ML20217H598

Forwards Integrated Plant Assessment (IPA) Commodity & Sys Repts & time-limited Aging Analyses Evaluation for Review & Approval IAW 10CFR54,license Renewal Rule

ML20217H598
Person / Time
Site:	Calvert Cliffs
Issue date:	03/27/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9804030403
Download: ML20217H598 (89)

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L T Cn4mixs H. CuosE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 March 27,1998 U. S. Nuclear Regulatory Commission Washington,DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Review and Approval of Commodity and System Reports and the Time-Limited Aging Analyses Evaluation for License 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) Commodity and System Reports and the Time-Limited Aging Analyses Evaluation for review and approval in accordance with 10 CFR Part 54, the license renewal rule. Should we apply for License Renewal, we will reference these 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."

Attachment (3) discusses the Primary Containment Structure and contains information about recently discovered corrosion in certain tendons in the Post Tensioning System. This is a current, ongoing evaluation under the current license. A long-term plan with defined, scheduled actions is expected to be in place prior to Unit I restart followhy the spring 1998 refueling outage.

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'Ihe 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 affect the content of the license renewal application.

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

approval. In Reference (b), the Nuclear Regulatory Commission (NRC) concluded that the IPA Methodology is acceptable for meeting 10CFR 54.21(a)(2) 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 effects of aging will be adequately managed pursuant to 10 CFR 54.21(a)(3) that are concer/nlly sound and consistent with the intent of the s

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 accordance with these guidance documents. We look forward to your comments on the reports as they are submitted and your continued cooperation with our license renewal efforts.

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Document Control Desk l

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l Should you have questions regarding this matter, we will be pleased to discuss them with you.

- Very truly yours,

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STATE OF MARYLAND  :

TO WIT:

COUNTY OF CALVERT  :

I, Charles H. 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 information has been reviewed in accordance with company practice and I believe it to be reliable.

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W Sgbscribed and sworn before me, a Notary Public in d for the State of Maryland and County of

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.this 37& lay of .1998.

WITNESS my Hand and Notarial Seal: jab 2 Notary Public My Commission Expires: b / U~0 h Date  !

CHC/DLS/ dim Attachments: (1) 2.1 Time-Limited Aging Analysis

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(2) 3.l A Piping Segments That Provide Structural Support (3) 3.3A Primary Containment Structure (4) : 3.3E . Auxiliary Building and Safety-Related Diesel Generator Building  ;

Structures "

(5) 5.10 ' Fire Protection System

l. (6) 6.2 Electrical Commodities cc: R. S. Fleishman, Esquire H. J. Miller, NRC

- J. E. Silberg, Esquire Resident Inspector,NRC Director, Project Dimetorate I-1, NRC R. I. McLean, DNR A. W. Dromerick, NRC J. H. Walter, PSC I

' D. L Solorio, NRC I L

I ATTACHMENT (1)

APPENDIX A - TECHNICAL INFORMATION 2.1 - TIME-LIMITED AGING ANALYSES t

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. Baltimore GaTand' Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998

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ATTACIIMENT (1)

APPENDIX A - TECHNICAL INFORMATION L1-TIME-LIMITED AGING ANALYSES 2.1 Time-Limited Aging Analyses i This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing time-limited aging analyses (TLAAs). The TLAAs were 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.

2.1.1 Introduction As required for the LRA, BGE has identified and evaluated analyses in the current licensing basis (CLB) l which may be valid only during the original 40-year license. These TLAAs are defined in 10 CFR 54.3 /

as:

. . . those licensee calculations and analyses that:

1) Involve systems, structures, ard compone.nts within the scope oflicense renewal, as delineated in f54.4(a);

2) Consider the effects ofaging:

3) Involve time-limited assumptions defined by the current operating term,for example, 40 years;

4) Were determined to be relevant by the licensee in making a safety determination:

5) Involve conclusions orprovide the basisfor conclusions related to the capability ofthe system, structure, and component to perform its intendedfunctions, as delineated in f54.4(b); and

6) Are contained or incorporated by reference in the CLB.

2 This definition was clarified by Statements of Consideration accompanying issuance of the License Renewal Rule. An analysis is relevant to license renewal if it provides the basis for a safety  ;

determination and, in the absence of the analysis, a different conclusion may have been reached.

The License Renewal Rule Section {54.21(c)(1) requires a list of TLAAs (as defined above) be provided  ;

in the LRA. The TLAAs were identified through a search of the CLB by performing a keyword search of BGE's electronic files of docketed correspondence and the Updated Final Safety Analysis Report (UFSAR). A list of potential TLAAs was developed using words and phrases indicative of time constraints. This initial list was supplemented by a funher search using a list of codes and standards governing design of systems, structures, and components at nuclear power plants as the input query.

Potential TLAAs thus identified were then screened to determine whether they met the definition presented in 10 CFR 54.3.

Section 54.21(c)(1) also requires the license renewal applicant to demonstrate that one of the following is true for each TLAA: l

1) The analyses remain validfor the period ofextended operation;

2) The analyses have been projected to the end ofthe period ofextended operation; or i 3) The effects ofaging on the intendedfunction(s) will be adequately managedfor the period of extended operation.

Application for License Renewal 2.1-1 Calvert Cliffs Nuclear Power Plant

i ATTACHMENT (1) l APPENDIX A - TECHNICAL INFORMATION l 2.1-TIME-LIMITED AGING ANALYSES -

The TLAAs that were determined to be subject to license renewal review were evaluated in order to demonstrate that each analysis will meet one of the three conditions listed above. The demonstration of how each TLAA meets one of these three criteria is provided in Section 2.1.3 below.

The BGE IPA Methodology and associated N,RC Safety Evaluation Report define the process used at CCNPP to satisfy the requirement to evaluate TLAAs for license renewal. [ Reference 1] Upon applying j this process, only one TLAA, the main steam piping fatigue analysis discussed in Section 2.1.3.4 below, has been demonstrated to meet Criteria 1 above. Criteria 2 has been demonstrated true for 6 out of 14 TLAAs, including the analyses related to irradiation embrittlement of the reactor vessel. In accordance with 654.29(a) of the License Renewal Rule, TLAAs that meet Criteria 2 have either been projected, or will be projected, through the period of extended operation. For those TLAAs that will be projected, i BGE states in Section 2.1.3 below when those actions will be completed. Criteria 3 has been i demonstrated true for the remaining TLAAs. These remaining TLAAs have been evaluated as part of the  ;

IPA for systems, structures, and components, which also requires a demonstration that the effects of '

aging are adequately managed. For these cases, a " pointer" to a distinct section of the BGE LRA where the TLAA is evaluated is provided in Section 2.1.3 below.

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Section QS4.21(c)(2) of the License Renewal Rule requires that the license renewal applicant provide a  :

list of all exemptions granted under {$0.12 that are determined to be based on TLAA. Exemptions are I discussed in Section 2.1.4 below.

2.1.2 List of TLAAs Table 2.1-1 presents a summary list of the TLAAs identified in the CLB. It identifies the analysis involved, the section number in the BGE LRA where additional detail is provided, the subject area of the TLAA(s), and the disposition of the TLAA. [ Reference 2, Table 5 1]

2.1.3 Demonstration of TLAA Dispositions 2.1.3.1 Environmental Qualification  !

The Environmental Qualification (EQ) Program is identified as a TLAA for the purposes of License Renewal. The TLAA aspect of EQ cncompasses all long-lived equipment in the scope of the EQ Program, whether active or passive. At CCNPP, each EQ File for a group of long-lived components includes a qualified life calculation that is considered a TLAA. [ Reference 2, Appendix A)

Environmentally-qualified equipment is replaced with qualified new equipment prior to the end of its qualified life. Preventive maintenance is scheduled to initiate and execute these replacements. Qualified life re-evaluations are an ongoing activity and consider actual normal operating conditions as compared to design maximums (e.g., actual ambient temperatures are below the maximum design temperature that was used as the basis for the current qualified life). Qualified lives are adjusted up or down accordingly.

Qualified life re-evaluations are performed now under the cunent EQ Program and will continue to be performed during the period of extended operation. Refer to Section 6.3 of the BE LRA, Environmental Qualification, for a demonstration of how the effects of aging on the intended functions of electrical equipment in the EQ Program will be adequately managed for the period of extended operation.

[ [ Reference 2, Appendix A]

Calvert Clifts is a Division of Operating Reactors (DOR) Guideline plant. The BGE LRA does not change our CLB relative to EQ. Baltimore Gas and Electric Company has DOR Guideline, NUREG-Application for License Renewal 2.1-2 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT (1)

APPENDIX A - TECHNICAL INFORMATION 2.1-TIME-LIMITED AGING ANALYSES 0588, and 10 CFR 50.49 qualified equipment. Calvert Cliffs will continue to function in the period of extended operation as it does today relative to EQ, except as required by changes to regulatory requirements. Equipment that must be replaced due to the approaching end ofits qualified life will be replaced in accordance with regulatory constraints associated with 10 CFR 50.49 Guidelines, NUREG-0588, or 10 CFR 50.49. Note that equipment qualified to the requirements of the DOR Guidelines or NUREG-0588 will not necessarily be replaced during the period of extended operation.

TABLE 2.1-1 LIST OF TLAAs ANALYSIS SECTION A6E6 EFFECT l Each EQ file is a TLAA 2.1.3.1 EQ-related lleatup and cooldown curves 2.1.3.2 Irradiation embrittlement Power-operated relief valve setpoint for low temperature over pressurization Pressurized thermal shock analyses Reactor vessel fatigue analyses 2.1.3.3 NSSS* fatigue Reactor Coolant System piping fatigue analyses Steam generator fatigue analyses Pressurizer fatigue analyses Pressurizer auxiliary spray line l fatigue analyses l Pressurizer surge line thermal stratification -(fatigue portion of the stress and fatigue calculations)

Main steam piping to turbine driven 2.1.3.4 Fatigue auxiliary feedwater pumps fatigue analysis Containment liner plate fatigue 2.1.3.5 Fatigue analysis Prestress loss calculations 2.1.3.6 Prestress loss Criticality calculation for the spent 2.1.3.7 Loss of fuel pool neutron absorption

• Nuclear Steam Supply System (NSSS) l Application for License Renewal 2.1-3 Calvert Cliffs Nuclear Power Plant a

A'ITACHMENT (1)

' APPENDIX A - TECHNICAL INFORMATION 2.1-TIME-LIMITED AGING ANALYSES l 2.1.3.2 Irradiation Embrittlement This group of TLAAs concerns the effect ofirradiation embrittlement on the reactor pressure vessel and how this mechanism afTects analyses that provide operating limits or address regulatory requirements for CCNPP. These calculations use predictions of the cumulative effects on the reactor vessel from irradiation embrittlement. The calculations are based on periodic assessments of the neutron fluence and resultant changes in reactor vessel material fracture toughness.

Three analyses are affected by embrittlement concerns and are considered in these TLAAs:

(1) Pressurized thermal shock requirements (10 CFR 50.61,10 CFR Pan 50 Appendix G);

(2) Low temperature overpressure protection power-operated relief valve setpoints Q administrative controls,(Technical Specification Figure 3.4.9-3);

(3) Plant heatup/cooldown (pressure / temperature or "PT" limit) curves (Technical Specification Figures 3.4.9-1 and 3.4.9-2). l The pressurized thermal shock analyses have been projected to the end of the period of extended operation. As described in Section 4.1.4.5.4 of the UFSAR and Section 4.2 of BGE's LRA, after the latest revision to regulations addressing the reference temperatures for pressurized thermal shock, BGE showed that both CCNPP reactor vessels will continue to meet pressurized thermal shock screening criteria for 60 years of operation. Baltimore Gas and Electric Company has also augmented its surveillance program to obtain embrittlement information that will bound the period of extended operation. As documented in a series of Safety Evaluation Reports, the NRC has concurred with this demonstration, noting that future test results may change this assessment. [ Reference 3, Section 4.1.4.5.4; References 4,5,6, and 7]

Title 10 CFR Part 50, Appendix G, requires the calculation and use of operational pressure and temperature limits during plant heatups, cooldowns, and inservice hydrostatic tests. The plant heatup/cooldown curves and associated low temperature overpressure protection pressure setpoint curves in the plant Technical Specifications provide for overpressure protection during these operating modes.

Currently, the Unit I curves are valid beyond 48 effective full power years, while the Unit 2 curves are valid to approximately 30 effective full power years. The Technical Specifications will continue to be updated either as required by 10 CFR Part 50, Appendices G and H, to assure the operational limits remain valid at the current cumulative neutron fluence levels, or on an as needed basis to provide appropriate operational flexibility.

2.1.3.3 NSSS Fatigue Analyses Components in the NSSS are subject to a wide variety of varying mechanical and thermal loads that

contribute to fatigue accumulation. The Reactor Coolant System components were designed in l

accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Sectionin, and the American National Standards Institute (ANSI) Standard USAS B 31.7, Nuclear Power Piping Code. These codes require the design analysis for Class I components to address fatigue and establish limits such that initiation of fatigue cracks is precluded. Portions of UFSAR Section 4.1 and the certified design specification identify the difTerent design cyclic transients used in the fatigue analysis required by code for various major components of the Reactor Coolant System including the reactor vessel, Reactor Coolant System piping, steam generatois, pressurizer, pressurizer auxiliary spray piping, and pressurizer surge line.

Application for License Renewal 2.1-4 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (1)

APPENDIX A - TECHNICAL INFORMATION l 2.1-TIME-LIMITED AGING ANALYSES The CCNPP Fatigue Monitoring Program trackt the number of critical thermal and pressure test transients, and monitors the cycles and fatigue usage for the limiting components of the NSSS.

Locations in these systems have been selected for monitoring for fatigue usage; they represent the bounding locations for critical thermal and pressure transients and operating cycles. In order to stay withia the design basis, corrective action is initiated well in advance of the cumulative fatigue usage factor approaching 1.0 or exceeding the number of design cycles, so that appropriate corrective actions can be taken in a timely and coordinated manner.

Baltimore Gas and Electric Company's demonstration that the effects of fatigue on the intended function (s) of NSSS components will be adequately managed for the period of extended operation is provided in the following sections of the BGE LRA: Section 4.1, Reactor Coolant System; Section 4.2, ,

Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System; and Section 5.2, Chemical and Volume Control System. The NRC staff concerns about fatigue for license renewal as identified in Generic Safety Issue 166, Adequacy of Fatigue Life of Metal Components, are also addressed in these referenced sections of BGE's LRA.

2.I.3.4 Main Steam Piping Fatigue Analysis The main steam supply lines to the auxiliary feedwater pump turbines provide the system pressure boundary function and are subject to thermal loadings. According to the UFSAR Chapter 10A discussion,21,999 rapid full temperature cycles have been considered. However, even if the number of assumed cycles were limited to 7000 equivalent full temperature cycles, which is much more limiting, this piping would have to be cycled approximately once every 3 days over an extended plant life of 60 years. Under current plant operating practices, the system is cperated only occasionally during plant heatups and cooldowns, during plant transients, and for periodic (monthly) testing. Plant heatups and cooldowns are limited to 500 each, and reactor trips are limited to 400 over plant life. Monthly testing over 60 years would contribute another 720 cycles. These actual and potential cycles combined equal '

slightly more than 2000 cycles for the auxiliary feedwater steam supply. It is, therefore, unlikely that the 7000 assumed cycles will be approached during the period of extended operation. Thus, the existing analysis is considered to remain valid for the period of extended operations, and there is reasonable assurance that the intended function will be maintained. Generic Safety Issue 166 does not apply to the main steam supply lines to the auxiliary feedwater pump turbines. [ Reference 2, Appendix C) 2.1.3.5 Containment Liner Plate Fatigue Analysis American Society of Mechanical Engineers codes require that the containment liner material be prevented from experiencing significant distortion due to the thermal load and that the stresses be considered from a fatigue standpoint. The following fatigue loads were considered in the design of the i liner plate: [ Reference 3, Section 5.1.4.3]

• The annual outdoor temperature variation, assumed to be 40 cycles during the plant's 40-year life;

• The interior tempe:ature variations during the startup and shutdown of the Reactor Coolant System, assumed to be 500 cycles; and i

e Thermal cycling due to a loss-of-coolant accident, assumed to occur once during plant life.

(Note: American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,  !

"aragraph 412(n), Figure N-415(A) and its appropriate limitations have been used as a basis for r

Application for License Renewal 2.1-5 Calvert Cliffs Nuclear Power Plant j

.O ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 2.1 -TIME-LIMITED AGING ANALYSES establishing allowable liner plate strains. Since the graph in Figure N-415(A) does not extend below ten cycles, ten loss-of-coolant accident cycles was used for the analysis.)

The design of the liner plate and penetration sleeves included consideration of thermal stress and fatigue for which there was an assumed number and severity of thermal cycles. Since this assumption was partly based on a 40-year operating life, the fatigue analyses must be reviewed to assure they remain valid during the period of extended operation. This review or re-analysis will be projected to the end of the period of extended operation by the year 2012. Generic Safety Issue 166 does not apply to the containment liner plate.

2.1.3.6 Containment Tendons Prestress Loss The prestress on the containment tendons decreases over plant life as a result of clastic deformation, creep and shrinkage of concrete, anchorage seating losses and tendon wire friction, stress relaxation, and corrosion. The extent of these losses over the original expected plant life was predicted to verify the Containment design. Regulatory Guide 1.35 requires periodic monitoring of cur ent prestress values to ensure that prestress loss predictions of the original design remain valid. For a complete discussion of the aging management review performed on th: tendons for license renewal refer to Section 3.3A, Primary Containment, of the BGE LRA. Section 3.3A includes a discussion on the recent operating

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experience regarding the discovery of corrosion and hydrogen-induced cracking.

Tendon prestress losses are determined by measuring tendon lift-off force. Technical Specification 4.6.1.6.1 (Unit 1) establishes the surveillance schedule for measuring lift-off forces of selected tendons. This measurement is performed in accordance with Surveillance Test Procedure STP-M-663-1. Technical Specification Figures 3.6.1-1 (Hoop),3.6.1-2 (Vertical), and 3.6.1-3 (Dome) provide the normalized lift-off forces required to be achieved during the surveillance test procedure as a I function of plant service life after initial prestressing. These curves presently cover 40 years of plant life. !

(Note that these curves are not included in the Unit 2 Technical Specifications as the lift-off tests have not been required for that Unit.) These curves will be recalculated by the year 2012 to accommodate the ]

projected 20-year period of extended operation.

2.1.3.7 Poison Sheets in Spent Fuel Pool The criticality analyses for the Units 1 and 2 spent fuel pools credit the existence of poison (i.e., neutron absorbing) sheets located between spent fuel assemblies. The criticality calculations assume the neutron absorbing material has a minimrm concentration of Boron.10. [ References 8 and 9] If th.:re was a reduction in the amount of neutron absorbing material to below that assumed, the calculation may become non-conservative.

The criticality analysis for Unit I contains an assumption of the boron concentration that accounts for a potential loss of boron carbide due to aging. This conservative assumption was made based on experiments showing that the Carborundum sheets that are installed in Unit I may experience a loss of boron content due to aging. Therefore, the Unit 1 analysis is considered a TLAA. [ Reference 8) The neutron absorbing sheets installed in Unit 2 are constructed of a material called Boraflex. The Unit 2 criticality analysis does not contain any assumption of a loss of boron concentration due to aging and, therefore, is not considered a TLAA. [ Reference 9)

Application for License Renewal 2.1-6 Calvert Cliffs Nuclear Power PJant

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ATTACHMENT (1) l APPENDIX A - TECHNICAL INFORMATION  !

2.1 -TIME-LIMITED AGING ANALYSES i The spent fuel pool contains high-density spent fuel storage racks that consist of a base structure supporting storage cells prunarily fabricated from stainless steel. For Unit 1, a neutron-absorbing sheet, i fabricated by He Carborundum Company and consistmg of a boron carbide powder in a fiberglass matrix, is sandwiched between the inner and outer walls on the four sides of each storage cell. The

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original neutron absorbing-sheet was specified to contain a minimum concentration of 0.024 grams per  !

square centimeter of Boron-10. [ Reference 10] I The Unit I criticality analysis contains an assumption of the boron concentration that accounts for a potential loss of boron carbide due to aging. This conservative assumption was made based on l experiments showing that the Carborundum sheets may experience a loss of boron content, following  !

exposure to gamma radiation equivalent to 40 years of service and a spent fuel pool water environment,  !

due to degradation of the matrix material in which the boron carbide is bonded. The loss of boron I carbide reduces the Boron-10 concentration experienced by an average value of 15%, with the maximum reduction in Boron-10 concentration experienced by any single test sample being 19.2%. The criticality analysis accounted for this potential degradation by assuming a minimum concentration of 0.020 grams per square centimeter of Boron-10. Since the degradation rate used in the current analysis was based on a radiation exposure sufficient to accommodate at least a 40-year pool lifetime, it must be updated to reflect the total exposure for 60 years. This analysis is currently being updated and will accommodate l

the period of extended operation. A service life of 70 years will be demonstrated for the Carborundum i sheets, which will permit at least 10 years of usage beyond the period of extended operation. The update will be completed by 1999. [ References 8 and 10] l l

Baltimore Gas and Electric Company has performed an aging analysis for the poison sheets and has i determined there are plausible aging mechanisms for both Units I and 2 at CCNPP. Baltimore Gas and l Electric Company's demonstration that the effects of aging are being adequately managed for the period  ;

of extended operation is provided in Section 3.3E, Auxiliary Building and Safety-Related Diesel '

Generator Building Structures, of the BGE LRA. Operating experience with these poison sheets is also discussed in that section of the report.

2.1.4 List of Exemptions Based on TLAA l Section 54.21(c)(2) of the License Renewal Rule requires a list of all exemptions granted under 10 CFR 50.12 that are determined to be based on a TLAA. These exemptions must be evaluated and justification provided for the continuation of the exemption during the period of extended operation.

Baltimore Gas and Electric Company found no exemptions that were based on a TLAA.

2.1.5 Conclusions Baltimore Gas and Electric Company has identified and evaluated the TLAAs important to license renewal, in accordance with 10 CFR 54.21(c). This evaluation demonstrates that 1 out of the 14 TLAAs remain valid,6 out of the 14 have been (or will be) projected to the end of the period of extended operation, and for the remaining 7, the effects of aging on the intended function (s) will be adequately managed for the period of extended operation.

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l Application for License Renewal 2.1-7 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECEINICAL INFORMATION l 2.1 -TIME-LIMITED AGING ANALYSES i 2.1.6 References

1. 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

2. CCNPP License Renewal Tect'nical Report," Time Limited Aging Analysis Review Report,"

Revision 0, November 1997

3. CCNPP Updated Final Safoty Analysis Report, Revision 21

4. Letter from Mr. D. G. M( Donald, Jr. (NRC) to Mr. G. C. Creel (BGE), dated July 15,1992, Response to the 1991 Pressurized Thermal Shock Rule,10 CFR 50.61, Calvert Cliffs Nuclear Power Plant, Unit 1 (TAC No. M82504) and Unit 2 (TAC No. M82505)

5. Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BGE), dated May 24,1993, Response to the 1991 Pressurized Thermal Shock Rule,10 CFR 50.61, Calvert Cliffs Nuclear Power Plant, Unit 2 (TAC No. M82505)

6. Letter from Mr. M. L. Boyle (NRC) to Mr. R. E. Denton (BGE), dated July 29,1994, Request for Approval to Use Plant Specific Data for Reactor Vessel Fracture Toughness Analysis, Calvert Cliffs Nuclear Power Plant, Unit No.1 (TAC No. M88316)

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7. Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BGE), dated January 2,1996,

" Updated Valves for Pressurized Thermal Shock Reference Temperatures - CCNPP Units I and 2 (TAC Nos. M93230 and 93231)

8. Combustion Engineering Design Analysis A-CCl-FE-0005," Reanalysis of Calvert Cliffs Unit 1 Spent Fuel Pool Criticality Calculations," April 14,1992

9. Combustion Engineering Design Analysis A-CC2-FE-0003," Reanalysis of Calvert Clifts Unit 2 Spent Fuel Pool Criticality Calculations," Revision 2, July 2,1992

10. Letter from Mr. A. E. Lundvall, Jr. (BGE) to Mr. Robert W. Reid (NRC), dated January 15,1980, " Spent Fuel Pool Modification Supplementary Information" 1

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l APPENDIX A - TECHNICAL INFORMATION 3.1A - PIPING SEGMENTS THAT PROVIDE STRUCTURAL i

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f l Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998

ATTACHMENT (2)

APPENDIX / - TECHNICAL INFORMATION 3.1A PIPING SEGMENTS TIIAT PROVIDE STRUCTURAL SUPPORT 3.1A Piping Segments that Provide Sructural Support This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA) addressing piping segments that provide structural support. This section should be reviewed in conjunction with Section 3.1, Component Supports. The items in this section have been evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment Methodology described in Section 2.0 of the BGE LRA. Rese sections are prepared independently and will, collectively, comprise the entire BGE LRA.

3.lA.1 Scoping 3.1 A.I.1 Piping Segments that Provide Structural Support Commodity Seoping Commodity Descriotion/Concentual Boundaries The purpose of this section is to document the commodity approach used to evaluate the aging management of the pipe beyond the safety-related/non-safety-related (SR/NSR) boundary to the first seismic restraint (s), which provide the structural support for the functional boundary isolation valve (s) or isolation points. Figure 3.l A-1 below is a simplified drawing that illustrates the piping of concern between points A and B.

A B

SR l NSR Seismic Non-Seismic Category i

, piping of concern

• F 1 h >4 A I E SR weld per seismic anchor or code equivalent Figure 3.1 A-1 (simplified)

(* piping supports in this region are covered in Section 3.1 of the BGE LRA, Component Supports) l The SR/NSR functional boundary includes a transition in safety and, in some cases, piping classifications. The structural integrity of the boundary valve, which functions as the system pressure boundary, must not be compromised. Therefore, the system's seismic structural boundary extends beyond the valve to the first seismic anchor or equivalent. In some instances, the valve itself may be anchored. However, in most cases, the anchor is beyond the valve; and the support system includes the piping segments that provide structural support for the boundary valve. Collectively, these components l Application for License Renewal 3.1 A-1 Calvert Cl;ffs Nuclear Power Plant i.

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ATTACHMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.lA PIPING SEGMENTS THAT PROVIDE STRUCTURAL SUPPORT act as a single " support system." These components ensure the integrity of the SR/NSR functional boundary under all design basis loading conditions.

The design loading condition: for these piping segments include factors such as dead loads, thermal loads, and seismic loads. :. ipporting information for loading conditions of specific supports is maintained onsite. [ Reference 1, Appendix 5A] Basic design basis information for piping segments is discussed in UFSAR Chapters I (Principal Architectural and Engineering Criteria for Design) and SA (Structural Design Basis).

Intended Functions The piping segments beyond SR/NSR boundaries have the intended function of providing structural support under all current licensing basis design loading conditions for SR components within the scope oflicense renewal.

Pining Seements Reauiring Aging Management Review Because the intended function listed above is provided without moving parts or without a change in configuration or properties, it is a passive intended function. Therefore, the piping segments between the SR/NSR boundary and the seismic anchor are within the scope oflicense renewal and are also subject to aging management review.

The scope of this section includes all piping segments beyond the SR/NSR functional boundary that perform the intended function of providing structural support to the SR piping and boundary isolation valve or isolation point (see Figure 3.l A-1).

All fluid systems containing SR piping are within the scope of license renewal, as shown in Table 3.l A-1. These systems have the potential for SR/NSR functional boundaries with Seismic Category I boundaries extending beyond them.

1 l

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1 l

I l

l l

Application for License Renewal 3.lA-2 Calvert Cliffs Nuclear Power Plant

e 3 ATTACHMENT (2) l APPENDIX A - TECHNICAL INFORMATION 3.1A PIPING SEGMENTS THAT PROVIDE STRUCTURAL w SUPPORT TABLE 3.1A-1 SYSTEMS WITHIN THE SCOPE OF LICENSE RENEWAL WITH POTENTIAL FOR CONTAINING PIPING SEGMENTS BEYOND SR/NSR BOUNDARIES j (CCNPP system numbers are shown in parentheses)

System Associated BGE LRA Section (011) Service Water Cooling Section 5.17 (012) Saltwater Cooling Section 5.16 l (013) Fire Protection Section 5.10 l (015) Component Cooling Section 5.3 (019) Compressed Air Section 5.4 ,

(023) Diesel Fuel Oil Section 5.7 (024) Emergency Diesel Generators Section 5.8 (029) Plant Heating Section 5.5 (036) Auxiliary Feedwater Section 5.1 (037) Demineralized Water / Condensate Storage Section 5.10 (038) Sampling System (Nuclear Steam Supply System) Section 5.13 (041) Chemical and Volume Control Section 5.2 (045) Feedwater Section 5.9 (046) Extraction Steam Section 5.12 (051) Plant Water Section 5.5  ;

(052) SafetyInjection Section 5.15 (053) Plant Drains Section 5.5 (061) Containment Spray Section 5.6 (064) Reactor Coolant Section 4.1 (067) Spent Fuel Pool Cooling Section 5.18 (069) Waste Gas Section 5.5 (071) Liquid Waste Section 5.5 l (074) Nitrogen and Hydrogen Section 5.12 (077/79) Area and Process Radiation Monitoring Section 5.14 (083) Main Steam Section 5.12 3.1 A.2 Aging Management The plausible age-related degradation mechanisms (ARDMs) for each piping segment beyond the SR/NSR boundary are the same ARDMs as those identified in their respective fluid system discussions listed in Table 3.l A-1. These are the ARDMs that could lead to degradation and the potential for loss of the passive intended structural support function.

The piping segments beyond the SR/NSR boundary are classified as Seismic Category I up to and including the first seismic anchor. Given the similarity of the piping materials for piping within the SR pressure boundary, to those outside this boundary that are designed and maintained to SR requirements, any material degradation identified on the pipe segments within the SR pressure boundary would lead to an evaluation for generic implications on the NSR side of this boundary. In a.!dition, the aging management programs credited in the fluid systems listed in Table 3.lA-1 in conjunction with the CCNPP Corrective Actions Program will ensure that the intended function of providing structural support to the SR pipe and boundary isolation valves or isolation points will be maintained under current licensing basis design loading conditions during the period of extended operation. The programs for

~

Application for License Renewal 3.l A-3 Calvert Cliffs Nuclear Power Plant

1

• e t i NITACHMENT (2)

APPENDIX A - TECHNICAL INFORMATION 3.1A PIPING SEGMENTS THAT PROVIDE STRUCTURAL SUPPORT aging management of the fluid systems listed in Table 3.lA-1 are referenced in previous BGE LRA submittals as noted in this table.

3.1A.3 Conclusion The piping segments beyond the SR/NSR functional boundary will be managed by the programs already  !

credited in the Sections of the BGE LRA for the SR portion of the systems listed in Table 3.l A-1. These l programs will manage the aging mechanisms such that the intended structural function of the piping I segments beyond the SR/NSR boundary up to the first anchor point or equivalent will be maintained consistent with the current licensing basis during the period of extended operation.

3.1A.4 Reference

1. "CCNPP Updated Final Safety Analysis Report," Revision 21 1

I l

Application for License Renewal 3.l A-4 Calvert Cliffs Nuclear Power Plant

c. s ~ -

ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE I

l Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998 j J

l ATTACHMENT G) '

APPENDIX A - TECHNICAL INFORMATION  !

3.3A - PRIMARY CONTAINMENT STRUCTURE I i

- 3.3A Primary Containment Structure This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application

! (LRA), addressing the Primary Containment. The Primary Containment was evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology l described in Section 2.0 of the BGE LRA. These sections are prepared concurrently and will, l collectively, comprise the BGE LRA. 1 3.3A.1 Scoping l 1

The IPA Methodology system level scoping describes conceptual boundaries for plant systems and l

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. The Primary )

Containment consists of two categories of components, the Containment Structure and the Containment '

System. The Containment Structure includes the majority of structural components such as beams, ,

columns, walls, and liners. The Containment System includes penetrations, hatches, air locks, and  !

associated instrumentation.

1 For the Containment Structure, the component level scoping to determine which components are within

{

the scope of license renewal was accomplished utilizing the scoping process for structures as described l in Section 4.2 of the BGE IPA Methodology. This was done because the features and documentation of  !

structures are distinct from that of systems at CCNPP, and therefore, the component level scoping ,

process for structures differs from that applied to systems. In the structural component scoping process, 1 scoping is conducted using a generic listing of structural component types. Additional structural component types not included in the generic listing because they are unique to the Containment Structure ,

are also identified. Scoping is implemented by determining which structural component types are  !

required for performance of the passive intended functions of the structure. The results of the  ;

Containment Structure scoping are merged with the results of the Containment System scoping to present  !

a combined scoping result for Primary Containment.

For the Containment System, the component level scoping to determine which components are within the scope of license renewal was accomplished utilizing the scoping process for systems as described in Section 4.1 of the BGE IPA Methodology. This scoping step begins with a listing of passive intended ftmetions. Subsequently, component types are dispostioned as either only associated with active functions, subject to replacement, or subject to AMR either in this section of the BGE LRA or another section. The component level scoping includes a determination of which components are 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. j Section 3.3A.I.1 presents the results of the structure / system level scoping,3.3A.I.2 the results of the component level scoping, and 3.3A.1.3 the results of scoping to determine components subject to an

( AMR.

(

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

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE j 3.3A.1.1 System L,evel Scoping l This section begins with a description of the Containment Structure and Containment System, which includes the boundaries of the Primary Containment as it was scoped. A brief summary is presented of l the overall operating experience related to aging. Finally, the results of the system level scoping process are presented with a listing ofintended functions of Primary Containment.

Descrintion/Concentual Boundaries l

' Figure 3.3A-1 is a simplified layout of site structures showing the structures that are within the scope of license renewal. The CCNPP site arrangement consists of numerous structures that are shown on Updated Final Safety Analysis Report (UFSAR) Figures 1-2 through 1-30, with further discussion of l their design features in Chapter 5. [ Reference 1, Chapters 1 and 5; References 2,3, and 4] A general description, boundary, and design discussion of the Containment Structures and Containment System follows: [ Reference 1, Chapters 1,5 and 8]

The twin Containment Structures are located northwest and southwest of the Auxiliary Building with a connective boundary to the Auxiliary Building formed by the cylindrical shape of each Containment Structure. Each Containment Structure houses a reactor and other Nuclear Steam Supply Syeem components consisting of steam generators, reactor coolant pumps, a pressurizer, and some of the reactor auxiliaries that do not normally require access during power operation. The containment consists of a shallow domed roof and a reinforced concrete cylinder that rests on a reinforced concrete foundation slab. The concrete cylinder and dome incorporate a post-tensioned contraction design. A carbon steel liner is attached to the inside of the Containment Structure to assure a high degree of leak tightness.

There are three personnel and equipment access openings in the containment: a two-door personnel air lock, a large diameter single door equipment hatch, and a two-door personnel escape hatch. The primary containment has numerous penetrations for piping and electrical connections. These penetrations are pressure-resistant, leak-tight assemblies, which are welded to the containment liner. A fuel transfer tube penetration in the containment is provided to permit fuel movement between the refueling pool in the containment and the spent fuel pool in the Auxiliary Building. A normal and an emergency sump are provided in the containment floor. [ Reference 1, Section 1.2.5; Reference 5, Section 1-1]

The Containment Structure and its structural components provide structural / functional support and shelter / protection to safety-related (SR) and non-safety-related equipment inside the Containment Structure.

The Containment Structure also serves as a pressure boundary or a f'ssion product retention barrier to protect public health and safety in the event of postulated Design Basis Events (DBEs). In addition, the Containment Structure provides a missile, flood, end fire barrier for SR equipment. The boundary addressed by this scoping and evaluation includes all in-containment structural components serving such functions and components comprising the containment pressure boundary, but does not include commodity items such as pipe supports and snubbers. [ Reference 5, Section 1.1.2]

The Containment Structure is designed to withstand an internal pressure of 50 psig, a coincident concrete surface temperature of 276*F, and limit leakage to no more than 0.20% by weight per day at the design temperature and pressure. The Containment Structure is designated a seismic Category I structure and is designed for all loading combinations described in Section SA.3 of the UFSAR. [ Reference 1, Sections 1.2.5 and 5.1.1]

f Application for License Renewal 3.3 A-2 Calvert Cliffs Nuclear Power Plant i

4 ATTACHMENT (3) l APPENDIX A - TECIINICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE N

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oc l Expansion Joint STRUCTURES WITHIN THE SCOPE OF o seloerwetors Elev 4F.6" h LICENSE RENEWAL Figure 3.3A-1 Simplified Layout of Structures Application for License Renewal 3.3A-3 Calvert Cliffs Nuclear Power Plant

l ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE ne Containment System includes those components of the primary Containment Structure that are listed in the CCNPP equipment list as individual components with unique equipment identifiers. Components within the Containment System boundary include the following major component types: Penetration, Door, Pressure Indicator, Pressure Switch, and Position Switch. The Containment System is in scope for license renewal based on 10 CFR 54.4(a) criteria. He Containment System components perform one or more of the following functions: provide closure on containment air lock and access / egress hatches, maintain functionality of electrical components as addressed by the Environmental Qualification (EQ)

Program, and maintain the pressure boundary for the system. [ Reference 6, Sections 1.1.2 and 1.1.3]

Containment penetrations range in size from the small closure pieces for electrical and piping penetrations to larger components such as air locks and the equipment hatch. All containment penetrations are pressure resistant, leak-tight, welded assemblies designed, fabricated and tested in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Class B, Nuclear Vessel Code. [ Reference 1, Section 5.1.4.4]

The conceptual boundaries of this evaluation include the Containment Structure and all of its structural components such as foundations, walls, slabs, and steel beams. Component suppons 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 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 Supports Commodity Evaluation and the evaluation for the structural component. Evaluations for structural components consider 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 (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 7, Section 1.1.1]

Cranes and fuel handling equipment that are connected to structures are evaluated for the effects of aging in the Fuel IIandling Equipment and Other 11eavy Load llandling Cranes Commodity Evaluation in Section 3.2 of the BGE LRA. The polar crane, reactor vessel head lift rig, transfer machine jib crane, fuel upending machine, and reactor refueling machine were evaluated in the Cranes and Fuel llandling Commodity Evaluation and are not included in this section. The polar crane girders are included herein.

Ooerating Exnerience An inspection of the Unit 1 Primary Containment was performed in 1992 to support the license renewal screening and AMR activities for the structures at CCNPP. The inspection was to evaluate the overall condition of the Primary Containment. A representative sample of internal and external structural components were examined, to the extent practical, in accordance with industry standards. The methodology employed meets the intent of the industry standard, " Rules for Inservice Inspection,Section XI, ASME Boiler and Pressure Vessel Code," which gives rules for the inspection of concrete.

Under these rules, the exterior and interior surfaces of the Containment Structure and components in the Containment System were found to be in good to excellent condition. The responsible engineer Application for License Renewal 3.3 A-4 Calven Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE determined by visual examination that there is no evidence of damage or degradation sufficient to warrant funher evaluation or repair.

In 1997, during performance of the 20 year Technical Specification tendon surveillance on Unit 1 ,

Containment Structure, broken wires were discovered. The discovery of broken wires initiated an l

expansion of the vertical tendon inspection scope to perform visual inspection and hit-off testing on all i venical tendons for Unit 1. At the completion of the expanded scope, a number of tendons were identified as having severe corrosion (pitting greater than 0.003 inches) and/or broken wires. A root cause analysis concluded that tendon wire failures and corrosion problems resulted from a combination of water and moist air intrusion, and inadequate initial grease coverage of wires below the upper stressing washer. This combination created a corrosive environment, which in turn, caused wire failure either by general corrosion or by hydrogen-induced cracking. [ References 8 and 9] Further details on tendon corrosion and other operating experience is provided in the Group discussions, where appropriate.

System / Structure Scooing Results i

The Containment Structure and the Containment System were both determined to be within the scope of I license renewal based on 10 CFR 54.4(a) as a result of executing the screening process described in Section 3 of the BGE IPA Methodology. The following intended functions of the Primary Containment l were determined based on the requirements of Q54.4(a)(1) and (2): [ References 5,6,10, and 11]

1. Support a pressure boundary or a fission product retention barrier function to protect public health and safety in the event of any postulated DBEs;

2. Provide shelter / protection to SR equipment (this function includes radiation protection for EQ ,

equipment and high energy line break-related protection equipment); I

3. Provide structural and/or functional support to SR equipment;

4. Serve as a missile barrier (internal or external);

5. 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; and j 6. Provide flood protection barrier (internal flood event).

l The following intended functions of the Containment Structure and Containment System were determined based on the requirements of 54.4(a)(3): [ References 5,6,10, and i1]

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

8. Provide closure of containment air lock and access / egress hatches during a station blackout (active); and

9. Maintain the functionality of electrical components as addressed by the EQ program.

3.3A.I.2 Component Level Scoping During the scoping process, a list of generic structural component types were identified for the Containment Structure. Additional structural component types, not included in the generic listing because they are unique to the Containment Structure, were also identified. Each structural component type is a category of components that is comprised of one or more structural components based on design and function. A total of 34 structural component types were identified as within the scope of license Application for License Renewal 3.3 A-5 Calvert Cliffs Nuclear Power Plant

A'ITACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE renewal because they contribute to at least one of the intended functions of the Containment Structure.

All of these structural component types are addressed below in Section 3.3A.I.3. These structural

! component types were further combined into the following four structural component categories based l on their design and materials. [ References 5 and 10]

• Concrete Components;

• Structural Steel Components; i e Architectural Components; and e

Unique Components (e.g., post-tensioning system, containment liner, refueling pool liner and permanent cavity seal ring, and emergency sump cover and screen).

l The components comprising the Containment System were identified through the CCNPP equipment database. The purpose of the component level scoping was to identify all system components that support one or more of the intended functions of the system. The intended functions of the Containment System are Functions 1,3, 8, and 9 as listed above in section 3.3A.I.l. Components that support these intended functions were categorized into the following three component types: [ References 6 and 11]

Comnonent Type Comoonent Descriotion DOOR Air locks and equipment hatch PEN Containment penetrations and fuel transfer tube ZS Limit switches Some components in the Containment System 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 following: (Reference 6, Section 3.2] l e Structural supports for piping, cables, and most components are evaluated for the effects of aging l in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA. Supports for l the steam generators and pressurizer are evaluated in the Reactor Coolant System evaluation in l Section 4.1 of the BGE LRA. Supports for the reactor vessel are evaluated in the Reactor l Pressure Vessels and Control Element Drive Mechanisms / Electrical System in Section 4.2 of the BGE LRA.

• Electrical control and power cabling are evaluated for the effects of aging in the Cables Evaluation in Section 6.1 of the BGE LRA.

l 3.3A.1.3 Components Subject to AMR l This section contains a discussion of structural component types for the Containment Structure and

! Containment System that are subject to AMR. In accordance with Section 5.0 of the BGE IPA Methodology, components that support only active functions, or that are subject to periodic replacement based on a qualified life or specified time period do not require AMR. Tables 3.3A-1 and 3.3A-2 includes the 44 component types for the Containment Structure and Containment System, respectively, and lists, by function number, the passive intended function (s) that each one supports. [ References 6 and 11]

(

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

e. *.

l-ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION i

3.3A - PRIMARY CONTAINMENT STRUCTURE The intended functions of the Containment Structure are Functions 1 through 7 as listed above in Section 3.3A.I.l. All of these intended functions are passive. Additionally, none of the structural component types are subject to periodic replacement based on a qualified life or specified time period.

Therefore, all of the Containment Structure components that comprise the 37 structural component types l within the scope oflicense renewal require AMR. [ Reference 5]

The intended functions of three of the seven Containment System component types, i.e., DOOR, PEN, and ZS, are Functions 1, 8, and 9 as listeo above in Section 3.3A.l.l. Function 8, provide closure of containment air lock and access / egress hatches during a station blackout, is an active function performed manually by an operator. This active function is the only intended function for component type ZS, limit switches. Therefore, the component type ZS does not require AMR. [ References 6 and 11]

The other component type that supports the active function is DOOR. The component type DOOR is comprised of the containment personnel air lock, the containment emergency air lock, and the

~

containment equipment hatch. These components also have a passive intended function of the Containment System, so they require AMR.

