ML20199C807
| ML20199C807 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/14/1997 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9711200146 | |
| Download: ML20199C807 (114) | |
Text
_;
Cuant.ts 11. Caost Baltimore Cias and Electric Company Vice President Cahert Chtts Nuclear Power Plant Nuclear Energy 165nahert Chits Parkway Luc , Maryland 20h57 410 495-4455 November 14,1997 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear power Plant Unit Nos.1 & ".; Docket Nos. 50-317 & 50-318 Request for Review and Approval of System and Commodity Reports for License Renewal
REFERENCES:
(a) Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated August 18,1995,Integra:ed Plant Assessment Methodology (b) Letter from Mr. D. M. Crutchlicld (NRC) to Mr. C. H. Cruse (BGE),
-. dated, April 8,1996, Final Safety Euuation (FSE) Concerning The Baltimore Gas and Electric Company Report entitled, Integrated Plant Assessment Methodology J (c) Letter from Mr. S. C. Flanders (NRC), dated March 4,1997, "St.mmary of Meeting with Baltimore Gas and Electric Company (BGE) on BGE License Renewal Activities" This letter forwaris the attached Integrated Plans Assessment (IPA) System and Commodity Reports for
, review and approval in accordance with 10 CFR Part 54, the license renewal rule. Should we apply fot License Renewal, we will reference IPA System and Commodity 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."
]2 The information in this report is accurate as of the dates of the references listed therein. P< r 10 CFR 54.21(b), an amendment or amedments will be submitted that identify any changes to ie 1
_ current licensing basis that materially affect the content of the license renewal application. j
,\1 9711200146 971114 5 ;" ^ " '
8# llll!illBilllilllil;!II.illlll!
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Document Control Desk-
- November 14,1997 Page 2
- In Reference (a), Baltimore Gas and Electric Company submitted the IPA Methodology fer review and-approval. :In Reference (b), the Nuclear Regulatory Commission (NRC) concluded that the IPA -
Methodology is acceptable for. meeting 10 CFR 54.21(a)(2) of: the license renewal rule, and- if:
_ implemented, provides reasonable assurance that all structures and components subject to an aging (managemen' review pursuant to 10 CFR 54.21(aXI) 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 conceptually sound and consistent with the intent of the license renewal rule, in Reference (c), the NRC stated that if the format and content of these reports met the requirements of
- the template developed by BGE, the NRC could begin the technical review. This report has been produced and formatted in 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.
Document Control Desk November 14,1997.
- Pcge 3 1
Should you have questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours,
-i pu V ~ ch&
STATE OF MARYLAND :
- TO WIT:
COUNTY OF CALVERT :
1, Charles II. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, rialt;more Gas and Electric Company (BGE), and that I am duly authorized to execute and file this i
= 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 statemems 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
- be reliable.
/
//vt)
,/w ha-Subgriyd and sworn before ipe, a Notary public in and for the State of hiaryland and County of U4_tt>fAf.) ,this /4 LLdayof 7teMan1W,1997.
WITNESS my lland and Notarial Seal: fAUAI >
Notary Public RJd h1y Commission Expires: _.-
N Date CilC/DLS/ dim Attachments: (1) 5.2 Chemical Volume and Control System (2) 3.5 Containment isolation Group (3) 5.13 Nuclear Steam Supply System Sampling System cc: R. S. Fleishman, Esquire 11. J. hiiller, NRC i .- J. E. Silberg, Esquire Resident inspector,NRC l: Director, Project Directorate I-1, NRC R.1. hiclean, DNR A. W. Dromerick, NRC J. II. Walter, PSC D. L, Solorio, NRC L
.$ i l
l ATTACHMENT (1) l i
APPENDIX A - TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM 4
Baltimore Gas...;u E ectric Company
' Calvert Cliffs Nuclear Power Plant November 14,1997 a . __ - - _,
ATTACllh*ENT m APPENDIX A - TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM 5.2 Chemical and Volume Control System This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing the Chemical and Volume Control System (CVCS). The CVCS was evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA)
Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire LRA.
5.2.1 Scoping System level scoping describes conceptual boundaries for plant systems and structures, develops screening tools that capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify systems and structures within the scope of license renewal Component level scoping describes the components within the boundaries of those systems and structures that contribute to the intended functions. Scop'mg to determine components subject to Aging Management Review (AMR) begins with a listing of passive intended functions and then dispositions the component types as either only associated with active functions, subject to replacement, or subject to AMR either in this report or another report.
Section 5.2.1.1 presents the results of the system level scopint,,5.2.1.2 the results of the component level scoping, and 5.2.1J the result,s,of scoping to determine components subject to an 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 scarches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.
5.2.1.1 System Level Seoping This section begins with a description of the system that includes the boundaries of the system as it was scoped. The intended functions of the system are listed and are used to define what portions of the system are within the scope oflicense ienewal.
System Descrintion/Concentual Boundaries The purpose of the CVCS is to perform the following functions: (Reference 1, Section 1.1.1; Reference 2, Secti on 9.1.1)
. Maintain reactor coolant activity at the desired level by removing corrosion and fission products;
. Inject chemicals into the Reactor Coolant System (RCS) to control coolant chemistry and minimize corrosion;
- Control the reactor coolant volume by compensating for coolant contraction or expansion from changes in reactor coolant temperature and other coolant losses or additions;
- Provide means for transferring fluids to the Radioactive Waste Processing System;
- Inject concentrated boric acid into the RCS upon a safety injection actuation signal;
- Control the reactor coolant boric acid concentration; Application for License Renewal S.21 Calvert Cliffs Nuclear Power Plant
4 ATTACIIMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM
. Provide auxiliary pressurizer spray for operator control of RCS pressure during startup and shutdown;
. Provide continuous on-line trending of reactor coolant boron concentration and fisrion product -
activity; and
. Provide a means for degasifying the RCS prior to maintenance outages and during normal operations.
The CVCS automatically adjusts the volume of water in the RCS using a signal from level instrumentation located on the pressurizer. 'Ihe system reduces the amount of Guid that must be transferred between the RCS and the CVCS during power changes by employing a programmed pressurizer level setpoint that varies with reactor power level. The CVCS also purifies and conditions the coolant by means of ion exchangers, filters, degasification, and chemical additives. [ Reference 2, Secdons 9.l.2.2 and 9.1.2.3J CVCS is composed of two subsystems: letdown and charging, and makeup. The letdown and cnarging subsystems' major components are: [P.eference 3, Table 1, Page 15 of 47]
. letdown stop valves; e regeneration heat exchanger; e excess How check valves;
. letdown How control valves; e letdown heat exchanger; e letdown backpressure control valves; e purincation filters; e ion exchangers; e volume control tank; ,
e charging pumps; e boronometer;
- process radiation monitor; and
. reactor coolant pump bleed off containment isolation valves (to the volume control tank).
The nakeup subsystem's major components are: [ Reference 3, Tai:le 1, Page 15 of 47]
. boric acid batching tank; e boric acid storage tanks; e botic acid pumps;
. reactor coolant makeup pumps; e chemical addition tank; Application for License Renewal 5.2-2 Calvert Cliffs Nuclear Power Plant
4 ATTACHMFNT (1)
APPENDIX A - TECilNICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM
. chemical addition metering tank; and
. chemical addition metering pump.
Figures 5.21 and 5.2 2 are simplified diagrams of the CVCS and are provided for information only.
Rese figures show the portion of the system within the scope oflicense renewal. (Reference 4. Table 2; References 5 through 10]
Calvert Cliffs' operating experience relative to age-related degradation of CVCS components has included occurrences of charging pump block cracking. Cha ging pump block cracking in reciprocating pumps is also a recognizeo industry problem with multiple incidents occurring industry wide. The cracks are a result of high-cycle mechanical fatigue caused by normal pump operation. He frequency of occurrences of block cracking r; CCNPP and other utilities has prompted CCNPP design improvements to the CVCS. In addition, CCNPP operating practices have been modified .o run one charging pump during normal operation (rather than two). This change will help extend pump life. (Reference 1, Attachment 6 for Group ID 041-PUMP 02, Page 4]
System Interfaces The CVCS interfaces with the fo!!owing systems: (Reference 1, Section 1.1.2; k, t.ences 5 throubh 10]
. RCS; e Nitrogen and Ilydrogen System;
. Waste Gas System;
. Reactor Coolant Waste Processing System;
. Compressed Air System;
. Safety injection System; e Containment Spray System;
. Spent Fuel Pool Cooling System;
- Process Radiation Monitoring System;
. Component Cooling (CC) System;
. Demineralized Water and Condensate System; and
. Engineered Safety Features Actuation System.
Interfaces in the major flow path c.f the CVCS are indicated on Figures 5.2-1 and 5.2 2.
Application for License Renewal 5.2-3 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CIIEMICAL AND VOLUMF CONTROL SYSTEM Component Coolmg System (WSLR - Refer to DGE LRA Secten 53) Safety Inyction System i Back h (Not WSLR) f Control Valves ByTass
, l
[-~4llent t,too.,, ___c: ._ _ _ 3 i V'
---dw_-- ,
e-r - -*- -
g- - - - -------l------ 7-------]
E uhangn _]
-__1 i } < g
, 4 4 _4
- I 1
? Letdown Process I g i g I
p - _ _ _ J Nw Nrtrogen &
t- --
- Radaman g%
- -*) i g g
" ~~l f Valve s i I Exclumger Exchanger Exchanger 21Controt / N #3W L-t L_.a.- - -*I l f Valve x, g Outside Containment L__________; - , 7 . , _,.-
...-------- ----------- -----------, , p-_-_--_4------ y i
Inside Containment l yg g y g-+ C + - - -
m , '** Waste Pmcessmg g
M.--A g 7--- 3,,,
Sysiem 4 -- X ---
Check (Not WSLR)
RC Pump Bleed v CVCS To
- e. nes Mcate Compnenu
"" T !iM(, within the scope of license renewal (WSLR) oniso:ston [' _ _ j i Sarcty valves . outlet : for the CVCS. Dashed lines indicate
.--- 2 Ser l l Val inseton
- components not WSLR for the CVCS.
,,, j W V*I "3 l 8 I Annotations are provided at interfaces L ' I n ie,, *
- + E between the CVCS and other systems to l Exchanger f ,__y -
p,c,,, ,,,,, l indicate if the other system is WSLR at the n interface pomt only.
l1 g] (Na WStR) I Aux Ch"'8"8 3
" u l % :
8 5 y, T C ~~~~"
%.n l Chargmg Pumps Safc*y Inycten System e, /
Check f
. /.g I
S, : I _ _ _
mm.Rer..oGE mA - s is, y,, , g yagy,, l Valve l l t d ,
i 8 i A FIGURE 5.2-1 5-_x_v V
i 1
Reactor Coolant System CCNPP CVCS - LETDOWN & CHARGING SUBSYSTEM (WSLR. Refer to BGE LRA Section 4 I) (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY)
Application for License Renewal 5.2-4 Calvert ClifrS Nuclear Power Plant
A'ITACIIMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM RCMU Flow Control ge- -.J
V'
Makeup sior Valve no. Dement I {*,i - - ~1 Vol '
4 - - -. - - - -D< : , A - :: . - - - - - - * - - - - - - riher * - -
- wa::r s i + -; s I t
, ,f. (4 - - - J l I Chemical ~ ~
Chemscal Wasse Gas system
^"
f/, ouoci
/p Stop Valve y4 umi Tank ma wstR) 4 ,----- 2 l
a-Acid Manual Td ,p----sl_ g-----
Bonc Acal N 7.S Batchmg '
To,Trnm 5'"P II*C**'"3bt' _I 8. Tank Refuehng Valves Valve a I
_I _Y] [Li _
Wmer Tank PX -- X - l l l Acid AW I
(%'sLR - Refer to Recirc Valse O , l C Rmre Valve BGE LRA sectum l MN :
) .
WXt ,
5 15) Do"
_ g 4 _1" g
\
nj Qj 1r " I
, g Chemical Ad6 tion %g pg pg 4_-----.------------ - ---- ----
', y %
Donc Acid C Basket Wasse Processmg Pump Y,7) O g Done Acid (R) stramer sysiem ,% g
', L_ _ (Not Wsut) s_ a 4 ,
Direct reca tiender stop i
Valve a ! y ,,
l ,,
b icas FIGURE .5.2-2 Note: Solid lines indicate components WSLR for the CVCS. Dashed lines indicate components not WSLR for CCNPP CVCS - MAKEUP SUBSYSTEM the CVCS. Annotations are provided (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY) at interfaces between the CVCs and other systems to indicate if the other system is WSLR at the interface point
! only.
t
! Application for License Renewal 5.2-5 Calvert Cliffs Nuclear Power Plant
ATTACilMENT (1) l APPENDIX A - TECIINICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM System Scaping Results The CVCS is in scope for license renewal based on 10 CFR 54.4(a). The following intended functions of the CVCS were detennined based on the requirements of 654.4(a)(1) and (2) in accordance with the CCNPP IPA Methodology Section 4.1.1: (Reference 1, Section 1.1.3; Reference 4, Table 1]
. To provide containment isolation of the CVCS during a loss-of-coolant accident (LOCA);
Note - This function also applies to station blackout (10 CFR 50.63) based on 954.4(a)(3).
- To inject concentrated boric acid into the RCS for reactivity control and RCS pressure and level control during design basis events; Note - This function also applies to pressurized thermal shock (10 CFR 50.61) and fire protection (10 CFR 50.48) based on 654.4(a)(3).
. To provide radiological release control by inlating the RCS letdown line during a LOCA;
. To maintain the pressure boundary of the CVCS (liquid and/or gas);
- To provide long-term core flush via pressurizer auxiliary spray; Note - This function also applies to pressurized thermal shock (10 CFR 50.6') based on l54.4(a)(3).
. To maintain electrical continuity and/or provide protection of the electrical system;
. To maintain mechanical operability and/or provide protection of the mechanical system;
. To restrict now to a specified value in support of a design basis event response; and
. To provide seismic integrity and/or protection for safety-related components.
The following intended functions of the CVCS were determined based on the requirements of 654.4(a)(3): [ Reference 4, Table 1]
. For post-accident monitoring - To provide information used to assess the environs and plant condition during and following an accident;
. For are protection (10 CFR 50.48)- To provide RCS pressure and inventory control to ensure safe shutdown in the event of a postulated severe fire; and
. For environmental qualification (10 CFR 50.49) - To maintaia functionality of electrical components as addressed by the Environmental Qualification Program.
