss 11. Cutwe; Italtimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power i'lant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby. htaryland 20657 410 495-4455 March 3.1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Clifts Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50 317 & 50-318 Request for Review and Approval of Commodity and System Reports for Liccnie Renewal

REFERENCES:

(a)

Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated August 18,1995, Integrated Plant Assessment Methodology (b)

Letter from Mr. D. M. Crutchfield (NRC) to Mr. C. H. Cruse (BGE),

dated, April 4,1996, Final Safety Evaluation (FSE) Concerning The Baltimore Gas and Electric Company Report entitled," Integrated Plant Assessment Methodology" (c)

Letter from Mr. S. C. Flanders (NRC), dated March 4,1997," Summary of Meeting with Baltimore Gas and Electric Company (BGE) on BGE License Renewal Activities" This letter forwards the attached Integrated Plant Assessment (IPA) Comn Mity and System Reports for review and approval in accordance with 10 CFR Part $4, the license renewal rule. Should we apply for License Re.cwal, we will reference IPA Commodity and System Reports as meeting the requirements of 10 CFR 54.21(a), " Contents of application technical information," and the demonstration required by 10 CFR 54.29(a)(1), " Standards for issuance of a renewed license."

The information in this report is accurate as of the dates of the references listed therein. Per 10 CFR 54.21(b), an amendment or amendments will be submitted that identify any changes to the current licensing basis that materially afTect the content of the license renewal rpplication.

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Document Control Desk March 3,1996 l

Page 2 l

In Reference (a), Baltimore Gas and Electric Company submitted the IPA Methodology for review and approv '

in Reference (b), the Nuclear Regulatory Commission (NRC) concluded that the IPA Methodology is acceptable for meeting 10 CFR 54.21(aX2) of the license renewal rule, and if implemented, provides reasonable assurance that all structures and components subject to an aging management review pursuant to 10 CFR 54.21(a)(1) will be identified. Additionally, the NRC concluded that the methodology provides processes for demonstrating that the ef.ects of aging will be adequately managed pursuant to 10 CFR 54.21(aX3) that are cnnceptually sound and consistent with the intent of the license renewal rule.

In Reference (c), the NRC stated that if the format and content of these reports met the requirements of the template developed by BGE, the NRC could begin the technical review. This report has been produced and formatted in accordence with these guidance documents. We look forward to your 4

comments on the reports as they are submitted and your continued cooperation with our license renewai efYons.

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! e Document Control Desk March 3,1998 Page 3 Should you have quections regarding this matter, we will be plet.. ta discuss them with you.

W / truly yours, f

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STATE OF MARYLAND

TO WIT:

COUNTY OF CALVERT

1. Charles 11. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other BGE employees and/or consultants. Such informatior. has been reviewed in accordance with company practice and I believe it to.be reliable.

f 26 t-

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Spscribed and sworn before e,

Notary 4ublic in aM for the State of Maryland and County of I Q >uw f"

,this ay of /h41 AJ.1998.

WITNESS my lland and Notarial Scal:

Notary Public V

My Commission Expires:

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D'le e

CilC/DLS/ dim Attachments: (1) 3.3B Turbine Building Structure (2) 3.3C Intake Structure (3) 3.3D Miscellaneous Tank and Valve Enclosures (4) S.11B Primary Containment (S) S.15 Safetyinjection System cc:

R. S. Fleishman, Esquire

11. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC Director, Project Directorate 1-1, NRC R.1. McLean, DNR A. W. Dromerick, NRC J.11. Walter, PSC D. L. Solorio. NRC

i ATTACIIMENT (1)

APPENDIX A - TECHNICAL INFORMATION i

3.3B - TURBINE BUILDING STRUCTURE S

Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 3,1998 k

A'ITACHMENT (1) l APPENDIX A - TECliNICAL INFORMATION 3.3H - TURHINE BL.'LDING STRUCTURE 3.38 Turbine Juilding Structure This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Applicat w i

(LRA), addressing the Turbine Building Structure (henceforth called the Turbine Building). The Turbine Building was evaluated in accordance with the Calvert ClitTs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LRA.

3.3H 1 Scoping The system level scoping process defines conceptual boundaries for plant systems and structures, develops screening tools that capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify syrams and structures within the scope of license renewal. Systems and structures that are within the scope of license renewal are then scoped further on a component level. [ Reference 1, Section 3.0]

The component level scoping procese for systems is described in Section 4.1 of the CCNPP IPA Methodology, and the component level scoping process for structures is described in Section 4.2 of the methodology. Components with unique equipment identiners in the site equipment database are scoped using the component level scoping process for systems. Structural components such as walls do not have unique equipment identifie,s. Therefore, the component level :: coping process for structees utilizes a generic listing of gructural component types. [ Reference 1, Section 4.0]

The component level scoping process for structures identifies structural type components a= being within the scope oflicense renewal if they perform one or more of the following generic structural functions:

[ Reference 1, Section 4.2.2]

Provide structural and/or functional support to safety-related (SR) equipment; Provide shelter / protection to SR equipment; Serve as a pressure boundary or a Hssion product retention barrier to protect public health and safety in the event of any postulated Design Basis Events; Serve as a missile barrier (internal or external);

Provide structural and/or functional support to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions; Provide flood protection barrier (internal flooding event); and Provide a rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of e

the plant.

The remainder of Section 3.3B.) provides a description of the Turbine Building including the conceptual boundaries from the system level scoping results, the results of the component level scoping, and the results of the scoping to determine the components subject to aging management review (AMR).

Representative historical operating experience pertinent to aging is included in appropriate areas to provide insight supporting the aging management demonstrations. This operating experience was obtained through key-word searches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.

Application for License Renewal 3.3B-1 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (1)

APPENDIX A - TECHNICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE Structure neseriptiowconceptual Boundaries Figure 3.3B-1 is a simplified layout of site structures, including the Turbine Building, showing the structures that are within the scope of license renewal, ne CCNPP site arrangement consius of numerous struc,ures as shown on Figure 12 of the CCNPP Updated Final Safety Analysis Report (UFSAR). Design features of the CCNPP structures are discussed in UFSAR Cha;'ter 5. (Reference 2; Reference 3, Table 2; References 4,5, and 6]

The Turbme Building is oriented parallel to the Chesapeake Bay shoreline between the North Service Building (which is located on the east, or bay side) and the Auxiliary Building (which is located on the west, or landward side). The Turbine Building is common to CCNPP Units 1 and 2. The Turbine Building houses the turbine generators, codenscrs, feedwater heaters, condensate and feed pumps, turbine auxiliaries, and switchgear assembiv (Reference 2, Figure 12, Section 1.2.2]

The Turbine Building is a steel structure, with metal siding, supported on reinforced concrete foundations. The circulating water intake and discharge conduits are incorporated into the spread footings. The turbine-generators are separated by an expansion joint.n the superstructure. (Reference 2, Section 5.6.3; Reference 7, Section 1.1.1]

The Turbine Building ir a Seismic ' ategory 11 structure with the exception of the Auxiliary Feedwater (AFW) Pump Rooms, which are Sciaic Category 1. All of the structural steel columns, beams, and roof trusses of the building have been designed as indept 'ent members and in accordance with the American Institute of Steel Construction " Specification for the I sign, Fabrication and Erection of Structural Steel for Buildings," 1963 Edition. Two bridge cranes are located in the turbine-generator section of the building. The Units I and 2 turbine generators are mounted on their own concrete pedestals that project up through the building to the operating deck at Elevation 45' (Reference 2, Section 5.6.3; Reference 7, Section 1.1.1)

As shown in Figure 3.3B-1, electrical ductbanks that run under the Turbine Building are connected between the AFW Pump Rooms and the intake Structure. These ductbanks contain electrical conduits used for outing of the cables that power the Saltwater Pumps. He ductbanks are Seismic Category I and are constructed of reinforced concrete that encases the conduits. The ductbanks are sloped downward toward the Intake Structure to fac.litate drainage of any groundwater that may seep into the conduits. (Reference 2, Section S A.2.1; Refcrence 4]

The Turbine Building sidirig is classified as non-safety related, while the siding clips that hold the siding in place are classified as SR. The siding clips are designed to fail when the differential pressure acioss the siding reaches a pre-determined pressure. This design allows the siding to " blow-ofi" and thereby provide venting after a postulated break of a main steam line in the Auxiliary Building or the Turbine Building. The venting function is provided in order to protect vital equipment and structures.

(Reference 2, Page 10A.1-31; Reference 8, Page 77; Reference 9. Table 3S (Sheet 8)]

A wall at the end of the Main Steam Pipe Tunnel (se,)arating the Auxiliary Building and the Turbine Building) is designed to fail at 0.5 psi [ pounds per square inch] so that it will vent pressure into the Turbine Building if a main steam line breaks near the Main Steam Pipe Tunnel. The wall is also designed to fail when subjected to a hydraulic pressure of 3 feet of water from a main feedwater line rupture in the Main Steam Fiping Area. The method of failure is that 4 retainer clips (2K x 2% x 11 gauge angles) on the Application for License Renewal 3.3B-2 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECilNICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE Turbine Building side of the wall will fait plastically (i.e., go beyond the ultimate strength of the material) when a pressure of 0.5 psi or 3 feet of water is exerted on the wall from the Main Steam Pipe Tunnel side.

Since the retainer clips are the controlling mechanism of failure, they are cHfied as SR. [ Reference 2, Section 10A.l.20; Reference 8, Page 74]

For all major structures below finish grades, a heavy waterproofing membrane of 40 mils thickness is provided at the exposed face of the exterior walls and below the base slab. Rubber waterstops are also provided at all construction joints up to grade elevation. Subsurface drains are provided to lower the l

clevation of groundwater around the plant. [ Reference 2, Section SA.5]

The conceptual boundaries of the Turbine Building evaluation include the AFW Pump Rooms because they are Seismic Category I. As described in the CCNPP !PA Methodology, all CCNPP Category I structures are designated as SR; therefore, all Category I structures are screened as within the scope of license reuewal. The Turbine Building evaluation included the AFW Pump Rooms and their associated structural components, but did not include commodity items such as component supports as discussed below. The electrical ductbanks that run under the Turbine Building between the AFW Pump Rooms and the intake Structure are also included in the conceptual boundaries of this evaluation because they are Seismic Category 1. In addition, the Turbine Building siding clips and retainer clips are within the scope of license renewal because they are SR [ Reference 1, Section 3.4; Reference 2, Sections 5.6.3 and SA 2.1; Reference 3, Tables 1 and 2; Reference 9, Table 3S)

Component supports that are connected to the structural components are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA. Component supports are defined as the connection between a system, or component we 4 a system, and a plant structural member. An example of a component support is the fixed base that supprts a pump. The pump would be scoped with its respective system evaluation. The component support is the fixed base that connects the concrete equipmet.t pad to the pump. The fixed base is scoped with the Component Supports Commodity Evaluation and the concrete equipment pad is scoped with the evaluation for the structure, if anchor bolts are used, there is overlap between the Component Supports Commodity Evaluation and the evaluation for the structural component. Evaluations for structural components considered the effects of aging caused by the surrounding environment, while the Component Supports Commodity Evalution considered the effects of aging caused by the supported equipment (thermal expansion, rotating equipment, etc.) as well as the surrounding environment. Si!pports for structural components such as platform hangers are not " component supports"in this sense because any support for a structural component is itself a structural component and is ine!uded in the scope of its respective structure. [ Reference 10, Section 1.1.1]

Application for License Renewal 3.3B-3 Calvert Cliffs Nuclear Power Plant

ATTAcilMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE

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.O LICENSE RENEWAL FIGURE 3.33-1 CCNPP SITE STRUCTURES (SIMPLIFIED D7AGRAM - FOR INFORMATION ONLY)

Application for License Renewal 3.3B-4 Calven Cliffs Nuclear Power Plant

ATTACHMENT (1)

APPENDIX A - TECliNICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE Structure Scoping Results The Turbine Building is within the scope oflicense renewal based on 10 CFR 54.4(a). Six out of seven of the generic structural functions listed above are applicable to the Turbine Building as shown in Table 3.3D l.

The intended functions for the Turbine Building were determined based on the requirements of 654.4(a)(1), $54.4(a)(2), and $54.4(a)(3), in accordance wi'h CCNPP IPA Methodology

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Section 4.2.2. [ Reference 7, Section 1.l.3; Reference 9 Table it }

TABLE 3.3B-1 INTENDED FUNCTIONS OF STRUCTURES Function Applicable to Applicable Turbine 10 CFR 54.4(a)

Building?

• Criteria 1.

Provide structural and/or functional.iupport to SR Yes Q54.4(a)(1) equipment 2.

Provide shelter / protection to SR equipment Yes 54.4(a)(1) 3.

Serve as a pre:;sure boundary or a fission prwiuct retention No 54.4(a)(1) barrier to protect public health and safety in the event of any postulated Design Basis Events 4.

Serve as a missile barrier (internal or external)

Yes s54.4(a)(1) 5.

Provide structural and/or functional support to non-safety-Yes 54.4(a)(2) related equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions 6.

Provide Hood protection barrier (internal flooding event)

Yes

%54.4(a)(2) 7.

Provide rated fire barriers to confine or retard a fire from Yes Q54.4(a)(3) spreading to or from adjacent areas of the plant Functions are shown as being applicable if they apply to any portion of the structure (e.g., AFW Pump Rooms.

Comnonents Subject to AMR As discussed above, the component level scoping process for structures utilized a generic list of structural component types. The generic list started with structural component typer ontained in industry technical reports addressing containment structures and other Category I strusares. Other structural component types were added to the list to ensure completeness. Additionally, any structural component types that are unique to the particular structure being secped, such as the prestressed tendons in the Coatainment and the sluice gates in the intake Structure, are noted. These structural components

' vere combined into four structural categories based on tbir design and materials as follows:

(Reference 1, Section 4.2.3; Reference 9]

Concrete Components; Structural Steel Components; Architectural Components; and Unique Components.

e Application for License Renewal 3.3B-5 Calvert Cliffs Nuclear Power Plant l

1

4 ATTACHMFNT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE During the scoping process, those structur2l component types actually contained in the Turbine Building were Mentined, Within the four structural component categories,24 structural component types were determined to contribute to at least one of the Turbine Building intended functions listed in Table 3,3B-1.

Table 3,3B-2 lists these 24 structural component types and their associated intended functions, Structural component types that are part of the Turbine Building, but do not contribute to any of the ir. tended functions of the structure, are not listed in the table, All of the structural component types that were identined as requiring AMR for the Turbine Building (except for the ductbanks, building siding clips, and retainer clips) are associated with the AFW Pump Rooms, (Reference 7. Table 2-1 and Appendix K, Section 2,2; Reference 9 Table 3S]

TABLE 3.3B-2 STRUCTURAL COMPONENT TYPES REQUIRING AMR FOR THE TURBINE BUILDING Structural Camponr[t Type l

Applicable Function (s)

Concrete (including Reis forcing Steel)

Walls I, 2, 4, 6, 7 Ground Floor Slabs and liquipment Pads I,2,4,6,7 tilevated I'loor Stabs I, 2, 6, 7 Cast in Place Anchors /limbedments' l 2, 6, 7 Ductbanks I,2 Grout 1, 2, 6, 7 Fluid Retainmg Walls and Slabs 1, 2, 6, 7 Pmt-Installed Anchors' 4, 5 StructuralSteel licams'

l. 2, 7 Ilaseplates' I,2,4,5,7 Floor Framing

• 1, 2, 7 Platform llanFers'

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5 Decking

• I, 2, 7 Jet Impingement llarriers' 4

4 Floor Grating

• 5 Stairs and Ladders

• 5 Architectural Conponents lluildmg Siding Clips 2

Retainer Clips 2

Fire Doors, Jambs, and liardware' 2,6,7 Access Doors, Jambs, and liardware' 2,6,7 Caulking and Sealants 6, 7 Unique Congonents Watertight Doors' 2, 6, 7 Pipe Whip Restraints

• 2 Pipe Encapsulations (See note below) 2

(* ',

Asterisk in " Structural Component Type" column indicates that the component type is included under the heading " Steel Components" in Table 3.3D-3,

(#)

Numbers in " Applicable Function (s)" column correspond to the associated intended functions as listed in Table 3.3B-,1, Note:

Pipe encapsulations are scoped as part of the enclosing structure but are es aluated for the efTects of aging in the hiain Steam AhtR Report.

Application for License Renewal 3.3 B-6 Calvert Cliffs Nuclear Power Plant n

ATTACilMENT m APPENDIX A - TECliNICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE As discussed in the CCNPP IPA Methodology Section 5.4, all seven of the generic structural functions are considered to be passive. In addition, plant stiuctural components are not normally subject to periodic replacement programs. Herefore, structural components are considered to be long lived unless specific justification is provided to the contrary. Based on the above, all of the structural component types listed in Table 3.3B.2 are subject to AMR for the Turbine Building. (Reference 7, Section 3.0]

Baltimore Gas and Electric Company may elect to replace components for which the AMR identifies that further analysis or examination is needed. In accordance with the License Renewal Rule, components subject to replacement based on qualified life or specified time period would not be subject to AMR.

3.3B.2 Aging Management The list of potential Age-Related Degradation Mechanisms (ARDMs) identified for the Turbine Building components is given in Table 3.3B-3, with plausible ARDMs identilied by a check mark (/) in the appropriate column. [ Reference 7, Attachments I and 2, Appendices C, E, K, and 0] For efficiency in presenting the results of these evaluations in this report, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all components. Table 3.3D-3 also identifies the group to which each ARDM/ device type combination belongs. Exceptions are noted where appropriate. The following groups have been selected for the Turbine Building:

G oup 1: Includes caulking and scalants subject to weathering; and Group 2: Includes steel components subject to corrosion.

Application for License Renewal 3.3 B-7 Calvert Cliffs Nuclear Power Plant

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u ATTACHMENT fn APPENDIX A -' TECHNICAL INFORMATION 33B - TURBINE BUILDING STRUCTURE TABLE 33B-3 POTENTIAL AND PLAUSIBLE ARDMs FOR THE TURBINE BUILDLNG STRUCTF" E Ground Elevased enummr Siding Chps '

-Steel Potential ARDMs Concrete Eqwpment h

N Grout Fluid Retammg Floor g

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Walls Pads Walls and Slabs Components

• gg SW.

CIp IMing of Calcium ifydrnxide Aggressive Chemical Attack on Concrete Corrosion of Embedded Steet/Rebar Settlement Corrosion

/(2)

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/(1)

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• " Steel Components" represent all structural wmev.c.t types marked with an asterisk (*) in Table 33B-2

/ - Indicates plausible ARDM determination

(#)- Indicates the group (s) in which the ARDM/comnonent type combination is evaluated Application for License Renewal 33B-8 Calvert Cliffs Nucicar Power Plant

ATTACHMENT (1)

APPENDIX A - TECHNICAL INFORMATION 3JB - TURBINE BUILDING STRUCTURE Aging mechanisms that are not plausible are generally not discussed further in these BGE LRA sections, unless they are considered noteworthy. For the Turbine Building, settlement is considered noteworthy and is discussed below.

l An industry technical report concluded that settlement is a potentially significant ARDM for Category I structures. Settlement occurs both during construction and after congruction. The amount of settlement depends on the physical properties of the foundation material. These properties range from rock (with little or no settlement likely) to compacted soil (with some settlement expected). Settlement may occur during the design life of the structure from changes in environmental conditions, such as lowering of the groundwater table. Settlement can occur in two stages; elastic expansion and time-dependent settlement. Elastic expansion of the confined soil occurs due to excavation unloading and results in a slightly upward movement. During construction, the soil moves downward as load is applied. 'Ihis clastic movement should be small and is complete when construction is completed. It has no effect on the structure and is not considered an aging mechanism. The excavation unloading and structural loading cause a small change in the void retio of the soil. This change results in a small amount of time-dependent sett!cment. He settlement rate will decline after completion of construction. Concrete and steel structural members can be affected by differential settlement between supporting foundations, within a building, or between buildings. Severe settlement can cause misalignment of equipment and lead to overstress conditions within the structum. When buildings experience significant settlement, cracks in structural members, differential elevations of supporting members bridging between buildings, or both,.nay be visibly detected. Settlement was deter...ined to be not plausible for the CCNPP Turbine Building based on the following site specific justification: [ Reference 7, Appendix J, Sections 1.0 and 2.5; Reference i1, Section 5.1)

The foundation for Turbine Building is situated on an engineered soil structure consisting of compacted soil on top of the site's Miocene deposit. The Miocene soil is very dense to extremely dense and is capable of supporting loads on the order of 15,000 to 20,000 pounds per square foot (psf) with slight consolidation. The design contact pressure of the Turbine Building foundation is only 5000 psf. This contact pressure is about the same as the overburden pressure removed due to excavation.

[ Reference 2, Sections 2.7.3.2, 2.7.5, and 2.7.6.2; Reference 7, Appendix J, Section 2.5; Reference 12]

Nuclear Regulatory Commission IE Circular No. 81-08 discusses operating experience at a number of plants with respect to insufficient compaction of foundation and backfill material during plant construction. The insufficient compaction resthd in excessive settlement of plant structures at a number of sites. The N"C recommended act.ons included verification that quality assurance and quality control measures, including procedures, test results, inspection personnel, and audits, were in effect during construction to assure that the soil was adequately compacted.

The backfill supporting the CCNPP Turbine Building was placed and compacted to requirements for density, moisture, and layer thickness in accordance with the quality assurance provisions in a specification used during plant construction. A continuous program of soil testing during construction was used to assure compliance with the specification. The soil supporting the Turbine Building foundation was compacted to a density of 97 percent compaction based on the standard Proctor compaction method. [ Reference 12; Reference 13, Sections 2.0,4.0, and 5.0; Reference 14, Sections 2.0 and 10.0, Reference 15]

I l

Application for License Renewal 3.38-9 Calvert Cliffs Nuclear Power Plant

1 ATTACHMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE The excavation for the Turbine Building was below the groundwater table. A permanent pipe e

drain system surrounding the plant was installed during plant constmetion to minimize fluctuation of the groundwater table, thus providing stable geological conditions.

[ Reference 2, Section 2.7.3.2; Reference 7, Appendix J, Sections 2.4 and 2.5]

Since the Turbine Building is situated on an exceptionally dense soil, the structure tends to uniformly settle. Most of the predicted settlement is expected in terms of uniform settlement, which has no adverse effect on the structural components of the Turbine Building. Any differential settlement is expected to be small and have negligible effect on the Turbine Building structural components. (Reference 7, Appendix J, Section 2.4]

Based on the above, settlement was determined to be not plausible for any structural components of the

(

Turbine Building. This conclusion is supported by a walkdown of the Tt.rbine Building, performed in 1994, which found no indication of structural damage due to settlement. (Reference 7, Appendix J Section 3.0]

The following is a discussion of the aging management demonstration process for each group identified above. It is presented by group and includes a discussion of materials and environment, aging mechanism effects, methods of managing aging, aging management program (s), and aging management demonstration.

Group 1 -(caulking and scalants subject to weathering)- Materials and Environment Group 1 includes caulking and scalants nbiect to weathering. These structural components provide flood protection barriers (internal flooding event) and provide rated fire barriers to confine or retard a fire from spreading to or from adjacent areas of the plant. [ Reference 7, Appendix O, Section 2.2 The material requirements for the caulking and scalants used during construction of CCNPP were governed by a constructMn specification. Specific manufacturers and brand names (or approved equals) were specified for difTerent applications. [ Reference 16]

The caulking and scalants located indoors will be subject to the ambient conditions within the Turbine Building. The Turbine Building ambient temperature is controlled by a plant heating and ventilation system as descri' sed in UFSAR Chapter 9. The caulking and sealants located outdoors will be subject to the temperature and humidity changes, rain, snow, etc. expected at the CCNPP site. ', Reference 2, Section 9.8.2.4, Tabie 9-18; Reference 7, Appendix O, Section 2.1]

Group 1 -(caulking and sealants subject to weathering)- Aging Mechanism Effects Caulking and sealants that are exposed to ambient conditions (indoor or outdoor) are susceptible to degradation due to weathering. Exposure to sunlight (ultraviolet exposure), changes in humidity, ozone cycles, temperature and pressure fluctuations, and snow, rain, or ice contribute to the weathering ARDM.

The eft-ets of 'veathering on most caulking and sealant materials are evidenced by a decrease in elasticity (drying out), an increase in hardness, and shrinkage. [ Reference 7, Appendix L, Section 1.0]

Weathering was determined to be plausible for the Turbine Building caulking and sealants due to their exposure to the environmental conditions that contribute to this ARDM. [ Reference 7, Appenoix O, Sections 2.1 and 2.5]

Application for License Renewal 3.3 B-10 Calvert Cliffs Nuclear Power Plant

A'ITACIIMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE This aging mechanism, if unmanaged, could eventually result in the caulking and sealants not being able to perform their intended functions under current licensing basis (CLB) conditions. Therefore, weathering was determined to be a plausible ARDM for which the aging effects must be managed for the Turbine Building caulking and scalants. [ Reference 7, Appendix 0, Section 2.3]

Group 1 -(caulking and scalants subject to weathering)- Methods to Manage Aging MitigaO: Because weathering of caulking and scalants is affected by exposure to environmental conditions that are not feasible to control (e.g., light, heat, oxygen, ozone, v ster, radiation), there are no practical methods to mitigate its etTects.

Discoverv: Caulking and scalants degrade over time and should be replaced as needed. An inspecti)n program that provides requirements and guidance for the identification, inspection, and maintenance af caulking and scalants can ensure that their condition is maintained at a level that allows them to perform their intended functions. An effective program will provide for baseline inspection, along w th periodic future inspections at appropriate intervals, depending upon the degree of harshness of the environment the caulking or scalant is in, items that are in a harsh exterior ervironment would be inspected more frequently. This program would involve visual inspection and probing to determine that the caulking or scalant is satirfactorily attached to the surface and is flexible.

Group 1 -(caulking s{ sealants subject to weathering)- Aging Management Programs Mitiention: TF

..o CCNPP programs credited for mitigation of weathering.

Diy~

.aulking and sealants that perform a fire barrier function are managed under an existing prograra. The Penetration Fire Barrier inspection Program, implemented through CCNPP Survei!!ance Tc:,t Procedure (STP), STP-F-592-1/2, is adequate to manage the effects of aging for caulking and sealants that function as fire barriers without modification. [ Reference 7, Appendix 0, Section 2.6].

The purpose of STP-F-592-1/2 is to provide instructions for visual inspection of fire barrier penetration seals in fire area boundaries that protect safe bitdown areas in Units I and 2. The scope of this procedure is to visually inspect the following type of fire barrier penetration seals for operability: [ References 17 and 18, Sections 1.0 and 2.2]

Electrical conduit and cable tray penetration seals; lleating, ventilating, and air conditioning duct penetration seals (ducts without dampers); and e

Mechanical pipe and expansionjoint penetration seals.

Procedure STP-F-592-1/2 was developed based on CCNPP Technical Specifications 3.7.12 and 4.7.I2 a, 10 CFR Part 50, Appendix R, the CCNPP Fire Protection Plan, NRC Generic Letter 86-10, and various plant drawings. [ References 17 and 18, Section 3.1]

The procedure is currently performed at least once per 18 months in accordance with Technical Specification 4.7.12.a. The procedure requires that the fire barrier penetration seals be visually inspected to determine if they are operable based on specific criteria that were developed for each type of fire barrier component, in general, the procedure inspects the penetration seals for damace. c 1.cking, voids, and proper 6

Application for License Renewal 3.3 B-11 Calvert Cliffs Nu:: lear Power Plant

NITACHMFNT4) 7 APPENDIX A - TECHNICAL INFORMATION t

3.3B - TURBINE BUILDING STRUCTURE installation. He procedure provides separate " failure criteria" and " repair criteria." The " failure criteria" are used to determine if the pene5 tion seal is considered to be inoperable. The " repair criteria" are used to determine if the penetration seal is operable but in need of repair. [ References 17 and 18, Sections 2.1 and 6.0, and Attachmere.]

If a fire barrier penetration seal is determined to be inopeiable based on the procedure criteria, plant personnel determine if actions are required in accordance with Technical Specification 3.7.12.a.

in addition, any conditions adverse to quality discovered during the inspection are documented on Issue Reports in accordance with the CCNPP C,...mtive Actions Program. [ References 17 and 18, Section 6.5 and Attachment B]

f The Fire Protection Program at CCNPP (which includes STP-F-592-1/2) is subject to periodic internal assessment in accordance with the requirem ts in BGE's Quality Assurance Policy. Audits are required for the Fire Protection Program and implementing procedures every two years. In addition, an independent fire protection and loss prevention program inspection and audit utilizing either qualified offsite BGE personnel or an outside fire protection firm is also required every two years. The Quality Assurance Policy also requires an inspection and audit of the fire protection and loss prevention program by a qualified outside fire consultant at least once every three years. An auait and inspection performed in 1996 (using an outside consultant as well as BGE personnel) concluded that the CCNPP Fire Protection Program is providing a level of safety consistent with good fire protection practices and NRC regulatory criteria. The inspection included plant walkdowns of some of the fire barrier penetration seals.

No age-related degradation issues for the seals were identified. [ Reference 19, Section 10.18]

The Fire Protection Program also undergoes periodic inspection by the NRC as part of their routine licensee assessment activities. An inspection of the program in 1994 included a review of procedure STP-F-592-1 and a plant tour that included inspection of some of the fire barrier penetrations. The NRC concluded that the Fire Protection Program complies with program requirements provided in the Technical Specifications and licensing documents. [ References 20 and 21]

Operating experienc< related to this program has shown that aging is a minor contributor to fire barrier penetration seal failures at CCNPP. The greatest contributor to degradation of these seals is believed to be due to inadequacies in the original installation of the seat materials. For example, degraded seals have been found to be the result of incomplete installation of the seat material (i.e., openings left in the penetrations) or due to improper grout installation.

The corrective actions taken as a result of the Penetration Fire Barrier inspection Prcgram will ensure that the Turbine Buildmg caulking and scalants that perform a fire barrier function will remain capable of performing their intended function under all CLB conditions.

Caulking and scalants that are not fire barriers are typically replaced upon identification of their degraded condition. Visual examinations of the caulking and scalants in the plant concluded that an inspection program was needed to adequately manage the aging of these architectural components.

[ Reference 7, Appendix 0, Section 2.4]

For the caulking and scalants that do not perform a fire barrier function, a new CCNPP Caulking and Sealant inspection Program will provide requirements and guidance for the identification, inspection frequencies, and acceptance criteria for caulking and scalant used in the Turbine Building to ensure that Application for License Renewal 3.3 B-12 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE their condition is maintained at a level that allows them to perform their intended functions. The new program will establish acceptance criteria and require a baseline inspection to determine the material condition of the caulking and scalants for the Turbine Building. If unacceptable degradation exists,

]

corrective actions will be taken. A technical basis will be developed for determining the periodicity of future inspections. [ Reference 7, Attachment 8]

Group 1 -(caulking and scalants subject to weathering)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to weathering of caulking and scalants:

)

The caulking and scalants provide flood protection barriers (internal flooding event) and provide e

g rated fire barriers to confine or retard a fire from spreading to or from adjacent areas of the plant.

'Q Weathering was determined to be a plausible ARDM for the caulking and sealants. This ARDM, if unmanaged, cculd eventually result in the caulking and scalants not being able to.>erform iheir intended functions under CLB conditicas.

g For caulking and scalants that function as fire barriers, the Penetration Fire Barrier !nspection e

m Program performs periodic visual inspections of fire barrier penetration seals, and contains

.j acceptance criteria that ensure corrective actions v ill be taken such that the fire barrier intended function will be maintained.

For caulking and scalants that do not perform a fire barrier function, a new Caulking and Scalants inspection Program will conduct inspections to dete.. age-related degradation, and will contain acceptance criteria that ensure corrective actions will be taken such that the intended functions will be maintained.

Therefore, there is reasonable assurance that the effects of weathering will be adequately managed such that the caulking and scalants will be capable of performing their intended functions, consistent with the CLB, during the period of extended operation.

Group 2 -(steel components subject to corrosion)- Materials and Environment Group 2 includes the Turbine Building steel components marked with an asterisk in Table 3.3B-2. These components are all subject to corrosion. They each contribute to 'ne or more of the various Turbine Building intended functions as shown in Table 3.3B 2. [ Reference 7,.' ppendix K, Section 2.2]

Since corrosion was recognized as a potential degradation mechanism for all structural steel components of the Turbine Building, its effects were considered in the original design. As a roult, all exposed structural steel surfaces in the Turbine Building, except grating and metal decking, which are galvanized steel, were shop-painted or field-painted during plant construction.

[ Reference 7, Appendix K, Section 2.4]

The steel components located indoors will be subject to the ambient conditions within the Turbine Building. The Turbine Building ambient temperature is c..ntrolled by a plant heating and ventilation system as described in UFSAR Chapter 9. The steel components locawd outdoors will be subject to the temperature and humidity changes, rain, snow, etc. expected at the CCNPP site. [ Reference 2, Section 9.8.2.4, Table 9-18]

t Application for License Renewal 3.3 B-13 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE gup 2 -(steel components subject to corrosion)- Aging Mechanism Effects Steel corrodes in the presence of moisture and oxygen as a result of electrochemical reactions, initially, the exposed steel surface reacts with oxygen and moisture to form an oxide film as rust. Once the protective oxide film has been formed and if it is not disturbed by erosion, attemating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from saltwater, the atmosphere, or groundwater, increase the mte of corrosion by increasing the electrochemical activity.~ If steel is in contact with another metal that is more noble in the galvanic series, corrosion of the steel may accelerate. [ Reference 7, Appendix K Section 1.0]

Corrosion products such as hydrated oxides of iron (rust) form on exposed, unprotected surfaces of the steel and are readily visible. 'Ihe alTect:d surface may degrade to such an extent that visible perferation may occur. In the case of exposed surfaces of steel with protective coatings, corrosion may cause the protective costings to lose their ability to adhere to the corroding surface. In t. 3 case, damage to the coatings can be visually detected well in advance of significant degradation of the steel. [ Reference 7, Appendix K, Section 1.0]

An inspection of the AFW Pump Rooms, which are located inside the Turbine Building, was performed in 1994. The interior and exterior of the pump rooms were inspected and minor areas of rust on steel components were identified. [ Reference 7, Attachment 7]

This aging mechanism, if unmanaged, could eventually result in the steel components not being able to perform their intended functions under CLB conditions. Therefore, corrosion was determined to be a plausible ARDM for which the aging effects must be managed for the Turbine Building steel components.

[ Reference 7, Appendix K, Section 2.3]

Group 2 -(steel components subject to corrosion)- Methods to Manage Aging

- Mitigation: The effects of corrosion cannot be completely prevented, but they can be mitigated by minimizing the exposure of external surfaces of the steel components to an aggressive environment and protecting the external surfaces with paint or other protective coating. Coatings serve as a protective layer, preventing moisture and oxygen from directly contacting the steel surfaces.

Discoverv: The effects of general corrosion / oxidation of steel are detectable by visual inspection. A visual examination by a person familiar with the components can be used to determine general mechanical and structural condition and check for rust. Observing that significant degradation of protective coatings has not occurred is an effective method to ensure that corrosion ha: not affected the intended function of the structural component. Since the coating does not contribute to the components' intended functions, degradation of the coating provides an alert condition that triggers corrective action before the occurrence of corrosion that would affect the components' ability to perform their intended functions. The degradation of the protective coating that does occur can be discovered and monitored by periodically inspecting the steel structural components. Corrective action for failed protective coatings and any actual metal degradation can be carried out as necessary.

[ Reference 7, Appendix K, Section 3.0]

Application for License Renewal 3.38-14 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 3.3H - TURHINE BUILDING STRUCTURE Group 2 -(steel components sIbht to corrmion)- Aging Management Programs Mitigation: No programs are credited for mitigation. The exposed surfaces of structural steel components are covered by protective coatings that mitigate the effects of corrosion. The discovery prograrns discussed below verify that the protective coatings are maintained.

Discoverv: Calvert Cliffs Administrative Procedure MN-1-319, " Structure and System Walkdowns,"

provides for discovery of corrosion of steel (or conditions that would accelerate corrosion, such as x

pooled water) for the Turbine Building by performance of visual inspections during plant walkdowns.

The purpose of the program is to provide direction for the performance of structure and system walkdowns and for the documentation of the walkdown results. This program is applicable to the Turbine Building steel components. [ Reference 22, Sect %n 1.l]

Under this program, responsib> personnel perform periodic walkdowns of their assigned structures and systems. Walkdowns may also be performed as required for reasons such as: material condition assessments; system reviews before, during, and after outages; start-up reviews (i.e., when the system is initially pres.udzed, energized, or ploced in service); and as required for plant modifications.

[ Reference 22, Section 5.1)

One of the objectives of the program is to assess the conomon of the CCNPP structures, systema, and components such that any abnormal or degraded condition will be identiGed, documented, and corrective actions taken before the condition proceeds to failure of the structures, systems, and components to perform their intended functions. Conditions adverse to quality are documented and resolved by the CCNPP Corrective Actions Program. [ Reference 22, Sections 5.1.C,5.2.A.1, and 5.2.A.5]

The program provides guidance for identification of specific types of degradation or conditions when performing the walkdowns. Inspotion items related to aging management include the following:

[ Reference 22, Section 5.2 and Attachments I through 13]

items related to specific ARDMs such as corrosion; Effects that may have been caurd by ARDMs such as damaged supports; concrete degradation, e

anchor bolt degradation, or leakage of fluids; and Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage e

of fluids, presence of standing water or accumulated moisture, or inadequate support of components (e.g., missing, detached, or loose fasteners and clamps).

A structure performance assessment is currer.:ly required for Category I structures at CCNPP at least once every six years. The assessment includes a review of each structural component that could degrade the overall performance of the structure. The program will be modified to add guidance regarding approval authority for significant departures from the walkdown scope / schedule specified.

[ Reference 22, Section 5.3]

The corrective actions taken as a result of the program described above will ensure that the Turbine Building steel components will remain capable of performing their intended functions under all CLB conditions.

Application for License Renewal 3.3 B-15 Calvert Cliffs Nuclear Power Plant l

1

A'ITACHMENT q)

APPENDIX A - TECHNICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE Group 2 -(stecI components subject to corrosis,m - Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of steel components for the Turbine Building:

The Turbine Building steel components contribute to one or more of the various Turbine Building intended functions as shown in Table 3.3B-2.

The steel components are subject to corrosion due to the normal ambient environmental conditions. This ARDM, if unmanaged, could eventually result in the steel components not being able to perform their intended functions under CLB conditions.

Corrosion is mitigated by applying protective coatings to the steel components and by periodically examining the components for degradation of that coating or conditions that could accelerate degradation.

Calu rt Cliffs procedure MN-1-319 provides for periodic visual inspections of these components during walkdowns of the Turbine Building. If any degradation is found, the appropriate corrective actions are taken to ensure that the intended functions will be maintained.

Therefore, there is reasonable assurance that the effects of aging due to corrosion of steel will be managed such that the steel components of the Turbine Building will be capable of performing their intended functions, consistent with the CLB, during the period of extended operation.

3.38.3 Conclusion The aging management programs discussed for the Turbine Building are listed in the following table, These programs are (or will be for new programs) administratively controlled by a formal review and approval process. As demonstrated above, these programs will manage the aging mechanisms and their effects such that the intended functions of the Turbine Building will be maintained during the period of extended operation consistent with the CLB under all design loading conditions.

The analysis / assessment, corrective action, and confirmation / documentation process for license renewal is in accordance with QL-2, " Corrective Actions Program." QL-2 is pursuant to 10 CFR Part 50, Appendix B, and coveis all structures and components subject to AMR.

(

Application for L; cense Renewal 3.3 B-16 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE TABLE 3.3B-4 LIST OF AGING MANAGEMENT PROGRAMS FOR Tile TURBINE BUILDING STRUCTURE Program Credited As Existing CCNPP Technical Procedure Discovery of weathering effects for caulking and STP-F-592 1/2, " Penetration Fire scalants that function as fire barriers for the Barrier Inspection" Turbine Building. (Group 1)

Modified CCNPP Administrative Procedure Discovery of corrosion effects for steel MN-1-319, " Structure and System components in the Turbine Bailding. (Group 2)

Walkdowns" New Caulking and Sealant inspection Discovery of weathering effects for caulking and Program scalants that do not function as fire barriers for the Turbine Building. (Group 1)

Application for License Renewal 3.3 B-17 Calvert Cliffs Nuclear Power Plant

i NITACHMENT m APPENDIX A TECilNICAL INFORMATION 3.3B - TURBINE HUILDING STRUCTURE 3.38.4 References 1.

CCNPP " Integrated Plant Assessment Methodology," Revision 1, January 11,1996 2.

CCNPP Updated Final Safety Analysis Report, Revision 21 3.

CCNPP " Life Cycle Management System and Structure Screening Results, Revision 5, September 10,1997 CCNPP Drawing 61230, " Salt Water Systems Underground Ducts Plan and Sections,"

Revision 6, October 15,1990 5.

CCNPP Drawing 63874Sil0004, "SR Ductbank Under West Plant Road Plan," Revision 0, April 4,1995 6.

CCNPP Drawing 63874Sil0005, " Underground Conduit West of Turbine Building Plan,"

Revision 0, July 15,1996 7.

CCNPP " Aging Management Review Report for the Turbine Building Structure," Revision 3, February 12,1997 8.

CCNPP Engineering Standard ES-Oll," System, Structure and Component (SSC) Evaluation,"

Revision 2, September 15,1997 9.

CCNPP " Component Level Scoping Results for the Turbine Building Structure," Revision 2, February 12,1997 10.

CCNPP " Aging Management Review Report for Component Supports," Revision 3, February 4,1997 11.

Electric Power Research Institute Report TR-103842, " Class i Structures License Renewal Industry Report," Revision 1, July 1994 12.

CCNPP Drawing 60119," Compacted Fill Areas," Revision 0, April 24,1970 13.

Bechtel Specification No. 6750-C-4A," Specification for Placement and Control of Compacted Fill-CCNPP Units I and 2," Revision 3, August 7,1970 14.