The containment personnel and emergency air locks and the equipment hatch are installed with resilient gaskets to help assure a leak tight barrier for the Primary Containment Structure. The equipment hatch and personnel air lock gaskets are currently scheduled for replacement every four years and are, therefore, not subject to AMR. [ References 12 and 13]

The emergency air lock gaskets are replaced based on condition. The gaskets are currently scheduled for inspection every two years. The inspection is performed visually and any indication of nicks, tears, or other damage is recorded. The door gasket is then measured to determine the amount of penetration, i.e.,the gasket protrusion as measured from the door face minus the door-face-to-bulkhead gap. If the calculation results are unsatisfactory, the door mechanism is adjusted to account for the permanent set of the gasket. When the permanent set becomes excessive, the gasket is replaced. Specific guidance is provided to the tester to promote maximum scaling, prevent unnecessary wear, and avoid metal-to-metal contact. The final results are verified by plant supervision. [ Reference 14]

The periodic inspections discussed above lead to replacement of the emergency air lock gaskets based on condition and provides reasonable assurance that the intended function of these gaskets will be maintained in the period of extended operation. These inspections are called for by the CCNPP Preventive maintenance Program.

The components that comprise the component type PEN are containment electrical penetrations, containment mechanical penetrations, containment fuel transfer tube / bellows, and containment sump recirculation penetrations. Each of these components have passive intended functions and require AMR.

It should be noted that some of the electrical penetrations that are required to support the EQ intended function are partly addressed in Section 6.3, EQ, of the BGE LRA. General corrosion of these penetrations is addressed in this section of the LRA and radiation damage and thermal damage are addressed in the EQ section of the LRA.

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

e, ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION l

t 3.3A - PRIMARY CONTAINMENT STRUCTURE TABLE 3.3A-1 CONTAINMENT STRUCTURE COMPONENT TYPES REQUIRING AMR i

Component Type Applicable Function i Concrete (including Reinforcing Steel)

Columns 3, 5 Beams 3, 5 Concrete Slabs and Equipment Pads 3, 5 Elevated Floor Slabs 3, 5 Cast-In-Place Anchors

• 3, 5 Grout 3, 5 Sumps 3 Post-Installed Anchors

• 3 Structural Steel Columns

• 3, 5 Beams' 3,5 Baseplates* 3, 5 Floor Framing

• 3, 5 Bracing

• 3, 5 Platform Hangers

• 3, 5 Decking

• 3, 5 Floor Grating

• 3, 5 Checkered Plates

• 3,5 Stairs and Ladders

• 5 Architectural Components Coatings (including galvanizing) 2 Partitions & Ceilings 7 Unique Components Basemat Liner 1 Containment Liner 1 Concrete Basemat 1,2,3,4,5,6,7 Concrete Dome 1,2,3,4,5,7 Concrete Containment Wall 1,2,3,4,5,6,7 Primary Shield Wall 2,3,4 Secor.dary Shield Wall 2,3,4 Refueling Pool Concrete 3 Refueling Pool Liner 1 Refueling Pool Permanent Cavity Seal Ring (PCSR) 1 Removable Missile Shield 2,4,5 Post-Tensioning System 1, 2, 3, 4 Trisodium Phosphate (TSP) Baskets

• 1, 2 Crane Girder

• 5 Lubrite Plates

• 3, 5 Pipe Whip Restraints

• 2 j Emergency Sump Cover and Screen

• 2, 3

(#l-9) numbers correspond to the associated intended functions as listed in Section 3.3A.I.1 ,

indicates that the component type is included under the heading " Steel Components" in Table 3.3 A-3  ;

Application for License Renewal 3.3 A-8 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT STRUCTURE TABLE 3.3A-2 CONTAINMENT SYSTEM COMPONENT TYPES REQUIRING AMR Consponent Type Applicable Function Device Type PEN Electrical Penetrations 1, 9 Mechanical Penetrations 1 l Fuel Transfer Tube / Bellows 1 Emergency Sump Recirculation Penetration 1 l

Device Type DOOR Containment Personnel Air ock 1, 8 Containment Emergency Air lock 1, 8 Containment Equipment Hatch 1, 8

(#1-9) . numbers correspond to the associated intended functions as listed in Section 3.3 A.I.  !

l 3.3A.2 Aging Management The list of potential Age-Related Degradation Mechanisms (ARDMs) identified for Containment Structure and Containment System components is given in Tables 3.3A-3 and 3.3A-4, respectively, with plausible ARDMs identified by a check mark (/) in the appropriate column. [ Reference 5, Attachments 1 and 2; Reference 6, Table 4-2]

For efficiency in presenting the results of these evaluations, structural component type /ARDM combinations are grouped together where there are similar characteristics and the discussion is applicable to the structural components within that group. Exceptions are noted where appropriate. Table 3.3A-2 also identifies the group to which each structural component type /ARDM combination belongs. The following groups have been selected:

Group 1 - Corrosion of tendons /prestress losses; Group 2 - Corrosion of steel; Group 3 - Corrosion of the containment wall and dome liners; Group 4 - Corrosion of the refueling pool liner and permanent cavity seal ring (PCSR); and Group 5 - Weathering of grout.

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

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APPENDIX A -TECHNICAL INFORMATION 3.3A- PRIMARY CONTAINMENT Settlement Industry technical repo ts conclude that settlement is a potentially significant ARDM for pressurized water reactor Containment Structures at some plants. [ Reference 15, Section 5.5] Settlement occurs both during construction and after construction. The amount of settlement depends on the physical properties of the foundation material. [ Reference 5, Appendix JJ Excavation unloading and structural loading during construction caused a small change in the void ratio of undisturbed soil. This change results in a very small or negligible amount of time-dependent settlement. [ Reference 1, Section 2.7.6.2; Reference 5, Appendix J, Section 1.0] Compacted soil is subject to some degree of settlement in the first several months after construction. [ Reference 15, Section 4.5.3.l] 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 table. Sites with soft soil and/or sites with significant changes in underground water conditions over a long period of time may be susceptible to significant settlement. [ Reference 15, Section 4.5.3.2] 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 structure.

When buildings experience significant settlement, cracks in structural members, differential elevations of supporting members bridging between buildings, or both may be visibly detected. [ Reference 5, Appendix J, Section 1.0] At CCNPP, long-term settlement was determined to be not plausible for the Containment Structure based on the following site-specificjustification:

The basemats for the Containment Structures are situated primarily on the site's Miocene deposit, which is an exceptionally dense soil that is capable of supporting loads on the order of 15,000 to 20,000 pounds per square foot (psf). [ Reference 1, Section 2.7.3; Reference 5, Appendix J] The ultimate bearing capacity of the foundation strata is in excess of 80,000 psf, and the allowable bearing capacity is in excess of 15,000 psf. [ Reference 1, Section 2.7.6.2] The design bearing pressure of the basemat for the Containment Structure is 8,000 psf, which is about the same as the removed overburden due to excavation. [ Reference 5, Appendix J, Section 2.1]

e A permanent pipe drain system surrounding the plant is designed to maintain the groundwater table below Elevation 10'-0", which minimizes the fluctuation of the groundwater table, thus providing stable geological conditions around the Containment Structure. Stable geological conditions minimize the susceptibility of the Containment Structure to settlement. [ Reference 5, Appendix J, Section 2.5]. The basemat for the Containment Structure is located between 1%- and 28-feet below the groundwater table. [ Reference 1, Section 2.7.3.2; Reference 1, Figure 5-3]

e The basemat for the Containment Structure tends to uniformly settle as a rigid body. Most of the predicted %-inch settlement is in terms of uniform settlement, which has no adverse effect on structaral components of the Containment Structure. Any differential settlement is expected to be small and have negligible effect. [Refe>ence 5, Appendix J, Section 2.4]

'Ihe 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.

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

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT Group 1 - corrosion of tendons /prestress losses - Materials and Environment Corrosion of tendons and prestress losses are two separate ARDMs grouped together because they both affect the containment Post-Tensioning System. The Post-Tensioning System is designed to contain a total of 876 tendons, including 204 dome tendons,468 hoop tendons, and 204 vertical tendons. Each tendon was designed to contain ninety %-inch-diameter steel wires (American Society for Testing and Materials

[ ASTM] A-421-65T), two anchor heads, and two sets of shims. Each tendon is stressed to 80% of ultimate strength during installation and performs at approximately 50 to 60% of ultimate strength during the life of the structure. He tendon sheathing system consists of spiral wound carbon steel tubing connecting to a trumplate (bearing plate and trumpet) at each end. The sheath was installed for the initial construction concrete pour and does not provide an intended function. The bearing plates were fabricated from steel plate conforming with ASTM A6-66 and the trumpets from American Iron and Steel Institute (AISI)

C1010-C1020 material. After fabrication, the tendon was shop dipped in a corrosion protection material, bagged, and shipped. After installation, the tendon sheathing was filled with a corrosion preventive grease providing the tendon with a grease environment for protecting the sheathing from a corrosive environment.

He ends of all tendons are covered with grease-filled caps for corrosion protection. [ Reference 1, Section 5.1.2.1; Reference 5, Appendices M and N, Sections 2.4]

Group 1 - corrosion of tendons /prestress losses - Aging Mechanism Effects Corrosion of Tendons - When corrosion of prestressing tendons occurs, it is generally in the form f localized corrosion. Most corrosion-related failures of prestressing tendons have been attributed to pitting, stress corrosion, hydrogen embrittlement, or some combination of these. [ Reference 5, Attachment M, Section 1.0]

Pitting is a highly localized form of corrosion. The primary parameter affecting its occurrence and rate is the environment surrounding the metal. The presence of halide ions, particularly chloride ions, is associated with pitting corrosion. [ Reference 5, Attachment M, Section 1.0]

Stress corrosion cracking results from the simultaneous presence of a conducive environment, a susceptible material, and tensile stress. The environmental factors known to contribute to stress corrosion cracking in carbon steels are hydrogen sulfide, ammonia, and nitrate solutions. Prestressed tendon anchor heads, which are constructed of a high strength, low alloy steel bolting material, are vulnerable to stress corrosion cracking. [ Reference 5, Attachment M, Section 1.0]

Hydrogen embrittlement occurs when hydrogen atoms, produced by corrosion or excessive cathodic protection potential, enter the metal lattice. Hydrogen produced by corrosion is not usually sufficient to result in hydrogen embrittlement of carbon steel. Cathodic polarization is the usual method by which this hydrogen is produced. The interaction between the dissolved hydrogen atoms and the metal atoms results in a loss of ductility manifested as brittle fracture. [ Reference 5, Attachment M, Section 1.0]

Corrosion is a plausible aging mechanism for the Post-Tensioning System, including the % inch diameter prestressing wires, the anchor heads, the shims, and the bearing plates, because they could be exposed to a corrosive environment fiom a combination of water and moist air intrusion and inadequate ir.itial grease coverage of wires. [ Reference 8] Corrosion of prestressing wires causes cracking or a reduction in the wires' cross-sectional area. In either case, the prestressing forces applied to the concrete are reduced. If the Application for License Renewal 3.3A-13 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A-TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT prestressing forces are reduced below the design level, a reduction in design margin results. [ Reference 5, Attachment M, Sections 1.0 and 2.5]

Prestress Losses - As the plant ages, tendons that were prestressed during construction tend to lose tension.

Defined as prestress losses, these reductions in tensile force are not readily observable. Several factors contribute to prestress losses:

• Stress relaxation of prestressing wires;

• Shrinkage, creep, and elastic deformation of concrete; e Anchorage seating losses;

• Tendon friction;and e Reduction in wire cross-section due to corrosion that leads to the wire reaching its point of yield.

With the exception of effects due to corrosion-induced wire cross-sectional loss, predictions of prestress losses were calculated during design and margins incorporated at the time of installation of the post-tensioning system to ensure that the Containment Structure can withstand the internal pressure developed during postulated DBEs with no loss ofintegrity. [ Reference 5, Appendix N, Section 1.0]

If the effects of conosion and prestress losses are allowed to progress unmanaged for an extended period of time, these aging mechanisms could affect the ability of the tendons to support the pressure boundary or fission product retention barrier function of the post-tensioning system by seducing its ability to resist loads imposed by design basis events. Other intended functions (stmetural or functional support to SR equipment, shelter / protection of SR equipment, and missile barrier) will not be affected because those functions will be provided by the containment wall itself. [ Reference 5, Appendix N, Section 2.3]

In 1997, during performance of the 20-year Technical Specification tendon surveillance on Unit 1 Containment Structure, conditions which may represent abnormal degradation of the Containment Strucaire were found. During testing of selected vertical tendons to determine the lift-off forces, broken wires were  ;

~

discovered. He discovery of broken wires initiated an expansion of the vertical tendon inspection scope to perform visual inspection and lift-off testing on all vertical tendons for both Units 1 and 2. At the completion of the expanded scope, a number of tendons (32% of the vertical tendons or less than 14% of all the tendons) were identified as having severe corrosion (pitting greater than 0.003 inches) and/or broken wires. [ Reference 8]

A root cause analysis concluded that tendon wire failures and corrosion problems resulted from a I combination of water and moist air intrusion, and inadequate initial grease coverage of wires below the j upper stressing washer. His combination created a corrosive environment, which in turn, caused wire '

failure either by general corrosion or by hydrogen-induced cracking. To slow corrosion and prevent further degradation of the tendon wires, BGE took an immediate shoit-term compensatory action by localized regreasing of the tendon wires and sealing off the potential moisture leak paths. Baltimore Gas and Electric Company has under consideration a number of options for the long-term corrective actions as outlined in the Containment Tendon Engineering Evaluation Report submitted to the NRC on October 28,1997. 1 Baltimore Gas and Electric Company has also completed an inspection of all of the veitical tendons in  ;

Unit 2. He condition of the Unit 2 tendons are similar to the condition of the Unit 1 tendons. The NRC has indicated that a long-term plan with clearly defined and scheduled actions should be in place prior to restart from the Calvert Cliffs Unit I spring 1998 refueling outage. [ Reference 9]

l Application for License Renewal 3.3 A-14 Calvert Cliffs Nuclear Power Plant

ATTACHMENT LM APPENDIX A -TECHNICAL INFORMATION i 3.3A - PRIMARY CONTAINMENT I i

Group 1 - corrosion of tendons /prestress losses - Methods to Manage Aging Mitigation: The effects of tendon corrosion can be mitigated by minimizing the exposure of the post- l tensioning system to moisture. Maintaining a good coating of grease on the tendon steel subcomponents would help protect the tendons. [ Reference 5, Appendix M, Section 2.1] Prestress losses in tendons were considered in the initial design of the prestress tendon system. [ Reference 5, Appendix N, Section 3.0]

Discoverv: The effects of tendon corrosion can be detected through visual examination [ Reference 5, Appendix M, Section 3.0] Prestress losses in tendons can be discovered by periodically measuring and then monitoring the tendon lift-off forces. [ Reference 5, Appendix N, Section 3.0]

Group 1 - corrosion of tendons /prestress losses - Aging Management Program (s)

Mitigation: The design of the containment Post-Tensioning System included provisions for minimizing exposure to water through the use of a petroleum-based grease packed into the tendon sheathing.

Maintenance of the grease quality and extent of coverage is performed through periodic inspections of a sample population of tendons in accordance with the Surveillance Test Procedure (STP)-M-663-1/2,

" Containment Tendon Surveillance." Refer to the discussion below under Discovery for a detailed description of the wiveillance inspection. [ Reference 5, Appendix M, Section 2.4] Since there are no methods recommended to mitigate prestress losses at this time, there are no programs credited with mitigating this ARDM.

Discoverv: A containment tendon surveillance is periodically performed on the Post-Tensioning System which includes visual examination, lift off measurements, wire tensile testing, and analysis of the sheath filler grease. He tendon surveillance is performed in accordance with CCNPP STP-M-663-1 for Unit I and STP-M-663-2 for Unit 2.

Procedure STP-M-663-1 provides instructions for the Unit 1 Containment Tendon Surveillance which includes: [ Reference 16, Section 6.0]

e Determining that for a representative sample of dome, vertical, and hoop tendons, each tendon retains a lift-off force equal to or greater than its lower limit expected range for the time of the test.

• Removing one wire from each of a dome, vertical and hoop tendon checked for lift off forte, and determining the extent of corrosion and the minimum tensile strength.

• Performing a chemical analysis of the sheath filler grease from the selected surveillance tendons to detect changes in its chemical properties.

Le prestressed tendons in CCNPP Unit I containment have been tested at 1,3,5,10,15 and 20 years in accordance with the testing procedure and acceptance criteria specified in STP-M-663-1. For the selected tendon a measurement of the lift-off point pressure is made and converted to lift-off force. This value is compared against a lower bound individual lift-off value. Selected wires are also removed for visual examination and testing. The testing determines the yield strength, ultimate tensile strength, and elongation at ultimate tensile strength. [ References 8 and 16]

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

1 ATTACHMENT (3)

APPENDIX A-TECHNICALINFORMATION I 3.3A - PRIMARY CONTAINMENT .

The visual inspections include an examination of the selected surveillance tendon ends to determine the extent of coverage of the sheathing filler and to detect the presence of water, an examination of all anchorage components for indications of corrosion, pitting, cracking, distortion, or damage, an examination of the surrounding concrete, and an examination of the removed tendon wire for signs of gross corrosion or damage. [ References 16 and 17]

A chemical analysis of the sheath filler grease is performed as part of STP-M-663-1 for meeting the inspection requirements of Regulatory Guide 1.35, Revision 2. One sheathing filler grease sample is obtained from each surveillance tendon to be tested for chemical analysis in accordance with ASTM i standards. Results of this analysis are evaluated to ensure that the concentration of water soluble impurities and water in the grease sample do not exceed the criteria for chlorides, nitrates, sulfides and water, j

[ Reference 5, Attachment 5; References 16 and 17] l Since both Units' initial one year surveillances, all testing is typically conducted on Unit 1 and visual examination and sheath filler grease analysis is typically conducted on both units. The visual examination and sheath filler grease analysis for Unit 2 is accomplished as described above for Unit 1. [ References 16 l and 17]

The prestress force data and physical condition data obtained during each surveillance test is evaluated in accordance with the guidance in Position 7 of Regulatory Guide 1.35, Revision 2, so that the integrity of  !

the prestressed tendon system is ensured. The prestressed tension system is a passive, and highly  !

redundant system. Historical data from CCNPP and from the industry reported very few incidents of random malfunction of the tendons or its components. The criteria provided in the Regulatory Guide 1.35, Revision 2, and adopted by CCNPP will ensure that the tendon system will perform its i intended functions through the time interval to the next surveillance. [ Reference 5, Attachment 5; References 16 and 17]

The program was altered in 1983 as a result of a Technical Specification chang due to the issuance of Regulatory Guide 1.35, Revision 3. The changes were minor and affected the surveillance sample size  ;

and the Technical Specification value to which the surveillance results were . compared. Another Technical Specification change will result from the new rule recently listed in 10 CFR 50.55(a),

incorporating ASME Section XI, Subsection IWE/IWL requirements. Under the new rule, the units will be tested alternately such that 5 tendons (currently 3) per group (hoop, dome, and vertical) are tested in I one unit while only a visual inspection is performed for the other unit. In the unit that is tested, the tendon forces are to be measured, one of each type detensioned for wire sample removal, and chemical and material analysis performed on these samples. The visual inspection in the other unit consists of removing end caps, checking the tendon condition and regreasing. Then during the next surveillance, the units will be reversed for tendon testing and visual inspection.

The tendon surveillance inspection program must be revised to extend the lift-off force versus time curve for a 60-year operating life. As a result of this curve revision, the retensioning of selected tendons may be required to meet its resultant revised lift-off force requirements. The existing Technical Specification lift-off force curves were developed for 40 years of operation and do not provide acceptance criteria for any extended period of operation. This is a Time-Limited Aging Analysis issue that will be addressed by re-evaluating the existing curves to reflect i required prestress levels and acceptance criteria for the renewal period. [ Reference 18]

3 Application for License Renewal 3.3 A-16 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (3)

APPENDIX A-TECHNICALINFORMATION 3.3A- PRIMARY CONTAINMENT Group 1 - corrosion of tendons /prestress losses - Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of tendons and prestress losses:

Containment tendons support the containment pressure boundary or a fission product retention barrier functions; therefore, their integrity must be maintained under current licensing basis (CLB) design loading conditions.

Containment tendons are susceptible to corrosion and prestress losses, which can affect the ability of the tendons to wpport the pressure boundary or a fission product retention barrier functions.

Other intended functions (structural or functional support to SR equipment, shelter / protection of SR equipment, and missile barrier) will not be affected because those functions will be provided by the containment wallitself.

Grease coatings mitigate the effects of corrosion by providing a protective layer, preventing moisture and oxygen from contacting the steel.

A containment tendon surveillance (STP-M-663-1/2) is periodically performed on a sample population of tendons that includes visual examination, lift-off measurements, wire tensile testing, ,

and analysis of the sheath filler grease. '

The existing tendon lift-off force curves will be re-evaluated to reflect the required prestress levels for the period of extended operation.

Therefore, there is reasonable assurance that the effects of corrosion and prestress losses of the containment Post-Tension System will be managed in such a way as to maintain the structures' integrity, consistent with the CLB, during the period of extended operation.

Group 2 -(corrosion of steel)- Materials and EnviroAnent Group 2 is comprised of components that are fabricated from steel, which corrodes in the presence of moisture and oxygen as a result of electrochemical reactions. The Containment Structure component types listed in Table 3.3A-1 and marked with an asterisk are all included within this group. These structural steel components were shop-painted or field-painted during the construction phase, with the exception of grating, wire mesh, checkered plates, and metal decking, which are constructed of galvanized or stainless steel. [Refei:nce 5, Attachment 2 and Appendix K, Sections 1.0 and 2.4]

l The following Containment System components are also included in Group 2; containment electrical penetrations, containment mechanical penetrations, containment fuel transfer tube / bellows, containment emergency sump recirculation penetrations, containment personnel air lock, containment emergency air lock, and containment equipment hatch. The electrical penetrations have subcomponents constructed of carbon steel, stainless steel, and non-metallic materials, i.e., epoxy, sealants, and adhesives. The mechanical penetrations, containment personnel air lock, containment emergency air lock, and containment equipment hatch are constructed of carbon steel. The containment fuel transfer tube, containment sump recirculation penetrations, and TSP baskets are constructed of stainless steel and the containment fuel transfer tube bellows is constructed ofinconel. [ Reference 6, Attachments 3]

The environment to which these components are subjected varies with their location. In the Containment Structure and Auxiliary Building (where containment penetrations are located), a climate-controlled Application for License Renewal 3.3 A-17 Calvert Clifts Nuclear Power Plant

ATTACHMENT m '

APPENDIX A - TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT  !

environment is normally maintained. The ambient temperature is controlled by a plant ventilation system as described in UFSAR Chapter 9. [ Reference 1, Table 9-18] The steel components located outdoors will be subject to the temperature and humidity changes, rain, snow, etc. expected at the CCNPP site. In those places where pockets that can harbor liquids are formed by structural components, steel may be subjected to standing water, which is, comtined with oxygen, a corrosive environment.

[ Reference 5, Appendix K, Section 2.1] Penetrations, air locks, and hatches can be exposed to conditions .

that cause condensation (warm air or water flowing through the component). In the presence of oxygen, such condensation can lead to corrosion if the surface coating is degraded. Some steel components may be exposed to elevated temperatures that could cause the coating to fail. [ Reference 6, Attachments 6]

Group 2 -(corrosion of steel)- 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, alternating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from seawater, the atmosphere, or groundwater, increase 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 may accelerate.

Corrosion is plausible for all components and subcomponents constmeted of carbon steel or galvanized steel. Corrosion is not plausible for the containment fuel transfer tube and bellows, containment emergency sump recirculation penetrations, and TSP baskets because they are constructed of stainless steel or Inconnel, which are highly resistant to general corrosion. Corrosion is also not plausible for the containment emergency sump cover and screen mesh and grating because they are constructed of stainless steel. Corrosion is plausible for the containment emergency sump cover and screen structural steel because it is constructed of carbon steel. If corrosion is left unmanaged for an extended period of time, the resulting loss of material could lead to the inability of the Group 2 steel components to perform their intended functions under CLB design loading conditions. [ Reference 5, Attachment 2 and Appendix K, Section 1.0; Reference 6, Attachments 5 and 6]  ;

In some cases, corrosion of carbon steel that is in contact with water may be microbiologically induced due to the presence of certain organisms. These organisms, which include microscopic forms such as bacteria l

and macroscopic types such as algae and barnacles, may influence corrosion of steel under broad ranges of j pressure, temperature, humidity, and pH. Microbiologically-induced corrosion is plausible for structural i steel components where water may collect and could result in random pitting and general corrosion.

[ Reference 5, Appendix K, Sections 1.0 and 2.1] Microbiologically-induced corrosion is not plausible for l

the electrical penetrations, mechanical penetrations, containment sump recirculation penetrations, containment personnel air lock, containment emergency air lock, and containment equipment hatch because they are not wetted surfaces. Microbiologically-induced corrosion is not plausible for the l containment fuel transfer tube due to insufficient tensile stresses, use of corrosion resistant materials,  !

I controlled refueling water chemistry, and because it is maintained dry except during refueling periods.

[ Reference 6, Attachments 5 and 6] If microbiologically-induced corrosion is left unmanaged for an  !

extended period of time, the resulting loss of material could lead to the inability of the Group 2 steel components to perform their intended functions under CLB design loading conditions.

Corrosion products such as hydrated oxides ofiron (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, corrosion may cause the protective Application for License Renewal 3.3 A-18 Calvert Cliffs Nuclear Power Plant i

v l

ATTACHMENT m t

APPENDIX A -TECHNICALINFORMATION 3.3A - PRIMARY CONTAINMENT coatings to lose their ability to adhere to the corroding surface. In this case, damage to the coatings can be l visually detected well in advance of significant degradation of the steel. [ Reference 5, Appendix K, Section 1.0)

In 1992, a visual inspection was performed for the interior of containment, including some containment

.' penetrations, structural steel floor frames, and equipment support interfaces. The coatings were observed to be in good condition and only minimal corrosion was observed. The exterior of the Containment Structure was also visually inspected. Some minor rust stains were observed on the concrete from metal components such as uncoated steel anchors and skyclimber structural steel suppons. There was also evidence of some deterioration of the protective coating on the buttress bearing plates and tendon grease caps. The methodology employed for the inspection was consistent with the industry standard, " Rules for Inservice Inspection,Section XI, ASME Boiler and Pressure Vessel Code," which gives rules for the inspection of concrete. Under these rules, the exterior and interior surfaces of the Containment Structure and components in the Containment System were found to be in good to excellent condition. The responsible engineer determined by visual examination that there was no evidence of damage or degradation sufficient to warrant further evaluation or repair.

A routine walkdown of containment penetrations was also performed in the Unit 2 Containment in 1995 in accordance with Administrative Procedure MN-3-100, " Painting and other Protective Coatings."

Some penetrations showed indications of rust / scale / corrosion inside the containment. An issue Report was initiated in accordance with the CCNPP Corrective Actions Program, and it was determined that the corrosion would be monitored and corrected under MN-3-100. In 1997, the rust / scale / corrosion was removed and the surfaces repainted in accordance with MN-3-100.

l Group 2 -(corrosion of steel)- Methods to Manage Aging Mitiption: The effects of corrosion can be mitigated by minimizing the exposure of external surfaces of steel to an aggressive env.' ,nment md protecting the external surfaces with paint or other protective coating. Coatings serve as a protective layer, preventing moisture and oxygen from directly contacting l the steel surfaces.

Discoverv:- The effects of corrosion of steel are detectable by visual inspection. 'Ihe external metal surfaces of structural steel components are covered by a protective coating, and observation that significant degradation has not occurred to this coating is an effective method to ensure that corrosion l

has not affected the intended function of the structural component. Coatings degrade slowly over time, i- allowing visual detection as pan of routine walkdowns during normal plant operations. Coatings that are l blistered or showing rust stains have degraded to the point where the steel is being exposed to moisture and oxygen. Since the coating does not contribute to the components' intended functions, degradation of the coating provides an aled condition that triggers corrective action before corrosion that affects the components' ability to perform its intended function can occur. The degradation of the protective coating that doer occur can be discovered and monitored by periodically inspecting the carbon steel structural components and by carrying out corrective action as necessary. [ Reference 5, Attachment 8 and Appendix K, Section 3.0]

1 l

l Application for License Renewal 3.3A-19 Calven Cliffs Nuclear Power Plant L

4 ATTACHMENT m APPENDIX A -TECHNICAL INFORMATION 3.3A- PRIMARY CONTAINMENT

' Group 2 - (corrosion of steel)- Aging Management Program (s)

Mitioatina: 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 ensure that the protective coatings of carbon steel structural components are maintained.

Discoverv: Corrosion of steel can readily be detected for Group 2 components through visual

, examination. Additionally, degradation of protective coatings, which help mitigate corrosion, can also

. be visually detected so that corrective actions can be taken to restore the coatings. An inspection prograrn can provide the assurance needed to conclude that the effects of corrosion are being effectively managed for the period of extended operation. Routine walkdowns of the Primary Containment would provide for discovery and management of corrosion of Group 2 components. Because the wire mesh on the containment emergency sump cover and screen blocks the view of the structural steel supports, a separate visual examination is conducted for that component. [ Reference 5, Appendix K, Attachment 4; Reference 6, Attachment 2]

Structure and System WaHeAnwns Periodic walkdowns that are performed for CCNPP structures and systems provide cpportunities to visually inspect the condition of plant equipment and to identify degraded conditions. Two procedures that specifically control walkdown activities that could identify and correct potential corrosion or degraded protective coatings for the Containment Structure and Containment System are CCNPP Administrative Procedures MN-3-100 and MN-1-319, " Structure and System Walkdowns." Each of these procedure:: are discussed below.

Administrative Procedure MN-3-100 provides for discovery of corrosion of steel or of conditions that would allow' corrosion to occur, such as degraded paint, for the inside of containment through performance of visual inspections during plant walkdowns. The purpose of Procedure MN-3-100 is to control painting and protective coatings activities performed inside containment to ensure they comply with Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to {

Water-Cooled Nuclear Power Plants," and American National Standards Institute (ANSI) N101.4 - 1972,

'" Quality Assurance for Protective Coatings Applied to Nuclear Facilities." Containment walkdowns performed in accordance with this procedure are credited for discovery and management of corrosion of l

Group 2 steel components inside of containment. [ Reference 5, Appendix K; Reference 19, Section 1.1] )

Procedure MN-3-100 requires that the responsible engineer perform a walkdown of the inside of  !

containment at the start of each scheduled refueling or maintenance outage to verify the condition of all Service Level I coatings. Service Level I coatings are those where failure could adversely affect the operation of post-accident fluid systems and, thereby, impair safe shutdown. MN-3-100 also controls the identification of correct Service Level and specifies minimum requirements for coating quality and work practices. [ Reference 19, Sections 3,5.2 and 5.3]

During the containment walkdown, a general visual inspection is performed on all readily accessible surfaces. A more thorough inspection is performed on all coatings near sumps or screens associated with l the Emergency Core Cooling System. The inspector develops a list of all areas inside containment  !

exhibiting deterioration. Repair areas are evaluated to ensure timely corrective action is taken. All )

routine and restorative coatings work is prioritized and implemented in accordance with the CCNPP Corrective Actions Program. [ Reference 19, Section 5.2]

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

=

i ATTACHMENT Lh t

[ APPENDIX A -TECHNICAL INFORMATION 3.3A- PRIMARY CONTAINMENT Controls for painting and other protective coatings have been in existence since initial plant startup.

' During that time, changes to the program have been made largely due to an existing qualified paint becoming unavailable. There have been no changes made specifically to address particular aging-related or coating-related problems or failures.

Administrative Procedure MN-1-319 provides for discovery of corrosion of steel or of conditions that would allow corrosion to occur, such as degraded paint, for the outside of containment through performance of visual inspections during plant walkdowns. The purpose of the procedure is to provide direction for the performance of structure and system walkdowns and for the documentation of the walkdown results. System walkdowns are credited for discovery and management of corrosion of Group 2 steel components outside of containment. [ Reference 6, Attachment 1; Reference 201 Under MN-1-319, responsible personnel perform periodic walkdowns of their assigned m ctures 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. Currently, i structure walkdowns should be performed every refueling outage and scheduled to ensure that every structure will receive a walkdown as a minimum every third outage. [ Reference 20, Sections 5.1 and 5.3]

One of the objectives of the walkdowns 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. Conditions adverse to quality are documented and resolved by the CCNPP Corrective Actions Program and in accordance with MN-3-100. [ Reference 20, Sections 5.1.C,5.2.A.1, and 5.2.A.5]

The walkdown procedure 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 20, 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, anchor bolt degradation, or leakage of fluids; and e Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage of fluids, presence of standing water or accumulated moisture, or inadequate support of components (e.g., missing, detached , or loose fasteners and clamps).

The procedure includes a walkdown checklist specifically for the Containment Structure. The checklist includes a section targeted at structural steel components. Checklist items include visual inspection for i corrosion, rust stains, and flaking / bubbling of protective coatings. A checklist for moisture barriers includes visual inspection for standing water, water intrusion, water marks, or corrosion of metal components. [ Reference 20, Attachment 3]

The rtructure and system walkdowns enhance the familiarity of responsible personnel with their assigned systems and structures and provides extended attention to plant material co6Sion beyond that afforded Application for License Renewal 3.3 A-21 Calvert Cliffs Nuclear Power Plant

4.

ATTACHMENT (3)

APPENDIX A-TECHNICALINFORMATION 3.3A- PRIMARY CONTAINMENT by Operations and Maintenance personnel alone. The procedure has been improved recently through incorporation of significant additional guidance on specific activities to be included in the scope of the structures walkdowns. Further enhancements will be made to provide guidance to help the person performing the walkdown in determining whether the intended functions will continue to be met as required by the CLB and guidance regarding approval authority for significant departures from scope or schedule. ,

The walkdowns described above will ensure that degraded conditions due to corrosion of steel are identified and corrected such that Group 2 components will be capable of performing their intended functions under all CLB conditions.

i Emergency Sumo Cover and Screen Insnection The containment emergency sump cover and screen contains a wire mesh for trapping debris that could j potentially be swept up by the safety injection pumps when containment spray is in the recirculation mode. This wire mesh obstructs the view of the structural steel that frames the grating and mesh screen.

Therefore, the routine walkdowns performed under MN-3-100 are not relied on for discovery of corrosion for this component. A Surveillance Test Procedure, STP-M-661-1/2, " Containment l Emergency Sump Inspection," is performed every refueling outage in accordance with plant Technical Specifications, and is credited for the required visual inspection.

i Procedure STP-M-661-1/2 provides for discovery of corrosion of the structural steel components of the cover and screen through performance of visual inspections. These inspections would also detect any degraded paint conditions that, if left uncorrected, could lead to the steel being exposed to a corrosive  ;

environment. Specifically, the procedure requires an inspection for signs of corrosion, debris, or structural distress to the screens. Any deficiencies are noted and corrective actions initiated in accordance with the CCNPP Corrective Actions Program. The visual inspections performed by STP-M-661-1/2 will ensure that degraded conditions due to corrosion of the cover and screen structural steel are detected and corrected such that the cover and screen will be capable of performing its intended function under all CLB conditions. [ References 21 and 22]

, Group 2 -(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 Group 2 components:

• The structural steel components, penetrations, air locks, and hatches that comprise Group 2 provide passive intended functions (refer to Table 3.3A-1) and their integrity must be maintained under all CLB conditions.

• Corrosion of steel is plausible due to possible degradation of the external protective coatings and exposure to water and oxygen.

• Periodic visual inspections of coated surfaces inside of containment will continue to be performed in accordance with MN-3-100. Procedure MN-3-100 also controls the assessment, prioritization, and corrective action for degraded coatings discovered outside containment. Corrective actions are taken in accordance with the CCNPP Corrective Actions Program.

• Periodic visual inspections of coated surfaces outside of containment will continue to be performed in accordance with CCNPP Administrative Procedure MN-1-319. The procedure will Application for License Renewal 3.3A-22 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A-TECHNICAL INFORMATION 3.3A-PRIMARY CONTAINMENT be modified to provide guidance to assist in functional adequacy determinations and for authority to deviate from scope or schedule. These walkdowns will identify and document significant coating degradation and/or presence of corrosion.

Periodic visual inspections of the structural steel members of the containment emergency sump cover and screen will continue to be performed in accordance with CCNPP Surveillance Test Procedures STP-M-661-1/2. Any significant coating degradation or corrosion will be discovered and corrective actions will be taken in accordance with the CCNPP Corrective Actions Program.

Therefore, there is reasonable assurance that the effects of aging due to corrosion of steel will be managed in such a way that Containment Structural and System components will be capable of performing their intended functions consistent with the CLB during the period of extended operation.

Group 3 -(corrosion of containment wall and dome liners)- Materials and Environment The containment wall and dome liners at CCNPP are ASTM A36 carbon steel. These liners were constructed from a series ofindividual steel plates welded together. Both the plate material and the welds are subject to the same potential degradation mechanisms. The significance of potential degradation of the liners is considered to apply equally to the plate material and the welds. For corrosion protection, the inside face of the containment wall and dome liners were covered with a protective coating during the construction phase. The containment wall and dome liners do not have dissimilar metals; therefore, they are not subject to galvanic corrosion. [ Reference 5, Attachment L, Sections 2.0 and 2.4]

The containment wall and the containment dome are 3'-9" and 3'-3" thick, respectively, and are subject to compressive stress due to dead weight and prestress load under normal plant operating conditions. Any cracks that do occur would be more tightly closed due to the prestress forces, allowing less potential for groundwater to penetrate to the containmee wall liner. This configuration minimizes cracks in the concrete that allow penetration of moisture, oxygen, and chlorides, which cause corrosion degradation. , Therefore, the containment wall liner from the concrete side and the liner anchors are not exposed to aggressive chemicals from the outside environment, such as Md rain, salt-cestaining atmospheres, and groundwater.  !

[ Reference 5, Attachment L, Sections 2.1.1 and 2.5]

The interior surfaces of the containment wall and dome liners are exposed to the containment internal environment. The Primary Containment Heating and Ventilation System maintains a climate-controlled environment inside the containment with design conditions as described in Table 9-18 of the UFSAR.

Corrosion of the internal surfaces could occur in the presence of moisture and oxygen as a result of electrochemical reactions unless the existing coating is maintained by an effective coating management program. [ Reference 5, Attachment L, Section 2.1.1]

Group 3 - (corrosion of containment wall and dome liners) - Aging Mechanism Effects Carbon 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 ifit is not disturbed by erosion, alternating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from bay water, the atmosphere, or groundwater, increase the rate of corrosion by increasing the electrochemical activity.

Corrosion products such as hydrated oxides ofiron (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 Application for License Renewal 3.3A-23 Calvert Cliffs Nuclear Power Plant i

I

s. 4- j i

A*ITACHMENT- Lh  !

j APPENDIX A-TECHNICAL INFORMATION 3.3A -PRIMARY CONTAINMENT 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 can be visually detected well in advance of significant degradation of the steel. [ Reference 5, Appendix K, Section 1.0 and Appendix L, Section 1.1]

Corrosion is plausible for the internel surfaces of the containment wall and dome limrs because they are potentially subject to moisture and oxygen and are constructed of carbon steel. If corrosion of the internal surfaces of the containment wall and dome liners is left unmanaged for an extended period of time, the resulting loss of material could lead to the inability of these steel components to perform their intended functions under CLB design loading conditions. [ Reference 5, Appendix K, Section 2.3]

Group 3 - (corrosion of containment wall and dome liners)- Methods to Manage Aging Mitigation: The effects of corrosion can be mitigated by minimizing the exposure of external surfaces of steel to an aggressive environment by protecting the external surfaces with paint or other protective coating. Maintaining the protective coating on the interior surfaces of the liner prevents moisture and oxygen from directly contacting the steel surfaces of the containment wall and dome liners, l

(Reference 5, Appendix K, Section 2.6]

j Discoverv: The effects of corrosion of steel are detectable by visual inspection. A visual examination by a person familiar with the liners 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 can trigger 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 through visual inspections.

Corrective actions for failed protective coatings and any actual metal degradation can be carried out as necessary. [ Reference 5, Appendix L, Section 3]

Group 3 - (corrosion of containment wall and dome liners)- Aging Management Program (s)

Mitigation: The containment wall and dome liners are covered by protective coatings that mitigate the  !

effects of corrosion. The discovery programs discussed below ensure that the protective coatings of carbon steel structural components are maintained. [ Reference 5, Appendix L, Section 3]

Discoverv: Corrosion of the containment wall and dome liners can readily be detected through visual examination. Additionally, degradation of protective coatings, which help mitigate corrosion, can also be visually detected so that corrective actions can be taken to restore the coatings. An inspection program can provide the assurance needed to conclude that the effects of corrosion are being effectively managed for the period of extended operation. [ Reference 5, Appendix L, Section 3]

l An examination of the containment liner plate is periodically performed in accordance with CCNPP '

Surveillance Test Procedure STP-M-665-1, " Containment Liner Plate Surveillance," for Unit I and STP-M-665-2, " Containment Liner Plate Surveillance," for Unit 2. STP-M-665-1/2 are currently performed as specified in the Containment Leakage Rate Testing Program, i.e., prior to the performance of an Integrated Leak Rate Test and during two other refueling outages before the next Integrated Leak Application for License Renewal 3.3A-24 Calvert Cliffs Nuclear Power Plant

.. s l ATTACHMENT m l

APPENDIX A-TECHNICALINFORMATION 3.3A - PRIMARY CONTAINMENT Rate Test, .in accordance with ASME Section XI,1992 Subsections IWE and IWL, if the interval for the Integrated Leak Rate Test has been extended to 10 years. The tests specifically address Technical Specification Surveillance Requirement 4.6.1.6.3 to visually inspect the exposed accessible interior and exterior surfaces of the Containment Structure. If any portion of the liner plate shows any signs of bulges, wrinkles, cracks, corrosion, flaking paint, or indication of other types of deterioration, corrective actions are required. Corrective actions are completed in accordance with the CCNPP Corrective Actions Program. [ References 23,24, and 25]

Recent inspections of the containment liners in accordance with STP-M-665 revealed that the liner plate was in good condition, except in the immediate vicinity of the expansion joint between the floor slab and liner plate. After approximately 20 years of operation, the original joint sealant, Thiokol Polysulfide, was deteriorating. B!istering of the paint and corrosion of the liner was observed at the joints as evidence of water intrusion. No through-liner flaws were found. Due to these discoveries, the Thiokol Polysulfide base sealant is being replaced with a liigh Density Silicone Elastomer (HDSE) that provides an effective barrier against water as well as smoke, gas, pressure, and fire. High Density Silicone Elastomer has high thermal stability and structural strength. Besides using a different sealant material, the design of the joint seal is also being modified. Originally, the Thiokol Polysulfide base sealant was applied to a shallow depth at the top of the compressible material in thejoints and made flush with the nominal 10-foot elevation base slab. He new HDSE sealant is being used to form a small curb above the joint to shed water in addition to providing a seal. Also, to improve the seal, the HDSE is being placed a minimum of three inches into the joint after removing some compressible material. A polyethylene backer rod is being placed in the joint between the HDSE and compressible material to separate them. Any required repair, cleaning, and repainting of the liner is being performed as part of this activity. This repair was already performed on the Unit I moisture barrier during the 1996 refueling outage. At that time, all of the liner plate was determined to be of adequate thickness in the area of the moisture barrier. Repairs to the expansion joints in Unit 2 are currently scheduled for the refueling outage in 1999. Hus, based on these actions, it can be concluded that the existing program, STP-M-665, adequately inspects the liner in the area of the seal and compressible material.

Group 3 - (corrosion of containment wall and dome liners)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of the containment wall and dome liners:

• The containment wall and dome liners support a pressure boundary or fission product retention barrier function to protect the public health and safety in the event of any postulated DBEs, and I their integrity must be maintained under all CLB conditions. I e Corrosion of the inside surface of the containment wall and dome steel liner plates is plausible due to possible degradation of the external protective coatings and exposure to water and oxygen. l If left unmanaged, corrosion could eventually result in the liner plates not being able to perform  !

their intended functions under CLB conditions f

!