5.2.1.2 Component Level Scoping Based on the intended functions listed above, the portion of the CVCS that is within the scope oflicense renewal includes all components (electrical, mechanical, and instrumentation) and their supports along the following system Cowpaths, as indicated on Figures 5.21 and 5.2 2: [ Reference 4, Table 2; References 5 through 10]
. From the RCS interface at the letdown stop valves through the regenerative heat exchanger to the letdown flow control valves. The letdown heat exchanger is also within the scope of license renewal due to its safety-related pressure boundary for the CC System, although the piping between the letdown flow control valves and the letdown heat exchanger is not within the scope oflicense renewal; e From the volume control tank outlet stop valve through the charging pumps and regenerative heat exchanger to the auxiliary spray and charging line check valves; Application for License Renewal 5.2-6 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM e From the reactor coolant pump bleed off isolation valves inside containment through the containment penetration to the isolation valve outside containment;
- From the boric acid storage tanks through the boric acid pumps to the charging pumps header and to the makeup s.op valve; and e From the boric acid storage tanks through the gravity feed valves to the charging pumps header.
All piping within the scope of license renewal for the CVCS is identified as being within the cafety-related pressure boundary as shown on the system piping and instrumentation diagrams. He piping and instrumentation diagrams also denote that all equipment within this boundary are considered safety-related pressure boundary components. Calvert Cliffs' components designated as safety-related pressure boundary are deaigned as Seismic Category 1 and are subject to the applicable loading conditions identified in UFSAR Section SA.3.2 for Seismic Category I systems and equipment design.
[ References 5 through 11]
A total of 53 device types within the CVCS were designated as within the scope of license renewal because they have at least one intended function. These device types are listed in Table 5.2-1.
[ Reference 1, Table 2.1; References 12 and 13]
Application for License Renewal 5.2-7 Calvert Cliffs Nuclear Power Plant
ATTACllMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM-TABLE 5.2-1 1
CVCS DEVICE TYPES WITillN THE SCOPE OF LICENSE RENEWAL Device Code Device Description
- CC Stainless Steel Piping (1500 psig rating)
- 11C Stainless Steel Piping (150 psig rating)
ACC Accumulator BS Basket Strainer CKV Check Valve COIL Coil CS Control Switch CV Control Valve CVOP Control Valve Operator DISC Disconnect Switch / Link FE Flow Element FIA Flow Mdicator Alarm FO Flow Orifice FT Flow Transmitter IU Fuse FY Flow Device (Relay) lilC lland Indicator Controller ils llandswitch IIV Hand Valve 11X lleat Exchanger 1/P Current / Pneumatic Device 11 Ammeter JL Power Lamp indicator LC 1.evel Controller LIA Levellndicator Alamt LIT Level Indicating Transmitter LS Level Switch LY Level Device (Relay)
M 430V Motor (Feed from Motor Control Center)
MB 480V Motor MD 125/250VDC Motor MOV Motor Operated Valve MOVOP Motor Operated Valve Operator PC Pressure Controller PCV Pressure Control Valve PDI Pressure Differential Indicator Pl Pressure Indicator -
PNL Panel PS Pressure Switch Application for License Renewal 5.2-8 Calvert Cliffs Nuclear Power Plant
. ~~ _ _ - ._. -- ~ . _ . -.
ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM TABLE 5.21 (continued)
CVCS DEVICE TYPES WITillN Tile SCOPE OF LICENSE RENEWAL Device Code Device Description -
PT Pressure Transmitter PUMP Pump' Driver Assembly QHT lleat Tracing Controller RV Relief Valve RY Relay SV Solenoid Valve TE Temperature Element TIC Temperature Indicating Controller TK Tank TS Temperature Switch U licater XL Miscellaneous Indicating Lamp ZL Position Indicating Lamp ZS Position Switch in addition, some components withm the scope of license renewal are common to many plant systems and perform the same passive functions regardless of system. These components are not included in the above table and are as follows:
. Structural supports for piping, cables, and components;
. Electrical cabling; and
. Instrun ent lines (l.c., tubing and small bore piping), tubing supports, instrument valves (e.g., equalization, vent, drain, isolation), and fittings.
5.2.1.3 Components Subject to AMR This section describes the components within the CVCS that are subject to an AMR. It begins with a listing of passive intended functions and then dispositions the device types as either only associated with active functions, subject to replacement, evaluated in other reports, evaluated in commodity reports, or remaining to be evaluated for aging management in this section.
Passive Intended Functions l
- in accordance with CCNPP IPA Methodology Section 5.1, the following CVCS functions were I- determined to be passive
- [ Reference 1, Table 3 1; Reference 14, Attachment 1]
- To maintain the pressure b undary of the CVCS (liquid and/or gas);
- To maintain electrical continuity and/or provide protection of the electrical system; e- To restrict tiow to a specified value in support of a design basis events response; and
. To provide seismic integrity and/or protection for safety-related components.
Application for License Renewal 5.2-9 Calvert Cliffs Nuclear Power Plant
NITACIIMENT (1)
APPENDIX A - TECHNICAL INFORMATiON 5.2 - CHEMICAL AND VOLUf1E CONTROL SYSTEM Device Tynes Subject to_ AMR Of the 53 device types within the scope oflicense renewal for the CVCS shown in Table 5.21:
.. Twenty eight d:vice types have only active functions. These device types are: COIL, CS, CVOP, DISC, FI A, FU, FY, lilC, ilS, I/P, II, JL, LC, LIT, LY, M, MB, MD, MOVOP, PC, QllT, RY, TIC, TS, U, XL, ZL, and ZS. [ Reference 1. Table 2 1. Table 3 2]
. None of the device types are subject to periodic replacement. [ Reference 1, Table 3 2; Reference 14, Attachment 2]
. Eight device types are evaluated in commodity AMRs. Device types FT, LIA, LS, PDI, PI, PS, and PT are evaluated in the instrument Lines Commodity Evaluation in Section 6,4 of the BGE LRA, and device type PNL is evaluated in the Electrical Panels Commodity Evaluation in Section 6.2 of the BGE LRA. Note Some liVs and some CKVs are evaluated in the Instrument Lines Commodity Evaluation and the remaining ilVs and CKVs are included in this report.
[ Reference 1, Table 3 2; Reference 14, Attachments 3,4,4A)
Of the 53 device types within the scope oflicense renewal for the CVCS, the remaining 17 device types, listed in Table 5.52, are subject to AMR and are included in the scope of this report. [ Reference 1, Table 3 2; References 12 and 13)
Some components in the CVCS are common to many plant systems and perform the same pMye ft.netion regardless of system (i.e., structural supports, electrical cabling, and instrument Fnes as discussed in Section 5.2.1.2 above). Therefore, these components are not included in the 53 CVCS device types discussed above, and they were evaluated as follows:
. Structural supports for piping, cables and components in the CVCS that are subject to AMR are evaluated for the effects of aging in the Component Supports Commodity Evaluation in Section 3.1 of the BGE LRA. This commodity evaluation, in conjunction with the Electrical Panels Commodity Evaluation discussed above, completely addresses the CVCS passive intended function,"To provide seismic integrity and/or protection of safety related components."
. Electrical cabling for components in the CVCS that are subject to AMR are evaluated for the effects of aging in the Electrical Cables Commodity Evaluation in Section 6.1 of the BGE LRA.
This commodity evaluation completely addresses the CVCS passive intended function, "To maintain electrical continuity and/or provide protection of the electrical system."
. Instrument lir;es (i.e., tubing and small bore piping), tubing supports, instrument valves (e.g., equalization, vent, drain, isolation), and fittings for components in the CVCS .that are subject to AMR are evaluated for the effects of aging in the Instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA. This commodity evaluation addresses the CVCS passive intended function,"To maintala the pressure boundary of the CVCS (liquid and/or gas)"
for instrument lines, and the associated supports, instrument valves, and fittings.
The only passive intended functions associated with the CVCS that are not completely addressed by one of the commodity evaluations d:scussed above are as follows:
. To provide containment isolation of the CVCS during a LOCA:
Application for License Renewal 5.2 10 Calvert Cliffs Nuclear Power Plant
A'ITACIIMENT (1)
APPENDIX A - TECHNICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM e To maintain the pressure boundary of the CVCS (liquid and/or gas); and
. To restrict Dow to a specified value in support of a design basis event response.
Therefore, only the 3 fur.ctions listed above for the 17 device types listed in Table 5.2 2 are addressed by the remainder of this report.
TABLE 5.2-2 CVCS DEVICE TYPES REQUIRING AhfR Device Code Device Description
- CC Stainless Steel Piping (1500 psig rating)
- llc Stainless Steel Piping (150 psig rating)
ACC Accumulator 11 S Basket Strainer CKV Check Valve CV Control Valve FE Flow Element FO Flow Orifice HV iland Valve _
llX lleat Exchanger MOV Motor Operated Valve PCV Pressure Control Valve _
PUMP Pump / Driver Assembly RV Relief Valve SV Solenoid Valve TE Temperature Element TK Tank Daltimore Gas and Electric Company may elect to replace components for which the AMR identifies further analysis or examination is needed. In accordance with the License Renewal Rule, components subject to replacement based on a qualified life or speci0ed time period would not be sub;cct to AMR.
5.2.2 Aging Management The list of potential Age-Related Degradation Mechanisms (ARDMs) identified for the CVCS components is given in Table 5.2 3, with plausible ARDMs identined by a check mark (/) in the appropriate device type column. [ Reference 1, Attachment 1, Attachment 5s] A check mark indicates that the ARDM applies to at least one component for the device type listed.
Application loi oicmse Renewal 5.2-11 Calvert Cliffs Nuclear Power Plant
. . . . - - -~ . ..
4 ATTACilMENT (1)
APPENDIX A - TECliNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM For ef0ciency in~ presenting the results of these evaluations in this report, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all components within that group. Exceptions are noted where appropriate. Table 5.2-3 also identifies the group to which each ARDM/ device type combination belongs. The following groups have been selected for the CVCS:
Group 1 - includes the device types subject to thermal fatigue.
Group 2 - includes the device types with borated water or boric acid internal environments subject to crevice corrosion, general corrosion, and pitting.
Group 3 - Includes the device types with air intunal environments subject to general carrosion.
Group 4 - Includes shell side of heat exchangers (i.e., cooling wate .8nternal environment) subject to crevice corrosion and pitting.
Group 5 - Includes the device types subject to wear.
Group 6 - Includes the device types subject to vibrational fatigue.
Group 7 - locludes the device types subject to stress corrosion cracking.
l l
l Application for License Renewal 5.2 12 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (I)
APPENDIX A - TECIINICAL INFORMATION -
5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM TABLE 5.2-3 POTENTIAL AND PLAUSIBLE ARDMs FOR TIIE CVCS i Device Types 4 .. Not Potentral ARDMs #CC- #11C ACC BS CKV CV CV cE FO HV IlX MOV PCV PUMP RV. SV TE TK Plausible Air Water pe, yg ,,,,
Cavitation Erosion x Corrosion Fatigue x Crevice Corrosion / / / / /
(2) (2) (2) (4) (2)
Dynamic Leading x Erosion Corrosion x Fatigue / / / / / / / / /
(thermal or viitational) (1) (6) (I) (1) (I,6) (1) (6) (6) (I)
Foulmg x Galvanic Corresion x General Corrosion / / / / / / / / / / / / /
(2) (2) (2) (2) (3) (2) (2) (2) (2) (3) (2) (2) (2)
Ilydrogen Damage x Intergranular Attack u MIC x Particulate Wear Erosion x Pittmg / / / / /
(2) (2) (2) (4) #2)
Radiation Damage x Rubber Degradation x ScJine Water Attack x Selective Leaching x Stress Corrosion Cracking / / / / / / /
(7) (7) (7) (7) (7) (7) (7)
Ihermal Damage ] x
, lhermal Embrittlement x
% car / /
(5) (5)
/ -indicates plausible ARDM determination g <)- indicates the group (s) in which this ARDM' device type combination is evaluated MIC = Microbiologically-induced Corrosion Application for License Renewal 5.2-13 Cahert Cliffs Nuclear Power Plant ,
ATTACHMENT (1) )
APPENDIX A TECHNICAL INFORMATION
,(2 - CHEMICAL AND VOLUME CONTROL SYSTEM 1he following is a discussion of the aging management demonstration process for each group identified almve. It is presented by group and includes a discussion of materials and environment, aging mechanisrn effects, methods to managing aging, aging management program (s), and aging management demonstration.
Group 1 (Device types subject to thermal fatigue)- Materials and Environment As shown in Table 5.2 3, Group 1 applies to device types #CC, CKV, CV (Water), HV, HX, and TE that are subject to fatigue.
Group I consists of the following CVCS components, piping, control valves, and hand valves in the RCS letdown line to the tube side of the Regenerative lleat Exchanger, the Regenerative Heat Exchanger, the Regenerative Heat Exchanger discharge temperature element, the Letdown lleat Exchanger, and the check valves at the CVCS/RCS interface downstream of the Regenerative Heat Exchanger. [ Reference 1. Attachment 1, Attachment 3s for Group lDs #CC-02, CKV-09, CKV-14, CV-04, ilV-08: liX-01, liX-02, TE-02; Reference 6]
All of the Group I components have the passive intended function to maintain pressure boundary integrity. The Group 1 :ontrol valves also have a containment isolation passive intended function.
[ Reference 1, Attachment 1]
Fatigue is plausible for the following subcomponent pa ts: [ Reference 1, Attachment 1]
= #CC - pipe, fittings, flanges, studs, nuts, and welds; e CKV - body and bonnet;
. CV - body / bonnet, stem, studs, nuts, plug, and bushing;
. IIV - body / bonnet, stem, studs, nuts, and disc - i seat;
. IlX - tubes /tubesheet, studs and nuts (for Letdown lleat Exchanger), channel / cover (for Letdown lleat Exchanger), channel and welds (for Regenerative Heat Exchanger); and
. TE - element.