Bechtel Specification No. 6750-C-Il-B," Specification for Testing of Concrete, Reinforcement and Soil-CCNPP Units I and 2," Revision 1, May 9,1975 15.

NRC IE Circular No. 81-08, " Foundation Materials," May 29,1981 16.

CCNPP Specification A-0010 (Bechtel Specification No. 6750-A-10), " Specification for Furnishing Delivery and Application of the Caulking and Sealants," Revision 1, March 3,1971 17.

CCNPP Technical Procedure STP-F-592-1, "Pene ration Fire Barrier Inspection," Revision 3, August 26,1997 18.

CCNPP Technical Procedure STP-F-592-2, " Penetration Fire Barrier Inspection," Revision 2, August 26,1997 19.

BGE " Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant," Revision 48, March 28,1997 20.

Letter from Mr. L. T. Deerflein (NRC) to Mr. C. H. Cruse (BGE), dated May 14,1997, " Plant Performance Review (PPR)- Caivert Cliffs" Application for Lkense Renewal 3.3 B-18 Calvert Cliffs Nuclear Power Plant

NITACllMENT (1)

APPENDIX A - TECIINICAL INFORMATION 3.3B - TURBINE BUILDING STRUCTURE 21.

Letter from Mr. J. T. Trapp (NRC) to Mr. R. E. Denton (BGE), dated May 6,1994, " Combined inspection Report Nos. 50-317/94-15 and 50-318/94 15"

- 22.

CCNP? Administrative Procedure Mi41219 " Structure and System Walldowns," Revision 0, September 16,1997 Application for License Renewal 3.3 B-19 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.3C - INTAKE STRUCTURE Baltimore Gas and Electric Company c

Calvert Cliffs Nuclear Power Plant March 3,1998

ATTACIIMENT (2)

I i

APPENDIX A - TECIINICAL INFORMATION 3.3C - LNTAKE STRUCTURE 3.3C Intake Structure This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing the intake Structure. The intake Structure was evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LRA.

TC.I Scoping she system level scoping process defines conceptual boundaries for plant systems and structures, develops screening tools which capture the 10 CFR 54.4(a) sceping criteria, and then applies the tools to identify systems and structures within the scope of license renewal. Systems and structures that are within the scope of license renewal are then scoped further on a component level. [ Reference 1, Section 3.0]

The component level scoping process for systems is described in Section 4.1 of the IPA Methodology, and the component level scoping process for structures is described ir. Section 4.2 of the methodology.

Components with unique equipment identifiers in the site equipment database are scoped using the component level scoping process for systems. Structural components such as walls do not have unique equipment identifiers. Therefore, the component level scoping process for structures utilizes a generic listing of structural component types. [ Reference 1, Section 4.0]

The CCNPP Intake Structure contains system type components with unique equipment identifiers in the -

site equipment database, as well as structural-type componer.ts that do not have ualque equipment identifiers. The system-type components include a variety of noa-safety related mechanical, electrical, and instrumentation components associated with equipment such as the traveling screens and screen wash pumps. The components with Intake Structure unique equipment identifiers were scoped using the component level scoping process for systems, and it was determined that none of these components are within the scope of license renewal. The component level scoping results for the structural-type components are discussed below. [ Reference 2]

The component level scoping process for structures identifies structural-type components as being within the scope of license renewal if they perform one or more of the following generic structural functions:

[ Reference 1, Section 4.2.2]

Provide structural and/or functional support to safety-related (SR) equipment; Provide shelter / protection to SR equipment; Serve as a pressure boundary or a fission product retention barrier to protect public health and safety in the event of any postulated Design Basis Events; Serve nr. a missile barrier (internal or external);

Provide structural and/or tunctional support to non-safety-related equipment whose failure could

~

d:rectly prevent satisfactory accomplishment of any of the required SR functions; Provide flood protection barrier (internal flooding event); and e

Application for License Renewal 3.3 C-1 Calvert Cliffs Nuclear Power Plant

---

__-._-._u

ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE Provide a rated fire barrier to confine or retard a fire from Epreading to or from adjacent areas of the plant.

He remainder of Section 3.3C.1 provides a description of the Intake Structure including the conceptual boundaries from the system level scoping results, the results of the component level scoping, and the results of the scoping to determine the components subject to aging management review (AMR).

Representative historical operating experience pertinent to aging is included in appropriate areas to provide insight supporting the aging management demonstrations. This operating experience was obtained through key-word searches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.

Structure Descriptien/Conceotual Boundaries Figure 3.3C 1 is a simplified layout of site structures showing the structures that are within the scope of license renewal, including the Intake Structure. The CCNP." site arrangement consists of numerous structures as shown on Figure 1-2 of the CCNPP Updated Final Safety Analysis Report (UFSAR).

Design features of the CCNPP structures are discussed in UFSAR Chapter 5. [ Reference 3; Reference 4, Table 2; References 5,6, and 7]

The intake Structure is situated to the cast of the main plant between the North Service Building and the Chesapeake Bay shoreline. The structure houses twelve circulating water pumps that supply water from the Chesapeake Bay to the condensers, and six saltwater pumps that provide cooling water to various plant equipment. Trash racks and traveling screens are provided to protect the condensers from foreign bodies present in the bay water. Running the full length of the structure is a gantry crane having a lifting capacity of 35 tons. [ Reference 3, Figure 1-2, Section 5.6.2.1, Reference 8, Section 1.1.1]

The intake Structure is approximately 90' (width) x 385' (length) and is constructed primarily of reinforced concrete. The foundation slab varies in elevation from -26'-0" to -14'-3". The total effective load due to the structure is approximately 42,000 tons. As a result, net soil pressures due to the structure are approximately 2500 pounds per square foot (psf). [ Reference 3, Sections 2.7.5.1 and 5.6.2.1]

For all major structures below finish grades, a heavy waterproofing membrane of 40 mils thickness is provided at the exposed face of the exterior walls and below the base slab. Rubber waterstops are also provided at all construction joi* up to grade elesation. Subsurface drains are provided to lower the elevation of groundwater around the plant. [ Reference 3, Section SA.5]

Since the intake Structure houses the saltwater pumps that are essential for safe shutdown of CCNPP, the structure was designed as a Category I structure for seismic, tornado, and hurricane conditions. The intake Structure is also designed to protect the saltwater pump motors from external flooding due to the maximum hypothetical hurricane tide and storm surges, including wave action. The Intake Structure design loads and conditions are shown in UFSAR Table 5-7 and UFSAR Section SA.S. [ Reference 3, Sections 2.8.3.6,5.6.2.2, and 5 A.5]

The structure is designed in accordance with American Concrete Institute (ACI) standards and the structural steel with American Institute of Steel Construction standards. The total length of the structure is divided into three sections above the base slab by two expansion joints. The high level roof at Application for License Renewal 3.3 C-2 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (2) l APPENDIX A - TECilNICAL INFORMATION 3.3C - INTAKE STRUCTURE Elevation 28'.6" is comprised of a reinforced concrete slab supported on a structural steel frame. Within l

this roof are access covers to each of the saltwater and circulating water pumps. [ Reference 3, Section 5.6.2.2]

The conceptual boundaries of this evaluation include the intake Structure and all of its structural components such as foundations, walls, slabs, and steel beams. Component supports that are connected to th structural components are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA, Component supports are defined as the connection between a system, or component within a system, and a plant structural member. An example of a component support is the fixed base that supports a pump. The pump would be scoped with its respective system evaluation. The component support is the fixed base that connects the concrete equipment pad to the pump. The fixed base is scoped with the Component Supports Commodity Evaluation and the concrete equipment pad is scoped with the evaluation for the structure. If anchor bolts are used, there is overlap between the Component Suppons Commodity Evaluation and the evaluation for the structural component. Evaluations for structural components considered the effects of aging caused by the surrounding environment, while the Component Supports Commodity Evaluation considered the effects of aging caused by the suppcrted equipment (thermal expansion, rotating equipment, etc.), as well as the surrounding environment. Supports for structural components such as platform hangers are not

" component supports" in this sense because any support for a structural component is itself a structural component and is included in the scope ofits respective structure. [ Reference 9, Section 1.1.1)

Cranes and fuel handling equipment that are connected to structures are evaluated for the effects of aging in the Cranes & Fuel llandling Commodity Evaluation in Sectica 3.2 of the BGE LRA. The intake Structure Gantry Crane rails, girders, and other structural support members were evaluated in the Cranes and Fuel llandling Commodity Evaluation and are not included in this section.

As shown in figure 3.3C-1, electrical ductbanks that run under the Turbine Building are connected between the Auxiliary Feedwater Pump Rooms and the intake Structure. The ductbanks are Seismic Category I and are constructed of reinforced concrete. These ductbanks contain electrical conduits used for routing of the cables that power the saltwater pumps. The conduits in the ductbank connect to electrical pull boxes that are mounted on the west wall of the intake Structure. These boxes provided a convenient pull point during construction for the saltwater pump motor cables. The pull boxes have experienced significant corrosion due to groundwater that drains into them through the conduits. The pull boxes are not within the scope oflicense renewal since they do not perform any intended functions as described in 10 CFR 54.4(a). The water leakage into the pull boxes is considered normal and is not considered a safety concern since the cables for the saltwater pump motors are suitable for submerged operation. The ductbanks are sloped downward toward the intake Structure, and the pull boxes are provided with weep holes to facilitate drainage of the conduits. The ductbanks are evaluated for the effects of aging in the Turbine Building Structure Evaluation in Section 3.3B of the BGE LRA. The cables are evaluated for the effects of aging in the Cables Commodity Evaluation in Section 6.1 of the BGE LRA. [ Reference 3, Section SA.2.1; Page 9.5 29; Reference 5]

Application for License Renewal 3.3C-3 Calvert Cliffs Nuclear Power Plant

A'ITACilMENT (2)

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- ~,_w 1-FUELOIL STORAGE TANK Encmcm them

• w No. 21 1A Deeel Genermor 1A ENCLOSURE me== F.de) ll Expansson Joint y

STRUCTURES WITHIN THE SCOPE OF L. _ LICENSE _ RENEWAL FIGURE 3.3C-1 CCNPP SITE STRUCTURES (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY)

Application for License Renewal 3.3C-4 Calvert Cliffs Nuclear Power Plant

l ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE 4

Structure Sconing Result 3 The Intake Structure is in the scope oflicense renewal based on 10 CFR 54.4(a). Six out of seven of the generic structural functions listed above are applicable to the Intake Structure as shown in Table 3.3C-1.

The intended functions for the intake Structure were determined based on the requirements of

$54.4(a)(1), l54.4(a)(2), and $54.4(a)(3), in accordance with CCNPP IPA Methodology Section 4.2.2.

[ Reference 2, Table IS; Reference 8, Section 1.1.3)

TABLE 3.3C-1 INTENDED FUNCTIONS OF STRUCTUHF5; Applicable to Applicable Funetion Intake 10 CFR 54.4(a)

Structure?

Criteria 1.

Provide structural and/or functional support to SR Yes Q54.4(a)(1) equipment 2.

Provide shelter / protection to SR equipment Yes

$54.4(aX1) 3.

See e as a pressure boundary or a fission product retention No

{54.4(a)(1) barrier to protect public health and safety in the event of any postulated Design Basis Events 4.

Serve as a missile barrier (internal or external)

Yes

@54.4(aX 1)

5. Provide structural and/or functional support to non-safety-Yes 54.4(a)(2) related equipment whose fa: lure could directly prevent satisfactory accomplishment of any of the required SR functions 6.

Provide flood protection barrier (intemal flooding event)

Yes 54.4(a)(2) 7.

Provide rated fire barriers to contine or retard a fire from Yes

{54.4(a)(3) spreading to or from adjacent areas of the plant Comnonents Subject to AMR As discussed above, the component level scoping process for structures utilized a generic list of structural component types. The generic list started with structural component types contained in industry technical reports nddressing containment structures and other Category I structures. Other structural component types were added to the list to ensure completeness. Additionally, any structural component types that are unique to the particular structure being scoped, such as the prestressed tendons in the Containment and the sluice gates in the intake Structure, are noted. These structural components were combined into four structural categories based on their design and materials as follows:

[ Reference 1, Section 4.2.3; Reference Z]

Concrete Components;

+

Structural Steel Components; Architectural Components; and Unique Components.

+

Application for License Renewal 3.3C-5 Calvert Cliffs Nuclear Power Plant

ATTACHMENT d)

APPENDIX A - TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE During the scoping process, those structural component types actually contained in the Intd.e Structure were identified. Within the four structural component categories, twenty-seven structural component types Were determined to contribute to at least one of the intake Structure intended functions listed in Table 3.3C l. Table 3.3C 2 lists these 27 structural component types and their associated intended functions. S'.ructural component types that are part of the intake Structure, but do not contribute to any of the intended functions of the structure, are not listed in the table [ Reference 2, Table 3S; Reference 8, Table 2.ll TABLE 3.3C-2 STRUCTURAL COMPONENT TYPES REOUIRING AMR FOR THE INTAKE STRUCTURE Structural Component Type l

Applicable Function (s)

Concrete (including Reinforcing Steel)

Foundations (I ootmgs, beams, and mats)

I,2 Columns 1, 2, 4, 7 Walls I, 2, 4, 7 licams 2, 4 Ground Floor Stabs and Equipment Pads I,2 Elevated iloor Stabs I,2 Roof Slabs 2

Cast in-Place Anchors,Embedments*

I, 2, 6, 7 Grout I,2 Fluid retaining Walls and Stabs I, 2, 6 Post Installed Anchors' 2, 5 StructuralSteel licams'

l. 2 liaseplates' I,5 Floor Framing

• 1, 5 Roof Frammg*

2 Ilracmg*

2, 5 Platform llangers*

5 Decking' 2

Floor Gratmg' 5

Checkered Plate

• 2 Stairs and Ladders

• 5 Architectural Congponents Fire Doors, Jambs, and liardware*

2, 7 Access Doors, Jambs, and liardware*

2 Caulkmg and Scalants 6, 7 Unique Congponents Watertight Doors

• 2, 6 Sluice Gates i

Expansion Joints 2, 7

(*)

Asterisk in " Structural Component Type" column indicates that the component type is included under the heading " Steel Components" in Table 3.3C-3.

(#)

Numbers in " Applicable Function (s)" column correspond to the associated intended functions as listed in Table 3.3C-l.

Application for License Renewal 3.3C-6 Calvert Cliffs Nuclear Power Plant

ATTACHMFNT m APPENDIX A - TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE As discussed in IPA Methodology Section 5.4, all seven of the generic structural functions are considered to be passive. In addition, plant structural components are not normally subject to periodic replacement programs. Therefore, structural components are considered to be long-hved unless specific justification is provided to the contrary. Baied on the above, all of the structural component types listed in Table 3.3C 2 are subject to AMR for the intake Structure. [ Reference 8, Section 3.0]

Daltimore Gas and Electric Company may elect to replace cc,mponents for which the AMR identifies that further analysis or examination is needed, in accordance with the License Renewai Rule, components subject to replacement based on qualified life or specified time period would not be subject to AMR.

3.3C.2 Aging M:nagement ne list of potential Age-Related Degradation Mechanisms (ARDMs) identified for the intake Structure components is given in Table 3.3C-3, with plausible ARDMs identified by a check mark (/) in the appropriate column. [ Reference 8. Attachments 1 and 2, Appendices C, E, K, and 0] For efficiency in presenting the results of these evaluations in this report, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all components.

Table 3.3C 3 also identifies the group to which each ARDM/ device type combination belongs.

Exceptions are noted where appropriate. The following groups have been selected for the Intake Structure:

Group 1: Includes caulking, scalants, and expansion joints subject to weathering; Group 2: Includes fluid-retaining walls and slabs subject to aggressive chemical attack on concrete and corrosion of embedded steel /rebar; Group 3: Includes steel components subject to corrosion; and Group 4: Ir des sluice gates subject to corrosion.

Application for License Renewal 3.3C-7 Calvert Cliffs Nuclear Power Plant

I ATTAC1G2ENI_G)

APPENDEX A - TECHNICAL INFORMATION 3.3C - INTAKE STRUCTURE TABLE 3.3C-3 POTENTIAL AND PLAUSIBLE ARDMs FOR THE LSTAKE STRUCTURE Concrete Ground C*I""'S*

Potential ARDMs Foundanons Elewsted Roof Grout FinMtetaming Caulking and Expansma Steel Sluice

Walls, Stabs and Door Sas Walls ar.d Slabs Scalar =ts Joi'-ts Cee-Gases s

ami Egnpment Stabs seams Pads Freeze-Thaw Leachmg of Calcmm Ild wide t

a,

/(2)

At l nive Chemical Attack on Concrete l

Corrosion of

/(2)

Embedded j

i Steel 1tebar Abrasen and Cavitation Settlement Cormsion

/(3)**

<( 4 )**

Weathermg

<(1)

/(1)

Fatigue Steel Cm.,

wr represent all uructural m,.,,.ea types marked with an aste 1(*) in Table 3.3C-2

• • - Group 3 u.u,~.. cats are subject to general cormsson. Group 4 w. 3.o._s are subject to crevice corrosion. microt AA'ty-induced cormsen (MIC). ard pittmg.

/ - Indicates plausible ARDM detennination

(#) - Indicates the group (s) in which the ARDM/wog.v.wa type combmation is evaluated s

t i

Application for License Renewal 3.3C-8 Calvert Cliffs Nuclerr Power Pint t

ATTACllMENT (2)

APPENDIX A TECliNICAL INFORMATION 3.3C - I',7 AKE STRUCTURE Aging mechanisms that are not plausible are generally not discussed funher in these BGB LRA sections, unless ti ey are considered noteworthy. For the intake Structure, settlement is considered noteworthy and is discussed below.

An industry technical report concluded that settlement is a potentially sigt.ificant ARDM for Category I structures. Settlement occurs both during construction Ci aller construction. 'lhe amount of settlement depends on the physical propenies of the foundation material. These propenies range from rock (with little or no settlement likely) to compacted soil (with some settlement expected). Settlement may occur during the design life of the structure from changes in environmental conditions, such as lowering of the groundwater table. Settlement can occur in two stages: elastic expansion and time dependent settlement. Elastic expansion of the confined soil occurs due to excavation unloading and results in a slightly upward movement. During construction, the soll moves downward as load is applied. This clastic movement should be small and is complete when construction is completed. It has no effect on the structure and is not considered an aging mechanism. The excavation unloading and structural loading cause a str;all change in the void ratio of the soll. This change results in a small amount of time.

dependent settlement. The settlement rate will decline after completion of construction Concrete and steel structural members can be affected by differential settlement between supporting foundations, within a building, or between buildings. Severe settlement can cause misalignment of equipment and lead to overstiess conditions within the structure. When buildings experience significant settlement, i

cracks in structural members, differential elevations of supporting members bridging between buildings, or both may be visibly detected. Settlement was determined to be not plausible for the CCNpP Intake Structure based on the folloving site specific justification: [ Reference 8, Appendix J, Sections 1.0 and 2.5; Reference 10, Section 5.1J The basemat elevation of the intake Structure is approximately 110 feet below the original ground e

elevation. The basemat is situated on Miocene soil, which is exceptionally dense and will support heavy foundation loads. The ultimate bearing capacity of the foundation strata is in excess of 80,000 psf, and the allowable bearing capacity is 15,000 psf. The design contact pressure of the intake Structure foundation is only 2500 psf. h contact pressure is only 23% of the overburden pressure removed due to excavation. [ Reference 8, Appendix J, Section 2.1]

in addition to soll bearing capacity, settlement was also investigated in the design of the Intake e

Structure. A maximum post construction settlement of 1/2 inch was predicted in the original Intake Structure design. Since the intake structure is situated on an exceptionally dense soil, the structure tends to uniformly settle. Most of the pudicted 1/2 inch settlement is in terms of uniform settlement, which has no adverse effect on the structural compom nts of the intake Structure. A small fraction of the 1/2-inch settlement will be in terms of mfierential settlement.

It is so small that the effect on the structural components ' negligible.

[ Reference 3, Section 2.7.6.2; Reference 8, Appendix J, S:ction 2.4]

Based on the above, settlement was determined to be not piausible for any structural components of the intake Structure. This conclusion is supported by a walkdown of the intic Structure, performed in 1994, which "ound no indication of structural damage due to settlement. [ Reference 8, Appendix J, Section 3.0]

Application for License Renewal 3.3C 9 Calvert Cliffs Nuclear Power PMnt

ATTACllMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.3C INTAKE STRUCTURE The following is a discussion of the aging management demonstration process for each group identified above. it is presented by group and includes a discussion of materials and environment, aging mechanism effects, methods of managing aging, aging management program (s), and aging management demonstration.

Group 1 - (caulking, sealants, and expansion joints subject to went*

• ng) - Materials and Environment Group 1 includes caulking, scalants, and expansion joints subject to weathering. These structural components provide intake Structure intended functions as shown in Table 3.3C 2.

(Refeience 8, Appendix 0, Section 2.2]

De material requirements for the caulking, scalants, and expansion joints used during construction of CCNPP were governed by construction specifications. Specific manufacturers' and brand names (or approved equals) were specified for different applications. (References 11 and 12]

He caulking, scalants, and expansion joints located indoors will be subject to the ambient conditions within the intake Structure. The intake Structure ambient temperature is controlled by a plant ventilation system as described in UFSAR Chapter 9. The caulking, scalants, and expansion joints located outdoors will be subject to the temperature and hu.nidity changes, rain, snow, etc. expr -ted at the CCNPP site.

[ Reference 3, Section 9.8.2.6, Table 918; Reference 8, Appendix 0, Section 2.1]

Group 1 - (caulking, scalants, and expansion joints subject to weathering) - Aging Mechanism Effects v

' ing, scalants, and expansion joints that are exposed to ambient conditions (indoor or outdoor) are o eeptible to degradation due to weathering. Exposure to sunlight (ultraviolet expocure), changes in midity, owne cycles, temperature and pressure fluctuations, aid snow, rain, or ice contribute to the weathering ARDM ne efTects of weathering on most caulking, sealant, and expansian joint materials are evidenced by a decrease in elasticity (drying out), an increase in hardness, and shrinkage. (Reference 8, Appendix 0, Section 1.0]

Weathering was determined to be plausible for the intake Structure caulking, sealants, and expansion joints due to their exposure to the ensironmental conditions that contribute to this ARDM. (Reference 8, Appendix 0, Sections 2.1 and 2.5]

Expansion joints that run along the intake Structure floor ham experienced age-related degradation in the past. ne degradation allowed water seepage up through the joints. The affected joints were subsequently repaire6 using approved sealant materials.

His aging mechanism, if unmanaged, could eventually result in the caulking, scalants, and expansion joints not being able to perform their intended functions under current licensing basis (CLB) conditions.

Therefere, wea.hering was determined to be a plausible ARDM for which the aging effects must be managed for the intake Structure caulking, scalants, and expansion joints. (Reference 8, Appendix 0, Section 2.3]

Application for License Renewal 3.3C-10 Calvert Cliffs Nuclear Powu Plant

ATTACHMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.3C INTAKE STRUCTURE Group 1 -(enutking, scalants, and expansion joints subject to weathering) Methods to Manage Aging hiitigation: Because weathering of caulking, scalants, and expansion joints is rNsted by exposure to environmental conditions that are not feasible to control (e.g., light, heat, oxygen, ozone, water, radiation), there are no practical methods to mitigate its effects.

Discoverv: Caulking, scalants, and expansion jcints degrade over time and should be replaced as needed.

An inspection program that provides requirements and guidance for the identification, inspection, and maintenance of caulking, scalants, and expansionjoints can ensure that their condition is maintained at a level that allows them to perfonn their intended functions. An effective program will provide for baseline inspo ' ion along with periodic future inspections at appropriate intervals depending upon the degree of harw.aess of the environment of the caulking, scalant, or expansion joint, items that are in a harsh exterior environment would be inspected more frequently. This program would involve visual inspection and probing to determine that the caulking, sealant, or expansionjoint is satisfactorily attached to the surface and is ficxible.

t Group 1 - (caulking, scalants, and expansion joints subject to weathering) - Aging Management Programs Mitiption: There are no CCNPP programs credited for mitigation of weathering.

Discoverv: Caulking, scalants, and expansion joints that perform a fire barrier function are managed under an existing program. The Penetration Fire Barrier inspection Program, implemented through a CCNPP Surveillance Test Procedure (STP), STP F 592-1/2, is adequate to manage the effects of aging for caulking, scalants, and expansion joints that function as fire barriers without modification. [ Reference 8, Appendix 0, Section 2.6].

The purpose of STP-F.5921/2 is to provide instructions for visual inspection of fire barrier penetration seals in fire area boundaries that protect safe shutdown areas in Units 1 and 2. He scope of this procedure is to visu lly inspect the following type of fire barrier penetration scals for operability: (References 13 and 14, Sections 1.0 and 2.2]

Electrical conduit and cable tray penetration seals; e

IIcating, ventilation, and air conditioning duct penetration seals (ducts without dampers); and Mechanical pipe and expansion joint penetration seals.

Procedure STP F 592-1/2 was developed based on CCNPP Technical Specificatims 3.7J2 and 4.7.12.a.

10 CFR Part 50 Appendix R, the CCNPP Fire Protectior. Plan, NRC Generic Letter 8f 10, and various plant drawings. [ References 13 and 14, Section 3.1]

4 The procedure is currently performed at least once per 18 months in accordance with Technical Specification 4.7.12.a. De procedurc requires that the fire barrier penetration seals be visually inspected to determine if they are operable based on specific criteria that were developed for each type of fire barrier component. In general, the procedure inspects the penetration seals for damage, cracking, voids, and proper installation. De procedure provides separate " failure criteria" and " repair criteria." The " failure criteria" are used to determine if the penettetion seal is considered to be inoperable. The " repair criteria" are used to Application for License Renewal 3.3C-11 Calvert Cliffs Nuclear Power Plant

ATTACliMENL12)

APPENDIX A - TECHNICAL INFORMATION 3.30 - INTAKE STRUCTURE determine if the penetration seal is operable but in need of repair. [ References 13 and 14, Sections 2.1 and 6.0,and Attachment A]

If a fire barrier penetration seal is determined to be inoperable based on the procedure criteria, plant personnel determine if actions are required in accordance with Technical Specification 3.7.12.a.

In addition, any conditions adverse to quality discovered during the inspection are documented on issue Reports in accordance with the CCNPP Corrective Actions Program. [ References 13 and 14, Section 6.5 and Attachment B)

The Fire Protection Program at CCNPP (which includes STP F 5921/2) is subject to periodic internal assessment in accordance with the requirements in BGE's Quality Assurance Policy, Audits are required for the Fire Protection Program and implementing procedures every two years. In addition, an independent fire protection and loss pvention program inspection and audit utilizing either qualified offsite BOE personnel or an s 'ide L protection firri is also required every two years. The Quality Assurance Policy also requires an inspection and audit c f the fire protection and loss prevention program by a qualified outside fire consultant at least once every three years. An audit and inspection performed in 1996 (using an outside consultant as well as B(iE personnel) concluded that the CCNPP Fire Protection Program is providing a level of safety consistent with good fire protection practices and NRC regulatory criteria. The inspection included plant walkdowns of some of the fire barrier penetration seals.

No age-related degradation issues for the seals were identified. (Reference 15, Section 18.18)

"Ihe Fire Protection Program also undergoes period:c it.:pection by the NRC as part of their routine licensee assessment activities. An inspection of the program in 1994 included a review of procedure STP F 592-1 and a plant tour that included inspectio n of some of the fire barrier penetrations. The NRC concluded that the Fire Protection Program complies with program requiremeats provided in the Technical Specifications and licensing documents. [ References 16 and 17]

Operating experience related to this program has shown that aging is a minor contributor to fire barrier penetration sent failures at CCNPP. The greatest contribtcor to degradation of these seals is believed to be due to inadequacies in the original installation of the seal materials. For example, degraded seals have been found to be the result of incomplete installation of the seal material (i.e., openings left in the penetrations) or due to improper grout installation.

The corrective actions taken as a result of the Penetration Fire Barrier inspection Program will ensure that the intake Structure caulking, scalants, and expansionjoints that perform a fire barrier function will remain capable of performing their intended function under all CLB conditions, Caulking, scalants, and expansion joints that are not fire barriers are typically replaced upon identification of their degraded condition. Visual examinations of the caulking, scalants, and expansion joints in the plant concluded that an inspection program was needec.o adequately manage the aging of these structural components. [ Reference 8 Appendix 0, Section 2.4J For the caulking, sealants, and expansion joints that do not perform a fire barrier function, a new CCNPP Caulking and Sealant inspection Program will provide requirements and guidance for the identification, inspection frequencies, and acceptance criteria for caulking, scalar?s, and expansion joints used in the intake Structure to ensure that their condition is maintained at a level that allows them to perform their Application for License Renewal 3.3C 12 Calvert Cliffs Nuclear Power Plant

.~ - --.. - _

ATTAC11 MENT m APPENDIX A TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE A

intended functions. The new program will establish acceptance uiteria and require a baseline inspection to determine the material condition of the caulking, realants, and expansion joints for the Intake Structure. If unacceptable degradation exists, corrective 2n ons will be taken. A technical basis will be developed for determining the periodicity of future inspections. [ Reference 8, Attachment 8]

j Group 1 - (caulking, sealants, and espansion joints subject to weathering) - Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to weathering of caulking, scalants, and expansion joints:

• The caulking, scalants, and expansionjoints provide intake Structure intended functions as shown in Table 3.3C 2.

Weathering was determined to L a plausible ARDM for the caulking, scalants, and expansion joints. This ARDM, if unm..naged, could eventually result in the caulking, scalantr, and i

expansionjoints not being able to perform their intended functions under CLB conditions.

For caulking, scalants, and expansion joints that function as fire barriers, the Penetration Fire i

Barrier inspection Program performs periodic visual inspections of fire barrier penetration seals, and contains acceptance criteria that ensure corrective actions will be taken such that the fire i

i barrier imended function will be maintained.

For caulking, scalants, and expansion joints that do not perfonn a fire barrier fimetion, a new i

Caulking and Sealants inspection Program will conduct inspections to detect age related degradation, and will contain acceptance criteria that ensure corrective actions will be taken such that the intended functions will be maintained.

Therefore, there is reasonable assurance that the effects of weathering will be adequately managed such that the caulking, scalants, and expansion joints will be capable of performing their intended functions, consistent whh the CLR, during the period of extended operation.

Group 2 - (fluid retaining walls and slabs subject to aggressive chemical attack on concrete and corrosion of embedded steet/rebar)- Materials and Environment Group 2 includes the intake Structure fluid-retaining walls and slabs that could be subject to aggressive chemical rttack on concrete and corrosion of embedded steel. The fluid-retaining walls and slabs provide structural and/or functional support to SR equipment, provide shelter / protection to SR equipment, and provide flood protection barriers (internal flooding event). [ Reference 8, Appendices C and E, Section 2.2]

The embedded steel /rebar is covered and prote.ted by concrete. At CCNPP, embedded steel is used in composite structural members and as anchorage for concrete surface attachments. Reinforcing steel (rebar) and cast in-place anchors are all treated as embedded steet/rebar in this evaluation. (Reference 8, Appendix E, Section 2.0]

The intake Structure concrete was constructed in accordance with a CCNPP design specification that adheres to relevant ACI codes and American Society for Testing and Materials specifications for a concrete structure of low permeability, in addition, suflicient concrete cover over embedded steellrebs was Application for License Renewal 3.3C-13 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (2)

APPENDIX A TECilNICAL INFORMATION 3.3C. INTAKE STRUCTURE l

specified in accordance with the ACI 318 code to provide corrosion protection. [ Reference 8, Appendix E, Section 2.4]

The fluid-retaining walb and slabs are exposed to intake water from the Chesapeake Bay, in June and August of 1968 and 1969, the chemical characteristics of the Chesapeake Bay surface water were analyzed at kwations in the vicinity of CCNPP. Some of the results obtained are as follows: [ Reference 8, Appendices C and E Section 2.5; Reference 18, pages 11 12 and 11 13]

pil:

range of 7.3 to 8.4; e sulfates:

range of 770 to 1150 parts per rr.illion (ppm); and e chlorides:

range of $800 to 7800 ppm.

It is assumed that there have been no signl0 cant changes in the above chemical characteristics since the original analysis was performed.

Group 2 - (fluid retalning walir and slabs subject to aggressive chemical attack on concrete and correalon of embedded steel /rebar)- Aging Mechanism Effects Aggressive chemical attack on concrete and corrosion of embedded stect/rebar was determined to be plausible for the intake Structure fluid-retaining walls and slabs since the intake water could contain chemicals that might attack the concrete or cause corrosion of the embedded steel /rebar. [ Reference 8, Appendices C and E, Section 2.5)

Aggressive Chemical Attack on Concrete Concrete, being highly alkaline (pil > 12.5), is vulnerable to degradation by strong acids. Acid attack can increase porosity and permeability of concrete, reduce its alkaline nature at the surfsce of the attack, reduce strength, and render the concrete sub)ct to further deterioration. A dense concrete with low permeability and a low water to-ccment ratio may provide an acceptaUe degree of protection against mild acid attack. [ Reference 8, Appendix C, Section 1.0; Reference 10, Section 4.1.3.1)

Chlorides and sulf ates of sodium, potassium, and magnesium may attack concrete depending upon their concentrations. Sulfate attack can produce significant expansive stresses within the concete, leading to cracking, spalling, and strength loss. Once establishea, these conditions allow further exposure to aggressive chen9cals. Use of adequate cement content, low water to-cement ratio, and thorough consolidation and curing contribute to low permeability and provide efTective protection against sulfate and chloride attack. Based on an industry report, minimum degradation threshold limits for concrete are 500 ppm chlorides or 1,500 ppm sulfates. [ Reference 8, Appendix C, Section 1.0; Reference 10, Section 4.1.3.1]

[

Application for License Renewal 3.3C 14 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (2)

APPENDIX A - TECilNICAL INFORMATION 3JC INTAKE STRUCTURE Corrosion of EmbaAAad Steel /Rebar Concrete's high alkalinity (pil > 12.5) provides an environment around embedded steel /rebar that protects it from corrosion. Ilowever, when the pil is reduced by the intrusion of aggressive ions, l

corrosion can occur. The corrosion rate is insignificant until a pil of 4.0 is reached. A reduction in pli can be caused by the leaching of alkaline products through cracks, the entry of acidic materials, or carbonation. Chlorides can be present in constituent materials of the original concrete mix (i.e., cement, aggregates, admixtures, and water), or they may be ir.troduced environmentally (e.g., from the intake water). De severity oicorrosion is influenced by the propenles and type of cement and aggregates as well as the concrete moisture content.

(Reference 8, Appendix E, Section 1.0; Reference 10, Section 4.1.5.1)

Corrosion products add volume to the original metal. The presence of suflicient corrosion products on embedded steel or rebar subjects the concrete to tensile stress that eventually causes hairline cracking, rust staining, spalling, and more severe cracking. These actions will expose more embedded steel /rebar to a potentially corrosive environm mt and cause funher deterioration in the concrete. A loss of bond between the concrete and embedded steet/rebar will eventually occur, along with a reduction in steel cross section. Dese conditions can ultimately impair structural integrity. (Reference 8, Appendix E, Section 1.0; Reference 10, Section 4.1.5.1) ne degree to which concrete will provide satisfactory protection for embedded steet/rebar depends in most instances on the quality of the concrete and the depth of concrete cover over the steel, ne permenbility of the concrete is also a major factor afTecting corrosion resistance. Concrete of low penneability contains less water under a given exposure and, hence, is more likely to have lower electrical conductivity and better resistance to corrosion. Such concrete also resists absorption of salts i

and their penetration into the embedded steel and provides a barrier to oxygen, an essential element of the corrosion process. Iww water to-ccment ratios and adequate air ehtrainment increase the resistance to water penetration and thereby provide greater resistance to corrosion. [ Reference 8, Appendix E, 4

Section 1.0 Reference 10, Section 4.1.5.1]

Rese aging mechanisms, if unmanaged, could eventually result in the intake Structure fluid-retaining walls and slabs not being able e perform their intended functions under CLB conditions. Therefore, aggressive chemical attack on conuete and corrosion of embedded steet'rebar were determined to be plausible ARDMs for which the aging effects must be managed for the intake Structure. [ Reference 8, Appendices C and E, Section 2.3]

Group 2 -(fluid retaining walls and slabs subject to aggressive chemical attack on concrete and corrosion of embedded steel /rebar)- Methods to Manage Aging Mitigation: The intake Structure concrete is designed to have a low permeability, in addition, sufficient concrete cover over embedded steel /rebar was specified to provide corrosion protection. These design considerations help to mitigate aggressive chemical attack on concrete and corrosion of embedded steel /rebar. Ilowever, to provide further assurance that degradation is not occurring, the discovery methods

}

described below are deemed necessary to manage these ARDMs. (Reference 8, Appendix E, Section 2.4]

Discoverv: Visual inspections of the fluid-retaining walls and slabs can be performed to provide assurance that degradation of the cor. crete (i.e., concrete cracking, rust staining, spalling) is not Application for License Renewal 3.3C 15 Calvert Cliffs Nuclear Power Plant

ATTACintENT (2)

APPENDIX A - TECilNICAL INFORMATION 3.3C - INTAKE STRUCTURE V an-occurring. If any significant degradation is found, appropriate corrective actions can be taken to ensure that the Guld retaining walls and slabs will continue to perform their intended functions during the period of extended operation.

Group 2 -(Huld retaining walls and slabs subject to aggressive chemical attack on contrete and corrosion of embedded steet/rebar)- Aging Management Program (s)

Mitigation: The design considerations discussed above help to mitigate aggressive chemical attack on concrete and corrosion of embedded steel /rebar for the Intake Structure foundation and ground Hoor slab.

Here are no programs credited with mitigating these ARDMs.

Discoscry:

Preventive Maintenance (PM) tasks are currently in place at CCNPP that call for the periodic draining of the intake Structure cavities during refueling outages to scrape and wash the saltwater tunnet/ cavity walls (i.e., the Guid retaining walls and slabs). Visual inspections are performed aller cleaning and repairs are made as required. These PM tasks will be modined to provide more speci0c guidance on inspecting for degradation (e.g., cracking, rust staining, spalling) that may be a result of ARDMs. The corrective actions taken as a result of these PM tasks ensure that the intake Structure Guid retaining walls and slabs remain capable of performing their intended functions under all CLB conditions. Operating experience associated with performance of these PM tasks is that no significant age-related degradation of the concrete has been identiDed. [ Reference 19]

These PM tasks dewater the intake Structure cavities and associated saltwater tunnels by the installation of stop logs in the vicinity of the trash racks. This allows inspectio's of the Guid retaining walls and slabs downstream of the stop logs. The podion of the intake Structure Guld-retaining walls and slabs upstream of the stop logs are not inspected by these PM tasks. Ilowever, all of the intake Structure Guld-retaining walls and slabs are subject to the same environmental conditions. Therefore, the portions that are inspected are considered representative of all of the intake Structure Guid-retaining walls and slabs.

Any conditions adverse to quality discovered during these inspections are dc.cuments on issue Repons in accordance with the CCNPP Corrective Actione Program. Issue Reports are required to identify the extent of the issue, including the suspected boundary of the problem. Corrective actions are taken as required as part of the issue Report resolution process. For the intake Structure Guid retaining walls and slabs, this cerrective action would include the use o divers to inspect the portion upstream of the stop r

logs if deemed necessary. [ Reference 19; Reference 20, Attachment l}

The PM tasks described above are performed in accordance with the CCNPP PM Program. This program has been established to maintain plant equipment, structures, systems, and components in a reliable condition for normal operation and emergency use, minimize equipment failure, and extend equipment and plant life. [ Reference 21, Section 1.1)

The program is govemed by CCNPP Administrative Procedure MN 1-102, " Preventive Maintenance Program," and covers all PM activities for nuclear power plant structures and equipment within the plant.

Guidelines drawn from industry experience and utility best practices were used in the development and enhancement of this program. [ Reference 21, Section 2.1]

The PM Program includes periodic inspection of specinc structures and components through various maintenance activities. These activities provide an effective means to discover and manage the age.

Application for License Renewal 3.3C 16 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECilNICAL INFORMATION 3.3C - INTAKE STRUCTURE ulated degradation effects on these structures and components. The program requires that an issue Report be initiated according to CCNPP Procedure QL 2100, " Issue Reporting c ed Assessment," for deficiencies noted during performance of PM tasks. The corrective actions taken ensure that the affected structures and components remain capable of performing their intended functions under all CLB conditions. [ Reference 21 Section 5.2.B.l.fj Specific responsibilities are assigned to BGE personnel for evaluating and upgrading the PM Program and for initiating program improvements based on system performance. Issue Reports are initiated according to CCNPP Procedure QL-2100 to request changes to the program that could improve or correct plant reliability and performance. Changes to the PM Program that require issue Reports include changes to the PM task scope, frequency, process changes, results from operating experience reviews, as well as other types of changes. [ Reference 21. Sections 5.1.A and 5.4.B]

The PM Program is subject to periodic internal assessment. Internal audits are performed to ensure that activities and procedures established to implement the requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program. These audits provide a comprehensive independent verification and evaluation of quality related activities and procedures. Audits of selected aspectr of operational phase activities are performed with a frequency commensurate with their strength of perfomiance and safety significance, and in such a manner as to assure that an audit of all safety.

related functions is completed within a period of two years. An audit performed in 1997 of the CCNPP Maintenance Program (which includes the PM Program) concluded that the program is effectively implemented at CCNPP.

No age related degradation issues were identified.

[ Reference 15, Section 18.18]

Group 2 -(fluid retaining walls and slabs subject to aggressive chemical attack on concrete and corroslon of embedded steel /rebar)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to aggressive chemical attack on concrete and corrosion of embedded steel /rebar for the intake Structure fluid-retaining walls and slabs:

ne fluid-retaining walls and slabs provide structural and'or functional support to SR equipment, e

provide shelter / protection to SR equipment, and provide flood protection barriers (internal flooding event).