• Periodic visual inspections of the containment wall and dome liners will continue to be performed l in accordance Surveillance Test Procedures STP-M-665-1/2. These inspections will identify and I document significant coating degradation and/or presence of corrosion. Corrective actions are l taken in accordance with the CCNPP Corrective Actions Program.

l Application for License Renewal 3.3 A-25 Calvert Cliffs Nuclear Power Plant

s. t ATTACHMENT m APPENDIX A-TECHNICALINFORMATION 3.3A - PRIMARY CONTAINMENT Therefore, there is reasonable assurance that the effects of aging due to corrosion of the containment wall and dome liners will be managed in such a way that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation.

Group 4 - (corrosion of the refueling pool liner and PCSR)- Materials and Environment ne refueling pool liner at CCNPP is constructed of Type 304 stainless steel material. The liner was constructed from a series ofindividual steel plates welded together. All exposed metal parts of the PCSR are constructed of Type 304 stainless steel. The hatch bolts for the PCSR are stainless' steel. He refueling pool liner and PCSR do net have dissimilar metals; therefore, they are not subject to galvanic corrosion.

[ Reference 1, Section 1 L2.2.4; Reference 5, Attachment L, Sections 2.0 and 2.4]

The stainless steel liner and PCSR are not load-bearing structural components, since the induced strains in the liner and PCSR are negligible under normal plant operating conditions. The heat affected zones of the welds are potential sites for " sensitization." The internal surfaces of the refueling pool liner and PCSR are normally dry. However, during refueling outages they are exposed to wetcr that contains boric acid. Sensitized Type 304 stainless steel is susceptible to IGSCC in boric acid solution. [ Reference 5, Appendix L, Section 2.1.3] 1 Group 4 -(corrosion of the refueling poolliner and PCSR)- Aging Meehanism Effects SA-240 Type 304 stainless steel is resistant to electrochemical corrosion in the refueling pool environment.

The corrosion rate of this steel ranges from 0.05 mil in 100 years (virtually no corrosion) to less than 0.001 mil per year in a borated fuel pool water environment. Therefore, the electrochemical corrosion is negligible for the refueling pool liner and PCSR. Furthermore, since the liner and PCSR are not exposed to a corrosive environment and the induced strains are negligible under normal operating conditions, the conditions for stress corrosion cracking to occur do not exist for the refueling pool liner or PCSR.

[ Reference 5, Appendix L, Section 3.1.3)

The heat affected zones of the welds are potential sites for sensitization. This is because of the changes in the microstructure that take place due to the welding heat, rendering the heat affected zones sensitized, and because of the potential for high residual stresses in and around the welds. Sensitized Type 304

' stainless steel may be susceptible to IGSCC in boric acid solution (13,000 ppm) at temperatures of 180'F and low pH (less than 4). Conditions that may contribute to the occurrence ofIGSCC include elevated temperatures, chloride content, boric acid concentration, oxygen concentration, and degree of sensitization. [ Reference 5, Appendix L, Section 2.1.3; Reference 26, Section 4.5.1.1]

Intergranular stress corrosion cracking of the stainless steel liner and PCSR would be expected to result in initiation and propagation of a crack that can eventually lead to detectable leakage. If IGSCC of the refueling pool liner and PCSR is left unmanaged for an extended period of time, the resulting cracks could lead to the inability of these stainless steel components to perform their intended functions under CLB design loading conditions. [ Reference 5, Appendix L, Sections 2.1.3 and 2.3]

The general condition of the refueling pool is good with no apparent signs of aging. There has been minimal leakage to date, where at one location in Unit 2, it may originate from drain pipes, not from the liner or PCSR. The exact location of this leakage is still being investigated. However, the leakage was

. evaluated for its effects on the rebar in the concrete and was shown to have a negligible effect on slab Application for License Renewal 3.3A-26 Calvert Cliffs Nuclear Power Plant

u < l ATTACHMENT (3J APPENDIX A-TECHNICALINFORMATION I 3.3A - PRIMARY CONTAINMENT strength. The refueling pool drain pipe is addressed in Section 5.18, " Spent Fuel Pool Cooling System," of the BGE LRA. One additional leak is currently being investigated and the leak source will be repaired if necessary. The refueling pool is typically dry and is filled during refueling outages only. There has been no evidence ofIGSCC identified in the liner or PCSR welds. A visual inspection of two welds closest to the highest stress region of the PCSR was performed one cycle (two years) after the PCSR was installed. No service-induced indications were discovered.

Group 4 -(corrosion of the refueling poolliner and PCSR)- Methods to Manage Aging Mitigation: The effects ofIGSCC can be mitigated by minimizing the exposure of external surfaces of steel to an aggressive environment. However, the discovery methods described below are deemed adequate to manage these ARDMs.

Discoverv: The effects of IGSCC of heat affected welds could result in material cracking and could cause the refueling pool liner or PCSR to leak. Because the liner and PCSR only serve as fluid retaining boundaries and do not provide a structural integrity function, detecting and, if needed, measuring and trending leakage from the refueling pool provides for effective aging management. Corrective measures can be taken to restore the refueling pool's integrity if any significant leakage is detected. [ Reference 5, Appendix L, Section 3]

Group 4 -(corrosion of the refueling poolliner and PCSR)- Aging Management Program (s)

Mitigation: Since there are no methods recommended to mitigate corrosion at this time, there are no programs credited with mitigating this ARDM. [ Reference 5, Appendix L, Section 3]

Discoverv: Detecting, and if needed, measuring and trending leakage from the refueling pool and PCSR ,

provides for discovery of IGSCC so that corrective measures can be taken prior to loss of intended  !

function. Routine inspections are performed on system components in accordance with CCNPP Administrative Procedure MN-1-319. These walkdowns provide for discovery and management of the effects of corrosion through visual inspections, reporting any leakage detected, and initiating corrective action. Under this procedure, any evidence of fluid leakage would be considered adverse to quality and, therefore, addressed in accordance with the CCNPP Corrective Actions Program. [ Reference 20] Refer ,

to the discussion on Aging Management Programs for Group 2 for a detailed description of MN-1-319.

Group 4 -(corrosion of the refueling poolliner and PCSR)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of the refueling pool liner and PCSR:

1

• The refueling pool liner and PCSR support a pressure boundary or fission product retention '

barrier function to protect the public health and safety in the event of any postulated DBEs and their integrity must be maintained under all CLB conditions.

e IGSCC on the inside surface on the refueling pool liner plates or PCSR is plausible due to potential sensitization of the steel at weld locations and a corrosive environment of borated water. Ifleft unmanaged, IGSCC could eventually result in the refueling pool liner or PCSR not being able to perform their intended function under CLB conditions Application for Licendnewal 3.3A-27 Calie'rt Cliffs Nuclear Power Plant

ATTACHMENT m i

APPENDIX A-TECHNICALINFORMATION l 3.3A - PRIMARY CONTAINMENT Periodic walkdowns of the containment will continue to be performed in accordance with Procedure MN-1-319. These inspections will identify and document the presence of leaks that may be due to IGSCC. I l

l Therefore, there is reasonable assurance that the effects of aging due to IGSCC of the refueling pool liner and PCSR will be managed in such a way that they will be capable of performing their intended function consistent with the CLB during the period of extended operation.

~

Group 5 -(weathering of grout)- Materials and Environment The Containment Structures at CCNPP are prestressed, post-tensioned concrete structures. During the post- i tensioning activities during initial construction, reports were made that 11 of the top bearing plates of the l venical tendons on Unit I and I of these plates on Unit 2 depressed into the concrete. Further investigations verified that voids exist under many of the vertical tendon bearing plates. Refer to Appendix SD," Study of Upper Vertical Tecdon Bearing Plates," of the UFSAR for a detailed discussion of the vertical tendon bearing plate repairs. Grout was used to make repairs to all bearing plates that showed indications of voids in the concrete below. [ Reference 1, Appendix SD.l.1]

Since the grout is located on top of the Containment Building, it is subject to the normal outside atmosphere conditions at the CCNPP site. The 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 site. [ Reference 5, Appendix 0]

Group 5 -(weathering of grout)- Aging Mechanism Effects ne grout at the containment tendon bearing plates is exposed to outdoor weather conditions and is susceptible to weathering. Aging mechanisms associated with weathering include exposure to sunlight (ultraviolet exposure), changes in humidity, ozone cycles, temperature and pressure fluctuations, and snow, rain, or ice. The effects of weathering on grout is evidenced by cracking and spalling. [ Reference 5, Appendix 0]

He durability of grout may affect the passive intended function it provides. Although aggressive environments may contribute to deterioration of the grout, most degradation results from water entering cracks in and around the grout and freezing. His is an ongoing process throughout the plant's life that results in cracking and spalling at susceptible locations. Some of the grout locations are in flat areas of the structure and are exposed to standing water.

An inspection of the Unit 1 Containment Structure was performed in 1992, which included a representative sample of the grout at the vertical tendon base plates. It was observed that the mortar generally showed some tight cracks, visually estimated to be about 0.005 inch, which is just big enough to show a moisture trail when water is present. He mortar appeared to be in good condition with no immediate repairs needed.

l l

1 Application for License Renewal 3.3A-28 Calvert Cliffs Nuclear Power Plant

ATTACHMENT G)

APPENDIX A -TECHNICAL INFORMATION 3.3A- PRIMARY CONTAINMENT Group 5 -(weathering of grout)- Methods to Manage Aging Mitigation: No methods for mitigating weathering of grout are needed; the discovery methods described below are deemed adequate to manage this ARDM.

Discoverv: The effects of weathering of grout are detectable by visual inspection. Periodic visual examinations would provide for discovery of degraded grout so that corrective actions could be taken to preclude the effects of weathering from affecting the intended function of the grout.

Group 5 -(weathering of grout)- Aging Management Program (s)

Mitigation: Since no methods for mitigation are needed at this time, there are no mitigation programs required.

Discoverv: The effects of weathering on the containment tendon baseplate grout is readily detected through visual observation. A containment tendon surveillance is periodically performed on the Post-Tensioning System that includes visual examination, lift off measurements, wire tensile testing, and analysis of the sheath filler grease. The tendon surveillance is performed in accordance with CCNPP Surveillance Test Procedure STP-M-663-1 for Unit I and STP-M-663-2 for Unit 2.

Procedures STP-M-663-1/2 are currently performed at five-year intervals in accordance with the plant Technical Specifications. The tests specifically addresses Technical Specification Surveillance Requirement 4.6.L6.2 to verify the structural integrity of the end anchorages and adjacent concrete exterior j surfaces. If any adjacent concrete shows indications of abnormal material behavior (e.g., cracking or l spalling), an engineering evaluation to demonstrate the ability of the Containment Structure to continue to perform its design function is completed in accordance with Technical Specification 3.6.1.6.b. Refer to the discussion under Aging Management Programs in Group 1 for a detailed discussion of these STPs.

[ References 16 and 17] j Group 5 - (weathering of grout)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to weathering of grout at the containment vertical tendon base plates:

• The grout provides structural support to the containment tendons and its integrity must be maintained under all CLB conditions.

• Weathering is plausible due to exposure to severe weather conditions and, if left unmanaged, it could eventually result in the grout not being able to perform its intended function under CLB conditions.

• Periodic visual inspections of the grout will continue to be performed in accordance with CCNPP .

Smveillance Testing Procedure STP-M-663-1/2. These inspections will discover the effects of weathering and assure that the necessary evaluations are performed to demonstrate the ability of the containment structure to continue to perform its design function.

Therefore, there is reasonable assurance that the effects of aging due to weathering of containment vertical tendon baseplate grout will be managed in such a way that it will be capable of performing its intended function consistent with the CLB during the period of extended operation.

Application for License Renewal 3.3 A-29 Calvert Cliffs Nuclear Power Plant

o .

ATTACHMENT (3)

APPENDIX A-TECHNICAL INFORMATION 3.3A - PRIMARY CONTAINMENT 3.3A.3 Conclusion The aging management programs discused for the Primary Containment Structure and System are listed in the following Table 3.3A-4. 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 in such a way that the intended functions of the components of the Primary Containment will be maintained during the period of extended operation consistent with the CLB under all design loading conditions.

The analysis / asses m it, corrective action, and confirmation / documentation process for license renewal is in accordance wita 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.

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Application for License Renewal 3.3 A-30 Calvert Cliffs Nuclear Power Plant

I . ,

i l ATTACHMENT (3) l APPENDIX A - TECHNICAL INFORMATION i

3.3A - PRIMARY CONTAINMENT l TABLE 3.3A-4 LIST OF AGING MANAGEMENT PROGRAMS FOR PRIMARY CONTAINMENT Program < Credited As .

Existing Preventive Maintenance Aging management of emergency air lock ga.skets by Program Procedure PAL-2 replacement based on condition.

Existing Containment Emergency Discovery and management of the effects of corrosion on the Sump Inspection Procedures containment emergency sump cover and screen by visual 1 STP-M-661-1 for Unit 1, inspection. (Group 2) and STP-M-661-2 for Unit 2 l

Existing Painting and Other Discovery and management of degraded protective coatings Protective Coatings and the effects of corrosion on steel components inside the (MN-3-100) Containment Structure. (Group 2) i Existing Liner Plate Surveillance Discovery and management of degraded protective coatings Test Procedures and the effects of corrosion on the containment wall and dome STP-M-665-1 for Unit I and liners. (Group 3)

STP-M-665-2 for Unit 2 1

Modified Containment Tendon Discovery and management of the effects of corrosion on the Surveillance Test Procedures containment tendons by visual inspection and analysis of the STP-M-663-1 for Unit I and filler grease. (Group 1)

STP-M-663-2 for Unit 2 Discovery and management of prestress losses for the containment tendons by performance oflift-off tests and wire tensile tests. The existing tendon lift-off force curves will be re-evaluated to reflect the required prestress levels for the period of extended operation. (Group 1)

Discovery and management of the effects of weathering of grout by visual inspection. (Group 5)

Modified Structure and System Discovery and management of the effects of corrosion on steel Walkdowns (MN-1-319) components outside the Containment Stru::ture. Guidance will be added to assist in fimetional adequacy. determinations and for authority to deviate from scope or schedule. (Group 2)

Discovery and management of the effects of corrosion on the refueling pool liner and PCSR by monitoring leakage.

(Group 4)

L l

Application for License Renewal 3.3 A-31 Calvert Cliffs Nuclear Power Plant

e o A*ITACHMENT m APPENDIX A-TECHNICALINFORMATION 3.3A- PRIMARY CONTAINMENT 3.3A.4 References

1. CCNPP Updated Final Safety Analysis Repon, Revision 21

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

Revision 6, October 15,1990

3. CCNPP Drawing 63874SH0004, "SR Ductbank Under West Plant Road Plan," Revision 0, April 4,1995

4. CCNPP Drawing 63874SH0005, " Underground Conduit West of Turbine Building Plan,"

Revision 0, July 15,1996

5. CCNPP Repon " Aging Management Review Report for the Containment Structure (System 059)," Revision 4, February 1997

6. CCNPP Report " Aging Management Review Repon for the Containment System (059)," l Revision 1, May 1996

7. CCNPP " Aging Management Review Repon for Component Supports," Revision 3, February 4,1997

8. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated October 28,1997,

" Containment Tendon Engineering Evaluation Report"

9. Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated January 23,1998, Review of Containment Tendon Evaluation Report - CCNPP Unit Nos. I and 2 (TAC Nos. M99880 and M99881) 1

10. CCNPP Report, " Component Level Scoping Results for the Containment Stmeture," Revision 2, February 11,1997

11. CCNPP Report, " Component Level ITLR Screening Results for the Containment System," '

Revision 0, February 23,1993

12. CCNPP Preventive Maintenance Checklist MPM00109, " Replace Gaskets for Equipment Hatch" l

i

13. CCNPP Preventive Maintenance Checklist MPM00108, " Replace Gaskets for Personnel Air Lock Door"

14. CCNPP Technical Procedure PAL-2, " Containment Personnel Emergency Escape Air Lock I Adjustment, Lubrication, and Inspection (Units 1 and 2)," Revision 1, February 5,1993

15. Electric Power Research Institute Report TR-103835,"PWR Containment Structures License Renewal Industry Report," Revision 1, July 1994

16. CCNPP Surveillance Test Procedure STP-M-663-1, " Containment Tendon Surveillance,"

Revision 9, August 21,1997

17. CCNPP Surveillance Test Procedure STP-M-663-2, " Containment Tendon Surveillance,"

Revision 7, October 1,1997

18. CCNPP Report, Time Limited Aging Analysis Review Report," Revision 0, November 1997

19. CCNPP Administrative Procedure MN-3-100, " Painting and Other Protective Coatings,"

Revision 4, March 10,1997 Application for License Renewal 3.3A-32 Calvert Cliffs Nuclear Power Plant

o +, e ATTACHMENT m APPENDIX A-TECHNICALINFORMATION 3.3A- PRIMARY CONTAINMENT

20. CCNPP Administrative Procedure MN-1-319, " Structure and System Walkdowns," Revision 0, September 16,1997

21. CCNPP Surveillance Test Procedure STP-M-661-1, Containment Emergency Sump Inspection, Revision 0, April 29,1993

22. CCNPP Surveillance Test Procedure STP-M-661-2, Containment Emergency Sump Inspection, Revision 0, April 29,1993

23. CCNPP Administrative Procedure EN-4-105, " Containment Leakage Rate Testing Program,"

Revision 1, March 14,1997

24. CCNPP Surveillance Test Procedure STP-M-665-1," Containment Liner Plate Surveillance -

Unit 1," Revision 1, April 16,1986

25. CCNPP Surveillance Test Procedure STP-M-665-2, " Containment Liner Plate Surveillance -

Unit 2," Revision 3, December 9,1992

26. Electric Power Research Institute Report TR-103842, " Class I Structures License Renewal Industry Report," Revision 1, July 1994 Application for License Renewal 3.3 A-33 Calvert Cliffs Nuclear Power Plant

. J 7 4 *4 ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL BUILDING STRUCTURES l

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Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998 6

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES 3.3E Auxiliary Building and Safety-Related Diesel Generator Building Structures This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing the Auxiliary Building, the adjacent Emergency Diesel Generator (EDG) Rooms, the Refueling Water Tank (RWT) Pump Rooms, the Safety-Related (SR) Diesel Generator Building, and the ductbank for EDG 1A. These structures, herein referred to as the Auxiliary Building and SR Diesel Generator Building Structures, were 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.

3.3E.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 fimctions of the structure.

By their nature, structures within the scope of license renewal are constructed in accordance with predetermined design requirements to support performance 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 ifit performs one or more of these structural functions: [ Reference 1, Section 4.2.2]

• Provide structural and/or functional support to SR equipment; Provide shelter / protection to SR equipment; NOTE: This function includes: (a) protection from radiation effects for equipment addressed by the Environmental Qualification Program; and (b) protection from high energy line break effects.

l Serve as a pressure boundary or a fission product retention barrier to protect public health and l safety in the event of any postulated Design Basis Events (DBEs);  ;

l

=

Serve as a missile barrier (internal or external);

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 l (e.g., Seismic Category II over i design considerations);

I Provide flood protection banier (internal flooding event); and  ;

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

This section begins with a description of the Auxiliary Building and SR Diesel Generator Building Structures. The intended functions performed by each structure are listed ar.d used to identify the structural component types within the scope of license 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 3.3E-1 Calvert Cliffs Nuclear Power Plant

pl,'%

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES 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 Descriotion/Conceamal Boundaries Figure 3.3Fel is a simplified layout showing the site structures that are within the scope of license renewal, including the Auxiliary Building and SR Diesel Generator Building Structures. [ References 2 through 5] A comprehensive layout and description of all site structures is provided in the Updated Final Safety Analysis Report (UFSAR), Chapter 1, with further discussion of their design features in Chapter 5 and Appendix 5A. [ Reference 6, Chapters 1,5, and Appendix SA). A general description, boundary, and

' design discussion for the structures addressed in this section follows:

The Auxiliary Building is located between the Unit I and Unit 2 Containment structures, on the west side of and adjacent to the Turbine Building. The Auxiliary Building is common to CCNPP Units I and 2. [ Reference 7, Section 1.1.1] Major structural features related to the Nuclear Steam Supply System and located inside the Auxiliary Building include the Cont ol Room, nuclear waste treatment facilities, and facilities for new and spent fuel handling, storage, and 1 shipment (including the spent fuel pool [SFP], the SFP storage racks, and the new fuel racks).

l

[ Reference 6, Sections 1.2.2 and 5.6.1.1] Three EDG Rooms and each Unit's RWT Pump Room j are adjacent to the Auxiliary Building structure, and are supported on reinforced concrete j foundations that are separate from the Auxiliary Building foundation mat. [ References 8 and 9] l (The components inside the Auxiliary Building and adjacent rooms that are within the scope of license renewal are evaluated by system in Section 5 of the BGE LRA.) He Auxiliary Building and adjacent rooms, and their structural components, provide support and shelter to S~R and NSR equipment. All structural components enclosed within these structures that serve such functions are within the scope of license renewal. Those areas inside the Auxiliary Building that are specifically excluded from Seismic Category I requirements in the CCNPP Quality List (e.g.,

maintenance shops, stairways, kitchen, toilets, offices) are not within the scope of license renewal _ (Reference 10, pages 58-59] De conce'ptual boundary of the Auxiliary Building includes the areas that house SR systems, equipment, or components that must remain functional before, during, or after a safe shutdown earthquake. Additionally, the conceptual boundary of the Auxiliary Building includes structural or functional supports for NSR components whose failure during an abnormal (e.g., seismic) event could adversely affect the operability of SR components; the associated structural components in the Auxiliary Building provide support for SR mounting of such components. [ Reference 10, pages 46 and 57 through 59] The Auxiliary Building and adjacent rooms are primarily reinforced concrete structures, and their foundations support structural steel and reinforced concrete frames that codsist mainly of reinforced concrete walls and floors. [ Reference 6, Section 5.6.1.1] Structural components located within the conceptual boundary of the Auxiliary Building have been designed for the loads and conditions shown in Table 5-6 of the UFSAR. [ Reference 6, Section 5.6.1.2] As part of modifications completed in ,

1992 to upgrade the spent fuel cask handling crane to meet the single-failure-proof criteria of l NUREG-0554, " Single Failure Proof Cranes for Nuclear Power Plants," the capability of the associated structural components in the Auxiliary Building to withstand the postulated seismic loading was re-evaluated and found adequate. [ Reference 11] Evaluations of the floor of the SFP l

j I

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

e r.. ,

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR

- BUILDING STRUCTURES indicate that its structural integrity will not be impaired in the unlikely event of a cask drop.

[ Reference 6, Sections 5.6.1.4 and 5.6.1.5]

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The SR Diesel Generator Building is located northwest of the Auxiliary Building and is common to CCNPP Units I and 2. [ Reference 12, Appendix B4, Section 1.1] It houses EDG 1 A, which is one of four EDGs designed to provide a dependable onsite power source capable of automatically starting and supplying the essential loads necessary to safely shut down the plant and maintain it in a safe shutdown condition under all conditions. (The other three EDGs are l housed in the rooms adjacent to the Auxiliary Building described above.) [ Reference 6, Sections 1.2.2 and 8.4] The SR Diesel Generator Building also houses the fuel oil storage tank for EDO 1 A and other auxiliary equipment. [ Reference 6, Sections 1.2.2 and 8.4.2; Reference 12, Appendix B4, Section 1.2] (The components inside this structure that are within the scope of license renewal are evaluated as part of the EDG System in Section 5.8 of the BGE LRA.) Since the SR Diesel Generator Building houses SR systems, equipment, or components that must remain functional before, during, or after a safe shutdown earthquake, it is required to meet Seismic CategoryI criteria, and has been designed for associated loads and conditions.

[ Reference 6, Section 5.6.5.2; Reference 10, pages 46,57, and 58] The SR Diesel' Generator Building !s primarily a reinforced concrete structure supported on a mat foundation at grade level (i.e., Elevation 45'-0") with a panial basement in the area of the EDG pedestal. In addition, a one-story structure is provided on the east side of the building as missile protection for the main building entry and EDG area exhaust louver. [ Reference 6, Section 5.6.5) The conceptual boundary of the SR Diesel Generator Building includes all structural components such as concrete foundations, walls, and slabs, as well as a buried ductbank that runs between the SR Diesel Generator Building and the Auxiliary Building for the electrical distribution for EDG 1 A.

Portions of this buried ductbank are also common to the Station Blackout Diesel. [ Reference 6, '

Section SA.2.1.2; Reference 12, Appendix B4, Section 4.2.4]

A one-time procedure was used to evaluate aging management for structural component types within the conceptual boundary of the SR Diesel Generator Building. The evaluation produced a .

listing of structural component types subject to AMR grouped by materials and environment and related them to similar groupings of structural component types in the Auxiliary Building.

[ Reference 12, Appendix B1, Section 6.3.H] Since completion of construction in 1996, evidence of age-related degradation of the SR Diesel Generator Building has not been observed.

[ Reference 6, Appendix IC, Section IC.1) Due to similarity in function and structure to the  !

Auxiliary Building, operating experience related to aging mechanisms and their management for l the Auxiliary Building is expected to provide early warning to BGE for any aging of the SR Diesel Generator Building that will need to be managed. [ Reference 12, Appendix B4, Section 1.1]

Application for License Renewal 3.3E-3 Calvert Cliffs l'uclear Power Plant

ATTACHMENT (4)

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• • O LICENSE RENEWAL FIGURE 3.3E-I CCNPP SITE STRUCTURES (SIMPLIFIED DIAGRAM - FOR INFORMATIO.N ONLY)

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

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES Component supports that are connected to structural components in the Auxiliary Building and SR Diesel Generator Building Structures 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 supponing 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 equipmer (thermal expansion, otating equipment, etc.), as well as the surrounding environment. Sul .,rts for structural components (e.g., platform hangers) are not " component supports" in this sense because any support for a structural component is itself a structural component (i.e., included in the scope of the associated structure).

[ Reference 13, Section 1.1.1]

Cranes that routinely lift heavy loads over SR components in the Auxiliary Building and components involved in fuel handling and transfer in and around the SFP are evaluated for the effects of aging in the Fuel Handling Equipment and Other Heavy Load Handling Cranes Commodity Evaluation in Section 3.2 of the BGE LRA. [ Reference 1, Section 7.2.2] The following structural component types in the Auxiliary Building interface with cranes and fuel handling equipment and are evaluated in this section:

Spent fuel cask handling crane -

interface with carbon steel crane rails; rail / supports (girders)

SFP reinforced concrete -

interfaces with spent fuel handling machine carbon steel rails, transfer machine jib crane structural members; SFP stainless steel liner -

interfaces with structural members of the spent fuel shipping cask platform, incore instrumentation trash racks, SFP platform, spent fuel inspection elevators fuel upending machine and transfer carriage; and Spent fuel and new fuel storage racks -

interface with fuel assemblies.

Sconed Structures and Their Intended Functions l

The Auxiliary Building and SR Diesel Generator Building Structures are in scope for license renewal based on 10 CFR 54.4(a). All seven generic structural ftmetions listed above are applicable to these structures as shown in Table 3.3E-1. The intended functions for these enclosures were determined based on the requirements of f54.4(a)(1), Q54.4(a)(2), and 54.4(a)(3), in accordance with the CCNPP IPA Methodology Section 4.2.2. [ Reference 1; Reference 2, Table 2; Reference 7, Secticn 1.1.3; Reference 12, Appendix B4, Section 1.3]

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

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES TABLE 3.3E-I INTENDED FUNCTIONS FOR THE AUXILIARY BUILDING AND SR DIESEL GENERATOR BUILDING STRUCTURES Applicable Applicable Applicable

"# " to Auxiliary to SR , to . Applicable Building & Diesel ductbank 10 CFR 54.4(a)

Adjacent Generator .

for Criteria Roones? Building? EDG1A7

1. Provide structural and/or functional support to SR Yes Yes Yes 54.4(a)(1) equipment

2. Provide shelter / protection to SR equipment Yes Yes Yes Q54.4(aXI)

NOTE: The SR Diesel Generator Building and the ductbank for EDG 1A are not required to provide protection from radiation or high energy line break effects.

3. Serve as a pressure boundary or a fission product Yes No No {54.4(aXI) retention barrier to protect public health and safety in the event of any postulated DBEs

4. Serve as a missile barrier (internal or external) Yes Yes Yes 54.4(aXI)

5. Provide structural and/or functional support to NSR Yes Yes No {54.4(aX2) 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) Yes Yes No @54.4(a)(2)

7. Provide rated fire barriers to confine or retard a fire from Yes Yes No {54.4(a)(3) spreading to or from adjacent areas of the plant Comnonents Subject to AMR A generic list of structural component types was developed for use during the structural component scoping task. The 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) These structural components were combined into the following four structural categories based on their design and materials: [ Reference 12, Appendix B4, Section 2.0; . Reference 14]

Concrete components; Structural steel components; Architectural components; and Unique components.

During the scoping process, applicable structural component types actually contained in the Auxiliary Building and SR Diesel Generator Building Structures were identified. Within the four structural Application for License Renewal 3.3 E-6 Calven Cliffs Nuclear Power Plant

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ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES component categories,47 structural component types were determined to contribute to at least one of the structural intended functions listed in Table 3.3E-1 for the associated structure. Table 3.3E-2 lists the structural component types and the associated functions that apply to the Auxiliary Building, the adjacent EDO Rooms and RWT Pump Rooms, the SR Diesel Generator Building, and the ductbank for EDG 1 A.

Unless otherwise noted, structural components that are part of the structure, but do not contribute to any of the intended functions of the structure, are not listed in Table 3.3E-2. [ References 4 and 5; Reference 7, Table 2-1; Reference 12, Appendix B4, Section 2.0; References 14 through 17]

Per the license renewal rule, ". . . Structures and components subject to an aging management review shall encompass those structures and components (i) That perform an intended function, as described in 654.4 without moving parts or without a change in configuration or properties . . . 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 functions requires moving parts or a change in configuration or propenies. 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]

Of the 47 structural component types within the scope of license renewal for the Auxiliary Building and SR Diesel Generator Building Structures, one unique component type, Pipe Encapsulations, was evaluated in the AMR for the Main Steam System; the results of the AMR for these components are presented in Section 5.12 of the BGE LRA. The remaining 46 structural component types, listed in Table 3.3E-2, are subject to AMR and are evaluated within this section. [ Reference 7, Table 2-1; Reference 12, Appendix B4, Table 2-1 and Section 4.2.4]

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 3.3 E-7 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES f

3.3E.2 Aging Management

'Ihe list of potential Age-Related Degradation Mechanisms (ARDMs) identified for components in the Auxiliary Building and SR Diesel Generator Building Structures is given in Table 3.3E-3, with plausible {

ARDMs identified by a check mark (/)in the appropriate column. [ Reference 7, Attachments 1 and 2;

{

Reference 12, Appendix B4, Section 4.0] For efficiency in presenting the results of these evaluations in j

this report, ARDM/ component type combinations are grouped together where there are similar j characteristics and the discussion is applicable to all components. Table 3.3E-3 also identifies the group 1 to which each ARDM/ component type combination belongs. The following groups have been selected for the Auxiliary Building and SR Diesel Generator Building Structures:

Group I: weathering of caulking, scalants, and expansion joints; Group 2: corrosion of steel (for components marked with an asterisk in Table 3.3E-2);

Group 3: corrosion of the SFP liner (sensitized zones); and Group 4: degradation of neutron-absorbing materials (for SFP storage racks). I I

Aging mechanisms that are not plausible are generally not discussed further in these BGE LRA sections, unless they are considered noteworthy. For the Auxiliary Building and SR Diesel Generator Building Structures, settlement is considered noteworthy and is discussed below.

I Industry technical reports conclude that settlement is a potentially significant ARDM for pressurized- i water reactor Containment Structures and for other Category 1 Structures at some plants. [ Reference 18, Section 5.5; Reference 19, Section 5.1.2] Settlement occurs both during construction and after '

construction. The amount of settlement depends on the physical properties of the foundation material.

[ Reference 7, Appendix J] Excavation unloading and structural loading during construction caused a small change in the void ratio of undisturbed soil. This change results in a very small or negligible amount of time-dependent settlement. [ Reference 6, Section 2.7.6.2; Reference 7, Appendix J)

Compacted soil is subject to some degree of settlement in the first 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 table. Sites with soft soil and/or sites with significant changes in underground water conditions over a long period of time may be susceptible to significant settlement. [ Reference 7, Appendix J; Reference 18, Section 4.5.3.2; Reference 19, Sectior. 4.6.3.2] 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 i within the structure. When buildings experience significant settlement, cracks in structural members, differential elevations of supporting members bridging between buildings, or both may be visibly detected. [ Reference 7, Appendix J] At CCNPP, long-term settlement was determined to be not l plausible for the Auxiliary Building and SR Diesel Generator Building Structures based on the following site-specificjustification:

l 1

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

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i ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES The foundation materials.supponing the Auxiliary Building and SR Diesel Generator Building Structures have varying physical propenies. The foundation for the Auxiliary Building is situated primarily on the undisturbed soil of the site's Miocene deposit, which is an exceptionally dense soil that is capable of supponing loads on the order of 15,000 to 20,000 pounds per square foot (psf). [ Reference 6, Section 2.7.3; Reference 7, Appendix J] The design contact pressure of the Auxiliary Building mat is only 8,000 psf. This contact pressure is about the same as the weight of soil removed by site grading and pit excavation for the structure. [ Reference 1, Sections 2.7.5 and 2.7.6.2; Reference 7, Appendix J] For the SR Diesel Generator Building, the ductbank for EDG 1 A, the EDG Rooms adjacent to the Auxiliary Building, and each Unit's RWT Pump Room, engineered soil structures incorporating compacted fill materials (e.g., crushed stone, compacted soil and/or concrete) provide foundation support. [ Reference 7, Appendix J; References 8,9, and 20; Reference 21, Sections 2.3.2 and 2.3.3] Quality assurance and quality control measures  !'

imposed during backfill placement included specification of maximum lift thicknesses and verification of minimum fill compaction. Beneath the SR Diesel Generator Building, crushed stone was compacted to 95 percent of maximum dry density based on the modified Proctor compaction method; similarly, the bedding material supporting the ductbank for EDG 1 A was compacted to 90 percent. [ References 22 and 23] Under the rooms adjacent to the Auxiliary

. Building, fill was compacted to 97 percent based on the standard Proctor compaction method. .;

[ References 8, 9, 24, and 25] For each of the structures constructed on compacted fill, a continuous program of soil testing during construction assured uniform placement of the material.

[ References 23 and 26] These activities assured that the causes of excessive settlement of plant ' l structures at other nuclear power plant sites did not exist during construction at CCNPP. ]

[ Reference 27] Control of the placement and compaction for these engineered soil structures was {

used to obtain the engineering propenies required to satisfy foundation design requirements.

A stable groundwater table exists in the vicinity of the Auxiliary Building and SR Diesel Generator Building Structures. A permanent pipe drain system surrounding the plant, including j the Auxiliary Building and adjacent rooms, is designed to maintain the groundwater table below l Elevation 10'-0"; this minimizes any changes to the site conditions that could affect seulement of-the foundations for these structures. [ Reference 6, Section 2.7.3; Reference 28] The foundation for the SR Diesel Generator Building and the ductbank for EDG 1A are situated outside the boundary of the permanent pipe drain system. [ References 4,5,15,16, and 17; Reference 21, Section 2.3.2] 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, Section 2.5.3.3] However, significant deviations from the seasonal cycles and j occasional meteorological effects (e.g., drought conditions) observed over the past 25 years are i not expected during the period of extended operation.

• - The foundations for the Auxiliary Building and SR Diesel Generator Building Structures tend to .j uniformly settle as rigid bodies. [ Reference 7, Appendix J; Reference 12, Appendix B4, Table 4-1] Most of the predicted settlement is expected in terms of uniform settlement, which j has no adverse effect on structural components of the Auxiliary Building, the SR Diesl Generator Building, or the ductbank for EDG 1 A. [ Reference 7, Appendix J; Reference 12, l Appendix B4, Table 4-1] A sand pocket is incorporated into the design of the ductbank for EDG 1A at its junction with the SR Diesel Generator Building to accommodate the effects of differential settlement on the associated conduit. The effects of one-time building settlement are Application for License Renewal 3.3E-12 Calvert Cliffs Nuclear Power Plant  ;

I

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES l

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 observed to date. A walkdown inspection of the Auxiliary Building, performed in 1994, found no indication of structural damage due to settlement. [ Reference 7, Attachment 7] These observations suppon the conclusion that settlement of the Auxiliary Building and SR Diesel Generator Building {

Structures at CCNPP is not plausible.

)

l

\

The following is a discussion of the aging management demonstration process for each group identified in Table 3.3E-3. 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 - (weathering of caulking, sealants, and expansion joints)- Materials and Environment The structural component types affected by weathering include caulking, scalants, and expansion joints used in the Auxiliary Building and SR Diesel Generator Building Structures. By accommodating thermal and seismic movement without exceeding allowable stresses, expansion joints contribute to providing shelter / protection to the SR equipment inside reinforced concrete structures. Elastic caulking and seslant materials are used to fill these joints, as well as barrier penetration seals for piping or conduits that run through structural concrete surfaces. These materials prevent the passage of steam or water through building joints during a high energy line break or flooding event. When any of these structural component types are installed in a fire barrier, they also contribute to performance of the structure's fire protection function. [ Reference 7, Appendix 0; Reference 14, Table 3S]

The material requirements for the caulking, scalants, and expansion joints used during construction of CCNPP were governed by construction specifications. Individual products were specified by manufacturer and brand name (or approved equivalent) for particular applications. [ References 29 and 30]

l The caulking, sealants, and expansion joints located indoors are exposed to temperature and humidity conditions inside the Auxiliary Building and SR Diesel Generator Building Structures as described in UFSAR Chapter 9. [ Reference 6, Section 9.8.2.3 and Table 9-18] The caulking, sealants, and expansion joints located outdoors are subject to the normal outside atmosphere at the CCNPP site. The 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 site. [ Reference 7, Appendix 0]

l Application for License Renewal 3.3E-13 Calvert Cliffs Nuclear Power Plant

I I

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NITACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES l

Group 1 -(weathering of caulking, sealants, and expansion joints)- Aging Mechanism Effects Caulking, scalants, and expansion joints that are exposed to ambient conditions (indoor or outdoor) are susceptible to degradation due to weathering. Exposure to sunlight (ultraviolet exposure), changes in l humidity, ozone cycles, snow, rain, ice, and temperature and pressure fluctuations contribute to the weathering ARDM. The effects of weathering on most materials, including caulking, sealant, and expansion joint materials,'are evidenced by a decrease in elasticity (e.g., drying out), an increase in '

hardness, and shrinkage. [ Reference 7, Appendix 0]

Expansion joints between the Containment structures and the EDG Rooms adjacent to the Auxiliary Building have experienced age-related degradation in the past. The affected joints were subsequently repaired using approved sealant materials.

Weathering is plausible for the caulking, sealants, and expansion joints used in the Auxiliary Building and SR Diesel Generator Building Structures because they are exposed to the environmental conditions that contribute to this ARDM. Ifleft unmanaged for an extended period of time, these materials will become brittle and lose their capability to perform their intended functions under current licensing basis (CLB) conditions. [ Reference 7, Appendix 0]

Group 1 -(weathering of caulking, sealants, and expansion joints)- Methods to Manage Aging Mitigation: Because weathering of caulking, sealants, and expansion joints is affected by exposure to 1 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 joints 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, sealants, and expansion joints can ensure that their condition is I maintained at a level that allows them to perform their intended functions. An effective program will I provide for baseline inspection along with periodic future inspections at appropriate intervals depending upon the degree of harshness of the environment of the caulking, sealant, or expansion joint Jtems that l are in a harsh exterior environment would be inspected more frequently. This program would involve j visual inspection and probing to determine that the caulking, sealant, or expansion joint is satisfactorily '

attached to the surface and is flexible.

Group 1 - (weathering of caulking, scalants, and expansion joints) - Aging Management Program (s) l Mitigntion: There are no CCNPP programs credited for mitigation of weathering.

Discoverv: Caulking, scalants, and expansion joints that perform a fire barrier function in the Auxiliary Building, the adjacent EDG Rooms, and the RWT Pump Rooms, are managed under an existing program. The Penetration Fire Barrier Inspection Program, implemented through CCNPP Technical Procedure STP-F-592-1/2, is adequate to manage the effects of aging for caulking, sealants, and expansionjoints that function as fire barriers without modification. [ Reference 7, Section 5.2.1].

Applicatim for License Renewal 3.3E-14 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4)

APPENDIX A - TECIINICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES 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 shutdown areas in Units 1 and 2. The scope of this i procedure is to visually inspect the following type of fire barrier penetration seals for operability: l

[ References 31 and 32, Sections 1.0 and 2.2]

=

Electrical conduit and cable tray penetration seals; Heating, 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 Specifications 3.7.12 and 4.7.12.a, 10 CFR Part 50 Appendix R, the CCNPP Fire Protection Plan, NRC Generic Letter 86-10,

" Implementation of Fire Protection Requirements," and various plant drawings. [ References 31 and 32, Section 3.1]

The procedure is performed at least once every 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 damage, cracking, voids, and proper installation. The procedure provides separate " failure criteria" and " repair criteria." The j

" failure criteria" are used to determine if the penetration seal is considered to be inoperable. The " repair criteria" are used to determine if the penetration seat is operable but in need of repair. [ References 31 and 32, 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 l addition, any conditions adverse to quality discovered during the inspection are documented on Issue Reports in accordance with the CCNPP Corrective Actions Program. [ References 31 and 32, Section 6.5 and Attachment B]

The Fire Protection Program at CCNPP (which includes STP-F-592-1/2) 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 prevention program inspection and audit utilizing either qualified offsite BGE personnel or an outside fire protection firm is 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 audit 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 33, Section 18.18]

The Fire Protection Program has been evaluated by the NRC as part ofits 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 34 and 35]

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

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES Operating experience 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 inadequate technique used in the original installation of the seal materials.

The corrective actions taken as a result of the Penetration Fire Barrier Inspection Program will ensure that the caulking, scalants, and expansion joints in the Auxiliary Building and adjacent rooms that perform a fire barrier function will remain capable of performing their intended function under all CLB conditions.

Caulking, sealants, and expansion joints that are not included in the Penetration Fire Barrier Inspection Program 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 needed to adequately manage the aging of these structural component types. [ Reference 7, Appendix O]

For the caulking, scalants, and expansion joints in the SR Diesel Generator Building, as well as for those in the Auxiliary Building and adjacent rooms that do not perform a fire barrier function, a new CCNPP Caulking and Scalant Inspection Program will provide requirements and guidance for the identification, inspection frequencies, and acceptance criteria for caulking, scalants, and expansion joints to ensure that 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, scalants, and expansion joints for the Auxiliary Building and SR Diesel Generator 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, Section 5.2.3 and Attachment 8]

Group I - (weathering of caulking, sealants, and expansion joints) - 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:

Caulking, scalants, and expansion joints in the Auxiliary Building and SR Diesel Generator Building Structures contribute to providing shelter / protection to SR equipment inside the structures. Some caulking and sealants also provide flood protection barriers. When used in fire barriers, all of these component types also contribute to the structures' fire protection functions.

Therefore, the condition of caulking, sealants, and expansion joints must be maintained under all CLB design conditions.

=

Caulking. sealant, and expansion joint materials are subject to weathering when exposed to normal indoor and outdoor conditions at the CCNPP site. If unmanaged, this ARDM could result in these components losing their capability to perform their intended functions under CLB design loading conditions.

• For caulking, scalants, and expansionjoints that function as fire barriers in the Auxiliary Building and adjacent rooms, the Penetration Fire 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 barrier intended function will be maintained.

Application for License Renewal 3.3E-16 Calvert Clifts Nuclear Power Plant

I F

ATTACHMENT (4) l APPENDIX A - TECilNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES For caulking, scalants, and expansion joints in the SR Diesel Generator Building, as well as for those in the Auxiliary Building and adjacent rooms that do not perform a fire barrier function, a new 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 a reasonable assurance that the effects of weathering will be adequately managed for the caulking, scalants, and expansion joints in the Auxiliary Building and SR Diesel Generator Building Structures 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 2 -(corrosion of steel)- Materials and Environment Group 2 comprises those components marked with an asterisk in Table 3.3E-2. These components are

! all fabricated from carbon 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 Auxiliary Building and SR Diesel Generator Building Structures. The environment to which these components are I

subjected varies with their installed location. Component types located indoors are exposed to temperature and humidity conditions inside the Auxiliary Building and SR Diesel Generator Building Structures as described in UFSAR Chapter 9. [ Reference 6, Section 9.8.2.3 and Table 9-18] Those steel component types located outdoors are subject to the normal outside atmosphere at the CCNPP site.