The subcomponent parts for the Group 1 components that come in contact with the CVCS process fluid are primarily constructed of stainless steel. The Group i subcomponent parts external to the process stream (e.g., studs and nuts) are constructed of alloy steel, carbon steel, or stainless steel. [ Reference 1, Attachment 4s]
The internal environment for the Group 1 CVCS components is borated water. Loss of Letdown or Loss of Charging Flow transients can result in rapid increases or decreases in CVCS fluid temperature. These
' transients can result in rapid tempernture transitions between 120 F and 550'F (i.e., differential temperature of up to 430'F). [ Reference 1, Attachment 3s , Attachment 6s; Reference l$, Attachment 5; Reference 16, Section 4.4.2]
Application for License Renewal 5.2-14 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM Group 1 (Device types subject to thermal fatigue)- Aging Mechanism Effects Fatigue is the process of progressive localized structural change occurring in a material subjected to conditions that produce Ductuating stresses ar., strains at some point or points in the material. This process may culminate in cracks or complete fracture after a suf0cient number of Ductuations. 'lhe fatigue life of a component is the number of cycles of stress or strain that it experiences before fatigue failure occurs. Failures may occur at either a high or low number of cycles in response to various kinds of loads (e.g., mechanical or vibrational loads, thermal cycles, or pressure cycles). Low-cyclc fatigue involves stressing of materials, oRen into the plastic range, with the number of cycles usually being less than 10'. This mechanism is typically associatei with thermal gradients created in thick sections (e.g.,> 1") or ir, restrained members during rapid heatup or cooldown. [ Reference 1, Attachment 7s; Reference 17, Pr.ges 14,66]
Low-cycle thermal fatigue is plausible since frequent CVCS operation may produce a large number of thennal transients, with an additional number of these stress cycles expected during the period of extended operation. [ Reference 1 Attachment 6s]
This aging mechanism, if unmanaged, could eventually result in crack initiation and growth such that the Group I components may not be able to perform their pressure boundary and containment isolation functions under current licensing basis (CLB) conditions. Therefore, thermal fatigue was determined to be a plausible ARDM for which the aging effects must be managed for the Group i ..)mponents.
Eroup 1 (Device types subject to thermal fatigue) . Methods to Manage Aging Mitigation: The effects of thermal fatigue can be nitigated by reducing the number and severity of the thermal transients experienced by the system and by proper system design and material selection.
Discoverv: The effects of thermal fatigue can be managed by monitoring the total fatigue damage accumulated by critical C\ CS components as a result of all stress cycles that the components have experienced during their service lives. The accumulation of fatigue effects can be monitored by counting the number of the thermal transients and by perfonning analysis to predict the remaining life of the affected components.
Group 1 (Device types subject to thenEal fatigue)- Aging Management Program (s)
Mitigation: As discussed above, the effects of thermal fatigue can be mitigated by reducing the number and neeerity of the thermal transients experienced by the system. As a part of general operating practice, plaat ope ators minimize the length and ses erity of transitory operational cycles.
Discoverv: The CCNPP Fatigue Monitoring Program (FMP) is based on Reference 16 and has been established to monitor and track fatigue u; age of limiting components of the Nuclear Steam Supply System and the steam generators. Eleven fatigue critical locations in these systems hase been selected for monitoring of fatigue usage. These represent the most bounding locations for critical thermal transien:s. For the CVCS, the Charging inlet Nozzle has been identiDed as the most bounding location.
[ Reference 15, Sections 1,1,1.2.A,2.1.E,6.0; Reference 18]
Application for License Renewal 5.2-15 Calvert Cliffs Nuclear Power Plant
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- i ATTACilMENT (1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM The FMP utilized two mefuds to track fatigue usage. One method is to track the number of critical thermal and pressure tes' transients (i.e., cycle counting) and compare them to the number allowed in the piping design an%.is. The piping design analysis is performed assuming a particular number and wverity of .arious transients. In accordance with either American Society of Mechanical Engineers (ASME) Section 111 or American Natianal Standards Institute (ANSI) B31.7, the analysis demonstrates that the component has an acceptable design as long as the assumptions remain valid. Therefore, if the actual number and severity of transients experienced by the component remains below the number assumed in the analysis, then the component remains within its design basis.
The other method is to determine the fatigue life of a component using a calculated cumulative usage factor (CUF), which is defined as a normalized measure of total fatigue damage accumulated by a component as a reault of all stress cycles that the component has experienced during its service life. The FMP monitors actual fatigue usage. The CUF can be calculated and tracked through plant life using thermal cycle counting or stress-based analysis techniques. Both methods use actual plant operating data, ne usage factor for several locations, including the Charging Inlet Nozzle, is calculated through stress-based analysis, which is tae more rigorous method, and which provides a more realistic CUF. In accordance with the code, the component remains within its design basis for allowable fatigue life if the CUF remains less than or equal to 1.0, [ Reference 15, Sections 1.2.A,3.0.B 3.0.F; Reference 18]
The data for thermal transients is collected, recorded, and analyzed using software that evaluates input data from plant instrumentation. De software is used to analyze plant data associated wi6 real transients and to predict the number of thermal cycle trsnsients for 40 and 60 years of plant egration based on the historical records. For the CVCS, the allowable number of Loss of Charging cycles is 1400, and the allowable number of Loss of Letdown cycles is 50. The present analysis for the Loss of Letdown transient predicts that Unit I will experience 79 cycles for 40 years of plant operation and 118 cycles for 60 years, and that Unit 2 will experience 40 cycles for 40 years and 61 cycles for 60 years. Since some of the predicted values exceed the allowable number of 50 cycles for this transient, analysis is being performed to justify increasing the number of design allowable cycles. [ Reference 15, Section 3.0.F; Reference 18]
Plant parameter data is co! Meted on a periodic basis and reviewed to ensure that the data represents actual transients. Valid data is entered into the software, which coutits the critical transient cycles and calculates the CUFs. The transient data is evaluated and the CUFs are calculated on a semi annual basis, which provides a readily predictable approach to the alert value. The data is tracked in accordance with procedures that are governed by a quality assurance program that meets 10 CFR Part 50, Appendix B, criteria. A CUF of less than 1.0, and/or the number of cycles remaining below the design allowable number, aie acceptable conditions for any given component since no crack initiation would be predicted.
In order to .=tay within the design basis, corrective action is initiated well in advance of the CUF approaching 1.0 or the number of cycles approaching the design allowable, so that appropriate corrective actions can be taken in a timely and coordinated manner. [Referem 15, Sections !.2.A,5.0]
The CCNPP FMP has been inspected by the NRC, which noted that the program has been developed toward providing assurance that fatigue life usage of primary system components has not exceeded limits provided fcr in the ASME Boiler and Pressure Vessel Code for Nuclear Vessels - Section 111. In addition, the NRC noted that the FMP can be used to identify components where fatigue usage may 7pplication fv. License Renewal 5.2-16 Calvert Cliffs Nuclear Power Plant l
ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION E2 - CHEMICAL AND VOLUME CONTROL SYSTEM challenge the remaining and extended life of the components, and can provide a basis for corrective action where necessary. [ Reference 19]
Since the FMP has been initiated, no locations have reached the limit on fatigue usage and no cracking due to low-cycle fatigue has been discovered. The FMP has undergone several modifications since its inception. Stress based analysis was added to the software to calculate the CUFs for several locations due to unique thermal transients experienced and the unique geometries involved. Other modi 0 cations have been made to the FMP to reflect plant operating conditions more accurately. The plant design change process has also been modified to require notification to the Life Cycle Management Unit of any proposed changes to the critical locations being monitored.
To fully address fatigue for license renewal, CCNPP participated in an Electric Power Research Institute (FPRI) sponsored task to demonstrate the industry fatigue position. He task applied industry-developed methodologies to identify fatigue sensitive component locations that may require further evaluation or inspection for license renewal and evaluate environmental effects as necessary. The program objective included the development and justification of aging management practices for fatigue at various component locations for the renewal period. The demonstration systems were the Feedwater System, the pressurizer surge line, and the Charging / Letdown System. [ Reference 20 Page 3]
Ogneric Safety Issue 166 Generic Safety issue (GSI) 166, Adequacy of Fatigue Life of Metal Components, identifies concerns identified by the NRC, which must be evaluated as part of the license renewal process. The NRC staff concerns about fatigue for license renewal fall into five categories: (Reference 20, Page 2; Reference 21]
. The first category is the adequacy of the fatigue design basis when environmental effects are considered. This concern has been addressed by the EPRI sponsored Charging / Letdown fatigue task described above.
- The second category is the adequacy of both the number and severity of design basis transients.
This concern has been addressed by the EPRI sponsored Charging / Letdown fatigue task described above.
- The third category is the adequacy of inservice inspection requirements and procedures to detect fatigue indications. This concern does not apply to the CVCS because the CCNPP inservice inspection Program is not credited with managing fatigue for CVCS components.
. The fourth category is the adequacy of the fatigue design basis for Class I piping components designed in accordance with ANSI B31.1. This concern does not apply to the CVCS because the CVCS piping subject to fatigue is designed in accordance with ANSI B31.7, Class 11 requirements.
. The fifth category is the adequacy of actior.s to be taken when the fatigue design basis is potentially compromised. This concern is adequately addressed by the CCNPP FMP as discussed above.
Application for License Renewal 5.2-17 Calvert Cliffs Nuclear Power Plant
A1TACHMFNT (1)
APPENDIX A - TECHNICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM '
Group 1 (Device types subject to thermal fatigue)- Demonstration of Aging Management Based on the information 1, resented above, the following conclusions can be reached with respect to the Group I components subject to thermal fatigue:
. The Group I components have the passive intended functions to maintain pressure boundary integrity and to provide containment isolation under CLB conditions.
. Hermal fatigue is plausible for the Group I components which, if unmanaged, could eventually result in crack initiation and growth such that the components may not be able to perform their pressure boundary and containment isolation functions under CLB conditions.
. The CCNPP FMP monitors the fatir usage at bounding locations to ensure that the Group I components remain within their desig,n basis and includes acceptance criteria to ensure timely corrective action is taken prior to degradation that would compromise the pressure boundary and containmeni isolation functions.
Therefore, there is reasonable assurance that the effcets of thermal fatigue will be managed for the Group I components such that they will be capable of performing their pressure boundary and containment isolation functions, consistent with the CLB, during the period of extended operation.
Group 2 (Device types with borated water or boric acid internal environments subject to crevice corrosion, general corrosion, and pitting)- Materials and Environment As shown in Table ' 2-3, Group 2 applies to device types #HC, CKV, HV, and MOV that are subject to crevice corrosion, general corrosion, and pitting; and device types #CC, BS, CV (Water), ilX, PUMP, RV, and TK that are subject to general corrosion. Group 2 consists of miscellaneous CVCS components with borated water or boric acid intemal environments. [ Reference I, Attachment 1, Attachment 3s for Group ids #CC 02, #CC-03, #CC-04, #CC-05, #HC-01, #HC-02, #HC-03, #11C-05, #HC-06, BS-01, CKV-02, CKV-04. CKV-05, CKV-08, CKV-10, CKV-II, CKV 14, CV-01, CV-02, CV-05, CV-09, llV 01, ilV-02, ilV-06, IIV-07, ilV 08, IIV-II,11V-12, HV-13, HV-14, HV-IS, HV-16, HX-01, HX-02, MOV-01, MOV-02, MOV-03, MOV-04, PUMP-01, PUMP-02, RV-06, TK-01]
Nearly all of the Group 2 components have the passive intended function to maintain pressure boundary integrity. Some of the Group 2 control valves and check valves have the containment isolation passive intended function. [ Reference 1, Attachment 1]
Crevice corrosion and pitting are plausible for portions of the system that do not have hydrogen overpressure and do not have frequent circulation (i.e., flow ie stagnant). Crevice corrosion and/or pitting are plausible for the following subcomponent parts: [ Reference 1, Attachment 1]
. #11C - pipe (for some #HCs), fittings (for some #HCs), welds (for some #HCs), studs (for some
- HCs), nuts (for some #HCs), and flanges (for some #HCs);
- CKV - studs (for some CKVs), nuts (for some CKVs), and body /c;mnet (for some CKVs);
e HV body / bonnet (for some HVs), stem (for some HVs), studs (for some HVs), nuts (for some HVs), disc (for some ilVs), seat (for some HVs), and gland flange (for some HVs); and Application for License Renewa' 5.2-18 Calvert Cliffs Nuckar Power Plant
gITACHMENT l1)
. APPENDIX A - TECIINICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYS'IEM e MOV - body / bonnet (for some MOVs), stem (for some MOVs), studs (for some MOVs), and mits (for some MOVs).
General cerrosion is plausible for the following subcomponer parts: [ Reference 1, Attachment 1]
. #CC - studs and nuts; e # llc - studs and nuts; e BS - studs and nuts; e CKV - studs (for some CKVs), and nuts (for some CKVs);
e CV (Water)- studs and nuts; e IIV - studs (for some ilVs), nuts (for some IIVs), and packing flange (for some HVs);
e liX cradles, shell (for some HXs), studs (for some HXs), nuts (for some HXs), and welds (for some IIXs);
e MOV - studs and nuts;
- PUMP - bolts; e RV - studs and nuts; and
. TK studs and nuts.
The subcomponent parts for the Group 2 components that come in contact with the CVCS process fluid are primarily constructed of stainless steel. The Group 2 subcomponent parts external to the process stream (e.g., stud:, and nu:-) are primarily constructed of alloy steel or carbon steel. [ Reference 1, Attachmer> 4s]
The internal environment for the Group 2 components is borated water or boric acid, Stagnant How conditions may exist in portions of the system due to the physical geometry of the components, and due to idle operation of portions of the system. Stagnation of the flow may allow impurities in the process fluid to concentrate. [ Reference 1, Attachment 3s , Attachment 6s]
Group 2 (Device types with borated water or boric acid internal environments subject to crevice corrosion, general corrosion, and pitting)- Aging Mechanism Effects Stainless steel and carbon steel are susceptible to crevice corrosion and pitting in a stagnant, fluid environment. The aggressiveness of these corrosion mechanisms are particularly dependent on fluid chemistry conditions and oxygen levels. Crevice corrosion is intense, localized corrosion within crevices or shielded ar as. It is associated with a small volume of stagnant solution caused by holes, gasket surfaces, lap joints, crevices under bolt heads, and other mechanical joints which have a crevice geometry. The crevice must be wide enough to permit liquid entry and narrow enough to maintain stagnant conditions, typically a few thousandths of an inch or less. Crevice corrosion is closely related to pitting corrosion and can initiate pits (i.e., loss of material) in many cases. In an oxidizing environment, a crevice can set up a differential aeration cell to concentrate an acid solution within the crevice. Even in a reducing environment, alternate wetting and drying can concentrate aggressive ionic species to cause pitting and crevice corrosion. Pitting is a form of localized attack with greater corrosion rates at some
. Application for Licease Renewal 5.2-19 Calvert Cliffs Nuclear Power Plant
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ATTACHMD'TJ) 1 APPENDIX A - TECIINICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM locations than at others. This form of corrosion essentially produces holes of varying depth, liigh j
. concentrations of impurity anions such as chlorides and sulfates tend to concentrate in the pit region, ;
i giving rise to a potentially aggressive solution in this zone. Since the Group 2 components can be subject to stagnant fluid conditions that may allow impurities in the process fluid to concentrate, a potentially corrosive environment may exist. Therefore, crevice corrosion and pitting were determined to be plausible for the Group 2 components. [ Reference 1, Attachment 6s and 7s]
General corrosion is the thinning (wastage) of a metal by chemical attack (dissolutian) at the surface of the metal by an aggressive environment. The consequences of the damage are loss of load-carrying cross-sectional area. General corrosion requires an aggressive environment and materials susceptible to that environment. His ARDM is plausible for the external surfaces of the Group 2 components because susceptible materials of construction (e.g., alloy steel, carbon steel) are exposed to potential borated water or boric acid leakage from mechanicaljoints in the CVCS piping system. General corrosion is not plausible for stainless steel. Additionally, crevice corrosion and pitting can occur when crevices (e.g., under nuts and/or bolt heads) are exposed to leakage. (Reference 1, Attachment 6s and 7s]
These aging mechanisms, if unmanaged, could eventually result in a loss of material such that the Group 2 components may not be able to perform their pressure boundary and containment isolation functions under CLB conditions. Therefore, crevice corrosion, general corrosion, and pitting were determined to be plausible ARDMs for which the aging effects must be mentged for the Group 2 components.