Aggressive chemical attack on concrete and corrosion of embedded steel /rebar are plausible for the intake Structure fluid retaining walls and slabs since the ir.take water could contain chemicals that might attack the concrete or cause corrosion of the embedded steet/rebar. Rese ARDM3, if unmanaged, could eventually result in the fluid-retaining walls and slabs not being able to perform their intended functions under CLB conditions.

The PM Program conducts periodic inspections of the Intake Structure fluid-retaininc, walls and slabs through performance of various PM tasks that provide the means to discover and manage age-related degradation. Corrective actions are taken to correct any deficiencies that are found to ensure that the affected structures or components remain capable of performing incir intended functions under all CLB conditions.

Application for License Renewal 3.3C 17 Calvert Cliffs Nuclear Power Plant

ATTACllMENT (2) l APPENDIX A - TECilNICAL INFORMATION 3JC - lNTAKE STRUCTURE l

'Iherefore, there is reasonable assurance that the elTects of aggressive chemical attach on concrete and corrosion of embedded stect/rebar will be adequately managed such that the fluid-retahing walls and slabs will be capable of performing their intended functions, consistent with the CLB, during the period

. of extended operation.

Group 3 -(steel components subject to corrosion)- Materials and Environment Group 3 includes the intake Structure steel componems marked with an asterisk in Table 3.3C 2. These components are all subject to corrosion. They each contribute to one or more of the various intake Structure intended functions as shown in Table 3.3C 2. [ Reference 8, Appendix K, Section 2.2)

Since corrosion was recognized as a potential degradation mechanism for all structural steel components of the intake Structure, its effects were censidered in the original design. As a result, all exposed structural steel surfaces in the Intake Structure except galvanized steel such as grating, checkered plate, and metal decking, were shop-painted or field painted hring plant construction.

[ Reference 8, Appendix K Section 2.4]

The steel components located indoors will be subject to the ambient conditions within the intake Structure. The intake Structure ambient temperature is controlled by a plant ventilatien system as described in UFSAR Chapter 9.

The steel components located outdoors will be subject to the temperature and humidity changes, rain, snow, etc. expected at the CCNPP site. [ Reference 3, Section 9.8.2.6, Table 9-18]

Group 3 - (= teel components subject to corrosion)- Aging Mechanism Effects Steel corrodes in the presence of moisture and oxygen as a result of electrochemical re ctions. Initially, the exposed steel surface reacts with oxygen and moisture to form an oxide film as rust. Once the protective oxide film has been formed r.nd if it is not disturbed by erosion, alternating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from sal' water, the atmosphere, or groundweer, mmense the rate of corrosion by increasing the electrochemical activity. If steel is in contact with another metal that is more noble in the galvanic series, corrosion of the steel may accelerate. [ Reference 8, Appendix K, Section 1.0]

Corrosion products, such as Sydrated oxides of iron (rust), form on exposed, unprotected surfaces of the steel and are readily visible. The affected surface may degrade to such an extent that visible perforation may occur in the case of exposed surfaces of steel with protective coatings, conosion may cause th:

protective coatings to lose their ability to adhere to the corroding surface. In this case, dunage to the coatings can be visually detected well in advance of significant degradation of the steel. [ Reference 8, Appendix K, Section 1.0]

Visual inspection of accessible interior and exterior areas of the intake Structure was performed in 1994.

This inspection identified minor areas of rust on steel components. [ Reference S, Attachment 7]

Application for License Renewal 3.3C 18 Calven Cliffs Nuclear Power Plant

ATTACHMENT m Al'PENDIX A TECHNICAL INFORMATION 3.3C - INTAKE STRUCTURE His aging mechanism, if unmanaged, could eventually result in the steel components not being able to perfonn their intended fuuctions under CLB conditions. Herefore, corrosion was determined to be a plausible ARDM for which the aging effects must be managed for the intake Structure steel components.

[ Reference 8, Appendix K. Section 2.3)

Group 3. (steel components subject to corrosion). Methods to Manage Aging Mitigation: He effects of corrosion cannot be completely prevented, but they can be mitigated by minimizing the exposure of extcrnal surfaces of the steel components to an aggressive environment and protecting the external surfaces with paint or other protective coating. Coatings serve as a protective layer, preventing moisture and oxygen from directly contacting the steel surfaces.

Discovsty: The effects of general corrosion / oxidation of steel are detectable by visual inspection. A visual examination by a person familiar with the components can be used to determine general mechanical and,tructural condition and check for rust. Observing that s!gnificant degradation of protective coatings has not occurred is an effective method to ensure that corrosion has not affected the intended function of the structural component. Since the coating does not contribute to the components' intended functions, degradation of the coating provides an alert condition that triggers corrective action before the occurrence of corrosion that would affect the components' ability to perform their intended functions. The degradation of the protective coating that does occur can be discovered and monitored by periodically inspecting the steel structural components. Corrective action for failed protective coatings and any actual metal degradttion can be carried out as necessary.

[ Reference 8, Appendix K, Section 3.0)

Group 3 -(steel components subject to corrosion)- Aging Management Programs Mitigation: No programs are credited for mitigation. The exposed surfaces of structural steel components are covered by protective coatings that mitigate the effects of corrosion. The discovery programs discussed below verify that the protective coatings are maintained.

Discoverv: Calvert Cliffs Administrative ProcWure MN t 319, " Structure and System Walkdowns,"

provides for discovery of corrosion of steel (or conditions that would accelerate corrosion, such as pooled water) for the intake Structure by performance of visual inspections during plant walkdowns.

The purpose of the program is to provide direction for the perfonnance of structure and system walkdowm and for the documentation of the walkdown results. This program is applicable to the intake Structure steel componenk. [ Reference 22, Section 1.1)

Under this program, responsible personnel perform period:c walkdowns of their assigned structures and systems. Walkdowns may also be performed as requirei for reasons such as: material condition assessments; system reviews before, during, and afler outagts; start up reviews (i.e., when the system is initially pressurized, energized, or placed in sersice); aad as requir a for plant modifications.

[ Reference 22, Section 5.1)

One of the objectives of the program is to assess the condition of the CCNPP structures, systems, and components such that any abnormal or degraded condition will be identified, do< umented, and corrective actions taken before the condition proceeds to failure of the structures, syrems, and components to Application for License Renewal 3.3C-19 Calvert Cliffs Nuclear Power Plant

A*ITACilMENT (M APPENDIX A TECIINICAL INFORMATION 3.3C INTAKE STRUCTURE perform their intended functions. Conditions adverse to quality are documented and resolved by the CCNPP Corrective Actions Program. [ Reference 22, Sections 5.1.C. 5.2.A.1, and 5.2.A.5]

The program provides guidance for identification of specific tyoes of degradation or conditions when performing the walkdowns. Inspection items related to aging management include the following:

[ Reference 2, Section 5.2 and Attachments 1 through 13]

Items related to specific ARDMs such as corrosion; Effects that may have been caused by ARDMs such as damaged supports; concrete degradation, e

anchor bolt degradation, or leakage of fluids; and Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage e

of fluids, presence of standing water or accumulated moisture, or inadequate support of components (e.g., missing, detached, or loose fasteners and clamps).

The program includes a walkdown checklist specifically for the intake Structure. The checklist includes a section targeted at ttructural steel components. Checklist items include visual inspection for corrosion, rust stains, and I;aking/ bub'oling of protective coatings. [ Reference 22, Attachment 6)

A structure perfonnance assessment is cunently required for Category I structures at CCNPP at least once every six years. The aesessment includes a review of each structural component that could degrade the overall puformance of the structure. The program will be modified to add guidance regarding approval authority for significant departures from the walkdown scope / schedule specified.

[ Reference 22, Section 5.3)

The corrective actions taken as a result of the program described above will ensuce that the intake Structure steel components will remain capable of performing their intended functions under all CLB conditions.

Group 3 -(steel components subject to corrosion)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of steel components for the intake Structure:

The intake Structure steel components contribute to one or more of the various intake Structure intended functions as shown in Table 3.3C 2.

The steel components are subject to corrosion due to the normal ambient envircnmental conditions. This ARDM, if unmanaged, could eventually result in the steel components not being able to perform their intended functions under CLB conditions, Corrosion is mitigated by t.pplying protective coatings to the steel components and by e

periodically examining the components for degradation of that coating or conditions that could accelerate degradation.

Calven Cliffs procedure MN-t-319 provides for periodic visual inspections of these components during walkdowns of the intake Structure. If any degradation is found, the appropriate corrective actions are taken to ensure that the intended functions will be maintained.

Application for License Renewal 3.3C-20 Calvert Cliffs Nuclear Power Plant

NITACHMENT (2)

APPEN1JIX A - TECilNICAL INFORMATION 3.3C INTAKE STRUCTURE Rerefore, there is reasonable assurance that the effects of aging due to corrosion of steel will be managed such that the steel components of the intake Structure will be capable of performing their intended functions, consistent with the CLB, during the period of extended operation.

Group 4 -(sluice gates subject to corrosion)- Materials and Environment Group 4 includes the intake Structure sluice gates, which are subject to corrosion. Rese intake Structure unique tempments provide structural and/or functional support to SR equipment as shown in Table 3.3C 2. He speelHc. corrosion mechanisms that are plausible are crevice corrosion, MIC, and pitting. [ Reference 8, Appendix K, Section 2.2)

The sluice gates are used to isolate the cireulating water pump intet bays from the saltwater pump suction pits for maintenance purposes. There are a tota! of twelve sluice gates (six for each CCNPP unit, two associated with each saltwater pump). Each sluice gate consists of a gate, a gate frame, a stem, a lin mechanism, and two wire rope / chain assemblies. The gate slides vertically in the gate frame to shut or open the tunnel to the associated saltwater pump. The stem is attached to the top of the gate and extends up to the lin mechanism. De two wire rope / chain assemblies are provided to hold the sluice gate open in case the stem or lin mechanism falls. The SR function of each gate is to remain in a Oxed position above the tunnel so as to r.ot inhibit now to the Saltwater System. The two wire rope / chain assemblies and associated fittings are credited for this function. The wire rope is constructed of monel. The chain and Attings are constructed of Type 316 stainless steel. [ Reference 23; Reference 24. Page 65]

The lower portions of the wire rope / chain assemblics are subject to a saltwater environment (i.e., intake water from the Chesapeake Bay). The concentrations of sulfates and chlorides in the intake water are described above in the Aging Mechanism Effects section for Group 2. The upper portions of the wire rope / chain assemblies are subject to the outdoor envP mmental conditions expected at the CCNPP site.

Group 4 -(sluice gates subject to corrosion). Aging Mechanism Effects Crevice corrosion is intense, localized corrosion within crevices or shielded areas. It is associated with a small volume of stagnant solution caused by holes, gasket surfaces, lap joints, crevices under bolt hends, and other mechanicaljoints that have a crevice geometry. The crevice must be wide enough to permit liquid entry and narrow enough to maintain stagnant conditions, typically a few thousandths of an inch or less. Crevice corrosion is closely related to pitting corrosion and can initiate pits (i.e., loss of material) in many cases, in an oxidizing environment, a crevice can set up a differential aeration cell to concentrate an acid solution within the crevice. Even in a reducing environment, alternate wetting and drying can concentrate aggressive ionic species to cause pitting and crevice corrosion. [ Reference 25, Attachment 7 i

for Valves]

Pitting is a form of localized attack with greater corrosion rates at some locations than at others. This form of corrosion essentially produces holes of varying depth to diameter ratios in the metal. liigh concentrations of impurity anions such as chlorides and sulfates tend to concentrate in the oxygen depleted pit region, giving rise to a potentially concentrated aggressive solution in this zone.

[ Reference 25, Attachment 7 for Valves]

Application for License Renewal 3.3C-21 Calvert Cliffs Nuclear Power Plant

NITACHMENT m APPENDIX A - TECHNICAL INFORMATION i

3.3C - INTAKE STRUCTURE l

r_

Microbiologically induced corrosion is accelerated corrosion of materials resulting from surface microbiological activity. Sulfate reducing bacteria, sulfur oxidizers, and iron oxidizing bacteria are most commonly associated with corrosion effects. This ARDM most ollen results in pitting, followed by excessive deposition of corrosion products. Stagnant or low How areas are most susceptible, and i

sedimentation aggravates the problem. Temperatures from about 50'F to 120'F are most conducive to MIC. [ Reference 25, Attact ment 7 for Valves]

Crevice corrosion and pitting are plausible for monel and stainless steel in a saltwater environment. He monel and stainless subcomponents of the wire rope / chain assemblies and associated fittings are susceptible to crevice corrosion and pitting due to the presence of sulfates and chlorides in the intake water. [ Reference 25, Attachments 3,4,5, and 6 for Group ids llV.01 and RV Ol]

Microbiologically induced corrosion is plausible for monel and stairiless steel in a saltwater environment. The monel and stainless subcomponents of the wire rope / chain assemblies and associated tittings are susceptible to MIC since sulfate-reducing bacteria, sulfur oxidizers, and iron oxidizing bacteria may be present in the intake water. [ Reference 25, Attachments 3,4,5, and 6 for Group ids llV.01 and RV.01]

The sluice gates have experienced corrosion in the past. Moderate corrosion of the sluice gates was observed in 1986, and new sluice gates were subsequently installed. [ Reference 26) his aging mechanism, if unmanaged, could eventually result in a loss of material such that the sluice gates may not be able to perform their intended function under CLB conditions, nerefore, corrosion was determined to be a plausible ARDM for which the aging effects must be managed for the sluice gates.

[ Reference 8, Appendix K Section 2.3]

Group 4 -(sluice gates subject to corrosion)- Methods to Manage Aging Mitigation: The effects of corrosion cannot be completely prevented, but it can be mitigated by design through selection of materials appropriate for saltwater service.

Discoverv: The effica of corrosion are detectable by visual inspection. A visual examination by a person familiar with the components can be used to determine general mechanical and structural condition and check for corrosion. Corrective action for any metal degradation can be carried out as necessary.

Group 4 -(sluice gates subject to corrosion)- Aging Management Programs Mitigation: No programs are credited for mitigation of corrosion of the sluice gates.

Discoverv: The sluice gates are subject to periodic inspection through existing PM activities as part of the CCNPP PM Program, ne sluice gate inspections are urrently performed each refueling outage with the intake Structure cavities dewatered through the use of sNp logs. The wire rope / chain assemblies and associated Otting are typically inspected as part of the w gate inspections. However, the P_M tasks do not specifically identify the wire rope / chain assembhe.

.d associated Ottings as subcomponents of the sluice gates that require inspection. Therefore, the existing PM tasks will be modined or new PM tasks Application for License Renewal 3.3C-22 Calvert Cliffs Nuclew Power Plant

ATTACitMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.3C INTAKE STRUCTURE

\\

will be initiated to provide specific instructions for inspection of these subcomponents. Periodic inspection of the wite rope / chain assemblies and associated fittings verifies that no sign 10 cant corrosion is occuning and corrective actions are taken as required to casure that the sluice gates will remain capable of performing their intended function under all CLB conditions. [ References 27 and 28]

The CCNPP PM Program details are discussed above in the Aging Management Program section for Group 2.

Group 4 -(stulee gates subject to corrosion). Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of the sluice gates:

The sluice gates provide structural and/or functional suppet to SR equipment.

& sluice gates are subject to corrosion due to exposure to the saltwatei environment. This e

ARDM, if t..unanaged, could eventually result in the sluice gates not being able to perform their intended function under CLB conditions.

The PM Program conducts periodic inspections of the slu;ce gates through performance of e

various PM tasks that provide the means to discover and manage corrosion. Corrective actions are taken to correct any deficiencies that are found to ensure that the sluice gates remain capable of performh., ' heir intended function under all CLB conditions.

Therefore, there is reasonable assurance that the effects of aging due to corrosion will be managed such that the sluice gates will be capable of perforraing their intended function, consistent with the CLB, during the period of extended operation.

3.3C.3 Conclusion The aging management programs discussed for the htake Structure are li.ted in the following table.

These programs are (or will be for new programs) administratively controlled by a formal review and approval process. As demonstrated above, these programs will manage the aging mechanisms and their effects such that the intended functions of the intake Structure will be uaintained during the period of extended operation, consistent with the CLB, under all design loading conditions.

The analysis / assessment, conective action, and confirmation / documentation process for license renewal is in accordance with QL-2, " Corrective Actions Program." QL 2 is pursuant to 10 CFR Part 50, Appendix B, and covers all structures and components subject to AMR.

Application for License Renewal 3.3C-23 Calvert Cliffs Nuclear Power Plant

(

ATTACilME*'* [2]

APPENDIX A TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE TABLE 3.3C-4 LIST OF AGING MANAGEMENT PROGRAMS FOR TIIE INTAKE STRUCTURE Program Credited As Existing CCNPP Technical Procedure Discovery of weathering effects for caulking, STP.F 5921/2," Penetration Fire scalants, and expansionjoints that function as fire Barrier inspection" barriers for the intake Structure. (Group 1)

Modified PM Program

~

Discovery of aggressive chemical attack and Repetitive Tasks 10092042, C".si n f embedded steel /rebar effects on the "I'"

"'I"C

I ' "#"I' 0" "I "I"E ""U" 10092043,10092044,10092045, 10092046,10092047,20092039, and slabs for the intake Stmetum. (Group 2) 20092040,20092041,20092042, 20092043, and 20092044 for intake Structure Cavity Repairs and Cleaning during FJueling Outages Modified CCNPP Admin'iative Procedure Discovery of corrosion effects for steel MN.1 319,' structure and System components in the intake Structure. (Group 3)

Walkdowns" Modified PM Program Discovery of corrosion effects for sluice gates.

or New PM tasks for inspection of Sluice (Group 4)

Gates

=

New Caulking and Sealant inspection Discovery of weathering effects for caulking, Program calants, and expansion joints that do not function as fire barriers for the intake Structure. (Group 1)

Application for License Renewal 3.3C.24 Calvert Cliffs huclear Power Plant

ATTACHMENT (2)

APPENDIX A - TECitNICAL INFORMATION 3JC. INTAKE STRUCTURE 3.3C.4 References 1.

CCNPP " Integrated Plant Assessment Methodology " Revision 1, January ll,1996 2

CCNPP ' Component Level Scoping Results for the intake Structure," Revision 2 February 12,1997 3.

CCNPP Updated Final Safety Analysis Report, Revision 20 4.

CCNPP " Life Cycle Management System and Structure Screening Results," Revision 5 September 10,1997 5.

CCNPP Drawing 61230, " Salt Water Systems Underground Ducts Plan and Sections,"

Revision 6, October 15,1990 6.

CCNPP Drawing 63874S110004, "SR Ductbank Udr West Plant Road Plan," Revision 0, April 4,1995 7.

CCNPP Drawing 63874S110005, " Underground Conduit West of Turbine Building Plan,"

Revision 0, July 15,1996 8.

CCNPP " Aging Management Review Report for the Intake Structure," Revision 3A, February 12,1997 9.

CCNPP " Aging Management Review Report for Component Suppons," Revaion 3 February 4,1997 10.

Electric Power Research Institute Report TR 103842, " Class 1 Structures License Renewal industry Report," Revision I, July 1994 11.

CCNPP Specification A 0010 (Bechtel Specification No. 6750.A 10), " Specification for Furnishing, Delivery and Application of the Caulking and Sealants," Revision 1, March 3,1971 12.

CCNPP Specification C-0010 (Bechtel Specification No. 6750-C-10), " Specification for Forming, Placing, Finishing, and Curing Concrete," Revision 9, January 8,1976 13.

CCNPP Technical Procedure STP F 5921, " Penetration Fire Barrier inspection," Revision 3,

.ugust 26,1997 i

CCNPP Technical Procedure SV F 592 2, " Penetration Fire Barrier inspection," Revision 2, August 26,1997 15.

BGE " Quality Assurance Policy for the Calven Cliffs Nuclear Power Plant," Revision 48, March 28,1997 16.

Letter from Mr. L. T. Doerflein (NRC) to Mr. C. IL Cruse (BGE), dated May 14,1997, " Plant Performance Review (PPR)- Calvert Cliffs" 17.

Letter from Mr. J. T. Trapp (NRC) to Mr. R. E. Denton (DGE), dated May 6,1994, " Combined Inspection Report Nos. 50-317/94-15 and 50-318/94 15" 18.

" Final Environmental Statement related to Operation of Calvert Cliffs Nuclear Power Plant Units I and 2," Baltimore Gas and Electric Company, Dockets Nos. 50 317 and 50 318, United States Atomic Energy Commission, Directorate of Licensing, April 1973 Application for License Renewal 3.3C 25 Calvert Cliffs Nuclear Power Plant

ATTACifMENT (2)

APPENDIX A - TECIINICAL INFORMATION 3.3C - INTAKE STRUCTURE 19.

CCNPP NUCLEIS Database, Repetitive Tasks 10092042, 10092043, 10092044, 10092045, 10092046,10092047, 20092039, 20092040, 20092041, 20092042, 20092043, and 20092044 for intake Structure Cavity Repairs and Cleaning during Refueling Outages 20.

CCNPP Administrative Procedure QL 2100," Issue Reporting and Assessment," Revision 8, December 8,1997 21.

CCNPP Administrative Procedure MN 1 102," Preventive Maintenance Program," Revision 5, September 27,1996 22.

CCNPP Administrative Procedure MN 1319," Structure and System Walkdowns," Revision 0, September 16,1997 23.

CCNPP Drawing 61841," Intake Structure Sluice Gates and Stop legs," Revision 8, June 6,1996 24.

CCNPP Engineering Standard ES-Oll, " System, Structure, and Component (SSC) Evaluation,"

Revision 2, September 15,1997 25.

CCNPP " Aging Management Review Report for the Saltwater System," Revision 4, February 11,1997 26.

Letter from Mr. E. C. Wenzinger (NRC) to Mr. J. A Tiernan (BGF), dated December 19,1986, "NRC Resident inspection 50 317/86-18,50-318/86-18" 27.

CCNPP Technical Procedure IIE-48," Sluice Gate inspection," Revision 1. April 23,1996 28.

CCNPP NUCLEIS Database, Repetitive Tasks 10122016, 10122010, 10122012, 10122051, 10122052,10122015,20122041,201220$0,20122051,20122052,20122053, and 20122043 for Sluice Gate Inspections per Procedure llE-48 Application for License Renewal 3.3C 26 Calvert Cliffs Nuclear Power Plant

l NrrACIIMENT (3)

APPENDIX A - TECilNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES 3

Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 3,1998

ATTACllMENT (3) l APPENDIX A - TECHNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES l

3.3D Miscellaneous Tank and Valve Enclosure

• i This is a section of the Baltimore Gas and '

.c Company (BGE) License Renewal Application (LRA), addressing the No.12 Condensate Stu 1 Tank (CST) Enclosure, the No. 21 T -1 Oil Storage Tank (FOST) Enclosure, and the Auxiliary Feeds 'er (AFW) Valve Enclosure. These enc. Sures were evaluated in accordance with the Calvert Clin Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. ~hese sections are prepared independently and will, collectively, comprise the entire BGE LRA.

3.3D,1 Structures Scoping The systems and structures scoping task identifies structures within the scope of license renewal on the basis of how their design supports generic structural functions satisfying the 10 CFR 54.4(a) scoping criteria. The component level scoping process for structures is conducted on the basis of a generic listing of structural component types. Scoping is implemented by determining which structural component types are required for performance of the passive intended functions of the structure.

By their nature, structures within the scope of license renewal are constructed in accordance with predetermined design requirements to support perfonnance of specific structural functions. Civil engineers experienced with nuclear plant structures established the following generic list of structural functions for CCNPP. A structure is considered to be within the scope oflicense renewal if it performs one or more of these structural functions: [ Reference 1, Section 4.2.2]

Provide structural and/or functional support to safety related (SR) equipment; Provide shelter / protection to SR equipment; e NOTE: This function includes: (a) protection from radiation effects for equipment addressed by the Environmental Oualification Program; and (b) protection from High Energy Line Break

effects, Serve as a pressure boundary or a fission product retention barrier to protect public health and e

safety in the event of any postulated Desi n Basis Events; t

Serve as a missile barrier (intemal or external);

e Provide structural and/or functional support to non safety related (NSR) equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions (e.g., seismic Category 11 over i design considerations);

Provide flood protection barrier (internal flooding event); and e

Provide rated fire barriers to confine or retard a fire from spreading to or from adjacent areas of e

the plant.

This section begins with a description of the miscellaneous tank and valve enclosures. The intended functions performed by each enclosure are listed and used to identify the structural component types within the scope oflicense renewal (i.e., those required to perform the intended functions). Finally, the components subject to Aging Management Review (AMR) are identified and dispositioned in accordance with the CCNPP IPA Methodology.

Application for License Renewal 3.3 D-1 Calvert Cliffs Nuclear Power Plant

I ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.30 - MISCELLANEOUS TANK AND VALVE ENCLOSURES Representative historical operating experience pertinent to aging is included.. on ::,,riate areas to c

provide insight supporting the aging management demonstrations. His ortrating r.perience was obtained through key. word searches of BGE's electronic database of informat'on on th< CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP prst.nel.

Structure Descriotion/Concentual Boundaries Figure 3.3D 1 is a simplified layout showing the site structures that are within the scope of license renewal, including the No.12 CST, No. 21 FOST, and AFW Valve Enclosures. [ References 2 through 5)

A comprehensive layout and description of all site structures is provided in the Updated Final Safety Analysis Report, Chapter 1, with further discussion of their design features in Chapter 5 and Appendix SA. [ Reference 6 Chapters 1,5, and Appendix SA) A general description, boundary, and design discussi<n er the enclosures addressed in this scetion follows:

No.12 CST Enclosure is located in the tank farm area north of the Turbine Building and is common for both Units I and 2.

It houses and protects No.12 CST, which provides demineralized water for decay heat removal and cooldown of Units 1 and 2.

[ Reference 7, Section 1.1.1; Reference 8, Section 1.1.11 (ne No.12 CST is within the scope oflicense renewal for the AFW System, which is evaluated in Section 5.1 of the BGE LRA.) The structural boundary comprises all of the enclosure's structural components such as walls, foundation slab, and roof slab. [ Reference 8, Section 1.1.2] The No.12 CST Enclosure is required to meet Seismic CategoryI criteria because it houses SR systems, equipment, or components that must remain functional before, during, or afler a safe shutdown earthquake. [ Reference 9, pages 46,57, and 58] Additionally, the structural boundary includes structural or functional supports for NSR roof drains and tank vents. During an abnormal event such as a seismic event, failure of these NSR equipment supports must not adversely affect the operability of SR components. The enclosure is a reinforced concrete structure of sufficient thickness to stop tornado-generated missiles and to resist tornado wind pressures. Bursting pressures are relieved by bafiled, missile proof vents. [ Reference 6, Section 10.3.2]

No. 21 FOST Enclosure is located in the yard area west of the Unit 2 Containment Structure and e

is common for both Units I and 2. It houses and protects No.21 FOST, which provides a fuel supply for the three emergency diesel generators installed in the Auxiliary Building.

{ Reference 6, Sections 1.2.2 and 8.4.1.2; Reference 10, Section 1.1.1] (T.ie No.21 FOST is within the scope of license renewal for the Diesel Fuel Oil System, which is evaluated in Section 5.7 of the BGE LRA.) He structural boundary comprises all of the enclosure's structural components such as concrete foundations, walls, and slabs. Additionally, the structural boundary includes structural or functional supports for NSR stairs and platforms During an abnormal event such as a seismic event, failure of these NSR equipment supports must not adversely affect the operability of SR components. [Refi.Tnce 10, Section 1.1.2) The No.21 FOST Enclosure is required to meet Seismic CategoryI criteria because it houses SR systems, equipment, or components that must remain functional before, during, or after a safe shutdown earthquake.

[ Reference 9, pages 46,57, and 58] The enclosure is a reinforced concrete structure designed to protect No.21 FOST from tomadoes and tomado missiles. Bursting pressures are relieved by baffled, missile-proof vents. The structure will also withstand the impact of a transmission tower fal!ing on it without damage to the No. 21 FOST. The enclosure also acts as a dike in the event of a tank failure. { Reference 6, Section 8.4.1.2]

Application for License Renewal 3.3 D-2 Calvert Cliffs Nuclear Power Plant

ATTACllMENT 0)

APPENDIX A - TECilNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES Y

N w.c i c e.n

' ' 'u l# %

y'~~

INTAKE STRUCTURE

-l w%

CONDENSATE STORAGE TANK No.12

% g,

,,),,

• ENCLOSURE TURBINE BUILDING

%a

.r" Y

.r"

]

c-e

,. = -.

spersFueiPooll % l AUXILIARY FEEDWATER VALVE moom ENCLOSURE l Eh 4F.F ig

~ ~ "~ 9""'S ~ ~f:~::' t"""* ' P' "~ ~

n, ONTAINME ONTAINME we.%.,*

AUXILIARY BUILDING w %. t..

n T

7 SAFETY RELATED DIESEL OENERATOR

, se "U

'"8 nw w,.,* _f I

(% w,,,*

Pump Room w am FUELOIL STORAGE TANK s ex. oues.* so, No. 21 tA N

1A ENCLOSURE oc ll Expansion Joint STRUCTURES WITHIN THE SCOPE OF N,',*

,O LICENSE RENEWAL FIGURE 3.3D-1 CCNPP SITE STRUCTURES (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY)

Application for License Renewal 3.3 D-3 Calvert Cliffs Nuclear Power Plant I

ATTACHMENT di APPENDIX A TECHNICAL INFORMATION 3.3D MISCELLANEOUS TANK AND VALVE ENCLOSURES AFW Valve Enclosure is located in the tank farm area north of the Turbine Building and is common for both Units I and 2. It houses and protects the AFW pump suction valves and l

associated manifold piping, which provide a pressure boundary function for the AFW System, i

[ Reference 11, Table 2; Reference 12] (Th. compor. nts inside this enclosure are within the scope oflicense renewal for the AFW System, which is esaated in Section 5.1 of the BGE LRA.) The stmetural boundary comprises all of the enclosure's structural components such as concrete foundations, walls, and slabs. Additionally, the stmetural boundary includes structural or functional supports for NSR manhole steps and grating. During an abnonnal event such as a seismic event, failure of these NSR equipment supports must not adversely affect the operability of SR components. The AFW Valve Enclosure is required to meet Seismic Category I criteria because it houses SR systems, equipment, or components that must remain functional before, during, or after a safe shutdown carthquake. The enclosure is a reinforced concrete structure designed to withstand and protect its associated piping from design basis loadings (e.g., weight, thermal, seismic, and wind). [ References 9 and 13]

Component supports that are connected to structural components in the miscellaneous tank and valve enclosures are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA. A " component support" is the connection between a system, or component within a system, and a plant structural member. Component supports interface with the components they support in the applicable systems, and they interface with the structural component to which they are attached. For example, a fixed base supporting a pump is considered a component support since it connects the concrete equipment pad to the pump. The pump itself would be scoped within its associated system evaluation. The fixed base would be scoped within the Component Supports Commodity Evaluation, and the concrete equipment pad would be scoped within the evaluation for the associated structure, if anchor bolts are used at the interface with the structural member, there is overlap between the Component Supports Commodity Evaluation and the evaluation for the structural component.

Evaluations for structural components considered the effects of aging caused by the surrounding environment, while the Component Supports Commodity Evaluation considered the effects of aging caused by the supported equipment Neimal expansion, rotating equipment, etc.) as well as the surrounding dronment.

Supporta for structural components (e.g., platform hangers) are not

" component supports"in this sense tecause any support for a structural component is itself a structural component (l.c., included in the scope of the associated structure) [ Reference 14, Section 1.1.1]

Sconed Structures and Their Intended Functions The No.12 CST, No. 21 FOST, and AFW Valve Enclosures are in scope for license renewal based on 10 CFR 54.4(a). Four of the seven generic s:ructural functions listed above are applicable to the miscellaneous tank and valve enclosures as shown in Table 3.3D-l. The intended functions for these enclosures were determined based on the requirements of l54.4(a)(1), l54.4(a)(2), and IS4.4(a)(3) in accordance with the CCNPP IPA Methodology Section 4.2.2.

[ Reference it Reference 2, Table 2; Reference 8, Section 1.1.3; Reference 10, Section 1.1.3]

Application for License Renewal 3.3D-4 Calvert Clifts Nuclear Power Plant

ATTACitMENT (3)

APPENDIX A - IECIINICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES TABLE 3.3D -1 INTENDED FUNCTIONS FOR MISCELLANEOUS TANK AND VALVE ENCLOSURES Applicable to Applicable

' Function Miscellaneous 10 CFR 54.4(a)

Tank and Valve Criteria Enclosures?

1.

Provide structural and/or functional support to SR Yes 654.4(a)(1) equipment 2.

Provide shelter / protection to SR equipment Yes 654.4(a)(1)

NOTE: These structures are not required to provide protection from radiation or liigh Energy Line Break effects.

3.

Serve as a pressure boundary or a fission product retention No

$54.4(a)(1) barrier to protect public health and safety in the event of any postulated Design Basis Events 4.

Serve as a missile barrier (internal or external)

Yes 654.4(a)(1) 5.

Provide structural and/or functional support to NSR Yes

{$4.4(a)(2) equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions (e.g., seismic Category 11 over i design considerations) 6.

Provide flood protection barrier (internal flooding event)

No

$54.4(a)(2) 7.

Provide rated fire barriers to confine or retard a fire from No 954.4(a)(3) spreading to or from adjacent areas of the pbnt Comnonents Subject to AMR A generic list of structural component types was developed for use during the structural component scoping task. He generic list started with structural component types associated with SR functions contained in industry technical reports addressing Containment and Category i Structures. Other structural component types related to fire and flooding events were added to the list to ensure completeness. [ Reference 1, Section 4.2.3] nese structural components were combined into the following four struc'. ural categories based on their design and materials: [ References 15 and 16)

Concrete components; Structural steel components; e

Architectural components; and Unique comporants.

e During the scoping process, applicable structural component types actually contained in the No.12 CST, No 21 FOST, and AFW Valve Enclosures were identified. Within the four structural component categories,17 structural component types were determined to contribute to at least one of the structural intended functions listed in Table 3.3D 1 for the associated enclosure. Table 3.3D-2 lists the structurel component types and the associated functions that apply to the No.12 CST, No. 21 FOST, and AFW Valve Enclosures. Unless otherwise noted, structural components that are part of the structure, but do Application for License Renewal 3.3 D-5 Calvert Clifts Nuclear Power Plant

ATTACllMENT (3)

APPENDIX A - TECIINICAL INFOlG1ATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES not contribute to any of the intended functions of the structure, are not listed in Table 3.3D 2.

[ References 15,16, and 17]

Per the license renew ! rule,"... Structures and components subject to an aging management review shall encompass those structures and components (i)'Ihat perform an intended function, as described in

$54.4 without moving parts or without a change in configuration or propedies... and (ii) That are not subject to periodic replacement based on a qualified life or specified time period... " From reviewing the generic list of structural functions, it is clear that none of the intended structural funMions requires moving parts or a change in configuration or properties. Plant structural components are not normally subject to periodic replacement programs; therefore, they are considered to be long lived unless specific justification is provided to the contrary. [ Reference 1, Section 5.4)

Dased on the reselts of the process described above, the 17 structural component types, listed in Table 3.3D 2, are subject to AMR and are evaluated within this section. [ References 8 and 10, Table 21; Reference 17]

1 Daltimore Gas and Electric Company may elect to replace components for which the AMR identifies that further analysis or examination is needed in accordance with the License Renewal Rule, components subject to replacement based on qualified life or specified time period would not be subject to AMR.

Application for License Renewal 3.3D-6 Calvert Cliffs Nuclear Power Plant

_ _ ~..._

NITACilMENT (3)

APPENDIX A TECilNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES TABLE 3.3D.2 STRUCTURAL COMPONENT TYPES REQUIRING AMR FOR MISCELLANEOUS TANK AND VALVE ENCLOV3ES No. 21 FOST No.12 CSI AFW Valve Enclosure Enclosure Enclosure Concrete (inclutling Reinforcing Steel)

Foundations (Footings, beams, and mats)

I,2 1, 2 1, 2 Walls 2, 4, 5 1, 2, 4, 5 1, 2, 4 RoofSlabs 2, 4, 5 2,4,5 2, 4 Cast in Place Anchors /Embedments' I, 2, 4, 5 1, 2, 4 1, 2, 4 Grout 1, 2, 4, 5 2

NA Post Installed Anrhors' 5

1, 5 NA StructuralSteel

^

13eams' 2, 4, 5 2, J. 5 NA 13aseplates*

2, 4, 5 2,4,5 NA Roof Framing

• 2, 4, 5 2,4,5 NA 13 racing

• 5 NA NA Platform llangers' 5

none NA Decking

• 2,4,5 2, 4, 5 2, 4 Floor Grating

• 5 none 5

Stairs and Ladders

• 5 none 5

ArchiMtural Components Unique Components Anchor 13 rackets' 1

NA NA Manhole Framing

• NA NA 2, 4 Manhole Cover' NA NA 24

. Indicates the component type is included under the heading " Steel Components" in the discus?lon addressing the results of AMR in Table 3.3D-3

(#) - Indicates the component type provides the following intended function for the corresponding structure; I

- Provide structural and/or functional support to SR equipment; 2 - Provide shelter / protection to SR equipment; 4

Serve as a missile barrier (internal or external); and 5

Provide structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions.

none - Indicates the component type does not contribute to the intended functions of the structure NA Indicates the component type is not part of the corresponding structure Application for License Renewal 3.3 D-7 Calvert Cliffs Nuclear Power Plant

ATTAClIMENT (3)

APPENDIX A TECHNICAL INFORMATION 3.31'. MISCELLANEOUS TANK AND VALVE ENCLOSURES 3.3D.2 Aging Management The list of potential Age Related Degradation Mechanisms (ARDMs) identiDed for miscellaneous tank and valve enclosure components is given in Table 3.3D 3, with plausible ARDMs identiDed by a check mark (/) in the appropriate column. [ References 8 and 10, Attachments 1 and 2] For ef0ciency in presenting the results of these evaluations in this report, ARDM/ component type com*vinations are grouped together where there are similar characteristics and the discussion is :pplicable to all components. Table 3.3D 3 also identines the group to which each ARDM/ component type combination belongs. One group has been selected for the No.12 CST, No. 21 FOST, and AFW Valve

Enclosures:

Group 18 terrosion of steel (for components marked with an asterisk in Table 3.3D 2).

TAHLE 3.3D-3 POTENTIAL AND PLAUSlHLE ARDMs FOR MISCELLANEOUS TANK AND VALVE ENCLOSURES Potential ARDMs Foundations Walls Roof Slabs Grout Compo nts*

l'reeze Thaw Leaching of Calcium liydroxide Settlement Corrosion of Steel

/(1)

" Steel Components" represer.ts all items marked with an asterisk (*) in Table 3.3D-2

/

Indicater plausible ARDM determination

(#) - Indicates the group (s)in which the ARDM/ component type combination is evaluated Aging mechanisms that are not plausible are generally not discu2 sed further in these BGE LRA sections, unless they ate considered noteworthy. For the Miscellaneous Tank and Valve Enclosures, settlement is considered notewonhy and is discussed below, industry technkal reports conclude that settlement is a potentially signincant ARDM for pressurized water reactor Containment Structures and for other Category i Structures at some plants. [ Reference 18, Section 5.5; Reference 19, Section 5.1.2] Settlement occurs both during construction and aRer construction. Th amount of settlement depends on the physical properties of the foundation material.

'crerences 8 and 10, Appendix Js] Excavation unloading and structural loading during construction

.aused a small change in the void ratio of ' ndisturbed soil. This change results in a very small or tagligible amount ci time-dependent settlement. [ Reference 6, Section 2.7.6.2; References 8 and 10, Appendix Js] Compacted soil is subject to some degree of settlement in the Orst several months after construction. [ Reference 19, Section 4.6.3.1) Settlement directly related to construction work is readily evident early in the life of the structure and is not considered to be an ARDM. Settlement may occur during the design life of the structure from changes in environmental conditions, such as lowering of the groundwater tab:e. Sites with son soil and/or sites with signincant changes in underground water conditions over a long period of time may be susceptible to signincant settlement. [ References 8 and 10, Appendix Js; Reference 18, Section 4.5.3.2; Reference 19, Section 4.6.3.2] Concrete and steel structural members can be affected by differential settlement between supporting foundations, within a building, or Application for License Renewal 3.3 D-8 Calvert Cliffs Nuclear Power Plant

Ar(ACHMENT (3)

(

APPENDIX A - TECHNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES between buildings. Sewre settlement can cause misalignment of equipment and lead to overstress conditions within the structure. When buildings experience significant settlement, cracks in st:actural members, differential elevations of supporti:.g members bridging between buildingr, or both may be visibly detected. [ References 8 and 10, Appendix Js] At CCNPP, long-term settlement was determined to be not plausible for the miscellaneous tank and valve enclosures based on the following site specific justification:

The foundations for the miscellaneous tank and valve enclosures are situated on an engineered soil structure consisting of compacted soil on top of the site's Pliestocene deposit. [ Reference 6, Section 2.7.3; References 8 and 10, Appendix Js; Reference 20] Quality assurance and quality control measures imposed daring backfill placement included specification of maximum lift thicknesses and verification of a minimum fill compaction of 97% based on the standard Proctor compaction method. [ Reference 6, Section 2.7.6.1; References 20 and 21] A continuous program of soil testing during construction assured uniform placement or the compacted fil!.

[ Reference 22] These activities assured that the cat.ses of excessive settlement of plant structures at other nuclear power plant sites did not exist during construction at CCNPP. [ Reference 23]

Control of the placement and compaction for these engineered soil structures was used to obtain the engineering properties required to satisfy foundation design requirements.

j The foundations for the miscellaneous tank and valve enclosures are above the groundwater table.

[ References 13, 24, and 25] The elevation of the groundwater table beneath these structures changes with the surface topography and can be expected to fluctuate slightly as a result of climatic changes. [ Reference 6 'Ntion 2.5.3.3] Significant deviations from the seasonal cycles and occasional meteorologicJ frects (e.g., drought conditions) observed over the past 25 years are not expected during the period of extended operation.