[ Reference 7, Appendix K] There is no heavy industry nearby CCNPP to add chemicals to the

atmosphere but, due to the close proximity of the Chesapeake Bay, the steel components located l outdoors could be exposed to condensation. [ Reference 6, Sections 2.8 and 2.10; Reference 7, l

Appendix C] Some of the steel components are located near the SFP, where condensation in the presence of oxygen could lead to oxidation. Caibon steel located in these areas may be subjected to more severe local environments. [ Reference 7, Appendix K]

Since corrosion was recognized as a potential degradation mechanism for all carbon steel components of site structures, protective coatings were incorporated into the original design. Exposed structural steel surfaces in the Auxiliary Building and SR Diesel Generator Building Structures were coated during the l construction phase (e.g., shop-primed, field-painted, hot-dipped galvanized). [ Reference 7, Appendix K;  ;

References 36 through 43]

l l Group 2 -(corrosion of stee!)- Aging Mechanism Effects l 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, alternating wetting and l drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either l from saltwater, the atmosphere, or groundwater, increase the rate of corrosion by increasing the electrochemical activity. If steel is in contact, through an electrolytic solution, with another metal that is more noble in the galvanic series, corrosion of the steel may accelerate. [ Reference 7, Appendix K]

Corrosion products such as hydrated 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 l

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

l ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES 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 can be visually detected well in advance of significant degradation of the steel. [ Reference 7, Appendix K]

i 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 3.3E-2 to perform their intended functions under CLB design loading conditions. [ Reference 7, Appendix K]

Group 2 -(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 external 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. I 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 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 carbon steel structural components. Corrective action for failed protective coatings and any actual metal degradation can be carried out as necessary. [ Reference 7, Appendix K]

Group 2 -(corrosion of steel)- 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 Walkdowns,"

provides for discovery of corrosion of steel (or conditions that would accelerate corrosion, such as pooled water) for the structural components in the Auxiliary Building and SR Diesel Generator Building Structures by performance of visual inspections during plant walkdowns. [ Reference 7, Attachment 4; Reference 12, Appendix B4, Section 4.2.5] 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.

[ Reference 44, Section 1.1]

Under this program, responsible personnel perform periodic walkdowns of their assigned structures and i

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 Application for License Renewal 3.3 E-18 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES initially pressurized, energized, or placed in service); and as required for plant modifications.

[ Reference 44, 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, 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 44, 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 44, 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, anchor bolt degradation, or leakage of fluids; and Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage 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 material condition beyond 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 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. [ Reference 44, Section 5.3 and Attachments 4 and 8]

The program described above will be modified to: (a) specifically identify the carbon steel component types within the scope of the performance assessments (including those identified in Table 3.3E-2 as l

unique structural component types); and (b) add guidance regarding approval authority for significant I departures from the walkdown scope / schedule specified. The modified program will ensure that I degraded conditions due to corrosion of steel are identified and corrected such that carbon steel components of the Auxiliary Building and SR Diesel Generator Building Structures will be capable of performing their intended functions cer.sistent with CLB design conditions.

Group 2 - (corrosion of steel)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to l corrosion of steel in the structural components of the Auxiliary Building and SR Diesel Generator Building Structures:

The carbon steel compenents 'dentified in Table 3.3E-2 provide various passive intended i functions for the associated structures, and failure could directly prevent satisfactory accomplishment of functions tbt must be maintained under CLB design loading conditions. i Application for License Renewal 3.3E-19 Calvert Cliffs Nuclear Power Plant l l

ATTACHMENT (4)

APPENDIX A - TECIINICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES Components in this group are exposed to moisture and oxygen in their installed locations.

Carbon steel corrodes 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).

Coatings, specified during original construction, mitigate the effects of corrosion by providing a protective layer that prevents moisture and oxygen from contacting the steel.

+

'Ihe CCNPP Structure and System Walkdowns Program provides for periodic walkdowns of Group 2 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 2, 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 of aging due to corrosion of carbon steel will be managed in such a way that structural components of the Auxiliary Building and SR Diesel Generator Building Structures will be capable of performing their intended functions consistent with the CLB during the period of extended operation.

Group 3 -(corrosion of the SFP liner)- Materials and Environment The SFP liner at CCNPP is constructed of Type 304 stainless steel material. [ Reference 7, Appendix L Reference 45] The liner was assembled from a series of individual plates that were welded together. l The stainless steel liner is not a load-bearing structural component; it was designed only as a leaktight  !

barrier serving a pressure boundary function. [ Reference 7, Table 2-1 and Appendix L; Reference 45]  !

One side of the SFP liner is normally exposed to the borated water contained inside the pool; the other side of the SFP liner conforms to a reinforced concrete wall. [ Reference 7, Appendix L]

The concrete behind the welds was formed such that a small trough or channel was created to allow detection of any leakage that may occur through the welds or liner. Individual channels are grouped into leak chases that are each connected to a hand valve through a short piping system. There are a total of ten such " telltale" valves for the SFP (four for leak chases from vertical plates and one for the floor leak chase in each Unit's side of the SFP). [ Reference 7, Appendix L; References 46 through 49]

Water has been collected from these leak chases monthly for many years. [ Reference 7, Attachment 5 and Appendix L] The total leakage measured each month averages about 0.1 gallons in a 24-hour period.

Due to the small sample size and otur interfering factors, it is difficult to confirm the SFP liner as the source of the water collected during the monthly testing; frequently, no water leakage is observed during

, the test. [ Reference 7, Appendix L] in 1995, the source of water collected from one of the " telltale" valves was determined to be from the Unit 2 SFP.

l Group 3 -(corrosion of the SFP liner)- Aging Mechanism kJfects Both the plate material and the associated weld materials are susceptible to stress corrosion cracking, which is defined as cracking under the combined actions of corrosion and tensile stresses. The stresses may be applied (external) or residual (internal), and the cracks themselves may be transgranular or Application for License Renewal 3.3E-20 Calvert Cliffs Nuclear Power Plant

ATTACHMENT Q) f APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR 4 BUILDING STRUCTURES intergranular. [ Reference 7, Appendix L] Type 304 stainless steels are particularly prone to this ARDM in locations that are sensitized, such as heat-affected zones in and around welds and at crevice geometries. This is because of the changes in the material's microstructure that take place due to the welding heat, and because of high residual stresses in and around the welds. [ Reference 19, Section 4.5.1.1] Under such conditions, Type 304 stainless steel may develop intergranular stress l corrosion cracking (IGSCC). [ Reference 7, Appendix L]

)

The SFP liner was not designed to carry significant structural loads, and the strains induced by conforming to deformations in the concrete wall of the SFP are negligible under normal plant operating conditions. The liner is not exposed to a corrosive environment under normal operating conditions. l Therefore, the conditions necessary for stress corrosion cracking do not exist for the SFP liner, i

[ Reference 7, Appendix L]

Ilowever, the heat-affected zones in and around welds joining the liner plates are potential sites for sensitization. Conditions that may contribute to the occurrence oflGSCC include elevated temperatures, chloride content, boric acid concentration, oxygen concentration, and degree of sensitization.

[ Reference 7, Appendix L; Reference 19, Section 4.5.1.1]

Initiation and propagation of cracks in the stainless steel material are the typical efTects of IGSCC.

[ Reference 7, Appendix L] IfIGSCC of the SFP liner were left unmanaged for an extended period of time, the resulting leakage could lead to the inability to perform the intended pressure boundary function under CLB design loading conditions. [ Reference 7, Appendix L]

Group 3 -(corrosion of the SFP liner)- Methods to Manage Aging Mitigation: Because the factors affecting corrosion of the SFP liner (i.e., applied or residual stresses, SFP chemistry, and materials and methods of construction) are inherent in the structure's design and function, there are no practical methods to mitigate its effects. However, the discovery methods described below are considered adequate to manage this ARDM.

Discoverv: Degradation due to IGSCC at the sensitized zones of the SFP liner would result in increasing SFP leakage and can be detected by monitoring the rate ofleakage from the SFP at the " telltale" valves.

[ Reference 7, Appendix L] Because the SFP Cooling System provides a permanent make-up capability with suitable redundancy or back-up to prevent uncovering of fuel assemblies in the SFP, measuring and trending leakage from the SFP provides for effective aging management. [ Reference 6, Section 9.4.4; Reference 7, Appendix L]

Group 3 -(corrosion of the SFP liner)- Aging Menagement Program (s)

Mitigation: There are no CCNPP programs credited for mitigation of corrosion of the SFP liner.

[ Reference 7, Appendix L]

Discoverv: The Calvert Cliffs Operating Manual, NO-1-201, establishes the requirements for implementing and using Operating Instructions as approved, preplanned methods of conducting operations. [ Reference 50] The CCNPP Performance Evaluation Program, NO-1-203, has been established to perform periodic operational checks and obtain readings to determine equipment Application for License Renewal 3.3 E-21 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES performance, as determined by manufacturers' recommendations, System Engineers' recommendations, and operating needs. [ Reference 51] These programs address controls for activities conducted as part of daily shin operations, and apply to operators and others who interact with them. [ Reference 52]

The Performance Evaluation Program provides for determination of SFP leakage on a monthly frequency. [ Reference 51, Section 5.3.B.4; Reference 53] Calvert Cliffs procedure PE 0-67-2-0-M,

"#11 & #12 Spent Fuel Pools - Determine Liner Leakage," directs performance of the SFP leakage test in accordance with OI-24D, " Spent Fuel Pool Cooling - Infrequent Operations," which provides detailed instructions for leakage monitoring of the SFP Cooling System. During this test, the " telltale" valves are opened, drained, and are monitored for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with catch devices installed at the outlet of each

" telltale" valve. [ Reference 54, Section 6.1] If the total leakage from all " telltale" valves exceeds one gallon in a 24-hour period, an engineering evaluation of the condition is performed. [ Reference 54, Section 6.1]

As part of the plant's administrative procedures hierarchy, the plant's nuclear operations procedures have numerous levels of controls and reviews, including assignment of responsibility for conducting performance evaluations as required, reviewing all the evaluations for accuracy and completeness, and j analyzing data for trends, if applicable. Specific responsibilities are assigned to BGE personnel for monitoring these programs through periodic audits. The Operating Manual and the Performance Evaluatioa Program have also been evaluated by the NRC as part of its routine licensee assessment l activities. These controls provide reasonable assurance that the associated activities will continue to be an effective method of monitoring the SFP liner for the effects ofIGSCC. [ References 50,51, and 52] i Group 3 -(corrosion of the SFP liner)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to corrosion of the SFP liner:

The SFP liner serves as a pressure boundary or fission product retention barrier to protect the ,

public health and safety in the event of any postulated DBEs and its integrity must be maintained under all CLB conditions.

]

a The SFP liner is constructed from stainless steel plates and exposed to a borated water environment.

Heat-affected zones in and around the SFP liner welds may be sensitized and susceptible to l IGSCC. If len unmanaged, IGSCC could eventually result in the SFP liner not being able to i perform its intended function under CLB conditions

• Periodic monitoring of leakage from the SFP under the Performance Evaluation Program will identify and document the presence of leakage that may be due to IGSCC, and ensure that appropriate corrective actions are taken if total leakage exceeds acceptance criteria.

Therefore, there is reasonable assurance that the effects of aging due to IGSCC of the SFP liner will be managed in such a way that it will be capable of performing its intended function consistent with the CLB during the period of extended operation.

Application for License Renewal 3.3E-22 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECIINICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES Group 4 -(degradation of neutron-absorbing materials)- Materials and Environment In 1980, the Unit I side of the SFP was modified with the installation of high-density spent fuel stora3 racks. These structural components consist of a base structure supporting storage cells primarily fabricated from stainless steel. A neutron-absorbing sheet, fabricated by The Carborundum Company and consisting of a boron carbide powder in a fiberglass matrix, is sandwiched between the inner and outer walls on the four sides of each storage cell. The Unit 2 side of the SFP was similarly modified in 1983; however, a different neutron-absorbing sheet, consisting of fine particles of boron carbide in a silicon polymer matrix, was used. This neutron-absorbing material is Boraflex, which is a trademark sheet form of a proprietary material manufactured by Brand Industrial Services, Inc.

l The intended function of the SFP storage racks is to provide structural / functional support for fuel  !

assemblics by maintaining a subcritical geometry in the SFP. The neutron-absorbing sheet materials are I not load-bearing structural components; however, they contribute to the intended function of the storage racks by absorbing neutrons in the SFP. Absorption of neutrons by these materials is assumed in the SFP criticality calculations. [ Reference 6, Section 9.7.2.1] The SFP storage racks are immersed in the borated water contained inside the pool. [ Reference 7, Appendix L]

Group 4 -(degradation of neutron-absorbing materials)- Aging Mechanism Effects Experiments have shown that unencapsulated Carborundum sheets experience a loss of boron carbide when exposed to gamma radiation in a water environment. [ Reference 55, Enclos are 3, Section 5.1.7]

Embrittlement and weakening of the polymeric bond phase reduces the tenacity with which the boron carbide particles are held. After exposures at higher levels of gamma odiation, unencapsulated Carborundum sheets are susceptible to spalling and surfac,e abrasion, which may result in loss of boron carbide material. These losses would be reflected in an observable decrease in sheet thickness and weight. [ Reference 56]

Several utilities have observed significant loss of Boraflex material from sample coupons. When the Boraflex material is subjected to gamma radiation in the SFP environment, the silicca polymer matrix becomes degraded, which may result in: (a) release of the silica filler and boron carbide from the sheet; and (b) shrinkage of the polymer and development of gaps in the material. The loss of boron carbide from Boraflex is characterized by slow dissolution of the Boraflex matrix from the surface of the sheet and gradual thinning of the material. The access of water to and around the Boraflex sheets is a significant factor influencing the rate of silica dissolution from Boraflex. [ Reference 57]

The use of adhesives to bond Boraflex sheets to steel in the storage cells is a significant factor leading to gap formation. Since the installation process for individual storage racks in the Unit 2 side of the SFP involved only single sheets of Boraflex material that were not fastened or permanently glued onto any surface or structure, gap formation due to Boraflex shrinkage is not expected. [ Reference 58]

Experimental data from industry test programs support conservative assumptions in SFP criticality analyses that encompass the gapping phenomenon within the design basis of the storage racks.

[ Reference 6, Section 9.7.2.1; Reference 57]

Degradation of the neutron-absorbing materials used in the SFP storage racks is plausible because they are exposed to gamma radiation and borated water in the SFP environment. A reduction in the amount of I

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

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APPENDIX A - TECIINICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES boron in the sheets (through spalling and surface abrasion of the Carborundum material, or dissolution of silica from the Boraflex material) could result in an increase in the reactivity of the SFP configuration.

Group 4 - (degradation of neutron-absorbing materials)- Methods to Manage Aging Mitigation: The SFP storage rack designs incorporate a vented, form-fitted wrapper that minimizes water ingress and gas accumulation. Because the factors affecting degradation of the neutron-absorbing '

materials (i.e., exposure to borated water and gamma radiation) are inherent in the design and function of the storage racks, there are no additional methods to mitigate its effects. However, the discovery methods described below are considered adequate to manage this ARDM.

Discoverv: Industry experience indicates that degradation in neutron absorption performance has not been observed in materials other than Boraflex. [ Reference 59] Although predictive computer models and areal density measurement techniques are under development in the industry, practical methods to directly monitor degradation of Boraflex material in the Unit 2 SFP storage racks are not currently available for use. [ Reference 60] However, degradation of either type of neutron-absorbing material can  ;

be monitored by periodic testing of sample coupons that are representative of the materials installed in '

the SFP storage racks. [ Reference 61]

Group 4 -(degradation of neutron-absorbing materials)- Aging Management Program (s)

Mitigation: There are no CCNPP programs credited for mitigating the degradation of neutron-absorbing materials in the SFP storage racks.

)

i Discov.my: Calvert Cliffs Administrative Procedun EN-4-101, " Coordination of Testing," provides the administrative process for the use of Engineering TN Procedures (ETPs). This program was established to ensure a comprehensive and integrated approach to testing activities. These activities require identification of testing requirements, specification of required plant conditions to perform the tests, development of procedures, integration of tests into a schedule based on required plant conditions, and accomplishment of testing according to schedule. [ Reference 62, Section 5.1.A]

Calvert Cliffs ETP 86-03R, " Analysis of Neutron Absorbing Material in Spent Fuel Storage Racks," was developed on the basis of vendor recommendations for detecting degradation of neutron-absorbing materials. [ Reference 63, Section 3.1.D] This program is designed to permit samples of the materials used in the SFP storage racks to be periodically removed from the SFP for examination. Through specific positioning of designated sample packets, both accelerated and long-term exposure to gamma i radiation and borated water is provided. The long-term sample packets are surrounded by typical fuel j assemblies, while the accelerated sample packets are placed next to the freshly-discharged fuel every i refueling outage. [ Reference 61] The sample coupons are a conservative representation of the neutron-absorbing materials in the SFP storage racks. The sample packet holders have a small gap between the top, bottom, and side spacer bars. Additionally, the spacer bars are 0.01 inches thicker than the sample coupons. These dimensional differences actually allow ingress of water to the sample coupon material; this effect is minimized in the SFP storage racks by the wrapper encapsulating the neutron-absorbing material. [ Reference 61] Each sample packet contains coupons of either Carborundum or Boraflex material. Sufficient samples are available so that the principal physical properties (i.e., sample weight for the Carborundum material, and sample hardness for the Boraflex Application for License Renewal 3.3E-24 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES material) can be determined as a function of exposure on a regularly scheduled basis. Visual condition is assessed on a graded scale, and the results of physical property analyses are compared to historical results. Vendor documents provide guidelines for interpreting test results. [ Reference 63] Unacceptable results are documented, reported, and corrected in accordance with the CCNPP Corrective Actions Program. [ Reference 62, Section 5.5.B.11]

He cumulative results of the coupon surveillance program indicate that the neutron-absorbing sheets have experienced no significant deterioration after more than 12 years of service. Evidence of erosion has been observed in sample coupons in the vicinity of inspection holes in the associated sample holder.

This erosion was determined to be the result of water flow on the surface of the material. This flow is due to thermal gradients produced by the spent fuel in the racks. Since the inspection holes in the SFP storage racks themselves are located above the level of the active fuel, erosion of material in their vicinity would not result in loss of the neutron-absorption function. There is no evidence that such erosion is occurring at locations other than the immediate vicinity of the sample holder inspection holes.

As part of the plant's administrative procedures hierarchy, ETPs have numerous levels of controls and reviews, including assignment of responsibility for identifying tests and associated plant conditions, conducting tests as required, and performing independent technical review of test results. [ Reference 62]

Specific responsibilities are assigned to BGE personnel for monitoring these programs through periodic audits. [ Reference 33] Calvert Cliffs ETPs have also been evaluated by the NRC as part of its routine licensee assessment activities. [ Reference 64] Calvert Cliffs will continue using the coupon surveillance program as the primary activity for monitoring condition of the neutron-absorbing materials in the SFP storage racks. As other methods are developed and determined to be effective (i.e., industry initiatives such as predictive computer models and areal density measurement techniques), they will be considered for applicability at CCNPP. [ Reference 61]

Currently, the coupon surveillance program requires removal of one long-term Carborundum sample packet and one long-term Boraflex sample packet from the SFP every four years; one accelerated sample packet of each material type is removed from the SFP every two years. [ Reference 63, Attachment 1]

The basis for this original timetable was e 40-year service life for the SFP storage racks; for each material type, two additional long-term and two additional accelerated sample packets were provided for contingency use. The program described above will be modified to: (a) reevaluate the adequacy of the sampling intervals in monitoring Carborundum and Boraflex condition through the period of extended operation; and (b) refine the process for scheduling sample packet removal from the SFP. He modified program will ensure that degradation of neutron-absorbing material is identified and corrected such that the SFP storage racks will be capable of performing their intended functions consistent with CLB design conditions.

Group 4 -(degradation of neutron-absorbing materials)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to degradation of Carborundum and Boraflex materials in the SFP:

The SFP storage racks incorporate neutron-absorbing materials to maintain the required suberitical margin for fuel assemblies in the SFP environment, and their function must be maintained under all CLB conditions.

Application for License Renewal 3.3E-25 Calvert Cliffs Nuclear Power Plant

A'ITACHMENT (4)

APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES

=

The Carborundum material (in the Unit 1 SFP storage racks) and the Boraflex material (in the Unit 2 SFP storage racks) is exposed to a borated water environment.

Experiments have shown that the Carborundum sheets can experience spalling and surface abrasion, which result in a loss of boron carbide, after exposures at higher levels of radiation.

Industry experience indicates that the silica filler in the Boraflex material can dissolve in the SFP environment and release the boron carbide neutron absorber. Ifleft unmanaged, degradation of these neutron-absorbing materials could eventually result in the SFP storage racks not being able to perform their intended function under CLB conditions.

The coupon surveillance program provides for periodic monitoring of the condition of neutron-absorbing materials in the SFP. The program will be modified to reevaluate the timetable and refine the scheduling process for removal of sample packets from the SFP. The program will identify and document degradation of the Carborundum and Boraflex materials, and ensure appropriate actions are taken in a timely manner if significant loss of neutron-absorbing capability is occurring.

Therefore, there is reasonable assurance that the effects of neutron-absorbing material degradation will be managed in such a way that the SFP storage racks will be capable of performing their intended function consistent with the CLB during the period of extended operation.

3.3E.3 Conclusion The aging management programs discussed for the Auxiliary Building and SR Diesel Generator Building Structures are listed in Table 3.3E-4. These programs are administratively controlled by a formal review ,

and approval process. As demonstrated above, these programs will manage the aging mechanisms and I their effects in such a way that the intended functions of the components of the Auxiliary Building and SR Diesel Generator Building Structures 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.

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Application for License Renewal 3.3E-26 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES TABLE 3.3E-4 AGING MANAGEMENT PROGRAMS FOR THE AUXILIARY BUILDING AND SR DIESEL GENERATOR BUILDING STRUCTURES Program ' Credited As Existing CCNPP Technical Procedure Program for discovery and management of STP-F-592-1/2, " Penetration Fire weathering effects for caulking, sealants, and Barrier Inspection" expansion joints that function as fire barriers in the Auxiliary Building and adjacent rooms by visual inspection. (Group 1)

Existing Operations Section Performance Program for discovery and management ofIGSCC Evaluation PE 0-67-2-O-M, "#11 & effects for the SFP liner by performing periodic

1. 12 Spent Fuel Pools - Determine leakage monitoring. (Group 3)

Liner Leakage," and associated Operating Instruction OI-24D, " Spent Fuel Pool Cooling - Infrequent Operations" Modified Structure and System Walkdowns Program for discovery and management of (MN-1-319) corrosion effects for carbon steel components in the

• Specify scope and control of Auxiliary Building and SR Diesel Generator periodic structure performance Building Structures. (Group 2) assessments Modified Engineering Test Procedure 86-03R, Program for discovery and management of neutron-

" Analysis of Neutron Absorbing absorbing material degradation for the SFP storage Material in Spent Fuel Storage racks by performing analysis of sample coupons.

Racks" (Group 4)

Reevaluate timetable and refine scheduling process for sample coupon removal from SFP New Caulking and Sealant Inspection New program for discovery and management of Program weathering efTects for caulking, sealants, and expansion joints in the SR Diesel Generator Building, as well as for those in the Auxiliary Building and adjacent rooms that do not function as fire barriers. (Group 1) i Application for License Renewal 3.3E-27 Calvert Cliffs Nuclear Power Plant

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l ATTACHMENT (4) j

' APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR l BUILDING STRUCTURES 3.3E.4 Referenees

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 Drawing 63874SH0004, "SR Ductbank Under West Plant Road Plan," Revision 0

5. BGE Drawing 63874SH0005, " Underground Conduit West of Turbine Building Plan,"

Revision 0

6. CCNPP Updated Final Safety Analysis Report, Units 1 and 2, Revision 21

7. CCNPP Aging Management Review Report," Auxiliary Building," Revision 3

8. BGE Drawing 61988, " Auxiliary Building at 45'-0" Emergency Generator Room Piping Ducts Unit 1," Revision 16

9. BGE Drawing 63988, " Auxiliary Building at Elevation 45'-0" Solid Waste Processing Area Unit 2," Revision 13

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

Revision 2

11. Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. G. C. Creel (BGE) dated January 17,1992,

" Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No.1 (TAC No. M71241) and Unit No. 2 (TAC No. M71242)"

12. CCNPP AMR Report, " Emergency Diesel Generator System (024)," Revision 1

13. CCNPP AMR Report," Component Supports," Revision 3

14. CCNPP Component Level Scoping Results for the Auxiliary Building, Revision 2

15. BGE Drawing 63874SH0001, " Underground Conduit West of Turbine Building Plan,"

Revision 3

16. BGE Drawing 63874SH0002, " Diesel Generator Project SR & SBO Manholes & Ductbank Plans, Sections & Details," Revision 2

17. BGE Drawing 63874SH0003, " Underground Conduit West of Turbine Building Plan," l Revision 0

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

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19. Electric Power Research Institute, "ClassI Structures License Renewal Industry Report; Revision 1," July 1994 l l 20. Bechtel Specification No. 6750-C-9, " Specification for Furnishing and Delivery of Concrete -

CCNPP Units 1 and 2," Revision 22

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I 21. BGE Diesel Generator Project Civil Engineering Design Report, Revision 1

22. BGE Design Specification No.SP-702, " Site Preparation and Earthwork Construction,"

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES

23. BGE Design Specification No. SP-700, " Materials Testing Services," Revision 6

24. Bechtel Specification No. 6750-C-4A, " Specification for Placement and Control of Compacted Fill- CCNPP Units 1 and 2," Revision 3

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

26. Bechtel Specification No. 6750-C-11-B, " Specification for Testing of Concrete, Reinforcement and Soil - CCNPP Units 1 and 2," Revision 1

27. NRC Inspection and Enforcement Circular 81-08," Foundation Materials," May 29,1961

28. BGE Drawing 61993, " Auxiliary Building Roofs Over Emergency Generator at Elevation 69'-0"," Revision 2

29. Bechtel Specification No. 6750-A-10, " Specification for Furnishing, Delivery and Application of the Caulking and Scalants," Revision 1

30. Bechtel Specification No.6750-C-10, " Specification for Forming, Placing, Finishing, and Curing Concrete," Revision 9

31. CCNPP Technical Procedure STP-F-592-1," Penetration Fire Barrier Inspection," Revision 3

32. CCNPP Technical Procedure STP-F-592-2," Penetration Fire Barrier Inspection," Revision 2

33. BGE " Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant," Revision 48

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

35. 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"

36. BGE Design Specification No. SP-706, " Purchase of Category 1 Structural and Miscellaneous Steel," Revision 3

37. BGE Performance Specification No. SP-707, " Erection of Category 1 Structural and Miscellaneous Steel," Revision 1

38. Bechtel Specification No.6750-C-31, " Specification fct Furnishing, Detailing, Fabricating, Painting, and Delivering Containment and Auxiliary Buitding Structural Steel - CCNPP Units 1 and 2," Revision 3

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

40. BGE Technical Requirements Document TRD-A-1000, " Coating Application Performance Standard," Revision 14

41. BGE Design Specification No. SP-717-NSR," Shop Applied Coating," Revision 2

42. Bechtel Specification No.6750-C-19, " Specification for Furnishing, Detailing, Fabricating, Delivering, and Erecting Structural Steel - CCNPP Units 1 and 2," Revision 3 k 43. Bechtel Specification No. 6750-A-24, " Specification for Painting and Special Coatings -

CCNPP Units 1 and 2," Revision 12

44. CCNPP Administrative Procedure MN-1-319," Structure and System Walkdowns," Revision 0 Application for License Renewal 3.3E-29 Calvert Cliffs Nuclear Power Plant

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,' APPENDIX A - TECHNICAL INFORMATION i

3.3E - AUXILIARY BUILDING AND SAFETY-RELATED DIESEL GENERATOR 1 BUILDING STRUCTURES

45. Bechtel Specification No. 6750-C-28," Specification for Stainless Steel Liner Plate and Spent Fuel Pool Bulkhead Gate - CCNPP Units 1 and 2," Revision 6

46. BGE Drawing 61706SH0001, " Auxiliary Building Spent Fuel Pool Liner Plan and Sections Sheet 1," Revision 18

47. BGE Drawing 61707S110002, " Auxiliary Building Spent Fuel Pool Liner Sections and Details," Revision 19

48. BGE Drawing 61708SH0003, " Auxiliary Building Spent Fuel Pool Liner Bulkhead Gates Sheet 3," Revision 14

49. BGE Drawing 61972SH0004," Auxiliary Building Spent Fuel Pool Liner Sections and Details Sheet 4," Revision 12 ,

50. CCNPP Administrative Procedure NO-1-201,"Calven Clifts Operating Manual," Revision 7

51. CCNPP Administratise Procedure NO-1-203," Operations Section Performance Evaluations,"

Revisioi. 3

52. CCNPP Administrative Procedure NO-1-100, " Conduct of Operations," Revision 9

53. CCNPP Operations Performance Evaluation Requirements Routine No. 0-67-2-0-M, "#11 &

1. 12 Spent Fuel Pools - Determine Liner Leakage," Revision 2

54. CCNPP Operating Instructions, OI-24D, " Spent Fuel Pool Cooing - Infrequent Operations,"

Revision 4

55. Letter from Mr. A. E. Lundvall, Jr. (BGE) to Mr. Robert W. Reid (NRC), dated January 15,1980," Spent Fuel Pool Modification Supplementary Information"

56. The Carborundum Company, " Handbook of the Effects ofIn-Pool Exposure on Properties of Boroa Carbide-Resin Shielding Materials,"(undated)

57. NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks,"

June 26,1996

58. Letter from Mr. S. A. McNeil (NRC) to Mr. G. C. Creel (BGE), dated March 7,1989,

" Amendment to increase Enrichment Limits for the Spent Fuel Storage Racks (TAC Nos.

68416 and 68417)

59. Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated September 18,1996,

" Resolution of Spent Fuel Storage Pool Safety Issues: Issuance of Final Staff Report and Notification of Staff Plans to Perform Plant-Specific, Safety Enhancement Backfit Analyses, Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 (TAC Nos. M96516 and M96517)"

60. Electric Power Research Institute,"A Synopsis of the Technology Developed to Address the Boraflex Degradation issue," November 1997 l 61. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated October 24,1996, "120-Day Response to Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks"

62. CCNPP Administrative Procedure EN-4-101," Coordination of Testing," Revision 1 f

l Application for License Renewal 3.3E-30 Calvert Cliffs Nuclear Power Plant I

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APPENDIX A - TECHNICAL INFORMATION 3.3E - AUXILIARY BUILDLNG AND SAFETY-RELATED DIESEL GENERATOR BUILDING STRUCTURES

63. CCNPP Engineering Test Procedure 86-03R, " Analysis of Neutron Absorbing Material in Spent Fuel Storage Racks," Revision 2

64. Letter from Mr. J. C. Linville (NRC) to Mr. G. C. Creel (BGE), dated January 30,1991, "NRC Region i Resident Inspection Report Nos. 50-317/90-34 and 50-318/90-34 (November 25,1990

- January 12,1991)"

Application for License Renewal 3.3E-31 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION l 5.10 - FIRE PROTECTION i

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Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10 Fire Protection This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA) addressing Fire Protection (FP). Fire Protection was evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology for commodities as described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LRA.

As discussed in Section 7.2.4 of the CCNPP IPA Methodology, due to the unique circumstances pertaining to the systems that perform FP intended functions (e.g., the degree of component level scoping completed elsewhere, similarity of FP functions, and degree of aging review already completed elsewhere), an aging management review (AMR) process separate and unique from that used for plant systems and structures was used.

Section 5.10.1 presents the results of the system and component level scoping process. Section 5.10.2 describes the methods used for AMR, and Section 5.10.3 provides summaries of the AMR results for each system evaluated for aging management in this section of the BGE LRA.

5.10.1 Scoping Forty-two systems are credited with performing FP functions. All components required for FP in 26 of these systems are safety-related (SR), and those systems are fully addressed in other sections of the BGE LRA. Some of the remaining 16 systems also have SR components that are addressed in other sections of the BGE LPA. Thus, the focus of this section is limited to the non-safety-related (NSR) pressure-retaining pertions of the remaining 16 systems. Scoping details are provided in the following subsections.

5.10.1.1 System Level Scoping System level scoping of the 122 systems and structures at CCNPP identified that 66 were within the

! scope oflicense renewal. [ Reference 1, Table 1] For these 66 systems and structures, those with FP functions were identified during the scoping process using the FP Screening Tool. The FP Screening Tool defines two categories of FP functions as follows: [ References 1 and 2]

FP Function:

This function includes equipment and facilities important to safety that provide for detecting, fighting, and extinguishing fires. These systems are necessary to protect SR equipment and structures from fire or explosion. This function does not include FP equipment or facilities protecting NSR equipment or structures.

Safe Shutdown Function:

This function applies to systems that provide for safe shutdown of the plant in the event of a severe fire. Calvert Cliffs' current licensing basis (CLB) requires compliance with 10 CFR Part 50, Appendix R, Sections Ill.G, DI.J, Ill.L, and 111.0. Therefore, the evaluations pertaining to safe shutdown identified those components that are required for compliance with these regulations. The safe shutdown function includes the capability to provide for:

[ Reference 2, Section 2.0]

e Reactor Coolant System (RCS) pressure and inventory control; Application for License Renewal 5.10-1 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT $

APPENDIX A - TECHNICAL INFORMATION

$.10 - FIRE PROTECTION

• Reactivity control; e lieat removal (hot standby or cold shutdown) from the RCS; and e Process monitoring.

The CCNPP Updated Final Safety Analysis Report, the FP Program licensing basis documentation, and the CCNPP Interactive Cable Analysis for each unit were reviewed to identify the system functions that address regulations on FP and BGE's commitments for implementation of those regulations. [See Section 2.0 of the BGE LRA, IPA, Section 3.3.2.1]

The FP Screening Tool identified that 42 of the 66 systems and structures within the scope of license renewal have one or more FP intended fun:tions. These 42 systems and strue:ures are listed in Table 5.10-1. [ Reference 1, Table 2]

5.10.1.2 Systems and Structures Addressed In Other Sections of the BGE LRA Evaluation of all components required for FP for 26 of the 42 CCNPP systems and structures with passive FP intended functions are included within their respective SR system or structural AMR or in a commodity evaluation. These systems and structures fall into one of the following three categories:

1) structures with components that provide a fire barrier; 2) fluid systems with components that provide part of a pressure boundary (PB) in systems with only SR PB components; and 3) electrical systems with components that perform only active electrical functions.

Structures The only passive FP intended functions performed by components in five systems and structures listed in Table 5.10-1 are to provide rated fire barriers to confine or retard fires from spreading to or from adjacent areas of the plant. Rated fire barriers include doors, walls, floors (and curbs), ceilings, penetration seals, and cable tray fire barrier materials. These components are in the following systems and structures: [ References 3 through 8]

System 009 - Intake Structure System 059 -

Primary Containment System 120 - Barriers and Barrier Penetrations N/A - Auxiliary Building N/A - Turbine Building The results of the AMR for the components in these systems and structures that perform this passive FP intended function are provided in Sections 3.3A,3.3B,3.3C, and 3.3E of the BGE LRA. These five systems and structures are not addressed further below. Note that the fire barrier components for the Barriers and Barrier Penetrations System are pan of the four structures listed above and are, therefore, not addressed as a " system" in the IPA process, but as part of those structures. [ Reference 1]

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION TABLE 5.10-1 SYSTEMS AND STRUCTURES WITHIN THE SCOPE OF LICENSE RENEWAL WITH FIRE PROTECTION FUNCTIONS System # ' Description 1 002 Electrical 125 Volt DC Distribution 2 004 Electrical 4 kV Transformers and Buses 3 005 Electrical 480 Volt Transformers and Buses 4 006 Electrical 480 Volt Motor Control Centers 5 008 Well and Pretreated Water ,

6 009 Intake Structure 7 011 Service Water (SRW) 8 012 Saltwater 9 013 FP 10 015 Component Cooling (CC) 11 017 Instrument AC 12 018 Vitalinstrument AC 13 019 Compressed Air 14 023 Diesel Fuel Oil 15 024 Emergency Diesel Generators 16 026 Annunciation 17 029 Plant Heating 18 030 Control Room Heating, Ventilation and Air Conditioning (HVAC) 19 032 Auxiliary Building and Radwaste Heating and Ventilation (H&V) 20 036 Auxiliary Feedwater(AFW) 21 037 Demineralized Water and Condensate Storage 22 041 Chemical and Volume Control 23 044 Condensate 24 045 Feedwater 25 052 Safety Injection -

26 053 Plant Drains 27 055 Control Rod Drive Mechanism and Electrical 28 059 Primary Containment 29 060 Primary Containment H&V 30 061 Containment Spray 31 064 Reactor Coolant 32 071 Liquid Waste 33 074 Nitrogen and Hydrogen Gas System 34 078 NuclearInstrumentation 35 083 Main Steam 36 093 Main Turbine 37 096 Fire and Smoke Detection 38 097 Lighting and Power Receptacles 39 100 Plant Communications Y

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APPENDIX A - TECHNICAL INFORMATION 5.I0 - FIRE PROTECTION TABLE 5.I0-1 SYSTEMS AND STRUCTURES WITHIN THE SCOPE OF LICENSE RENEWAL WITH FIRE PROTECTION FUNCTIONS System # Description -

40 120 Barriers and Barrier Penetrations 41 N/A Auxiliary Building 42 N/A Turbine Building Fluid Systems The only passive FP intended function performed by components in eight systems and structures listed in Table 5.10-1 is the PB function (e.g., piping, valve bodies, etc.). All of these components, with one exception noted below, also provide the SR PB function. Therefore, AMR of these components is included in the AMR for the passive SR PB intended functions. These systems, and the sections of the BGE LRA where the AMR results for each is provided, are as follows: [ References 9 through 24]

System 012 - Saltwater LRA Section 5.16 System 024 - Emergency Diesel Generators LRA Section 5.8 System 030 - Control Room HVAC LRA Section 5.1IC System 032 - Auxiliary Building H&V LRA Section 5.11 A System 045 - Feedwater LRA Section 5.9 System 052 - Safety Injection LRA Section 5.15 System 060 - Primary Containment H&V LRA Section 5.11B System 061 - Containment Spray LRA Section 5.6 The exception is that part of the passive FP-related PB of System 030 is not SR. Ilowever, the results of the AMR provided in Section 5.11C of the BGE LRA addresses this NSR portion, as well as the SR parts. The eight systems listed above are not addressed further below.

Electrical Systems There are 13 electrical systems listed in Table 5.10-1 with components that perform FP intended functions that are active. Those systems require no further evaluation since the remaining intended functions that are passive, i.e., electrical continuity and component support, are addressed in other commodity evaluations. These systems are: [ References 2 and 25 through 48]

System 002 - Electrical 125 Volt DC Distribution System 004 - Electrical 4 kV Transformers and Buses System 005 - Electrical 480 Volt Transformers and Buses System 006 - Electrical 480 Volt Motor Control Centers System 017 - Instrument AC System 018 - VitalInstrument AC l l

System 026 - Annunciation i System 055 - Control Rod Drive Mechanism and Electrical System 078 - Nuclear Instrumentation System 093 - Main Turbine System 096 - Fire and Smoke Detection Application for License Renewal 5.10-4 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION System 097 - Lighting and Power Receptacles System 100 - Plant Communications l

These 13 systems are not addressed further in this report.

5.10.1.3 Systems and Structures Addressed as Part of FP As described above, all components required for FP in 26 of the 42 CCNPP systems and structures with passive FP intended functions are fully addressed within their respective SR system or structural AMR or in a' commodity evaluation. The remaining 16 systems are in the scope of the FP AMR and are addressed in this report. These systems fall into one of the following two categories: 1) systems that have undergone component level scoping; and 2) systems that have not been previously scoped because they primarily have only NSR functions.

5.10.1.3.1 Systems With Prior Component Level Scoping Nine systems that perform passive FP in; ended functions have both SR and NSR PB components. The SR portions of these systems are addressed in other sections of the BGE LRA since these SR systems had component level scoping and AMR performed due to their non-FP intended functions. The NSR PB portions of these systems are addressed later in this report. These nine systems, and the sections of the BGE LRA where the SR PB portion ofeach i. addressed, are as follows:

System 0ll - SRW Section 5.17 l System 015 - CC Section 5.3 l System 019 - Compressed Air Section 5.4 l System 023 - Diesel Fuel Oil Section 5.7 )

System 036 - AFW Section 5.1 l System 041 - Chemical and Volume Control Section 5.2  !

System 064 - Reactor Coolant Section 4.1 l System 074 - Nitrogen and Hydrogen Gas System Section 5.12 i System 083 - Main Steam Section 5.12 j 5.10.1.3.2 Systems 'Without Prior Component Level Scoping  !

Seven systems rely almost entirely on NSR components to perform their passive FP intended functions,  ;

and there was no component level scoping or AMR performed for each individual system. These systems are addressed in this section of the BGE LRA and are as follows: 4 System 008 - Well and Pretreated Water System 013 - FP

• System 029 - Plant Heating

• System 037 - Demineralized Water and Condensate Storage

• l System 044 - Condensate  ;

! System 053 - Plant Drains

• System 071 -

' Liquid Waste *

• These five systems also have a passive intended function, i.e., containment isolation, that is SR. The AMR results below address the NSR FP intended functions. The AMR results for the SR components in these systems are provided in the Containment Isolation Group, Section 5.5 of the BGE LRA.

Application for License Renewal 5.10-5 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION l 5.10 - FIRE PROTECTION 5.10.1.4 Component Level Scoping For most systems and structures within the scope of license renewal, a detailed list of components contributing to an intended function of the system or structure was produced. For some systems with passive FP intended functions, component level scoping was performed the same way, but for others, it was performed by characterizing the extent of the system that supports such functions. This was accomplished by defining the boundary (or envelope) of the important pressure-retaining features of the system in terms of major components or interfaces with other systems, and by identifying the specific l device types that fell within that boundary (or envelope). [ Reference 2, Appendix B, Task 1, Section 6.3]  !

This is an acceptable method of component level scoping since the components subject to AMR can be l readily determined from review of the device type lists and drawing references. [ Reference 49]

5.10.1.4.1 Components Addressed in Commodity Evaluations Some components with FP functions 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. i These components include the following: [ Reference 2, Section 2.0)

Structural supports for piping, cables, and components are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA.

• 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.

• Electrical Pnels that support and/or protect electrical components are evaluated for the effects of P ,

aging in the Electrical Panels Commodity Evaluation in Section 6.2 of the BGE LRA. J e

Instrument tubing and piping and the associated tubing supports, instrument valves and fittings (generally everything from the outlet of the final root valve up to and including the instrument),

and the PBs 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.

5.10.2 AMR Methods During normal operation, SR systems and components typically do not operate under their design conditions. The demands placed on most SR systems and components during normal operation are much less than the demands placed on them during mitigation of design basis events. Some SR systems, such as Safety Injection and Containment Spray, do not normally operate and are maintained in a continuous standby mode. The SR systems and components on standby do not demonstrate they are capable of performing any intended functions during normal day-to-day operations. Therefore, functional tests of SR systems and components have been devised to demonstrate their ability to perform active intended functions. But these tests are not suitable for demonstrating the ability to perform passive intended functions. This is because the tests are incapable of simulating the loading conditions (e.g., seismic accelerations or other dynamic loading) under which SR components are required to perform. Therefore, methods of demonstrating aging management, such as condition assessments or inspections, are required for SR systems and components.