Group 2 (Device types with borated water or boric acid internal environments subject to erevice corrosion, general corrosion, and pitting)- Methods b Manage Aging -
hiitigatiam The effects of crevice corrosion and pitting that occur due to fluid stagnation, can be mitigated by minimizing the exposure of the internal surfaces of the Group 2 components to an aggressive environment. Maintaining system chemistry conditions to minimize impurities will limit the rate and effects of degradation due to these ARDMs. [ Reference 1, Attachment 6s, Attachment 8]
The etTects of general corrosion, crevice corrosion, and pitting that occur due to leakage of borated watet or boric acid leakage, can be mitigated by performing inspections of the external surfaces of the Group 2 components for signs of leakage or boric acid residue and taking appropriate corrective action (e.g., removal of boric acid residue) prior to the onset of corrosion degradation. (Reference 1, Attachment 6s, Attachment 8]
Discoverv: The degradation of the Group 2 components that does occur can be discovered and monitored by performing visual inspections of the subcomponent parts subject to crevice corrosion, general corrosion, and pitting. Visual inspections would need to be performed on the internal surfaces of the Group 2 components to detect corrosion associated with fluid stagnation, and on the external sm faces to detect corrosion associated with leakage of borated water or boric acid.
Application for License Renewal 5.2-20 Calvert Cliffs Nuclear Power Plant
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ATTACHMENT m APPENDIX A - TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM Group 2 (Device types with berated water or borie acid internal environments subject to crevice corrosion, general corrosion, and pitting)- Aging Management Program (s)
Mitigation: Calvert Cliffs Technical Procedure CP 204, " Specification and Surveillance Primary Systems," is the program credited with managing the effects of crevice corrcaion and pitting that occur due to fluid stagnation on the internal surfaces of the Group 2 components. The program provides for monitoring and maintaining the RCS ar.J associated synems (including the CVCS) chemistry. 'The chemistry controls provided by CP 204 have been established to: minimize impurity ingress to plant systems; reduce corrosion product generation, transport, and deposition; reduce collective radiation exposure through chemistry; improve integrity and availability of plant systems; and extend component and plant life. Maintaining system chemistry conditions to minimize impurities limits the rate and effects of component degradation. Calvert Clifts Technical Procedure CP-204 is based on the Technical Specifications, BGE's interpretation of industry standards, and recommendations made by Combustion Engineering. [ Reference 1, Attachment 8; Referenct '22, Sections 1.0,2.0; Reference 23, Section 6.1.A]
The scope of CP-204 includes the following systems / components: [ Reference 22, Section 2.0, Attachments I through 13]
. Reactor Coolant (Modes 1 through 6);
. Spent Fuel Pool (Modes I through 6);
. Refueling Water Storage Tank (Modes I through 6);
e Refueling Pool (Mode 6);
- Safety Injection Tanks (Modes 1 through 6);
. liigh Pressure Safety injection Pump Discharge (Modes 1 through 6);
. Boric Acid Storage Tank (Modes I through 6);
e Reactor Coolant Waste Receiver Tank (Modes I through 6);
e Reactor Coolant Waste Evaporator Bottoms (Modes I through 6);
. Boric Acid Batching Tank (Modes I throug' 5);
e CVCS lon Exchangers (Modes I through 6); and
. Spent Fuel Pool lon Exchangers (Modes I through 6).
Calvert Cliffs Technical Procedure CP-204 describes the surveillance and specifications for monitoring the primary systems' fluid chemistry. CP-204 lists the parameters to monitor (e.g., chloride, fluoride, sulfate, oxygen, pil), the frequency of monitoring these parameters, and the acceptable value or range of values for each parameter. The primary chemistry parameters are measured at procedurally-specified frequencies (e.g., daily, weekly, monthly), and are compared against " target values" that represent a goal or pretietermined warning limit. If a target value is not met, corrective actions are taken as prescribed by the procedure, thereby ensuring timely response to chemical excursions.
[ Reference 22, Sections 3.0.C,4,6.0]
The chemistry program at CCNPP (which includes CP-204) is subject to periodic internal assessment.
Internal audits are performed to ensure that activities and procedures established to implement the Application for License Renewal 5.2 21 Calvert Cliffs Nuclear Power Plant 1
l ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM j l
requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program.' l These audits provide a comprehensive ' independent verification and evaluation of quality related actisities and procedures. Audits of selected aspects of operational phase activities are performed with a frequency commensurate with their strength of performance and safety significance, and in such a manner as to assure that an audit of all safety related functions is completed within a period of two years.
[ Reference 24, Section 1B.18] .
1 Operating experience relative to the chem 4try program at CCNPP is that it has been effective in its 8
function of minimizing corrosion and corrmion-related failures and problems.
Calvert Cliffs Technical Procedure CP-204 provides for a prompt review of primary system chemistry parameters so that steps can be taken to return chemistry parameters to normal levels and thus minimizing impurnies which will limit the rate and effects of degradation due to corrosion mechanisms.
! Reference 1, Attachment 3; Reference 22, Section 2.0]
The CCNPP "Bovic Acid Corrosion Inspection Program," (MN-3-301) is credited with mitigatmg the effects of general corrosion on the external surfaces of the Group 2 components through discovery of leakage and removal of any boric acid residue that is found. Removal of any boric acid residue from component external surfaces mitigates the corrosion effects on the Group 2 components prior to the onset of corrosion degradation. Further details on the Boric Acid Corrosion Inspection (BACl) Program are detailed in the Discovery section below. (Reference 25]
Discoverv: To verify that no significant crevice corrosion or pitting is occurring on the internal surfaces of the Group 2 components, a new plant program will be developed to provide requirements for inspections of representative components. The program is considered an Age-Related Degradation inspection (ARDI) Program as defined in the CCNPP IPA Methodology (reference Section 2.0 of the BGE LRA).
The elements of the ARDI Program willinclude:
. Determination of the examination sample size based on plausible aging effects;
. Identification of inspection locations based on plausible aging effects and consequences ofloss of component intended function;
- Determination of examination techniques (including acceptance criteria) that would be effective, considering the aging effects for which the component is examined; e Methods for interpretation of examination results; e Methods for resolution of adverse examination findings, including consideration of all design loadings required by the CL B 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 taken as part of the ARDI Program will ensure that the Group 2 components remain capable of performing their passive intended functions under all CLB conditions.
. Application for License Renewal 5.2-22 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION I 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM l
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1 Calvert Cliffs Administrative Procedure MN-3 301 provides systematic requirements to ensure that boric
- acid corrosion does not degrade the reactor coolant pressure bocadary and thereby increase the
' probability of abnormal leakage, rapidly propagating failure, or rupture. - The program establishes programmatic guidelines in response to NRC Generic Letter 88-05," Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR [pressurl:cd water reactorf Plants."
[ Reference 25, Section 1.1; Reference 26)
The scope of the program is threefold: (1) It provides examination locations where leakage may cause degradation of the primary pressure boundary by boric acid corrosion; (2) It provides examination requirements and methods for the detection of leaks; and (3) It provides the responsibilities for initiating engineering evaluations and the subsequent proposed corrective actions. [Refen nce 25, Section 1,2]
The program requires a containment walkdown following er .eactor shutoown (as soon as possible l
after attaining Hot Shutdown condition) in order to identify and quantify any leakage found in specific l areas of the Containment Building. A second walkdown is performed during heatup prior to plant startup (afler attaining normal operating pressure and temperature) if leakage was identined and corrective actions were taken. Only locations where leakage haa previously been identified need to be included in this second walkdown. The walkdowns are performed in accordance with CCNPP Administrative Procedure MN-3110," Inservice Inspection of ASME Section XI Components," and procedure MN 301. A containment walkdown is not necessary if a reactor shutdown occurs within 30 days of a previous shutdown, unless the reason for the shutdown is excessive RCS leakage. [ Reference 25, l
Section 5.1)
The program also requires examination of specific components for discovery of leakage during each refueling outage. Some of the components examined include carbon steel bolting on Class I valves and l valves in systems containing borated water that could leak onto Class I carbon steel components.
l [ Reference 25, Section 5.1.B]
l Leakage or corrosion discovered by the BACI Program require that an issue Report be initiated l according to CCNPP procedure QL-2-100," Issue Reporting and Assessments," in order to document and resolve the deficiency. The program requires that Issue Reports written due to discovery of leakage address the removal of boric acid residue and inspection of the components for corrosion. Issue Reports are initiated on discovery of corrosion and are required to address the evaluation of the component for continued service and corrective actions to prevent recurrences. [ Reference 25, Sectior.= 5.2 and 5.3]
l The BACI Program is subject to periodic internal assessmer.t. Internal audits are perforrned to ensure that activities and procedures established to implement the re.luiremems of 10 CFR Put 50, Appendix B, comply with BGE's overall Quality Assurance Program. These audits provide a comprehensive independent verincation and evaluation of quality-related activities and procedures. Audits of selected aspects of operational phase activities are perfermed with a frequency commensurate with their strength
- of performance and safety .;ignificance, and in such a manner as to assure that an audit of all safety-j related functions is completed within a period of two years. [ Reference 24, Section IB.18]
The BACI Program has evolved to account for operational experience. For example, in 1989, an inservice inspection of Unit 2 pressurizer discovered evidence of reac:or coolant leakage from approximately 28 of the 120 pressurizer heater penetrations and one upper level instrument nozzle. A Application for 1,icense Renewal 5.2-23 Calvert Cliffs Nuclear Power Plant l
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A*ITAClIMENT (1)
APPENDIX A - TECilNICAL INFORMATION 5.2 - CIIEMICAL AND VOLUME CONTROL SYSTEM
- Safety Evaluation.was performed and es a result, CCNPP modified the inspection plan for the Unit 1
_ pressurizer heater sleeves and instrument nozzles. [ Reference 25, Section 5.1.D; Reference 27]
Both CCNPP Units have had occurrences of boric acid leakage through the incore Instrumentation flange connections, in March 1993 (Unit 2), and February 1994 (Unit 1), evidence of boric acid leakage and corrosion were discovered on the Incore Instmmentation flanges and flange nuts. The BACI Program existed at the time of these events, but only required specific inspection for leaks at the beginning and end of each outage. The program'did not address leaks discosered outside of normal inspections. A:: a corrective action, the BACI Program was revised to ensure that all boric acid leaks are evaluated.
[ References 28 and 29]
The corrective actions taken as a result of the issue Reports initiated by the BACI Program will ensure that the Group 2 components remain capable of performing their passive intended functions under all CLB condi'.lons.
Group 2 (Device types with borated water or boric acid internal environments subject to crevice corre-lon, general corrosion, and pitting)- Demonstration of Aging Management Based on the information presented ai>ove, ths following conclusions can be reached with respect to the Group 2 components subject to crevice corrosian, general corrosion, and pitting:
. The Group 2 components have the passive intended functions to maintain pressure, boundary integrity and provide containment isolation under CLB conditions.
. Crevice corrosion, general corrosion, and pitting are plausible for the Group 2 components which, if unmanaged, could eventuaily result in loss of material such that the components may not be able to perform their passive intended functions under CLB conditions.
. Calvert Cliffs Technical Procedure CP 204 wil! mitigate the effects of crevice corrosion and pitting on the internal surfaces of the Group 2 components by maintaining primary system chemistry conditions such that impurities will be minimized, and contahs acceptance criteria that ensure timely correction of adverse chemistry parameters.
. The CCNPP ARDI Program will conduct inspections of representative components to detect the effects of crevice corrosion and pitting, and will contain acceptance criteria that ensure corrective actions will be taken such that there is reasonable assurance that the passive intended functions will be maintained.
. The BACI Program will mitigate the effects of general corrosion by performing inspections on the extemal surfaces of the Group 2 components for signs of leakage or boric acid residr and taking appropriate corrective action (e.g., removal of bcric acid residue) prior to the onset of corrosion degradation.
. The BACI Program will conduct inspections to detect the effects of corrosion on the external surfaces of the Group 2 components, and will ensure corrective actions will be taken such that there is reasonable assurance that the passive intended functions will be mairitained.
Application for License Renewal 5.2 24 Ca! veit Cliffs Nuclear Power Plant
ATTACilMENT (1)
APPENDIX A TECHNICAL INFORMATION 5.2 . CHEMICAL AND VOLUME CONTROL SYSTEM Therefore, there is reasonable assurance that the effects of crevice corrosion, general corrosion, and pitting will be managed for the Group 2 components such that they will be capable of performing their passivs intended functions, consistent with the CLB, during the period of extended operation.