The foundations for the No.12 CST, No. 21 FOST, and AFW Valve Enclosures tend to uniformly settle as rigid bodies. Most of the predicted settlement is expected in terms of uniform settlement, which has no adverse effect on structural components of the miscellaneous tank and valve enclosures. [ References 8 and 10, Appendix Js] The effects of one-time building settlement are included in the stresses allowed by design codes and standards for piping systems.

Any differential settlement is expected to be small and have negligible effect.

At CCNPP, no cracking or other evidence of settlement that would affect structural integrity has been observcd to date. Walkdown inspections of the miscellaneous tank and valve enclosures, performed in 1994, found no indication of structural damcge due to settlement. [ References 8 and 10, Attachment 7s]

An opportunity to inspect the below-grade concrete of the No. 21 FOST Enuosure in 1996 also revealed no indications of concrete crecking. These observations support the conclusion that settlement of the miscellanecus tank and valve enclosures at CCNPP is not plausible.

The following is a discussion of the aging management demonstration process for each group identified above. It is presented by group and includes a discussion of materials and environment, aging mechanism effects, methods of managing agir.g, aging management program (s), and agiag management demonstration.

Application for License Renewal 3.3 D-9 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.3D - MISCr~:LLANEOUS TANK AND VALVE ENCLOSURES Group 1 -(corrosion of steel)- Materials and Environment Group I comprises those components marked with an asterisk in Table 3.3D-2. These components are all fabricated from :arbon steel, which is subject to general corrosion when exposed to moisture and oxygen. They each contribute to one or more of the various passive intended functions for the No.12 CST, No. 21 FOST. and AFW Valve Enclosures. For the purposes of this aging evaluation, the internal environmental ccaditions for these enclosures, which are not " weather tight" facilities, are considered to be the same as the external environment. [ References 8 and 10, Appendix Ks] 'lle CCNPP site is located in a geographic region subject to severe weather conditions. All outdoor components will experience the extreme temperature ranges, rain, snow, and changes in humidity expected at the CCNPP de. Since the air inside these enclosures is not conditioned, the interior components will experience similar temperature and humidity changes throughout the life of the plant. [ References 8 and 10, Appendix Os]

Since corrosion was recognized as a potential degradation mechanism for all carbon steel components of site structures, protative coatings were incorporated into the original design. Exposed structural steel surfaces in the No.12 CST, No.21 FOST, and AFW Vatic Enclosures were coated during the construction phase (e.g., shop-primed, field-painted, hot-dipped galvanized). [ References 8 and 10, Appendix Ks; References 26 through 29]

Group 1 -(corrosion of steel)- Aging Mechanism Effects Steel corrodes in the presence of moisture and oxygen as a result of electrochemical reactions. Initid.ly, the exposed steel surface reacts with oxygen and moisture to form an oxide film as rust. Once the protective oxide film has been formed, and if it is not disturbed by erosion, alternating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from saltwater, the atmosphere, or groundwater, increase the rate of corrosion by increasing the electrochemical activity. If steel is in contact, through an t lectrolytic solution, with another metal that is more noble in the galvanic series, corrosion of the steel may accelerate. [ References 8 and 10, Appendix Ks]

Corrosion products such as hydrated oxides ofiron (rust) form on exposed, u protected surfaces of the steel and are readily visible. The afTected surface may degrade to such an extent that visible perforation may occur, in the case of exposed surfaces of steel with protective coatings, corrosion may cause the protective coatings to lose their ability to adhere to the corroding surface. In this case, damage to the coatings :an be visually detected well in advance of significant degradation of the steel.

[ References 8 and 10, Appendix Ks]

Visual inspection of accessible interior and exterior areas of the No.12 CST and No.21 FOST Enclosures were performed in 1994. [ References 8 and 10, Attachment 7s] The inspectbns identified minor surface corrosion on steel beams of insufficia.t magnitude and severity to be considered a structural integrity issue. No corrosion was observed to be causing damage to protective coatings (e.g., sprayed-on fire-proofing material for roof beams in the tank room in the No. 21 FOST Enclosure; galvanized decking material; painted carbon steel structural components).

Application for License Renewal 3.3 D-10 Calvert Cliffs Nuclear Power Plant

[

ATTACHMENT (3) l

(

APPENDIX A - TECIINICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES If corrosion is left unmanaged for an extended period of time, the loss of carbon steel material can result in a reduction in the load bearing capability of the corroded parts and increased likelihood of mechanical failure. This could lead to the inability of components identified in Table 3.3D-2 to perform their intended functions under CLB design loading conditions. [ References 8 and 10, Appendix Ks]

Group 1 -(corrosion of steel)- Methods to Manage Aging Mitigation: The effects of corrosion cannot be completely prevented, but they can be mitigated by minimizing the exposure of externel surfaces of steel to an aggressive environment and protecting the external surfaces with paint or other protective coating. Coatings serve as a protective layer, preventing moisture and oxygen from directly contacting the steel surfaces.

Discoverv: The effects of general corrosion / oxidation of carbon steel are detectable by visual inspection.

A visual examination by a person familiar with the components can be used to determine general mechanical and structural condition and check for rust. Observing that significant degradation of protective coatings has not occurred is an effective method to ensure that corrosion has not affected the intended function of the structural component. Since the coating does not contribute to the components' intended functions, degradation of the coating provides an alert condition that triggers corrective action before the occurrence of corrosion that vould affect the components' ability to perform their intended functions. The degradation of the protective coating that does occur can be discovered and monitored by periodically inspecting the carbon steel structural components. Corrective action for failed protective coatings and any actual metal degradation can be carried out as necessary. [ References 8 arid 10]

Group 1 -(corrosion of stect)- Aging Management Programs Mitigation: The exposed metal surfaces of carbon steel structural components are covered by protective coatings that mitigate the effects of corrosion. The discovery programs discussed below verify that the protective coatings of carbon steel structural components are maintained.

Discoverv: Calvert Cliffs Administrative Procedure MN-1-319, " Structure and System Walkoowns,"

provides for discovery of corrosion of steel (or conditions that would accelerate corrosion, such as pooled water) for the structural components in the No.12 CST, No.21 FOST, and AFW Valve Enclosures by performance of visual inspections during plant walkdowns. [ References 8 and 10, s] The purpose of the program is to provide direction for the performance of structure and system walkdowns and for the documentation of the wa$down results. [ Reference 30, Section 1.1]

Under this program, responsible personnel perform periodic walkdowns of their assigned structures and systems, Walkdowns may also be performed as required for reasons such as: material condition assessments; system reviews before, during, and after outages; start-up reviews (i.e., when the system is initially pressurized, energized, or placed in service); and as required for plant modifications.

[ Reference 30, Section 5.l]

One of the objectives of the program is to assess the condition of the CCNPP structures, systems, and components such that any a'; normal or degraded condition will be identified, docum,nted, and corrective actions taken before the condition proceeds to failure of the structures, systems, and components to 5 plication for License Renewal 3.3D-11 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

AI'PENDIX A - TECHNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES perform their intended funedons. Conditions adverse to quality are documented and resolved by the CCNPP Corrective Actions Program. [ Reference 30, Sections 5.1.C,5.2.A.1, and 5.2.A.5]

The program provides guidance for identification of specific types of degradation or conditions when performing the walkdowns. Inspection items related to aging management include the following:

[ Reference 30, Section 5.2 and Attachments I through 13]

items related to specific ARDMs such as corrosion; e

Effects that may have been caused by ARDMs such as damaged supports; concrete degradation, 2

anchor bolt degradation, or leakage of fluids; and Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage e

of fluids, presence of standing water or accumulated moisture, or inadequate support of components (e.g., missing, detached, or loose fasteners and clamps).

The Structure and System Walkdown Program enhances the familiarity of responsible personnel with their assigned systems and provides extended attention to plant meterial condition beyond that afforded by Operations and Maintenance personnel alone. The program has been improved recently through incorporation of s:gnificant additional guidance on specific activities to be included in the scope of structures walkdowns. A structure performance assessment is currently required for Category I structures at CCNPP at least once every six years. The assessment includes a review of each structural component that could degrade the overall performance of the structure (including the Group I caroon steel components for the miscellaneous tank and valve enclosures). [ Reference 30, Section 5.3 and ]

The program described above will be modified to: (a) specifically identify the structures within the scope of the performance assessments (including the No.12 CST, No.21 FOST, and AFW Valve Enclosures); and (b) add guidance regarding approval authority for significant departures from the walkdown scope / schedule specified. The modified program will ensure that degraded conditions due to corrosion of steel are identified and corrected such that carbon steel components of the No.12 CST.

No.21 FOST, and AFW Valve Enclosures will be capable of performing their intended functions consistent with CLB design conditions.

Group 1 -(corrosion of steel)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of steel in the structural components of the No.12 CST, No. 21 FOST, and AFW Valve Enclosures.

The carbon steel components identified in Table 3.3D-2 provide various passive intended functions for the associated enclosures, and failure could directly prevent satisfactory accomplishment of fuactions that must be maintained under CLB design loading conditions.

Components in this group are exposed to moisture and oxygen in their installed locations.

Carbon steel cerrodes in the presence of moisture and oxygen, which leads to a loss of material.

This could eventually result in inability of the affected components to perform their intended function (s).

Application for License Renewal 3.3D-12 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORh1ATION 3.3D - h11SCELLANEOUS TANK AND VALVE ENCLOSURES Coatings, specified during original construction, mitigate the effects of corrosi m by providing a protective layer that prevents moisture and oxygen from contacting the steel.

l The CCNPP Structure and System Walkdowns program provides for periodic walkdowns of Group 1 components. The program will be modified to specify more clearly the scope and control of periodic performance assessments. The program will provide for the discovery of corrosion of steel (or conditions that would accelerate corrosion) for the components in Group 1, and ensure appropriate actions are taken in a timely manner to correct degraded components or protective coatings.

Therefo,e, there is reasonable assurance that the effects of aging due to corrosion of carbon steel will be managed in such a way that structural components of the No.12 CST, No. 21 FOST, and AFW Velve Enclosures will be capable of performing their intended functions consistent with the CLB during the period of extended operation.

3.3D.3 Conclusion The aging management programs discussed for t!" No.12 CST, No. 21 FOST, and AFW Valve Encicsures are listed in the Table 3.3D-4. These programs are administratively controlled by a formal review and approval process. As demonstrated above, these programs will manage the aging mechanisms and their efTects in such a way that the intended functions of the components of the No.12 CST, No.21 FOST, and AFW Valve Enclosures will be maintained during the period of extended operation consistent with the CLB under all design loading conditions.

The analysis / assessment, corrective action, and confirmation / documentation process for license renewal is in accordance with QL-2, " Corrective Actions Program." QL-2 is pursuant to 10 CFR Part 50, Appendix B, and covers all structures and components subjNt to AMR.

TABLE 3.3D -4 AGING MANAGEMENT PROGRAMS FOR MISCELLANEOUS TANK AND VALVE ENCLOSURES Program Credited As Modified Structure and System Walkdowns Program for discovery and management of (MN-1-319) corrosion effects for carbon steel components in the No.12 C '

No. 21 FOST, and AFW Valve

. Specify scope and control of Enclosurs up1) periodic structure performance assessments Application for License Renewal 3.3 D-13 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES 3.3D.4 References l

1.

CCNPP IPA Methodology, Revision 1 2.

CCNPP System and Structure Screening Results, Revision 5 3.

BGE Drawing 61230," Salt Water Systems Underground Ducts Plan and Sections," Revision 6 4.

BGE Urawing 63874SH0004,"SR Ductbank Under West Plant Road Plan," Revision 0 j

5.

BGE Drawing 63874SH0005," Underground Conduit West of Turbine Building Plan," Revision 0 6.

CCNPP Up '.ted Final Safety Analysis Report, Units 1 and 2, Revision 21 7.

CCNPP Aging Management Review Report," Auxiliary Feedwater System (036)," Revision 1 8.

CCNPP Aging Management Revii w Report, " Condensate Storage Tank No.12 Enclosure,"

Revision 2 9,

CCNPP Engineering Standard ES-011, " System, Structure, and Component (SSC) Evaluation,"

Revision 2 10.

CCNPP Aging Management Review Report, " Fuel Oil Storage Tank No.21 Enclosure,"

Revision 2 11.

CCNPP Component Level Scoping Results for System 036 - Auxiliary Feedwater System, Revision 2 12.

BGE Drawing 60717SH0001, "Well Water, Pretreated Water, Demineralized Water and Ccndensate Storage System," Revision 71 13.

BGE Drawing 60481, "Well Water, Pretreated Water & Condensate Storage Tank Piping, 3

Partial Plans and Sections," Revision 20 14.

CCNPP Aging Management Review Report," Component Supports," Revision 3 15.

CCNPP Component Level Scoping Results for Enclosure for Condensate Storage Tank #12, Revision 1 16.

CCNPP Component Level Scoping Results for Enclosure for Fuel Oil Storage Tank #21, Revision 1 17.

BOE Drawing 63798," Yard Condensate Valve Pit Plans and Sections," Revision 1 18.

Electric Power Research Institute, "PWR Containment Structares License Renewal Industry Report; Revision 1," July 1994 19.

Electric Power Research Institute, "ClassI Structures License Renewal Industry Report; Revision 1," July 1994 20.

BGE Drawing 60119," Compacted Fill Areas," Revision 0 21.

Bechtel Specification No. 6750-C-4A," Specification for Placement and Control of Compacted Fill - CCNPP Units I and 2," Revision 3 22.

Bechtel Specification No. 6750-C-Il-B," Specification for Testing of Concrete, Reinforcement and Soil-CCNPP Units I and 2," Revision 1 23.

NRC Inspection and Enforcement Circuin 81-08," Foundation Materials," May 29,1981

. Application for License Renewal 3.3D-14 Calvert Cliffs Nuct,:ar Power Plant

ATTACHMENT d)

APPENDIX A - TECHNICAL INFORMATION 3.3D - MISCELLANEOUS TANK AND VALVE ENCLOSURES 24.

BGE Drawing 63755S110001," Yard Tank Enclosures," Revision i 25.

BGE Drawing 63756SH0003," Yard Tank Enclosures," Revision 2 26.

BGE Drawing 63754S110001," Yard Tank Enclosures," Revision 1 27.

Bechtel Speci0 cation No.6750-C-61(Q), " Technical Specification for Furnishing and Delivering Structural Steel-CCNPP Units 1 and 2," Revision 0 28.

BGE Technical Requirements Document TRD-A 1000, " Coating Application Performance S' adard," Revision 14 29.

Bechtel Specification No. 6750-A-24, " Specification for Painting and Special Coatings - CCNPP Units 1 and 2," Revision 12 30.

CCNPP Administrative Procedure MN-1-319," Structure and System Walkdowns," Revision 0 Application for License Renewal 3.3 D-15 Calvert Cliffs Nuclear Power Plant l

1

ATTACHMENT (4) l APPENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 3,1998 l

4 4

NITACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM 5.11B Primary Containment Heating and Ventilation System This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA) addressing the Piaary Containment Heating and Ventilating (H&V) System. The Primary Containment II&V System was evaluated in accordance wi h the Calvert Cliff Nuclear Power Plant (CCNPP) t Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LR A.

5.11B.1 Seoping System level scoping describes conceptual boundaries for plant systems and structures, develops screening tools which capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify systems and structures within the scope of license renewal. Component level scoping describes the components within the boundaries of those systems and structures that contribute to the intended functions. Scoping to determine components subjet to aging management review (.AMR) begins with a listing of passive intended functions and then dispositions the device types as either only associated with active ftmetions, subject to replacement, or subject to AMR either in this report or another report.

Representative historical operating experience pertinent to aging is included in appropriate areas, to provide insight supporting the aging management demonstrations. This operating experience was obtained through key word searches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.

Section 5.11B.l.1 presents the results of the system level scoping,5.11B.1.2 the results of the component level scoping, and 5.118.1.3 th results of scoping to determine components subject to an AMR.

5.11B.I.1 System Level Seoping This section begins with a description of the system, which includes the boundaries of the system as it was scoped. The intended functions of the system are listed and are used to define what portions of the system are within the scope oflicense renewal.

System Descriotion/Concentual Boundaries The Primary Containment H&V System consists of the six subsystems described below. Also included within the system boundaries is pressure monitoring equipment for the containment and penetration room atmospheres. The pressure for the containment atmosphere is measured to provide signals for the Engineered Safety Features Actuation System (ESFAS) protective actuation and for post-accident monitoring. Penetration room pressure is monitored to provide signals upon high pressure to isolate letdown during a loss-of-coolant accident or a letdown line rupture (high energy line break).

Containment dome temperature, containment cooler fan status, and containment hydrogen purge inside and outsioe containment isolation valve positions are monitored for post-accider. monitoring. Other monitoring equipment is provided to support operations and testing. (References I and 2]

The containment air recirculation and cooling subsystem removes heat from the containment atmosphere during normal plant operations and accident conditions. The subsystem is independent of the Containment Spray and Safety Injection Systems. The system consists of four cooling units, an air mixing plenum, and the distributing ductwork and piping, all located inside of containment. Service Application for License Renewal 5.1IB-1 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4)

APPENDIX A - TECIINICAL INFORMATION

5. IIB - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM water (SRW) is circulated through the air cooling coils to remove heat. Engineered Safety Features Actuation Signal provides startup signals for accidents that require containment heat removal.

[ References 2 through 5]

The containment penetration room ventilation subsystem processes air passing through the hydrogen purge subsystem following a Design Basis Event while that subsystem is in operation. The penetration room exhaust fans draw air through the hydrogen purge lines and through a prefilter, a high efficiency particulate air (IIEPA) filter, and an activated charcoal filter to remove radioactive particulates and iodines before discharging the air to the environs. The containment penetration room ventilation subsystem can also collect and process containment penetration leakage during the post-accident period by drawing air from ae containment penetration rooms. No credit is taken for the latter method of post-accident dose reduction. [ References 2,4, and 6]

The containment iodine removal subsystem removes iodine from the containment following a loss-of-coolant accident. The subsystem consists of three banks of filters made up of maisture :eparators, HEPA filters, and activated charcoal filters. An electric-driven induced-draft fan is used to pull the containment atmosphere through the filters and to discha ge back into the containment. [ References 2 and 3]

The hydrogen purge subsystem is designed to control hydrogen concentrations inside containment below 4.0 % by volume should both hydrogen recombiners fai' to function properly. During power operations, the exhaust portion of this system is used to vent the containment to control containment pressure and airborne radioactivity levels. The containment atmosphere is drawn from containment through a moisture separator by the penetration room exhaust blowers. The purged etmosphere is carried to the containment penetration room ventilation subsystem's IIEPA and charcoal filters before being discharged to the environs. A separate blower is provided to supply purge replacement air to the containment through a separate penetration. This blower is not used during normal operations. Motor-operated valves (MOVs) are located in the supply and exhaust penetration piping on the outboard side of the containment penetration to provide containment isolation when necessary. [ References 4 and 6]

The containment purge subsystem supplies filtered air to the containment via a supply fan and penetration piping. One exhaust fan for each containment draws air from the containment through penetration piping and HEPA filters, and discharges the air into the respective main plant exhaust plenum and ultimately to the plant vent. Air-operated butterfly valves are located in the supply and exhaust penetration piping on cach side of the containment penetration to provide containment isolation when the system is in operation. During normal operation, the Unit 2 containment penetrations are sealed with blind flanges on the outboard side of the penetration. [ References 2,3, and 5] The outboard side butterfly valves for Unit I will be replaced with sealed blind flanges during normal operation following the next refueling outage.

The control element drive mechanism cooling subsystem draws air from the containment atmosphere and through the reactor head cooling shroud and into two cooling coils of the control element drive mechanism cooler, which is located on the missile shield above the reactor. From there,100% redundant fans discharge the cooled air upward and back into the containment atmosphere. Feur ducts connect the shroud to the cooler coil house. One pair of ducts directs air to one cooling coil and the other pair supplies air to the opposite coil. Component cooling water is pumped through the water-air coils. A Application for License Rer.ewal 5.1IB-2 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4) f

(

APPENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM motor operated damper, located between each fan and the cc,il house, prevents short-circuiting of air around the cooler when only one fan is operating. [ References 2,3, and 5]

System Interfacts The Primary Containment il&V System has an interface with the following systems and components:

[ References 2,7, and 8]

Containment Spray Actuation System; e

Containment Isolation Actuation System; e

Main Plant Vent; e

Radiation Monitoring System; Reactor Protective System; e

Safety injection Actuation System; Service Water System; and e

Component Cooling Water System.

e System Sconing Resuhs The Primary Containment H&V System is within the scope oflicense renewal based on 10 CFR 54.4(a).

The following intended functions of the Primary Containment li&V System were determined based on the requirements of {54.4(a)(1) and (2), in accordance with the CCNPP IPA Methodology, Section 4.1.1:

[ Reference 9, Table 1]

To control containment temperature and piessure; To provide containment atmosphere f'Itration and radiation control; To collect and process containment penetration leakage into the penetration rooms; e

To filter hydrogen purge air for radiation control following Design Basis Events; To measure pressure in the containment penetration rooms; To provide containment atmosphere pressure source to ESFAS instrumentation for protective actuation; To isolate the containment; To maintain electrical continuity and/or provide protection of the electrical system; e

To maintain the pressure boundary of the system (liquid and/or gas); and To provide seismic integrity and/or protection of safety-related components.

e The followir3 Primary Cantainment il&V System intended function was determined based on the requirements of Q54.4(a)(3): [ Reference 9, Table 1]

For environmental qualification (Q50.49)- To maintain functionality of electrical equipment as addressed by the Environmental Qualification Program, and to provioe information used to assess the plant and environs condition during and following an accident.

Application for License Renewal 5.11B-3 Calvert Cliffs Nuclear Power Plant

r ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM All components of the Primary Containment 11&V System that support the above functions are considered safety-related and Seismic Category I and a.e srbject to the applicable loading conditions identified in Calvert Cliff's UFSAR Section SA3.2 for Seismic Category I systems and equipment design. [ References I and 6]

QpsEt ing Exneriencs t

Over 20 years of operating experience has shown the H&V systems at CCNPP to be highly reliable in maintaining their passive functions. Some fasteners have become loose due to dynamic loading.

Vibration related aging concerns are minimized through system design and maintenance practices.

Vibration isolators, i.e., flexible collars or rubber boots, are installed to minirnize fan vibration being transferr:d to other equipment. [ Reference 1 Attachments 6] furthermore, fans are monitored for vibration whenever the fan belts are retensioned or replaced. [ Reference 10]

Corrosion has been discovered below the cooling coils in some coolers. These areas are routinely inspected to assess the corrosion rates and adequacy of the system pressure boundary. Other than the limited amount of degradation experienced due to vibration, wear, and corrosion, no other significant aginj concerns have been identified that could affect the ability of the Primary Containment il&V System components to pe. form their passive functions.

The CCNPP Containmeat Leakage Rata Testing Program has been inspected by the NRC on numerous occasions through routine inspections and during eviews of Technical Specification amendment requests. These inspections have not identified any aging-related concerns that need to be addressed in the AMR of Primary Containment H&V System components. Overall, the Containment Leakage Rate Testing Program has meintained the containment isolatian portions of the system within the requirements of 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

Baltimore Gas and Electric Company has requested and received Technical Specification amendments for revising the containment Type C testing schedule equired under 10 CFR Part 50, Appendix J; e.g., to adopt the performance-based requirements of Option B to Appendix J. During the reviews of these requests, significant analysis of past operating experience was performed for CCNPP and the industry as a whole. The NRC has indicated, based on their reviews of Type C performance history, that the wear-out portion of the component life has not been reached, and 3

may not be reached provided good maintenance practices continue to be followed. [ References !!

through 18] Additional information on operating experience is provided in the Group i discussion of aging management programs.

t Application for License Renewal 5.118-4 Calvert Cliffs Nuclear Power Plant

4 ATTACHMENT (4)

APITNDIX A - TECIINICAL INFORMATION 5.11H - PRIMARY CO*'~ \\INMMT IIEATING AND VENTILATION SYSTEM 5.11B.I.2 Component Level Scoping Based on the intended system functions listed above, the po: hns of the Primary Containment H&V System that are within the scope of license renewal include all efety-related components in the system j

(electrical, mechanical, and instrument), and their supports. Safety-related portions of the Primary Containment H&V System include the following: (References 1,3,4, and 5]

Subsystem Portion Within Scone Containment Air Recirculation and Cooling Cooling units, fans, and connecting ductwork up to and including the fusible dropout plates Containment Penetratic. Room Ventilation Entire subsystem Containment Iodine Removal Entire subsystem liydrogen Purge All components from the inboard containment isolation valve to the connection with the containment penetration exhaust subsystem Containment Purge Containment penetrations only Control Element Drive Mechanism Cooling None Pressurizer Compartment Cooling None Instrumentation and Controls Containment pressure instrumentation and intgrated leak rate testing containment penetrations The following 38 device types in the Primary Containment H&V System were designated e within the scope of license renewal because they have at least 1 intended function [ Reference 1, Tr.tiv 1-l]:

Device Tvoe Device Description Device Tvoe Device Descriotion CKV Check valve MB 480V motor COLL Coil MD 125/250VDC motor CV Control valve MOVOP Motor-operated valve operator DAMP Damper MOV Motor-operated valve DISC Disconnect switch / link NB 480V local control station DUCT Heating, ventilation and air PDI Pressure differential indicator conditioning duct PDIS Pressure differential indicator switch E!I Voltage / current device PI Pressure Indicator FAN Fan PIA Pressure Indicating alarm FL Filter PO Piston operator FT Flow transmitter PT Pressure transmitter FU Fuse RCMB llydrogen recombiner GD Gravity damper RY Relay HB Piping (Code HB)

SV Solenoid valve llS Handswitch TE Temperature element HV Hand valve TI Temperatuu indicetor HX lleat exchanger YX Power supply 11 Ammeter ZL Position indicating lamp JL Power lamp indicator ZS Position switch M

480V motor (feed from MCC)

Application fr License Renewal 5.1 l B-5 Calvert Cliffs Nuclear Power Plant a

4 ATTACHMENT (4)

. "" JDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM I

l Some components in the Primary Containment H&V System are common to many other plant systems and have been included in separate sections of the BGE LRA that addre.,s those components as commodities for the entire plant. These components include the following: [ Reference 1, Section 3.2]

Structural supports for piping, cables, and components are evaluated for the e:Tects of aging in the Comoonent Supports Commodity Evaluation in Section 3.1 of this application.

Electrical control and power cabling are evaluated for the effects of aging ia the Electricta Cabler Commodity Evaluation in Section 6.1 of this application. This commodity evaluation completely addresses the passive inter. ' function entitled " maintain electrical continuity and/or provide protection of the electrical s, J' for the Primary Containment H&V System.

Process and instrument tubing and tubing supports are evaluated for the effects of aging in the Instrument Line Commodity Evaluation in Section 6.4 of this application.

5.11B.I.3 Components Subject to AMR This section describes the components within the Primary Containment H&V System that are subject to an AMR. It begins with a listing of passive intended functions and then dispositions the device types as either only associated with ac*ive functions, subject to replacement, evaluated in other reports, evaluated in commodity reports, or remaining to be evaluated for aging management in this section.

Passive Intended Functions In acccrdance with CCNPP IPA Methodology Section 5.1, the following Primary Containment H&V System intended functions were determined to be passive. [ Reference 1, Table 3-1]

To isolate the containment; Maintain the pressure boundary of the system; Maintain electrical continuity and/or provide protection of the electrical system; and Provide seismic integrity and/or protection of safety-related components.

Device Tynes Subject to AMR Of the 38 device types within the scope oflicense renewal: [ Reference 1, Table 3-2; Reference 19]

21 Device types have only active functions and do not require AMR; coil, voltage /cmrent device, fuse, handswitch, ammeter, power lamp indicator,480V motor (feed from MCC),480V rnotor, 125/250VDC motor, motor-operated valve operator, pressure differential indicator, piessure indicator, pressure indicating alarm, piston operator, hydrogen recombiner, relay, temperature element, temperature indicator, power supply, position indicating lamp, position switch.

5 Devices types are evaluated in another section of this application.

'oisconnect switches / links and 480V local control stations, i.e., cabinets, panels, and enclosares, are evaluated for the effects of aging in the Eiectrical Commodities Evaluation in Section 6.2 of this application. The disconnect switches / links and 480V local control stations are the only device types in the system that have the pass.'ve intended function er. titled " provide seismic Application for License Renewal 5.11 B-6 Calvert Cliffs Nucl:ar Power Plant

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM integrity and/or protection of safety related components" Therefore, this commodity evaluation completely addresses that passive intended function for the Primary Containment H&V System.

Pressure differential inaicator switch, pressure transmitter, and Onw transmitter are evaluated for the effects of aging in the Instrument Lines Coramodity Evaluation in Section 6.4 of this application.

The remaining 12 device types listed in Table 5.11B-1 are subject to AMR and are included in this section. For AMR, some device types have a number of subgroups associated with them because of the diversity of materials used in theh fabrication or differences in the environments to which they are subjected.

Containment and system pressure boundary integrity are the only passive mtended functions associated with the Primary Containment H&V System not addressed by one of the commodity evaluations referred to above. Therefore, only the pressure-retaining function for the 12 device types listed in Table 5.11B-1 is considered in the AMR for the Primary Containment H&V System. Unless otherwise annotated, all components of each listed device type are subject to AMR and included in this section.

TABLE 5.11B-1

- PRIMARY CONTAINMENT H&V SYSTEM DEVICE TYPES REQUIRING AMR Check valve Gravity damper Control valve Piping (Code HB)

Damper Hand ValveW Duct Heat exchanger Fan MOV Filter Solenoid valve (1) Instrument line manual drain, equalization, and isolation valves in the Primary Containment H&V System that are subject to AMR are evaluated for the effects of aging in the Instrument Lines Commodity Evaluation in Section 6.4 of this application instrument line manual root valves are evaluated in this report. [ Reference 19, Attachment 3]

5.11B.2 Aging Management A list of potential age-related degradation mechanisms (ARDivis) identiDed for the Primary Containment H&V System comporents is given in Table 5.11B-2. The plausible ARDMs are identified in the Table by a check mark (/) in the appropriate device type column. A check mark indicates that the ARDM applies to at least one group for the device type listed. For efficiency in presenting the results of the evaluations in this section, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all components within that group. Exceptions are noted where appropriate. Table 5.11B-2 identifies the group in which each ARDM/ device type combination belongs. [ Reference 1, Table 4-2]

Application for License Renewal 5.11 B-7 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4)

' APPENDIX A - TECHNICAL INFORMATION

~

5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM TABLE 5. IIB-2 POTENTIAL AND PLAUSIBLE ARDMs FOR THE PRIMARY CONTAINMENT H&V SYSTEM Primary Costalement H&Y System Device Types Fotentist ARDM llB CKV CV DANP DUCT FAN FL GD HV

!!X MOV SV Cavitation Erosion Corrosion Fatigue Creep / Shrinkage Crevice Corrosion

/(2)

/(2)

/(2,5)

/(2)

Dynamic Loading

/(3)

Erosion Corrosion Fatigue l'ouling Galvanic Corrosion General Corrosion

/(2)

/(2)

/(2)

/(2) flydrogen Damage Intergranular Attack Irradiation Embrittlement Microbiologically-Induced

/(2)

Corrosion (MIC)

Oxidation Particulate Wear Erosion Pitting

/(2)

/(2)

/(2.5)

/(2)

Radiation Damage

/(4)

Elastomer Degradation

/(4)

/(4)

/(4)

Saline Water Atts -

Selective Leaching Stress Corrosion Cracking Stress Relaxation Thermal Damage Thermal Embrittlement Wear

/(1)

/(1)

/(4)

/(4)

/(1)

/(4)

/(1)

/ - indicates that the ARDM is plausible for component (s) within the Note: Not every component within the device types listed here may be Device Type susceptible to a given ARDM. This is because components within a device type are not always fabricated from the same materials or subjected to the same

(#) - Indicates the Group in which this device type /ARDM combination environments. Exceptions for each device type will be indicated in the aging is eva!uated management section for each ARDM discussed in this report.

Application for License Renewal 5.11B-8 Calvert Cliffs Nuclear Power Plant

~

r ATTACHMENT m APPENDIX A - TECilNICAL INFORMATION 5.11H - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM The following groups have been selected for the Primary Containment H&V System:

Group 1 -

Includes wear for check valves, control valves, hand valves and MOVs; Group 2 -

includes crevice corrosion, general corrosion, MIC, and pitting for all components exposed to moisture; Group 3 -

Includes dynamic loading for fans; Group 4 -

Includes radiation damage, elastomer degradation, and wear for non-metallic subcomponent parts; and Group 5 -

Includes crevice corrosion and pitting for heat exchanger cooling coils.

The following is a discussion cf the aging management demonstration process for each group identified above, it is presented by group and includes a discussion on materials and environmer.t, aging mechanism effects, methods to manage aging, aging management program (s), and aging management demonstration.

Group 1 (wear for check valves, control valves, hand valves, and MOVs) - Materials and Environment n,-

The Primary Containment il&V System contains several types of valves that form part of the containment pressure boundary and whose disks / balls and seats are subject to wear. Check valves, control valves, hand valves, and MOVs provide containment pressure boundary for the containment penetrations in the containment purge subsystem (Unit I only, blank flanges are installed ir. Unit 2),

hydrogen purge subsystem, and integrated leak rate testing lines. Hand valves also provide containment pressure boundary (some are required to be open and some closed) for the containment pressure instrumentation lines used for ESFAS. Solenoid valves are also used for containment isolation of the containment pressure instrumentation lines, but are discussed in Group 2 below.

[ Reference 1, Attachments 3 and 5]

The disks / balls and seats of the check valves, hand valves and MOVs are constructed of alloy steel, sta, tiess steel, or stellited carbon steel. The control valves have a ('isk constructed of carbon steel and a seat of ethylene propylene. [ Reference 1, Attachment 4]

The interna! environment for the Primary Containment H&V System containment pressure boundary valves is containment air. The ambient air pressure variation is limited to -1.0 to +1.8 psig during normal plant operation. The maximum design service conditions regarding relative humidity and ambient air temperature for normal plant operation are 70% and 120*F, respectively. [ Reference 20,, Table 1]

Group 1 (wear for check valves, control valves, hand valves, and MOVs) - Aging Mechanism Effects Wear results from relative motion between two surfaces (adhesive wear), from the influence of hard, abrasive particles (abrasive wear) or fluid stream (erosion), and from small, vibratory or sliding motions under the influence of a corrosive environment (fretting). Motions may be linear, circular, or vibratory Application for License Renewal 5.11 B-9 Calvert Cliffs Nuclear Power Plant l

l

A'ITACIIMFNT (4)

APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM in inert or corrosive environments. Wear rates may accelerate as expanded clearances result in higher contact stresses. [ Reference 1, Attachment 7]

The internals of Group i valves subject to wear are relied on to maintain containment pressure boundary integrity. These valves are periodically cycled during testing and outages and the valve internals and seating surfaces may experience wear. Wear is considered plausible for the disk / ball and seat of check valves, control valves, hand valves, and MOVs because they may experience cyclic relative motion at the tight fitting surfaces. Movement of the disk against the seat can result in a gradual loss of material, which could result in a small amount of leakage. Ifleft unmanaged, wear could lead to a loss of pressure be ndary integrity. [ Reference 1, Attachment 6]

Calvert Cliffs has experienced some wear of the containment purge supply and exhaust containment isolation valves (control valves). Check valves have also experienced pressure boundary failures with several valves failing back leakage tests, including those tests performed in response to Generic Letter 88-14. However, the root cause of these failures is due to a combination of wear and misapplication of the valve for its intended function. [ Reference 1, Attachments 4 and 6; Reference 21]

Group 1 (wear for check valves, control vahes, hand valves, and MOVs) - Methods to Manage Aging Mitigation: Since the wear of valve disk / balls and seats is due to valve operation, decreased operation of the valves would slow the degradation of the valves seating surfaces. This is not a feasible mitigation technique because it would place unnecessary restrictions on plant operation. The restrictions are unaccessary because limited leakage through the valves will not significantly impact the intended function. Furthermore, the discovery methods discussed below are deemed adequate for verifying that significant degradation is not occurring. [ Reference 1, Attachment 6]

Discover Wear for valve disks / balls and seats can be detected by performing visual inspections or through leak rate testing. Since wear occurs gradually over time, periodic testing can be used to discever minor leakage of the valve seating surfaces so that corrective actions can be taken prior to the loss of the intended function. [ Reference 1 Attachment 6]

Group 1 (wear for check valves, control valves, hand valves, and MOVs) - Aging Management Program (s)

Mitigation: There are no feasible methods of mitigating wear of the valve disks / balls and seats; therefore, there are no programs credited with mitigating the aging effects due to this ARDM.

Discoverv: The containment isolation valves are subject to local leak rate testing under the CCNPP Containment Leakage Rate Testing Program. Hand valves are not subject to periodic leakage testing.

They will be included in the scope of an Age-Related Degradation Inspection (ARDI) Program.

Containment Leakace Rate Testing Program The valves performing the containment pressure boundary function are subject to local leak rate testing under the CCNPP Cantainment Leakage Rate Testing Program as required by 10 CFR Part 50, Appendix J. This testing is implemented through CCNPP Surveillance Test Procedures in accordance Application for License Renewal 5.1 I B-10 Calvert Cliffs Nuclear Power Plant

e NITACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.1IB - PRIMARY CONTAINMENT HEATING AhD VENTILATION SYSTEM with the plant Technical Specifications and CCNPP Administrative Procedure EN-4105," Containment Leekage Rate Testing Program." [ Reference 1, Attachment 6; References 22 through 26]

The CCNPP Containment Leakage Rate Testing Program was estaMished to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water Cooled Power React-rs," Option B "Performaace-Based Requirements." It follows the guidance provided in Regulatory Guide 1.163,"Performimee-Based Containment Leak-Test Progri.m," September 1985. Appendix J specifies containment leakage testing requirements, including the types of tests required, frequency of testing, test methods, test pressures, acceptance criteria, and reporting requirements. Containment leakage testing requirements include performance of Integrated Leakage Rate Tests, also known as Type A tests, and local leak rate tests (LLRTs), also known as Type B and C tests. Type A tests measure the overall leakage rate of the containment. Type B tests are intended to detect leakage paths and measure leakage for certain containment penetrations (e.g., airlocks, flanges, and electrical penetrations). Type C tests are intended to measure containment isolation valve leakage rates. (References 22,26,27, and 28]

The CCNPP Containment Leakage Rate Testing Program is based on 10 CFR Part 50, Appendix J, Option B, requirements and implements the requirements in CCNPP Technical Specifications 3.6.1.2, 4.6.1.2, and 6.5.6. The secpe of the program includes Type B and C testing of containment pen:trations.

[ Reference 1, Attachment 6; References 22 through 26] Per References 22 through 26, currently the LLRT includes the following procedural steps:

Leak rate monitoring test equipment is connected to the appropriate test point.

Test volume is pressurized to the LLRT pressme, which is conservative with respect to the 10 CFR Part 50, Appendix J, Option B, test pressure requirements.

Leak rate, pressure, and temperature are monitored at the frequency specified by the LLRT procedure and the results are recorded.

The maximum indicated !cak rate is compared against administrative limits that are more restrictive than the maximum allowable leakage limits, "As found' leakage equal to or greater than the administrative limit, but less than the maximum e

allowable limit, is ev>

' 9 octermine it further testing is reqeired and/or if corrective maintenance is to be n

• For "as found" lea'

.ceeds the maximum allowable limit, plant personnel deiermine if Technical Specification Limiting Condition for Operation 3.6,1.2.b has been exceeded. Technical Specification 3.6.1.2.b contains the maximum allowable combined leakage for all penetrations and valves subject to the Type B and C tests. Corrective action is taken as required to restore the leakage rates to within the appropriate acceptance criteria.

if any maintenance is required on a containment isolation valve that changes the closing e

characteristic of the valve, an "as left" test must be performed on the penetration to ensure leakage rates are acceptable.

The CCNPP Containment Leakage Rate Testing Program has been inspected by the NRC on numerous occasions through routine inspections and during reviews of Technical Specification amendment requests. Routine inspections at the site included procedure reviews, leakage test witnessing, test Application for License Renewal 5.118-11 Calvert Cliffs Nuclear Power Plant

e ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM reviews, and results evaluation of both integrated leakage rate tests and LLRTs. Inspectors noted when individual containment isolation valves failed their leakage tests and reviewed the repair and resetting actions taken by BGE. With some specific exceptions, the inspections typically noted acceptable conditions. No aging-related deficiencies were identified. [ References 11 through 14}

Baltimore Gas and Electric Company has requested, and received Technical Specification amendments for revising the containment Type C testing schedule required under 10 CFR Part 50, Appendix J. The requests were initiated to accommodate extending the fuel cycle to 24 months, and to recognize the added Option B under Appendix J. Currently CCNPP follows the schedule of Option B, which is a performance based scheduling process. During the reviews of these requests, significant analysis of past operating experience was performed for CCNPP and the industry as a whole. The NRC has indicated, based on their reviews of Type C performance history, that the wear-out portion of the component life has not been reached, and may not be reached provided good maintenance practices continue to be followed. Furthermore, reviews of site-specific data indicate that the leakage rate data at the end of the CCNPP Unit 1 operating cycles falls within a typical range. [ References 15 through 18]

These aviews demonstrate that CCNPP has normal and acceptable operating experience with respect to corapc.unt aging of components relied on the containment isolation. The corrective actions taken as part of the Containment Leakage Rate Testing Program will ensure that the containment isolation check valves, control valves, and MOVs remain capable of perfon.iing their containmer.. pressure boundary integrity function under all current licensing basis (CLB) conditions.

ARDI Program The Group i hand valves will be included within a new plant program to accomplish the needed inspections for wear. This program is considered an ARDI Program as defined in the CCNPP IPA Methodology presented in Section 2.0.

The elements of the ARDI program wia include:

Determination of the examination sample size based on plausible aging effects; Identification of inspection locations in the system / component based on plausible aging effects e

and consequences ofloss of compoi.ent intended function; Determination of examination techniques (including acceptance criteria) that would be effective, considering the aging effects for which the component is examined; Methods for interpretation of examination results; e

Methods for resolution of unacceptable examination findings, including consideration of all e

design loadings required by the CLB, and specificetion of required corrective actions; and Evaluation of the need for follow-up examinations to monitor the progression of any age-related degradation.