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION The above stipulations are not true for NSR components such as those being addressed in this section of the BGE LRA. He demands placed on most NSR systems and components during normal operation are the same as, or greater than, the demands placed on them during mitigation of fires. Because they are NSR, they are not designed to operate under postulated dynamic loading conditions such as seismic accelerations. Therefore, operation of the system during normal operations is an adequate test of the j system for FP design loading conditions. Demonstration that the active FP intended functions are

{'

capable of being performed also demonstrates that the passive FP intended functions are capable of being performed. In other words, it can be shown that most NSR systems and components demonstrate they are capable of performing their passive FP intended functions along with their active FP intended functions during normal routine operation, testing, or inspection activities. (Reference 2, Section 2.0]

As such, four different methods were applied to demonstrate aging management of these NSR components. Refer to Figure 5.10-1 on the following page for an illustration of this process.

The first three methods were applied in sequential order to demonstrate that aging effects for an entire system, or portions ofit, could be adequately managed without a specific determination of Age-Related Degradation Mechanisms (ARDMs). In this manner, the scope of the system requiring further review was reduced with application of each succeeding method. Device types not addressed by any of these first three methods required an AMR that identified the plausible ARDMs and the appropriate aging management programs. It should be noted that, in some cases, system components may have aging effects managed by more than one of the four methods. However, since the end result would be the same, the approach using successive incremental methods was used without identifying all possible management alternatives. Table 5.10-2 on the following page lists the 16 systems that are evaluated for aging management in this section. [ Reference 2, Table ES-1] The table has four additional columns identifying which of the four methods was used to demonstrate agina management. The methods are explained below. [ Reference 2]

Representative historical operating experience pertinent to aging is included where appropriate to provide insight supporting the aging management demonstrations. This operating experience was obtained through documented discussions with currently assigned cognizant CCNPP personnel.

Key word searches of BGE's electronic database of information on the CCNPP dockets was also performed in order to obtain pertinent operating experience.

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Figure 5.10-1 FP AMR Process NSR Pressure Boungary Portion of Systems Subject to FP AMR FPP Activities No Manage Aging Yes No Activities Manage Aging Yes SR PB No Activities manage Aging 1,

Yes AMR Conducted for Remaining Components FP AMR Complete Application for License Renewal 5.10-8 Calvert Cliffs Nuclear Power Plant

A'ITACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION TABLE 5.10-2 FP AMR

SUMMARY

OF RESULTS Performance FP Vrogram and SR PB System Activities Condition AMR AMR System Monitoring No. Manage Manages Conducted Aging Activities * "

Aging Manage Aging Well and Pretreated Water 008 No Yes isN/A)  ![N/A) 5.10.3.1 SRW 011 No Yes ' sN/Ns jN/A} 5.10.3.2 FP 013 Yes BN/A9> RN/Ai jijiN/A% 5.10.3.3 CC 015 No Yes c SN/As )N/A? 5.10.3.4 Compressed Air 019 No Yes jiN/Ai! s 1N/As 5.10.3.5 Diesel fuel Oil 023 Yes iN/A) iN/A5 (N/A!!!! 5.10.3.6 Plant Heating 029 No Yes  !!N/A] [N/At 5.10.3.7 AFW 036 Yes(Partial) Yes (Partial) inn /A? , iN/A7 5.10.3.8

'bmneralized Water & 037 No Yes j!N/A) > {N/Aj 5.10.3.9 Condemate Storage Chemical and Volume 041 No No Yes >

>sN/As. 5.10.3.10 Control W Condensate 044 No Yes (Partial) No Yes (Partial) 5.10.3.11 Plant Drains 053 Yes(Partial) Yes(Partial) $N/A$ RN/Ai 5.10.3.12 Reactor Coolant 064 No No Yes $N/Ai 5.10.3.13 Liquid Waste 071 No Yes liN/A? yN/As - 5.10.3.14 Nitrogen & ilydrogen Gas 074 Yes djN/A3 i[N/A) , ?N/A3 5.10.3.15 Main Steam 083 No Yes (N/Aj $N/A1 5.10.3.16 5.10.2.1 FP Program Activities Manage Aging This is the first step in the sequential process described above. This method demonstrates that the aging effects on a system's NSR pressure-retaining components are adequately managed by specific performance and/or condit;on monitoring activities required by the plant's FP Program. The Nuclear Program Directive S A-1, " Fire Protection Program," establishes requirements and assigns responsibilities for the FP Program at CCNPP. The FP Program is the integrated effort involving components, procedures, and personnel used to carry out all activities of FP Program and prevention. It contains maintenance, testing, and inspection criteria to provide reasonable assurance that various NSR systems are capable of performing their FP intended functions. Any abnormal condition would be j detected and investigated to ensure that it does not have the ability to impact safety or adversely affect  :

operation of the system. Any such condition would be repaired prior to impacting the passive FP intended function of the system in question. The site's FP Program is part of the plant's CLB. )

[ Reference 2, Section 6.1; Reference 50] 4 Application for License Renewal 5.10-9 Calvert Cliffs Nuclear Power Plant a

ATTACHMENT S APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5 order to demonstrate adequate aging management for each system's components through various FP Trogram performance and/or condition monitoring activities, the following tasks were performed:

[ Reference 2, Section 6.1]

• The system's intended functions were identified.

The applicable performance and/or condition monitoring activities (e.g., maintenance, testing, and inspection activities) required by the FP Program for the system were identified.

The performance and/or condition monitoring activities applicable to the system's passive FP intended functions were identified.

The NSR pressure-retaining components within the portion of the system tested by the FP Program performance and/or condition monitoring activities were identified.

For systems and components to which it applies, this method shows that the effects of aging will not impact FP intended functions during the period of extended operations. Where this type of demonstration was successful, the FP Program is credited as the appropriate aging management program.

[ Reference 2, Section 6.l]

The FP Program provides the necessary controls to protect the health and safety of CCNPP workers and the general public, satisfy NRC and Insurer requirements, meet applicable State of Maryland codes and standards, and safeguard BGE assets by preventing fires and minimizing the consequences of any fire that may occur. A discussion of the CCNPP FP Program is presented in Section 9.9 of the UFSAR.

[ Reference 51] The program addresses: [ Reference 50, Section 1.2]

• Fire protection aspects of structure system, and component design;

• Inspection and testing of FP systems and equipment;

• Procurement of FP equipment and material; e Controls for the prevention of fires; e Fire fighting; e Fire prevention and response training; e Monitoring and continuous assessment of the FP Program; and e Auditing of the FP Program.

Fire protection equipment and systems are inspected and tested upon initial installation and periodically thereafter. [ Reference 51] Inspections ensure that the installation, maintenance, and modification of FP equipment conform to design requirements. [ Reference 50] The inspection and testing is conducted following the guidance of applicable National Fire Protection Association Codes and Standards, as well as recommendations and requirements of the insurance carrier and the NRC. Plant procedures mandate test frequencies and the testing process. Applicability, compensatory actions, testing requirements, and ,

testing frequencies for those FP systems that protect safe shutdown and SR equipment are contained in l the CCNPP Technical Specifications. [ Reference 52] Plant procedures also identify compensatory actions to be taken when equipment required for 10 CFR Part 50, Appendix R, safe shutdown actions becomes inoperable. [ Reference 51]

Application for License Renewal 5.10 10 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Activities related to FP are performed within the applicable provisions of the BGE's Quality Assurance Program based on 10 CFR Part 50, Appendix B, and in accordance with the quality assurance guidance in Branch Technical Position 9.5-1, Appendix A, and the NRC's guidance document, " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance." [ Reference 51]

Internal assessments of the FP ProFram are conducted through periodic audits in accordance with Quality Assurance Policy requirements. [ Reference 50]

5.10.2.2 Performance and Condition Monitoring Activities During Normal Operation Manage Aging This is the second step in the sequential process and is applied if the FP Program method above is not completely successful in demonstrating aging management of a system's NSR components. As was noted previously, the demands placed on most NSR systems and components during normal operation are the same as, or greater than, the demands placed on them during mitigation of fires. Herefore, satisfactory performance of periodic functional tests can be used to demonstrate that aging is adequately managed for the passive FP functions of NSR components. A system that is in continuous operation during normal operation can be characterized as undergoing a continuous FP functional test if the system parameters (pressure, temperature, flow, etc.) encountered during performance of FP intended functions are bounded by the normal operating parameters of the system. The performance and condition monitoring activities conducted in accordance with procedures such as MN-1-319, Structure and System Walkdowns, and NO-1-100, Conduct of Operations, ensure detection of abnormal conditions. MN-1-319 stipulates in part that the intent of the walkdowns is to identify and record any new or existing condition that could prevent a system or component from performing its intended function. Conditions adverse to functionality, indicatiens of system or equipment stress or abuse, safety or fire hazards and housekeeping deficiencies are identified. Walkdowns are scheduled for plant conditions that provide good indications

{

of system functionality. NO-1-100 requires that operators be accountable for their immediate areas of responsibility. This includes performing general inspections and checking the condition of areas and -i equipment. Operators assess degraded equipment conditions to ensure personnel and affected equipment j safety while completing corrective actions. [ References 53 and 54] It should be noted that many of the i

systems are required to deliver water for their FP function, whether for fire fighting or for safe shutdown.

For those systems, absolute leak tightness is not required.

Where the above type of demonstration is successful, performance and condition monitoring activities j durir.g normal operation are credited for identifying the effects of system aging. Specific aging management programs are not necessary, and no further evaluation is required. [ Reference 2,  ;

Section 6.2] A more detailed description of MN-1-319 and NO-1-100 are provided below.

System Walkdawns

)

4 The Structure and System Walkdown Program has been established to standardize the general intent and l method of conducting walkdowns and of reporting the walkdown results. This procedure meets the

{

requirements for evaluating structure and system material condition in accordance with the (NRC) i Maintenance Rule at CCNPP. -Walkdown activities provide for discovery of many ARDMs by performing periodic visual inspections for evidence of aging.' When degraded conditions are identified, more detailed inspections are performed and/or corrective actions are taken to repair the deficiency.

[ Reference 53]

l l

~

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ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Under this program, 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 after outages; start-up reviews (i.e., when a system is pressurized, re-energized, or placed into normal service); and as required for plant modifications. Inspection items typically related to aging management include identifying unusual noises, leaks, corrosion, or degraded paint and identifying system and equipment stress or abuse, such as excessive vibrations, bent or broken component supports, loosened fasteners, etc. [ Reference 53, Sections 5.1 and 5.2]

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 intended functions. Conditions adverse to quality are documented and resolved by the Calvert Cliffs Corrective Actions Program. [ Reference 53, Sections 5.1.C,5.2.A.1, and 5.2.A.5; Reference 55]

The program provides guidance for specific types of degradation or conditions to inspect for when performing the walkdowns. General inspection items related to aging management include the following: [ Reference 53, Section 5.2 and Attachments I through 13]

e items related to specific ARDMs such as corrosion; e Effects that may have been caused by ARDMs such as damaged supports, concrete degradation, anchor bolt degradation, or leakage of fluids; and e Conditions that could allow progression of ARDMs such as degraded protective coatings, leakage 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 specific activities to be included in the scope of the walkdowns.

Conduct of Ooerations Administrative procedure NO-1-100 addresses the controls and basic standards for conduct of daily shift operations, Control Room access and conduct, special evolutions and tests, and briefings. This procedure requires that operators assess degraded equipment conditions to ensure personnel and affected equipment safety while completing corrective actions. [ Reference 54] For those system (s) and component (s) where the system parameters during performance of FP intended functions are bounded by the normal system operating parameters, performance and condition monitoring activities during normal operation provide for discovery of unspecified aging effects by visual inspection and assessment of degraded conditions.

Application for License Renewal 5.10-12 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Administrative procedure NO-1-100 establishes the responsibilities and authority of operating shift personnel for the daily conduct of plant operations. It serves as a governing procedure for a wide range of performance and condition monitoring activities during normal operation. Some of the performance and condition monitoring activities that are controlled by this procedure include the following:

[ Reference 54]

Operator Rounds - visual inspections of operating spaces each shift during plant operator rounds; )

Plant Logs - co!!ect selected data for operating equipment and analyze it to detect abnormal or degraded equipment performance; l

Operations Section Performance Evaluations - periodic checks to determine equipment performance, as determined by manufacturers' recommendations, System Engineers' recommendetions, and i operating needs; Surveillance Testing - surveillance requirements speciied by CCNPP Technical Specifications to verify that SR structures, systems, and components continue to function or are in a state of readiness to perform their functions; and Troubleshooting - diagnosing plant / equipment symptoms for the purpose of identifying / quantifying a degraded pararacter/ component or verifying the operability of a component. 1 Opuator rounds have historically been effective in identifying plant deficiencies. The documented guidance and expectations have been improved over the years as a result oflessons learned and the site emphasis on continual quality improvement. Plant operating practices are also periodically evaluated by )

the NRC as part of their Systematic Assessment of Licensee Performance efforts.

5.10.2.3 AMR of SR PB Components Manages Aging The third step in the sequential process is applied if the FP Program and normal operating condition ,

methods above are not completely successful in addressing a system's aging management. This method '

applies to the NSR portion of SR systems for which there is an AMR that determined plausible ARDMs and addressed management of the aging efTects. This method recognizes that similar materials subjected to similar process fluids and environmental service conditions can reasonably be expected to have the same plausible aging effects, and can be managed in the same manner regardless of whether a system's l components are classified as SR or NSR. [ Reference 2, Section 6.3]

This method involves the following tasks: [ Reference 2, Section 6.3]

e Review the results of a system's SR PB AMR with specific focus on the plausible ARDMs identified and the aging management programs.

• Determine if the plausible ARDMs are equally applicable to NSR PB components of the FP AMR through similarities in equipment types, device types, materials of construction, process fluids, exterior environments, operating conditions and other service conditions.

l

• Determine if the programs credited for the SR PB components are applicable to the NSR PB l components subject to the same ARDMs. If the Age-Related Degradation Inspection (ARDI) l Program is used to manage aging, add the NSR PB components to the scope of the ARDI.

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ATTACHMENT 8)

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION In completing the above steps, component make, model and other component-specific information was not always identified for components subject to the FP AMR solely for the purpose of determining the applicability of the SR review results. Rather, the similarities between the two portions of the system were characterized based on a review of design specifications (e.g.,the pipe class and valve type specification sheets) supplemented with a review of the system descriptions, other available design documents, and any other appropriate information. [ Reference 2, Section 6.3]

This method demonstrates that aging of NSR PB components with passive FP intended functions is adequately managed when they are subject to the same aging management activities as similar SR PB components.

. 5.10.2.4 AMR Condacted for Remainlag Composeats One system, the Condensate System, was found to have components not addressed by the three methods discussed above. For portions of the Condensate System, it was necessary to use the fourth method.

This method utilizes the normal IPA process used for most systems and structures, which includes identifying system intended functions, plausible ARDMs, and methods to manage aging effects.

5.10.2.5 Summary of FP AMR Results e

Aging of all components in scope for three systems is fully managed by the FP Program (Systems 013,023 and 074).

Aging of all components in scope for eight systems is fully managed by performance and condition monitoring activities during normal operation. (Systems 008,011,015,019,029,037, 071 and 083)

• Aging of all components in scope for two systems is fully managed by a combination of the activities associated with the FP Program and performance and condition monitoring activities during normal operation. (Systems 036 and 053)

Aging of all components in scope for two systems is fully managed by programs identified for similar SR PB components. (Systems 041 and 064)

Aging of all components in scope for one system is fully managed by a combination of the performance and condition monitoring activities during normal operation and the ARDI Program as determined by identification of plausible ARDMs. (System 044) 5.10.3 Systems The remainder of this report provides the results of the review for the 16 systems listed in Table 5.10-2.

For each, a brief discussion of the scoping is provided along with the aging management demonstration.

5.10.3.1 Well and Pretreated Water Syr. em [ Reference 2, Appendix A, System 008]

l The Well and Pretreated Water System wasists of three ground wells, three submersible pumps, tivo '

activated carbon filters, two pretreated water storage tanks (PWS1s), and two pretreated water booster pumps. Each of the storage tanks is equipped with a heat excheger and a circulating pump. The Fystem interfaces with the Domestic Water System, Demineralized Water and Condensate Storage System, the warehouse and switchyard control house domestic water subsystem, the FP System, and the Plant licating System.

Application for License Renewal 5.10-14 Calvert Clifts Nnclear Power Plant

ATTACHMENT m l

APPENDIX A - TECHNICAL INFORMATION j 5.10 - FIRE PROTECTION The fire pumps take suction from two 500,000 gallon capacity (each) PWSTs. He layout of the pump suction piping from the tanks is such that a minimum of 300,000 gallons (each tank) is always available to the FP System. %e remaining 200,000 gallons (each tank) may be used for other services and also is .

available for FP System supply backup.  :

l During normal operation, the Well and Pretreated Water System is the source of all makeup water for power production, fire fighting and potable water systems. The system pumps operate intermittently to provide makeup water to the PWSTs on an automatic basis to ensure minimum capacity requirements are maintained. The level of the PWSTs is monitored to provide continuous verification of the required capacity / There are two separate alarm annunciators (Control Room and fire pump house) to indicate if the level in either tank drops below 303,000 gallons.

The tanks are supplied by three well water pumps with a nominal combined capacity of 966 gallons per minute to ensure replenishment of the 300,000 gallons within eight hours. The valve in the interconnection piping between the tanks is maintained locked closed to preclude inadvertent draining of both tanks should a leak develop in one tank or its piping. Any of the well water pumps can be aligned to either or both tanks, the FP header and ultimately to the fixed fire suppression system, hoses, and hydrants.

5.10.3.1.1 Operating Experience In 1997, the No.13 well water header developed leaks due to corrosion. This one portion of the system was particularly susceptible to corrosion due to it being a carbon steel pipe without a protective wrap, lack of a cathodic protection system, and its location in an area with groundwater flow. Other portions of J'

the system that have been uncovered and inspected have been in excellent condition primarily due to adequate coating and wrapping. The corroded portion is currently being replaced.

IIcavy corrosion has been discovered on selected penetrations on the PWSTs frorr failed coatings due to heat from installed heat tracing. Some penetrations have been replaced as required, additional  ;

penetrations were inspected and sandblasted, and then all penetraticr.r. s.ere coated with a high j temperature coating.

l 5.10.3.1.2 Scoping Summary i The FP function of this system is to provide water to the FP System for suppression of fires in the plant.

- Dere is also a safe shutdown-related FP function to support RCS heat removal by providing an alternate  !

source of water to the steam generators via the FP and AFW Systems. At reduced steam generator l pressure, the diesel-driven fire pump can be used to supply steam generator inventory via a fire hose and spool piece connected to direct fire main water to the AFW System. Also, a well water pump can be l used to supply makeup water from a PWST to a condensate storage tank (CST), which is a water supply  !

source for the AFW System. De line-up is accomplished by connecting a fire hose between the fire pump house test manifold and a CST emergency hose connection.

De portion of the Well and Pretreated Water System within scope for FP includes components in the flow path from the well water pumps to the PWSTs and the associated pretreated water booster pumps. 3 The following passive FP intended function (not addressed in other evaluations) applies: j

• Maintain pressure-retention capability of the system (liquid and/or gas).

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.3.1.3 ~ Aging Management Demonstration As it relates to the requirement to retain system pressure, the parameters of the system while performing the required FP functions are no different than the normal operating parameters. Under certain fire scenarios, the system may have to provide water at a higher flow rate than that required for normal plant makeup requirements. In some cases water would have to be provided directly from the wells to refill tanks or to keep the supply header pressurized. The quantity or rate of water usage under fire suppression or safe shutdown scenarios, however, is not an issue since these are clearly active functions of the system. At issue is the assurance of system pressure-retaining capability. The fire suppression or safe shutdown functions supported by the system do not challenge the system pressure-retaining capability any more than normal operating conditions. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of all components in scope for this system is fully managed by performance and condition monitoring activities during normal  ;

operation.

5.10.3.2 SRW System [ Reference 2, Appendix A, System 011]

The SRW System is a closed loop system and uses plant demineralized water with a corrosion inhibitor. j The system removes heat from turbine plant components, blowdown recovery heat exchangers, l containment cooling units, spent fuel pool cooling heat exchangers, and emergency diesel generator heat l exchangers. The system is divided into two subsystems in the Auxiliary Building to meet single failure

)

criteria. Each subsystem has a head tank to maintain the subsystem's pressure and to allow for thermal expansion. The SRW additive tank is connected to both subsystems to allow chemical addition to  ;

control and minimize corrosion. I 5.10.3.2.1 Operating Experience Representative historical operating experience pertinent to aging is included in the AMR discussion for the SRW System in Section 5.17 of the BGE LRA.

5.10.3.2.2 Seoping Summary For nearly all fires, the required SRW flow path is the normal system line-up. The FP function of the SRW System is to provide cooling water to emergency diesel generators, containment air coolers, instrument air compressors, and plant air compressors to ensure safe shutdown in the event of a fire. '

Diesel Generators I B,2A, and 2B receive cooling water from SRW Headers 12, 21, and 22, respectively.

Diesel _ Generators OC and 1A have self-contained cooling systems and are not supplied by the SRW System. The containment air coolers maintain the containment temperature less than 120 F. Depending on the location of a fire, alternate SRW valve line-ups, including cross-connecting unit headers, may be required to provide a heat sink for the containment coolers. The system also supplies cooling water to the instrument air compressors and plant air compressors that are required to support various loads during shutdown following a fire. Normally, one of the instrument air compressors is used for supplying the control air for air-operated valves.

The portions of the system that provide cooling to the diesels and to the containment air coolers are all SR and are addressed in Section 5.17 of the BGE LRA. The NSR portion of the system in scope for the FP AMR includes the components that retain pressure in the cooling process flow paths to the instrument air and plant air compressors. Since most of the NSR loads are serviced by a common header, most of the NSR portions of the system must maintain pressure to allow cooling water to be supplied to the air Application for License Renewal 5.10-16 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION compressors. This includes open line connections to other NSR equipment that form the boundaries of l

the system subject to the FP AMR. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 56]

e Maintain the PB of the system liquid.

5.10.3.2 3 Aging Management Demonstration The operation of the NSR portion of the SRW System during a fire is the same as normal operation with respect to providing cooling to the air compressors. The parameters of the system while performing the required FP functions are no different than the normal operating parameters. The use of the system during a fire will not challenge the pressure-retention. ability of the system more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of all NSR components in scope for this system is fully managed by performance and condition monitoring activities during normal operation. i 5.1033 FP System [ Reference 2, Appendix A, System 013]

The FP System is designed using the guidance of National Fire Protection Association codes and in accordance with insurance requirements, NRC requirements, and applicable Maryland State codes. The FP System is made up of several subsystems: deluge water spray, preaction sprinklers, automatic ,

sprinklers, indoor and outdoor hose stations, Halon, foam, and portable extinguishers. l l

The deluge water spray systems protect the steam generator feed pumps, hydrogen seal oil unit, unit transformers, and service transformers. The preaction sprinkler systems protect the Diesel Generator Rooms, and manually-actuated systems protect the turbine generator bearings. The automatic sprinkler systems protect many areas / rooms coataining redundant trains of safe shutdown equipment located within the Auxiliary Building, as well as the following areas: Lube Oil Room, Warehouses, Service Buildings, Paint Shop, Baling and Drumming Room, Turbine Building under the operating floor and intermediate floor, and the Auxiliary Boiler Room. Dry pipe automatic sprinkler systems protect the equipment hatch access buildings. Hose stations provide protection for the Auxiliary Building, Intake Structure, Containment Structures, Turbine Building, and Service Buildings. The Halon system protects the Cable Spreading Rooms and contiguous cable chases, Switchgear and Electrical Equipment Rooms, and under the Computer Room floor. The foam system is manually released to protect the outdoor fuel storage tanks. The foam storage tank is located outdoors. Portable fire extinguishers are provided at convenient and readily accessible locations throughout the plant. 1 Fire protection water is supplied by two full-capacity fire pumps from the PWSTs. One is electrically-driven, the other is diesel-dr:ven. A jockey pump is provided to maintain the FP water system full and pressurized. A booster pump takes suction from plant SRW and discharges to the system to meet intermittent water usage requirements other than FP. All systems are enunciated in the Control Room.

5.1033.1 Operating Experience l Over the years, generally favorable system performance has been attributed in part to the use of well l water stored in a closed tank. The use of such water results in low levels of organic materials in the ]

piping, which helps to minimize microbiologically-induced corrosion. During the recent installation of 1 the new diesel generators, there was a need to tie into the existing main loop in the protected area. This l

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ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION loop is important to safety in that it supplies fire fighting water to SR structures, systems, and components. Opening of the water main allowed an inspection of the interior and exterior surfaces, which shovced no evidence ofcorrosion, even though it was installed in the early 1970s.

Cases of unacceptable leakage have occurred in a portion of the system that supplies water to the warehouses. The leaks were promptly isolated from the main header and repaired so as not to impact plant operation. The warehouse portion of the system was originally considered temporary, so it was installed underground without cathodic protection. Now that the warehouses have remained in use, leaks in this section of the system have occurred. Some leaks were attributed to corrosion of the piping that lacks cathodic protection, and some were attributed to damage from heavy loads (vehicles) passing over the buried pipes. Leaks are detected by the amount of time the jockey pump runs to maintain system pressure and from physical changes of the ground around the leak. Monitoring thejockey pump run time provides for a continuous test of system leakage. An increase in jockey pump run time would lead to initiation ofcorrective actions to identify and repair unacceptable system leakage.

5.10.3.3.2 Scoping Summary He FP functions of the FP System are:

e Protect personnel, structures, and equipment from fire utilizing fixed fire suppression equipment, including:

- Fire pumps, piping systems, and water supply;

- Automatic water suppression systems; 1

- Manual water and foam suppression equipment and systems; and l - Automatic Halon suppression systems. 4

• Provide water curtains as rated fire barriers for unrated hatches and doors.

l

• Provide pressurized fire fighting water to hose stations inside containment. j e Provide isolation for ventilation duct penetrations to limit the spread of fire (automatic fire dampers).

The safe shutdown functions of the FP System are:

• Provide alternate makeup water via fire hose connections to the CSTs to support RCS heat removal.

e Provide an alternate source of head tank makeup water for the CC and SRW Systems, via fire l hose connection, to support RCS heat removal.

• Provide an alternate water source to the steam generators via the AFW System through a spool piece and fire hose for decay heat removal, at reduced steam generator pressure.

• Provide an alternate source of cooling water to the instrument air and plant air compressors via l fire main hose connections.

The portion of the system in scope for the FP AMR includes the pressure-retaining fire fighting equipment that performs one or more of the intended functions listed above. This includes the AFW spool piece and hose stations in the protected area. The containment isolation intended function of the l

l l

Application for License Renewal 5.10-18 Calvert Cliffs Nuclear Power Plant !

ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION  !

5.10 - FIRE PROTECTION fire suppression water main is addressed in Containment Isolation, Section 5.5 of the BGE LRA. The l following passive FP intended function (not addressed in other evaluations) applies:

Maintain the pressure-retaining capability of the system (liquid and/or gas).

5.10.3.3.3 Aging Management Demonstration Fire Protection Program activities are credited with maintaining / verifying the ability of the FP System to perform active and passive FP intended functions. Aging effects on the NSR pressure-retaining j components are adequately managed by the following specific performance and/or condition monitoring activities required by the plant's FP Program:

• STP-M-021-0 Fire Pump Diesel Inspection (every 549 days)

• STP-F-076-0 Staggered Test of Electric Fire Pump (every 31 days)

• STP-F-077-0 Staggered Test of Diesel Fire Pump (every 31 days)

• STP-M-190-0 Diesel Fire Pump Battery Weekly Check (every 7 days)

• STP-F-290-0 Hose Station and Hydrant House Inspection (every 31 days)

• STP-F-291-0 Halon System Valve Position Verification (every 31 days) e STP-M-390-0 Fire Pump Battery Quarterly Check (every 92 days) e STP-F-489-0 Halon System Nozzle and Piping inspection (every 366 days)

• STP-F-492-0 Halon System Tank Level and Pressure Verification (every 184 days)

• STP-F-493-0 Fire Suppression Water System Flush Test (every 366 days)

• STP-F-495-0 Visual Inspection of Yard Fire Hydrants (every 184 days)

• STP-F-496-0 Yard Fire Hydrant Hose Hydrostatic Test and Gasket Inspection (every 366 days)

• STP-F-497-0 Yard Fire Hydrant Flow Check (every 366 days) e STP-F-690-0 Sprinkler System Inspection (every 549 days)

• STP-F-691-0 Fire Suppression System Flow Test (every 730 days)

• STP-F-692-0 Hose Station Operability Test (every 1095 days)

• STP-F-693-0 Fire Suppression System Valve Cycling Test (every 366 days)

• STP-F-694-0 Inspection & Hydrostatic Test of Fire Hoses Outside Containment (every 1095 days)

e. STP-F 695-0 Inspection & Rerack of Fire Hoses Outside Cont. (every 549 days) e STP-F-696-0 Fire Pump Flow Test (every 549 days) e STP-F-697-0 Fire Suppression System Functional Test (every 549 days) e STP-F-690-1 Hose Station Inspection (Shutdown)(every 31 days)

• - STP-F-690-2 Hose Station Inspection (Shutdown)(every 31 days) e STP-F-693 1 Removal & Replacement of Fire Hoses in Containment (every 730 days) e STP-F-693-2 Removal & Replacement of Fire Hoses in Containment (every 730 days)

• STP-M-699-1 Switchgear Rooms Halon System Functional Test (every 184 days) e STP-M-699-2 Switchgear Rooms Halon System Functional Test (every 184 days)

This extensive set of periodic performance and condition monitoring activities ensures the system can perform the passive intended PB function. Performance of these activities will provide opportunities for degradation to be detected before a loss of intended function can occur. Thus, aging of all NSR components in scope for this system is fully managed by the FP Program.

I

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ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION i 1

\

5.10.3.4 CC System [ Reference 2, Appendix A, System 015] j 1

Component cooling is a closed system consisting of three motor-driven circulating pumps, two heat exchangers, a head tank, associated valves, piping, instrumentation, and controls for each unit. During {

normal plant operation, one of the pumps and one of the heat exchangers are required for cooling service.

l Items cooled by the system include:

. Letdown heat exchanger; e Shutdown cooling heat exchangers;.

• {

Miscellaneous waste processing heat exchanger; j e Waste gas compressor aftercoolers and jacket coolers; e Control element drive mechanism coolers; e )

Reactor coolant pump mechanical seals and lube oil coolers;

. Low pressure safety injection seals and coolers; l'

• High pressure safety injection seals and coolers; e Containment penetration cooling; e Reactor support cooling; e Steam generator lateral support cooling; e Coolant waste evaporators; e Reactor Coolant and Miscellaneous Waste Sampling System; i e Degasifier vacuum pump cooler; e Post-Accident Sample System; and J e Reactor coolant drain tank heat exchanger.

5.10.3.4.1 Operating Experience t

Representative historical operating experience pertinent to aging is included in the AMR discussion for the CC System in Section 5.3 of the BGE LRA.

l 5.10.3.4.2 Scoping Summary The safe shutdown FP functions of the CC System are:

• Provide a heat sink for essential shutdown cooling loads to ensure safe shutdown in the event of a severe fire.

l

• Provide a heat sink for essential shutdown cooling loads of the alternate unit in the event it experiences a severe fire that debilitates its own CC System. l l In order to provide cooling to the shutdown cooling heat exchanger, one of three pumps and either of the l two CC heat exchangers with attendant flow path must be operable. Should the CC System in the l affected unit be disabled by a fire, cooling can be supplied from the unaffected unit through existing j piping. This requires backflow through the affected unit's reactor coolant waste evaporator supply and return lines. Also, if normal CC head tank makeup flow paths become unavailable due to a fire, makeup can be supplied from the fire main via a hose connection to the Condensate System.

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.. .. I ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION i The portion of the system in scope for the FP AMR includes the NSR components in the head tank make up flow paths and the flow paths to and from the reactor coolant waste evaporator. The following )

passive FP intended function (not addressed in other evaluations) applies: (References 57]

e Maintain the PB of the system liquid.

5.10.3.4.3 Aging Management Demonstration The CC System is in continuous operation during power production. All of the normally operating heat loads are critical to the production of power. The greatest heat loads on the system are the shutdown cooling heat exchangers when the RCS is cooled down to cold shutdown. The operation of the system during a fire is the same as normal or shutdowe cooling modes of operations with respect to providing cooling to the essential heat loads. The paramaers of the NSR portion of the system in scope while performing the required FP functions are no different than the normal operating parameters. The use of the system during a fire will not challenge the pressure-retention ability of the system more than normal.

Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of all NSR components in scope for this system is fully managed by performance and condition monitoring activities during normal operation.

5.10.3.5 Compressed Air System [ Reference 2, Appendix A, System 019]

The' Compressed Air System consists of the instrument air and plant air subsystems with a SR backup supply of air from the saltwater air compressors. The instrument air subsystem is designed to provide a reliable supply of oil-free dry air for the pneumatic instruments and controls and pneumatically-operated containment isolation valves. The plant air subsystem is designed to meet necessary service air requirements for plant maintenance and operation.

The instrument air subsystem incorporates two full-capacity, non-lubricated compressors, each having a separate inlet filter aftercooler and moisture separator. The instrument air compressors discharge to a single header that is connected to two air receivers. Both air receivers discharge to a common outlet header that supplies instrument air to the dryer and filter assemblies. The header then divides into branch lines supplying various plant areas. An emergency back-up tie from the plant air header automatically supplies air to the instrument air subsystem if the pressure to the instrument filter and I dryer assembly falls below a pre-determined setpoint. Local controls prevent plant air use when this occurs. For the transition from normal to emergency service, strategically-located air storage tanks provide an approximate 20-minute supply.

The plant air subsystem incorporates one full-capacity plant air compressor with an inlet filter, i aftercooler, and moisture separator that discharges to the plant air receiver. The receiver outlet header is connected to the prefilter assembly, which is followed by an outlet header branching into two separate air ]

headers, one to the instrument air dryers and filter assembly, and the other to various plant areas. A i system cross-tie between Units I and 2 has been provided for the plant air headers.

The Compressed Air System operates continuously during all plant operating modes. Normally, only i one of the two instrument air compressors are sufficient for maintaining adequate pressure on the l instrument air header of each unit. [ Reference 51, Section 9.10.4] Instrumentation and controls are provided to automatically maintain system operating pressure by initiating actions at predetermined pressure setpoints. The automatic actions include stading the standby instrument air compressor, cross-connecting the plant air subsystem to the instrument air header, isolation of plant air header loads, and Application for License Renewal 5.10-21 Calvert Cliffs Nuclear Power Plant

1 J

t ATTACHMENT LM APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION l.

L isolation of containment loads. [ References 58 through 63] Because of its importance to plant j-

' operations, the compressor load time is routinely tracked in order to discover any increase in system leakage so appropriate corrective actions can be taken. The design of the system and installed equipment redundancy assure a reliable source of compressed air to loads being supplied by the instrument air ,

l header. [ Reference 51, Section 9.10.5]

l The power supply for the instrument air compressors is the normal distribution system backed up by the emergency diesel generator. Additional emergency air compressors, known as the saltwater air I

- compressors, provide redundant air supply to most SR components if the normal air compressors are lost.

The saltwater air compressors are seismically-qualified, air-cooled, and oil-free. The instrument air portion of the Compressed Air System is primarily used for valve actuation and is not used in any reactor indication, control, or protective circuitry. These valve actuators are designed to fail in the safe position

!' after loss of the instrument air supply.

5.10.3.5.1 Operating Experiemee

)

Representative historical operating experience pertinent to aging is included in the AMR discussion for the Compressed Air System in Section 5.4 of the BGE LRA. f 4

5.10.3.5.2 Scoping Summary The safe shutdown FP functions of the Compressed Air System are: I e

Provide compressed air to the instrument air header from an instrument air compressor; Provide compressed air to selected SR equipment from a saltwater air compressor; Provide compressed air to the instrument air header from the plant air compressor via the back-up tie from the plant air header; or Provide compressed air to the instrument air header from the unaffected unit's plant air compressor via the back-up tie from the plant air header and the Unit 1/2 plant air cross-connect.

The Compressed Air System provides control air for essential loads to suppon safe shutdown. The l components include the instrument air, plant air, and saltwater air compressors, along with the associated system valves, piping and controls. This function also includes manual isolation of non-essential air loads. Only one air compressor is required to achieve safe shutdown. The portion of'the system in scope for the FP AMR includes all NSR components of the system. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 64]

• Maintain the PB of the system (liquid and/or gas).

5.10.3.5.3 Agle.g Management Demonstration Pneumatically-operated equipment can be found in every system that is crucial for power production.

Thus, the system is relied on daily. Normally, the plant air compressor and one instrument air compressor will cycle to maintain the desired pressure. The other instrument air compressor and the saltwater air compressors are on standby. [ Reference 51, Section 9.10.4] The demands placed on the Compressed Air System during a fire are the same as, or less than, the normal operating requirements.

Only a single air compressor is required to supply necessary loads following a fire. The parameters of the NSR ponion of the system while performing the required FP functions are no different than the normal operating parameters. The use of the system during a fire will not challenge the pressure-Application for License Renewal 5.10-22 Calved Cliffs Nuclear Power Plant

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ATTACHMENT $

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION retention ability of the system more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of all NSR components in scope for this system is fully managed by performance and condition monitoring activities during normal operation.

5.10.3.6 Diesel Feel Oil System [ Reference 2, Appendix A, System 023]

The Diesel Fuel Oil System supplies fuel to the emergency diesel generators, auxiliary boilers, and the diesel-driven fire pumps. Major components of the system are: two fuel oil storage tanks, a fuel oil unloading pump, an auxiliary boiler supply header, and two diesel generator supply headers.

5.10.3.6.1 Operating Experience Representative historical operating experience pertinent to aging is included in the AMR discussion for the Diesel Fuel Oil System in Section 5.7 of the BGE LRA.

5.10.3.6.2 Scoping Summary The Diesel Fuel Oil System provides the following FP support functions in the event of a fire:

• Provide diesel fuel to the emergency diesel generators; and

• Provide diesel fuel to the diesel-driven fire pump.

Most of the system is SR. The only NSR portion of the system in scope for the FP AMR includes the piping and components related to the diesel-driven fire pump. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 65]

• Maintain the PB of the system liquid.

5.10.3.6.3 Aging Management Demonstration Fire Protection Program activities are credited with maintaining / verifying the ability of the Diesel Fuel Oil System to perform active and passive FP intended functions. Aging effects on the NSR pressure-retaining components are adequately managed by the following specific performance and condition monitoring activities required by the plant's FP Program:

• STP-F-77 Staggered Test of Diesel Fire Pump; and

• STP-F-6% Fire Pump Flow Test.

The diesel-driven fire pump is periodically tested to verify operability / availability through valve lineups, flow and discharge pressure testing, sequential starting capabilities, and controller functions. The pump

( is under observation during pcrformance of the above tests, and degradation of the fuel oil supply lines would be immediately evident. Additionally, the day tank is refilled as required following each test to maintain a minimum quantity of fuel oil in the tank. [ References 66 and 67] Thus, the integrity of the fuel oil supply piping to the tank is verified each time it is refilled. These periodic performance monitoring activities ensure the system can perform the passive FP intended function. Performance of these activities will provide opportunities for degradation to be detected before a loss of intended function can occur. Thus, aging of all NSR components in scope for this system is fully managed by the FP Program.

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i NITACHMENT m APPENDIX A - TECHNICAL INFORMATION t

5.10 - FIRE PROTECTION l 5.10.3.7 Plant Heating System [ Reference 2, Appendix A, System 029]

The Plant IIcating System consists of two main hot water pumps, two main hot water generators, a main circulating loop, an air removal subsystem, and various branch loops and booster pumps. It is a closed system with provision for automatic makeup from the pretreated water systems. The system is set up for balanced flow conditions, thus maintaining steady flow conditions even if unit heaters or branch loops are isolated.

5.103.7.1 Operating Experience In 1994, BGE discovered leakage of some plant heating piping due to corrosion. There was some corrosion adjacent to the penetration through the concrete wall of the pipe tunnel, but the worst corroded area was in the range of 5-10 feet from the wall. The leak was located approximately 5 feet from the wall. There was evidence of past excavation in the area where digging equipment struck the outside of the pipes in the same area. Furthermore, cathodic protection levels were noted to be weak in this area.

The piping has been replaced and wrapped, and new anodes were installed for the cathodic protection system.

5.103.7.2 Scoping Summary The Plant Heating System provides heating (freeze protection) to the PWSTs. The PWSTs are the source of water for the FP System. These tanks are provided with a recirculating-type heating system to maintain a minimum temperature of 45 F as protection against freezing. The portion of the system in scope for the FP AMR includes the components in the main process flow paths shown as normally open on the system drawings. The containment isolation intended function of the system is addressed in Containment Isolation, Section 5.5 of the BGE LRA. The following passive FP intended function (not addressed in other evaluations) applies:

• Maintain the PB of the system (liquid and/or gas).

5.103.73 Aging Management Demonstration The PWST heat exchangers are constructed of a tube bundle mounted in an enclosed shell. Hot water from the Plant Heating System circulates through the tubes during cold weather to keep the water in the PWST from freezing. The shell is physically located within the tank from which the PWST circulating pump takes suction. The circulating pump discharges near the top of the tank and ensures adequate mixing of the contents. All components of the heat exchanger not visible from the outside of the tank are constructed of monel, which is highly resistant to corrosion in this environment. Thus, corrosion of the heat exchanger in a location not normally visible is extremely unlikely. The carbon steel components are plainly visible from the outside of the tank, and any corrosion would be easily identified. [ References 68 and 69]

During periods of cold weather, the Plant Heating System i2 in continual use and is crucial to heating plant areas and selected plant equipment. The demands placed on the system during a fire are the same as the normal operating requirements. The parameters of the system while performing the required FP functions are no different than the normal operating parameters. The use of the system during a fire will not challenge the pressure-retention ability of the system more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging

{ of all NSR components in scope for this system is fully managed by performance and condition monitoring activities during normal operation.

Application for License Renewal 5.10-24 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.3.8 AFW System [ Reference 2, Appendix A, System 036]

The AFW System is designed to provide feedwater to the steam generators for the removal of sensible and decay heat, and to cool the primary system to 300 F in case the main condensate pumps or the main i feed pumps are inoperable. The turbine-driven AFW trains may also be used for normal system cooldown to 300 F. The motor-driven portion of the system is designated for emergency use only (i.e., not for use during normal plant startup or shutdown - except testing is allowed).

Three AFW pumps are installed in each unit, consisting of one motor-driven and two non-condensing steam turbine-driven pumps. For a shutdown, only one pump is required to be operating, the others are in standby. The steam generator's AFW System is initiated by remote manual control or on low level in either steam generator. Upon automatic initiation of AFW, one motor-driven and one turbine-driven pump automatically start. These pumps take suction from a 350.000 gallon CST that is protected against tornadoes and tornado-generated missiles.

The turbine driver is supplied with steam from the steam generator as long as the pressure is above 50 psig. Each turbine has a manually-set governor for controlling turbine speed. Once set for a certain speed, the governor is designed to maintain approximately constant speed with a minimum of 50 psig steam pressure. The steam supply can also be provided from the Auxiliary Boiler Steam System. In addition, in an emergency, the steam-driven train can operate independent of offsite power and the diesels for up to two hours. The AFW air accumulators provide a sufficient control air source until operators can manually regulate the system.

5.10.3.8.1 Operating Experience Representative historical operating experience pertinent to aging is included in the AMR discussion for the AFW System in Section 5.1 of the BGE LRA.

5.10.3.8.2 Scoping Summary The safe shutdown FP functions for the AFW System are as follows:

e Monitor essential AFW parameters to ensure safe shutdown in the event of a fire (CST level, steam generator level, steam generator pressure, and AFW pump discharge pressure);

• Provide control of the AFW System from the Control Room or the auxiliary shutdown panel to ensure safe shutdown in the event of a fire; e Provide heat removal to support hot standby and cold shutdown functions from either turbine-driven train of the affected unit;

• Provide heat removal to support hot standby and cold shutdown functions from the motor-driven train of either unit; and

. Provide heat removal to support hot standby and cold shutdown functions at low pressure situations via the motor-driven train using water from the diesel-driven fire pump.