Group i(Device types with air latere .1 environments subject to general corrosion)- Materials and Environment As sl.own in Table 5.2 3, Group 3 applies to device types CV (air) and PCV that are subject to general corrosion. Gro9p 3 consists of the pressuriwr auxi'lary spray control valve operator, the charging line to reactor coolant loop 1 A control valve operator, the chargiag line to reactor coolant loop 20 control valve operator, and the associated pressure control valves that regulate the instrument air (IA) supply for the three control valve operators. [ Reference 1, Attachment 1. Attachment 3s for GrouplDs CV 10, PCV.01; Reference 6)
All of the Group 3 components have the passive intended function to maintain pressure boundary integrity. [ Reference 1, Attachment 1]
General corrosion is plausible for the following subcorrponent parts: [ Reference I, Attachment 8):
- CV - yoke, adjusting screw; and
- PCV boltin The CV yoke is constructed of paosphated painted datile iron and the adjusting screw is constructed of rinc plated steel. The PCV bolting is carbon steel. [ Reference 1. Attachment 4s]
The internal environment for the Group 3 components is IA. The IA supply is nonnally supplied from the IA compressors and is very dry, filtered, oil free air, particle size, dew point, and oil hydrocarbons are controlled in accorunce with industry r:andards. Occasionally, tir that does not meet the same air quality standards may enter the I A System due to operation of the plant air compressors (minimal drying capacity) or the saltwater air unpressors (no dryer) which serve as backups to the IA compressors. 1 Therefore, there is a possibility that moisture may enter the IA supply, although its effect is expected to be limited since the bnkup compressors are operated on a short term basis. An ir.spection performed on the piping immediately downstream of the saltwater air compressors, where the worst case of general corrosio.: is expected, revealed only very light surface rust on the inside of each piece. After more than 20 years in operation, approximately 60% of the pipe interior contained no rust and appeared similar to the inside of new pipe. Thickness measurements showed negligible loss of wall thickness. (Reference 1 Attachments 3 and 8; Reference 2, Section 9.10; Reference 30 Attachment 8]
Group 3 (Device types with. air internal environments subject to gancral corrosion) - Aging Mect anism Effects General corrosion is the thirming (wastage) of a metal by chemical attack (dissolution) at the surface of the metal by an aggressive environment. The consequences of the damage are loss of load-carrying cross-sectional area General corrosion requires an aggressive environment and materials susceptible to
- that environment, tus ARDM is plausible for the Group 3 components because susceptible materials of construction are egosed to potentially moist air. Ilowever, the exposure of these components to
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Application for License Renewal 5.5.25 ert Clift Nuclear Power Plant
ATTACitMENT 10 APPENDIX A . TECHNICAL INFORMATION 5.2 C!!EMICAL AND VOLUME CONTROL SYSTEM ,
moisture is expected to be minimal and short term and is not expected to result in significant levels of
- degradation. [ Reference 1 Attachment 6s, Attachment 7 for Valve Operators, Attachment 8) j nis aging mechanism, if unmanaged, could eventually result in a loss of material such that the Group 3 i components may not be able to perform their pressure boundary function under CLB conditions.
Therefore, general corrosion was determined to be a plausible ARDM for which the aging effects must i be managed for the Group 3 components.
Group 3 (Device types with air internal environments subject to general corrosion) . Methods to Manatt Aging Mitigation: The effects of general corrosion for the Group 3 components can be mitigated by minimizing their exposure M an aggressive environment (i.e., minimizing moisture in the l A supply). As discussed above, the exposure of these components to moisture is expected to be minimal and short term and is not expected to result in cignificant levels of degradation. Continued maintenance of the lA System air quality to ir,dustry standards will ensure minimal component degradation. [ Reference 1, Attachments 6s, Attachment 8]
Disamy: There are no methods deemed racessary to discover general corrosion since the aging effects can be mitigated by continued maintenance of the I A System air quality.
Group 3 (Device types with air Internal environments subject to general corrosion) - Aging Management Program (s)
Calvert Cliffs initiated Preventive Maintenance Checklists IPM 10000(10001), " Check Unit 1(2) '
Instrument Air Quality," following a review of recommendations from industry operating experience.
The industry operating experience recommends maintaining the air quality within the requirements of Instrument Society of America (ISA) Standard ISA S-7.3, " Quality Standard for Instrument Air."
Standard ISA S 7.3 recommends limits for maximum particle size, dew paint temperature, and oil content. He checklists are performed in accordance with CCNPP Repetitive Tasks 10191024 l (20121022)," Check Unit 1(2) Instrument Air Quality at System Low Points." Preventive Maintenance Checkli<ts IPM 10000 (10001), check IA quality at three locations in the I A System: at the dryer outlet, at the furthest point from the dryer, and at the approximate mid-point between the other two.
Measurements of dew point and particulate count are taken every 12 weeks. According to procedure, dew point data and particulate sample results are reviewed and trended. The responsible plant personnel t
detennine if the air quality is abnormal, and initiate corrective action to return the air quality to normal and to investigate the condition of the dependent load internals as appropriate. This process ensures I A quality is maintained in accordance with industry standards for moisture (dew point). Operating experience relative to IA quality control has shown that the air normally provided is very dry and contains little particulate matter, [ References 31 and 32)
Discoverv: Since there are no methods deemed necessary to discover general corrosion, there are no j programs credited with discovery of the aging effects due to this ARDM.
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l Application for License Renewal 5.2-26 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (1)
APPENDIX A TECilNICAL INFORMATION 5.2 - CllEMICAL AND VOLUME CONTROL SYSTEM Group 3 (Device types with air laternal environments subject to general corrosion) -
Demonstration of Aging Management liased on the information presented above, the following conclusions can be reached with respect to the Group 3 components subject to general corrosion:
- The Group 3 components have the passive intended function to maintain pressure boundary integrity ut. der CLD conditions.
. General corrosion is plausible for the Group 3 components wh'ch, if unmanaged, could eventually result in loss of material such that the components may not be able to perform their pressure boundary function under CLB conditions.
. CCNPP Preventativo Maintenance Checklists IPM 10000 (10001) will wriodically monitor IA System air quality to maintain it in accordance with industry standuds.
Therefore, there is reasonable assurance that the effects of general corrosion will be managed for the Group 3 components such that they will be capable of performing their pressure boundary function, consistent with the CLil, during the period of extended operation.
Group 4 (Shell side of heat exchangers subject to crevice corrosion and pitting) - Materials and Environment As shown in Table 5.2 3, Group 4 applies to device type llX that is subject to crevice corrosion and pitting. Group 4 applies to the shell side of the Letdown lleat Exchanger. [ Reference 1. Attachment 3 fc GroupIDllX Ol]
The Letdown llcat Exchanger has the passive intended function to maintain pressure boundary integrity.
[ Reference 1, Attachment 1]
Crevice corrosion and pitting is plausible for the shell and welds. These subcomponent parts are constructed of carbon steel, [ Reference 1. Attachments 4 and $)
The internal environment on the shell s:Je of the Letdown licat Exchanger is CC System water.
Stagnant flow conditions may be present in idled sections of the CC System. [ Reference t, Attachments 3 and 6]
Group 4 (Shell side of heat exchangers subject to crevice corrosion and pitting) Aging Mechanism Effects Carbon steel is susceptible to crevice corrosion and pitting in a stagnant, Ould environment. The aggressiveness of these corrosion mechanisms are particularly dependent on Ould chemistry conditions and oxygen levels. 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, and other mechanicaljoints that have a crevice geometry. The crevice must !
be wide enough to permit liquid entry and narrow enough to maintain stagnant conditions, typically a j few thousandths of an inch or less. Crvvice corrosion is 4.losely related to pitting corrosion and can ;
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1 Application for License Renewal 5.2 27 Calvert Cliffs Nuclear Power Plant !
AIIACllMENT_(1)
APPENDIX A - TECIINICAL INFORMATION 5.2 - CilEMICAL AND VOLUME CONTROL SYSTEM initiate pits (i.e., loss of material) in many cases. In an oxidizing environment, a crevice can set up a differential aeration cell to concentrate an acid solution within the crevice. Even in a. reducing environment, alternate wetting and dnjing can concentrate aggressive ionic species to cause pitting and crevice corrosion. Pitting is a form oflocalized attack with greater corrosion rates at some locations than at others. His form of corrosion essentially produces holes of varying depth. liigh concentrations of impurity anions such as chlorides and sulfates tend to concentrate in the pit region, giving rise to a potentially aggressive solution in this zone. (Reference 1, Attachment 7 for lleat Exchangers]
Since the Group 4 components can be subject to stagnant fluid conditions which may allow impurities in the process fluid to concentrate, a potentially corrosive environment may exist. Therefore, crevice corrosion and pitting v/ere determined to be plausible for the Group 4 components. [ Reference 1 Attachment 6)
%ese aging mechanisms, if unmanaged, could eventually result in material loss such that the Group 4 components may not be able to perform their pressure boundary function under CLB conditions.
%crefore, crevice corrosion and pitting were determined to be plausib!c ARDMs for which aging effects must be managed for the Group 4 components.
Group 4 (Shell side of heat eschangers subject to erevice corrosion and pitting) - Methods to Manage Aging Mitigation: The effects of crevice corrosion and pitting for the Uroup 4 components can be mitigated by minimizing their exposure to an aggressive environment. Maintaining system chemistry conditions to minimize impurities will limit the rate and effects of degradation due to these ARDMs. [ Reference 1, Attachment 6, Attachment 8]
Discoverv: The degradation that does occur can be discovered and monitored by performing visual inspections of the heat exchanger shell and welds.
Group 4 (Shell side of heat exchangers subject to crevice corrosion and pitting) Aging Management Program (s) hiitigation: Calvert Cliffs Technical Procedure CP.206," Specification and Surveillance for Component Cooling / Service Water Systems," is the program credited with managing the effects of crevice corrosion and pitting for the Group 4 components. The piogram provides for monitoring and maintaining CC System and Service Water System chemistry to control the concentrations of oxygen, chlorides, other chemicals, and contaminants. The water is treated with hydrazine to minimize the amount of oxygen in the water, which aids in the prevention and control of most corrosive mechanisms. Continued maintenance of system water quality will ensure minimal piping or component degradation. Calvert Cliffs Technical Procedure Cp.206 is based on the Technical Specifications, BGE's interpretation of industry standards, and recommendations made by Combustion Engineering. [ Reference 33, Section 2.0; Reference 34, Attachment 8]
Calvert Cliffs Technical Procedure CP.206 describes the surveillance and specifications for monitoring the CC System fluid. CP 206 lists the parameters to monitor, the frequency of monitoring these parameters, and the target and action levels for the CC System fluid parameters The parameters Application for License Renewal 5.2 28 Calvert Clifts Nuclear Power Plant
NITAC11 MENT (1)
APPENDIX A - TECifNICAL INFORMATION 5.2 CHEMICAL AND VOLUME CONTROL SYSTEM I monitored by CP 206 are pit, hydrarine, chloride, dissolved oxygen, dissolved copper, dissolved iron, suspended solids, gamma activity, and tritium activity (normally not a radioactive system). i
[ Reference 33, Attachment 1]
These chemistry parameters are currently monitored on a frequency ranging from three times per week to once a month. A si of the parameters listed in CP 206 currently have target values that give an acceptable range or limit for the anociated parameter. Two of the parameters, pil and hydrazine, have action levels associated with them. If a target value or action level is not met, corrective actions are prescribed by the procedure, thereby ensuring timely response to chemical excursions. [ Reference 33, Section 6.0.C. !
Attachment l]
Operational experience related to CCNPP Technical Procedure CP.206 has shown no problems related to use of this procedure with respect to the CC System. In 1996, CP 206 was revised to include diss(lved iron as a chemistry parameter. Dissolved iron was added to CP.206 to act as a method to discover any unusual corrosion of the CC System components. [ Reference 35)
The chemistry program at CCNPP (which includes CP 206) is subject to periodic internal assessme..t.
Internal audits are performed to ensure that activities and procedures established to implement the requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program.
These audits provide a comprehensive independent verification and evaluation of quality related activities and procedures. Audits of selected aspects of operational phase activities are performed with a frequency commensurate with their strength of performance and safety significance, and in mch a manne: as to assure that an audit of all safety related functions is completed within a period of two years.
[ Reference 24 Section 18.18]
Calvert Cliffs Technical Procedure CP 206 provides for a prompt review of CC System chemistry parameters so that steps can be taken to return chemistry parameters to normal levels and thus minimizing impurities which will limit the rate and efTects of degradation due to corrosion mechanisms.
[ Reference 1, Attachment 8; Reference 33, Section 6.0.C]
DiscIy: To verify that no significant crevice corrosion or pitting is occurring for the Group 4 components, a new plant program will be developed to provide requirements for inspections of representative components. The program is considered an ARDI Program as defined in the CCNPP IPA Methodology (reference Section 2.0 of the BGE I".A). He program details are discussed above in the Aging Management Program section for Group 2.
Group 4 (Shell side of heat eschangers subject to crevice corrosion and pitting) . Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the Group 4 components subject to crevice corrosion and pitting:
. The Group 4 components have the passive intended function to maintain pressure boundary integrity under Cl.B conditions, Application for License Renewal 5.2 29 Calvert Cliffs Nuclear Power Plant
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AITACllMENT f1)
APPENDIX A TECHNICAL INFORMATION ,
5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM
_m c Cnivks wosion and pitting are plausible for the Group 4 components which, if unmanaged, ,
cw4 uentually result in loss of material such that the components may not be able to perform
, usu pressure boundary function under CLB conditions, f
Calvert Cliffs Technical Procedure CP 206 will mitigate the effects of crevice corrosion and pitting by maintaining CC System chemistry conditions such that impurities will be minimized, and contains acceptance criteria that ensure timely correction of adverse chemistry parameters.
e '!he CCNPP ARDI Program will conduct inspections of representative components to detect the effects of crevice corrosion and pitting and will contain acceptance criteria that ensure corrective ,
actions will be taken such that there is reasonable assurance that the pressure boundary function will be maintained.
Therefore, there is reasonable assurance that the clTects of crevice corrosion and pitting will be managed for the Group 4 components such that they will be capable of performing their pressure boundary function, consistent with the CLB, during the period of extended operation.
Group 5 (Devlee types subject to wear)- Materials and Environment As shown in Table 5.2 3, Group 5 applies to device types CKV and CV (Water) that are subject to wear.
Group 5 consists of check valves and control valves whose internals are required to maintain a pressure boundary at a safety related/non safety related interface or that have a containment isolation function.
[ Reference 1, Attachment 1, Attachment 3s and 4s for Group ids CKV-03, CKV-07, CKV.10, CKV 11, CV-02, CV 07, CV 09]
Wear is plausible for the following subcomponent parts: [ Reference 1. Attachment 1]
e CKV disc and spring (for some CKVs), and internals (for some CKVs); and e CV . disc / cage (for some CVs), spindle and seat ring (for some CVs), and plug / cage (for some CVs).
The subcomponent parts that are subject to wear are constructed of stainless steel or stellite.