The corrective actions will be taken in accordance with the CCNPP Corrective Actions Program and will ensure that the components will remain capable of performing the system pressure boundary integrity function under all CLB conditions.

Application for License Renewal 5.1IB-12 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (O APPENDIX A - TECHNICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM Group 1 (wear for check valves, control valves, hand valves, and MOVs)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to wear of check valves, control valves, hand valves and MC /s for the Primary Containment Il&V System.

The valve internals maintain containment pressure boundary and their integrity must be maintained under all CLB conditions, Wear is plausible for valve disks / seats and results in material loss which, iflef1 unmanaged, could e

lead to leakage.

The containment isolation valves are subject to local leak rate testing in accordance with the CCNPP Containment Leakage Rate Testing Program.

Leak rate testing activities will provide reasonable assurance that significant leakage that could be the result of wear of the seating surfaces is discovered and appropriate corrective actions taken.

The hand valves are not subject to periodic leak rate testing so they will be included within a new ARTA program to accomplish the needed inspections.

Therefore, there is reasonable assurance tliat the effects of wear for Primary Containment H&V System valves will be managed in such a way as to maintain the components' pressure boundary integrity, consistent ivith the CLB, during the period of extended operation.

Group 2 (crevice corrosion, general corrosion, MIC, and pitting for all components exposed a moisture)- Materials and Environment Group 2 is comprised of components that are povntially exposed to moist air and condew. tion. These include the piping, hand valves, and MOVs in the hydrogen purge subsystem exhaust path. This portion of the sys:em is po'entially exposed to warm humid air from containment. The moisture in the air could condense upon contact with the cooler pipe and valves, partularly outside of containment. Also included in Group 2 ere the containment coolers that may be exposed to condensed moisture from the cooling coils. If the drains become plugged, there may also be standing water in t!". drain pan that could spill over onto other sections of the eauipment base. Surfaces that are painted or galvanized are protected from corrosion by keeping tne component surfaces from being exposed to the moisture.

However, where the coating is damaged, corrosion may take place. [ Reference 1, Attachment 6s and ]

The subject piping, fittings, flanges, and welds are all cor.structed of carbon steel. The body / bonnet of the hand valves is constructed of carbon steel and the stems, disks, and seats are either alloy steel, stellited carbon steel, or stainless steel. The body / bonnet of the MOVs are constructed of carbon steel, the stems of stainless stee!, the wedge / disk of stellited carbon steel or stainless steel, and the ser.t of stellite stainless steel or ethylene propylene. All of the studs and nuts are extemal and are not exposed to the moisture. The valve disks / seats of he containment isolation valves are relied upon for containment pressure boundary. The disk seat of the flow control MOV does not serve the pressure boundary function. [ Reference 1, Attachments 4,5, and 6)

Application for License Renewal 5.11B-13 Calvert Cliffs Nuclear Power Plant m

l NITACHMENT (4)

AlfENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT llEATING AND VENTIMTION SYSTEM The containment cooler housing is constructed of carbon steel. The boot between the cooler and the fan is constructed of rubber.

The cooling coils are addressed below in Group 5.

[ Reference 1, Attachments 4,5, and 6]

Group 2 (crevice corrosion, general corrosion, MIC, and pitting for all components exposed to moisture)- Aging Mechanism Effects Crevice cortosion is intense, localized corrosion within crevices or shielded areas, it is associated with a small volume of stagnant solutica caused by holes, gasket surfaces, lapjoints, crevices under bolt heads, surface deposits, designed crevices sbr attaching thermal sleeves to safe-ends, and integral weld backing rings or back-up bars. The crevice must be wide enough to permit liquid entry and narrow enough to maintain stagnant conditions, typically a few thousandths of an inch or less. Crevice corrosion is closely related to pitting corrosion and can initiate pits in many cases, as well as lead to stress corrosion cracking. [ Reference 1,

Attachment:

6 and 7]

General corrosion is the thinning (wastage) of a metal by chemical attack (dissolution) at the surface of the metal by an aggressive environment. General cormsion requires an aggressive environment and materials susceptible to that environment. Wastage u n" a concern for austenitic stainless steel alloys.

The consequences of the damage are loss of load carrying cross-sectional area.

[ Reference 1, Attachments 6 and 7]

Microbiologically induced corrosion is accelera corrosion of materials resulting from surface microbiological activity. Sulfate reducing bacteria, sulfur oxidizers, and iron oxidizing bacteria are most commonly associated with corrosion effects. Microbiologically-induced corrosion often results in pitting followed by excessive deposition of corrosion products. Stagnant or low flow areas are most susceptible.

Essentially all sysams using untreated water and most commonly used materials are susceptible.

Conseqt.ences range from leakage to excessive differential pressure and flow blockage.

Microbiologically-induced corrosion is generally observed in SRW applications utilizing raw, untreated water. [ Reference 1, Attachments 6 and 7]

Pitting is another form of localized attack with greater corrosion rates at some locations than at others.

Pitting can be very insidious and destructive, with sudden failures in high pressure applications (especially in tubes) occurring by perforation. Thh foi of corrosion essentially produces holes of varying depth-to-diameter ratios in the steel. Deep pitting is more common with passive metals, such as austenitic stainless steels, than with non passive metals. Pits aie generally elongated in the direction of gravity. In many cases, crosion corrosion, fretting corrosion, and crevice corrosion can also lead to pitting. [ Reference 1, Attachments 6 and 7]

For Group 2 components, long-term exposure to the moist environment may result in localized and/or general area material loss of the internal surfaces of the components and, if unmanaged, could eventually result in loss of the pressure-retaining capability under CLB design loading conditions. The areas where there are crevices and/or stagnant conditions, e.g., body / bonnet joint, stem to bonnet / packing area, valve seat area, low points in the pipe, etc., are the locations most susceptible to these corrosion mechanisms.

Crevice corrosion, general corrosion, and pitting are plausible for the subcomponents constructed of carbon steel or alloy steel. Crevice corrosion, general corrosion, and pitting are not plausible for subcomponents constructed of stainless steel or for the stellited surfaces of carbon steel components due e

Application for License Renewal 5.118-14 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (4)

APPENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM to the inherent corrosion resistance of the meterials and the non-aggiessiveness of the environment. The containment cooler housing is also subject to MIC due to the potential existence of standing stagnant water. Since the valves are required to maintain pressure boundary while in the closed position, degradation of the internal surfaces of all subcomponents required for the pressure-retaining function must be managed. [ Reference 1, Attachments 4,5, and 6)

Group 2 (crevice corrosion, general corrosion, MIC, and pitting for all components exposed to moisture). Methods to Manage Aging Mitigation: Since there is no design feature to control the humidity of the air the Group 2 components are exposed to, the only feasible method of mitigating the effects of corrosion is to replace the components with components constructed of more corrosion resistant materials.

However, this mitigation technique is not necessary because the discovery techniques discussed below are deemed adequate to manage aging due to crevice corrosion, general corrosion, MIC, and pitting. [ Reference 1, ]

Discoverv: The effects of corrosion (crevice corrosion, general corrosion, MIC, and pitting) on Group 2 components can be discovered and monitored through con-destructive examination techniques such as visual inspections. Representative samples at susceptible locations can be used to assess the need for additional inspections at less susceptible locations. Periodic preventive maintenance would lead to the discovery of corrosion of components that are readily observable dering the activity. If corrosion is occurring on valve seating surfa.es, the degradation can be detected through pressure tests of the valves in the closed position. Corrosion would cause loss of material that can lead to valve leakage.

[ Reference 1, Attachment 8]

Group 2 (crevice corrosion, general corrosion, MIC, and pitting for all components exposed to moisture)- Aging Management Program (s)

Mitigatic : Since there are no mitigation techniques deemed necessary at this time, there are no a

mitigation programs credited for managing corrosion of Group 2 components.

Discoverv: For Group 2 components, crevice corrosion, general corrosion, MIC, and pitting can be readily detected through non-destructive examination techniques. Periodic preventive maintenance will provide assurance that the effects of corrosion are not threatening the pressure-retai.dng capability of the heat exchangers. The remaining Group 2 components will be included in the scope of an ARDI Program. In addition, the containment isolation valves are periodically leak tested. This will provide an early indication of degradation of the valve seating surfaces. [Refeience 1, Attachment 8]

Freventive Maintenance Program The CCNPP Preventive Maintenance Program has been established to maintain plant equipment, structures, systems, and components in a reliable condition for normal operation and emergency use, minimization of equipment failure, and extension of equipment and plant life. The program covers all preventive maintenance activities for nuclear power plant structures and equipment within the plant, including the Primary Containment H&V System components within the scope of license renewal.

[ Reference 29] Guidelines drawn from industry experience and utility best practices were used in the development and enhancement of this nrogram.

Application for L.icense Renewal 5.1IB-15 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (4)

APPENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM e

Calvert Cliffs currently has preventive maintenance activities scheduled for every two years (i.e., every refueling outage) to perform inspections and cleaning of the containment air coolers. These activities

]

specincally include a complete insxction of the coolers for cleanliness, corrosion, and leaks.

[ References 30 and 31] This inspection would discover signincant corrosion of the containment air cooler housings if it were occurring. Corrective actions are taken in accordance with the CCNPP Corrective Actions Program.

The plant maintenance program has numerous levels of management review, all the way down to the specific implementation procedures. For example, there are specine responsibilities assigned to BGE personnel for evaluating and upgrading the Preventive Maintenance Program. [ Reference 29] The Preventive Maintenance Program has also undergone evaluation by the NRC as part of their routine licensee assessment activities. These assessments and controls provide reasonable assurance that the Preventive Maintenance Program will continue to be an effective method of managing the effects of cotrosion for the heat exchangers.

ARDi Program The Group 2 piping, hand valves, and MOVs will be included within a new plant program designed to provide the needed inspections for corrosion. This program is considered an ARDI Program as defined in the CCNPP IPA Methodology presented in Section 2.0.

The corrective actions will be taken in accordance with the CCNPP Corrective Actions Program and will ensure that the components will remain capable of performing the system pressure boundary integrity function under all CLB conditions.

Refer to tl: Group 1 discrssion on aging nanagement programs for a detailed discussion of the ARDI Program.

Containment Leakage Rate Testing Program in addition to the ARDI Program, the contaimant isolation valves are subject to periodic pressure testing for valve leakage at the seating surface. These components are subject to local leak rate testing under the CCNPP Surveillance Test Procedures in accordance with 10 CFR Part 50, Appendix J.

[ Reference 1, Attachment 8; References 23 and 24] Continued local leak rate testing on a periodic basis will assure acceptable leak tightness of the seating surfaces of these valves and will also ensure that any leakage remains within the guidelines of the Technical Specifications.

The LLRT is part of the overall CCNPP Containment Leakage Rate Testing Program, which is implemented through Surveillance Test Procedures. The CCNPP Containment Leakage Rate Testing Program is discussed in detail above for Group 1.

The corrective actions taken as part of the Containment Leakage Rate Testing Program will ensure that the containment isolation valves are capable of performing their containment pressure boundary integrity function under all CLB conditions.

Application for License Renewal 5.11 B-16 Calvert Cliffs Nuclear Power Plant

A'ITACHMENT (4)

APPENDIX A - TECilNICAL INFORMATION 5.118 - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM Group 2 (crevice corrosion, general corrosion, MIC, and pitting for all components exposed to moisture)- Demonstration of Aging Management Based on the factors presented above, the following conclusions can be reached with respect to crevice corrosion, general corrc,sion, MIC, and pitting of Group 2 components:

The Group 2 components provide the system pressure boundary and their integrity must be maintained under all CLB conditions, Crevice corrosion, general corrosion, MIC, and pitting are plausible for the components, and e

result in material loss which, ifleft unmanaged, can lead to loss of pressure boundary integrity.

Cleaning and inspection activities performed in accordance with the Preventive Maintenance Program will provide reasonable assurance that the effects of corrosion are discovered prior to threatening the pressure retention capability of the containment air coolers, inspections will be performed, and appropriate corrective action will be taken if significant corrosion is discovered.

To provide the assurance needed to conclude that the effects of corrosion are not threatening the pressure retention capability of the piping, hand valves, and MOVs, they will be included in the scope of an ARDI Program. Inspections will be performed, and appropriate corrective action will be taken if significant corrosion is discovered.

in addition to the ARDI Program, the containment isolation valves are subject to periodic e

pressure testing for valve leakage as part of the CCNPP Containment Leakage Rate Testing Program. Pressure testing for valve leakage would provide an early indication of degradation of the valve seating surfaces.

Therefore, there is reasonable assurance that the effects of crevice cor osion, general corrosion, MIC, and pitting on Group 2 components will be managed in such a way as to maintain the compomnts' pressure boundary integrity, consistent with the CLB, during the period of extended operation.

Group 3 (dynamic loading for tans)- Materials and Environment Group 3 is comprised of fans because dynamic loading is a concern for the fasteners. Nonnat bearing wear and dirt buildup can cause imbalancec, in the rotating parts of the fans, thereby inducing vibration.

Flexible collars are installed on the fans to provide dynamic isolation for adjacent components, which minimizes the dynamic loading for those components. The fan casings and fasteners are constructed of carbon steel. The fans are located indoors and have an internal environment of air with minimal humidity / moisture. (Reference 1, Attachments 3,4, and 7)

Group 3 (dynamic loading for fans)- Aging Mecha. ism Effects y

Dynamic loading (vibration) is created in blowers by rotating parts with imbalances due to dirt buildup and normal bearing wear. Tnere is a history ofloosened mechanical fasteners due to vibration in fans at CCNPP. This mechanism is plausible for the fans, but is not considered plausible for adjacent heating, ventilation and air conditioning equipment due to the dynamic isolation provided by flexible collars. If dynamic loading was left unmanaged, it could result in the loss of pressure boundary integrity of the Group 3 components under CLB design loading conditions. [ Reference 1, Attachments 5 and 6]

Application for License Renewal 5.11B 17 Calvert Cliffs Nuclear Power Plant i

ATTACllMFNT (4)

APPENDIX A - TECIINICAL INFORMATION 5.llB - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM Group 3 (dynamic loading for fans)- Methods to Manage Aging Mitigation: Dynamic loading can be mitigated by minimizing the mechanical loading due to vibration.

The system is designed to rninimize vibration by using equipment support isolators and equipment-to-duct isolators, such as Dexible collars and boots. Visual inspections during routine walkdowns weuld provide for detection of vibration so that corrective actions could be taken to minimize vibration and, thereby, mitigate the effects of dynamic loading. [ Reference 1, Attachment 8]

Discoyny: The effects of dynamic loading, e.g., loosened fasteners, can be detected through visual insocctions. Periodic visual inspections and system walkdowns would provide for detection of the efiects of dynamic loading, as well as vibration problems, which can cause this ARDM to occur.

[ Reference 1, Attachment 8]

s Group 3 (dynamic loading for fans)- Aging Management Program (s)

Mitigation: The CCNPP Structure and System Walkdown Program provides for periodic visual inspections of the external surfaces of Primary Containment H&V System components. During these walkdowns, any vibration problems would be detected so that corrective actions could be taken to minimize the vibration. [ Reference 1, Attachment 8] Refer to the discussions below in Discovery for a description of the CCNPP Structure and System Walkdown Program.

Discoverv: The effects of dynamic loading can be detected thrc'igh visual inspections and walkdowns.

Loosened fasteners may be detected by visual observation or by hearing unusual noise and vibration. For the equipment located outside containment, routine system walkdowas are adequate for detection of the effects of dynamic loading.

For the equipment located inside containment, routine preventive maintenance activities are adequate for detection of the effects of dynamic loading.

System Walkdowns Visual inspections are performed on system fans located outside containment as part of the routine system walkdowns in accordance with CCNPP Administrative Procedure MN-1-319, " Structure and System Walkdowns." MN-1-319 provides for discovery of the effects of dynamic loading, as well as abnormal or excessive vibration, which can cause this ARDM to occur. This procedure requires routine system walkdowns that include visual inspections, reporting the walkdown results, and initiating corrective action. [ Reference 32]

Under this program, BGE personnel, with assigned responsibility for specific structures and systems, perform periodic walkdowns. Walkdowns may also be performed as required for reasons such as material condition assessments; system reviews before, during, and afler outages; start-up reviews (i.e.,when a system is re-energized or placed in service); and as required for plant modifications.

Inspection items typically related to aging management include identifying unusual noises and identifying system and equipment stress or abuse, such as excessive vibrations. bent or broken component supports, loosened fasteners, etc. [ Reference 32, Section 5.1]

One of the objectives of the program is to assess the condition of the CCNPP structures, systems, and components such that any degraded condition will be identified, documented, and corrective actions taken before the degradation proceeds to failure of any structure, system, and component to perform its Application for License Renewal 5.11 B-18 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (4)

APPENDIX A - TECIINICAL INFORh1ATION 5.11B - PRIh1ARY CONTAINh1ENT IIEATING AND VENTILATION SYSTEh!

intended functin Conditions adverse to quality are documented and resolved by the Calvert Cliffs Corrective Actions Program. (Reference 32, Sectiens 5.1.C,5.2.A.1, and 5.2.A.5]

]

The program provides guidance for specinc types of degradation or conditions to inspect for when performing the walkdowns. General inspection items related to aging management include the following: (Reference 32, Section 5.2 and Attachments 1 through 13]

items related to specific ARDMs such as corrosion; e

Effects that may have been caused by ARDMs such as damaged supports, concrete degradation, e

anchor bolt degradation, or leakage of fluids; and Conditions that could aliow progression of ARDMs such as degraded protective coatings, leakage e

of fluids, excessive vibration, or inadequate support of components (e.g., missing, detached, or loose fasteners and clamps).

This program promotes familiarity of the systems by the responsible personnel and provides extended attention to plant material condition beyond that afforded by Operations and Maintenance alone. The program has been improved over time, based on past experience, to provide guidance on speciOc activities to be included in the scope of the walkdowns.

Preventive Maintenance Calvert Cliffs currently has preventive maintenance activities that help keep the subsystems operating reliably. These routine activities will allow for early detection of vibration problems for equipment located inside containment that can then be fixed prior to loss of fastener tightness, if dynamic loading does affect the fans, it will also be discovered during the performance of those activities. Preventive maintenance tasks are currently in place for performing inspections of the containment air coolers every two years (i.e.,every refueling outage)

These activities specifically include an inspection of the fasteners to ensure they are installed and not loose. [ References 30 and 31] Preventive maintenance tasks are also in place for lubricating the containment iodine removal fan motor every two years (i.e., each refueling outage). This activity typically includes running the fan after lubrication to verify operability. [ References 33 and 34) Refer to the discussion above in Group 2 under Aging Management Programs for a detailed discussion of the Preventive Maintenance Program.

The corrective actions required as a result of the system walkdowns and preventive maintenance activities will be taken in accordance with the CCNPP Corrective Actions Program, and will ensure that the fans will remain capable of performing the system pressure boundary integrity function under all CLB conditions.

Group 3 (dynamic loading for fans)- Demonstration of Aging 51anagement Based on the factors presented above, the following conclusions can be reached with respect to dynamic loading for fans:

Primary Containment H&V System fan housing provide a system pressure boundary function and the integrity of their fasteners must be maintained under CLB design conditions.

Application for License Renewal 5.1IB-19 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (f)

APPENDIX A - TECllhlCAL INFORMATION 5.118 - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM Dynamic loading is a plausible ARDM for the fans due to potentially excessive vibration e

resulting from fan operation. Dynamic loading is considered plausible only for the fans due to the use of vibration isolators for each of the fans, which prevents excessive vibration from being

]

transmitted to other components.

g If len unmanaged, dynamic loading can result in loosened fasteners, which could lead to loss of pressure boundary integrity.

I Existing visual inspections provide reasonable assurance that signs of loosened fasteners, as well e

as such conditions as unusual noises or vibration on fans located outside containment, would be detected as part of routine system walkdowns E

Existing activities perfo med as part of the Preventive Maintenance Program for equipment e

located inside containment provide reasonable assurance that the effects of dynamic loading will be detected. These activities include specific inspections of fasteners for the containment air I

cooler fans.

Therefore, there is reasonable assurance that the effects of dynamic loading for fans will be managed in

'5 such a way as to maintain the components' pressure boundary integrity, consistent with the CLB, during the period of extended operation Group 4 (radiation damage, elastomer degradation, and wear for non-metallic subcomponent parts)- Materials and Environment The Primary Containment Fl&V System galvanized carbon steel ducting was installed with Dexible collars in connections between fans and ducts or casings to prevent ex;essive movement of long ducts.

These flexible collars are made of clastomers and are installed with sufficient slack to prevent transmission of v% ration. Collars are secured to fans and ducts with galvanized steel bars fastened with bolts for an ai.-tight construction. There are no flexible collars located inside containment that are within the scope oflicense renewal. [ Reference 1, Attachment 4; Reference 35]

The penetration room exhaust dampers are required to maintain system pressure boundary while in the closed positica, and are constructed with compressible seals to provide leak tightness. These seals are constructed of neoprene sponge material, which is an elastomer. The dampers are located outside of containment where exposure to low radiation levels is not sufficient to cause degradation of the material.

[ Reference 1, Attachment 4s and 6s]

Each containment air cooler has a rubber boot installed between it and the fan to prevent the transfer of vibration to the cooler. Normal bearing wear and dirt buildup cause imbalances in the rotating parts of the fans, thereby inducing vibration. The containment air coolers and associated rubber boots are exposed to radiation because they are located inside containment. [ Reference 1, Attachment 6s]

The design service sonditions for ducting and equipment located inside containment are discussed above in the Materials and Environment section of Group 1. The penetration room exhaust dampers have an internal environment of air from the penetration rooms or con *ainment. The maximum environmental service conditions regarding relative humidity and ambient air temperature for the penetration rooms during normal plant operation are 70% and 140 F, respectively. [ Reference 20, Attachment 1, Table 1]

Application for License Renewal 5.11 B-20 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (4)

I APPENDIX A TECilNICAL INFORhfATION 5.11H - PRIMARY CONTAINMENT llEATING AND VENTILATION SYSTEM l

Grcup 4 (radiation damage, elastomer degradation, and wear for non metallie subcomponent parts). Aging Mechanism Effects Radiation can cause ionization and excitation of atoms in organic materials, which leads to damage through chemical reaction of the excited ions and/or free radicals. For clastomers, exposure to radiation r

vaa result in degradation of material properties, such as tensile strength, elongation, and compressibility.

hiaterial susceptibility is dependent upon the strength and type of the radiation Geld, and upon the specinc material composition. hiost materials exhibit a threshold level below wSich signi0 cant degradation of t echanical properti,;s of the waterial does not occur. [ Reference 5, Attachment 7; Reference 36]

An clastomer is a material that can be stretched to signincantly r* ater than original length and, upon release of the stress, will re; urn with force to approximately its origmal length. When an elastomer ages, there are :hree mechanisms primarily involved:

Scission the process of breaking moleculst bonds, typically due to ozone attack, UV light, or radiation; Crosslinki; g the process of creating molecular bonds between adjacent long-chain molecules, e

ypically due to oxygen attack, heat or curing; and

' mpound ingredient evaporation, leaching, mutation, etc.

e Natural aging tests indicate that where there is a signi0 cant property change in an clastomer, it appears that it occurs wnhin the Dist Ove to ten years ahu initial formulation / curing. Elastomers generally harden as they age, making scaling more dif0ct,lt. [ Reference 1, Attachment 7s]

Wear results from relative motion between two surfaces (adhesive wear), from the in0uence of hard, abrasive particles (abrasive weer), or fluid stream (erosion), and from small, vibratory or sliding motions under the inauence of a corrosive environment (fretting). hiotions m1y be linear, circuler, or vibratory in ir,ert or corrosive environments. Fretting is a wear phenomenon that occurs between tight Stting surfaces subjected to a cyclic, relative motion of extreme 4 small amplitude. Common sites for fretting are in joints that are bolted, keyed, pinned, press 6t, or riveted. [ Reference I, Attachment 7s]

Elastomer degradation and wear are plausib'e for the Dexible collars in the duct, and rubber boots on tne heat exchangers, since the clastomers will degrade due to the relatiu motion caused by vibrating equipment, pressure variations and turbulence, and exposure to moderate heat, oxygen, and czone.

These stressors could result in eventual tearing of the colla s and boot. Elastomer degradation and wear are plausible for the damper reals because the nenr..ne wiil degrade /ue to. elative motion between the blade and sleeve during damper operatior...u, xposure to moderate heat, oxygen, and ozone. These stressors will result in eventual breakdown of the seal. Radiation damage is plausible for the rubber boots on the containment air coolers because they are locate.1 inside containment and are exposed to radiation._ Radiation damage is not plausible for the duct Oexible collars or damper seals because they are located outside of contamment where the radiation levels are relatively low.

[ Reference 1, s and 7s] If len unmanaged, these ARDhis could eventually resu't in the loss of pressure boundary integrity of the duct Dexible collars, damper seals, and heat exchanger rubber boots under CLB design loriing conditions 3pplication for License Renewal 5.11B 21 Caivert Cliffs Nuclear Power Plant

4 A*ITACllMENT (4)

APPENDIX A. TECHNICAL INFORMATION 5.11H PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM 1

Group 4 (radiation damage, elastomer degradation, and wear for -en metallic ubcomponent parts). Methods to Manage Aging Mitigation: Radiation damage can be mitigated by reducing the component's exposure to radiation through shielding. Elastomer degradation can be mitigated by utilizing materials that are less susceptible to heat and oxygen. Wear can be n.itigated by midmizing vibration of the duct and heat exchangers and by minimizing operation of the dampers to slow degradation of the seating surfaces, which lead to a loss ofleak tightness.

Discoverv: Periodic visual inspections can be performed for the equipment in Group 4 to detect the efTects of radiation damage, elastomer degradation, and wear. Degradation of the Dexible coll 1rs and rubber boots located outside containment can be detected through periodic system walkdowns because the collars and boots are readily accessible. Degradation of rubber boots for the containment air coolers can be detected during routine preventive maintenance activities. Degradation of damper seals can be detected through periodic inspections. Periodic walkdowns and preventive maintenance activities may also detect loss of the damper seal because leakage would become apparent by the backward rotation of the opposite non-operating fan. If significant degradation is discovered, the Dexible collars, damper seals, or rubber boots can be repaired or replaced a appropriate. [ Reference 1, Attachment 8]

3roup 4 (radiatlan damage, elastomer degradation, and wear for non-metallic subcomponent parts). Aging Management Program (s)

Mitigation: The system was designed to minimize vibrations by using equipment support isolators and equipment to-duct isolators, such as the Dexible collars and rubber boots. Changes to materials or to system operating practices are not deemed necessary to mitigate the effects of these ARDMs.

Implementing the discovery n.ethods discussed below wHl be adequate to manage these s.RDMs. Since there are no additional methods of mitigating radiation damage, elastomer degradation, and wear, there are no programa credited with mitigating the aging effects due to these ARDMs. [ Reference 1, s ana 8]

Discoverv: Radiation damage, clastomer degradation, and wear can be readily deteved for Group 4 components through visunt examination. Routine system walkdowns would discover the effects of these ARDMs on the external sarfaces of the Group 4 components located outside containment. Periodic preventive maintenance would lead to the discovery of the effects of these ARDMs on the internal surface. of components and for components located inside containment where routine walkdowns are not feasible. [ Reference 1, Attachment 8]

System WMidQFns Calvert C!iffs Administrative Procedure MN 1319 provides for discovery of the efTects of elastomer degradation and wear of the Primary Containment H&V System components located outside of eentainment tc performance of visual inspections during plant walkdowns. The purpose of the procedure is to provide direction for the performance cf structure and system walkdowns and for the dxumentation of the walkdown results. Under this program, inspection items typically related to aging management include identifying unusual noises ud identifying system and equipment stress or abuse, such as excessive vibrations, bent or broken component supports, etc. Specifically, signs of cracking or tearing of duct Ocxible collars would be detected during these walkdowns. In addition, loss of the Application for License Renewal 5.11B 22 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (4)

APPENDIX A TECIINICAL INFORMATION f

5.118 - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM damper seal may be detected due to the leakage becoming apparent by the backward rotation of the opposite non-operating fan. [Referenec32] Refer to the discussion above in Group 3 under Aging Manage.nent Programs for a detailed oiscussion of MN 1319.

Prgygntive Maintenance ProEAm The CCNPP Pieventive Maintenance Program has been established to maintain plant equipment, structures, systems, and components in a reliable condition for normal operation and emergency use, minimization of equipment failure, and extension of equipment and plant life. Preventive maintenance activities are scheduled every 48 weeks to perform lubrication and fan belt incpections of the penetration room exhaust fans and lubrication of the penetration room exhaust dampers. Fellowing maintenance activities, the fans are test run to assure proper operation. [ Reference 10] Personnel responsible for the maintenance activities would discover if significant elastomer degradation or wear to the penetration room exhaust damper seals is occurring. In addition, loss of damper seal may be detected during the test run by the backward rotation of the opposite non operating fan due to leakage past the failed seal.

Preventive maintenance activities currently in place will also provide for visual inspection of the containment air cooler boots. Inspections and cleaning of the containment air coolers and their associated fans are performed every two years (i.e., every refueling outage). These activities specifically include an inspection of the rubber boot for tears, holes, or deterioration. [ References 30 and 31]

Corrective actions are taken in accordance with the CCNPP Corrective Actions Program. Refer to the discussion of aging management programs for Group 2 for a detailed description of the Preventive Maintenance Program.

Group 4 (radiation damage, clastomer degradation, and wear for non metallie subcomponent parts). Demonstration of Aging Management Based on the above discussions, the following conclusions can be reached wkh respect to radiation damage, elastomer degradation, and wear for duct flexible collars, damper seals, and rubber boots:

Primary Containment il&V System ducts, dampers and heat ex: hangers provide a system o

pressure boundary function and their integrity must be maintained under CLB design conditions, Radiation damage, elastomer degradation, and wear are plausible for the flexible collars and e

rubber boots. Elastomer degradation and wear are plausible for the damper seals.

if left unmanaged, radiation damage, elastomer degradation, and wear can result in material loss, e

tearing, or cracking which could lead to loss of pressure boundary integrity.

Visual inspections conducted in accord ace with CCNPP Administrative Proceoue MN-1319 e

provide reasonable assurance that the effects of these ARDMs would be discovered for components located outside containment. Signs of cracking or tea:ing of duct collars would be detected during these walkdowns, as well as su;h conditions as excessive vibrations and leakage of dampers, which would become apparent by the opposite non-operating fan rotating backwards.

Existing routine preventive maintenance activities to periodically lubricate and inspect the fan belts on the penetratioa room exhaust fans and to lubricate the penetration room exhaust dampers provide reasonable assurance that the effects of these ARDMs on the damper seals would be detected.

Applicatbn for License Renewal 5.11 B-23 Cak ert Cliffs Nuclear I om.

l 1

ATTACHMENT (4)

APPENDIX A - TECilNINL INFORMATION i

5.118 - PRIMARY CONTAINMENT HEAWNG AND VENTILATION SYSTEM l

Existing routine preventive maintenance activities to periodically inspect the containment air e

coc! cts and associated fans provide reasonable assuran

" he effects of these ARDMs on the rubber boots would be detected.

%erefore, there is reasonable assurance that the effects of radiation damage, elastomer degradation, and -

wear for duct flexible collars, damper seals, and heat exchanger rubber boots will be managed in such a way as to maintain the components' pressure boundary integrity, consktent with the CLB, during the s

period ofextended operation.

Group 5 (crevice corrosion and pitting of heat exchanger cooling coils) - Materials and Environment The cooling coils in the containment air coolers provide a system pressure boundary for a safety related portion of the SRW System. The cooling coils are constructed of 90/10 copper nickel. The internal environment is SRW that is chemically treated water. The external environment is air with some humidity. Since the purpose of this component is to col the air, condensation occurs on the outside surface of the coil. [ Reference 1. Attachments 4 and f j Group 5 (crevice corrosion and pitting of hea[cx6 anger cooling coils)- Aging Mechanism Effects Crevice corrosion is intense, localized corrosion within crevices or shielded areas. It may occur in areas presenting a crevice g:ometry that allows the process fluid to stagnate and/or environmentally produced impurities to concentrate. The crevice must be wide enough to permit liquid entry and narrow enough to maintain stagnant conditions, typically a few thousandths of an inch o less. Crevice corrosion is closely related to pitting corrosion and can init ee pits in many cases, as well as lead to stress corrosion i

cracking. [ Reference 1, Attachments 6 ana 7]

Pitting is another form oflocalized attack with greater corrosion rates at some locations than at others.

Pitting can be very insidior2 and destructive. with sudden failures in high pressure applications (especially in tubes) occurring by perforation. This form of corrosion essentially produces holes of varying depth to diameter ratios in the steel. Deep pitting is more common with passive metals, such as austenitic stainless steels, than with non passive metals. Pits are generally elongated in the directio.. of gravity. In many cases, crosion corrosion, fretting corrosion, and crevice corrosion can also lead to pitting. [ Reference 1, Attachments 6 and 7]

For the containment air cooler cooling coils, long term exposure to the moist environment ms.y result in localized material loss of the internal and/or external surfaces of the coils and, if unmanaged, could eventually result in loss of the pressure-retaining capability under CLB design loading conditions.

Crevice corrosion and pitting tre plausib'e for the cooling coils particularly in areas where there are crevices and/or stagnant conditions. [ Reference 1 Attachments 5 and 6]

Group 5 (crevice corrosion and pitting of heat exchanger cooling coils)- Methods to Manage Aging hiitigation: The efTects of corrosion on the internal surface cannot be completely prevented, but they can be mitigated by minimizing the exposure of the cooling coils to an aggressive environment. Maintaining an environment of purified water with controls o:. pit, oxygen, suspended solids, and chlorides during i

Application for License Res.ewal 5.118-24 Calvert Cliffs Nuclear Power Plant

AITACilMENT (4) l l

APPENDIX A

'fECHNICAL INFORMATION l

5.11H - PRIMARY CONTAINMENT IIEATING AND VENTILATION WSTEM normal plant operation can mitigate these ARDMs. Since there is no design feature to control the humidity of the air the cooling coils are exposed to, the only feasible methods of mitigating the efTects of corrosion of the external surface is to replace the coils with coils constructed of a more corrosion resistant material and to periodically clean the coils to remove corrosive impurities. [ Reference 1,

' ]

Discoverv: The effects of corrosion (crevice corrosion and pitting) on the containment air cooler cooling coils can be discovered and monitored through non destructive examination techniques such as visual inspections. Periodic inspections / cleaning would lead to discovery of corrosion of components that are re.dily observabb during the activity. [ Reference 1 Attachment 8] If corrosion is occurring on the internal sarfaces, the degradation can be detected through planned visua' inspections or testing (e.g., ultrasonic testing, eddy current testing, pressure testing).

Group 5 (crevice corrosion and pitting of heat exchanger co; ling coils) - Aging Management Program (s)

Mitigatim: The CCNPP Chemistry Program is relied upon for monitoring and maintaining SRW chemistry to control the concentrations of oxygen, chlorides, other chemicals and contaminants. For example, the water is chemically treated to minimize the amount of oxygen in the water, which aids in the prevention and control of most corrosion mechanisms. Continued maintenance of system water quality will ensure minimal degradation of the cooling coil internals. Routine cleaning of the coils to remove corrosive impurities will continue to be done in conjunction with the preventive maintenance inspections discussed below in Discovery. [Referene: 1, Attachment 8; Reference 37]

Calvert Cliffs Chemistry Procedure CP 206, " Specifications and Surveillance for Component Cooling / Service Water Systems," describes the surveillance and specifications for monitoring the SRW System fluid CP 206 lists the parameters to monitor, the frequency of monitoring these parameters, and the target and action levels for the SRW System fluid rarameters. These chemistry parameters are currently monitored on a frequency ranging from three times per week to once a month. All of the paremeters listed in CP 206 currently have target values that give an accepable range or limit for the associated parameter. [ Reference 37, Attachment 1]

De CCNPP Chemistry Program has been subject to periodic internal assessment activities. Internal audits are performed to ensure that activities and procedures established to implement the requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program. Thes: audits provide a comprehensive independent verification and evaluation of quality related activities and procedures. Audits of selected aspects of operational phase activities are performed with a frequency commensurate with their strength of performance and safety significance, and in such a manner as to nssuie that an audit of all safety related functions is completed within a period of two years.

[ Reference 38, Section 18.18] nese activities, as well as other external assessments, help to maintain highly effective chemistry control. Continuous improvement is also achieved through monitoring industry initiatives and trends in the area of corrosion control. For cxample, in 1996, CP 206 was revised to include dissolved iron as a chemistry parameter. Dissolved iron was added to CP-206 to act as a method to discover at y abnormal corrosion of carbon steel components.

% plication for License Renewal 5.11B 25 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4)

APPENDIX A - TECIINICAL INFORMATION 5.11B - PRIMARY CONTAINMENT IIEATING AND VENTILATION SYSTEM Discoverv: For the containment air cooler cooling coils, crevice corrosion and pitting can be readily detected through non-destructive examination techniques. Periodic inspections as part of the Preventive Maintenance Program will provide assurance that the effects of corrosion of the cuernal surface are not threatening the pressure-retaining capability of the cooling coils. The internal aurface cf the cooling coils will be included in the scope of an ARDI Program. [ Reference 1 Attachment 8J Preventive Maintenance Proeram The CCNPP Preventive Maintenance Program has been established to maintain plant equipment, structures, systems, rid components in a reliable condition for normal operation and emergency use, minimization of equipment failure, and extension of equipment and plant life. Calvert Cliffs currently has preventive maintenance activities for performing inspections and spray washing the containment air coolers every four years (i.e.,every other refueling outage). These activities specifically include an inspection of the cooling coils for leaks, which would lead to the discovery of any crevice corrosion or pitting on the external surface of the coils. (References 39] Corrective actions are taken in accordance with the CCNPP Corrective Actions Prograrr.. Refer to the discussion of aging management programs for Group 2 for a detailed description of the Preventive Maintenance Program.

ARDI Pronram The containment air cooler cooling coils will be included within a new plant program to accomplish the needcd inspections for corrosion of the internal surfaces. This program is the ARDI Program as defined in the CCNPP IPA Methodology presented in Section 2.0, Corrective actions will be taken in accordance with the CCNPP Corrective Actions Program and will ensure that the components will remain capable of performing their pressure boundary integrity function under all CLB conditions. Refer to the Group i discussion on aging management programs for a detailed discussion of the ARDI Program.

Group 5 (crevic[:orrosion and pitting of heat exchanger cooling coils)- Demonstration of Aging Management Based on the factors presented above, the following conclusions can be reached with respect to crevice l

corrosion and pitting of the containment air cooler cooling coils;

'Itc cooling coils provide e system pressure boundary for the SRW System and their integrity e

must be maintained under all CLB conditions.

Crcvice corrosion and pitting are plausible for the cooling coils, and result in material loss which, ifleft unmanaged, can lead to loss of pressure boendary integrity.

l The CCNPP Chemistry Program controls fluid (hemistry of the SRW System to minimize the corrosiveness of the enviromr.ent for components exposed to SRW.

The containment air coolers will be included in a new ARDI Program to complete the necessary inspection for the effects of corrosion on the internal surfaces.

Existing preventive maintenance activities to clean and inspect the containment air coolers provide reasonable assurance that the effec.s of corrosion on the external surfaces will be discovered.

Application for License Renewal 5.11 B-26 Calvert Cliffs Nuclear Power Plant

4 ATTACllMENT (4) l APPENDl!! A - TECliNICAL INFORMATION f,.11B - PRIMARY CONTAINMENT llEATING AND VENTILATION SYSTEM nerefore, there is reasonable anurance that the effects of crevice corrosion and pitting on the containment air cooler cooling coils will be managed in such a way as to mainta%.he components' pressure boundary integrity, consistent with the CLB, during the period of extended operation.

5.11H.3 Conclusion ne programs discussed for t'se Primary Containment il&V System are listed in Table 5.1lB-3. These programs are (and will be for new programs) administratively controlled by a formal review and approval process. As has been demonstrated in the above section, these programs will manage the aging mechanisms and their effects such that the intended functions of the components of the Primary Containment il&V System will b: maintained, consistent with the CLB, during the period of extended operation.

He analysis / assessment, corrective action, and confirmation / documentation process for license renewal is in accordance with QL-2, " Corrective Actions Program." QL-2 is pursuant to 10 CFR Part 50, Appendix B, and covers all structures and components subject to AMR.

Table 5.11B-3 AGING MANAGEMENT PROGRAMS FOR Tile PRIMARY CONTAINMENT II&V SYSTEM Progm m Credited For Existing CCNPP Cont inment Lenkage Rate Discovery and management of leakage that Testing Program could be an effect of seating surface wear of the check valves, control valves, and MOVs that Surveillance Test Procedures provide containment pressure boundary STP M 5711 1, STP M 57112. and (Group 1)

Discovery and management of leakage that

/

could be an effect of crevice corrosion, general corrosion, MIC, and pitting on the seating surfaces of conti.nment isolation valves that are potentially exposed to moisture (Group 2)

Existing CCNPP Preventive Maintenance Program Preventive Maintenance Checklists Discovery and management of the effects of MPM09150 and MPM09151 crevice corrosion, general corrosion, MIC, and pitting for the containment air cooler housings (Group 2)

Preventive Maintenance Checklists liitigation, discovery, and management of the MPM09150 and MPM09151 t Tects of dynsmic loading of the containment air cooler far.$ (Group 3)

Preventive Maintenance Checklists Mitigatien, discovery, and management of the MPM04112 and MPM04197 effects of dynamic loading of the containment iodine removal fans (Group 3) i Application for License Renewal 5.11 B-27 Calvert Cliffs Nuclear Power Plant

. ~.