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! 1 ATTACHMENT (5)

Ii APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Alternate sources of water to support the AFW heat removal functions include CSTs 11 and 12, which can be aligned to the AFW System through manipulation of manual valves as required. Another source of water is the FP System. Specifically, when steam generator pressure has been reduced to less than L 100 psig, the diesel-driven fire pump can be used to provide steam generator inventory using FP System l water sources. The procedure requires the motor-driven AFW pump on the affected unit to be isolated l

and drained. An AFW spool piece is installed at the discharge of the pump and connected to two fire

. hoses supplied from hose stations that are aligned to the diesel-driven fire pump. The NSR portion of the j system in scope for the FP AMR includes: the AFW spool piece for the fire hose connections, AFW isolation valves from CSTs 11 and 21 and the piping between the isolation valves and the CSTs. The CSTs are included within the scope of the Demineralized Water and Condensate Storage System discussed below in Section 5.10.3.9. The following passive FP intended function (not addressed in other 1 L evaluations) applies: [ Reference 70]

]1 e Maintain the PB of the system (liquid and/or gas).

{

j 5.103.8.3 Aging Management Demonstration Per the FP Program, the AFW spool piece and associated hardware is prestaged equipment that is inventoried and inspected each quarter. This activity ensures that the spool piece can perform the l passive FP intended function, and it will provide opportunities for degradation to be detected before a ,

loss ofintended function can occur. Thus, aging of the spool piece is fully managed by the FP Program.

The CSTs are used to store makeup water for the Condensate System during normal operation. The NSR . j section of piping and valves in scope are open to the CSTs and pressurized by the height of water in the tanks. The parameters of this portion of the system while performing the required FP functions are no different than the normal operating parameters. The use of the system during a fire will not challenge the pressure-retention ability of the system more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of this portion of the system is fully managed by performance and condition monitoring activities during normal operation.

5.103.9 '- Demineralized Water & Condensate Storage System [ Reference 2, Appendix A, System 037]

The Demineralized Water and Condensate Storage System stores demineralized water from the Makeup Demineralizer System for normal plant operations and emergency conditions. The system consists of a demineralized water storage tank, two demineralized water transfer pumps, two CSTs, and the associated valves, piping, and controls. The system interfaces with the:

• Auxiliary Boilers;

• AFW System; e Chemical and Volume Control System (CVCS); >

e CC System; e Condensate Demineralizer; e Condenser Air Removal System; e Miscellaneous Waste System;

• Post Accident Sampling System; e Reactor Coolant and Waste Processing Sampling System; e RCS; Application for License Renewal 5.10-26 Calvert Cliffs Nuclear Power Plant

ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Reactor Coolant Waste Processing System;

• SRW System;

• Stator Winding Cooling System;

• Steam Generator Blowdown System;

• Various Lab Stations and Faucets; and

• Waterbox Priming System.

5.103.9.1 Operating Experience Leaks have been discovered on penetrations on CSTs 11 and 21 due to galvanic corrosion. The penetrations have a carbon steel nipple that is welded to a stainless steel half coupling. The lines are insulated, and over time, were brought into electrical contact by wet insulation. No other damage was observed on the pipe or tank walls. The penetrations will be replaced and protectively coated and/or wrapped before being reinsulated.

5.103.9.2 Scoping Summary The safe shutdown FP function of the Demineralized Water and Condensate Storage System is to provide a backup source of water to the AFW System from CSTs 11 and 21. The AFW System is normally aligned to CST 12. At a predetermined low level in the tank, an alternate CST (11 or 21) is placed in service. The CSTs are normally provided makeup water from the demineralized water storage tank. As .

I a last resort, the CSTs can be filled with water from the PWSTs. The line up is accomplished by connecting a fire hose between the fire pump house test manifold and emergency hose connections on the tanks. The portion of the system in scope for the FP AMR is limited to CSTs 11 and 21, associated level instruments, emergency hose connections, and all pressure-retaining piping and components up to the first isolation valve on all headers to and from the tanks. The containment isolation intended function of the system is addressed in Containment Isolation, Section 5.5 of the BGE LRA. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 71)

• Maintain the PB of the system (liquid and/or gas).

5.103.9.3 Aging Management Demonstration The CSTs are used to store makeup water for the Condensate System during normal operation. The NSR components in scope are pressurized by the height of water in the tanks. The parameters of this portion 1 of the system while performing the required FP functions are no different than the normal operating 1 parameters. The use of the system during a fire will not challenge the pressure-retention ability of the I system more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of this portion of the system is fully managed by performance and condition monitoring activities during normal operation.

5.103.10 CVCS [ Reference 2, Appendix A, System 041]

The CVCS is composed of two subsystems: letdown and charging, and makeup. The system performs the following functions:

• Maintain reactor coolant activity at the desired level by removing corrosion and fission products;

• Inject chemicals into the RCS to control coolant chemistry and minimize corrosion;

• Control the reactor coolant volume by compensating for coolant contraction or expansion resulting from changes in reactor coolant temperature and other coolant losses or additions; Application for License Renewal 5.10-27 Calvert Cliffs Nuclear Power Plant l

ATTACHMENT @

l APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION e

Provide means for transferring fluids to the Radioactive Waste Processing System; e

Inject concentrated boric acid into the RCS upon a safety injection actuation signal; I e Control the reactor coolant boric acid concentration; Provide auxiliary pressurizer spray for operator control of RCS pressure during shutdown; Provide a means for functionally testing the check valves that isolate the Safety Injection System from the RCS, and for hydrostatic and leak testing of the RCS; and e

Provide continuous on-line measurement of reactor coolant boron concentration and fission product activity.

5.103.10.1 Operating Experience ,

Representative historical operating experience pertinent to aging is included in the AMR discussion for the CVCS System in Section 5.2 of the BGE LRA.

5.103.10.2 Scoping Summary The safe shutdown FP function for the CVCS is to provide primary makeup in support of RCS pressure and inventory control. If it becomes necessary to conserve RCS inventory during a fire, all letdown, including NSR controlled reactor coolant pump bleedoff flow, is isolated. The pressure-retaining capability of these components must be maintained intact if the RCS inventory control strategy is to be I successful. Thus, the portion of the system in scope for the FP AMR is limited to the NSR piping and valves comprising the flow path from the reactor coolant pump controlled bleedofflines to the letdown l

subsystem. The following passive FP intended function (not addressed in other evaluations) applies:

[ Reference 72]

e Maintain the PB of the system (liquid and/or gas).

5.103.103 Aging Management Demonstration Under certain fire conditions, RCS inventory control requirements may lead to the isolation of the controlled bleedoff flow to the volume control tank. Depending on the RCS pressure and the condition of the reactor coolant pump seals, the pressure in the isolated lines could rise high enough to cause a relief valve to lift for short periods of time. The end result is that the temperature and pressure in this portion of the system is likely to be slightly higher than during normal operation.

l The AMR for the SR PB components of the system includes an evaluation of the letdown line from the l

RCS, the charging line into the RCS, and associated components in these flow paths. The materials of construction are predominately stainless steel with alloy and carbon steel fasteners at mechanicaljoints.

The chemistry in this part of the system is the same as RCS chemistry, including hydrogen overpressure.

The NSR piping and components from the controlled bleedofflines are constructed of the same materials and exposed to the same environmental conditions as these SR portions of the system. Thus, the same conclusions apply to the NSR components in scope for FP. The only plausible ARDM is general corrosion of the alloy and carbon steel fasteners due to boric acid leakage. Aging of those NSR subcomponents for the period of extended operation will be managed by the Boric Acid Corrosion Inspection Program.

Application for License Renewal 5.10-28 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION 5

5.10 - FIRE PROTECTION 5.10.3.11 Condensate System [ Reference 2, Appendix A, System 044]

The exhaust steam from the !cv pressure turbines is discharged into the main condenser shells where the latent heat of vaporization is removed and condensate is formed. Condensate from the hotwells is pumped by two electric, motor-driven condensate pumps through the gland steam condenser, the Condensate Demineralizer and Precoat Filtering System, the lowest feedwater heating stage drain coolers, and the two lowest pressure feedwater heating stages (three heaters per stage), to the suction of the three condensate booster pumps. These pumps deliver the condensate to the two turbine-driven feed pumps through two parallel sets of three feedwater heaters.

5.10.3.11.1 Operating Experience No significant age-related degradation has been identified that would affect the pressure-retaining portions of the Condensate System. The water chemistry controls used to minimize steam generator corrosion are generally efTective in controlling corrosion of components in the Condensate System.

5.10.3.11.2 Scoping Summary The safe shutdown function of the Condensate Sya er i < te provide an alternate flow path for makeup water to the SRW and CC head tanks from the fim W Ns is accomplished via a condensate system fire hose connection to a Turbine Building hose status The portion of the Condensate System in scope for the FP AMR includes the NSR components in the makeup flow path to the SRW and CC head tanks from the fire hose connection. It is comprised of the following two parts:

1. The piping tapping off the condensate pump discharge header that is normally pressurized, up to and including the normally closed manual isolation valves of the makeup lines going to the SRW and CC Systems; and

2. The makeup lines downstream of the normally closed manual isolation valves to the SRW and CC Systems, up to the check valves of the head tank demineralized water makeup lines. This includes a section of piping only; the check valves and downstream piping to the head tanks are evaluated with the SRW and CC Systems in Sections 5.10.3.2 and 5.10.3.4, respectively.

The following passive FP intended function (not addressed in other evaluations) applies:

Maintain the pressure-retaining capability of the system (liquid and/or gas).

5.10.3.11.3 Aging Management Demonstration The part of the system identified in Item I above is pressurized by condensate pump discharge pressure during normal operation. The demands placed on this part of the system during a fire are the same as, or less than, the normal operating requirements. Therefore, use of this part of the system during a fire will not challenge its pressure-retention ability more than normal. Thus, for this part of the system, parameters during performance of FP intended functions are bounded by the normal operating parameters, and aging of this part of the system is ful!y managed by performance and condition monitoring activities during normal operation.

The system piping identified in Item 2 above is isolated under normal operating conditions and, therefore, it is not managed by performance and condition monitoring activities during normal operation.

Thus, an AMR was conducted for this device type. The list of potential ARDMs identified is provided in Application for License Renewal 5.10-29 Calvert Cliffs Nuclear Power Plant

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l ATTACHMENT (5) {

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Table 5.10-3. The plausible ARDMs are identified by a check mark (/) in the device type column.

[ Reference 2, System 044, Attachment 5]

TABLE 5.10-3 POTENTIAL AND PLAUSIBLE ARDMs Dwice Type Potestial ARDMs Piping Cavitation Erosion Corrosion Fatigue Crevice Corrosion /

Erosion Corrosion Fatigue Fouling Galvanic Corrosion General Corrosion /

Hydrogen Damage intergranular Attack Microbiologically-Induced Corrosion Particulate Wear Erosion Pitting /

Radiation Damage Rubber Degradation Saltwater Attack Selective Leaching Stress Corrosion Cracking Stress Relaxation Thermal Damage Thermal Embrittlement l Wear The following paragraphs contain a discussion of the AMR for the piping. It includes a discussion on materials and environment, aging effects, methods to manage aging, aging management programs, and aging management demonstration.

Crevice corrosion. general corrosion. and nitting for nining - Materials and Environment The material of the device type is carbon steel with subcomponents consisting of the pipe, fittings, flanges, studs, and nuts. [ Reference 2, System 044, Attachment 3, Attachment 4]

The internal environment for the device type during power generation is stagnant condensate at a temperature less than 200 F. [ Reference 2, System 044, Attachment 3] The stagnant conditions result because the piping is downstream of normally shut valves up to the check valves on the head tank makeup lines. [ Reference 2, System 044, Section 5.0]

Application for License Renewal 5.10-30 Calvert Cliffs Nuclear Power Plant t

A'ITACHMFNT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Crevice corrosion. general corrosion. and nitting for eining - Aging Mechanism Effects This section describes each ARDM and the effects on the susceptible subcomponents. [ Reference 2, System 044, Attachments 6 and 7]

General corrosion is a thinning (wastage) of a metal by chemical attack (dissolution) at the surface of the metal by an aggressive environment. The consequences of the dam ge are loss of load-carrying cross-sectional area. General corrosion requires an aggressive environment and materials susceptible to that environment.

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 heads, surface deposits, 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. In an oxidizing environment, a crevice can set up a differential aeration cell to concentrate an acid solution within the crevice. Crevice corrosion is closely related to pitting corrosion and can initiate pits in many cases.

Pitting is a form oflocalized 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 steel. These pits are, in many cases, filled with oxide debris, especially for ferrous materials such as carbon 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 direction of gravity.

These forms of corrosion are plausible for the pipe, fittings, and flanges of this device type since they have geometry that allows process fluids (condensate) to stagnate and environmentally-produced impurities to concentrate causing localized corrosion of these components. Since this piping is not exposed to the main flow stream, local fluid chemistry conditions may deviate substantially from bulk fluid chemistry of the Condensate System. The direct effect of these three corrosion mechanisms is a localized loss of material that, if left unmitigated, could result in degradation of the pressure-retaining ability either by through-wall leakage or loss of mechanicaljoint fasteners.

General, crevice, and pitting corrosion mechanisms are also plausible for the nuts and bolts of this device type since corrosion of the mechanicaljoint sealing surfaces and leakage onto the surrounding fasteners can occur. The fasteners are susceptible to these corrosion mechanisms when the internal fluid

. (condensate) escapes onto them. Corrosion of the fasteners could lead to their failure and breech of the mechanicaljoint and loss of pressure-retaining ability.

Crevice corrosion. general corrosion. and nittina for nining - Methods to Manage Aging Mitigation: The effects of corrosion cannot be completely prevented, but they can be mitigated by minimizing the exposure of the carbon steel material to an aggressive chemical environment. Although CCNPP has a Secondary Chemistry Program, the line is not normally filled, so proper water chemistry cannot be relied on.

Discoverv: The effects of corrosion on system components can be disec,vered and monitored through non-destructive examination techniques such as visual inspections. Inspections at susceptible locations can be used to assess the need for additional inspections at less susceptible locations. Based on piping Application for License Renewal 5.10-31 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION and component geometry and fluid flow conditions, areas most likely to experience corrosion can be determined and evaluated.

Crevice corrosion. general corrosion. and nitting for pining - Aging Management Programs Mitigation: The occurrence of corrosion is expected to be limited and not likely to affect the intended function due to the relatively benign environmental operating conditions. No additional mitigation programs are needed at this time. l l

Discoverv: Corrosion can be readily detected through non-destructive examination techniques. As such, an inspection program to identify occurrence oflocalized corrosion is an effective means of determining  ;

if corrective actions are required for managing this aging mechanism.

The subject piping will be included within a new plant program to accomplish the needed inspections for (

general corrosion, crevice corrosion, and pitting. This program is considered an ARDI Program as defined in the CCNPP IPA Methodology presented in Section 2.0 of the BGE LRA.

The elements of the ARDI program will include:

. . Determination of the examination sample size based on plausible aging effects; e Identification of inspection locations in the system / component based on plausible aging effects and consequences ofloss ofcomponent intended function; e Determination of examination techniques (including acceptance criteria) that would be efTective, l considering the aging effects for which the component is examined; l

• Methods for interpretation of examination results; e Methods for resolution of unacceptable examination findings, including consideration of all design loadings 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.

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 PB integrity function under all CLB conditions.

Crevice corrosion. general corrosion. and nitting for eining - Demonstration of Aging Management Based on the factors presented above, the following conclusions can be reached with respect to crevice corrosion, general corrosion, and pitting of the subject piping:

• The piping provides a pressure-retaining boundary and the integrity must be maintained under CLB design conditions.

• Crevice corrosion, general corrosion, and pitting are plausible for this piping, and result in material loss which, ifleft unmanaged, can lead to loss of pressure-retaining boundary integrity.

• Due to the relatively benign environmental operating conditions, significant degradation of this piping is not expected, llowever, to provide assurance that these ARDMs are being managed in this portion of piping, these components will be included in the scope of an ARDI Program.

Application for License Renewal 5.10-32 Calvert Cliffs Nuclear Power Plant

4 1 ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Inspections will be performed, and appropriate corrective action will be taken if significant corrosion is discovered.

Therefore, there is reasonable assurance that the effects of crevice corrosion, general corrosion, and pitting of this piping will be managed in such a way as to maintain the components' PB integrity, consistent with the CLB, during the period of extended operation.

5.10.3.12 Flant Drains System (Reference 2, Appendix A, System 053]

The Plant Drains System is commonly referred to as the plant sump system. It includes the floor and equipment drain piping, sump piping, sump pumps, and instrumentation associated with all areas of the plant. Plant drains include the Turbine Building clean and oily waste sumps, the Service Buildings clean and oily waste sumps, the Auxiliary Building sumps, the intake Structure sumps, the yard area sumps, and the Fuel Oil System drainage. The yard area sumps include the lube oil storage tank area sump, the yard manhole sump, the fire pump house sumps, the acid storage tank area sump, the yard sump, the yard waste oil collection tank, and the yard oil interceptor. Fuel oil system drainage includes the Diesel Generator Room oil drainage subsystem, Fuel Oil Storage Tank l1 area dike, and valve pit drainage subsystem. Containment sumps are treated separately in other sections of the BGE LRA.

5.10.3.12.1 Operating Experience No pertinent operating experience was discovered for this system.

5.10.3.12.2 Scoping Summary Sumps and floor drains collect and remove fire fighting water from areas containing SR equipment where fixed fire suppression systems are installed or where fi e hoses may be used. Drains discharging i to a common header from SR areas containing five or more gallons of combustible liquids were designed with check valves to prevent backflow of combustible liquids. The drains provided with such check valves are from the Charging Pump Rooms, ECCS Pump Rooms, Diesel Generating Rooms, and Auxiliary Feed Pump Rooms. The portion of the system in scope for the FP AMR includes the NSR piping and valves in the floor drain lines from rooms containing SR equipment. The containment isolation intended function of the system is addressed in Containment Isolation, Section 5.5 of the BGE LRA. The following passive FP intended functions (not addressed in other evaluations) apply:

• Provide drainage of fire fighting water in rooms containing SR equipment, and e Maintain the pressure-retaining capability of the system (liquid and/or gas).

5.10.3.12.3 Aging Management Demonstration The drain system piping (and associated components) from rooms containing SR equipment must maintain pressure to ensure drainage of fire fighting water. The demands placed on the system to drain water during a fire are the same as those during normal drain system operation. Use of the drains during a fire will not challenge this pressure-retention ability more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters of the system, and aging of this part of the system, with the exception of the backflow prevention check valve disks, is fully managed by performance and condition monitoring activities during normal operation.

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ATTACHMENT (M APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION l Aging management of the disks of the backflow prevention check valves from the Charging Pump Rooms, ECCS Pump Rooms, Diesel Generator Rooms, and AFW Pump Rooms will be accomplished through the FP Program. Baltimore Gas and Electric Company is currently evaluating the most 1 appropriate method of accomplishing this aging management. The activity selected will ensure the valves can perform their FP intended function, and it will provide opportunities for degradation to be detected before a loss ofintended function can occur. Thus, aging management of the check valves will be fully managed by the FP Program and performance and condition monitoring activities during normal operation.

5.10.3.13 RCS [ Reference 2, Appendix A, System 064]

The RCS is the primary coolant loop, located entirely within the containment, consisting of two heat transfer loops connected in parallel across the reactor pressure vessel. Each loop contains one steam generator, two circulating pumps, connecting piping, and flow and temperature instrumentation. Coolant system pressure is maintained by a pressurizer connected to one of the loop hot legs. During operation, the four pumps circulate wates through the reactor vessel where the water serves as both coolant and moderator for the core. The heated water enters the two steam generators, transferring heat to the secondary (steam) system, and then returns to the pumps to repeat the cycle. A lube oil collection system  ;

for the reactor coolant pump motors is considered a subsystem of the RCS. The lube oil collection tanks j are sized for the entire oil contents of two reactor coolant pumps. There are four tanks to handle the  ;

eight reactor coolant pumps (four reactor coolant pumps per unit).

l 5.10.3.13.1 Operating Experience Representative historical operating experience pertinent to aging is included in the AMR discussion for the RCS in Section 4.1 of the BGE LRA.,

5.10.3.13.2 Scoping Summary The FP function for the RCS is:

• Provide a lube oil collection system for reactor coolant pump motors, sized to accommodate the largest potential oil leak.

The safe shutdown functions for the RCS are:

• Provide monitoring of essential parameters;

• Provide a means for removal of decay heat; e Serve as a fission product barrier; and e Control inventory loss.

Parameters monitored include pressurizer pressure and level indication and hot / cold leg temperature indication. Monitoring is included in the Control Room and at the alternate shutdown panel. All monitoring components in scope are SR PB and have been addressed by the SR PB AMR for the RCS or other commodity evaluations. Also, the reactor coolant pump motor lube oil collection system is SR and is addressed in the SR PB AMR as well.

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ATTACHMENT m l

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION Decay heat removal, fission product boundary, and controlling inventory loss are all related to the PB of the RCS. Decay heat removal and fission product boundary functions are entirely accomplished within the SR PB portion of the system. However, ifit becomes necessary to conserve RCS inventory during a fire, all letdown, including NSR controlled reactor coolant pump bleedoff flow, is isolated. [ Reference 2, Appendix A, System 041, Section 5.0] The pressure-retaining capability of these components must be maintained intact if the RCS inventory control strategy is to be successful. Thus, the portion of the system in scope for the FP AMR is limited to the NSR piping and associated components in the controlled bleedoff lines from the reactor coolant pumps to the CVCS. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 73]

e Maintain the PB of the system (liquid and/or gas).

5.10.3.13.3 Aging Management Demonstration I Under certain fire conditions, RCS inventory control requirements may lead to the isolation of the I controlled bleedoff flow to the volume control tank. Depending on the RCS pressure and the condition of the reactor coolant pump seals, the pressure in the isolated lines could rise high enough to cause a relief valve to lift for short periods of time. The end result is that the temperature and pressure in this portion of the system is likely to be slightly higher than during normal operation.

The AMR for the SR PB components of the CVCS includes an evaluation of the letdown line from the RCS, the charging line into the RCS, and associated components in these flow paths. The materials of construction are predominately stainless steel with alloy and carbon steel fasteners at mechanicaljoints.

The chemistry in this part of the system is the same as RCS chemistry including hydrogen overpressure.

The NSR piping and components of the controlled bleedofflines are constructed of the same materials and exposed to the same environmental conditions as these SR portions of the CVCS. Thus, the same conclusions apply to the NSR components of the RCS in scope for FP. The only plausible ARDM is general corrocion of the alloy and carbon steel fasteners due to boric acid leakage, and aging for the period of extended operation will be managed by the Boric Acid Corrosion Inspection Program.

5.10.3.14 Liquid Waste System [ Reference 2, Appendix A, System 071]

The Liquid Waste System consists of two subsystems - miscellaneous waste processing and reactor coolant waste processing. The miscellaneous waste processing subsystem provides controlled handling and disposal of various liquid wastes from both units. The subsystem consists of.

• Miscellaneous waste receiver tank; e Miscellaneous waste monitor tank; e Miscellaneous waste receiver tank pump; e Miscellaneous waste monitor tank pump; e Miscellaneous waste filters; e Miscellaneous waste ion exchanger; e Miscellaneous waste metering pump; and e Associated strainers, piping, valves, and instrumentation.

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ATTACHMENT (5) t APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION l The miscellaneous waste processing subsystem receives liquid waste from three major sources:

• Auxiliary building gravity drains; e Hot laboratory and soapy drains; and e Containment normal sump and pumped sumps.

Additional sources are:

• SRW System; e CC System; e Blowdown Recovery System; e Refueling water tanks; e Refueling water tank room sump pump; and I e Spent fuel pool.

The reactor coolant waste processing subsystem provides controlled handling and disposal of radioactive liquid wastes from both reactor units. The subsystem provides temporary storage for reactor coolant wastes and processes wastes prior to disposal. The reactor coolant waste processing subsystem consists  ;

of: i e Two reacter coolant drain tanks; I e Three cartridge filters;

• Two degasifiers; i e Four reactor coolart waste ion exchangers;

• Two reactor coolant waste receiver tanks;

• Two evaporators;

• Two reactor coolant waste monitoring tanks; e Reactor coolant drain tank pumps;

• Degasifier vacuum pumps; e Reactor coolant waste receiver tank pumps; e Reactor coolant waste receiver tank metering pumps; and

• Associated piping, valves, controls, and instrumentation.

5.10.3.14.1 Operating Experienee No significant age-related degradation has been identified that would affect the pressure-retaining portions of the Plant Drains System. The system is comprised of components constructed of stainless steel thereby minimizing any corrosion related concerns.

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APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.3.14.2 Scoping Summary The Liquid Waste System supports the Plant Drains system by ensuring fire suppression water is drained from rooms containing SR equipment. The portion of the system in scope for the FP AMR is limited to the NSR components in the flow paths from the sump pump discharge check valves serving areas containing SR equipment to the waste processing subsystems. The containment isolation intended function of the system is addressed in Containment Isolation, Section 5.5 of the BGE LRA. The following passive FP intended functions (not addressed in other evaluations) apply:

e Provide drainage of fire suppression water in rooms containing SR equipment; and Maintain the pressure-retaining capability of the system (liquid and/or gas).

5.10.3.14.3 Aging Management Demonstration The Liquid Waste System receives and processes the drains from the rooms containing SR equipment through the interface with the Plant Drains System. Piping (and associated valves) located within rooms containing SR equipment must maintain pressure-retaining capability to ensure drainage from the rooms.

The demands placed on the system to drain water during a fire are the same as those during normal system operation. Use of the system during a fire will not challenge the pressure-retention ability more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters of the system, and aging of this part of the system is fully managed by performance and condition monitoring activities during normal operation.

5.10.3.15 Nitrogen and Hydrogen System The Nitrogen and Hydrogen System consists of two independent subsystems supplying gasses for normal operation. Portions of the nitrogen gas subsystem provide containment isolation so they are SR and are addressed :n Section 5.5 of the BGE LRA. A portion of the hydrogen gas subsystem is important for FP and is inch ded herein. The hydrogen gas subsystem is common to both units and provides hydrogen gas to the folloWag equipment:

. Two main generators (cooling medium);

• Two volume coritrol tanks (RCS chemistry control); and e Radiation-chemistry chemical cabinet (gas standard and burn gas).

The hydrogen gas subsystem consists of hydrogen gas bottles, a truck fill connection, pressure control unit, distribution header, and the associated piping, valves, and controls. [ Reference 1]

5.10.3.15.1 Operating Experience No significant age-related degradation has been identified that would affect the pressure-retaining portions of the hydrogen supply lines for the Nitrogen and Hydrogen System. Cracks were discovered where a hydrogen supply line enters the main generator. However, this failure was due to high vibration on the main generator and is not considered an age-related concern. To prevent this from recurring, a new support was added for the hydrogen supply line to the main generator. Periodic pressure testing of the main generators has identified only minor system leakage due to valve packing and other mechanical joint leakage.

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ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.3.15.2 Scoping Summary The FP function of the hydrogen gas subsystem is to isolate hydrogen flow to the Auxiliary Building in the event of a downstream piping rupture. Excess flow check valves are installed upstream of the Auxiliary Building supply lines to each unit (one per unit). The two valves were designed to close if flow exceeds 75 cfm or if the differential pressure across the valves exceeds 75 psi. The portion of the system in scope for the FP AMR is limited to the NSR excess flow check valves. The following passive FP intended functions (not addressed in other evaluations) apply:

Maintain the pressure-retaining capability of the system gas.

5.10.3.15.3 Aging Management Demonstration Aging management of the excess flow check valves will be accomplished through the FP Program.

Baltimore Gas and Electric Company is currently evaluating the most appropriate method of accomplishing this aging management The activity selected will ensure the valves can perform their FP intended function, and it will provide opportunities for degradation to be detected before a loss of intended function can occur. Thus, aging management of the check valves will be fully managed by the FP Program.

5.10.3.16 Main Steam System [ Reference 2, Appendix A, System 083]

The Main Steam System is designed to transfer steam from the steam generators to the turbine throttle stop valves, the reheaters, and the turbine-driven pumps. The Main Steam System also controls steam generator pressure by means of steam bypass, dump, or safety valves (high pressure) and main steam isolation valves (MSIVs)(low pressure).

Major components of the system are: flow restrictors, safety valves, MSIVs, turbine throttle stop valves, steam dump valves, turbine bypass valves, MSIV-bypass valves, AFW pump turbine steam supply isolation valves and bypass valves, moisture separator reheater isolation valves, moisture separator reheaters, and associated piping and controls.

Overpressure protection for the shell side of the steam generators and the main steam line piping up to the inlet of the turbine stop valve is provided by 16 spring-loaded American Society of Mechanical Engineers Code main steam safety valves that discharge to the atmosphere. Eight of these safety valves are mounted on each of the main steam lines upstream of the MSIVs and outside of the containment.

The Steam Dump and Bypass System is used to rapidly remove RCS stored energy and to limit secondary steam pressure following a turbine-reactor trip. The Atmospheric Steam Dump System consists cf two automatically-actuated atmospheric dump valves (ADVs) that exhaust to the atmosphere.

The Turbine Bypass System consists of four turbine bypass valves that exhaust to the main condenser.

The power-operated steam dump valves and steam bypass valves minimize the need for opening of the main steam safety valves following turbine and reactor trips from full power. -

The Main Steam System also provides a means of heat removal during hot standby and during a plant cooldown. The ADVs are capable of removing reactor decay heat when the condenser is not available.

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ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.3.16.1 Operating Experience Representative historical operating experience peninent to aging is included in the AMR discussion for the Main Steam System in Section 5.12 of the BGE LRA.

5.10.3.16.2 Scoping Summary The safe shutdown functions for the Main Steam System in the event of a fire are:

Provide controlled RCS heat removal to maintain the affected unit at hot standby and to cool the i affected unit down to cold shutdown conditions if required.

Provide monitoring of steam generator pressure in support of RCS heat removal function.

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I Heat removal is accomplished using the ADVs or main steam safety valves. Due to the limited capacity of the ADVs at low steam generator pressures, use of both steam generators is required to reach cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the room where the fire is located meets 10 CFR Part 50, Appendix R, Paragraph III.G.2 requirements, then the affected unit does not need to be in cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this case, only one steam generator is necessary. If the ADVs are disabled by a fire, they are manually isolated, and RCS heat is removed through the main steam safety valves until manual operation of the ADVs can be established. If main steam isolation is required to control the RCS heat removal rate, it can be accomplished by closure of the MSIVs and associated bypass valves and isolating the Steam Generator Blowdown System. If the MSIVs are inoperable due to fire, an alternate method of isolating the main steam header is to manually isolate the turbine bypass valves, gland seal steam, and second stage steam to the moisture separator reheaters. Steam generator pressure is monitored using SR pressure instrument loops.

The portion of the system in scope for the FP AMR is limited to the NSR pressure-retaining piping and components located downstream of the MSIVs up to the next isolation valves, i.e., turbine bypass valves, moisture separator reheater isolation valves, main turbine stop valves, main feed pump turbine stop valves, and steam seal isolation valve. All other portions of the system used for FP intended functions

! are SR and included in the AMR in Section 5.12 of the BGE LRA. The following passive FP intended function (not addressed in other evaluations) applies: [ Reference 74]

• Maintain the PB of the system (liquid and/or gas).

l 5.10.3.16.3 Aging Management Demonstration This portion of the Main Steam System is subjected to normal system conditions during operation since it is part of the main steam flow path to the high pressure turbines. The demands placed on the system during a fire are the same as, or less than, those during normal system operation. Use of the system during a fire will not challenge the pressure-retention ability more than normal. Thus, system parameters during performance of FP intended functions are bounded by the normal operating parameters of the system, and aging of this part of the system is fully managed by performance and condition monitoring activities during normal operation.

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ATTACHMENT 8)

APPENDIX A - TECIINICAL INFORMATION 5.10 - FIRE PROTECTION 5.10.4 Conclusion i

Table 5.10-4 lists the programs credited in this section of the BGE LRA. *t*, a programs will be administratively controlled by a formal review and approval process. As has been demonstrated in the

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above sections, these programs will manage the aging mechanisms and their effects such that the j intended function of the components of these systems will be maintained, consistent with the CLB, J during the period of extended operation.

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.

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ATTACHMENT (5)

APPENDIX A - TECIINICAL INFORMATION 5.10 - FIRE PROTECTION TABLE 5.10-4 LIST OF AGING MANAGEMENT PROGRAMS FOR FIRE PROTECTION Program - Credited As i Existing FP Program, Nuclear Program Directive Managing the effects of aging for the following  ;

SA-1 systems; FP, Diesel Fuel Oil, Auxiliary l Feedwater (partial), Plant Drains (partial), and Nitrogen and Hydrogen Gas.  ;

Existing Structure and System Walkdown Program, Managing the effects of aging for the following CCNPP Administrative Procedure systems; Well and Pretreated Water, SRW, CC, I MN-1-319 Compressed Air, Plant Heating, Auxiliary Feedwater (partial), Demineralized Water and Condensate Storage, Condensate (partial), Plant Drains (partial), Liquid Waste, and Main Steam.

Existing Conduct of Operations, CCNPP Managing the effects of aging for the following Administrative Procedure NO-1-100 systems; Well and Pretreated Water, SRW, CC, Compressed Air, Plant Heating, Auxiliary l Feedwater (partial), Demineralized Water and Condensate Storage, Condensate (partial), Plant Drains (partial), Liquid Waste, and Main Steam.

Existing Doric Acid Corrosion Inspection Program Mitigation, detection, and management of the (Refer to the discussions in Sections 5.2, effects of general corrosion in a portion of the CVCS, and 4.1, RCS, of the BGE LRA for CVCS, i.e., the fasteners in the NSR piping and detailed information regarding this associated components in the controlled program) bleedofflines for the reactor coolant pumps.

Mitigation, detection, and management of the effects of general corrosion in a portion of the RCS, i.e.,the fasteners in the NSR piping and associated components in the controlled bleedofflines for the reactor coolant pumps.

New ARDI Program Detection and management of the effects of crevice corrosion, general corrosion, and pitting on a portion of the piping in the Condensate System (as identified in Section 5.10.3.11).

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ATTACHMENT (5)

APPENDIX A - TECHNICAL INFORMATION l 5.10 - FIRE PROTECTION References

1. )

System and Structure ITLR Screening Results, Life Cycle Management Renewal Program, )

BGE, Revision 4, April 4,1995 I

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2. CCNPP Fire Protection Aging Manapment Review, Revision 1, January 30,1997

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

4. CCNPP Component Level Scoping Results for the Primary Containment System (059),

Revision 0, February 23,1993

5. CCNPP Pre-Evaluation Results for the Containment System (059), Revision 1, May 9,1996

6. CCNPP Component Level Scoping Results for the Primary Containment Structure (SYS 059),

Revision 1, March 25,1996

7. CCNPP Component Level Scoping Results for the Auxiliary Building, Revision 2, February 14,1997

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

9. CCNPP Component Level Scoping Results for the Salt Water Cooling System, Revision 3, July 15,1996

10. CCNPP Component Pre-evaluation for the Salt Water System (012), Revision 4, December 30,1996

11. CCNPP Component Level Scoping Results for the Emergency Diesel Generators, Revision 1, January 5,1993

12. CCNPP Pre-evaluation Results for the Emergency Diesel Generator System (024), Revision 0, January 30,1995

13. CCNPP Component Level Scoping Results for the Feedwater System (045), Revision 2, December 16,1996

14. CCNPP Pre-evaluation Results for the Main Feedwater System (045), Revision 1, l March 11,1996

15. CCNPP Component Level Scoping Results for System 052 Safety Injection System, Revision 2, April 8,1996

16. CCNPP Pre-evaluation Results for the Safety injection System (052), Revision 1, j December 1,1995

17. CCNPP Component Level Scoping Results for System 061 - Containment Spray System, Revision 1, April 17,1996

18. CCNPP Component Pre-evaluation Results for the Containment Spray System (061),

Revision 2, March 6,1997

19. CCNPP Component Level Scoping Results for the Control Room HVAC System, Revision 1, July 17,1996 Application for License Renewal 5.10-42 Calvert Cliffs Nuclear Power Plant L

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ATTACHMENT &

APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION

20. CCNPP Component Pre-evaluation for the Control Room HVAC System (030), Revision 1, April 2,1997

21. CCNPP Component Level Scoping Results for the Auxiliary Building and Radwaste H&V System, Revision 2, July 17,1996

22. CCNPP Pre-evaluation Results for the Auxiliary Building and Radwaste H&V System 032, Revision 0, June 2,1995

23. CCNPP Component Level Scoping Results for the Primary Containment H&V System, Revision I, August 15,1996

24. CCNPP Pre-evaluation Results for the Primary Containment Heating & Ventilating System (060), Revision 0, May 1,1995

25. CCNPP Component Level Scoping Results for the Electrical 125 Volt DC Distribution System, Revision 3, August 8,1995

26. CCNPP Pre-evaluation Results for the 125 VDC System (002), Revision 0, September 7,1994

27. CCNPP Component Level Scoping Results for the Electrical 4 kV Transformers and Buses System, Revision 1, August 5,1995

28. CCNPP Pre-evaluation Results for the Electrical 4kV Transformers and Buses System (004),

Revision 1, October 13,1994

29. CCNPP Component Level Scoping Results for the Electrical 480V Transformers and Buses System, Revision 1, August 5,1995

30. CCNPP Pre-evaluation Results for the Electrical 480V Transformers and Buses System (005),

Revision 1, October 15,1994

31. CCNPP Component Level Scoping Results for the Electrical 480V Motor Control Centers System, Revision 1, August 5,1995

32. CCNPP Pre-evaluation Results for the Electrical 480V Motor Control Centers System (006),

Revision 0, September 12,1994

33. CCNPP Component Level Scoping Results for the Instrument AC System, Revision 1, June 30,1995

34. CCNPP Pre-evaluation Results for the Instrument AC System (017), Revision 0, September 12.1994

35. CCNPP Component Level Scoping Results for the Vital Instrument AC System, Revision 1, August 5,1995

36. CCNPP Pre-evaluation Results for the Vital Instrument AC System (018), Revision 0, September 12,1994 f 37. CCNPP Component Level Scoping Results for the Annunciation System, Revision 1, August 4,1995

38. CCNPP Pre-evaluation koults for the Annunciation System (026), Revision 0, October 12,1994

39. CCNPP Component Level Scoping Results for the Control Rod Drive Mechanism and Electrical System, Revision 2, May 22,1995 i

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40. CCNPP Pre-evaluation Results for the Control Rod Drive Mechanism and Electrical System (055), Revision 0, September 28,1994

41. CCNPP Component Level Scoping Results for the Nuclear Instrumentation System, Revision 1, January 14,1997

42. CCNPP Pre-evaluation P.esults for the Nuclea'r Instrumentation System (078), Revision 0, February 2,1995

43. CCNPP Component Level Scoping Results for the . Main Turbine System, Revision 0, j October 11,1994

44. CCNPP Pre-evaluation Results for the Main Turbine System (093), Revision 0, October 14,1994

45. CCNPP Component Level Scoping Resulu for the Lighting and Power Receptacle System, Revision 1, August 5,1995

46. CCNPP Pre-evaluation Results for the Lighting and Power Receptacle System (097),

Revision 0, October 27,1994

47. CCNPP Component Level Scoping Results for the Plant Communications System, Revision 0, October 14,1995

48. CCNPP Pre-evaluation Results for the Plant Communications System (100), Revision 0, October 27,1994

49. Memorandum from Mr. J. P. Moulton (NRC) to Baltimore Gas and Electric Company, dated November 20,1996, " Summary of Meeting with Baltimore Gas and Electric Company (BGE) on BGE License Renewal Activities"

50. CCNPP Nuclear Program Directive SA-1, " Fire Protection Program," Revision 2, January 20,1998

51. CCNPP Updated Final Safety Analysis Report, Revision 21

52. CCNPP Technical Specifications, Units I and 2

53. CCNPP Administrative Procedure MN-1-319, " Structure and System Walkdowns," Revision 0, September 16,1997

54. CCNPP Administrative Procedure NO-1-100, " Conduct of Operations," Revision 10, January 9,1998

55. CCNPP Administrative Procedure QL-2-100," Issue Reporting and Assessment," Revision 10, March 9,1998

56. CCNPP Component Pre-evaluation for the Service Water System (011), Revision 1, October 7,1996

57. CCNPP Component Pre-evaluation for the Component Cooling Water System (015),

Revision 1, November 4,1996.

58. CCNPP Operations Drawing No. 60712SH0001," Compressed Air System, Instrument Air and Plant Air, Unit 1" Revision 46, December 5,1996

59. CCNPP Operations Drawing No. 62712SH0001," Compressed Air System, Instrument Air and Plant Air, Unit 2" Revision 33, February 10,1998 Application for License Renewas 5.10-44 Calver; Cliffs Nuclear Power Plant

3. < A ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.10 - FIRE PROTECTION

60. CCNPP Electrical Logic Diagram No. 62622SH0001," Instrument Air Compressor, Units I and 2" Revision 7, January 14,1991

61. CCNPP Electrical Logic Diagram No. 62622SH0002, " Plant Air Compressor, Units 1 and 2" Revision 8, December 18,1996

62. CCNPP Electrical Logic Diagram No. 62622SH0006, " Containment Instrument Air Header Control Valve, Units 1 and 2" Revision 4, December 21,1990

63. CCNPP Electrical Logic Diagram No. 62622SH0004, " Instrument / Plant Cross Connect Plant Air Header Isolation, Units 1 and 2" Revision 3, December 21,1990

64. CCNPP Pre-evaluation Results in Compliance with BGE Procedure LCM-10 for Compressed Air System (019), Revision 0, November 30,1995 3 65. CCNPP Pre-evaluation Results for the Diesei Fuel Oil System (023), Revision 1, March 11,1996

66. CCNPP Surveillance Test Procedure STP-F-77-0, " Staggered Test of Diesel Fire Pump,"

Revision 3, December 1,1997

67. CCNPP Surveillance Test Procedure STP-F-696-0, " Fire Pump Flow Test," Revision 3, January 1,1997

68. CCNPP Drawing No. 12352-0001, "Eight Inch Heat Exchanger for Pretread Water Storage Tank," Revision 0, July 10,1970

69. CCNPP Specification No. M-0024, "Pretreated Water Storage tanks Heat Exchangers and Miscellaneous Waste Processing System Heat Exchanger, Units 1 and 2," Revision 3, October 4,1973

70. CCNPP Pre-evaluation Results for the Auxiliary Feedwater System (036), Revision 1, April 3,1996.

71. CCNPP Pre-evaluation Results for the Demineralized Water & Condensate Storage System I (037), Revision 0, October 11,1994

72. CCNPP Pre-evaluation Results for the Chemical and Volume Control System 041, Revision 2, October 17,1996

73. CCNPP Component Pre-evaluation for the Reactor Coolant System (064), Revision 1, December 20,1996.

74. CCNPP Pre-evaluation Results for the Main Steam System (083), Revision 1, July 11,1996 1

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ATTACHMENT (6) i APPENDIX A - TECHNICAL INFORMA* ION 6.2 - ELECTRICAL COMMODITIES 4

Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant March 27,1998 i

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APPENDIX A - TECIINICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES 6.2 Electrical Commodities This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing Electrical Commodities (ECs), which have been 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 LRA.

6.2.1 Scoping 6.2.1.1 Electrical Commodities Scoping Electrical components are associated with most plant systems. The scoping process, performed separately for each system within the scope of license renewal, identified passive electrical structural enclosures / supports (e.g., panels, racks, etc.) from 26 systems that were included in the Electrical Commodities Evaluation (ECE). Since the component materials and environments are common to numerous systems, it was determined that a commodity evaluation approach would be more efficient i

rather than evaluating these electrical commodities (ECs) in each system aging management review {

(AMR). [ Reference 1, Sections 1.1) I 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 keyword searches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.

Commodity Descriotion/Concentual Boundaries Electrical commodities are within the scope of license renewal because they support and protect various plant electrical components that are required to perform the functions described in 10 CFR 54.4(a)(1),

(2), and (3). As discussed in Section 5.0 of the CCNPP IPA Methodology, system components are l assigned to the scope of the ECE during the system pre-evaluation process. As a result of that process,

several types of passive, long-lived electrical components were considered electrical commodities.

l These components typically were either conductive equipment (such as distribution buses) or l panels / cabinets, which support and/or protect safety-related electrical equipment and terminal blocks.