[ Reference 1, Attachment 4s]
The internal environment for the Group 5 components is borated water. [ Reference 1, Attachment 3s]
Group 5 (Devke types subject to wear)- Aging Mechanism Effects The Group 5 valve internals are subject to wear due to the relative motion between the internal subcomponent parts during normal valve operation. Wear is dependent on frequency of valve operation and may cause the leak tightness of the valve to decrease with time. [ Reference 1. Attachment 6s, Attachment 7 for Valves)
-This aging mechanism, if unmanaged, could eventually result in a loss of leak tightness such that the Group 5 components may not be able to perform their pressure boundary and containment isolation function, under CLB conditions. Therefore, wear was determined to be a plausible ARDM for which the
, aging efTects must be managed fcr the Group 5 components.
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l Application for License Renewal 5.2 30 Calvert Cliffs Nuclear Power Plant I
l I . _- - . _ . . _- -. --
. l ATTACllMENT (1)
APPENDIX A TECHNICAL INFORMATION 5.2 - CHEMICAI, AND VOLUME CONTROL SYSTEM Group 5 (Device types subject to wear). Methods to Manage Aging Mitigallon: Since the wear of the valve internal subcomponent puts is due to valve operation, decreased use of the valve would slow the degradation of the valve leak tightness. Proper material selection for the valve internal parts would also slow the effects of wear.
Dixomy: for the valves required to maintain a pressure boundary at a safety related/non safety related interface, the degradation that does occur due to wear can be discovered and monitored through visual inspections of the valve internals. For the valves that have a containment isolation function, the effects of wear can be managed by discovery ofleakage during periodic leak rate testing.
Group 5 (Device types subject to wear) . Aging Management Program (s)
Mitigation: As discussed above, the effects of wear can be mitigated by decreased valve operation and by proper material selection. Decreased valve operation is not feasible from a plant operations standpoint. Therefore, it is concluded that there are no additional specific means deemed necessary to mitigate the effects of wear (in addition to proper material selection) because the inspection activities discussed in the Discovery section below are deemed adequate for effectively managing wear for the Group 5 components.
D k o m y: To verify that no significant wear is occurring for the Group 5 valves required to maintain a pressure boundary at a safety related/non safety related interface, a new plant program will be developed to provide requirements for inspections of representative components. The program is considered an ARDI Program as defined in the CCNPP !PA Methodology (reference Section 2.0 of the DGE LRA).
The program details are discussed above in the Aging Management Program section for Group 2.
[ Reference 1 for Group ids CKV-07, CKV-10, CV 07, CV 09]
The Group 5 valves that have a containment isolation function are associated with containment penetrations 1C and 211, which are periodically tested as pan of the CCNPP Local Leak Rate Test t
(LLRT) Program. [ Reference 1. Attachment 1, Attachment 3s for Group ids CKV 03, CKV ll, CV 02; Reference 2, Figure 5 10, Sheets 2 and 4; References 36 through 39]
The CCNPP LLRT Program is part of the overall CCNPP Containment Leakage Rate Testing Program.
The CCNPP Containment Leakage Rate Testing Program was established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors," Option it Appendix J specifies containment leakage testing requirements, including the types of tests required, frequency of testing, test methods, test pressures, acceptance criteria, and reporting requirements. Containment leakage testing requirements include performance of Integrated Leakage Rate Tests, also known as Type A tests, and LLRTs, also known as Type D and C tests, Type A tests measure the overall leakage rate of the containment. Type 11 tests are intended to detect leakage paths and measure leakage for l certain containment penetrations (e.g.,airlocks, flanges, and electrical penetrations). Type C tests are intended to measure containment isolation valve leakage rates. [ Reference 40, Section 6.5.6; References 4I and 42]
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. Application for License Renewal 5.2 31 Calvert Cliffs Nuclear Power Plant
I 1
ATTACllMENT m APPENDIX A TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM 1he CCNPP LLRT Program is based on the requirements of CCNPP Technical Specifications 3.6.1.2,4.6.1.2, and 6.5.6. The scope of the program includes Type B and C testing of containment penetrations. The LLRT is performed on a performance based testing schedule in accordance with Option D of 10 CFR Part 50, Appendix J, as implemented by CCNPP Technical Specifications. [ References 40,41, and 42]
Local leak rate testing presently includes the following procedural steps:
- Leak rate monitoring test equipment is connected to the appropriate test point
- The test volume is pressurized to the LLRT Program test gressure, which is conservative with respect to the 10 CFR Part 50, Appendix J, test pressure requirements. Appendix J requires l testing at a pressure "P,." which is the peak calculated containment internal pressure related to the ,
design basis accident.
- Leak rate, pressure, and temperature are monitored at the frequency specified by the LLRT procedure and the results are recorded.
- The maximum indicated leak rate is compared against administrative limits that are more restrictive than the maximum allowable leakage limits.
. "As found" leakage equal to or greater than the administrative limit, but less than the maximum allowable limit, is evaluated to determine if further testing is required and/or if corrective maintenance is to be performed.
- For "as found" leakage that exceeds the maximum allowable limit, plant personnel determine if Technical Specification Limiting Condition for Operation 3.6.l.2.b has been exceeded. Technical Specification 3.6.1.2.b contains the maximum allowable combined leakage for all penetrations and valves subject to the Type H and C tests. Corrective action is taken as required to restore the leakage rates to within the appropriate acceptance criteria.
e if any maintenance is required on a containment isolation valve that changes the closing characteristic of the valve, an "as left" test must be performed on the penetration to ensure leakage rates are acceptable.
The corrective actions taken as part of the LLRT Program will ensure that the Group 5 valves that have a containment isolation function will remain capable of performing their intended function under all CLB conditions during the period of extended operation.
Group 5 (Device types subject to wear)- Demonstration of Aging Management Based on the infonnation presented above, the following conclusions can be reached with respect to the Group 5 components subject to wear:
. The Group 5 components have the passive intended functions to maintain the pressure boundary and to provide containment isolation under CLB conditions, e Wear is plausible for the Group 5 valve internals which, if unmanaged, could eventually result in a loss of leak tightness such that the Group 5 components may not be able to perfonn their pressure boundary and containment isolation functions under CLB conditions.
Application for License Renewal 5.2 32 Calvert Cliffs Nuclear Power Plant
ATTACllMFNT f1)
APPENDIX A TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM e The CCNPP ARDI Program will conduct inspections of representative components to detect the effects of wear and will contain acceptance criteria that ensure corrective actions will be taken such that there is reasonable assurance that the pressure boundary function will be maintained.
- The CCNPP LLRT Program performs periodic testing to detect leakage, whis may be a result of wear on the valve internals, and contains acceptance criteria that ensure corrective actions will be taken such that there is a reasonable assurance that the containment isolation function will be maintained.
Therefore, there is reasonable assurance that the effects of weat will be managed for the Group $
components such that they will be capable of performing their pressure boundary and containment isolation functions, consistent with the CLil, during the period of extended operation.
Group 6 (Device types subject to vibrational fatigue) . Materials and Environment As shown in Table $.2 3, Group 6 applies to device types #11C, ilV, PUMP, and RV that are subject to fatigue. [ Reference 1, Attachment 1]
Group 6 consists of the following CVCS components: Charging Pumps, and the piping, hand valves, and relief valves between the Charging Pumps' suction stabiliter and the Charging Pumps' discharge desurger. These components are subjected to significant vibrational transients due to normal operation of the Charging Pumps. [ Reference 1 Attachments 3,5, and 6 for Group ids #11C 03, ilV-II,11W12.
PUMP 02, R%04; Reference 6)
All of the Group 6 components have the passive intended function to maintain pressure boundary integrity. [ Reference I, Attachment 1)
Fatigue is plausible for all of the above components due to vibrational fransients. Calvert Cliffs Units I and 2 hase experienced cases of fatigue failures in CVCS piping (in the late 1970s) that were attributed to vibrational loads imposed by operation of the Charging Pumps. The piping failures were typically small cracks in various welds, mostly in the Charging Pump suction lines. The cracks sometimes caused leaks in the CVCS piping, in response to these vibrational fatigue occurrences, BGE performed piping design modifications to reduce vibration and improve the CVCS reliability. [ Reference 1, Attachment 1; Reference 43]
Fatigue is plausible for the following subcomponent parts: [ Reference 1. Attachment 1. Attachment 5 for RWO4)
. # llc pipe, Attings, Danges, studs, nuts, and welds; e llV body / bonnet, stem, studs, nuts, disc and seat (for some llVs), packing Dange (for some llVs);
- PUMP - block and bolts; and e RV - case, cylinder, adjusting bolt, spindle, disc, and spring.
The subcomponent parts for the Group 6 components that come in contact with the CVCS process Huid are primarily constructed of stainless steel. The Group 6 subcomponent parts external to the process
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Application for License Renewal 5.2 33 Calvert Cliffs Nuclear Power Plant
i ATTACitMENT m APPENDIX A TECHNICAL INFORMATION 5.2 CHEMICAL AND YOLUME CONTROL SYSTEM j stream (e.g., studs and nuts) are primarily constructed of alloy steel or carbon steel. The internal environment for the Group 6 CVCS components is horated water. [ Reference 1, Attachment 3s and 4s)
Group 6 (Device types subject to vibrational fatigue) . Aging Mechanism Effects Fatigue is the process of progressive locallred structural change occurring in a material subjected to conditions that produce Ductuating stresses and strains at some point or points in the material. This :
process may culminate in cracks or complete fracture aller a sufUclent number of Ductuations. The fatigue life of a component is the number of cycles of stress or strain that it experiences before fatigue failure occurs. Failures may occur at either a high or low number of cycles in response to various kinds of loads (e.g., mechanical or vibrational loads, thermal cycles, or pressure cycles). liigh cycle fatigue failure occurs when the component cyclic stresses (including modifying factors such as stress concentrations, surface conditions, and plating) exceed the material fatigue strength for the number of cycles. [ Reference 1, Attachment 7s; Reference 17 Pages 14,66]
liigh cycle vibrational fatigue is plausible for all of the Group 6 components since they are subject to vibrational transients due to normal operation of the Charging Pumps [ Reference 1, Attachment 6s]
'ihis aging mechanism, if unmanaged, could eventually result in crack initiation and growth such that the Group 6 components may not be able to perform their pressure boundary function under CLB conditions.
Therefore, vibrational fatigue was determined to be a plausible ARDM for which the aging effects must be managed for the Group 6 components. ,
Group 6 (Device types subject to vibrational fatigue) . Methods to Manage Aging hittipation* The efTects of vibrational fatigue can be mitigated by reducing the severity of the vibrational transients experienced by the system by proper system design and material seleulon.
Dissmcry: The degradation that does occur due to vibrational fatigue can be discovered and monitcred through visual inspections of the internal and external surfaces of the Group 6 components.
, Group 6 (Device types subject to vibrational fatigue) . Aging Management Program (s)
Mitigation: As discussed above, the effects of vibrational fatigue can be mulgated by reducing the sewrity of the vibrational transients experienced by the system by proper system design and material sel~ tion. Design modifications were made to the CVCS to address vibrational fatigue problems experienced in the late 1970s (as described in the Materials and Environment section above). Therefore, there are no additional specific means deemed necessary to mitigate the effects of vibrational fatigue (in addition to proper system design and material selection) because the inspection activities discussed in the Discovery below are deemed adequate for effectively managing vibrational fatigue for the Group 6 components, Discoverv: To verify that no significant vibrational fatigue is occurring for the Group 6 components, a new plant program will be developed to provide requirements for inspections of representative components. The program is considered an ARDI Program as defined in the CCNPP IPA Methodology l
Application for License Renewal 5.2 34 Calvert Cliffs Nuclear Power Plant
4 ATTACllMENT (1)
APPENI)IX A TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM r
(reference Section 2.0 of the BGE LRA). 'lhe program details are discussed above in the Aging l Management Program section for Group 2. e Group 6 (Device types subject to vibrational fatigue)- Demonstration of Aging Management Hased on the information presented above, the following conclusions can be reached with respect to the r Group 6 components subject to vibrational fatigue:
e - The Group 6 components have the passive intended function to maintain pressure boundary integrity under CLB conditions.
. Vihrational fatigue is plausible for the Group 6 components uhich, if unmanaged, could eventually result in crack initiation and growth such that the components may not be able to '
perform their pressure boundary function under CLB conditions.
e %e CCNPp ARDI Program will conduct inspections of representative components to detect the .
effects of vibrational fatigue, and will contain acceptance criteria that ensure corrective actions will be taken such that there is reasonable assurance that the pressure boundary function will be maintained.
Therefore, there is reasonable assurance that the effects of vibrational fatigue will be managed for the Group 6 components such that they will be capable of perfonning their pressure boundary function, consistent with the CLB, during the period of extended operation.
Group 7 (Devlee types subject to stress corrosion cracking)- Materials and Environment As shown in Table 5.2 3, Group 7 applies to device types # llc, CKV, CV (Water), FE, ilV, MOV, and RV that are subject to stress corrosion cracking. Group 7 consists of piping, valves, and flow elements that contair, boric acid and have heat tracing. [ Reference 1, Attachment 1, Attachment 3s and Attachment 6s for GrouplDs #11C 02, #11C 05, #11C 06, CKV 02, CKV 06, CKV-07. CKV 08, CKV 10, CV 01, CV 09, FE 02,llV 06,ilV-07,ilV-14, MOV 03, MOV 04, RV 03, RV 05, RV 06]
All the Group 7 components have the passive intended function to maintain pressure boundary integrity.
(Reference 1. Attachment 1]
Stress corrosion cracking is plausible for the following subcomponent parts: [ Reference 1 Attachment 1]
- ilC pipe, fittings, welds, and flanges; CKV body / bonnet; rY (Water)- body / bonnet; FE - element; ilV body / bonnet; MOV - body / bonnet; and RV - case and cylinder (for some RVs); nonle and body / bonnet (for some RVs).
Application for License Renewal 5.2 35 Calvert Clifts Nuclear Power Plant
I A'ITACilMFNT (1)
L APPENDIX A TECHNICAL INFORMATION 5.2 CHEMICAL AND VOLUME CONTROL SYSTEM ne subcomponent parts that are subject to stress corrosion cracking are constructed of stainless steel.