ATTACllMENT (4)

APPENDIX A - TECHNICAL INFORMATION 5.llH - PRIMARY CONTAINMENT HEATING AND VENTILATION SYSTEM

~

Program Credited For Preventive Maintenance Checklist Discovery and management of the effects of MPM04111 radiation damage, elastomer degradation, and wear of damper seals (Group 4)

Preventive Maintenance Checklists Discovery and management of the effects of MPM09150, and MPM09151 radiation damage, clastomear degradation, and wear of rubber boots for the containment air coolers (Group 4)

Preventive Maintenance Checklists Mitigation, discovery, and management of the MPM09007 efTects of crevice corrosion and pitting for the external surface of the containment air cooler cooling coils (Grour 5)

Existing CCNPP Administrative Procedure Mitigation, discovery, and management of the MN-1 319, " Structure and System effects of dynamic loading of fans located Walkdowns" c,utside containment (Group 3)

Discovery and management of the effects of elastomer degradation and wear of duct flexihte collars and damper seals located outside containment (Group 4)

Existing CCNPP Chemistry Program Procedure Mitigation of the effects of crevice corrosion CP 206," Specifications and and pitting for the internal surface of the Surveillance for Component containment air cooler cooling coils (Group 5)

Cooling / Service Water Systems" New ARDI Program Discovery and management of the effects of seating surface wear of the hand valves (Group 1)

Discovery and management of the effects of

=

crevice corrosion, general corrosicn, MIC, and pitting for piping, hand valves, and MOVs that are potentially exposed to moisture (Group 2)

Discovery and management of the effects of crevice corrosion and pitting for the internal surfaces of the containment air cooler cooling coils (Group 5)

Application for License Renewal 5.11 B-28 Calvert Cliffs Nuclear Power Plant

ATTACilMENT (4) l APPENDIX A TECIINICAL INFORMATION J.llB - PRIMARY CONTAINMENT llEAT!NG AND VENTILATION SYSTEM 5.11H.4 References 1.

"CCNPP Primary Containment il&V System Aging Management Review Report," Revision 1 Feb-uary 7,1997 2.

CCNPP Report " System Level Screening Hesults,", Revision 4 April 6,1995 3.

CCNPP Drawing No. 60723Sil001, " Ventilation Systems:

Containment, Turbine, and Penetration Rooms," Revision 39, April 25,1997 4.

CCNPP Drawing No. 60723Sil002, " Ventilation Systems:

Containment, Turbine, and Penetration Rooms," Revision 29, September 8,1997 5.

CCNPP Drawing No. 60723S11003, " Ventilation Systems:

Containment," Revision 17, September 8,1997 6.

"CCNPP Updated Final Safety Analysis Report," Revision 21 7.

CCNPP Drawing No. 60710S110002, "Compnent Cooling System, Unit 1," Revision 31, January 17,1996 8.

CCNPP Drawing No. 62710S110002, " Component Cooling System, Unit 2," Revisior,19, January 17,1996 9.

" Component Level Screening Results for the Containment il&V System, System No. 060, CCNPP," Revision 1, July 23,1996 10.

CCNPP Preventive Maintenance Checklist MPM04111 " Lubricate Containment Penetration Room Exhaust Fans" 11.

Letter from Mr. T. T. Martin (NRC) to Mr. A. E. Lundvall, Jr. (BGE) dated June 30,1982, inspection No. 50-317/8215 (Routine, Unannounced Inspection of the Containment Penetration Leakage Testing Program, the Containment Integrated Leakage Rate Test, Tours of Facility, and Follow up on Previous inspection Findings, June 16,17,18,21,22,1982) 12.

Letter from Mr. T. T. Martin (NRC) to Mr. A. E. Lt.ndvall, Jr. (BGE) dated January 20,1983,

" Inspection No. 50-318/82 26" (Routine, Unannounced inspection of Procedure Review, Witnessing and Results Evaluation of Local Leak Rate Test and Integrated Leak Rate Test, December 15 through 18,1982) 13.

Letter from Mr. / U. Ebneter (NRC) to Mr. A, E. Lundvall, Jr. (BGE), dated June 25,1985,

" Inspection No. 50 317/85-10"(Routine, Announced Inspection of the Containment Leakage Testing Program including Procedure Review of Containment Integrated Leakage Rate Test (CILRT) and Local Leak Rate Test [LLRT] Procedures, CILRT and LLRT Witnessing. CILRT and LLRT Test Review, On Line Primary Containment Leakage Monitoring, and General Tours of the Facility, April 29 May 2, and May 17 - 21,1985) 14.

Letter from Mr. S. I'.

Ebneter (NRC) to Mr. A. E. Lundvall, Jr. (BGE), dated December 24,1985, " Combined inspection Nos. 50 317/85 33 and 50 318/85 33" (November 18 through 25,1985) 15.

Letter from Mr. S. A. McNeil (NRC) to Mr. G. C. Creel (BGE), dated March 15, 1989,

" Issuance of Technical Specification Amendment and Temporary Exemption Concerning Application for License Renewal 5.11B 29 Calvert Clift iclear Power Plant

l l

NITACilMENT (4)

APPENDIX A - TECIINICAL INFORMATION 5.118 PRIMARY CONTAINMENT }{ EATING AND VENTILATION SYSTEM Retest Schedular Requirements of Appendix J to 10 CFR Part 50 for Types B and C Local Leak t

Rate Tests (TAC No. 71589)"[ Amendment No. I 18, Unit 2]

16.

Letter from Mr. D. G. Mcdonald, Jr. (NRC) Mr. Mr. G. C. Creel (BGE), dated February 19,1992, " Issuance of Amendments for CCNPP Unit No.1 (TAC No. M82213) and Unit No. 2 (TAC No. M82212)" [ Amendment Nos. 168/147]

17.

Letter from Mr. A. W. Dromerick (NRC) Mr. C.11. Cruse (BGE), dated February 11,1997,

" Issuance of Amendments for CCNPP Unit No.1 (TAC No, M97341) and Unit No. 2 (TAC No. M97342)" [ Amendment Nos. 219/196]

18.

Letter from Mr. C.

11. Cruse (BGE) to NRC Document Control Desk, dated November 26,1996, " License Amendment Request; Adoption of 10 CFR Part 50, Appendix J, Option B for Types B and C Testing" 19.

CCNPP Report " Component Pre Evaluation for the Primary Containment il&V System,"

Revision 1, January 24,1997 20.

CCNPP Engineering Standard ES 014, " Summary of Ambient Enviionmental Service Conditions," Revision 0, November 8,1995 21.

Letter from Mr. G. C. Creel (BGE) to NRC Document Control Desk, dated May 25,1989,

" Response to Request for AdditionalInformation Generic Letter 8814, instrument Air Supply System Problems Affecting Safety Related Equipment" 22.

Letter from Mr. A. W. Dromerick (NRC) to Mr. C.11. Cruse (BGE), " Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit 1(TAC No. M92549) and Unit 2 (TAC No M92550)," dated December 10,1996 [ Amendment Nos. 217/194]

23.

Surveillance Test Procedure STP-M 5711 1," Local Leak Rate Test, Penetrations ID (Oxygen Sampling),47A,47B,47C,47D,48A,48B,49A,49B,49C (llydrogen Sampling)," (Unit 1),

Revision 0, May 16,1991 24.

Surveillance Test Procedure STP-M 57112," Local Leak Rate Test, Penetrations ID (Oxygen Sampling),47A,47B,47C,47D,48A,48B,49A,49B,49C (llydrogen Sampling)," (Unit 2),

Revision 1, March 18,1997 25.

Surveillance Test Procedure STP M 6711, " Containment Purge Isolation Valve Leak Rate Test, Pcnetrations 13 and 14,"(Ur.it 1), Revision 4, July 30,1991 26.

CCNPP Administrative Procedure EN-4105," Containment Leakage Rate Testing Program,"

Revision 1, March 14,1997 27.

10 CFR Part 50, Appendix J " Primary Reactor Containment Leakage Testing for Water.

Cooled Power Reactors" 28.

Letter from Mr. R. E. Denton (DGE) to NRC Docurrent Control Desk, " License Amendment Request: Adoption of 10 CFR Part 50, Appendix J, Option B for Type A Testing,"

January 10,1996 29.

CCNPP Administrative Procedure MN 1-102,"Preventise Maircenance Program," Revision 5, September 27,1996 30.

CCNPP Preventive Maintenance Checklist MPM09150," Inspect Containment Air Coolers" Application fr.r License Renewal 5.11 B-30 Calvert Cliffs Nuclear Power Plant

ATTACilMENT (4)

APPENDIX A - TECilNICAL INFORMATION 5.llH - PRIMARY CONTAINMENT llEATING AND VENTILATION SYSTEM 31.

CCNPP Preventive Maintenance Checklist MPM09151," Inspect Containment Air Coolers" 32.

CCNPP Administrative Procedure MN 1319," Structure and System Walkdowns," Revision 0, September 16,1997 33.

CCNPP Preventive Maintenance Checklist MPM04197, " Lubricate Containment lodine Removal Fan Motor" 34.

CCNPP Preventive Maintenance Checklist MPM04112. " Lubricate Containment lodme Removal Fan Motor" 35.

CCNPP Specification No. 6750 M 196, "Specliication for lleating, Ventilating, and Air Conditioning Ducts," Revision 4, June 14,1974 36.

EPRI Report NP 2129, " Radiation Effects on Organic Materials in Nuclear Plants," Final Report, November 1981 1

37.

CCNPP CP 206," Specifications and Surveillance Component Cooling / Service Water System,"

Revision 3, November 4,1996 38.

BGE " Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant," Revision 48 39.

CCNPP Preventive Maintenance Checklist MPM09007, " Inspect / Clean Containment Air Cooler Coils"

)

e Application for License Renewal 5.118-31 Calvert Cliffs Nuclear Power Plant

.c, l

L, ATTACIIMENT (5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 3,1998 i

I NITACHMENT (5)

APPENDIX A - TECliNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 5.15 Safety injection System This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing the Safety injection (SI) System. The SI System was evaluated in accordance with the Calvert Cliffs Nucl ar Power Plant (CCMPP) Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LRA.

5.15.1 Scoping System ievel scoping describes conceptual boundaries for plant systems and structures, develops screening tools which capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify systems and structures within the scope of license renewal. Component level scoping describes the components within the boundaries of those systems and structures that contribute to the intended functions. Scoping to determine components subject to aging management review (AMR) begins with a listing of passive intended functions and then dispositions the component types as either only arsociated with active fuwtlons, subject to replacement, or subject to AMR either in this report or another report.

Representative historical operating experience pertinent to aging is included in appropriate areas to provide insight supporting the aging management demonstrations. This operating experience was obtained through key. word searches of BGE's electronic database of information on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.

Section 5.15.1.1 presents the results of the system level scoping,5.15.1.2 the results of the component level scoping, and 5.15.1.3 the results of scoping to determine coraponents subject to an AMR.

5.15.1,1 System Level Scoping 0

This section begins with a description of the system, which includes the boundaries of the system as it was scoped. The intended functions of the system are listed ard are used to define what portions of the system are within the scope oflicense renewal.

System Descriotion/Concentual Boundaries He major functions of the SI System are to: (a) supply emergency core cooling in the unlikely event of a Loss-of Coolant Accident; and (b) increase shutdown maigin following the rapid cooldown of the Reactor Coolant System (RCS) caused by a rupture of a main steam line. These functions are performed by injecting borated water into the RCS. ;* Reference 1, Section 1.1.1; Reference 2, Section 6.3.1) The SI System is also utilized to: (a) remove heat from the RCS during plant cooldown once RCS temperature is below 300'F; (b) maintain suitable RCS temperatures during refueling and maintenance gerations; and (c) provide storage capacity for borated water needed for spent fuel pool (SFP) and rerueling pool operations. (Reference 1, Section 1.1,1; Reference 2, Sections 63.1,6.3.2.4, and 9.2.4) During normal plant operations, the SI System is maintained in a standby mode with components aligned for injection to the RCS. [ Reference 3, Section 5.0)

Appli.ation for License Renewal 5.15 1 Calvert Cliffs Nuclear Power Plant I

ATTACIIMENT (5)

APPENDIX A TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM The Si System consists of high pressure and low pressure subeystems that provide borated water to four Si headers, each connected to associated cold leg piping in the RCS. In addition to the associated piping, valvn, controls, and instrumentation, the S: System for each CCNPP Unit comprises the following major components: (Reference 2, Sections 6.3.1 and 6.3.2]

Three electric motor driven high pressure safety injectior WPSI) pumps, each with an associated seal coolcr; Two electric motor driven low pressure safety injection (LPSI) pumps, each with an associated e

seat heat exchanger (llX);

Four safety injection tanks (SITS); and e

A refueling water tank (RWT), with an associated electric motor driven pump and heat exchanger e

(RWTilX).

In 19(19, a bent vertical support in the Unit i LPSI suction piping was identined, investigation detemiined that securing one LPSI pump while others were running resulted in s!amming of the secured pump's discharge check valve (CKV) and a subsequent water hammer transient of sufUcient magnitude to detorm several supports. Following a design evaluation, the damaged supports were replaced with upgraded supports. [ Reference 4] Operating procedures were modified to require throttling How prior to securing a pump if two LPSI pumps are running.

In 1994, a non isolable leak at a welded Otting in the discharge test piping for 22A SIT resulted in plant shutdown. Metallurgi:al analysis of the cracked weld concluded that high cycle fatigue caused crackir.3 at the weld joint. [ Reference 5] Changes had been made to the support installed at this connection during replacement of the SIT outlet CKV in 1993. Other locations were evaluated and found to be acceptable. [ Reference 6] Vibration measurements and computer modeling subsequent to the failure showed that the pipe support configuration introduced a high potential for harmonic oscillation, coinciding with the excitation frequency of the RCS, to occur. [ Reference 5] The test connection was replaced and the pipe s ipport con 0guration was changed to prevent vibration at the critical frequency.

In 1998, damage was found on the stanchion and restraining steel of a pipe support located on the common LPSI pump discharge header in Unit 1. The cause of the damage is currently understood to be the result of water hammer loading caused by LPSI pump discharge CKV slam (s). The cause of the apparent CKV slam (s) is still under investigation.

In the past, cracking occurred at certain other welds in SI System piping. [ References 7,8, and 9] In each case, inspection and/or evaluation concluded that similar piping and components in the remainder of the SI System for both units were not experiencing similar effects. [ Referent 9 and 10] Analysis following each event determined that cracking was due to high-cycle fatigue causeo by vibration inherent in the design of the system. Reconfiguration of the affected piping and associated supports has prevented recurrence of cracking at each location. [ References 9 and 11] Since normal and design operating conditions result in neither excessive cycling nor signincant mechanical / vibrational loading conditions, high cycle fatigue is not plausible for the SI System.

Many other SI System compcnents have been disassembled over the plant's operating history with no unusual or unexpected signs of wear or degradation noted.

Application for License Renewal 5.15 2 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM i

ne SI System is composed of the following general categories of equipment and devices: (Reference 1 Section 1.1.2; Reference 2. Section 6.3.2.4; Reference 12. Table 2)

Piping Convey borated water to perfona system functions; Valves Control valves (CVs), CKVs, hand valves (llVs), motor-operated valves (MOVs), and relief valves (RVs), which provide containment isolation and system alignment / isolation / protection; instruments Measure system flow rates, tank levels, and temperatures; Tanks Store borated water used for injection into the RCS and for refueling pu uses; ilXs Provide a heat sink for seal cooling water for system pumps, and previ

• ireezing of borated water in the RWT; and Pumps Provide motive force to move borated water into the RCS, into the SITS, and into the RWT.

System Interfacss The SI System has interfaces with 11 plant systems. These interfaces and their applicability for license renewal are discussed below.

He following interfaces with the SI System are within the scope oflicense renewal:

Engineered Safety Features Actuation System (ESFAS) [ Reference 1. Section 1.1.2) The ESFAS e

supplies control signals to SI System components in response to Design Basis Event (DBE) conditions. When a Safety injection Actuation Signal (SIAS) has been generated, the ESFAS provides a signal to: (a) start two llPSI pumps and both LPSI pumps: (b) open the LPSI header, main IIPSI header, auxiliary llPSI header, and SIT outlet MOVs; and (c) close the SlT CKV leakage and leakofTto reactor coolant drain tank CVs. When the inventory in the RWT is nearly depleted, s{gnals from SI System level switches cause a Recirculation Actuation Signal (RAS) to be generated in the ESFAS. In response, the ESFAS provides a signal to: (a) open the containment sump discharge MOVs;(b) shut down the LPSI pumps; and (c) close the mini flow return to RWT isolation MOVs.

[ Reference 2, Section 6.3.3)

This interface involves cables / conduits associated with transmittin;, signals between the ESFAS and the SI System, as well as controls associated with the pumps and valves.

RCS [ Reference 1, Section 1.1.2] De SI System has several interfaces with the RCS. The RCS supplies control signals to SI System MOV5 in response to RCS pressure. When pressurizer pressure exceeds preset values, the RCS provides signals to: (a) open the SIT outlet MOVs; and (b) prevent opening of the shutdown cooling (SDC) header return isolation MOV outside containment. [ Reference 2, Sections 6.3.1 and 9.2.6) This interface involves cablesmonduits associated with transmitting signals to the SI System, and controls associated with the MOVs.

He Sl 3ystem piping interfaces with the RCS at the discharge of the loop inlet CKVs. Each of these four connections allows injection of borated water to the reactor pressure vessel through an inlet nozzle in the associated RCS cold leg piping in the event emergency core cooling is needed.

[ References 13 and 14) During a normal plant shutdown, this piping is also used to supply ?DC flow from the LPSI pumps. He SI System piping also interfaces with the RCS piping at the Application for License Renewal 5.15-3 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (5) l APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM l

outlet of the SDC header return isolation MOV inside containment, allowing return of SDC How.

[ Reference 2, Section 4.1.2]

Containment Spray (CS) System [ Reference 1, Section 1.1.2] The SI System piping has several -

interfaces with the CS System. On CS pump start in response to a SIAS, the SI System provides borated water to the CS pump suction header from he RWT. Prior to initiation of a Contalument Spray Actuation Signal, How from the CS System returns to the RWT tigough the CS pump mini Dow return CKVs. [ Reference 2, Section 6.4.4] After a RAS initiation, operators may choose to divert a portion of the cooled water from the CS headers to the suction of the llPSI pumps. [ Reference 2, Section 6.3.1) in the SDC mode of operation, borated water from the LPSI pump discharge header Hows through the shutdown cooling heat exchanger (SDCHX) to the LPSI return header on its way into the RCS cold leg. [ Reference 2, Section 9.2.1) Additional connections to the CS System allow Dow:

(a) from the SDC retum header to Dow instrumentation in the SI System flowpath used during purification; (b) from the SDC return header to the RWT retum header; and (c)from the CS header to the RWT return header.

[ Reference 3, Section 6.5; Reference 15, Sections 6.11 and 6.12; Reference 16, Section 6.2]

Component Cooling (CC) System (Reference ! Section 1.1.2] The CC System provides How to

+

the LPSI pump seal llXs and ilPSI pump seal coolers, as well as cooling water piped directly to the bearing housings and stuffing box Jackets for these pumps, [ References 17 through 21]

SFP Cooling System [ Reference 1, Section 1.1.2] The SI System provides makeup water for the SFP through a piping connection end valve at the RWT.- (Reference 2, Section 6.3.2.4] The SI System piping also interfaces with the SFP Cooling System at the LPSI pump suction header to provide additional SFP cooling when the complete core is removed from the reactor vessel and temporarily stored in the SFP. [ Reference 2, Section 9.2.l; Reference 16, Sections 6.11 and 6.12; Reference 22]

Chemical and Volume Control System (CVCS) [P+mnce 1, Section 1.1.2) The SI System e

piping has several interfaces with the CVCS. One..ection allows the charging pumps to discharge to the auxiliary HDSI header as an attemate charging path to the RCS. [ References 23 and 24] Two other connections provide for reactor coolant purification during SDC operations.

Another connection allows transfer of borated water between the CVCS and the RWT.

[ References 25 and 26]

Primary Containment System [ References 27 and 28] Two recirculation headers in the SI System connect to the emergency sump inside containment. After a RAS initiation, the llPSI pumps take suc' ion directly from the emergency samp, [ Reference 2, Section 6.3.1)

The following interfaces with the SI System are not within the scope oflicense renewal:

Compressed Air System [ Reference 1, Section 1.1.2] The Compressed Air System provides instrument air for operation of CVs in the SI Sy stem. [ References 29 and 30]

Nitrogen and Hydrogen System [ Reference 1, Section 1.1.2) A distribution header in the Nitrogen and liydrogen System provides nitrogen gas used to pressurize the SITS. [ Reference 2, Section 6.3.2.3]

Plant lleating System [ Reference 1, Section 1.1.2] The Plant Heating System provides hot water now through the shell side of the RWTHX; borated water from the RWT is circulated through the Application for License Renewal 5.15-4 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15. SAFETY INJECTION SYSTEM tube side of the RW111X to prevent freezing of the RWT contents in winter. (Reference 2, Section 6.3.2.4) 1.iquid Waste System [ References 31 through 34) The SI System has several interfaces with the e

Liquid Waste System. Drainage from the RWT, the RWT circulating pump casing, and the RWT circulating pump RV is collected by the miscellaneous waste processing subsystem.

[ Reference 2, Sectiou 11.1.2.1.3) Flow from the SIT leakoff return lines is directed to the reactor coolant drain tank during normal operation. [ Reference 2, Section 11.1.2.1.2)

Figures 5.15-1 and 5.15 2 comprise a simpli0cd diagram of the St System and are provided for information only. These figures depict the major St System components and interfaces discussed above.

System Sconing Results The SI System is in see - for license renewal based on 10 CFR 54.4(a). The following intended functions of the SI System were determined based on the requirements of {54.4(a)(1) and (2).n accordance with the CCNpP IPA Methodology Section 4.1.1: [ Reference 12, Table 1]

To provide borated water to the RCS for reactivity control, and pressure arid level control in response to DBEs upon SIAS; To provide borated water to the RCS for reactivity control, and pressure and level control (passively) when RCS pressure drops below 200 psig; To recirculate lost coolant back to the RCS and CS System (Recircuiation Mode);

To send a signal to ESFAS for RAS To provide long term core flush via hot leg injection; To provide containment isolation of the Si system during a Loss-of Coolant Accident; e

To maintain the pressure boundary of the system (liquid and/or gas);

To maintain mechanical operability and/or provide protection of the mechanical system; To provide borated water from the RWT to the CS pumps; e

To maintain electrical continuity and/or provide protection of the electrical system; To provide makeup water from the RWT to the SFP during a fuel handling incident; and To restrict flow to a specified value in support of a DBE response.

e Application for License Ren:wal 5.15 5 Calvert Cliffs Nuclear Power Plant J.AA

ATTACHMENT f5)

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Application for License Renewal 5.15-6 Calvert Clifts Nuclear Power Plant I

ATTACILMENT f5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM CoNTANMENT '

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FIGURE 5.15-2 SAFETY INJECTION SYSTEM (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY)

Applicatic.e for License Renewal 5.15-7 Calvert ClilTs Nuclear Pourr Plant x

m a ma --

4 A'ITACIIMENT (5)

{

APPENDIX A - YECIINICAL INFORMATION 5.15. SAFETY INJECTION SYSTEM The following intended functions of the SI System were determined based on the requirements of 654.4(a)(3): [ Reference 12, Table 1)

For Ore protection (650.48) To: (a) provide RCS pressure and inventory coatrol to ensure safe shutdown in the event of a postulated severe Gre;(b) provide RCS heat removal by realigning and operating in the SDC mode; and (c) prevent inadvertent dumping of the SITS when RCS temperature is less than 300'F.

For environmental quali0 cation (650.49) - To: (a) provide information used to assess the environs and plant condition during and following an accident; and (b) maintain functionality of electrical components as addressed by the Environmental Qualification Program.

For station blackout (650.63)

To: (a) provide valve position indication and closure of containment isolation va'ves; and (b) provide RCS isolation to maintain RCS inventory.

All components of the SI System that meet the environmental qualification or station blackout criteria of 54.4(a)(3) are also safety related. [ Reference 12, Table 2] Sorae of the components that meet the fire protection criteria are ron safety related and are within the scope oflicense renewal only because of the 54.4(a)(3) criteria. [Ref-rence 12, Table 2)

All components of the SI System that support the above functions, with the exception of the non safety related instrumentation and controls supporting the Ore protection functions, meet Seismic Category I requirements. [ Reference 12, Table 2; References 17,18,22 through 28, and 31 through 35]

The SI System piping provided with the llPSI and LPSI pumps is designed in accordance with United States of America Standard Code for Pressure Piping, Power Piping, B31.1.0-1967. [ References 36,37, and 38] All other piping in the SI System complies with the following criteria in American National Standards Institute (ANSI) Nuclear Power Piping Code B31.7,1969: [ Reference 2, Section 6.3.2.6]

Design requirements for Class I piping apply to: (a) all S1 System piping between the Si header e

CKVs, the SIT out!et MOVs, the SIT CKV leakage CVs, and the RCS interface at the loop inlet CKVs; (b) SIT discharge piping from the SIT outlet CKVs to the SIT outlet MOVs; and (c) SDC piping between the interface with RCS (i.e., at the outlet of the SDC header return isolation MOV inside containment) and the SDC header return isdation MOV outside containment.

[ Reference 39, Piping Class CC 4; Refeience 40, Piping Classes CC-13 and CC-14]

Design requirements for Class til piping apply to: (a) all SI System piping from the interfaces e

with the SFP Cooling System and CVCS to the spoolpiece connecting to the LPSI discharge header; and (b) all piping associated with the RWT heating Dowpath. [ Reference 41, Piping Class liC-4; Reference 42, Piping Class llc 23]

Design requirements for Class 11 piping apply to all remaining pipine i the SI System.

[ Reference 41, Piping Class llCA; Reference 43, Pipmg Classes DC 1 and ;

2; Reference 44, Piping Classes GC-1, GC-3, GC-4, GC 5, GC-7, and GC-9; Reference 39, Piping Class CC-6; Reference 40, Piping Class CC-13)

The design parameters for major SI System components are presented in Section 6.3.2 of the CCNPP Updated Final Safety Analysis Report.

Application for License Renewal 5.15 8 Calvert Cliffs Nuclear Power Plant

_ _ _ ~. _ _ _ _ _.____ _

9 NITACHMENT (5)

APPFNDIX A - TECHNICAL INFOlG1ATION 5.15 - SAFETY INJECTION SYSTEM 5.15.1.2 Component Level Scoping Dased on the irnended functions listed above, the portion of the SI System that is within the scope of license re.cwal includes all components (electrical, mechanical, and instrument) and their supports assxiated with the storing and deliverh.c of borated water to the RCS. The following system Dowpaths allow transfer of borated weter to the RCS interface at each of the four loop inlet CKVs (rerer to Figure 5.151): [P.eference 2, Sections 6.3.1 and 9.2.2; Reference 12, Table 2; References 27,28, and 31 through 34) injection mode Dowpath (post DBE operations after a SIAS; motive force provided by llPSI e

pumps) from the RWT, through the running IIPSI pumps (i.e., two of the three installed pumps),

and from there to the Si beader CKVs and loop inlet CKVs by way of both: (a) a main llPSI header and four main llPSI header MOVs; and (b) an auxiliary llPSI header and four auxiliary IIPSI header MOVs; injection mode Dowpath (post-DF.E operations after a SIAS; motive force provided by LPSI e

pumps)- from the RWT, through both Li'S1 pumps, into a common discharge header, through the LPSI now CV, the four LPSI header MOVs, to the Si header CKVs and loop inlet CKVs; injection mode Howpath (post DBE operations aner RCS pressure drops below approximately e

200 psig; motive force provated by pressurized SITS) from each of the four SITS, through the open SlT outlet CKVs and SIT outlet MOVs, to the loop inlet CKVs; Recirculation mode flowpath (post DBE operatioin, after a RAS; motive force provided by llPSI e

pumps) - from the interface with the containment emergency sump, through the containment sump discharge MOVs, and through the llPSI injection mode flowpath described above; and SDC mode nowpath (motive force provided by LPSI pumps)- from the RCS interface at the e

ou'.let of the SDC header return isolation MOV inside containment, through the SDC header return is.* tion MOV outside containment, and through the LPSI injection mode Dowpath described above, with a portion of the borated water pass;ng through the SDCliX LPSI inlet MOV into the CS System. Afler passing through the SDCilXs and the SDC temperature / flow CV in the CS System, this fluid re-enters the SI System on the downstream side of the LPSI flow CV, rejoining the remainder of the borated water in the SDC mode Dowpath.

The following system flowpaths form parts of the system pressure boundary and are also included within the scope of license renewal for the SI System: [ Reference 2, Section 6.3.2; Reference 12, Table 2; References 25 through 28, and 31 through 34)

Minimum flow recirculation Howpaths for pumps (motive force provided by associated pumps)-

e from the discharge headers for each liPSI and LPSI pump through an associated How orifice and mini flow return CKV, and from the CS System interface at the outlet of each CS pump mini flow return CKV, through tl.e mini flow return to RWT isolation MOVs, through the comi "WT recirculation header back into the RWT; Cire con flowpath for the RWT-from the RWT, through the RWT circulating pump, the tubes e

ir. ue RWTliX, and back into the RWT; NOTE: The RWTIIX Mnnets/ covers / tubes and associated stainless steel welds form a part of the Si System pressure boundary. The intended function for other subcomponent parts of the RWTIIX (i.e., the shell and asso3!ated carbon steel welds, fittings, studs, nuts, and vessel Application for License Renewal 5.15-9 Calvert Cliffs Nuclear Power Plant

A*ITACJIMENT (5) j APPENDIX A - TECilNICAL INFORMATION i

5.15 - SAFETY INJECTION SYSTEM 1

supports)is to provide structu al support for the tube assembly. A pressure boundary breach of the Plant lleating System will not impact this support function.

[ Reference 1 for ilXs)

SDC recirculation flowpaths from the CS System interfaces in the SDC return header up to and e

including the following:

Through the SDCilX recirculation stop CV to the LPSI pump suction header; or The common RWT recirculation header; or Through the SI to-CVCS flow instrumentation to the CVCS interface at the outlet of the SDC supply to CVCS backup ilV; and Leakoff return flowpaths for each SIT - from the SI System piping betweei the SIT outlet and e

loop inlet CKVs, through the SIT CKV leakage CV, through common leakoff return piping and a flow orifice up to and including the following:

To the Liquid Waste System interface :t the outlet of the leakoff to-reactor coolant drain tank CV, or Through the normally closed Si leakoff return header isolation llVs to the common RWT recirculation header.

Additional components that form parts of the system pressure boundary along these flowpaths (e.g., piping, instruments, seal coolers and ilXs for pumps, SIT fill and-drain CVs, normally closed ilVs, RVs, solenoid-operated valves in instrument air supply piping) and their supports are also included within the scope oflicense renewal for the S1 System. [ Reference 12, Table 2; References 17,18,22,25 through 28, and 31 through 34]

Application for License Renewal 5.15-10 Calvert Cliffs Nuclear Power Plant

ATTACHMENT _1$)

l APPENDIX A TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM i

~~

'the following 53 device types in the SI System have at least i intended function: (Reference 1 Table 21]

Class "CC" Piping (stainless steel, primary rating 1500 psig at 1125'F)

Pressure Indicator Class "DC" Piping (stainless steel, primary rating 900 psig at 1125'F)

Pressure Switch Class "GC" Piping (stainless steel, primary rating 300 psig at il25'F)

Pressure Transmitter Class " llc" Piping (stainless steel, primary rating 150 p6ig at 500'F)

Pump / Driver Assembly Circuit Breaker lland Valve ReliefValve Check Valve lleat Exchanger Relay Coil Current / Pneumatic Device Solenoid Control Valve Ammeter Solenoid Valve Control Valve Operator Power Lamp indicator Temperature Element Disconnect Switch / Link LevelIndicator Temperature Indicator Voltage / Current Device LevelIndicator Alarm Tank Flow Element Level Switch Temperature Recorder Flow Indicator Level Transmitter Temperature Transmitter Flow Indicator Controller level Device (Relay)

Power Supply Flow Orifice 4kV hiotor Expansic 1 int Flow Transmitt:r 125/250Vdc Motor Position Indicator Fuse Motor Operated Valve Position Indicating Lamp Flow Device (Relay)

Motor Operated Valve Operator Position Switch llandswitch 4kV Local Control Station (Disconnect / Link)

Position Transmitter Some components in the SI S ystem are common to many other plant systems and have been included in separate sections of the BGE LRA that address those components as commodities for the entire plant.

These components include the fobowing: (Reference 1, Section 3.2)

Except for the RWTifX supports that are addressed in this report, structural supports for piping, cables, and components are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Sectior. 3.1 of the BGE LRA. The RWTilX supports are evaluated as subcomponents of the llX devbe type. [ Reference 1, Attachment 5 for ilXs.]

Electrical control and power cabling are evaluated for the effects of aging in the Electrical Cables Commodity Evaluation in Section 6.1 of the BGE LRA. This commodity evaluation completely addresses the passive intended function entitleu mintain electrical continuity and/or provide protection of the electrical r.ystem" for the SI Syr instrument tubing and piping and the associated tubing supporn, instrument valves and fittings e

(generally everything from the outlet of the final root valve up to and including the instrument),

and the pressure boundaries of the instruments themselves, are all evaluated for the effects of aging in the Instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA. This commodity evaluation partially addresses the passive intended function entitled "maintai, the pressure boundary of the system (liquid and/or gas)" Sr the SI System.

5.15.1.3 Components Subject to AMR This section describes the components within the St System that are subject to AMR. It begins with a listing of passive intended functions and then ciispositions the device types as either only associated with I

Application for License Renewal 5.15-11 Calvert Clifts Nuclear Power Plant

ATTACilMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM active functions, subject to replacement, evaluated in other reports, evaluated in commodity reports, or remaining to be evaluated for aging management in this section.

Passive Intended Functions in accordance with CCNPP IPA Methodology Section 5.1, the following St System functions were

'etermined to be passive: [ Reference 1, Table 3.l]

To maintain the pressure boundary of the system (liquid and/or gas);

To taaintain electrical continuity and/or provide protection of the electrical system; and To restrict flow to a specified value in s'ipport of a DDE response.

e Device Tyocs Subject to AMR Of the 53 device types within the scope oflicense renewal for the SI System: (Reference 1. Table 3 2 and Attachment 3 for XJs; Reference 45, Attachment 4A]

Twenty nine device types were associated with only active fur.:tions: Circuit Breaker, Coil, CV e

Operator, Disconnect Switch / Link, Voltage / Current Device, Flow Indicator Controller, Fuse, Flow Device (Relay), llandswitch, Current / Pneumatic Device, Ammeter, Power Lamp indicator, Level indicator, Level Indicator Alarm, Level Device (Relay),4kV Motor, 125/250Vdc Motor, MOV Operator,4kV Local Control Station (Disconnect / Link), Relay, Solenoid, Solenoid Valve, Temperature Recorder. Temperature Transmitter, Power Supply, Position Indicator, Position Indicating Lamp, Position Switch, and Position Transmitter; Level Transmitters associated with the containment emergency sump are subject to periodic e

replacement based on a quali0ed life or specified time period; One device type in this system, Expansion Joint, was evaluated in the AMR for the Containment e

System, addressed in S:ction 3.3A of the BGE LRA; and Seven device types are associated with a separate cor Mity evaluation. Level Transmitters associated with the RWTs and SITS, as well as Flow.

• ors, Flow Transmitters, Level Switches, Pressure Indicators, Pressure Switches, and Pressure.

mitters in the SI System are evaluated separately in the Instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA.

The remaining 16 device types, listed in Table 5.151, are subject to AMR and are included in the scope of this section. [ Reference 1. Table 3 2] Except for llVs, all components of each listed device type are addressed in it is section. Mrnual drain, equalization, and isolation valves in Si instrument lines that are

- subject to AMR are evaluated for the effects of aging in the Instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA. The manual root valves that are used to isolate these components are evaluated in this section. [Referenes 45, Attachments 4 and 4A]

Baltimore Gas and Electric Company may elect to replace components for which the AMR identifies that further analysis or examination is needed, in accordance with the License Renewal Rule, components subject to replacement based on qualified life or specified time period would not be subject to AMR.

Application for License Renewal 5.15-12 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (5)

APPENDIX A - TECilNICAL INFORMATION 5.15 SAFETY INJECTION SYSTEM Table 5.15-1 SAFETY INJECTION SYSTEM DEVICE TYPES SUBJECT TO AMR Class CC Piping ( CC) lland Valve (ilV)

Class DC Piping ( DC) lleat Exchanger (llX)

Class GC Piping ( GC)

Motor Operated VMa (MOV)

Class llc Piping ( llc)

Pump / Driver Assembly (PUMP)

Check Valve (CKV)

Relief Valve (RV)

Control Valve (CV)

Temperature Element (TE)

Flow Element (FE)

Temperature Indicator (TI)

Flow Orifice (FO)

Tank (TK) 5.15.2 Aging Management The list of potential Age-Related Degradation Mechanisms (ARDMs) identified for the St System components is given in Table 5.15 2, with plausible ARDMs identified by a check mark (/) in the appropriate device type column. [ Reference 1 Table 4-2 and Attachment 1) A check mark indicates that the ARDM applies to at least one component for the device type listed. For efficiency in presenting the results of these evaluations in this report, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all components. Tabic 5.15-2 also identifies the group to which each ARDM/ device type combinat'on belongs. Exceptions are noted where appropriate. The following groups have been selected for the SI System:

Group 1:

general corrosion of external surfaces dua to leakage of borated water; Group 2:

general corrosion. crevice corrosion. and/or oitting of internal surfaces exposed to chemically-treated water; Group 3:

niicrobiologically-induced corrasion (MIC) of internal surfaces for recirculation header piping connected to the emergency sump inside containment; Group 4:

fatigue for piping and valves exposed to thermal transients; Group 5:

stress corrosion cracking (SCC) near RWT penetrations and associated welds; and Group 6:

Etathering of RWT perimeter seal.

Application for License Renewal 5.15-13 Calvert Clifts Nuclear Power Plant

w..

ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Table 5.15-2 POTENTIAL AND PLAUSIBLE ARDMs FOR THE SAFETY INJECTION SYSTEM Plausible l Not Device Types

- i C

C F

F H

H M

P R

T T

T Potential ARDMs C,D G

H K

V E

O V

X 0

U V

E I

K for S 88*

• C C

C C

V V

M 7

P Cavitation Erosion j

x Crevice Corrosion

/(2)i/(2) /(2)!/(2) /(2) /(2) 42) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) x Dynamic Loading x

Erosion Corrosion Fatigue

/(4)

/(4) /(4)

/(4)

/(4)

Fouling i

x x

Galvanic Corrosion General Corrosion

/(1) /(1) /(1) /(1) /(1) /(1)

/(1) /(1) /(I) /(1) /(1)

/(1) x Ilydrogen Damage x

Intergranular Attack M!C

/(3) x Particulate Wear Erosion i

Pitting

/(2)'/(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) /(2) x Radiation Damage x

Rubber / Elastomer Degradation x

Selective Leaching SCC i

/(5)

Dermal Damage i

x Hermal Embrittlement x

x Wear Weathering

/(6)

/

indicates plausible ARDM determination for this device type Note: Not every component within the device types listed here may be susceptible to a given ARDM. This is because

. group in which this ARDM/ device type components (and subcomponents) withm a device type are the (number) indicates comb.mation is evaluated.

not always fabricated from the same materials or subjected to the same environments. Exceptions for each device type will be indicated in the aging manage.nent subsection for each ARDM discussed in this report.

Application for License Renewal 5.15-14 Calvert Clifts Nuclear Power Plant

ATTACIIMENT d)

APPENDIX A - TECIINICAL INFORMATION 5.15. SAFETY INJECTION SYSTEM ne following is a discussion of the aging management demonstration process for each group identified above, it is presented by group end includes a discussion of materials and environment, aging mechanism effects, methods of managing aging, aging management program (s' and aging management demonstration.

Group 1 - (general corrosion of external surfaces)- Materials and Environment Group I comprises the various SI System components whose external surfaces are subject to general corrosion. The components in this group are included in the CC, DC, GC,11C, CKV, CV, ilV, llX, MOV, PUMP, RV, and TK device types. All of these components provide the passive intended function of maintaining the system pressure boundary.

[ Reference 1. Attachment 1)

The applicable subcomponents in these device types are constructed of the following materials:

[ Reference 1, Attachments 4 and 5 for all device types except FEs, fos, TEs, and Tis)

Piping carbon steel nuts and alloy steel studs; e

CKVs carbon steel nuts and alloy steel studs; e

CVs carbci, steel nuts and alloy steel studs; e

llVs carbon steci nuts and alloy steel studs for all llVs except the Unit 1 RWT retum from SFP e

cooling IIV; llVs - external surfaces of carbon steel body for the Unit 1 RWT return from SFP cooling IlV; e

ilXs - carbon steel nuts and alloy steel studs; carbon steel vessel supports for RWTilXs; external e

surfaces of the cast iron case / cover for pump seal coolers and ilXs; MOVs - carbon steel nuts and alloy steel studs; e

PUMPS - carbon steel nuts and alloy steel studs for liPSI, LPSI pumps; e

RVs - carbon steel nuts and alloy studs for the Si leakoff and SDC return header RVs outside e

containment; and TKs - carbon steel nuts and alloy studs for the KWTs; extemal surfaces of the carbon steel shell and skirt plate for the SITS.

The external surfaces evaluated in Group 1 are not normally exposed to a corrosive environment, but may be exposed to boric acid as a result ofleakage from the associated components or nearby systems and components that contain borated water. The possible effects of such leakage include general corrosion of susceptible external surfaces. A potential source of borated water leakage is the internal environment for the components in Group 1, with some normal service conditions as high as 2235 psig and 604'F.

[ Reference 1, Attachment 6s for all device types except FEs, fos, TEs, and Tis; Reference 39, Piping Classes CC-4 and CC 6; Reference 40, Piping Classes CC-13 and CC-14; Reference 41, Piping Classes llc-3 and ilC-4; Reference 42, Piping Class liC-23; Reference 43 Piping Classes DC-1 and DC 2; Reference 44, Piping Classes GC-1, GC-3, GC 4, GC-5, GC-7, and GC-9;)

For the RWTs, ti.e external surfaces are exposed to the normal outside atmosphere at the CCNPP site, including fluctuating temperatures and humidity, sunlight, rain, freezing rain, ice, and snow.