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Cables were excluded from this evaluation and have been addressed in the Cables Commodity l Evaluation in Section 6.1 of the BGE LRA. The ECE is composed of the following structural enclosures i for electrical equipment, which provide %pport and/or protection of the electrical equipment within them: [ Reference 1, Section 1.2.2]

l e Miscellaneous panels; e Motor control center (MCC) cabinets;

• Switchgear/ disconnect cabinets;

• Bus cabinets; e Circuit breaker cabinets; e Local control stations panels; e Battery terminals and charger cabinets; and

. Inverter cabinets.

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES The supporting cabinets and panels of the identified equipment above are evaluated in this report along with terminal blocks and other structural subcomponents of these enclosures. Most of the supporting panels and cabinets covered by this report are identified by the name of the equipment they support because they do not have their own unique equipment identifiers. The panels are included here as a device type because some of the panels do have unique equipment identifiers. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the cabinets or panels that house them and are included in this evaluation. They are phenolic material subject to the effect of electrical stresscrs.

Electrical commodities are assigned to a number of systems in the CCNPP equipment database because they are functionally related to the system components. In all cases, the passive intended function of such electrical commodities equipment is to provide structural suppon to active system components contained in this equipment, and/or to ensure electrical continuity of power, control, or instrumentation signals. The conceptual boundaries for the ECs includes panels and the enclosures / supports (i.e., cabinets, etc.) for MCCs, sv ahgears, buses, disconnect switches / links, local control stations, batteries, chargers, and inverters for the systems shown in Table 6.2-1. The CCNPP system numbers and the applicable BGE LRA Sections for the systems are also shown in the table. [ Reference 1, Section 3.0]

The design basis and associated loading conditions for the CCNPP electrical systems (which include ECs) are described in Updated Final Safety Analysis Report Section 8.1.1. All ECs vital to plant safety l are designated as Class IE so that their integrity is not impaired by a Safe Shutdown Earthquake, high l

winds, or disturbances in the external electrical system. [ Reference 2, Section 8.1.1]

Ooerating Exnerience

! The following historical operating experience is included to provide insight in supporting the aging management demonstrations provided in Section 6.2.2 of this report.

The ECs are usually not subject to extreme conditions or excessive loads; however, some CCNPP ECs are subject to corrosive environments. For example, there have been EC components that have been rusted and corroded due to exposure to saltwater spray. [ Reference 3] Other instances of corrosion have occurred in the Nos. I1 A and B traveling screen control panels. Though these panels are not within the scope of license renewal, they are made of materials and exposed to environments similar to panels within the scope oflicense renewal addressed in this report. The legs of these panels were corroded due to exposure to the atmosphere and were replaced with stainless steel support legs.

The cathodic protection system panels for the intake structure baffle walls experienced corrosion, and have been replaced with new, upgraded panels made of fiber reinforced plastic. Though the cathodic protection system bafile walls are not within the scope license renewal, they are made of materials and exposed to environments similar to panels within the scope oflicense renewal addressed in this report.

Any panel onsite that is either outside or subject to salty air or high humidity could be subject to external corrosion, and corrosion product buildup on contacts or terminations if panel door and penetration seals leak, l

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES l The discovery of these anomalies and the actions taken subsequent to discovery demonstrates that CCNPP inspects and maintains the ECs subject to harsh environments to ensure that these components remain capable of performing their intended function (s) under current licensing basis (CLB) conditions.

TABLE 6.2-1 SYSTEMS CONTAINING ELECTRICAL PANELS WITHIN THE SCOPE OFTHE ELECTRICAL COMMODITIES EVALUATION System f System Name BGEgRA

Number Section

! 002 Electrical 125 VDC Distribution -

004 Electrical 4 kV Transformers and Buses -

005 Electrical 480V Transformers and Buses -

006 Electrical 480V MCCs -

011 Service Water 5.17 012 Saltwater 5.16 013 Fire Protection 5.10 017 Instrument AC (alternating current) -

018 VitalInstrument AC -

019 Compressed Air 5.4 020 Data Acquisition -

024 Emergency Diesel Generators 5.8 j 030 Control Room Heating. Ventilation, and Air Conditioning (HVAC) 5.11.C 032 Auxiliary Building and Radwaste Heating and Ventilation 5.11.A 036 Auxiliary Feedwater 5.1 038 Nuclear Steam Supply System Sampling 5.13 048 Engineered Safety Feature Actuation -

052 Safety injection 5.15 058 Reactor Protective -

060 Containment Heating and Ventilation 5.11.B 062 Control Boards -

073 Hydrogen Recombiners -

074 Nitrogen and Hydrogen Gas -

i 078 Nuclear Instrumentation - l 079 Radiation Monitoring 5.14 l

094 Plant Computer -

097 Lighting and Power Receptacle - l 103 Diesel Generator Building HVAC 5. llc l

Scooed Structures and Comoonents and Their Intended Functions l The conceptual boundaries for the ECs include the panels and enclosures / supports (i.e., cabinets, etc.) for MCCs, circuit breakers, switchgear, buses, local control stations, disconnect switch / links, battery terminals, chargers, and inverters as described above for the systems shown in Table 6.2-1. All of these l panels and enclosures / supports perform passive intended functions and are subject to AMR. Active 1

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES electrical devices are explicitly excluded from AMR based on Q54.2)(aXIXi). Based on the discussion in Section 4.1.1 of the CCNPP IPA Methodology, ECs that perform the following passive intended functions are within the scope of li:ense renewal based on Q54.4(aXI) and Q54.4(aX2): [ Reference 1, Section 1.2.3, 3.1, 3.2]

To maintain electrical continuity and/or provide protection of the electrical system; and e

To provide seismic integrity and/or protection of safety-related components.

Based on the above, the EC components that support passive functions and that are long-lived are subject to AMR. The EC enclosure device types that are subject to an AMR are listed in Table 6.2-2.

[ Reference 1, Sections 3, Table 3-1] The device types in Table 6.2-2 are used to identify the panels and cabinets (subject to AMR in the ECE) that might not have a unique equipment identification number in the equipment database.

The aging of non-metallies used in electrical cabinets is discussed below. Non-metallic subcomponents in electrical cabinets and panels fall into one of three categories.

Subcomoonent Category 1: Subcomoonent of Active Device - Excluded from AMR The first category includes subcomponents of active electrical devices such as teflon-coated sleeve bearings, polyester glass arc chutes, and polyester glass bus stand-off insulators used in 480 VAC circuit breakers. Active devices are excluded from the requirement for an AMR.

Subcomoonent Category 2: Conductor Insulation - Subject to AMR The second category includes the organic insulation of wiring or buswork. For example, crosslinked polyethylene insulated wiring is used in the 4 kV switchgear cabinets,480 VAC load centers, and 480 VAC MCCs. The internal operating temperatures for insulated connections in 4 kV switchgear cabinets and 480 VAC load centers is limited to 70 Centigrade (C) based on an ambient temperature of 40C per American National Standards Institute / Institute of Electrical and Electronic Engineers C37.20.2 - 1987. This is more than 10C below the 60-year service limiting temperatures for polyolefin and ethylene propylene rubber materials. Therefore, thermal aging is not plausible for the wiring contained in the 4 kV switchgear cabinets or the 480 VAC load centers. Internal operating temperatures in 480 VAC MCCs can approach 60-year service limits for polyolefm insulated wiring.

Therefore, thermal aging is plausible for polyolefin insulated wiring in 480 VAC MCCs and will require aging management. This wiring will be included in the cables aging management program discussed in Section 6.1, Cables, of the BGE LRA. Organically insulated control panel wiring is not subject to plausible thermal aging since operating temperatures in this service are well below 60-year service limiting temperatures for any organic insulating material, except polyvinyl chloride (PVC),

which is not used in this service. This category of subcomponents subject to aging applies to the 480 VAC MCCs of Group 5.

The bus splices in 4 kV switchgear cabinets are insulated with PVC boots. The operating temperature of the PVC boots is within SC of ambient. Ilowever, PVC is sensitive to thermal aging even at relatively low temperatures. Failure of air conditioning for an extended period of time could impact these boots. Therefore, thermal aging is plausible for the PVC boots in 4 kV switchgear cabinets and will require aging management. The aging of buswork insulation applies to the 4 kV switchgear cabinets in Group 3. These boots will be included in the cables aging management program discussed in Section 6.1, Cables, of the BGE LRA.

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Subcomoonent Categorv 3 Subcomnonent of Panel- Subject to AMR The third category includes subcomponents of the housing / cabinet such as insulating stand-off supports. Such subcomponents provide structural support for buswork and active ungrounded devices contained in the housing / cabinets. These subcomponents are subject to plausible aging if certain elastomers are used. They are not subject to plausible aging if a thermoset or thermoplastic material is used. For example, polyester-glass is used to support and insulate the buswork in the 4 kV switchgear cabinets,480 VAC load centers, and 480 VAC MCCs. However, polyester-glass is not subject to plausible aging. Elastomers, other than silicone rubber, used to insulate and support ungrounded devices are subject to plausible embrittlement and loss of insulation resistance due to thermal aging. Such aging could result in failure of the support under seismic loading or short circuit forces. The critical characteristic is the loss of flexibility since changes in mechanical properties typically precede changes in electrical properties for insulating clastomers. The 125 VDC chargers, inverters, and power distribution panels of Group 3 and Group 7 will be examined to explicitly identify any such clastomeric insulating supports. These supports will be explicitly included in the Age-Related Degradation Inspection (ARDI) Program or the existing " clean and inspect" Preventive Maintenance (PM) Procedure. Then, the PM or ARDI will be adjusted to monitor the support for the discovery ofloss of flexibility. There are other non-metallic subcomponents of the panels that do not have a license renewal function. 'Ihese subcomponents include dust shields and wiring penetration sleeves.

Terminal blocks are also considered to be subcomponents of the housing / cabinets. They are hard plastic, typically phenolic material, and are not subject to plausible thermal aging due to exposure to normal ambient temperatures. However, they are subject to electrical stress. The plausible aging identified with this stressor is the ohmic heating effects brought about by increased termination resistance associated with loosened connections. This aging effect and the management of it is addressed explicitly in this report. All groups of panels contain terminal blocks subject to electrical stressor aging. -

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-2 DEVICE TYPE WITHIN THE SCOPE OF LICENSE RENEWAL FOR THE ELECTRICAL COMMODITIES EVALUATION Device Type Component Supported and/or. Typical Associated EC

' Protected for Support / Protection BATT Batteries Terminals '

BKR Circuit Breakers Cabinet BUS Electrical Buses Cabinet CHGR Chargers Cabinet

{

DISC Disconnect Switch / Links Cabinet INV Inverters Cabinet  !

MCC MCCs Cabinet I NA 4 kV Local Control Stations Panel NB 480V Local Control Stations Panel l ND 125/250 VDC Local Control Stations Panel  !

PNL Panel Panels 6.2.2 Aging Management l The potential age-related degradation mechanisms (ARDMs) for the EC components are listed in l

Table 6.2-3. The plausible ARDMs are identified in the table by a check mark (/) in the appropriate column. The device types listed in Table 6.2-3 are those previously identified in Table 6.2-2 as passive and long-lived. For efficiency in presenting the results of these evaluations ir. this report, the components  :

here are grouped together based on device types. [ Reference 1, Section 4.4] The groups addressed are:

• Group 1 - Battery Terminals / Charger and Inverter Cabinets (electrical stressors, general l corrosion, and wear);

e Group 2 - Breaker Cabinets (electrical stressors, wear, and fatigue);

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• Group 3 - Bus Cabinets (electrical stressors, wear, and fatigue);

i e Group 4 - Disconnect Cabinets (electrical stressors, wear, and fatigue);

e Group 5 - MCC Panels (electrical stressors, wear, fatigue, and dynamic loading);

e Group 6 - 4 kV, 480 VAC, and 125/250 VDC Local Control Station Panels (electrical stressors, wear, fatigue, and general corrosion); and e Group 7 - Miscellaneous Panels (electrical stressors, wear, fatigue, and dynamic loading).

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

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APPENDIX A - TECHNICAL INFORMATION l 6.2 - ELECTRICAL COMMODITIES l

The clean and inspect PM procedures (as opposed to testing PMs) cover most of the electrical panels at CCNPP from 4 kV down to 125 VDC. However, some of the lower voltage electrical panels are not covered by clean and inspect PMs even if they are subject to testing PMs. These electrical panels will be subject to an ARDI Program to inspect for signs of age-related degradation.

TABLE 6.2-3 POTENTIAL PLAUSIBLE ARDMs Ene80 sures for Support and/or Protection of Electrical Commodities

. Device Types Potential ARDMs BA1T .SKR- 5118 . CHGR DISC . . INV MCC NA .NB ND: FNL terminal cabinet cabinet cableet cableet cabinet panel panel panel panel Corrosion Fatigue Crevice Corrosion Dynamic Loadmg (1)

/(5) /(7)

Electncal Stressors (2) /(2) /(3) /(1) /(4) /(1) /(5) /(6) /(6) /(6) /(7)_

Erosion Corrosion f atigue (1) /(2) /(3) /(4) /(5) /(6) /(7)

Fretting General Corrosion /(1) /(6)

Hydrogen Damage Intergranular Attack Microbiologically.

Induced Corrosion Neutron Embrittlement Oxidation Pstting Saline Water Attack Shrinkage / Creep Stress Corrosion Cracking Thermal Embrittlement Uniform Attack Wear /(2) /(3) /(1) /(4) /(1) /(5) /(6) /(6) /(6) /(7)

Note 1: Dynamic loading and fatigue are plausible for some but not all components associated with the indicated device type. Reference 1 contains the detail for the plausibility of these aging mechanisms for each individual component.

Note 2: Electrical stressors applies to phenolic terminal blocks.

Group I (battery terminals / charger and laverter cabinets)- Materials and Environment As Table 6.2-3 shows, the battery terminals are subject to corrosion, while the charger and inverter cabinets and associated terminal blocks are susceptible to the effects of electrical stressors and wear.

This group consists of the components for the 125 VDC Electrical Distribution System. The battery terminals are made of aluminum, lead, and copper, while the charger and inverter cabinets are constructed from carbon steel. [ Reference 1, Attachment 1, BATT-01, CHGR-01, INV-01, Attachments 2,5]

The environment that the battery terminals / charger and inverter cabinets experience is that of a mild controlled atmosphere within the CCNPP Auxiliary Building. The ECs in this group are subject to Application for License Renewal 6.2-7 Calvert Cliffs Nuclear Power Plant

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES 1

op,erational and maintenance activities. Terminal blocks attached to the cabinets are used for the l termination of electrical connections. Rese blocks are considered to be part of the cabinets that house ,

them and are included in this evaluation. They are phenolic material subject to the effects of electrical I stressors. The battery terminals are subject to the potentially corrosive environment of battery acid. j

[ Reference 1, BATT-01, CHGR-01, INV-01, Attachments 5,6]

Group 1 (battery terminals / charger and inverter cabinets)- Aging Mechanism Effects Electrical stressors (e.g., local ohmic heating) occur most commonly as a result of loose or improper terminations, which result in the degradation of organic materials and terminal block hardware. Loose terminations can occur as a result of the operation of panel components, as well as non-seismic vibration produced externally to the electrical components, which causes connections and terminals to loosen.

Degradation of organic material can occur as ohmic heating and elevated ambient temperatures cause terminal blocks to degrade. Terminal blocks are generally made of organic material that may lose its mechanical integrity (e.g., cracking and embrittlement may cause loss of support and insulating capabilities). [ Reference 1, Attachment 7s] The terminal blocks are subject to the above conditions and are, therefore, susceptible to the aging effects of electrical stressors.

Wear results from relative motion between two surfaces and from small, vibratory or sliding motions under the influence of a corrosive environment (fretting). Cabinet components, such as door hinges and circuit breaker racking mechanisms and other sliding parts, can wear from repeated openings for maintenance and testing. [ Reference 1, Attachments 6,7] Therefore, wear was determined to be plausible for the charger and inverter cabinets for which aging effects must be managed during the period of extended operation.

General corrosion is the thinning (wastage) of a metal by chemical attack (dissolution) at the surface by an aggressive environment. The consequences of the damage are the loss of load carrying cross-sectional area of the metal. General corrosion requires an aggressive environment. An important concern for pressurized water reactors is the attack of boric acid on carbon steels. Borated water has been observed to leak from piping, valves, storage tanks, etc., and fall on other steel components and attack the component from the outside. In addition, systems that contain saltwater can also leak and corrode carbon steel components. Acid leakage from station batteries can result in the corrosion of battery terminals. Therefore, general corrosion was determined to be plausible for the 125 VDC battery terminals. [ Reference 1, NB-01, Attachments 6,7]

If unmanaged, these ARDMs could eventually result in the loss of seismic support capability and electrical continuity under CLB design loading conditions.

Group 1 (battery terminals / charger and inverter cabinets)- Methods to Manage Aging Mitiention: Here are no feasible ways of preventing electrical stressors on the terminal blocks in battery charger and invener cabinets other than through proper installation and maintenance. There are also no feasible ways of preventing wear on cabinet components other than not operating the associated equipment; therefore, there are no practical means of preventing wear from occurring. In addition, there are also no feasible ways of preventing general corrosion of battery terminals other than proper maintenance.

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Discoverv: The effects of these ARDMs are detectable by visual techniques. Inclusion of the battery racks, charger and inverter cabinets in regular maintenance / overhaul inspections of these components under maintenance / overhaul / inspection repetitive tasks would result in the discovery of signs of these ARDMs.

Group 1 (battery terminals / charger and inverter cabinets)- Aging Management Program (s)

Mitigation: There are no CCNPP programs credited with the mitigation of electrical stressors, general corrosion, or wear on the battery terminals / charger and inverter cabinets and associated terminal blocks.

Discoverv: The CCNPP PM Program is credited for the discovery of electrical stressors, general corrosion, and wear of the battery terminals / charger and inverter cabinets and associated terminal blocks.

This program is described below.

Calvert Cliffs PM Pronram The CCNPP PM 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 4, Section 1.1)

The program is governed by CCNPP Administrative Procedure MN-1-102, " Preventive Maintenance l Program," and covers all PM activities for nuclear power plant structures and equipment within the plant, l including the ECE structural components (i.e., panels, etc.) within the scope of license renewal.

Guidelines drawn from industry experience and utility best practices were used in the development and enhancement of this program.

The PM Program includes periodic inspection of specific components through various maintenance activities. These activities provide an effective means to discover and manage the age-related degradation effects on these components. The program requires that an Issue Report be initiated according to CCNPP Procedure QL-2-100, " Issue Reporting and Assessment," for deficiencies noted during performance of PM tasks. Corrective actions are taken to ensure that the afTected components remain capable of performing their passive intended functions under all CLB conditions. [ Reference 4, l Section 5.2.B.I.fj The PM Program has had numerous levels of management review, all the way down to the specific implementation procedures. Specific responsibilities are assigned for evaluating and upgrading the PM l Program and for initiating program improvements based on system performance. Issue Reports are l initiated according to CCNPP Procedure QL-2-100 to request changes to the program that could improve or correct plant reliability and performance. Changes to the PM Program that require Issue Reports

( included changes to the PM task scope, frequency, process changes, results from operating experier.ce reviews, as well as other types of changes. [ Reference 4, Sections 5.1.A and 5.4] The PM Program also has undergone periodic evaluation by the NRC as part of their routine licensee assessment activities.

[ Reference 5)

Under the PM Program, electrical stressors and wear (and fatigue, dynamic loading, and general corrosion for other groups in this report) of the EC components are managed through existing PM.

[ Reference 1, Attachment 1) The following repetitive tasks will be modified to inspect for these ARDMs and include additional specified components where they are not currently included in the PM task.

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES The following CCNPP PM repetitive tasks are credited with the discovery of the effects of general ,

corrosion on the indicated 125 VDC oattery terminals. These repetitive tasks reference other procedures to test and inspect the batteries and thn terminals. Each of these repetitive tasks is currently performed every 12 weeks: [ Reference 1, Attachme.1t 1; References 6 through 9]

l e Calvert Cliffs Repetitive Task 10020008,"1 BATil1;" l e Calven Cliffs Repetitive Task 10020009,"I BATT12;"

e Calvert Cliffs Repetitive Task 20020008,"2 BATT22;" and e Calvert Cliffs Repetitive Task 20020009,"2 BATT21."

i The following repetitive tasks are credited for discovery of electrical stressors and wear on the battery charger cabinets. The repetitive tasks direct the user to clean, inspect, and calibrate the chargers. These repetitive tasks are currently performed every 96 weeks. [ Reference 1, Tables 5-1,5-2,5-3; References 10 and 11] ,

• Calvert Cliffs Repetitive Task 10020006, " BATTERY CHARGER 11;"

• Calvert Clifts Repetitive Task 10020007," BATTERY CHARGER 12;" '

• Calvert Cliffs Repetitive Task 10020015, " BATTERY CHARGER 23;"

• Calvert Cliffs Repetitive Task 10020016," BATTERY CHARGER 24;"

• Calvert Cliffs Repetitive Task 20020002," BATTERY CHARGER 21;"

• Calven Cliffs Repetitive Task 20020003," BATTERY CHARGER 22;"

e Calvert Cliffs Repetitive Task 20020014," BATTERY CHARGER 13;" and, e Calvert Cliffs Repetitive Task 20020015," BATTERY CHARGER 14."

The following repetitive tasks will discover wear and electrical stressors on the inverter cabinets. These Electrical PMs (EPMs) direct the user to inspect the inverter cabinets and calibrate the meters. These repetitive tasks are currently performed every 96 weeks. [ Reference 1, Table 5-1, Table 5-3, References 12 through 19]

e Calvert Cliffs Repetitive Task 10180013," Inverter 14;"

e Calvert Cliffs Repetitive Task 10180012,"Invener 13;"

e Calvert Clifts Repetitive Task 20180011, " Inverter 22;"

• Calvert Cliffs Repetitive Task 20180012," Inverter 23;"

• Calvert Cliffs Repetitive Task 20180013," Inverter 24;"

• Calvert Cliffs Repetitive Task 10180010," Inverter 11;"

• Calvert Clifts Repetitive Task 20180010, " Inverter 21;" and,

• Calvert Cliffs Repetitive Task 10180011," Inverter 12."

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Any corrective actions that are required during these inspections are performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the battery terminals / charger and inverter cabinets will remain capable of performing their intended function under all CLB conditions.

Group 1 (battery terminals / charger and inverter cabinets)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the 125 VDC battery terminals, charger and inverter cabinets:

The 125 VDC battery terminals, charger and inverter cabinets, and associated terminal blocks, have the intended functions of maintaining the seismic integrity and'or protection of safety-related components and electrical continuity under CLB design conditions.

Electrical stressors, general corrosion, and wear are plausible for the 125 VDC battery terminals, charger and inverter cabinets, and associated terminal blocks, which could lead to loss of seismic integrity and/or protection of safety-related components and electrical continuity under CLB design conditions.

e The CCNPP PM Program will provide for the discovery of the aging effects of electrical stressors, wear, and general corrosion that may be of concern for the 125 VDC battery terminals, charger and inverter cabinets using repetitive tasks. The repetitive tasks will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected. Inspections will be performed and appropriate corrective action will be taken where any of these ARDMs are discovered.

Therefore, there is reasonable assurance that the effects of these ARDMs on the 125 VDC battery terminals, charger and inverter cabinets will be adequately managed to maintain their intended function under all design loadings required by the CLB during the period of extended operation.

Group 2 (breaker embinets)- Materials and Environment As Table 6.2-3 shows, breaker cabinets and associated terminal blocks are susceptible to the effects of electrical stressors, fatigue, and wear. This group consists of the Reactor Protective System (RPS) tnp switchgear cabinets, which are made of carbon steel. [ Reference 1, Attachment 1, BKR-01, Attachments 2,5)

The environment that the RPS trip switchgear cabinets experience is that of a mild controlled atmosphere within the CCNPP Auxiliary Building. These switchgear cabinets are subject to operational and maintenance activities. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the cabinets that house them and are included in this evaluation. They are phenolic material subject to the effects of electrical stressors. [ Reference 1, BK.R-01, Attachments 5,6)

Group 2 (breaker cabinets)- Aging Mechanism Effects The effects of electrical stressors and wear were previously discussed in Group 1 (battery terminals / charger and invener cabinets) under the Aging Mechanisms Effect section. The RPS trip switchgeir cabinets are subject to the conditions described in the Group 1 Aging Mechanism Effects section and are, therefore, susceptible to the aging effects of these ARDMs.

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Fatigue damage results from progressive, localized structural change in materials subjected to fluctuating stresses and strains. Associated failures may occur at either high or low cycles in response to various kinds of loads (e.g., mechanical or vibrational loads, thermal loads, or pressure cycles). Fatigue cracks initiate and propagate in regions of stress concentration that intensify strain. The fatigue life of a component is a function of several variables, such as stress level, stress state, cyclic wave form, environmert, and the metallurgical condition of the material. Failure occurs when the endurance limit number of cycles (for a given load amplitude) is exceeded. All materials are susceptible (with varying endurance limits) when subjected to cyclic loading. [ Reference 1, Attachment 7s]

Fatigue typically occurs in switchgear cabinets from low-level vibrational loading of electrical equipment (e.g., relays, contactors, transformers, etc.) induced by AC hum or from mechanical operation. Mechanical stresses can cause housing welds to crack. [ Reference 1, Attachments 6,7]

Therefore, fatigue was also determined to be plausible for the RPS trip switchgear cabinets for which aging effects must be managed during the period of extended operation.

If tamanaged, these ARDMs could eventually result in the loss of seismic support capability under CLB design loading conditions.

Group 2 (breaker embinets)- Methods to Manage Aging Mitigation: There are no feasible ways of preventing electrical stressors on the terminal blocks in the RPS trip switchgear cabinets other than through proper installation and maintenance. There are also no feasible ways of preventing fatigue and wear on the RPS trip switchgear cabinets other than not operating the contained equipment; therefore, there are no practical means to prevent fatigue or wear from occurring.

Discoverv: The effects of these ARDMs are detectable by visual techniques. Inclusion of the RPS trip switchgear cabinets and associated terminal blocks in a regular maintenance / overhaul inspection of these components would provide for the discovery of these ARDMs.

Group 2 (br~aker embinets)- Aging Management Program (s)

Mitigation. There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue or wear on the RPS trip switchgear cabinets and associated terminal blocks.

Discoverv:

The CCNPP PM Program is credited with the discovery of these ARDMs on the RPS trip switchgear and associated terminal blocks. Refer to the previous discussion of the CCNPP PM Program in Group 1 (battery terminals, charger and inverter cabinets) under Aging Management Programs. The EPM below will be modified to include these ARDMs, where they are not presently included, and additional specified components, where they are not presently inspected.

Calvert Cliffs EPM 58500, " Reactor Trip Circuit Breaker Inspection," is the particular PM checklist credited for the discovery of electrical stressors, fatigue, and wear. The PM checklist directs the user to other procedures (e.g., field test and evaluation procedures) in performing the necessary inspections to reveal the presence of electrical stressors, fatigue, and wear on the RPS cabinets. This checklist is currently performed every 48 weeks. [ Reference 1, BRK-01, Attachments,1,8, Reference 20]

Application for License Renewal 6.2-12 Calvert Cliffs Nuclear Power Plant

e ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions. [ Reference 1, BKR-01, Attachments 1,8]

Group 2 (breaker cabinets) - Demonstration of Aging Management

. Based on the information presented above, the following conclusions can be reached with respect to the RPS trip switchgear cabinets and associated terminal blocks:

. The RPS trip switchgear cabinets and associated terminal blocks, have intended functions of maintaining the seismic integrity and/or protection of safety-related components under CLB design conditions.

• Electrical stressors, fatigue, and wear are plausible for the carbon steel RPS cabinets and associated terminal blocks, which could lead to loss of seismic integrity and/or protection of safety-related equipment under CLB design conditions.

• The CCNPP PM Program will provide for the discovery of electrical stressors, fatigue, and wear that may be of concern for the RPS trip switchgear cabinets and associated terminal blocks. This EPM will be modified to include these ARDMs where they are not presently included.

Therefore, there is reasonable assurance that the effects of these ARDMs on the RPS trip switchgear cabinets will be adequately managed to maintain their intended function under all design loadings required by the CLB during the period of extended operation.

Group 3 (bus cabinets)- Materials and Environment As Table 6.2-3 shows, the bus cabinets' terminal blocks are susceptible to the effects of electrical stressors, and the bus cabinets are susceptible to fatigue and wear. The 125 VDC distribution bus cabinets and associated terminal blocks are susceptible to wear and electrical stressor.: while the 480V and 4 kV bus cabinets and associated terminal blocks are susceptible to all three ARDMs. The bus cabinets are made of carbon L.-l. [ Reference 1, BUS-01, Attachments 1,5,6)

The environment that the bus cabinets and associated terminal blocks experience is that of a mild controlled atmosphere within the CCNPP buildings. However, at operating voltages of 480V and above, AC hum can produce exposure to low-level vibration. There are also routine maintenance and/or modifications performed on these bus cabinets, which requires manipulation of the bus cabinet subcomponents. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the cabinets that house them and are included in this evaluation. They are phenolic material subject to the effect of electrical stressors. [ Reference 1 BUS-01, Attachments 5,6]

Group 3 (bus cabinets)- Aging Mechanism Effects The effects of electrical stressors and wear were previously discussed in Group 1 (battery terminals / charger and inverter cabinets) and fatigue was previously discussed in Group 2 (breakers) under the Aging Mechanisms Effects sections.

l Application for License Renewal 6.2-13 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT (6) l APPENDIX A - TECHNICAL INFORMATION l

6.2 - ELECTRICAL COM.MODITIES j

! The bus cabinets are subject to the conditions described in the Groups 1 and 2 Aging Mechanism Effects sections and are, therefore, susceptible to the aging effects of these ARDMs. If unmanaged, these ARDMs could eventually result in the loss of seismic support capability and maintenance of electrical l continuity under CLB design loading conditions Group 3 (bus cabinets)- Methods to Manage Aging Mitintion: There are no feasible ways of preventing electrical stressors on the bus cabinets' terminal blocks other than through proper installation and maintenance. There are also no feasible ways of preventing fatigue and wear on the bus cabinets, which must be opened for maintenance, other than not i operating the contained equipment; therefore, there are no practical means to prevent fatigue or wear from occurring.

Discoverv: The effects of these ARDMs are detectable by visual techniques. Regular maintenance / overhaul inspections of these components under maintenance / overhaul / inspection repetitive tasks would discover signs of these ARDMs. [ Reference 1, BUS-01, Attachments 1,'8]

' Group 3 (bus cabinets)- Aging Management Program (s)

Mitiention: There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue, or wear on the bus cabinets and associated terminal blocks.

Discoverv: The CCNPP PM Program is credited with the discovery of these ARDMs on the 125 VDC, 480 VAC, and 4 kV bus cabinets and associated terminal blocks. Refer to the previous discussion of the CCNPP PM Program in Group 1 (battery terminals, charger and inverter cabinets) under Aging Management Programs.

The following CCNPP PM repetitive tasks are credited with the discovery of the effects of electrical stressors and wear on the indicated 125 VDC bus cabinets and associated terminal blocks. These repetitive tasks call for the user to inspect and clean the bus and cabinet. Each of these repetitive tasks is ]

currently performed every 10 years: [ References 21 through 24]

e Calvert Cliffs Repetitive Task 10020004," Inspect DC Bus 11 Disconnects;"

e Calvert Cliffs Repetitive Task 10020005," Inspect DC Bus 12 Disconnects;"

• Calvert Cliffs Repetitive Task 20020006," Inspect DC Bus 21 Disconnects;" and e Calvert Cliffs Repetitive Task 20020007, " Inspect DC Bus 22 Disconnects."

The following summarizes the corresponding checklists or PMs that will manage ARDMs for the j 480 VAC and 4 kV bus cabinets. The PM checklists or repetitive tasks will be modified to include these l ARDMs where they are not presently included, and additional specified components where they are not presently inspected.

t Calvert Cliffs EPM 05900, "480V Load Center and Transformers Cleaning and Inspection," is credited l with the discovery of effects of electrical stressors, fatigue, and wear on the 480V bus cabinets and associated terminal blocks. This EPM requires the cleaning and inspection of the load center bus enclosure, transformer, and buswork. This checklist is currently performed every eight weeks.

[ Reference 1, Table 5-3,05 BUS-01, Attachment 1, Reference 25]

Application for License Renewal 6.2-14 Calvert Cliffs Nuclear Power Plant i

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ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES The following CCNPP PM repetitive tasks are credited with the discovery of the effects of electrical stressors, fatigue, and wear on the indicated 4 kV bus cabinets and associated terminal blocks. These repetitive tasks call for the user to clean, inspect, and test the buses and cabinets. Each of these repetitive tasks is currently performed every 10 years:. [ Reference 1, Table 5-1, Attachment 1; References 26 through 29]

• Calvert Cliffs Repetitive Task 10040016,"4 kV Bus 11;"

Calvert Cliffs Repetitive Task 10040018,"4 kV Bus 14;"

• Calvert Cliffs Repetitive Task 20040016,"4 kV Bus 21;" and

• Calvert Cliffs Repetitive Task 20040018,"4 kV Bus 24."

Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions.

Group 3 (bus cabinets)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the bus cabinets:

• The 125 VDC,480 VAC, and 4 kV bus cabinets and associated terminal blocks have the intended functions of maintaining the seismic integrity and/or protection of safety-related components and maintenance of electrical continuity under CLB design conditions.

• Electrical stressors and wear are plausible for the 125 VDC bus cabinets and associated terminal blocks, and these two ARDMs, plus fatigue, are plausible for the 480 VAC/4 kV bus cabinets. 1 These ARDMs could lead to a loss of seismic integrity and/or protection of safety-related components and loss of electrical continuity under CLB design conditions.

• The CCNPP PM Program will provide for the discovery of any effects due to electrical stressors and wear on the 125 VDC bus cabinets, and electrical stressors, fatigue, and wear on the 4 kV and i 480 VAC bus cabinets and associated terminal blocks through the use of EPMs and repetitive tasks. These EPMs and repetitive tasks will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected.

Therefore, there is reasonable assurance that the effects of electrical stressors, fatigue, and wear on the 125 VDC,480 VAC, and 4 kV bus cabinets and associated terminal blocks will be adequately managed to maintain their intended functions under all design loadings required by the CLB during the period of i extended operation.

Application for License Renewal 6.2-15 Calvert Cliffs Nuclear Power Plant

l ATTACHMFNT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES

( Group 4 (disconnect cabinets)- Materials and Environment l Table 6.2-3 shows that the disconnect cabinets' terminal blocks are susceptible to the effects of electrical i

stressors, and the disconnect cabinets are susceptible to fatigue and wear. This group also contains contactor panels and associated terminal blocks, which are susceptible to electrical stressors, fatigue, and wear. The contactor panels and disconnect cabinets are made of carbon steel. [ Reference 1, DISC-01/02, Attachments 1,5,6)

The environment that these device types experience is that of a mild controlled atmosphere within the Auxiliary Building. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the cabinets that house them and are included in this evaluation. They are phenolic material subject to the effects of electrical stressors. [ Reference 1, DISC-01/02, Attachments 1,3]

l Group 4 (discommect cabinets)- A@ag Mechanism Effects The effects of electrical stressors and wear were previously discussed in Group 1 (battery terminals / charger and inverter cabinets) and fatigue was discussed in Group 2 (breaker cabinets) under the Aging Mechanisms Effect section.

The contactor panels and disconnect cabinets and associated terminal blocks are subject to the conditions

! described in the Groups I and 2 Aging Mechanisms Effect sections and are, therefore, susceptible to the aging effects of these ARDMs. If unmanaged, these ARDMs could eventually result in the loss of seismic support capability under CLB design loading conditions Group 4 (disconnect cabinets)- Methods to Manage Aging

Mitigation; Here are no feasible ways of preventing electrical stressors oa the contactor panels and l disconnect cabinets' terminal blocks other than through proper installation ar.d maintenance. There are also no feasible ways of preventing fatigue and wear on the contactor panch and disconnect cabinets, that must be opened for maintenance, other than not operating the contained rquipment; therefore, there are no practical means to prevent fatigue or wear from occurring.

Discoverv: A program of regular maintenance and inspection would discover indications of fatigue in the housing welds, indications of wear in fasteners or portions of contactor panels / disconnect cabinets that are held together for long periods of time, and indications of electrical stressors in terminal blocks before these ARDMs prevent the EC components from performing their passive intended functions.

l [ Reference 1, DISC-01/02, Attachments 1,8]

[

l Group 4 (discomaect cabinets)- Aging Management Program (s)

Mitigation: There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue, and wear on the contactor panels and disconnect cabinets and associated terminal blocks.

Discoverv: The CCNPP PM Program is credited with the discovery of these ARDMs on the disconnect  !

cabinets and associated terminal blocks. Refer to the previous discussion of the CCNPP PM Program in ,

Group 1 (battery terminals, charger and inverter cabinets) under Aging Management Programs. The Application for License Renewal 6.2-16 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES following EPM and repetitive task will be modified to include these ARDMs, where they are not presently included, and additional specified components, where they are not presently inspected.

Calvert Cliffs EPM 60601, " Third Train Containment Filtu Motor and Control Inspection," is the particular PM checklist credited for the discovery of these ARDMs. The PM checklist directs the user to other procedures (e.g., field test and evaluation procedures) in performing the necessary inspections to reveal the presence of electrical stressors, fatigue, and wear on the cabinets. The checklist presently directs the user to look for wear and discoloration on the stationary cubicle stabs. This checklist is currently performed every 96 weeks. [ Reference 1, DISC-01, Attachments,1,8, Reference 30]

Calvert Cliffs Repetitive Task 20320008,"21 Switchgear HVAC Unit Motor and Breaker Inspection," is credited for the discovery of all three ARDMs on the contactor panel and associated terminal blocks for the switchgear room air conditioning compressor cabinets. The repetitive task calls for the user to inspect for dust, discoloration, and signs of overheating. It also directs the user to other procedures and checklists to perform the inspection and testing of the contactor panels. This repetitive task is currently performed every two years. [ Reference 1, Table 5-1, Reference 31]

Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions.

Group 4 (disconnect cabinets)- Demon:tration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the i contactor panels and disconnect cabinets:

• The contactor panels and disconnect cabinets and associated terminal blocks have the intended functions of maintaining the seismic integrity and/or protection of safety-related components and malatenance of electrical continuity under CLB design conditions.

• Electrical stressors, fatigue, and wear are plausible for the carbon steel contactor panels and disconnect cabinets and associated terminal blocks, which could lead to loss of seismic integrity and/or protection of safety-related equipment or loss of electrical continuity under CLB design conditions..

• The CCNPP PM Program will provide for the discovery of the effects of electrical stressors, fatigue, and wear on the contactor panels and disconnect cabinets and associated terminal blocks i through the use of EPMs and repetitive tasks. These EPMs and repetitive tasks will be modified l to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected.

Therefore, there is reasonable assurance that the effects of electrical stressors, fatigue, and wear on the contactor panels and disconnect cabinets and associated terminal blocks will be adequately managed to L maintain their intended function under all design loadings required by the CLB during the period of extended operation Application for License Renewal 6.2-17 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECIINICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Group 5 (MCC panels)- Materials and Environment Table 6.2-3 shows that the MCC panels' terminal blocks are susceptible to the effects of electrical stressors, and the MCC panels are susceptible to fatigue, wear, and dynamic loading. The MCC panels in this group include the emergency diesci generator (EDG) MCC and engine auxiliary MCC panels, and other 480 VAC MCC panels. Only the EDG engine auxiliary MCC panels are susceptible to dynamic loading. The EDG MCC, engine auxiliary MCC, and other MCC panels are made of carbon steel.

[ Reference 1, MCC-01/02, Attachments 1,5,6]

The environment that these device types experience is that of a mild controlled / ventilated atmosphere within the Auxiliary and Turbine Buildings. The EDG engine auxiliary MCCs are located near the EDGs. In this location there is exposure to high-level vibration from the operation of the EDGs. There is also routine maintenance performed on these device types, which requires manipulation of the their subcomponents. Terminal blocks attached to the panels are used for the termination of electrical connections. These blocks are considered to be part of the panels that house them and are included in this evaluation. They are phenolic material subject to the effects of electrical stressors. [ Reference 1, MCC-01/02, Attachments 5,6]

Group 5 (MCC panels)- Aging Mechanism Effects The effects of electrical stressors and wear were previously discussed in Group 1 (battery terminals, charger and inverter cabinets) and fatigue was discussed in Group 2 (breaker cabinets) under the Aging Mechanisms Effect section.

The EDG MCC, engine auxiliary MCC, and other MCC panels and associated terminal blocks are subject to the conditions described in the Groups I and 2 Aging Mechanisms Effect sections and are, therefore, susceptible to the aging effects of these ARDMs. If unmanaged, these ARDMs could eventually result in the loss of seismic support under CLB design loading conditions.

Dynamic loading is only plausible for the EDG engine auxiliary MCC panels. Power plant components and structures are designed to accommodate loads that are expected in service. Experience has shown that while expected loads have been properly treated, dynamic loads not explicitly considered during design have occurred in service, causing material degradation and component failure. Although dynamic loading occurs through phenomena such as water hammer and unstable fluid flow, switchgears are typically not located near the sources of these loads. Components can be subjected to loading due to vibration from equipment such as compressors, diesel generators, and large pumps, as well as vibration from seismic loading. The expected effects of dynamic loading include failure of welds, and failure or degradation of fasteners, hardware, and supports. [ Reference 1, MCC-02, Attachments 6,7]

Since the EDG auxiliary MCC panels are located near the EDGs, which produce vibration, they are considered to be susceptible to dynamic loading. If unmanaged, these plausible ARDMs could eventually result in the loss of seismic support capability under CLB design loading conditions Application for License Renewal 6.2-18 Calvert Cliffs Nuclear Power Plant i

ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Group 5 (MCC panels)- Metbods to Manage Aging Mitigation: There are no feasible ways of preventing electrical stressors on the these MCC panels' terminal blocks other than through proper installation and maintenance. There are also no feasible ways of preventing fatigue and wear on the MCC panels other than not operating the contained equipment; therefore, there are no means to prevent fatigue or wear from occurring. In addition, there are no feasible ways of preventing dynamic loading on the EDG engine auxiliary MCC panels other than not running the EDGs. The EDGs must be periodically operated to ensure that they are capable of performing their design function; therefore, there are no means to prevent dynamic loading from occurring.

Discoverv: A program of regular maintenance and inspection for the EDG MCC panel, and associated terminal blocks, could discover indications of fatigue in the panel housing welds, indications of wear in fasteners or portions of the MCC panels that are held together for long periods of time, and indications of electrical stressors and dynamic loading before these ARDMs prevent the EDG MCC panels from performing their passive intended function. [ Reference 1, MCC-01/02, Attachments 1,8]

The effects of these ARDMs, as described above, are detectable by visual techniques. Inclusion of EDG engine auxiliary MCC panel and associated terminal blocks in an inspection program that inspects a sample of representative equipment for the signs of these ARDMs during the period of extended operation would provide for the discovery of these ARDMs in the EDG MCC panels. [ Reference 1, MCC-02, Attachments 1,8]

If unmanaged, these ARDMs could eventually result in the loss of seismic support capability under CLB design loading conditions Group 5 (MCC panels)- Aging Management Program (s)

Mitigation: There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue, and wear on the EDG MCC panel and associated terminal blocks and the mitigation of these three ARDMs and dynamic loading on the EDG engine auxiliary MCC panels.

Discoverv: The CCNPP PM Program is credited with the discovery of these ARDMs on the MCC panel and associated terminal blocks. Refer to the previous discussion of the CCNPP PM Program in Group 1 (battery terminals, charger and inverter cabinets) under Aging Management Programs.

The PM checklists below are credited for the discovery of the electrical stressors, fatigue, and wear on the EDG MCC panels. The PM checklists direct the user to other procedures (e.g., field tests and evaluation procedures) in performing the necessary inspections to reveal the presence of electrical stressors, fatigue, and wear. These EPMs will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected. These checklists are currently performed every six years. [ Reference 1, Table 5-3, MCC-01, and Attachments 2,8, References 32,33, and 34]

e EPM 06093," Check MCC 2AG Feeder Breakers;"

e EPM 06047," Check MCC IBG;" and e EPM 06049," Check MCC 2BG."