[ Reference 1, Attachment 4s)
Electrical heat tracing is instaltd on the Group 7 components. He heat tracing is needed to maintain the boric acid above the saturation temperature. %c heat tracing is designed to maintain 160*F; however, the operating temperature may be set lower. [ Reference 2, Section 6.10]
Group 7 (Device types shbject to stress corrosion cracking) Aging Mechanism Effects Stress corrosion cracking is selective corrosive attack along or across material grain boundaries. Stress corrosion cracking requires applied or residual tensile stress, susceptible materials (such as austenitic stainless steels, alloy 600, alloy X750, SAE 4340, and ASTM A289), and oxygen and/or ionic species (e.g., chlorides / sulfates). Common sources of residual stress include thermal processing Lad stress risers created during surface finishing, fabrication, or assembly. The heat input during welding can result in a localized sensitized region that is susceptible to stress corrosion cracking. Transgranular stress corrosion cracking (i.e., across the material grains) may be a concern in low alloy and stainless steel if aggressive chemical species (caustics, halogens, sulfates, especially if coupled with the presence of oxygen) are present. [ Reference 1. Attachment 7s for pipe, Valves, and Elements)
Stress corrosion cracking is plausible for the Group 7 components because external surfaces of the stainless steel subcomponents are in contact with halogens and are subject to high temperatures due to the heat tracing. He halogens are found in the adhesives used to adhere the heat tracing to the component exterior surfaces. %e stress corrosion cracking could lead to minor breeches of the pressure boundary and subsequent boric acid leakage. [ Reference 1. Attachment 6s]
Operating experience at CCNPP includes at least one case of externally initiated stress corrosion cracking in CVCS heat traced piping . Transgranular cracking was observed on the boric acid recirculation line. Analysis of the cracked piping determined that the cause of the failure was due to the presence of chlorides under high temperature conditions due to the heat tracing. The potential chloride sources were determined to be from the heat tracing adhesive, insulation, or residual chloride contamination from construction. Nuclear industry operating experience has also identified heat tracing '
as contributing to cracking of stainless steel piping in the presence of chlorides. [ Reference 44]
This aging mechanism, if unmanaged, could eventually result in cracking such that the Group 7 components may not be able to perform their presst.re boundary function under CLB conditions.
Therefore, stress corrosion cracking was determined to be a plausible ARDM for which the aging effects must be managed for the Group 7 components.
Group 7 (Device types subject to stress cerrosion cracking). Methods to Manage Aging Mitigntis: Stress corrosion cracking can be mitigated by minimizing the exposure of the stainless steel subcomponents to an aggressive environment for corrosion. Removing the corrosive adhesive associated with the heat tracing (and choosing another method to adhere the heat tracing) will mitigate the effects of stress corrosion cracking.
Application for License Renewal 5.2 36 Calvert Cliffs Nuclear Power Plant
0 ATTACHMENT (1)
APPENI)IX A TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTRh SYSTEM Discoverv: Since the effects of stress corrosion cracking can be mitigated as described above, there are no additional methods deemed necessary to manage this ARDM. .
Group 7 (Device types subject to stress corrosion cracking) . Aging Management Program (s)
Mitigation: A plant modification was initiated in 1991 to replace the original heat tracing in the CVCS.
The existing hest tracing adhesive will be removed, eliminatng the halogen impurities that promote ,
stress corrosion cracking. The new heat tracing will be installed with an adhesive that contains no halogen impurities. Portions of the original heat tracing have already been replaced. The modification will be completely implemented prior to the start of the license renewal period. Implementation of this modification will render this ARDM as no longer plausible. (Reference 1, Attachment 6s, Attachment 8]
Discoverv: As discussed in the Mitigation section above, since the plant modification will render stress corrosion cracking as non plausible after the modification is completely implemented, there are no discovery programs deemed ntcessary to manage this ARDM.
'Gro sp 7 (Device types subject to stress corrosion cracking) . Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the Group 7 components subject to stress corrosion cracking:
- The Group 7 components have the passive intended function to maintain pressure boundary integrity under CLB conditions.
. Stress corrosion cracking is plausible for the Group 7 components which if unmanaged, could eventually result in cracking such that the Group 7 components may not be able to perform their pressure boundary function under CLB conditions.
- A plant modification was initiated to replace the original heat tracing and removes the potentially ,
corrosive adhesive from the Group 7 components. The modification will be completely implemented prior to the start of the license renewal period and will render this ARDM as no longer plausible.
Therefore, there is reasonable assurance that the efTects of stress corrosion cracking will be managed for the Group 7 components such that they will be capable of performing their pressure boundary function, consistent with the CLil, du ing the period of extended operation.
5.2.3 Conclusion The programs discussed for the CVCS are listed in the following table. These programs are administratively controlled by a formal review and approval process. As demonstrated above, these programs will manage the agli:g mechanisms and their effects such that the intended functions of the CVCS components will be maintained during the period of extended operation consistent with the CLB under all design loading conditions, ne analysis / assessment, corrective action, and confinaation/ documentation process for license renewal is in accordance with QL-2, "Coirective Actions Program." QL-2 is pursuant to 10 CFR Part 50, Appendix B, and covers all structur% and components subject to AMR.
Application for License Renewal 5.2-37 Calvert Clifts Nuclear Power Plant
=. . - . _ _ __ _ _ _ _ _ _ __ _ . _ _ _ _ _ _
l e
l A*ITACilMENT (1) j APPENDIX A - TECHNICAL INFORMATION 5.2 - CHEMICAL AND VOLUME CONTROL SYSTEM TAHLE 5.2-4 1IRT OF AGING MANAGEMENT PROGRAMS FOR Tile CVCS i Program Credited For Existing CCNPP Administrative Procedure EN 1300, Monitoring and management of the effects
" Implementation of Fatigue Monitoring" of thermal fatigue for the Group I components. .
Existing CCNPP Administrative Procedure MN.3 301, Mitigation, detection, and management of i
" Boric Acid Corrosion inspection Program" the effects of crevice corrosion, general corrosion, and pitting for the Group 2 components.
Existing CCNPP Preventative Maintenance Checklists Mitigation of the effects of general IPM 10000 (10001)," Check Unit 1(2) corrosion for the Group 3 components, instrument Air Quality" Existing CCNPP lechnical Procedure CP 206, Mitigation of the effects of crevice
" Specifications and Surveillance for corrosion and pitting for the Group 4 Component Cooling / Service Water Systems" components.
Existing CCNPP Technical Procedure CP 204, Mitigation of ti,c effects of crevice
" Specification and Surveillance Primary corrosion and pitting for 'the Group 2 Systems" components.
Existing LLRT Program Detection and management of leakage that CCNPP Surveillance Test Procedure could be the result of wear for the Group 5 M 571 A 1(2)," Local Leak Rate Test, components.
Penetrations l A,10, IC" CCNPP Surveillance Test Procedure M 571C-1(2)," Local Leak Rate Test, Penetrations 2A,2B" -
Existing Plant Modification Mitigation of the effects of stress corrosion cracking for the Group 7 components.
New ARDI Program Detection and management of the efTects of crevice corrosion end pitting for the Group 2 and Group 4 components.
Detection and management of the effects of vibrational fatigue for the Group 6 components.
Detection and management of the effects of wear for the Group 5 components.
Application for License Renewal 5.2-38 Calvert Clifts Nuclear Power Plant
, i A'[TActiMI:NT (1)
APPENDIX A TECHNICAL INFORMATION 5.2 CHEMICAL AND YOLUME CONTHOL SYSTEM i 5.2.4 References j
- 1. "CCNPP Aging Management Review Report for the Chemical and Volume Control System (System 04l)," Revision 2, March 19,1997
- 2. CCNPP Updated Final Safety Analysis Report, Revision 20
- 3. "CCNPP System Level Scoping Results " Revision 4, April 6,1995
- 4. "CCNPP Component Level Scoping Results for the Chemical and Volume Control System (System 04l)," Revision 2, April 9,1996
- 5. CCNPP Drawing 60730S11000)," Chemical and Volume Cantrol System," Revision 67
- 6. CCNPP Drawing 60730S110002," Chemical and Volume Control System," Revision 49
- 7. CCNPP Drawing 60730S110003," Chemical and Volume Control System," Revision 33
- 8. CCNPP Drawing 62730Sil0001," Chemical and Volume Control System," Revision 61
- 9. CCNPP Drawing 62730S110002, " Chemical and Volume Control System," Revision 39
- 10. CCNPP Drawing 62730S110003," Chemical and Volume Control System," Revision 32
Jevision I, August 30,1996
- 12. CCNPP Drawing 92767Sil llc 1,"M.600 Piping Class Sheets," Revision 55
- 13. CCNPP Drawing 93767Sil CC 1,"M 600 Piping Class Sheets," Revision 48
- 14. "CCNPP Pre Evaluation Results for the Chemical and Volume Control System (System 041),"
Revision 2, October 17,1996
- 15. CCNPP Administrative Procedure EN 1300, " Implementation of Fatigue Monitoring,"
Revision 0, February 28,1996
- 16. Combustion Engineering Owners Group Task 571, Report No. CE NPSD 634 P, " Fatigue Monitoring Program for Calvert Cliffs Nuclear Power Plants Units 1 and 2," April 1992
- 17. " Metal Fatigue in Engineering," 11. O. Fuchs and R.1. Stephens, John Wiley & Sons, Copyright 1980
- 18. CCNPP " Fatigue Monitoring Report for 1996," Final Report for 1996 generated by CCNPP Administrative Procedure EN 1300," Implementation of Fatigue Monitoring," March 25,1997
- 19. Letter from Mr. J. P. Durr (NRC) to Mr. C. Stoiber (sic) (DGE), dated February 11,1993,
" Inspection Report Nos. 50-317/92 32 and 50-318/92 32"
- 20. DGE Procurement Specification 6422284S," Technical Services to Evaluate Thermal Fatigue EITects on Calvert Cliffs Nuclear Power Plant Systems Requiring Aging Management Review for License Renewal," Revision 0, July 29,1996
- 21. NUREG 0933, Generic Safety issue 166," Adequacy of Fatigue Life of Metal Components,"
Revision I, June 30,1995
- 22. CCNPP Technical Procedure CP 204, "Speci0 cation and Surveillance Primary Systems,"
Revision 7, March 11,1997 Application for License Renewal 5.2 39 Calvert Cliffs Nue: car Power Plant
t NITACllMENT (1) t APPENDIX A TECilNICAL INFORMATION 5.2 . CilEMICAL AND VOLUME CONTROL SYSTEM
- 23. CCNPP Nuclear Program Directive, Cil 1, " Chemistry Program," Revision I, '
December 13,1995
- 24. IlGE " Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant," Revision 48, March 28,1997 ;
2$. CCNPP Administrative Procedure MN 3 301, " Boric Acid Corrosian Inspection Program."
Revision I, December 15,1994
- 26. NRC Ocneric Letter 88 05,"Horic Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," March 17,1988
- 27. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated July 29,1994,
" Pressurizer lleater Sleeves and instrument Nontes inspection Plan Modification / lleater Sleeves Nickel Plating"
- 2 8. Letter from Mr. C. II. Cruse (DGE) to NRC Document Control Desk, dated July 29, 1994,
" Licensee Event Report 94 004, Revision 1. Excessive Corrosion of incore Instrumentation Flange Components" ;
- 29. Letter from Mr. L. T. Doerdeln (NRC) to Mr. R. E. D(oton (BGE), dated October 16,199$,
"NRC Region i Inspection Report Nos. 504317/95 08 and 50 318/95 08"
- 30. CCNPP " Aging Management Review Report for the Compressed Air System," Revision 4
- August 11,1997
- 31. CCNPP Nuclels Database, Preventative Maintenance Checklists IPM 10000 (10001), " Check Unit I (2) Instrument Air Quality"
- 32. CCNPP Nucleis Database, Repetitive Tasks 10191024 (20191022), " Check Unit 1 (2)
Instrument Air Quality at Selected System Low Points"
- 33. CCNPP Technical Procedure CP 206, "Speci0 cations and Surveillance Component Cooling / Service Water System," Revision 3, November 4,1996
- 34. "CCNPP Aging Management Review Report for the Component Cooling System," Revision 1, November 7,1996
- 35. CCNPP 1996 Component Cooling and Service Water System Assessment, February 26,1997
- 36. CCNPP Unit 1 Surveillance Test Procedure M 571 A 1, " Local Leak Rate Test, Penetrations 1 A,111,1C," Revision 0, May 16,1991 37, CCNPP Unit 2 Surveillance Test Procedure M 571 A 2, " Local Leak Rate Test, Penetrations I A,111, IC," Revision 0, October 17,1991
- 38. CCNPP Unit 1 Surveillance Test Procedure M 571C 1, " Local Leak Rate Test, Penetrations 2A,2B," Revision 0, May 17,1991
~ 39. CCNPP Unit 2 Surveillance Test Procedure M 571C 2, " Local Leak Rate Test, Penetrations 2A,2D," Revision 0, October 17,1991
- 40. 1, citer from Mr. A. W. Dromerick (NRC) to Mr. C.11. Cruse (DGE), Ated March 7, IF
" Issuance of Amendments for CCNPP Unit No.1 (TAC No M96350) and Unit No. 2 (TAC No, M96351)"(Amendments 221/197)
Application for License Renewal 5.2-40 Calvert Cliffs Nuclear Power Plant
I ATTACHMENT f1)
APPENDIX A TEC11NICAL INFORMATION 5.2 CilEMICAL AND VOLUME CONTROL SYSTEM
- 41. 10 CFR Part 50, Appendix J. " Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" ,
- 42. letter from Mr. C. 11. Cruse (BGE) to NRC Document Control Desk, dated I November 26,1996," License Amendment Request; Adoption of 10 CFR Part 50, Appendix J.
Option !! for Types 11 and C Testing"
- 43. Letter from Mr. R. W. Reid (NRC) to Mr. A. E. Lundvall, Jr. (BGE) dated October 18,1980,
" Safety Evaluation Regarding Charging Pump Pipe Vibration"
- 44. NRC IE Information Notice No. 8544, "licat Tracing Contribute to Corrosion Failure of
- Stainless Steel Piping," April 30,1985 i Application for License Renewal- 5.2-41 Calvert Cliffs Nuclear Power Plant 1
- l 0
ATTACilMENT 12) 1 APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP ,
l l
l llaltimore Gas and Electric Comnany ,
1 - Calvert Cliffs NucIcar Power Plant ;
l November 14,1997 l
I NITACllMENT (D APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP 5.5 Containment Isolation Group I
%is is a section of the llaltimore Gas and Electric Company (IlGE) License Renewal Apphcation (LRA) addressing the Containment isolation (Cl) Group. He Cl Grcup was evaluated in accordance with the Calvert Cliff Nuclear Power Plant (CCNPP) Integrated Plant Assecsment ( PA) Methodology described in Section 2.0 of the BGE LRA. Rese sections are prepared independently and will, collectively, comprise the entire flGE LRA.