[ Reference 1, Attachment 3 for TKL) For all other components evaluated in Group 1, the external surfaces are exposed to an environment of climate-controlled air in either the Auxiliary Building or 'he Application for License Renewal 5.15 15 Calvert Cliffs Nuclear Power Plant

f ATTACIIMENT f5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Containment. [ Reference 1. Attachment 3s] During normal operation, temperature and humidity in the Auxiliary Buiiding do not exceed 160'F and 70%, respectively. [ Reference 46, page $4] For the general areas inside containment where SI System components are located, the maximum normal temperature and humidity values are 120'F and 70%, respectively. [ Reference 46, pages 29,30,62, and 63]

Group 1 -(general corrosion of externel surfaces) Aging Mechanism Effects General corrosion is thinning of a metal by the chemical attack of an aggressive environment at its surface. An important concem for pressurized water reactors is boric acid attack upon carbon steels and low alloy steels. General corrosion is not a concern for austenitic stainless steels. [ Reference 1, s for all device types except FEs, FOr, TEs, and Tis]

General corrorion is pinosible for all carbon steel, alloy steel, and cast iron subcomponents in this group.

Mechanicaljoints in pressure boundary subcomponents provide the opportunity for leakage of borated water onto external component surfaces. The carbon steel and alloy steel surfaces are panicularly susceptible to significant acceleration of corrosion when exposed to boric acid in the concentrations present in the SI System. [ Reference 1, Attachment 6s for all device types except FEs, fos, TEs, and Tis)

The result of this corrosion mechanism is a reduction in the integrity of the corroded parts and a resulting increase in the likelihood of mechanical failure, if unmanaged, long term exposure to general corrosion could eventually result in loss of the pressure retaining capability under current licensing basis (CLB) design loading conditions.

Group 1 -(general corrosion of external surfaces)- Methods to Manage Aging Mitigation: Boric acid corrosion can be mitigated by minimizing leakage. The susceptible areas of the SI System (i.e., mechanical joints) can be routinely observed for signs of borated water leakage, and appropriate corrective action can be initiated as necessary to climinate leakage, clean spill areas, and assess any corrosion. [ Reference 1, Attachment 6s for all device types except FEs, fos, TEs, and Tis]

Discoverv: The effects of corrosion are generally detectable by visual techniques. Visual inspections would need to be performed to detect ccrrosion associated with leakage of fluids onto the external surfaces of St System components. [ Reference 1. Attachment 6s for all device types except FEs, fos, TEs, and Tis]

Group I -(general corrosion of external surfaces)- Aging Management Program (s)

Mitigation: The CCNPP Boric Acid Corrosion Inspection (BACl) Program (MN 3 301) is credited with mitigating the effects of boric acid corrosion through timely discovery of leakage of borated water and removal of any boric acid residue that is found. [ Reference 1, Attachment 8] ' Itis program requires visual inspection of'.he components containing boric acid for evidence of leaks, quantification of any leakage indications, and removal of any leakage residue from component surfaces. [ Reference 47]

Further details on the BACI Program are detailed in the Discovery subsection below.

Application for License Renewal 5.15 16 Calvert Cliffs Nuclear Power Plant

t ATI%CIIMENT $

APPENDIX A - TECIINICAL INFOT MATION 5.15 - SAFETY INJEC". 3N SYSTEM Ay Discoverv:

Discovery of boric acid leakage.4 ensured by the BACI Program.

[ Reference :

r- ] This program also requires investigation of any leakage that is found. A visurd examination of external surfaces is performed for components containing boric acid. [ Reference 47]

1 The Inservice Inspection Program requ',d the establishment of the BACI Program to systematically 1

ensure that boric acid corrosion dus act degrade the primary system boundary. [ Reference 48,

$cction 5.8.A l.] The program also applies to " valves in systems containing borated water which could 1: ale onto Class I cerbon stec: components," and it identifies other plant areas to be examined

[ Reference 47, Section 5.lB] The program controls examination, test methods, and actions to r-k the loss of structural and pressure-retaining integrity of components due to boric acid i

[ Reference 47, Section 3.0.C] The basis for the c:tablishment of the program is Generic la

" Boric Acid Corroslon of Carbon Steel Reactor Pressure Boundary Components in PW1

.ats."

1

[ Reference 47, Section 1.1)

The scope of the program is threefold in that it: (a) identifies locations to be examined; (b) piovides examination requirements and methods for the detection ofleaks; and (c) pro A the rmnsibilities for

=

initiating engineering evaluations and necessary corrective actions. [ Refer 47, Section 1.2]

During each refueling m.hge, inservice inspection personnel perform a walkdown inspection to identify and qu9ntify any leaka.e,e found at specific locations inside the Containment and in the Auxiliary Building. The inservice inspection ensures that all components where boric acid leakage has been previously Jocumented are also examir.ed in accordance with the requirements of this program. A second inspection of these components is performed prior to plant staitup (at normal operating pressure and temperature) ifleakage was identlfied previously and corrective actions were taken. [ Reference 47, 5

Sections 5.1 and 5.2)

?

Unde-the B! C! Program, the walkdown inspections applicable to SI System components are type VT-2 (a type of v:.cial examination descrihcd in the American Society of Mechanical Engineers [ASME]

Boiler and Pressure Vesse! Code,Section XI, IWA-2212). The VT-2 visual exeminations include the accessible external exposed surfaces of pressure-retaining, n n4asulated components; floor areas or equipment stirfaces located underneath non-insulated components; vertical s.irfaces of insulation at the lowest elevation where leakage may be detected, md borizontal surfaces at each insulation joint for insulated coraponents; floor areas and equipment surfaces beneath components and other areas where water may be channeled for insulated components whose external insulation surfaces are inaccessible for direct ex:mination; and for discoloration or residue on any surface for evidenec. of boric acid accumulation. [Ref rence 47, Section 5.2]

if either 'eakege or corrosion is discovered, issue reports (irs) are genera (ed in accordance with CCNPP procedure QL-2-100, " Issue Reportii.g and Assessment," to document and resolve the deficiency.

Corrective actions address the n:moval of boric acid residue and inspection of the affected components for gene al corrosion. If general corrosion is found on a component, the IR provides for evaluation of the component for continued service and corrective c. ions to p en mt recurrence.

[ Reference 47, Section 5.3]

E The BACI Program has evolved with regard to boric acid leaks discovered during other types of walkdowns and inspections. The program specifies the minimum qualification level for inspectors Appiication for License Felewal 5.15-17 Calvert Clitfr Nuclear Power Plant

ATTACHMENT (5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM l

evaluating boric acid leaks.

Apparent leaks that are discovered during these other walkdowns/ inspections ree documented in irs by the individual discovering the leak. These irs are then routed to the inservice inspection group for closer inspection and evaluation by a qualified inspector.

This approach provides fr more boric acid leakage inspection coverage while still ensuring that appropriately qualified indiviouals asscss and quantify any resultant damage.

The corrective actions taken as a result ofIRs under this program will ensure that SI System components containing borated water remain capable of peiforming their imended function under all CLB conditions during the period of extended operation.

Group 1 -(general corresion of external surfaces)- Demonstration of Aging Man 9gement Based on the information presented above, the following conclusion: can be reached with respect to general corrosion of external surfaces for SI System components:

Re components in Group I contribute to maintaining the system pressure bour lary, and their integrity must be maintained under all CLB design conditions.

The materials of construction for subcomp3nents in this group cre carbon steel, alloy steel, or cast iron.

General corrosion is a plausible ARDM for this group because the susceptible external surfaces are exposed to potential boric acid leakage from mechanicaljoints. If unmanaged, this ARDM could c7entually result in the loss of pressure-retaining capability under CLB design loading l

conditions, The corrosive effects of boric acid leakage will be m.;naged by iaeans of the BACI Program.

e When boric ac!d leakage is identified, either through required program inspections or through irs resulting from other types of walkdowns and inspections, 'his program will ensure that corrosion induced by boric acid is discovered and that appropriate corrective action is taken.

Therefore, there is a reasonable assurance that the effects of general corrosion will be adequately managed fc external rurfaces of SI System components such that they will be capable of performing their intended functicns consistent with the CLB during the period of extended operation under all design loading coMitions.

Group 2 -(genera; corrosion, avice corrosion, and/or pitting ofinternal surfaces)- Materials and Environment Group 2 comprises the various SI System components that are xposed to che.nically-treated water and whose internal surfaces ar subject tc general corrosion, crevice corrosion, and/or pitting. The components in this group are included in all St System device types. The internals for the SI header CKVs, the loe, inlet CKVs, the SIT nutlet CKVs, and the SDC header return isolation MOV outside containment function to reduce the possibility of inter-system leakage when none of the associated system flowpaths allowing transfer of borated water to/from the RCS is active. [ Reference 49]

Likewise, the internals for the LPSI header MOVs, the main HPSI header MOVs, and the auxiliary HPSI header MOVs act as pressure-retaining boundari s for the Containment when their associated injection e

flowpaths are not actite. [ Reference 1, Attachment 2s for MOVs] The seating surfaces for the Si header l

Application for License Renewal 5.15-18 Calvert Cliffs Nuclear Power Plant s,

a ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM CKVs, the SDC header return isolation MOV outside containment, and the Si leakoff return isolation I

HVs also act as pressure-retaining ooundaries for the Containment. [ Reference 1, Attachment 3s for CKVs, MOVs, and HVs] The remaining components provide the passive intended function of

(

maintaining the system pressure boundary. [ Reference 1, Attachment 1] The applicable subcomponents in these device types are constructed of the following materials: [ Reference 1, Attachments 4 and 5 for all device types; Reference 50]

Piping - internal surfaces of stainless steel fittings and flanges, as well as hidden surfaces of e

carbon steel nuts and alloy steel studs; also, internal surfaces of stainless steel pipe and welds for smsll bore piping; CKVs - internal surfaces of stainless steel body / bonnet, as well as hidden surfaccs of carbon steel o

nuts and alloy steel studs; also, various cobalt based, nickel based, or iron-based alloy facing materials associated with t' e internals for the Si header CKVs; and stainless Jeet/ stellite internals for the loop inlet and SIT outlet CKVs; CVs - internal surfaces of stainless steel body / bonnet and r;em as well as hidden surfaces of carbon steel nuts and alloy steel studs; FEs - stainless steel; fos - stainless steel orifice plates; HW - internal surfaces of stainless s' eel body / bonnet and stem, as well as hidden surfaces of e

carbon steel nuts and alloy steel studs, for all HVs except the Unit 1 RWT return from SFP cooling HV; also, stainless steel disc / seat for those HVs that are normally closed; HVs - internal surfaces of the rebber i.ned carban steel body and stainiess steel stem for the Unit e

i RWT return from SFP cooling HV; HXs - hidden surfaces of carbon steel nute and alloy steel stud _ for all HXs; stainless steel tubing e

and weld connection, brass flex connector, and internal surfues of the cast iron case / cover for the HPSI pump seal coolers and L?SI pump seal HXs; stainless steel bonnets / covers / tubes and welds for the RWTHXs; MOVs - stainles steel wedge and seat rinF or the SDC header return isolation MOV outside f

a containment; stair.less steel / stellite disc / seat for the containment sump discharge MOVs; interr.at surfaces of stainless steel body / bonnet and st.m, as well a hidden surfaces of carbon seel nuts and alloy steel studs, for all MOVs; PUMPS - internal surfaces of the mechanical seals (nickel binder for tungsten carbide wearing surfaces; some with stainless steel seal ring body), stainless steel casing, and shaft, as well as hidden surfaces of ce%n steel nuts and alloy steel studs for the HPSI and LPSI pumps; PUMPS - internal surfaces of stainless steel casing and shaft and hidden surfaces of stainless steel nuts and studs for the RWT circulating pumps; RVs - Inconel or stenlite disc, stainless stec...pindle and spring / washer, and internal surfaces of a

stainless steel base / cylinder / adjusting bolt for pump and injection header RVs; RVs - stainless steel nozzle / disc, spindle point / adjusting bolt, spring / washer, hidden surfaces of e

carbon steel nuts and alloy steel studs, and internal surfaces of stainless steel body / bonnet for SI leakoff and SDC return header RVs outside conteinment; 4

Application for License Renewal 5.15-19 Calvert Cliffs Nucles-Power Plant

)

w ATTACHMENT m APPENDIX A - TECHNICAL INFtRMATION 5.15 - SAFETY INJECTION SYSTEM RVs - stellite disc, stainless steel spindle, adjusting bolt, spring / washer, and internal surfaces of e

stainless steel base / cylinder for SDC return header RVs inside containment; TEs - stainlass stee!;

a Tis - stainless steel well; TKs - internal surfaces of stainless steel shell, manhole, penetrations, and welds, as well as hidden surfaces of carbon steel nuts and alloy steel studs for the RWT; and TKs - internal surfaces of carbon steel shell (clad with stainless steel) and stainless steel manway for the SITS, as well as hidden surfaces of the associated stainless steel nuts and studs.

Except as noted below, the internal surfaces fer all components evaluated in Group 2 are exposed to the borated water environment described in subsection Group 1 - Materials and Environment, above. For Si

)

System IIXs, the inside of the shell and the outside of the tubes containing the borated water are exposed to the following fluids:

For the HPSI pump seal coolers and LPSI pump seal HXs, chemically-treated water from the CC C

System, with a design pressure of 150 psig and maximum operational temperature of 167'F.

[ Reference 1, Attachment 3s for HXs; Reference 51, Piping Class HB-23]

For the RWTHXs, untreated well water from the Plant Heating System, with a design pressure of 150 psig and maximum operational temperature of 200 F. [ Reference 1, Attachment 3s for HXs; Refermee 51, Piping Class HB 29]

Since the SI System is maintained in a standby mode during normal operations, stagnant internal conditions result in overall component temperatures close to ambient. These conditions may also allow impurities in the process fluid to concentrate. [ Reference 1, Attachment 6s for all device types]

Group 2 - (general corrosion, crevice corrosion, and/or pitting of internal surfaces) - Aging Mechanism Effects General corrosion is described in subsection Group 1 - Aging Mechanism Effects, above. Crevice corrosion and pitting are related forms of intensive, localized corrosion. Crevice corrosion occurs in crevices that are wide enough to permit liquid entry ard narrow enough to maintain stagnant conditions.

Such locations may include spaces under nuts and/or bolt heads, holes, gasket surfaces, lap joints, surface deposits, designed crevices for atiching thermal sleeves to sa#e-ends, and integral weld backing rings or back-up bars. Pitting occurs when corrosion proceeds at one small location at a rate greater than the ccrrosion rate of the surrounding area. Pitting is an autocatalytic process that produces conditions that stimulate the contiming.a/ivity of the pit. In either case, the stagnant fluid within the pit or crevice tends to accumulate con - P chemicals such as chlorides and sulfates, and thereby to accelerate the local corrosion psocess. wrevice corrosion can initiate pitting in many cases. P? ting can result in complete perforation of the material. [ Reference 1, Attachment 7s for all device types)

Crevice corrosion and pitting are plausible for all subcomponen'.s in this group. Additionally, general corrosica is plausible for the internal surfaces of all carbon steel and cast iron tubcomponents in this group. For the Unit 1 RWT return from SFP cooling HV, long-term exposure to b rated water can resu!t in permeation and cracking of the rubber liner designed to protect internal surfaes of its carbon steel Application for License Rene val 5.15-20 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM valve body from corrosion effects. These ARDMs are plausible at mechanical joints (e.g., flanges, body / bonnet joints, svelded joints in smalRore piping) since they present a crevice geometry at the sealing surface-that may allow process fluids to stagnate and cause concentration of environmentally-produced impurities. [ Reference 1, Attachment 6s for all device types] Similar stagnation and impurity deposits are possible at other component mterior crevices *at are formed by : lose-fitting interface points at interior suScomponents (e.g., %s/tubesheets in HXs, fittings in piping, pump shafts, valve stems, valve seating surfaces). [ Reference 1, Attachment 6s for -CC Piping, -DC Piping, -GC Piping, -HC Piping, CKVs, CVs, HVs, HXs, MOVs, PUMPS, RVs, TEs, Tis, and TKs]

Group 2 - (general corrosion, crevice corrosion, and/or pitting of internal surfaces) - Methods to Manage Aging hiitigation. Control of fluid chemistry in the SI System and interfacing systems can significantly limit 4

the effects of general corrosion, crevice corrosion, and pitting. [ Reference 1, Attachment 6s for all device types] The chemistry control program should nmitor pertinent chemical pr.rameters on a frequency that would allow for corrective actiou to mini..

creation of an environment conducive to corrosion.

Discoverv: The effects of corrosion are generally detectable by visual techniques. During repairs related to leakage in the loop inlet CKVs, visual methe is have identified minor pitting on valve discs and seats.

Lapping of the valve seats and replacement of valve discs has been used to restore the valves. Seating surface degradation can be discovered by testing the components that are suscept:ble to this ARDM.

Degradation of CKV seatir surfaces can be discovered by monitoring fe. system leakage and performing leak rate testing. Pressure testing of the SDC header return isolation MOV outside ccatainment and the Si leakoff return isolation HVs can provide for detectian of leakage that could be the result of crevice corrosion and pitting of the valve seating surfaces. Internal surfaces of components that are not routinely inspected can be subjected to inspection to determine the exteia of general and/or localized degradation that may be occurring. [ Reference I, Attachment 6s for all device types]

Group 2 - (general corrosion, crevice corrosion, and/or pitting of internal surfaces) - Aging Management "rogram(s)

Mitigation: Maintenance of proper fluid chemistry in the SI System and interfacing systems will limit the effects of general corrosion, crevice corrosion, and pitting on internal surfaces for Group 2 suMomponents. [ Reference 1, Attachment 8]

The CCNPP Chemistry Program has been established tc. minimize impurity ingress to plant systems; reduce corrosion product generation, transport, and deposition; reduce collective radiation exposure through chemistry; improve integrity and availability of plant systems; and extend component and plant life.

[ Reference 52, Section 6.1.A]

The program is based on Technical Specification 3, BGE's interpretation of industry standards, and recommendations made by Combustion Engineering.

[Re erence 53, Section 2.0]

r Cr. vert Cliffs Technical Procedure CP-204," Specification and Surveillance-Primary Systems," provides

'or monitoring and maintaining chemistry 'm the RCS and associated systems.

[ Reference 53, Application for License Renewal 5.15-21 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAI ETY INJECTION SYSTEM Section 2.0, Attachments 1 through 15] Control of primary water chemistry is credited with limiting the effects of crevice corrosion and pitting in SI System components. [ Reference 1, Attachment 8]

Calvert Cliffs Technical Procedure CP 206, " Specifications and Surveillance-Component Cooling / Service Water Systems," provides for monitoring or CC System chemistry to control tne con +entrations of oxygen, chlorides, and other chemicals and contaminants. [ Reference 54, Section 2.0, ] Control of the water chemistry provides an environment that limits the rate of degradation and its effects for the LPSI pump seal IIXs and ilPSI pump seal coolers, which are cooled by water from the CC System. [ Reference 1, Attachment 8]

Calvert Cliffs Technical Procedure CP-202, " Specifications and Surveillances-Demit.eralized Water, Safety Related Battery Water, & Well Water Systems," provides for monitoring of pli levels in well water. [ Reference 55, Section 2.0, Attachment 6] Observing that significant changes in well water chemistry have not occurred is an effective method to ensure that the environment on the shell side of the RWTilXs, which are heated by well water ftom the Plant lleating System, has not been altered to the extent that the rate of degradation is affected. [ Reference 1, Attachment 8]

Each of the program procedures describes the surveillance and specifications for monitoring fluid chemistry for the applicable systems. They list the parameters to be monitored, the frequency for monitoring of each parameter, and he acceptable value or range of vslues for each parameter.

[Referenc,: 53, Attachments 1 through 15; Reference 54, Attachment 1; Reference 55, Attachment 6]

Fach parameter is measured at a procedurally-specified frequency (e.g., daily, weekly, monthly) and compared against a target value that represents a goal or predetermined warning limit. [ Reference 53, Section 3.0; Reference 54, Section 3.0; Reference 55, Section 3.0] If a measured value is outside of its required range, corrective actions are taken (e.g., power reduction, plant shutdown) as prescribed by the procedure, thereby nsuring timely response to chemical excursions. The procedures provide for rapid assessment of off-normal chemistry parameters so that steps can be taken to return them to normal levels.

[ Reference 53, Section 6.0.C; Reference 54, Section 6.0.C; Reference 55, Section 6.0.C]

The CCNPP Chemistry Program has been subject to periodic interr Al assessment activities. Internal audits are performed to ensure that activities and procedures estaolished to implement the requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program. These audits provide a comprehensive independent verification and evaluatinn of quality-related activities and procedures. Audits of selected aspects of operational phase acuvities are performed with a frequency commensurate with their leul of performance and safety significance and in such a manncr as to assure that an audit of all safety-related functions is completed within a period of two years. [ Reference 56, Section 1B.18] These activities, as well as other external assessments, help to maintain highly effective chemistry control, and facilhate continuous improvement through monitoring industry ininatives and trends in the area of corrosion control.

A review of operating experience identified no site-specific problems or ever.ts related to general corrosion, crevice corrosion, or pitting that required significant changes or adjustments to the CCNPP Chemistry Program. It has been effective in its function of mitigating corrosion and controlling corros'on-related failures and pro'olems within acceptable limits. In 1996, CP-206 was revised to include monitoring of dissolved iron as a method for dis,:overing any unusual corrosion of carbon steel components. Self-assessments of chemistry control performance have resulted in activities to reluce the Application for License Renewal 5.15-22 Calvert Cliffs Nuclear Power Plant

ATTACHMENT ($)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM number of times that chemistry parameters exceed action levels (e.g., additional scheduling coordination for outage evolutions that could afket CC/ Service Water chemical parameters).

Discoverv:

General corrosion, cruice corrosion, and pitting of internal surfaces for Group 2 components are managed through a combination of monitoring, testing, and inspection activities at CCNI'P.

The loop inlet CKVs, SIT outlet CKVs, and SI header CKVs require leak rate testing under the CCNPP Pump and Valve Inservice Testing (IST) Program. [ Reference 57] This program was estabbshed to implemem IST in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR 50.55a(f).Section XI, Subsection IWV, directs each licensee to comply with the applicable ptions of ASME/ ANSI OM-10. American Society of Mechanical Engineers / ANSI OM 10 provides the rules and requirements for IST of CCNPP valves, including the types of tests required, frequency of testing, test methods, test pressun' acceptance criteria, and reporting requirements. In addition to the general Code requirements discussed above, there are additional interpretations and positions that have come about as a result of past regulatory and licensee acticns, including NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants." [ References 57, 58, and 59]

Pressure in the piping between the loop inlet CKVs and the SIT outlet CKVs is continuously monitored by instrumentcion, with an indication in the Control Room. [ References 32 and 34]

Monitoring these indications, which demonstrate functionally adequate seat tightness of the loop inlet CKVs on an ongoing basis in lieu of leak rate testing, is part of the overall CCNPP Pump and Valve IST Program. [ Reference 57, Attachment 1 - Section 5.4 and Relief Request No. VR-08] Excessive backleakage thNgh a loop inlet CKV would result in alarm actuation and assessment of the leakage. To ensure se:h leakage frc.m the RCS will be detected, observation of alarm function is documented as part of the SIT outlet CKV closure verification described below.

[ References 60 and 61, Section 6.5] If unidentified RCS leakage exceeds the acceptance criteric provided in the CCNPP Technical Specifications, the applicable abnormal operations procedures are implemented. Appropriate corrective actions are determined in accordance with the CCNPP Technical Specifications, Surveill nce Test Program procedures, and the CCleP Corrective Actions Program. [ References 59,62, and 63] Historically, this method has been effective in identifying the sources of RCS leakage.

Calvert Cliffs proceduies STP O-65J-1(2), which verify the closure and seat leakage integrity of the SlT outlet CKVs and the Si header CKVs, are also part of the overall CCNPP Pump and Valve IST Progrem. [Rcferences 60 and 61] Testing is implemented by CCNPP Technical Specification 4.0.5. [ References 58 and 64] The SIT outlet CKVs and the SI header CKVs are tested in sccordance with ASME/ ANSI OM-1987, including OMa-1988 Addenda. Leak testing of these valves is required every two years in accordance with the Pump and Valve IST Program -

Third Ten-Year Interval. [ Reference 57] Completion of the SIT outlet CKV and SI header CKV closure verification in accordance with STP O-65J-1(2) satisfies the biennial seat leakage measurement requirement. [ References 60 and 61, Section 2.0.B] Foi the SIT outlet CKVs, this verification presently includes the following procedural steps:

[ References 60 and 61, Section 6.5]

A pp*: cation for License Renewal 5.15-23 Calvert Clifts Nuclear Power Plant

l ATTACHM_ENT (M j

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Pressure is applied to the CKV in the reverse flow direction (i.e., the CKV is seated), and the instrumentation that monitors pressure between the loop inlet CKVs and the SlT outlet CKVs is checked for proper indication and alarm function.

Level in the associated SIT is monitored, and leakage is quuntified by recording the change in level during the period of the test (minimum duration of 20 minutes).

For the Si header CKVs, closure verification presently includes the following procedural steps:

[ References 60 and 61, Section 6.6]

Test equipment is connected to the appropriate test point.

Pressure is applied to the CKV in the reverse flow direction (i.e., the CKV is seated),

if a pressure increase on the upstream (Iow pressure) side of the valve indicates CKV leakage, leakage is quantified by: (a) establishing a reference pressure; and (b) recording the volume and duration ofleakage collected in a suitable container on the upstream side of the valve.

In both cases, the measured leak rate is corrected for test pressure and compared against acceptance criteria for each individual valve and a cumulative CKV leakage limit. If the corrected leakage is less than or equal to the acceptance criteria. the test is satisfactory. If not.. the affected equipment is declared inoperable and appropriate corrective actions are determined in s

accordance with CCNPP IST Program procedcres and the CCNPP Corrective Ac+ ions Program.

Calvert Cliffs procedures STP M-571G-1(2) and M-571L-1(2), which cover local leak rate testing (LLRT) for the Si leakoff return isolation HVs and the SDC header return isolation MOV outside containment, respectively, are part of the overall CCNPP Containment Leakage Rate Testing Program. [ References 65 through 63] The CCNPP Containment Leakage Rate Testing Program was established to implement the leakage testing of the Containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J specifies containment leakage testing requirements, including the types of tests required, frequenc/ of testing, test methods, test pressures, acceptance criteria, and reporting requirements.

Contai.iment leakage testing requirements include performance ofIntegrated Leakage Rate Tests, also known as Type A tests, and LLRTs, also known as Type B and C tests. Type A tests measure the overall leakage rate of the Containment. Type B tests are intended to detect leakage paths and measure leakage for certain containment penetrations (e.g., airlocks, flanges, and electrical penetrations). Type C tests are intended to measure containment isolatior: valve leakage rates. [ Reference 64, Section 6.5.6; References 69 and 70]

The CCNPP LLRT Program is based on the requi;ements of CCNPP Technical Specifications 3.6.1.2,4.6.1.2, and 6.5.6.

The scope of the program includes Type B and C testing of containment penetrations. The valves that isolate the containment penetration piping for the SDC return header and the SI leakoff return header are included in the scope of this program as part of the leakage testing for the associated containment penetrations.

[ Reference 64]

Application for License Renewal 5.15-24 Calvert Cliffs Nuclear Fower Plant

4 ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM The LLRT is done on a performance-based testing schedule in accordance with Option B of 10 CFR Part 50, Appendix J, as imple:nented by CCNPP Technical Specifications.

[ References 64,69, and 70) Local leak rate testing presently includes the following procedural steps:

Leak rate monitoring test equipment is connected to the appropriate test point.

The test volume is pressurized to the LLRT Program test pressure, which is conservative with respect to the 10 CFR Part 50, Appendix J, test pressure requirements. Appendix J requires testing at a pressure "P.," which is the peak calculated containment internal pressure related to the Design Basis Accident.

Leak rate, pressure, and temperature are monitored at the frequency specified by the LLRT procedure and the results are recorded.

The maximum indicated leak rate is compared against administrative limits that are more restrictive than the maximum allowable leakage limits.

"As found" leakage equal to or greater than the administrative limit, but less thaa the maimum allowable limit, is evaluated to determine if further testing is required and/or if cormtive maintenance is to be performed.

For "as found" leakage that exceeds the maximum allowable limit, plant personnel determine if Technical Specification Limiting Condition for Operation 3.6.1.2.b has been exceeded.

Techn: cal Specification 3.6.1.2.b contains the maximum allowable combined leakage for all penetrations and valves subject to the Type B and C tests. Corrective action is taken as required to restore the leakage rates to within the appropriate acceptance criteria.

If any maintenance is required on a containment isolation valve that changes the closing characteristic of the valve, an "as left" test must be perforn.ed on the penetration to ensure leakage rates are acceptable.

Baltimore Gas and Electric Company will include all Group 2 components in an Age-Related Degradation Inspection (ARDI) Program to verify that unacceptable degradation of internal surfaces is not occurring, thereby validating the effectiveness of the CCNPP Chemistry Program in mitigating the effects of general corrosion, crevice corrosion, or pitting. [ Reference 1, ]

The ARDI Program is defined in the CCNPP IPA Methodology presented in Section 2.0 of the BGE LRA.

The elemems of the ARDI Program will include:

Determination of the examination sample size based on plausible aging effects; Identification ofinspection locations in the system / component based on plausible aging effects and consequences of loss of component intended function; Determination of examinstion techniques (including acceptance criteria) that would be effective, considering ths ging ef'ects for which the component is examined; Methods for interpretation of examination results; l

Application for License Renewal 5.15-25 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 1

l Methods for resolution of adverse ext.mination findings, including consideration of all design loading conditions required by the CLB and specification of required corrective actions; and Evaluation of the need for follow up examinations to monitor the progression of any age-related degradation.

Any corrective actions that are required by the Pump and Valve IST Program, the LLRT Program, or the ARDI Program, will be taken in accordance with the CCNPF Corrective Actions Program and will ensure that the components will remain capable of performing their intended function undca all CLB conditions.

Group 2 - (general corrosion, crevice corrosion, and/or pitting of internal surfaces) -

1)emonstration of Aging Management Based on the information prese:'ed above, the following conclusions can be reached with respect to general corrosion, crevice corrosion, and pitting of internal surfaces for SI System components that are exposed to chemically-treated water:

The components in Group 2 contribute to maintaining the system pessure boundary.

Additionally, certain CKVs and MOVs function to reduce the possibility of inter system leakage when none of the associated flowpaths allowing trr.asfer of bcrated water to/from the RCS is active; some CKVs, MOV, and HVs also act p essure-retaining boundaries for the Containment. The integrity of these components must be maintained under all CLB design conditions.

The materials of construction for subcomponents in ti a group include cast iron, carbon steel, alloy steel, stainicss steel, and brass, as well as varior., facing matericts.

General corrosion, crevice corrosion, and pitting are plausible ARDMs for this group and, if unmanaged, these ARDMs could eventually result in the loss of pressure-retaining capability under CLB design loading conditions.

Maintenance of proper fluid chemistry in the SI System (in accordance with CP-204) will limit the effects of general corrosion, crevice corrosion, and pitting on susceptible pressure boundary suhomponents for Group 2 components. Controlling the chemistry of water in the CC System (in accordance with CP-206) and monitormg chemistry of the well water supplying the Plant Heating System (in accordance with CP-202) will ensure that the water supplied to SI System HXs is of an appropriate chemistry to minimize corrosion.

The CCNPP Pump and Valve IST Program incorporates monitoring pressure in fications associated with the 1%p inlet CKVs. This activity, combined with assessments performed in the event of alarm actuation, detects leakage that could result from crevice corrosion and pitting on the seating so-faces of the loop inlet CKVs. Corrective actions to address abnormal leakage are directed by Technical Specifications.

The CCNPP Pump and Valve IST Program performs ;eak testing of the SIT outlet CKVs and the SI header CKVs; the CCNPP LLRT Program perfonns leakage testing of the SDC header return isolation MOV outside containment and the SI leakoff return isolation HVs. These programs are credited with detecting leakage that could result from crevice corrosion and pitting on the seating Application for License Renewal 5.15-26 Calvert Cliffs Nuclear Power Plant

8 ATTACHMENT @

7 APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM sm-faces of the listed valves. These programs ensure that appropriate corrective actions will be taken if significant leakage is discovered.

All Group 2 compor.ents will be subjected tu a new ARDI Program. This program will examine a representative sample of the components for degradation, and ensure that appropriate corrective actions are initiated on the basis of the findings.

Therefore, there is a reasonable assurance that the -ffects of general corrosion, crevice corrosion, and pitting will be adequately managed for internal surfaces of SI System components exposed to chemically treated water such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under t.Il design loading conditions.

Group 3 - (MIC of internal surfaces)- Materials and Environment Group 3 consists of the two recirculation headers in the SI System connected to the emergency sump inside containraent wnose internals are subject to MIC. This stainless steel piping provides the passive intended function of maintaining the system pressure boundary. [ Reference 1, Attachments 4 and 5 for

-liC Piping]

This section of piping is filled from the RWT prior to starting up followmg a refueling outage; therefore, the internal surfaces of this piping are exposed to stagnant borated water in the emergency sump. Filling these lines provides a thermal barrier to prevent significant heatup of the containment sump discharge MOV bonnets caused by fluid antering the emergency sump inside containment after 3 DBE and prior to a RAS. [ Reference 3, Section 6.16; Reference 71) Since this body of weter is left open to the atmosphere inside containment for extended periods of time (i.e.,the refueling cycle), it is capable of supporting microbes introduced to the system.

[ Reference 1, Attachmer/ 6 for -11C Piping; Reference 72] Refer to subsection Group 1 - Materials and Environment, above, for discussion of the atmosphere inside containment.

Group 3 -(MIC ofinternal surfaces)- Aging Mechanism Effects Microbiologically-induced corrosion is accelerated corrosion of materials resulting from surface microbiological activity. Sulfate-reducing bacteria, sulfur oxidizers, and iron-oxidizing bacteria are most commonly associated with these corrosion effects. Microbiologically-induced corrosion most often results in pitting, followed by excessive deposition of corrosion products. Essentially any system that uses untreated wa:er and most commonly-used materials are susceptible to MIC. [ Reference 1, for Piping] Long-term exposure of the recirculation header piping to microbe activity may result in localized material loss and, if left unmanar d, could eventually result in :oss of pressure-retaining capability under CLB design loading conditions. [ Reference 1, Attachment 6 -lIC Piping]

Group 3 -(MIC ofinternal surfaces)- Methods to Manage Aging Mitigation: Industry studies indict.te that MIC can be prevented through adding biocides to the system fluid, increasing fluid temperature above 200*F, and/or using flow to eliminate longterm stagnation.

[ Reference 72]

t Application for License Renewal 5.15-27 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Discoverv: The effects of corrosion are generally detectable by visual techniques. [ Reference 72]

Internal surfaces of the recirculation headers can be subjected to inspection to determine the extent of localized degradation that may be occurring. [ Reference 1, Atuchment 6 for -HC Piping]

Group 3 -(MIC ofinternal surfaces)- Aring Management Program (s)

Mitigation: Maintaining the thermal barrier precludes elimination of the stagnant body of water at ambient temperature from the recirculation headers, Since BGE does not ticst the process Guid for micrabes, there are no programs credited with mitigating the effects of MIC for the components in this group. [ Reference 1 Attachment 8; Reference 72]

Ditcoverv: 13altimore Gas and Electric Company will include the recirculation headers connected to the emergency surap nside containment in an ARDI Program to verify that unacceptable degradation of i

internal surfaces by MIC is not occurring. [ Reference 1, Attachment 8] For a discussion of the elements of the ARDI Program, refer to subsection Group 2 - ;ing Management Programs, above.

Group 3 -(MIC ofinternal surface)- Demonstration of Aging Management x

Based on the information presented above, the following conclusions can be reached with respect to mig ofinternal surfaces for the SI System recirculation header.;:

The recirculation headers connected to the emergency sump inside containment contrib~; to e

maintaining the system pressure boundary, and their integrity must be maintained under all CLB design conditions.

This group cons:sts of stainless steel piping.

Microbiologically-induceo corrosion is a plausible ARDM bccause the stagnant water in this piping is exposed to the containment atmosphere for extended periods of time. If unmanaged, this ARDM could eventually result in the loss of pressure-retaining capability under CLB design leading conditions.

The recirculation header piping will be subjected to a new ARDI Program. This program will examine a representative sample of the components for degradation, and ensure that appropriate corrective actions are initiated on the basis of the findings.

Therefore, there is a reasonable assurance that the effects of MIC will be adequately managed for internal surfaces of the SI System recirculation headers such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under all design loading conditions.

i Appiication for License Renewal 5.15-28 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Group 4 - (fatigue for piping and valves exposed to thermal transients) - Materials and Environment 2

Group 4 comprises SI System components in the SIT injection and SDC mode flowpaths for which fatigue is a plausible ARDM. Specifically, this group consists of the following SI System components that are depicted in Figure 5.15-1:

Si header CKVs (l[2]CKN SI-il8, -128, -138, -148), SIT outlet CKVs (l[2]CKVSI-215, -225,

-233, -245), SIT outlet MOVs (l[2]MOV614, 624, 634, 644), loop inlet CKVs (l[2]CKVSI-217,

-227,-237,-247), and the intervening pipe segments (1#CC4-1001, -1002, -1003, -1004, -1009,

-1010, -1011, -1012 (2#CC4-2001, -2002, -2003, -2004, -2009, -2010, -2011, -2012);

l#CCl3-1009 -1010,-10'.1,-1012 [2#CC13-2009,-2010, -2011,-2012]);

SIT CKV leakage CVs (l[2]CVSI-618, -628, -638, -648), SIT CKV leakage isolation HVs (l[2]HVSI-699, -700, -701, -702), and the associated SlT recirculation piping connecting to the Si header (1 #CC4-1013, -1014, -1015, -1016 [2#CC4-2013. -2014, -2015, -2016]); and SDC return header piping from the RCS interface at the outlet of the SDC header return isolation MOV inside cantainment (1#CC14-1004 [2#CCl4 2004]), up to and including the SDC header return isolation MOV outside containment (l[2]MOV651).

All of these components provide the passive intended function of maintaining the system pressure boundary. [ Reference 1, Attachmem 1] Additionally, the SI header CKVs and the SDC header return isolation MOV outside containment are required to maintain both the reactor coolant pressure boundary and the containment pressure boundtry when none of the associated system flowpaths allowing transfer of borated water to/from the RCS is active.

[ Reference 1, Attachment 3s for CKVs, MOVs; Reference 49] The pressure-retaining subcoraponents in these device types are constructed of carbon steel, alloy steel, and stainless steel, es well as adaus facing materials, as described in subsections Group 1 and Group 2 - Materials and Environment, above. [ Reference 1, Attachments 4 and 5 for pipe, CKVs, CVs, HVs, MOVs; Reference 50]

Except fo piping between the SIT outlet CKVs and the SIT outlet MOVs, the original design code for the piping in this group is ANSI B31.l Class I. [ Reference 39, Piping Class CC-4; Reference 40, Piping Class CC-14] Components in this piping, which comprise the Class I portion of the SDC mode flowpath, were designed to withstand up to 500 SDC initiation transients during the anticipated life of the plant, consisting of rapid temperature rises from ambient (70 F) to 300 F. The piping between the SIT outlet 1

CKVs ans the SIT outlet MOVs was originally designed in accordance with ANSI B31.7 ClassIl requirements, but was subsequently upgraded to Class I requirements.

[ Reference 40, Piping Class CC-13]

Additional thermal stresses are imposed on the piping ard valves between the SIT outlet CKVs and the loop inlet CKVs due to thermal stratification effects. Stratificadon in the Si line has been shown to exist, and was found to be dependent primarily on the temperature difference between the RCS cold leg and the ambient temperature. These pipe segments have relatively hot fluid on the RCS side of the closed loop into CKVs and relatively cold fluid on the SIT side of the closed SIT outlet CKVs. Natural convection caused by the presence of these hot and cold b mdaries produces thermal stratification, evidenced by observed top-to-bottom wall temperature differences of up to 145 F in this piping. [ References 73]

l Application for License Renewal 5.15-29 Calvert Cliffs Nuclear Power Plant n

(

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Since the SI System vent / drain / test ilVs, instrument isolation }{Vs, and RVs connected to the piping in Group 4 are generally thin wall components, they do not experience the large temperatr, grr.dients that would be necessary to cause significant degradation. Likewise, the normal and design operating conditions applied to SI System corrponents that are not associated with the above Dowpaths result in neither the quantity of cycles nor the loading conditions (mechanical, vibrational, thermal, and/or pressure) necessary to cause significant degradation. Herefore, except for the components comprising Group 4, fatigue is not plausible for the SI System.

The internal surfaces for components in this group are exposed to the borated water environment described in subsection Group 1 - Materials and Environment, above. [Refereace 1, Attachment 6s for CC Piping, CKVs, CVs, liVs, MOVs; Reference 39, Piping Class CC-4; Reference 40, Piping Class CC-14] The external environment is climate-controlled air in the Containment. [ Reference 1, s] Since the SI System is maintained in a standby mode during normal operations, component temperatures range from ambient to approximately 300 F under the inuuence of the natural circulation Onw in the piping between the closed SIT outlet and loop inlet CKVs. [ Reference 73]

During reactor cooldown, the Group 4 components subjected to thermal stratification cool to temperatures approaching ambient. Initiation of SDC also causes a rapid temperature transient for other components in the SDC imode Downath from ambient (about 70 F) to RCS temperature (no greater than 300 F). [ Reference 2, Section 9.2.4.1 Greep 4 -(fatigue for piping and valves exposed to thermal transients)- Aging Mechanism Effects Fatigue is the process of geogressive localized structural change occurring in a material subjected to conditions that produce Cuctuating stresses and strains at some point or points in the material. This process may culminate in cracks or complete fracture aner a sufficient number of fluctuations. The fatigue life of a component is the number of cycles of stress or strain that it experiences before fatigue failure occurs. Failures may occur at either a high e low number of cycles in response M various kinds of loads (e.g., mechanical or vibrational loads, thermal cycles, or pressure cycles). Low-cycle fatigue involves stressing of rnaterials, often into the plastic range, with the number of cycles usually being less 5

than 10. This nicchanism is typically associated with thermal gradients created in thick sections (e.g.,> 1") or in restrained members during rapid heatup or cooldown. A component subjected to sufficient cycling with significant strain accumulates fatigue damage, which potentially can lead to crack initiation rnd crack growth. The cracks may then propagate under continuing cyclic stresses. For plant equipment operating in a corasive environment, growth of fatigue cracks may be subsequently dominated by co rosion advance. Such environmental effects must be considered during system design.