Application for License Renewal 6.2-19 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION l 6.2 - ELECTRICAL COMMODITIES Additionally, the following PM checklists are credited with the discovery of electrical stressors, fatigue, and wear on the indicated safety-related 480 VAC MCC panels and their associated terminal blocks.

These checklists direct the users in the use of field tests and evaluation pocedures for these panels. Each of these checklists is also to be modified in the same manner discussed above and is currently performed every six years. [ References 35 through 38]

e EPM 06067," Check MCC 104R and Feeder Breaker;"

e EPM 06038," Check MCC 114R and Feeder Breaker;"

l e EPM 06051, " Check MCC 204R and Feeder Breaker;" and e EPM 06039, " Check MCC 214R and Feeder Breaker."

A review of maintenance history has yielded no age-related failures of the equipment in this group.

Therefore, the ARDI Program will provide for the discovery of the effects of electrical stressors, fatigue, wear, and dynamic loading on the EDG engine auxiliary MCC panel and associated terminal blocks.

This program will examine representative samples of susceptible cabinets for signs of these ARDMs. The l

inspections will determine whether further action is needed and will be performed prior to the period of

- extended operation. [Refemnce 1, Attachment 8s) The ARDI Program is defined in the CCNPP IPA Methodology presented in Section 2.0 of this application.

The elements of the ARDI Program will include: [ Reference 1, MCC-02, Attachments 1,2,8]

• Determination of the examination sample size based on plausible aging effects and device types; e Identification of a sample of the device type population for inspection prior to the period of extended operation based on plausible ARDMs and the consequences of the loss of device type intended functions; e 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 adverse examination findings, including consideration of all design loadings required by the CLB and specification of required corrective actions; and e Evaluation of the need for follow-up examinations to monitor the progression of any age-related degradation.

~ Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions.

Group 5 (MCC panels)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the EDG MCC and engine auxiliary MCC panel and associated terminal blocks:

. These MCC panels and associated terminal blocks have an intended function of maintaining the seismic integrity and/or protection of safety-related components under CLB design conditions.

' Application for License Renewal 6.2-20 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (O APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Electrical stressors are plausible for the MCC panels' terminal blocks, while fatigue, and wear are plausible for the EDG and safety-related 480 VAC MCC carbon steel panels. Wear, fatigue, and dynamic loading are plausible for the EDG engine auxiliary MCC carbon steel panels. This susceptibility could lead to a loss of seismic integrity and/or protection of safety-related equipment.

The CCNPP PM Program will provide for the discovery of the effects of electrical stressors, fatigue, and wear on the EDG MCC and safety-related 480 VAC MCC panels and associated terminal blocks through the use of EPMs. 'Ihese EPMs ' vill be modified to assign additional l

specified components to the inspection / cleaning activitbs and to develop checklists for these '

other specified components. These PMs will also be modified to include these ARDMs where they are not presently included.

The CCNPP ARDI Program will provide for the discovery of the effects of electrical stressors, fatigue, wear, and dynamic loading that may be of concern for the EDG engine auxiliary MCC.

panel and associated terminal blocks. Inspections will be performed and appropriate corrective action will be taken if these ARDMs are discovered.

Therefore, there is reasonable assurance that the effects of electrical stressors, fatigue, wear, and l dynamic loading on the EDG MCC panels and engine auxiliary MCC panels will be adequately managed l in order to maintain their intended function under all design loadings required by the CLB during the  !

period of extended operation.

Group 6 (local control station panels)- Materials and Environment 1

I Table 6.2-3 shows that the 125/250 VDC,480 VAC, and 4 kV local control station panels and associated terminal blocks are susceptible to the effects of electrical stressors, fatigue, and wear. In addition to these ARDMs, the 480 VAC boric acid pump control panels and saltwater air compressor (SWAC) local l control station panels are also susceptible to the effects of general corrosion. All of the enclosures for  ;

these device types are constructed of carbon steel. [ Reference 1, Attachment 1, NA, NB, ND, i Attachments 5,6) '

The environment that these device types experience is that of a mild controlled atmosphere within the I CCNPP Auxiliary Building. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the panels that house them and are included in this evaluation. They are phenolic material subject to the effects of electrical stressors. The  ;

boric acid pump and SWAC local control stations are located in regions with piping that contains either i borated water or saltwater. [ Reference 1, NB-01, Attachments 6,7; Reference 1, MCC-02, PNL-01, Attachment 6]

Group 6 (local control station panels)- Aging Mechanism Effects The effects of electrical stressors, wear, and general corrosion were previously discussed in Group 1 (battery terminals, charger and inverter cabinets) and fatigue was discussed in Group 2 (breaker cabinets) under the Aging Mechanisms Effects sections.

Application for License Renewal 6.2-21 Calvert Cliffs Nuclear Power Plant

I ATTACHMENT @

L APPENDIX A - TECHNICAL INFORMATION l

6.2 - ELECTRICAL COMMODITIES l

l The local control station panels are subject to the conditions described in the Group 1 Aging Mechanisms Effects section and are, therefore, susceptible to the aging effects of these ARDMs. If unmanaged, these ARDMs could evemually result in the loss of seismic support capability under CLB design loading

- conditions.

General corrosion is plausible for the boric acid pump control panels and SWAC local control station panels. If unmanaged, general corrosion could eventually result in the loss of seismic support capability j under CLB design loading conditionsi [ Reference 1, NB-01]

Group 6 (local control station panels)- Methods to Manage Aging l Mitigation: There are no feasible ways of preventing electrical stressors on the these local control station panels other than through proper installation and maintenance. There are also no feasible ways of preventing fatigue and wear on the local control station panels other than not operating the contained equipment; therefore, there are no practical means to prevent fatigue or wear from occurring.

l Discoverv: A program of regular maintenance and inspection for the local control panel and associated

! terminal blocks would discover indications of fatigue in the housing welds, indications of wear in fasteners or portions of disconnect cabinets that are held together for long periods of time, and i indications of electrical stressors and general corrosion before these ARDMs prevent the local control

!. stations from performing their passive intended function. [ Reference 1, NA-01, NB-01, ND-01, l Attachments 1,8]

Inclusion in a program that examines a representative sample of susceptible local control panels for signs of these ARDMs could provide for the discovery of these ARDMs on these control panels.  !

Group 6 (local control station panels)- Aging Management Program (s) l Mitigation: There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue, L wear, and general corrosion on the local control station panel and associated terminal blocks.

Discoverv: The CCNPP PM Program is credited for the discovery of electrical stressors, fatigee, and wear on the local control station panels and associated terminal blocks. Refer to the previous discusion of the CCNPP PM Program in Group 1 (battery terminals, charger and inverter cabinets) under_ Aging Manag ment Programs. The following repetitive tasks and checklists will be modified to include these ARDMs where they are not presently included, and additional specified components, where they are not presently inspected. He various checklists and repetitive tasks credited here are listed in the following groups: )

4 kV Panels Calvert Cliffs EPM 04003,"nird Train 4 kV Breaker, Disconnect Switch, Relays, Meter, and Motor," is credited for the discovery of wear on the 4 kV panels and electrical stressors on some of these 4 kV ,

panels' terminal blocks. The PM checklist directs the user to other procedures (e.g., field test and j evaluation procedures) in performing the necessary inspections to reveal the presence of electrical j stressors and wear. This checklist is currently performed every two years. [ Reference 1, Table 5-1, Reference 39]

Application for License Renewal 6.2-22 Calvert Cliffs Nuclear Power Plant

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ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES The PM Program uses repetitive tasks to discover the effects of electrical stressors and wear on local control station panels and associated terminal blocks. The repetitive tasks credited for discovery of these ARDMs are: [ Reference 1, Tables 5-1,5-2,5-3]

CCNPP Repetitive Tasks 10120003, " Inspect 13 Salt Water Pump Motor, Normal Feed Breaker, {

Disconnect Switches, Calibrate Meters and Relays," and 20120003, " Inspect 23 Salt Water Pump Motor, Normal Feed Breaker, Disconnect Switches, Calibrate Meters and Relays," are both credited for the discovery of electrical stressors and wear on the saltwater cooling pump panel and associated terminal blocks. The PM checklist directs the user to perform inspections and to use other procedures (e.g., field test and evaluation procedures) in performing the necessary steps to reveal the presence of electrical stressors and wear. These repetitive tasks are currently performed every three years. [ Reference 1, Table 5-3, NA-01, Attachments 1,2,8, References 40 and 41] i e

CCNPP Repetitive Tasks 10520005, " Inspect 13 HPSI Pump Motor, Normal Feed Breaker, Disconnect Switches, Calibrate Meters and Relays," and 20520001, " Inspect 23 HPSI Pump j Motor, Normal Feed Breaker, Disconnect Switches, Calibrate Meters and Relays," are both credited for the discovery of electrical stressors and wear on the high pressure safety injection pumps panel and associated terminal blocks. The PM checklist directs the user to perform l inspections and to use other procedures (e.g., field test and evaluation procedures) in performing l

the necessary steps to reveal the presence of electrical stressors and wear. These repetitive tasks are currently performed every 96 weeks. [ Reference 1, Table 5-3, NA-01, Attachments 1,2,8, References 42 and 43]

480 VAC Local Control Stations A review of maintrwnee history has yielded no age-related failures of the equipment in this group.

{

Therefore, the ARDI Program is credited with the discovery of electrical stressors, fatigue, wear, and i general corrosion on the SWAC local control station panels and boric acid pump local control panels and i associated terminal blocks. De ARDI Program was previously described in Group 2 (breaker cabinets) under Aging Management Programs.

{

i EPM 30701, "CR/CSR Smoke Removal Damper Control Panel IC108 Inspection," is credited with the discovery of electrical stressors, fatigue, and wear on the Control Room HVAC compressor panel and associated terminal blocks, and an ARDI (as described above) for the same ARDMs on its heater panel and associated terminal blocks. The PM checklist directs the user to clean and inspect these panels to reveal the presence of ARDMs. This checklist is currently performed every 44 weeks. [ Reference 1, Table 5-1, NB-01, Attachments 1,2,8, Reference 44]

EPM 60600, " Containment Cooler Fan MTR/BKR/ Controller inspection," is credited with the discovery of electrical stressors, fatigue, and wear on the containment cooling fan local control station panel and l associated terminal blocks. The PM checklist directs the user to perform inspections and to use other l procedures (e.g., field test and evaluation procedures) in performing the necessary steps to reveal the presence of these ARDMs. This checklist is currently performed every 96 weeks. [ Reference 1, Table 5-1, NB-01, Attachments 1,2,8, Reference 45]

Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions.

Application for License Renewal 6.2-23 Calvert Cliffs Nuclear Power Plant

. ATTACHMENT (6)

I L

l APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES l

! 125/250 VDC Local Control Stations A review of maintenance history has yielded no age-related failures of the equipment in this group.

Therefore, the ARDI Program will provide for the discovery of the effects of electrical stressors and wear on the 125/250 VDC blowdown heat exchanger isolation and Auxiliary Feedwater Pump local control panel and associated terminal blocks if they are present. The ARDI Program was previously described in Group 5 (MCC panels) under Aging Management Programs. [ Reference 1, ND-01, l Attachments 1,2,8]

l Any corrective actions that are required will be performed in accordance with the CCNPP Corrective Actions Program, and will ensure that the components will remain capable of performing their intended I function under all CLB conditions.

Group 6 (local control station panels)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to electrical stressors, fatigue, wear, and general corrosion on the local control station panels susceptible to these ARDMs:

. The 4 kV, 480 VAC, and 125/250 VDC local control station panels' terminal blocks are susceptible to electrical stressors. These local control station panels are susceptible to fatigue, wear, and general corrosion and have an intended function of maintaining the seismic integrity and/or protection of safety-related components under CLB design conditions.

• De CCNPP PM Program will provide for the discovery of the effects of electrical stressors, fatigue, and wear on the 4 kV and 480 VAC local control station panel and associated terminal blocks through the use of repetitive tasks and EPMs. The repetitive tasks will provide for the discovery of electrical stressors and wear on the 4 kV local control station panel and associated terminal blocks, while the EPMs will provide for the discovery of electrical stressors, fatigue, and wear on the 480 VAC local control station panel and associated terminal blocks. These repetitive tasks and checklists will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected.

• The CCNPP ARDI Program will provide for the discovery of the effects of electrical stressors and/or wear that may be of concern for the 125/250 VIX: local control station panels and electrical stressors, fatigue, wear, and general corrosion on the 480 VAC SWAC local control station panels and boric acid pump local control panels. The 480 VAC Control Room HVAC compressor crankcase heater panel and associated terminal blocks are subject to electrical stressors, fatigue, and wear. Inspections will be performed and appropriate corrective action will be taken if these ARDMs are discovered.

Therefore, there is reasonable assurance that the effects of electrical stressors, fatigue, wear, and general corrosion on the 4 kV,480 VAC, and 125/250 VDC local control station panel and associated terminal blocks will be adequately managed to maintain the intended function of these components under all design loadings required by the CLB during the period of extended operation.

Application for License Renewal 6.2-24 Calvert Clifts Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECIINICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES Group 7 (miscellaneous panels)- Materials and Environment As Table 6.2-3 shows, the miscellaneous panels' terminal blocks are susceptible to the effects of electrical stressors, while the panels are susceptible to fatigue and wear. In addition, the EDG control panels in this group are also susceptible to dynamic loading. The panel enclosures in this group are made ofcarbon steel. [ Reference 1, PNL-01/02/03, Attachments 1,5,6)

The environment that these device types experience is that of a mild controlled or ventilated atmosphere within the CCNPP buildings. In some locations, there is exposure to high-level vibration from the operation of the EDGs. There is also routine maintenance performed on these device type, which requires manipulation of their subcomponents. Terminal blocks attached to the cabinets are used for the termination of electrical connections. These blocks are considered to be part of the panels that house them and are included in this evaluation. 'Ihey are phenolic material subject to the effects of electrical stressors.[ Reference 1, PNL-01/02/03, Attachments 5,6]

Group 7 (miscellaneous panels)- Aging Mechanism Effects The effects of electrical stressors and wear were previously discussed in Group 1 (battery terminals, charger and inverter cabinets), fatigue was discussed in Group 2 (breaker cabinets), and dynamic loading was discussed in Group 5 (MCC panels) under the Aging Mechanisms Effects sections.

The panels are subject to the conditions described in the Groups I and 2 Aging Mechanism Effects sections and are, therefore, susceptible to the aging effects of these ARDMs. If unmanaged, these ARDMs could eventually result in the loss of seismic support capability under CLB design loading conditions Group 7 (miscellaneous panels)- Methods to Manage Aging Mitigation: Theri, are no feasible ways of preventing electrical stressors on the panels' terminal blocks other than through proper installation and maintenance. There are also no feasible ways of preventing fatigue and wear on the panels other than not operating the contained equipment; therefore, there are no practical means to prevent fatigue or wear from occurring. In addition, there are no feasible ways of preventing dynamic loading on some of the panels (e.g., EDG control panels) other than not running the EDGs. The EDGs must be periodically operated to ensure that they are capable of performing their design function. Therefore, there is no practical means to prevent dynamic loading from occurring.

Discoverv: A program of regular maintenance and inspection for the panels would discover indications of fatigue in the housing welds, indications of wear in fasteners or portions of panels that are held together for long periods of time, indications of electrical stressors, and indications of dynamic loading (e.g., fasteners) before these ARDMs prevent the panels from performing their passive intended function.

[ Reference 1, PN1 01/02, Attachments 1,8]

Inclusion in a program that examines a representative sample of susceptible panels for signs of these ARDMs could provide for the discovery of these ARDMs on these panels.

Application for License Renewal 6.2-25 Calvert Cliffs Nuclear Power Plant

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.. .- 1 ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES I Gromp 7 (miscellaneens panels)- Aging Management Program (s) I j Mitiaatian: There are no CCNPP programs credited with the mitigation of electrical stressors, fatigue, l wear, and dynamic loading on the miscellaneous system panel and associated terminal blocks.

! )

Discoverv: The CCNPP PM Program is credited with the discovery of these ARDMs on the panel and

l. associated terminal blocks for the plant systems previously listed in the scoping section. Refer to the

)

l previous discussion of the CCNPP PM Program in Group 1 (battery terminals, charger and inverter j cabinets) under Aging Management Programs. The PM Program utilizes both EPMs and Instrument 4 i

PMs (IPMs) to discover the effects of electrical stressors, fatigue, and wear on miscellaneous system l

panel and associated terminal blocks. Rese EPMs and IPMs will be modified to include these ARDMs where they a:e not presently included, and additional specified components where they are not presently

!~ inspected. Currently,55% of the panels in this group are covered by PMs, while the remaining panels will be subject to an ARDI Program. He following are the EPMs and IPMs credited for discovery of these ARDMs: [ Reference 1, Tables 5-1,5-2,5-3, PNL-01/02/03, References 46 through 55] ,

i e

EPM 02800, " Clean and Inspect 125 VDC Distribution Panels," will provide for the discovery of wear and electrical stressors on the 125 VDC Electrical Distribution Panel and associated terminal blocks. This checklist is currently performed every 10 years.

• EPM 18800," Clean and Inspect 120V Vital Instrument AC Distribution Panels," will provide for i the discovery of wear and electrical stressors on the 120V Vital Instrument Distribution Panel l and associated terminal blocks. This checklist is currently performed every four years.

• EPM 32601, " Check SWGR RM HVAC Breakers & Motors," will provide for the discovery of electrical stressors, fatigue, and wear on the 11/12 and 21/22 switchgear room AC control cabinets and associated terminal blocks. This checklist is currently performed every two years.

i e EPM 73601, "H2 Recombiner Power Supply and Feeder BKR Inspection," will provide for the discovery of wear and electrical stressors on the 11,12,21 and 22 hydrogen recombiner power supply cabinets and associated terminal blocks. This checklist is currently performed every 96 weeks.

e IPM12104, " Clean and Inspect NSR DAS Fans, Filters, and Printers," will provide for the discovery of electrical stressors and wear on the data acquisition computer panel and associated terminal blocks. This checklist is currently performed every 24 weeks.

• IPM12103, " Clean and inspect SR DAS Fans and Filters," will provide for the discovery of electrical stressors and wear on the data acquisition computer panel and associated terminal blocks. This checklist is currently performed every 24 weeks.

• IPM13000," Clean and Inspect Unit 1 ESFAS Cabinet Filters," will provide for the discovery of electrical stressors and wear on the Unit 1 Engineered Safety Features Actuation System (ESFAS) cabinets and associated terminal blocks. This checklist is currently performed every 12v/ceks.

e IPM13001," Clean and Inspect Unit 2 ESFAS Cabinet Filters," will provide for the discovery of electrical stressors and wear on the Unit 2 ESFAS cabinets. This checklist is currently l performed every 12 weeks.

e IPM13118, " Clean and Inspect Control RM Panels / Cabinets and Vacuum RVLMS Filters,"

(RVLMS is Reactor Vessel Level Monitoring System) will provide for the discovery of electrical Application for License Renewal 6.2-26 Calvent Cliffs Nuclear Power Plant

,: o-

' ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES stressors and wear on the Unit 1 RPS cabinets, Control Room panels, and Nuclear Instrumentation Control Room panels and associated terminal blocks. This checklist is currently performed every two years, e IPM13119," Clean and Inspect Control RM Panels / Cabinets and Vacuum RVLMS Filters," will provide for the discovery of electrical stressors and wear on the Unit 2 RPS cabinets, Control Room panels, and Nuclear Instrument Control Room panels and associated terminal blocks.

This checklist is currently performed every two years.

The PM Program uses repetitive tasks to discover the effects of electrical stressors and wear on miscellaneous system panel and associated terminal blocks. These repetitive tasks will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected. The following repetitive tasks are credited for discovery of electrical stressors and wear for the indicated DG local control panels and plant computer panels. Each of these repetitive tasks is currently performed every five years: [ Reference 1, Tables 5-1,5-2,5-3, References 56 through 60]

• Repetitive Task 10240015, "lB DG Local Control Panel;"

• - Repetitive Task 20240007, "2B DG Local Control Panel;"

e Repetitive Task 10945001,"SSS CPU A PNL (Gould 9750);"

e Repetitive Task,20945001 "SSS CPU B PNL (Gould 9750);" and e Repetitive Task 20240009, "2A DG Local Control Panel."

A review of maintenance history has yielded no age-related failures of the equipment in this group.

Therefore, the ARDI Program will provide for the discovery of the effects of electrical stressors, fatigue, wear, and dynamic loading on the remainder of the miscellaneous panels and associated terminal blocks.

The ARDI Program was previously described in Group 5 (MCC panels) under Aging Management Programs. [ Reference 1, PNL-01/02/03, Attachments 1,2,8]

Any corrective actions that are required will be performed in accordance with the CCNPP Corrective

Actions Program, and will ensure that the components will remain capable of performing their intended function under all CLB conditions.

Groep 7 (miscellaneous panels)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to electrical stressors, fatigue, and wear on the miscellaneous panels susceptible to these ARDMs:

• - The CCNPP miscellaneous system panels' terminal blocks are susceptible to electrical stressors.

The miscellaneous panels are susceptible to fatigue, wear, and dynamic loading, and have the intended functions of maintaining the seismic integrity and/or protection of safety-related components and electrical continuity under CLB design conditions.

. .The CCNPP PM Program will provide for the discovery of the effects of electrical stressors, fatigue, wear, and dynamic loading on the miscellaneous system panel and associated terminal blocks through the use of repetitive tasks, EPMs and IPMs. The repetitive tasks and IPMs will

' discover electrical stressors and wear of the panel and associated termir.al blocks, while the EPMs will discover electrical stressors, fatigue, and wear of the panel and associated terminal blocks.

Application for License Renewal 6.2-27 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES These repetitive tasks and checklists will be modified to include these ARDMs where they are not presently included, and additional specified components where they are not presently inspected.

The CCNPP ARDI Program will provide for the discovery of the effects of electrical stressors, fatigue, wear, and dynamic loading that may be of concern for the miscellaneous system panels and associated terminal blocks. Inspections will be performed and appropriate corrective action will be taken if these ARDMs are discovered.

Therefore, there is reasonable assurance that the effects of these ARDMs on the miscellaneous system panels will be adequately managed to maintain their intended functions under all design loadings required by the CLB during the period of extended operation.

6.2.3 Conclusion The programs discussed for ECE 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 ECE will be maintained, consistent with the CLB, during the period of extended operation.

I l 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 aging management review.

l j

j Application for License Renewal 6.2-28 Calvert Cliffs Nuclear Power Plant

ATTACHMENT @

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE i Pmgrame ' Credited For-Modified Repetitive Task 10020008, Discovery of general corrosion on the 125 VDC Battery "I BATril" 11 terminals (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10020009, Discovery of general corrosion on the 125 VDC Battery "I BATT12" 12 terminals (Group 1). This repetitive task will be 3 modified to include these ARDMs where they are not l presently included and additional specified components j I

where they are not presently inspected.

Modified Repetitive Task 20020008, Discovery of general corrosion on the 125 VDC Battery -

"2 BATf22" 22 terminals (Group 1). This repetitive task will be

modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

l Modified Repetitive Task 20020009,"2 Discovery of general corrosion on the 125 VDC Battery l BATT21" 21 terminals (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components l where they are not presently inspected.

Modified Repetitive Task 20040016, Discovery of the effects of electrical stressors, fatigue, "4 kV Bus 21" and wear of 4 kV Bus 21 cabinets (Group 3). This i repetitive task will be modified to include these ARDMs where they are not presently included and

additional specified components where they are not  ;

presently inspected. j Modified Repetitive Task 20040018, Discovery of the effects of electrical stressors, fatigue, l "4 kV Bus 24" and wear of 4 kV Bus 24 cabinets (Group 3). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected. I Modified Repetitive Task 10040016, Discovery of the effects of electrical stressors, fatigue, l "4kV Bus 11" and wear of 4 kV Bus 11 cabinets (Group 3). This i repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Application for License Renewal 6.2-29 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE

Program . . Credited Fori l Modified Repetitive Task 10040018, Discovery of the effects of electrical stressors, fatigue, "4kV Bus 14" and wear of 4 kV Bus 14 cabinets (Group 3). This repetitive task will be modified to include these l

ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10240015 "lB Discovery of the effects of electrical stressors and wear DG Local Control Panel" of the IB EDG local control panel (Group 7). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20240007,"2B Discovery of the effects of electrical stressors and wear DG Local Control Panel" of the 2B EDG local control panel (Group 7). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20240009,"2A Discovery of the effects of electrical stressors and wear DG Local Control Panel" of the 2A EDG local control panel (Group 7). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20320008,"2 Discovery of the effects of electrical stressors, fatigue HVAC/A SWGR Room A/C and wear of the 21 and 22 switchgear room A/C Compressor" compressor contactor panels (Group 4). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10120003, Discovery of the effects of electrical stressors and wear

" Inspect 13 Salt Water Pump on the saltwater cooling pump panels (Group 6). This Motor, Normal Feed Breaker, repetitive task will be modified to include these Disconnect Switches, Calibrate ARDMs where they are not presently included and Meters and Relays" additional specified components where they are not presently inspected.

l Application for License Renewal 6.2-30 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE Program 1 Credited For Modified Repetitive Task 20120003, Discovery of the effects of electrical stressors and wear

" Inspect 23 Salt Water Pump on the saltwater cooling pump panels (Group 6). This Motor, Normal Feed Breaker, repetitive task will be modified to include these Disconnect Switches, Calibrate ARDMs where they are not presently included and Meters and Relays" additional specified components where they are not presently inspected.

Modified Repetitive Task 10520005,"I SI Discovery of the effects of electrical stressors and wear HPSI Pump 13 Motor" l of the 13 HPSI Pump disconnect panels (Group 6).

This repetitive task will be modified to include these ARDMs where they are not presently included and

{

additional specified components where they are not presently inspected.

Modified Repetitive Task 20520001,"23 Discovery of the effects of electrical stressors and wear Si HPSI Pump Motor" of the 23 HPSI Pump disconnect panels (Group 6).

This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10945001,"SSS Discovery of the effects of electrical stressors and wear i CPU A Panel (Gould 9750)" of the plant computer panels (Group 7). This repetitive I task will be modified to include these ARDMs where l they are not presently included and additional specified i components where they are not presently inspected. 1 Modified .Repetitiw Task 20945001, "SSS Discovery of the effects of electrical stressors and wear CPU B Panel (Gould 9750)" of the plant computer panels (Group 7). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Tasks Discovery of the efTects of electrical stressors and wear 10020006," Battery Charger 11" on the 125 VDC Battery Charger cabinets 11, 12, 23, 10020007," Battery Charger 12" and 24 (Group 1). These repetitive tasks will be 10020015," Battery Charger 23" modified to include these ARDMs where they are not 10020016," Battery Charger 24" presently included and additional specified components where they are not presently inspected.

Application for License Renewal 6.2-31 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE Program x Credited For Modified Repetitive Tasks Discovery of the effects of electrical stressors and wear 20020002," Battery Charger 21" on the 125 VDC Battery Charger cabinets 13, 14, 21, 20020003," Battery Charger 22" and 22 (Group 1). These repetitive tasks will be 20020014," Battery Charger 13" modified to include these ARDMs where they are not 20020015," Battery Charger 14" presently included and additional specified components where they are not presently inspected. 4 Modified Repetitive Task 10180013, Discovery of the effects of electrical stressors and wear

" Inverter 14" on Inverter 14 cabinet (Group 1). This repetitive task i will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10180012, Discovery of the effects of electrical stressors and wear

" Inverter 13" on Inverter 13 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20180011, Discovery of the effects of electrical stressors and wear l

" Inverter 22" on Inverter 22 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20180012, Discovery of the effects of electrical stressors and wear j

" Inverter 23" on Inverter 23 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20180013, Discovery of the effects of electrical stressors and wear

" Inverter 24" on Inverter 24 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10180010, Discovery of the effects of electrical stressors and wear l

" Inverter 11" on Inverter 11 cabinet (Group 1). This repetitive task j will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

l Application for License Renewal 6.2-32 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE Program Credited For Modified Repetitive Task 20180010, Discovery of the effects of electrical stressors and wear

" Inverter 21" on Invester 21 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

. Modified Repetitive Task 10180011, Discovery of the effects of electrical stressors and wear

" Inverter 12" on Inverter 12 cabinet (Group 1). This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10020004, Discovery of the effects of electrical stressors and wear

" Inspect DC Bus 11 of the 125 VDC bus 11 disconnect cabinets (Group 3).

Disconnects" This repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 10020005, Discovery of the effects of electrical stressors and wear

" Inspect DC Bus 12 of the 125 VDC bus 12 disconnect cabinets (Group 3).

Disconnects" This repetitive task will be modified to include these i ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20020006, Discovery of the effects of electrical stressors and wear

" Inspect DC Bus 21 of the 125 VDC bus 21 disconnect cabinets (Group 3).

Disconnects" His repetitive task will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified Repetitive Task 20020007, Discovery of the effects of electrical stressors and wear

" Inspect UC Bus 22 of the 125 VDC bus 22 disconnect cabinets (Group 3).

Disconnects" This repetitive task will be modified to include these ARDMs where they are not presently included and 1 additional specified components where they are not i presently inspected. j l

l l

l Application for License Renewal 6.2-33 Calvert Cliffs Nuclear Power Plant

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE 0Programa Credited Fori Modified EPM 04003," Third Train 4 kV Discovery of the effects of wear and electrical stressors Breaker, Disconnect Switch, on the 13 and 23 Servico Water Pump, Saltwater Pump, Relays, Meter, and Motor" and HPSI Pump disconnect panels (Group 6). This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified EPM 0606," Check MCC 104R Discovery of the efTects of electrical stressors, fatigue, and Feeder Breaker" and wear on the safety-related 480 V MCC panels EPM 06038," Check MCC ll4R (Group 5). These PMs will be modified to include and Feeder Breaker" these ARDMs where they are not presently included, EPM 06051," Check MCC 204R and additional specified components where they are not ,

and Feeder Breaker" presently inspected.

EPM 06039," Check MCC 214R and Feeder Breaker"  ;

Modified EPM 06049," Check MCC 2BG" Discovery of the effects of electrical stressors, fatigue, and wear on the EDG MCC 21G panel (Group 5). This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified EPM 06093," Check MCC 2AG Discovery of the effects of electrical stressors, fatigue, and Feeder Breaker" and wear on the EDG MCC panels 1IG (Group 5).

This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

-4 Modified EPM 60600," Containment Discovery of the effects of electrical stressors, fatigue, Cooler Fan and wear on the Containment Cooling Fan local control MTR/BKR/ Controller station panels (Group 6). This PM will be modified to j Inspection" include these ARDMs where they are not presently included and additional specified components where  !

they are not presently inspected.

Modified EPM 60601," Third Train Discovery of the effects of electrical stressors, fatigue,  !

Containment Filter Motor and and wear on the Containment Cooler disconnect I ControlInspection" cabinets (Group 4). This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

l l

i Application for License Renewal 6.2-34 Calvert Cliffs Nuclear Power Plant

e. ..

, ATTACHMENT (6) l l APPENDIX A - TECIINICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 l LIST OF AGING MANAGEMENT PROGRAMS FOR ECE I

Program  ? Credited For Modified EPM 73601,"H2 Recombiner Discovery of the efTects of electrical stressors and wear Power Supply and Feeder BKR on the Hydrogen Recombiner power supply cabinets Inspection" (Group 7). This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified IPM12104," Clean and Inspect Discover of the effects of electrical stressors and wear NSR DAS Fans, Filters, and on the Data Acquisition computer panels (Group 7).

Printers" This PM will be modified to include these ARDMs where they are not presently included and additional specified components where they are not presently inspected.

Modified IPM12103," Clean and Inspect Discover of the effects of electrical stressors and wear SR DAS Fans and Filters" on the Data Acquisition computer panels (Group 7).

This PM will be modified to include these ARDMs where they are not presently included and additional i specified components where they are not presently inspected.

l Modified IPM13000 " Clean / Inspect Unit 1 Discover of the effects of electrical stressors and wear ESFAS Cabinets" on the Unit 1 ESFAS cabinets (Group 7). This PM will be modified to include these ARDMs where they are l not presently included and additional specified I components where they are not presently inspected.

Modified IPM13001," Clean and Inspect Discover of the effects of electrical stressors and wear Unit 2 ESFAS Cabinet Filters" on the Unit 2 ESFAS cabinets (Group 7). This PM will be modified to include these ARDMs where they are .

not presently included and additional specified components where they are not presently inspected.

Modified EPM 02800," Clean and Inspect Discovery of the effects of electrical stressors and wear 125 VDC Distribution Panels" on the 125 VDC Electrical Distribution Panels. This PM will be modified to include these ARDMs, where they are not presently included, and any additional specified components where they are not presently

, inspected (Group 7).

I Application for License Renewal 6.2-35 Calvert Cliffs Nuclear Power Plant

ATTAC'I4 MENT (6) l APPENDIX A - TECHNICAL INFORMATION

, 6.2 - ELECTRICAL COMMODITIES l

l TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE '

Prograan-. . i. Credited For. 1

[ Modified EPM 05900,"480V Load Center Discovery of the effects of electrical stressors, fatigue and Transformer Cleaning and and wear on the 480V Bus cabinets. His PM will be Inspection" modified to include these ARDMs, where they are not presently included, and any - additional specified components where they are not presently inspected (Group 3).

Modified EPM 06047," Check MCC IBG" Discovery of the effects of electrical stressors, fatigue, and wear on the EDG MCC panel 12G. This PM will be modified to include these ARDMs, where they are not presently included, and any additional specified components where they are not presently inspected (Group 5).

Modified EPM 18800," Clean and Inspect Discovery of the effects of electrical stressors and wear 120VInstrument AC on the Service Water distribution panels. This PM wil:

Distribution Panels" be modified to include these ARDMs, where they are not presently included, and any additional specified components where they are not presently inspected (Group 7).

Modified EPM 30701,"CR/CSR Smoke Discovery of the effects of electrical stressors and wear Removal Damper Control Panel on the Control Room HVAC compressor. This will be IC108 Inspection" modified to include these ARDMs, where they are not _ i presently included, and any additional specified i components where they are not presently inspected (Group 6).

Modified EPM 32601," Check Switchgear Discovery of the effects of electrical stressors, fatigue, l Room HVAC Breakers and and wear on the 11/12 and 21/22 switchgear room AC  !

Motors" compressor contactor cabinets. This PM will be l modified to include these ARDMs, where they are not presently included, and any additional specified components where they are not presently inspected (Group 7).  !

Modified IPM13118," Clean and Inspect Discovery of the effects of electrical stressors and wear Control Room Panels / Cabinets on the Unit 1 RPS cabinets, Control Room panels, and and Vacuum RVLMS Filters" Nuclear Instrumentation Control Room panels. This PM will be modified to include these ARDMs, where they are not presently included, and any additional i specified components where they are not presently i mspected (Group 7).

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l' Application for License Renewal 6.2-36 Calvert Cliffs Nuclear Power Plant i

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APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES TABLE 6.2-4 LIST OF AGING MANAGEMENT PROGRAMS FOR ECE j L Program ; ' Credited For :

Modified IPM13119," Clean and Inspect Discovery of the effects of electrical stressors and wear Control Room Panels / Cabinets on the Unit 2 RPS cabinets, Control Room panels, and and Vacuum RVLMS Filters" J

Nuclear Instrumentation Control Room panels. This PM will be modified to include these ARDMs, where i they are not presently included, and any additional specified components where they are not presently inspected (Group 7).

Modified EPM 58500," Reactor Trip Discovery of the effects of electrical stressors, fatigue, Circuit BreakerInspection" and wear on the RPS switchgear cabinets (Group 2).

This PM will be modified to include these ARDMs, where they are not presently included.

New CCNPP ARDI Program Discovery of the effects electrical stressors, fatigue, wear, and dynamic loading on the EDG auxiliary MCC panels (Group 5); electrical stressors and/or wear for 125/250 VDC local control station panels and electrical stressors, fatigue, wear, and general corrosion on the SWAC and Boric Acid Pump local control panels j (Group 6); and electrical stressors, fatigue, wear, and i I

dynamic loading on miscellaneous panels (Group 7).

i Application for License Renewal 6.2-37 Calvert Cliffs Nuclear Power Plant t

ATTACHMENT (6)

APPENDIX A - TECHNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES 6.2.4 References

1. "CCNPP Aging Management Review Report for the B ew Commodities, Volumes I and 2" Revision 1, July 23,1997

2. Calvert Cliffs Nuclear Power Plant, Updated Final Safety Analysis Report, Revision 20

3. Letter from Mr. G. C. Creel (BGE) to Mr. T. T. Martin (NRC), dated May 29,1990, " Unit 1 Startup Assessment"

4. CCNPP Administrative Procedure MN-1-102," Preventive Maintenance Program," Revision 5,

, September 27,1996

> 5. Letter from Mr. R. W. Cooper II (NRC) to Mr. C. H. Cruse (BGE), dated May 31, 1996, "Calvert Cliffs Plant Performance Review Results"

6. CCNPP Repetitive Task 10020008,"I Battery 11"

7. CCNPP Repetitive Task 10020009,"I Battery 12"

8. CCNPP Repetitive Task 20020008,"2 Battery 22"

9. CCNPP Repetitive Task 20020009,"2 Battery 21"

10. CCNPP Repetitive Tasks 10020006, " Battery Charger 11;" 1002007, " Battery Charger 12;"

10020015, " Battery Charger 23;" 10020016, " Battery Charger 24"

11. CCNPP Repetitive Tasks 20020002, " Battery Charger 21;" 2002003, " Battery Charger 22;"

20020014, " Battery Charger 13;" 20020015, " Battery Charger 14"

12. CCNPP Repetitive Task 10180013," Inverter 14"

13. CCNPP Repetitive Task 10180012," Inverter 13"

14. CCNPP Repetitive Task 20180011, " Inverter 22"

15. CCNPP Repetitive Task 20180012 " Inverter 23"

16. CCNPP Repetitive Task 20180013," Inverter 24"

17. CCNPP Repetitive Task 10180010," Inverter 11"

18. CCNPP Repetitive Task 20180010, " Inverter 21"

19. CCNPP Repetitive Task 10180011," Inverter 12" 20 CCNPP EPM 58500 Checklist Sheet, " Reactor Trip Circuit Breaker Inspection," Revision 0, April 8,1993

21. CCNPP Repetitive Task 10020004," Inspect DC Bus 11 Disconnects"

22. CCNPP Repetitive Task 10020005," Inspect DC Bus 12 Disconnects"

23. CCNPP Repetitive Task 20020006 " Inspect DC Bus 21 Disconnects"

24. CCNPP Repetitive Task 20020007 " Inspect DC Bus 22 Disconnects"

25. CCNPP EPM 05900 Checklist Sheet, "480V Load Center and Transformer Cleaning and Inspection," Revision 0, September 3,1994

26. CCNPP Repetitive Task 10040016,"4 kV Bus 11" Application for License Renewal 6.2-38 Calvert Cliffs Nuclear Power Plant

1 ATTACHMENT (6)

APPENDIX A - TECliNICAL INFORMATION 6.2 - ELECTRICAL COMMODITIES

27. CCNPP Repetitive Task 10040018,"4 kV Bus 14"

28. CCNPP Repetitive Task 20040016,"4 kV Bus 21"

29. CCNPP Repetitive Task 20040018,"4 kV Bus 24"

30. CCNPP EPM 60601 Checklist Sheet, " Third Train Containment Filter Motor and Control Inspection" Revision 0, June 8,1994

31. CCNPP Repetitia Task 20320008, "21 Switchgear HVAC Unit Motor and Breaker Inspection"

32. CCNPP EPM 06093 Checklist Sheet, " Check MCC 2AG Feeder Breakers," Revision 1, August 14,1997

33. CCNPP EPM 06047 Checklist Sheet," Check MCC IBG," Revision 0, August 14,1997

34. CCNPP EPM 06049 Checklist Sheet," Check MCC 2BG," Revision 0, August 14,1997

35. CCNPP EPM 06067 Checklist Sheet, " Check MCC 104R and Feeder Breaker," Revision 0, January 8,1992

36. CCNPP EPM 06038 Check List Sheet, " Check MCC ll4R and Feeder Breaker," Revision 0, February 3,1992

37. CCNPP EPM 06051 Checklirt Sheet, " Check MCC204R and Feeder Breaker," Revision 0, January 3,1992

38. CCNPP EPM 06039 Checklist Sheet, " Check MCC214R and Feeder Breaker," Revision 0, January 3,1992

39. CCNPP EPM 04003 Checklist Sheet, " Third Train 4 V Breaker, Disconnect Switch, Relays, Meter, and Motor" Revision 0, March 20,1993

40. CCNPP Repetitive Task 10120003, " Inspect 13 Salt Water Pump Motor, Normal Feeder Breaker, Disconnect Switches, Calibrate Meters and Relays"

41. CCNPP Repetitive Task 20120003, " Inspect 23 Salt Water Pump Motor, Normal Feeder Breaker, Disconnect Switches, Calibrate Meters and Relays"

42. CCNPP Repetitive Task 10520005, " Inspect 13 HPSI Pump Motor, Normal Feeder Breaker, Disconnect Switches, Calibrate Meters and Relays"

43. CCNPP Repetitiw Task 20520001, " Inspect 23 HPSI Pump Motor, Normal Feeder Breaker, Disconnect Switches, Calibrate Meters and Relays"

44. CCNPP EPM 30701 Checklist Sheet,"CR/CSR Smoke Removal Damper Control Panel IC108 Inspection," Revision 0, January 6,1993

45. CCNPP EPM 60600 Checklist Sheet, " Containment Fan Motor / Breaker / Controller Inspection,"

Revision 0, July 30,1994

46. CCNPP EPM 02800 Checklist Sheet, " Clean and Inspect 125 VDC Distribution Panels,"

Revision 0, March 2,1992

47. CCNPP EPMIS800 Checklist Sheet, " Clean and Inspect 120V Vital Instrument AC Distribution Panels," Revision 0, March 2,1992 Application for License Renewal 6.2-39 Calvert Cliffs Nuclear Power Plant

.s o ATTACHMENT (6)

APPENDIX A - TECHNICAL'INFORMATION 6.2 - ELECTRICAL COMMODITIES

48. CCNPP EPM 32601 Checklist Sheet, " Check Switchgear Room IIVAC Breakers and Motors,"

Revision 0, January 3,1992

49. CCNPP EPM 73601 Checklist Shx4, "II2 Recombiner Power Supply and Feeder Breaker Inspection," Revision 0, April 16,1996  ;

50. CCNPP IPM12104 Checklist Sheet," Clean and Inspect NSR DAS Fans, Filters, and Printers," I Revision 0, February 15,1992

51. CCNPP IPM12103 Checklist Sheet, " Clean and Inspect SR DAS Fans, Filters," Revision 0, February 15,1992  !

52. CCNPP IPM13000 Checklist Sheet, " Clean and Inspect Unit 1 ESFAS Cabinet Filters,"

Revision 0, November 8,1991

53. CCNPP IPM13001 Checklist Sheet, " Clean and Inspect Unit 2 ESFAS Cabinet Filters,"

Revision 0, November 8,1991

54. CCNPP IPM13118 Checklist Sheet, " Clean and Inspect Control Room Panels / Cabinets and Vacuum RVLMS Filters," Revision 0, July 8,1996

55. CCNPP IPM13119 Checklist Sheet, " Clean and Inspect Control Room Panels / Cabinets and Vacuum RVLMS Filters," Revision 0, July 8,1996

56. CCNPP Repetitive Task 10240015,"IB DG Local Control Panel"

57. CCNPP Repetitive Task 20240007 "2B DG Local Control Panel"

58. CCNPP Repetitive Task 10945001,"SSS CPU A Panel (Gould 9750)"

59. CCNPP Repetitive Task 20945001,"SSS CPU B Panel (Gould 9750)"

60. CCNPP Repetitive Task 20020009, "2A DG Local Control Panel" i

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Application for License Renewal 6.2-40 Calvert Cliffs Nuclear Power Plant I