[ Reference 1, Attachment 7s foi %pe, valves; Reference 74]

Low-cycle thermal fatigue is plausible for the devices in this group since they experience cyclical thermal loading and pressurization ttut contribute to fatigue accumulation. [ Reference 1, Attachment 6s for pipe, CKVs, CVs, HVs, MOVs; Reference 73] The limiting locations for these transients are in the RCS piping (i.e., the Si nozzles and the SDC outlet nozzle). [Ref:rence 75, Section 3.3.2 and Table 5-1]

This aging mechanism, if unmanaged, could eventually result in crack initiation and growth such that the Group 4 components may not be able to perform their pressure boundary function under CLB design foading conditions.

Application for License Renewal 5.15-30 Calvert Cliffs Nuclear Power Plant i

9

ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Group 4 -(fatigue for piping and valves exposed to thermal transients)- Methods to Manage Aging Mitigation: 'Ihe effects of low-cycle fatigue can be mitigated by proper system design and material selection, and by operational pactices that reduce the number and severity of thermal traasients on the susceptible components. [ Reference 1, Attachment 5s for pipe, CKVs, CVs, HVs, MOVs]

Discoverv: Fatigue cracks can be discovered by inspecting components; the scope and frequency of inspections can be established based on the likelihood that fatigue cracks have initiated. As discussed above, low cycle fatigue was addressed in the original design for the components in the SDC mode flowpath by determination of an allowable number of full-range thermal cycles for the anticipated life of the plant. The accumulation of fatigue effects on these components can be monitored by counting the number of thermal transients and by performing analysis to predict the remaining life of the affected components.

Group 4 - (fatigue for piping and valves exposed to thermal transients) - Aging Management Program (s)

Mitigation: As part of general operating practice, plant operators minimize the duration and severity of transitory operational cycles. Further modification of plant operating practices to reduce the :nagnitude and/or frequency of thermal transients would unnecessarily place additional restrictions on plant operations.

P ury: The CCNPP Fatigue Monitoring Program (FMP) has been established to monitor and track

_ pe usage of limiting components of the Nuclear Steam Supply System, including the SI System, and the steam generators. Reference 75 was used in the development of this program. Eleven fatigue critical locations in these systems have been selected for monitoring of fatigue usage. These represent the most bounding locations for critical thermal transients. For components in the SI System, fatigue usage is bounded by the fatigue usage of the Si nozzles and the SDC outlet nozzle in the RCS. [ Reference 76, Sections 1.1,1.2.A,2.1.E,6.0 The FMP utilizes two methods to track fatigue usage:

One method is to track the number of critical thermal and pressure test transients (i.e., cycle counting) and compare them to the number allowed in the piping design analysis. The piping design analysis is performed assuming a particular number and severity of various transieras. In accordance with either ASME Section 111 or ANSI B31.7, the analysis demonstrates that the component has an acceptable design as long as the assumptions remain valid. Therefore, if the actual number and severi y of transients experienced by the component remains below the t

number assumed in the analysis, the component remains within its design basis.

The other method is to determine the fatigue life of a component using a calculated cumulative usage factor (CUF), which is defined as a normalized measure of total fatigue damage accumulated by a component as a result of all stress cycles that the component has experienced during its service life. The CUF can be calculated and tracked through plant life using thermal cycle counting or stress-based analysis techniques. In accordance with the ASME Boiler and Pressure Vesse; Code, the component remains within its design basis for allowable fatigue life if the CUF remains less than or equal to one rReference 76, Sections 1.2.A,3.0.B,3.0.F]

Application for License Renewal 5.15-31 Calvert Cliffs Nuclear Pcwer Plant

ATTACHMENT (5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYCEM Both methods use actual plant operating data. At CCNPP, the usage factor for several locations, including the si nozzles and the SDC outlet nozzle in the RCS, is calculated using thermal cycle counting. Since the FMP monitors ac'ual fatigue usage, a more realistic CUF is calculated. The data for thermal transients is collected, recorded, and analyzed using a computer program that evaluces input data from plant instrumentation. The computer software is used to analyze plant data associated with real trensients and to predict the number of thermal cycle transients for 40 and 60 years of plant operatica Lased on the historical records. For the S1 System, the allowable number ofinitiation of SDC cycles is 500. Based on actual occurrences to date, partial cycle analysis for the initiation of SDC transient predicts that Unit I will experience 30 effective full cycles for 40 years of plant operation and 45 effective full cycles for 60 years. Similarly, Unit 2 projections estimate 29 effective full cycles for 40 years and 43 effective full cycles for 60 years. [ Reference 76, Section 3.0.F]

Plant parameter data is collected on a periodic basis and reviewed to ensure that the data represents actual transients. Valid data is entered into the computer program that counts the critical transient cycles and calculates the CUFs. The data is tracked in accordance with procedures that are governed by a quality assurance progrun that meets 10 CFR Part 50, Appendix B, criteria. The transient data is evaluated and the CUFs are calculated on a semi-annual basis, which provides a readily predictable approach to the alert value. Acceptable conditions exist, since no crack initiation would be predicted, wh:n the calculated CUF for any given component is less than one, or when the design allowable number of cycles for the component has not been exceeded.- In order to stay within the design basis, corrective action is initiated well in advance of the CUF approaching cne or the number of cycles approaching the design allowable, so that appropriate corrective actions can be taken in a timely and coordinated manner.

[ Reference 76, Sections 1.2.A 5.0]

Since the FMP has been initiated, no locations have reached the limit on fatigue usage and no cracking due to low-cycle fatigue has been discovered. The FMP has undergone several modifications since its inception. Stress-based analysis was added to the computer software to calculate the CUFs for several locations due to unique thermal transients experienced and the unique geometries involved. Other modifications have been made to the FMP to rettect plant operating conditions more accurately. The plant design change process has also been modified to require noti 0 cation to the Life Cycle Management Unit of any proposed changes to the critical locations being monitored.

The CCNPP FMP has been inspected by the NRC, which noted that the program has been developed toward providing assurance that fatigue life usage of primary system components has not exceeded limits provided for in ASME Section Ill. In addition, the NRC noted that the FMP can be used to identify components where fatigue usage may challenge the remaining and extended life of the components and can provide a basis for corrective action where necessary. [ Reference 77]

Since its inception in 1988, BGE has participated in the extensive program undertaken by the Combustion Engineering Owners Group to address thermal stratificadan concerns. In response to NRC Bulletin 88-08," Thermal Stresses in Piping Connected to Reactor Coolant Systems," BGE identified the potential for thermal stratification in the piping between the SIT cettet CKVs and the loop inlet CKVs, and subsequently confirmed the natural convection phenomenon described in subsection Group 4 -

Materiak and Environment, above. (References 73 and 78] Since the current piping analysis for the affected portions of the SI System does not include the additional stresses imposed by thermal Application for License Renewal 5.15-32 Calvert Cliffs Nucles Power Plant i

ATTACHMENT (5)

APPENDlX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM stratification, BGE will complete an engineering review of the industry's task reports and determine: (a) any necessary changes to the piping analyses of record for the SI System, including the Group 4 pipe segments; and (b) the impact of such changes on fatigue usage parameters used by the FMP.

To further address fatigue for license renewal, CCNPP participated in a task, sponsored by the Electric i-Power Research Institute, to demonstrate the industry fatigue position.

The task applied industry-developed methodologies to identify fatigue-sensitive component locations that may require further evaluation or inspection for license renewal and evaluate environmental effects, as necessary.

The program objective included the development and justification of aging management practices for fatigue at various component locations for the renewal period. The demonstration systems were the Feedwater System, the pressurizer surge line in the RCS, and the letdown and charging subsystem in the CVCS. [ Reference 79, page 3]

Evaluation of Thermal Fatigue Effects to Address Generic Safety Issue 166:

Generic Safety issue 166, Adequacy of Fatigue Life of Metal Components, presents concerns identified by the NRC that must be evaluated as part of the !icense renewal process. The NRC staff concerns about j

fatigue for license renewal fall into the following five categories: { Reference 79, page 2: Refe ence 8v]

w The first category, adequacy of the fatigue design basis when environmemal effects a :

e considered, does not apply to the Group 4 components becaus: of stringent RCS water chemistry controls, exceptionally low oxygu concentrations (less than five parts per billion), and stainless steel materials used in fabrication of the affected piping and valve rubcomponents.

The second category concerns the adequacy of both the number and severity of design-basis transients. The engineering review addressing thermal strati'ication in the SIT injection mode flawpath, discussed above, will evaluate the inipact of this phenomenon on design basis transients considered in the piping analyses of secord for the SI System.

The third category, adequacy of inservice inspection requirements and procedures to detect fatigue indications, does not apply because CCNPP does not rely en inservice inspection as the sole means for detection of fatigue, The fourth category, adequacy of the fatigue design basis for Class I piping components designed a

in accordance with ANSI B31.1, does not apply because the piping in tnis group is designed in I

accordance with ANSI B31.7, Class I.

The final category, adequacy of actions to be taken when the fatigue design basis is potentially S-compromised, as discussed above, is adequately addressed by the CChPP FMP.

Group 4 -(fatigue for piping and valves exposed to thermal transients)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to fatigue for Group 4 components:

Piping and valves in the SIT injection and SDC mode flowpaths contribute to maintaining the r

reactor coolant pressure boundary, the containment pressure boundary, and/or the SI System pressure boundary. Their integrity must be maintained under all CLB design conditions.

L Apphcation for License Renewal 5.15-33 Calvert Cliffs Nuclear Power Plant

ATTACIIMFNT (5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM The materials of construction for subcomponents in this group are carbon steel, alloy steel, and stainless steel, as well as various facing materials.

Fatigue is a plausible ARDM for this group because the components are subject to severe thermal cycling during initiation of SDC andh RCS heatup/cooldown. If unmanaged, this ARDM could ever.tually result in crack initiation and growth such that the components may not be able to peri'orm their intended functions under CLB conditions.

The FMP monitors fatigue usage at bounding ocations to ensure that the Group 4 components i

remain within their design besis and includcs acceptance criteria to ensure timely corrective action is taken prior to degradation that would compromise the pressure boundary and containment isolation functions.

The results ofir.dustry studies to address the issue of thermal stratification in piping connected to the RCS will be reviewed, and changes in the des', analyses for piping in Group 4 will be made as necessary.

Therefore, there is a reasonable assurance taat the effects of fatigue will be adequately managed for susceptible components in the SI System such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under all design loading conditions.

Group 5 -(SCC near RWT penetrations and associated welds)- Materials and Environment Group 5 consists of heat-affected zones in the stainless steel base metal near the RWT penetrations and associated welds that are subject to SCC. [ Reference 1, Attachments 4 and 5] Nozzle penetrations for SI System piping connected to the RWT consist of stainless steel pipe pene+ rating the tank walljoined by r full penetration groove weld with a fillet cap. An additional reinfore plate or penetration seal plate is similarly welded to the outer diameter of the pipe and the taa wall, forming a narrow crevice

" Telltale" holes are drilled through these plates on the 1.orizontal centerline. [ Reference 81] The RWT provides the passive intended function of maintaining the system pressure boundary. [ Reference i ]

The internal environment at the RWT penetrations is borated water with normal operating parameters of up to 90 psig and 105 F. [ Reference 1, Attachment 6s for -HC Piping and TKs; Reference 42, Piping Class HC-23] The external surfaces of the RWT penetrations and welds are exposed to the normal outside atmosphere at the CCNPP site. [ Reference 1, Attachment 6 for RWTs] In the crev;ce formed between the reinforcement plate and the RWT wall, moisture from the outside etmosphere could accumulate.

Group 5-(SCC near RWT penetrations and.ssociated welds)- Aging Mechanism Effects d

Stress corrosion cracking refers to selective corrosive attack along or across material grain boundaries.

This ARDM requires applied or residual tensile stress, susceptible materials (such as austenitic stainless steels), and oxygen and/or ionic species (e.g., chlorides / sulfates). Common sources of residual stress include thermal processing and stress risers created during surface finishing, fabrication, or assembly.

The heat input during welding can result in a locally-sensitized region that is susceptible to SCC.

[ Reference 1, Attachment 7 for TKs]

Apr ication for License Renewal 5.15-34 Calvert Cliffs Nuclear Power Plant

l ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.I5 - SAFETY INJECTION SYSTEM Root cause evaluations of crack indicaticas at the RWT penetrations have concluded that residual stresses were introduced at these particular locations by the procedures and speci6 cations applied during tank fabrication. Due to the low operating pressures involved and the tough, ductile nature of the construction materials (i.e., Type 304 stainless s'ect), this ARDM is not expected to result in catastrophic failure. [ Reference 1, Atta:hment 6 for RWTs] Ilowever, this aging mechanism, if unmanaged, could result in initiation of cracks, through-wall propagation, and leakage, such that the affected RWT penetrations and welds may not be able to perform their pressure boundary function under CLB design loading conditions.

Since all pipe segments in the injection mode, recirculation mode, and SDC mede flowpaths may not have any Dow due to flushing or performance testing for periods of at least 30 days during normai reactor operation, they were recognized as portion of th: SI System with a high likelihood of containing stagnant oxygenated borated water. Except for the indications near the RWT penetrations noted above, inservice inspections and additional examinations have concluded that the integrity cf welds in these portion; of the SI System have not been affected by se-vice environment and residual stresses that have

nduced pipe cracking in the industry. [ References 82 r.nd 83]

Group 5 -(SCC near RWT penetrations and associated welds)- Methods to Manage Aging Mitigation: For heat-affected zones near the RWT penetrations and associated welds, weld repair of identified cracks can mitigate this ARDM by removing the material affected by residual stresses.

Additionally, control of fl'iid chemistry in the RWT can minimize the introduction of impurities at these locations; refer to subsection Group 2 - Methods to Manage Aging, above. [ Reference 1, Attachir.ent 6 for RWTs]

Discoverv: The effects of SCC near the RWT penetrations and associated welds are detectable by visual inspection. Leakage from cracks at locations susceptible to SCC can be detected by observation of the

" telltale" holes in the associated reinforcement plates. If necessary, internal surfaces of the RWTs can be subjected to oxamination to determine any localized degradation that may be occurring. [ Reference 1 for RWTs]

Group 5 -(SCC near RWT penetrations and as;ociated welds)- Aging Management Program (s)

Mitigation: Observation of a dried boric acid btildup during a system talkdown led to discovery of a pinhole leak inside an outlet nozzle on the Unit 2 RWT. Metallographic examination of material from the excava:ed area concluded that the leak was caused by SCC at the penetration weld. A weld repair of the affected nozzle was comp %ted in 1993.

Maintenance of proper fluid chemistry in accordance with CP-204, as discussed in subsection Group 2 -

Aging Management Programs, above, will limit the effects of SCC. [Re.Srence 1, Attachment 8]

Discoverv: Calvert Cliffs Administrative Procedure MN-1-319, Structure and System Walkdowns,"

provides for discovery of leakage that may result from SCC near the RWT penetrations and associated welds by performance of visual inspections during plant walkdowns. [ Reference 1, Attachment 8 for Application for License Ren:wal 5.15-35 Calvert Cliffs Nuclear Power Plant

{'

ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM RWTs] The purpose of the progre.m is to provide direction for the performance of structure and system wa.kdowns and for the documentation of the walkdown results. [ Reference 84, Section 1.1]

Under this program, responsible personnel perform periodic walkdowns of their assigned structures and systems. Wa!(dawns may also be performed as required for reasona such as: material condition assessments; system reviews before, during, and afler nutages; stet up reviews (i.e., when the system is initially pressuiized, energized, or placed in service); and as required for plant modifications.

[ Reference 84, Section 5.l]

One of the objectives of the program is to assess the condition of the CCNPP structures, systems, and components such that any abnormal or degraded condition will be identified, documented, and corrective actions taken before the condition proceeds to failure of the structures, systems, and components to perform their intended functions. Coaditions adverse to quality are documented an' resolved by the CCNPP Corrective Actions Program. [ Reference 84, Sections 5.1.C,5.2.A.1, and 5.2.A.5]

The prograin provides guidance for identification of specific types of degradation or conditions when performing the walkdowns. Inspection itercs related to aging maaagement include the following:

[ Reference 84, Section 5.2 anc attachments 1 through 13]

Items related to specific ARDMs such as corrosion; Effects that may have been caused by ARDMs such as damaged supports; concrete degradation, mchor bolt de<;radation, or leakage of fluids; and Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage e

of fluids, pre ence of standing water or accumulated moisture, or inadequate support of components (e.g., missing, detached, or loose fasteners and clamps).

The Structure and System Walkdown Program enhances the familiarity of responsible personnel with their assigned systems and provides extended atteation to plant material condition bcyond that afforded by Operations and Maintenance personnel alone. The program has been improved recently through incorporation of significant additional guidance on specific activities to be included in the scope of structures walkdowns. A structure performance assessment is currently required for fluid-rMaining tanks that serve equipment important to safety at CCNPP at least once every six years. The assessment includes a review of each structural component that could degrade the overall performance of the tank (including RWT penetrations and associated welds). [ Reference 84, Sections 1.2.B.7 and 5.3, and ]

The program described above will be modified to: (a) specifically identify the field-erected storage tanks within the scope of the performance assewaents (including the RWT); (b) provide additional visual inspection criteria specific to detecting leakage near the RWT penetrations; and (c) add guidance regerding approval authority for significant departures from the walkdown scope schedule specified.

Since the susceptible ' cations are not nonnally accessible for direct visual inspection, BGE will complete an engin :t g review of SCC at the RWT penetrations that will either: (a) confirm that detection of minor leakage from the " telltale" holes, by itself, will adequately manage SCC at the susceptible locations prior to a challenge to the structural integrity of the penetrations under design basis conditions (e.g., analyze using a " leak-before-break" methodology); or (b) include the RWT penetrations Application for License Renewal 5.15-36 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJFCTION SYSTEM and associated welds in an ARDI Program to verify that unacceptable degradation due to SCC at these locations is not occurring. [Refe:ence 1, Attachment 8] For a discussion of the elemeats of the ARDI Program, refer to subsection Group 2 - Aging Management Programs, above.

The modified Structure and System Walkdown Program, combined with the results of the engineering review described above, will ensure that degraded conditions due to SCC are identified and corrected such that the RWT penetrations and associated welds will be capable of performing their intended function consistent with CLB design conditions.

Group 5 - (SCC nerr RWT penetrations and associated welds) - Demonstration of Aging Management Based on the informatica proemed above, the following conclusions can be reached with.espect to SCC for heat-affected zonu near the RWT penetrations and associated welds:

The RWT penetrations and their associated welds contribute to maintaining the system pressure boundary. Their integrity must be maintained under all CLB design mditions.

Nozzle penetrations for SI System piping connected to the RWT are fabricated of stainless steel.

Residual stresses were introduced at these locations during tank fabrication.

Stress conosion cracking is considered to be a plausible ARDM for the heat-affected zones near the RWT penetraticns and associated welds, since moisture may accumulate in the narrow crevice between the reinfoicement plate and the RWT wall. If unmanaged, this ARDM could eventually result in the loss of pressure-retaining capability under CLB design loading conditions.

Maintenance of proper fluid chemistry in the SI System (in accordance with CP-204) will limit the effects of SCC at the susceptible locations of the RWT.

The CCNPP Structure and System Walkdowns Program provides for periodic walkdowns of SI System components, including the RWT. The program will be modified to specify more clearly the scope and control of periodic performance assessments as they apply to datet. tion of leakage that could result from SCC near the RWT penetrations and associated welds. The program will provide for the discovery of such leakage, and ensure appropriate actions are taken in a timely manner to prevent loss of function.

Based on the results of an engineering review of SCC at the RWT penetrations, these locations may be subjected to a new ARDI Program. This program will examine a representative sample of the components for degradation, and ensure that appropriate corrective actions are initiated on the basis of the findings.

Therefore, there is a reasonable assurance that the effects of SCC ivill be adequately managed for the RWT penetrations and associated welds such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under all design loading conditions.

Application for License Renewal 5.15-37 Calvert Cliffs Nuclev i)ower Plant

l

~

\\

ATTACHMENT m APP' SIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Group 6 -(weathering of RWT perimeter seal)- Materials and Environment The RWT perimeter seal is fibrated cold plastic coal tar pitch flashing, an elastomeric material.

[ Reference 85] This subcomponent supports maintenance of the system pressure boundary by preventing moisture penetration under the RWT. The RWT perimeter seal is exposed to the normal outside atmosphere at the CCNPP site. [Referen e 1, Attachment 6 for RWTs]

Group 6 -(weathering of RWT perimeter seal)- Aging Mechanism Effects Exposure to sunlight, changes in humidity, ozone cycles, snow, rain, ice, and temperature and pressure fluctuations contribute to the weathering ARDM. The effects of weathering on most materials, including the RWT perimeter seal, are evidenced by a decrease in elasticity (e.g., drying out), an increase in hardness, and shrinkage. [ Reference 1, Attachments 6 and 7]

Weathering is plausible for the RWT perimeter seal because it is exposed to the outside environment. Ifleft unmanaged for an extended period of time, the materials of construction will become brittle and lose their capability to prevent moisture penetration under the RWT.

5 Group 6 -(weathering of RWT perimeter scal)- Methods to Manage Aging Mitigation: Since weathering is caused by exposure of the susceptible materials to environmental conditions, which are not feasible to control (e.g., light, heet, oxygen, ozone, wate.-), there are no reasonable methods to mitigate its effects. The discovery method discussed below is deemed adequate to manage this ARDM. [ Reference 1, Attachment 8]

Discoverv: The RWT perimeter seal itself does not perform the passive intended function of maintaining the system pressure boundary; however, it plays a role in mitigating corrosion of the RWT bottom by providing a moisture barrier. [ Reference 1, Attachment 6 for RWTs] Periodic visual inspections can be made of the RWT perimeter seal to detect its degradation. Based on the results of the inspections, the perimeter seal can be repaired or replaced in order to maintain its sealing capabilities. [ Reference 1, ]

Group 6 -(weathering of RWT perimeter seal)- Aging Management Program (s)

Mitigation: There are no CCNPF programs credited for mitigation of weathering.

Disgacy: Calvert Cliffs procedure MN-1-319 is described in subsection Group 5 - Aging Management Programs, above. Performance of visual inspections during plant walkdowns specified in this procedure provides for discovery of weathering for the RWT perimeter seal. [ Reference 1, Attachw.ent 8 for RWTs; Reference 84, Sections 1.2.B.7 and 5.3, and Attachment 9] Modifications to this program will include additional visual inspection criteria specific to the perimeter seal. The modified program will ensure that degraded conditiona due to weathering are identified and corrected such that the RWT perimeter seal will be capable of performing its intended function consistent with CLB design conditions.

Application for License Renewal 5.15-38 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (5)

APPENDlX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Group 6 -(weethering of RWT perimeter seal)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to weathering of the RWT perimeter seal:

The RWT perimeter seal plays a role in mitigating corrosion of the tank bottom by providing a moisture barrier, thereby helping to maintain the pressure boundary function of the SI System.

His capability must be maintained during the period of extended operation.

The RWT perimeter seal is subject to weathering when exposed to the normal outside atmosphere at the CCNPP site.

If unmanaged, this ARDM could result in the loss of the seal's moisture-retaining capability and subsequent corrosion of the RWT.

The CCNPP Structure and System Walkdowns Program provides for periodic walkdowns of SI v

System components, including the RWT. The program will be modified to specify more clearly the scope and control of periodic performance assessments. The program will provide for the discovery of weathering for the RWT perimeter seal, and ensure appropriate actions are taken in a timely manner to correct degraded components or protective coatings.

Therefore, there is reasonable assurance that the effects weathering will be adequately managed for the RWT perimeter seal such that it will be capable of performing its intended function consistent with the CLB during the period of extended operation under all design loading conditions.

5.15.3 Conclusion The aging management programs discussed for the S1 System are listed in Table 5.15-3. These programs are administratively controlled by a formal review and approval process. As demonstra:ed above, these programs will manage the aging mechanisms and their effects in such a way that the intended functions of the components of the SI System will be maintained during the period of extended operation consistent with the CLB under all design loading conditions.

The analysis / assessment, corrective action, and confirmation / documentation process for license renewal is in accordance with QL-2, " Corrective Actions Program." QL-2 is pursuant to 10 CFR Part 50, Appendix B, and covers all structures and components subject to AMR.

Application for L ).ense Renewal 5.15-39 Calvert Cliffs Nuclear l'ower Plant

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ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Table 5.15-3 AC1NG MANAGEMENT PROGRAMS FOR THE SAFETY INJECTION SYSTEM Program Credited As Existing CCNPP Administrative Procedure Program for mitigation and discovery of general MN 3-301," Boric Acid Corrosion corrosion for external surfaces of piping, CKVs, inspection Program" CVs, HVs, HXs, MOVs, RVs, pumps, and tanks (included in Group 1) that are exposed to borated water (due to leakage) by performing visual inspections.

Existing CCNPP Technical Procedure CP-204, Program for mitigation of general corrosion, crevice

" Specification and Surveillance corrosion, and/or pitting for internal surfaces of all Primary Systems" SI System device types (included in Group 2) that are exposed to borated water (as process fluid) by controlling chemistry conditions.

Program for mitigation of SCC near RWT penetrations and associated welds (included in Group 5) that are exposed to borated water (as process fiuid) by controlling chemistry conditions.

Existing CCNPP Tt;chnical Procedure CP 206, Program for mitigation of general corrosion, crevice

" Specification and Surveillance coriosion, and/or pitting for imernal sudaces of the Component Cooling / Service Water LPSI pump seai HXs and IIPSI pump seal coolers System" (included in Group 2) that are exposed to chemically-treated water from the CC System by controlling chemistry conditions in the CC System.

Existing CCNPP Technical Procedure CP-202, Program for mitigation of crevice corrosion and

" Specification and Surveillances pitting for internal surfaces of the RWTHXs Demineralized Water, Safety Related (included in Group 2) that are exposed to well water Battery Water, & Well Water from the Plant Heating System by monitoring well Systems" water chemistry conditions.

Existing CCNPP Surveillance Test Procedure Program for discovery and management of leakage M 571G-l(2)," Local Leak Rate Test, that could be the result of crevice corrosion and Penetrations 9,10,23,24,37,39" pitting for seating surfaces of the SDC header return CCNPP Surveillance Test Procedure is lation MOV outside containment and the St M-571L-1(2)," Local Leak Rate Test, leakoff return isolation HVs (included in Group 2).

Penetration 41" i

Existing CCNPP Pump and Valve IST Program for discovery and management of leakage Program that could be the result of crevice corrosion and i

pitting for seating surfaces of the loop inlet CKVs, the Si header CKVs, and the SIT outlet CKVs j

(included in Group 2).

i s

Application for License Renewal 5.15-40 Calvert Cliffs Nuclear Power Plant l

i O

o ATTACHLWNT (5)

APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Program Credited As Existing CCNPP FMP Program for discovery and management of thermal fatigue for piping and valves in the SIT injection and SDC mode flowpaths (included in Group 4) by monitoring and evaluating low-cycle fatigue usage.

Modified Structure and System Walkdowns Program for discovery and management of SCC (MN-1-319) near RWT penetrations and associated welds included in Group 5) by performing visual (inspections,

. Specify scope and control of periodic structure performance assessments Prograru for discovery and management of weathering effects for the RWT perimeter seal (included in Group 6) by performing visual inspections.

New ARDI Program Program for discovery and management of general corrosion, crevice corrosion, and/or pitting for internal surfaces of all S1 System device types (included in Group 2) by identifying and correcting degraded conditions.

Program for discovery and management of MIC for internal surfaces of the recirculation header piping (included in Group 3) by identifying and correcting degraded conditions.

Progtr.m for discovery and management of SCC near RWT penetrations and associated welds (included in Group 5) by identifying and correcting degraded conditions.

NOTE: Susceptible locations near the RWT penetrations will be included in this program, as necessary, based on the results of an engineering review of SCC at the RWT penetrations.

Not CCNPP Engineering Review of Determine contribution of thermal stratification Applicable Combustion Engineering Owners elTects to thermal fatigue for piping and valves in Group Task Repcets related to NRC the SIT injection mode flowpath (included in Bulletin 88-08 Group 4) by evaluating industry studies and revising

= Review results ofindustry studies cunent analym.

to address the issue of.hermal stratification in piping connected to the RCS and determine changes to piping analyses and fatigue usa, e parameters, as necessary Application for License Renewal 5.15 41 Calvert Cliffs Nuclear Power Plant

ATTACIIMENT (5)

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM Program Credited As Not CCNPP Engineering Review of SCC Determine the acceptability of periodic inspections Applicable at the RWT Penetrations for leakage near the RWT penetrations and associated welds (included in Group 5) for

. Confirm that detection of minor leakage by visuai inspection veill discovery and management of SCC by engineering adequately manage SCC prior to a analysis.

challenge of the structural integrity of the RWT penetrations under design Lasis conditions, or include the susceptible locations in an ARDI Program, as necessary i

i l

l l

l Application for License Renewal 5.15-42 Calvert Cliffs Nuclear Power Plant

O ATTACHMENT (9 APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 5.15.4 References 1.

CCNPP Aging Management Review Report," Safety injection System," Revision 2 2.

CCNPP Updated Final Safety Analysis Report, Units I and 2, Revision 21 3.

CCNPP Operating Instructions, Ol 3A 1(2), " Safety injection and Containment Spray, Unit 1(2)," Revision 3 4.

Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated August 17,1990, forwarding Licensee Event Report 89-007 01, " Damaged LPSI/ Shutdown Cooling Suction Piping Restraint" 5.

Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated August 10,1994,

" Licensee Event Report 94-003, Reactor Shutdown for Leak From Fatigue Crack in Safety Injection Tank Line" 6.

Letter from Mr. C. J. Cowgill (NRC) to Mr. R. E. Denton (BGE), dated August 25,1994, "NRC ' legion I Resident Inspection Report Nos. 50-317/94 24 and 50-318/94-24 (July 3,1994

- August 6,1994)"

7.

Letter from Mr. J. R. Lemons (BGE) to NRC Document Control Desk, dated April 22,1987, fonvarding Licensee Event Report 87 003-00, " Failure of Inlet Piping to Relief Valve j

[2 RV-439]"

8.

Letter from Mr. W. V. Johnston (NRC) to Mr. J. A. Tiernan (BGE), dated June 15,1987,

" Inspection No. 50-318/87-15" 9

Letter from Mr. R. E. Denton (BGE)

NRC Document Control Desk, dated February 14,1991," Licensee Event Report 90-0.sd, Revision 00" 10.

Letter from Mr. E. C. Wenzinger (NRC) to Mr. J. A. Tiernan (BGE), dated May 6,1987,"NRC RI Inspection 50-317/87-06,50-318/87-06" 11.

Letter from Mr. J. R. Lemons (BGE) to NRC Document Control Desk, dated June 5,1987, forwarding Licensee Event Report 87-004-00, " Failure of Inlet Piping to Relief Valve

[2 RV-439]"

l 12.

CCNPP Component Level Scoping Results, " System CS2 - Safety Injection System,"

l Revision 2 13.

BGE Drawing 60729SH0001, " Reactor Coolant System," Rr n 62 14.

BGE Drawing 62729SH0001," Reactor Coolant System," Revision 73 15.

CCNPP Operating Instructions, OI 2D-1(2), " Purification System Operation, Unit 1(2),"

Revision 3 16.

CCNPP Operating Instructions,01-3B-1(2)," Shutdown Cooling, Unit 1(2)," Revision 7 17.

BGE Drawing 60710SH0001," Component Cooling System," Revision 36 18.

BGE Drawing 62710SH0001," Component Cooling System," Revision 35 Application for License Renewal 5.15-43 Calvert Cliffs Nuclear Power Plant

O ATTACHMENT W APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 19.

BGE Drawing 12103 0002, " Piping - Bearing & Stuffing Box Cooling with Sea: Circulation,"

Revision 1 20.

BGE Drawing 12102-0001," Seal Piping," Revision 2 21.

BGE Drawing 12102 0003," Manifold Piping-Series System," Revision 1 22.

BGE Drawing 60716," Spent Fuel Pool Cooling, Pool Fill & Drain Systems," Revision 47 23.

BGE Drawing 60730S110002," Chemical and Volume Control System," Revision 51 24.

BGE Drawing 62730SH0002," Chemical and Volume Control System," Revision 40 25.

BGE Drawing 60730SH0003," Chemical and Volume Control System," Revision 33 26.

BGE Drawing 62730S110003," Chemical and Volume Control System," Revision 33 27.

BGE Drawing 60731SH0003," Safety injection & Containment Spray Systems," Revision 21 28.

BGE Drawing 62731SH0003," Safety injection & Containment Spray Systems," Revision 17 29.

BGE Drawing 60712SH0001, " Compressed Air System, instrument Air and Plant Air,"

Revision 46 30.

BGE Drawing 62712SH0003, " Compressed Ai-System, Instrument Air end Plant Air,"

Revision 89 31 BGE Drawing 607313H0001," Safety injection & Containment Spray Systems," Revision 65 32.

BGE Drawing 60731Sil0002," Safety injection & Containment Spray Systems," Revision 36 33.

BGE Drawing 62731SH0001," Safety injection & Containment Spray Systems," Revision 63 34.

BGE Drawing 62731SH0002," Safety injection & Containment Spray Systems," Revision 34 35.

CCNPP Engineering Standard ES-011, " System, Structure, and Component (SSC) Evaluation,"

l Revision 2 36.

Combustion Engineering Specification No. 68-487-410, " General Engineering Speci0 cation for Safety System Pump," Revision 0 I

37.

Combustion Engineering Specification No. 8067-487-401, " Project Engineering Specification for a High Pressure Safety Injection Pump," Revision 2 38.

Combustion Engineering Specification No. 8067-487-402, " Project Engineering Specification for a Low Pressure Safety Injection Pump," Revision 2 39.

BGE Drawing 92769SH-CC-1,"M-601 Piping Class Summary," Revision 23 40.

BGE Drawing 92769SH-CC-2,"M-601 Piping Class Summary," Revision 20 41.

BGE Drawing 92769SH-HC-1,"M-601 Piping Class Summary," Revision 26 l

42.

BGE Drawing 92769SH-HC-3,"M-601 Pbing Class Summary," Revision 20 43.

BGE Drawing 92769SH-DC-1,"M-601 Piping Class Summary," Revision 19 44.

BGE Drawing 92769SH-GC-1,"M-601 Piping Class Summary," Revision 23 i

Application for License Renewal 5.15-44 Calvert Cliffs Nuclear Power Plant 1

o ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 45.

CCNPP Life Cycle Management Pre-Evaluation Results, " Safety Injection System (052),"

Revision 2 46.

CCNPP Engineering Standard ES-014, " Summary of Ambient Environmental Service Condition <

avision 0 47.

CCNPP AC.mnistrative Procedure MN-3-301, " Boric Acid Corrosion Inspection Program,"

Revision 1 48.

CCNPP Administrative Procedure MN 3-110, " Inservice Inspection of ASME Section XI Components," Revision 2 49.

Letter from Mr. J. A. Tiernan (BGE) to Mr. R. A. Capra (NRC), dated July 7,1987, " Response to Generic Letter 87-06, Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves" 50.

BGE Drawing 12320 0012," Forged Bolted Bonnet Swing Check Valve," Revision 23 51.

BGE Drawing 92769SH-HB-3,"M-601 Piping Class Summary," Revision 29 52.

CCNPP Nuclear Program Directive CH 1," Chemistry Program," Revision 1 53.

CCNPP Technical Procedure CP-204, " Specification and Surveillance Primary Systems,"

Revision 8 54.

CCNPP Technical Procedure CP-206, " Specifications and Surveillance-Component Cooling / Service Water System," Revision 3 55.

CCNPP Technical Procedure CP-202," Specifications and Surveillances-Demineralized Water, Safety Related Battery Water, & Well Water Systems," Revision 5 56.

BGE " Quality Assurance Policy for the Calvert Cliffs Nuen. r Power Plant," Revision 48 57.

Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated October 1,1997,

" Response to Questions on the Third Ten-Year Inservice Trst Program for Safety-Related Pumps and Valves" 58.

CCNPP Administrative Procedure EN-4-102,"ASME Pump and Valve Testing," Revision 1 59.

CCNPP Administrative Procedure EN-4-104," Surveillance Testing," Revision i 60.

CCNPP Surveillance Test Procedure O-65J-1, " Safety injection Check Valve Quarterly Operability Test"(Unit I), Revision 6 61.

CCNPP Surveillance Test Procedure O-65J-2, " Safety Injection Check Valve Quarterly Operability Test"(Unit 2), Revision 8 62.

CCNPP Surveillance Test Procedure 0-27-1, " Reactor Coolant System Leakage Evaluation" (Unit 1), Revision 16 63.

CCNPP Surveillance Test Procedure 0-27-2, " Reactor Coolant System Leakage Evaluation" (Unit 2), Revision 14 64.

Letter from Mr. A. W. Dromerick (NRC) Mr. C. H. Cruse (BGE), dated February 11,1997,

" Issuance of Amendments for CCNPP Unit No.1 (TAC No. M97341) and Unit No. 2 (TAC No. M97342)" (Amendment Nos. 219/196)

Application for License Renewal 5.15 45 Calvert Cliffs Nuclear Power Plant

4 ATTACHMENT $

APPENDIX A - TECIINICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 65.

CCNPP Surveillance Test Procedure M 571G-1," Local Leak Rate Test, Penetrations 9,10,23, 24,37,39"(Unit 1), Revision 0 f

66.

CCNPP Surveillance Test Procedure M-571G 2," Local Leak Rate Test, Penetrations 9, i0,23, 24,37,39"(Unit 2), Revision 1 67.

CCNPP Surveillance Test Procedure M 571L-1, " Local Leak Rate Test, Penetration 41" (Unit 1), Revision 0 4

68.

CCNPP Surveillance Test Procedure M 571L-2, " Local Leak Rate Test, Penetration F (Unit 2), Revision 0 2

69.

10 CFR Part 50, Appendix J,

" Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" 70.

Lettcr from Mr. C.

11. Cruse (BGE) to NRC Document Control Desk, dr.ted November 26,1996,"Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317

& 50-318 License Amendment Request: Adoption of 10 CFR Part 50, Appendix J, Option B for Types B and C Testing" l

71.

Letter from Mr. C. II. Cruse (BGE) to NRC Document Control Desk, dated February 13,1996, "Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Response to Request for Additional Information - Generic Letter 95-07, Pressure Locking and Thermal Binding of Safety-Related, Power-Operated Gate Valves, [ TAC Nos. M93444 & M93445]"

72.

Letter from Mr. L. E. Philpot (Gilbert / Commonwealth, Inc.) to Mr. J. Rycyna (BGE), dated August 29,1995,"MIC Position Paper" 73.

Combustion Engineering Owners Group Task 818, Report No. CE-NPSD-963-01,

" Temperature Distributions and Structural Analysis of Safety injection Piping Subject to Thermal Stratification," September 1995 74.

" Metal Fatigue in Engineering," 11. O. Fuchs and R. I. Stephens, John Wiley & Sons, Copyright 1980 75.

Combustion Engineering Owners i op Task 571, Report No. CE-NPSD-634-P, " Fatigue Monitoring Program for Calvert Cliffs Nuclear Power Plants Units I and 2," April 1992 76.

CCNPP Administrative Procedure EN-1-300, " Implementation of Fatigue Monitoring,"

Revisian 0 77.

Letter from Mr. J. P. Durr (NRC) to Mr. C. Stoiber (sic) (BGE), dated February 11,1993,

" Inspection Report Nos. 50-317/92-32 and 50-318/92-32" 78.

Letter from Mr. J.

A. Tiernan (BGE) to NRC Document Control Desk, dated September 29,1988, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 NRC Bulletin 88-08: Thermal Stresses in Piping Connected to Reactor Coolant Systems" 79.

BGE Procurement Specification 6422284S, " Technical Services to Evaluate Thermal Fatigue Effects on Calvert Cliffs Nuclear Power Plant Systems Requiring Aging Management Review for License Renewal," Revision 0 Application for License Renewal 5.15-46 Calvert Cliffs Nuclear Power Plant

- ~

1 c

s ATTACHMENT m APPENDIX A - TECliNICAL INFORMATION 5.15 - SAFETY INJECTION SYSTEM 80.

NUREG-0933, Generic Safety issue 166, " Adequacy of Fatigue Life of Metal Components,"

Revision 1 81.

BGE Drawing 12329B-0005, " Miscellaneous Nozzles & Sump Details - 41'-6" $ x 41'-6" HG.

Refueling Tank," Revision 2 82.

Letter from Mr. A. E. Lundvall, Jr. (BGE) to Mr. B.11. Grier (NRC), dated August 24,19i9, "lE Bulletin No. 79-17" 83.

Letter from Mr. A.

E. Lundvall, Jr. (DGE) to Mr. B. 11. Grier (NRC), dated November 27,1979, "lE Bulletin No. 79-17 Revision 1 (Pipe Cracks in Stagnant Borated Water Systems at PW.1 Plants)"

84.

CCNPP Administrative Procedure MN 1-319," Structure and System Walkdowns," Revision 0 85.

BGE Drawing 61813," Yard Tank Foundations Sheet 1," Revision 5 4

Application for License Renewal 5.15-47 Calvert Cliffs Nuclear Power Plant