5.5.1 Scoping System level scoping describes conceptual boundaries for plant systems and structures, develops screen:ng tools which capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify systems and structures within the scope of license renewal. Component level scoping descritys the components within the boundaries of those systems and structures that contribute to the intended functions. Scoping to determine components subject to aging management review (AMR) begins with a listing of passive intended functions and then dispositions the device types as either only associated with active functions, subject to replacement, or subject to AMR either in this report or another report.
Section 5.5.1.1 presents the results of the system level scoping,5.5.1.2 the resultr of the component level scoping, and 5.5.1.3 the results of scoping to determine components subject to an AMR.
Representative historical operating experience pertinent to aging is included in appropriate areas, to provide insight supponing the aging management demonstrations. His operating experience was obtained through key word searches of BGE's electronic databasm ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel.
5.5.1.1 System Level Scoping nis section begins with a description of the Cl Group, which includes the boundaries of the group as it was scoped. The intended functions of the Cl Group are listed and are used to define what portions of the group are within the scope of license renewal.
(
SysicmEcitription! Conceptual Boundaries There are numerous systems that have the Cl function and, therefore, conta!n containment isolation valves (CIVs), containment penetrations, and the associated piping and test connections. The components that perfonn the Ci function in systems that are evaluated in other sections of the BGE LRA i are included within those aging management sections; e.g.,the containment penetration portions of the Auxiliary Feedwater System are in Section 5.1 of the BGE LRA. The Cl Group largely consists of those CIVs, containment penetrations, and the associated piping and test connections in systems that either have no other components within the scope of license renewal or have components that are evaluated in other sections of the ilGE LRA (e.g., the Fire Protection [FP] sectian).
There are two systems, the Waste Gas (WG) System and the Demineralized Water (DW) System, that have a system pressure boundary intended function in addition to the Cl intended function. The scope of l
this report for the WG System includer the WG decay tanks and associated piping, isolation valves, and instrumentation that provide a pressure boundary for the stored WG. A ruptured WG d cay tank is a Design Basis Event unalyzed for CCNPP, ne scope of this repoit does not include DW System Application for License Renewal 5.5-1 Calvert Clifts Nuclear Power Plant
- g w w --
,--- -- y
ATTACilMl'NT (2) '
l APPENDIX A - TECilNICAL INFORMATION ,
5.5 - CONTAINMENT ISOLATION GROUP compor,er a beyond those that perfonn the Cl function. Cther DW System comporents support of Aniliary Feedwater System operation and are included in the AMR for the Auxiliary Feedwater System.
(Reference 1, Secthn 1.1.1; References 2 and 3]
'igure 5.51 is simplified diagram showing those portions of systems included in the Cl Group. It is proviued for information only.
Five of the systems, the FP, Plant llcating (Pil), DW, Plant Drains (PD), and Liquid Waste (LW)
Systems, have FP intended functions in addition to the Cl intended function. This report does not includo any additional romponents from these systems because of FP alone. Components in these systema that have only TP intanded functions are evaluated in Section 5.10 Fire Protection, of the 11GE GA. [Refe ence I, Section 1.1.11 The C Onup is comprised of portions of the following systems: [ Reference 1, Section 1.1.l]
FP PD Pit WG liW LW Plant Water (PW)
Each of the seven systems comprising the Cl Group provide Guid penetrations for the Containment llullding consisting of piping and valves, which rnect design basis doubic barrier criteria. The design basis governing isolstion valve requirement is to minimize Duld leakage through penetrations, not serving ngineered safety feature systems, by a double barrier so that no single, credible failure or malfunct.on of an active component can result in loss of isolation or intolerabl leakage. The installed double barriers take tne forr3 of closed piping systems, both inside and outside the Containment lluilding, and various types c.! isolation valves. [ Reference 2, Section 5.2.1)
Containment isolation valves are designed to ensure leak tightnas and reliability of operation.
Containment isolation globe, check, and gate valves meet the requirements of Manufacturers Standardin ion Society Speci0catin MSS SP 61, and Cl butter 0y valves meet the requirements of American Water Works Association C 504. Required valve clos!ng times are achieved by appropriate selection of valve, operator typ, and operator size. [ Reference 2, Section 5.2.1)
Sysicm Interfacss The Cl Group interfaces with the following plant systems and components: (Reference 1, Section 1.1.2]
- Containment Structure; e Containment Pen (trations; e Reactor Coolant System (RCS);
e Reactor Co.lant and Waste Process Sampling; e Gas Analyzing Sampling; and
. Engineered Safety Featuscs Actuation System.
Application for Ll cense Renewal 5.5 2 Calvert Cliffs Nuclear Power Plant
9 ATTACllMENT f1)
Al'?l:NDIX A . TI:CIINICAL INFORMATION 5.5 CONTAINMI:NT ISOLATION GROUl*
Containment f inside Outside (PT l' ire Protection
...@ l .
Q .-.
WSLR for Mk "
Instrument Lines ea Commodity Ls aluation WSLR for Containment Iwlation Group Plant Containment -
C Inside Outside fleating 1f Liquid, , , , N ,,
JL Waste l V us V A
Plant IIcatiria, C r2 WSLR for WSLR for i Demin
- Instrument Lines Containment J
- ~~~
Water "' ~ - Commodity isolation Y Lyaluation Group AN r A LJ PT
/ Weste
'~~ ~
I'*
Plant ,,,, ,,,, 7n Water Typical for instrumentation (typical) 4L us ( >
y Typicalfor 2 '
Plant ----- - - ---
A k Piping i lf 1 JL
- 6. a Note: Dashed lines represent Q portions not within the scope Oaseous _ _ _ NA V '
- M ]'
V ._ oflicense renewal (WSLR)
Waste V V JL JL
! Figure 5.5 1 l
Containment isolation Group Application for License Renewal 5.5-3 Calvert Clifts Nuclear Power Plant L
A1TAGIMENT f2)
APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP System Scoping Resuhs The Cl Group is within the scope of license renewal based on 10 CFR 54.4(a). He following intended functions of Cl Group systems were determined based on the requiremerts of $54.4(a)(1) and (2), in accordance with the CCNPP IPA Methodology: (Reference 1, Section 1.l.3]
e Provide CI (applies to all seven systems);
e Maintain the pressure boundary of the system (liquid and/or gas) for mitigation of Design Basis Events (applies to WG and DW Systems); and
- Maintain electrical continuity and/or provide protection of the electrical system.
The following intended functions o the r Cl Group systems were determined based on the requirements of 10 CFR 54.5(a)(3): [ Reference 1, Section 1.1.3]
e For Environmental Quali0 cation (10 CFR $0.49) Maintain functionality of electrical componcats as addressed by the Environmental Quali0 cation Program, provide information used to assess the environs and plant conditions during and following an accident, and provide CIV position indication.
All seven systems in the Cl Group also have FP functions that meet the requirements of 10 CFR 54.4(a)(3). Portion: of each of these systems are included in the FP evalur.tlon, discussed in Section 5.10 of the llGE LRA, for their role in performing passive FP functions. [ Reference 1 Sections 1.1.1 and 1.1.3]
All components of the Cl Group that support the above functions are safety related and Seismic Category 1, and are subject to the applicable loading conditions identined in the Updated Final Safety Analysis Report (UFSAR) Section 5A.3.2 for Seismic Category 1 systems and equipment design.
[ References 1,4, and 5]
The containment penetration piping up to the Orst outside Civ is designed in accordance with American National Standards institute (ANSI)ll31.7, Power Piping Code, Class 11. For those penetrations consisting of two outslie CIVs, the piping between the first and second outside valves is designed in accordance with ANS, G31.7, Class 111. Penetrations and associated piping are considered either Class 11 or Class MC for the purposes of the American Society of Mechanical Fngineers (ASME)Section XI Inservice inspection Program. The portion of the piping associated with the WG decay tanks that does not provide a Ci function is designed to ANSI B31.7, Class ill and is considered nm-Class for ASME Section XI. [ Reference 1, Attachment 3s for pipe; References 6 through 11]
Operating Exnerience The CCNPP Containment Leakage Rate ' resting Program has been inspected by the NRC on numerous occasions through routine inspections and during reviews of Technical Speci0 cation amendment requests. These inspections have not identined any aging related concerns that need to be addressed in the AMR of Cl Group components. Overall, the Containment Leakage Rate Testing Program has maintained the Cl portions of the Cl Group within the requirements of 10 CFR Part 50, Appendix J,
" Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors."
. Application for License Renewal S.5 4 Calvert Clifts Nuclear Power Plant i
. _ ~ . - - _ - _ . . .
ArrACitMENT (2)
APPENDIX A TECilNICAL INFORMATION 5.5 CONTAINMENT ISOLATION GROUP llaltimore Gas and Electric Company has requested and received Technical Specification amendments for revising the containment Type C testing schedule required under 10 CFR Part 50, Appendix J; e.g., to adopt the performance based requirements of Option B to Appendix J. During the reviews of these requests, significant analysis of past operating experience was performed for CCNPP and me industry as a whole. The NRC has indicated, based on their reviews of Type C performance history, and the wear.
out portion of the component life has not been reached, and may not be reacFed provided good maintenance practices continue to be followed. [ References 12 through 19] Additional infonnation on operating experience is provided in the Group 3 discussion of aging mt.nagement programs.
5.5.1.2 Component Level Scoping Based on the intended system functions listed above, the portion of the Cl Group that is within the scope of license renewal includes all safety.related components (electilcal, mechanical, and instrument), and their supports, comprising the containment penetration pressure boundary for the FP, Pil, PW, PD, LW, DW, and WG Systems. Also included are the safety related components (electrical, mechanical, and instrument), and their supports, associated with the WG decay tank pressure boundary. It should be noted that some non safety related portions of the FP, Pil, DW, PD, and LW Systems are included in the Fire Protection Evaluation, discussed in Section 5.10 of the BGE LRA, for their role in performing FP functions. [ References 3,4,5 and 20 through 27]
The Cl Group component types were scoped using controlled plant drawings, the UFSAR, and the NUCLEIS Equipment Technical Database. The UFSAR was used to determine which containment fluid penetrations were used by systems that do not perform a mitigating function other than Cl for Design Basis Events. For these systems, the purpose of the component level scoping was to identify all system components that suppo1 the intended functions of the system. This method and the process described in the CCNPP IPA Methodology are equivalent for ccmponent level scoping on systems. [ Reference 1, Section 2.1)
The following list of 10 device types are within the scope of license renewal for the Cl Group:
[ Reference 1. Table 21]
e Piping Class llB (carbon steel) e Level Switch e Piping Class llc (stainless steel) e Motor Operated Valve (MOV) e Check Valve e Pressure Transmitter e Control Valve e ReliefValve e lland Valve e Tank Some components in the Cl Grout 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 1, Section 3.2]
e Structural supports for piping 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 Cables Evaluation in Section 6.1 of the BGE LRA. This commodity evalua: ion completely addresses ti.e passive intended function entitled " maintain electrical continuity and/or provide protection of the electrical system" for the Cl Group.
Application for License Renewal 5.5 5 Calvert Cliffs Nuclear Power Plant
4 ATTACHMENT (2)
APPENDIX A - TECilNICAI, INFORhtATION 5.5. CONTAINMENT ISOLATION GROUP 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 pressure boundaries of the instruments themselves, are evaluated for the effects of aging in the Instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA.
5.5.1.3 Components Subject in AMR 1his section describes the components within the Cl Group that are subject to AMR. It begins with a listing of passive intended functions, and then dispositions the device types as either only associated with active functions, subject to replacement, evaluated in commodity evaluations, or remaining to be evaluated for aginF management in thl: section.
Passive Intended Functions in accordance with CCNPP IPA Methodology section 5.1, the following Cl Gioup functions were determined to be passive: [ Reference 1, Table 3 1) e Maintain the pressure boundary of the system (liquid and/or gas)(applies to WG System only);
- Provide CI (applies to all seven systems); and
- Maintain ebetrical continuity and/or provide protection of the electrical system (applies to DW and WG Systems only).
Dukclypes Subject to AMR Of the 10 device types within the scope oflicense renewal, two device types, level switch and pressure transmitter, are evaluated in the lustrument Line Commodity Evaluation in Section 6.4 of the BOE LRA.
[ Reference 1 Table 3 2,]
The remaining eight device types are listed in Table 5.51 and are subject to AMR. Unless otherwise annotated, all components of each listed device type are subject to AMR.
TAHLE 5.5-1 CLGROUP DEVICE TYPES REOUIRING AMR Piping Class 110 lland Valve
- Piping Class llc MOV Check Valve Relief Valve Control Valve Tank
- Instrument line manual drain, equalization, and isolation valves in the Cl Group that are subject to AMR are evaluated for the effects of aging in the Instrument Line Commodity Evaluation in Section 6.4 of the BGE LRA. Instrument line manual root valves are evaluated in this report. [ Reference 26, Attachment 3]
Application for License Renewal 5.56 Calvert Cliffs Nuclear Power Plant
i A*ITACilMENT (2)
APPENDIX A - TECilNICAL INFORMATION 5.5 CONTAINMENT ISOLATION GROUP ,
5.5.2 Aging Management ,
A list of potential age related degradation mechanisms (ARDMs) identified for the Cl +
Group comptnents is given in Table 5.5 2. The plausible ARDMs are identified in the table by a check mark (/) in the appropriate device type column. For AMR, some device types have a number of subgroups associated with them because of the diversity of material used in their fabrication or difTerences in the environments to which they are subjected. A check mark indicates that the ARDM applies to at least one subgroup for the device type listed.
For efliciency in presenting the results of the evaluations in this section, ARDM/ device type combinations are grouped together w here t!.ere are similar characteristics and the discussion is applicable to all con ponents within that group. Exceptiona are noted in the discussions where appropriate.
Table f 5 2 identifies the group in which each ARDM/ device type combination belongs. The following grourihave been selected for the Cl Group. [ Reference 2, Table 4 2)
Group i includes crevice corrosion, general corrosion, microbiologically induced corrosion (MIC),
and pitting for device types exposed to well water and subject to AMR. The affected device types include piping, check valves, hand valves, and MOVs.
Group 2 includes crevice corrosion, general ceirosion. and pitting for device types exposed to treated water or gaseous waste and subject to AMR, The affected device types include piping, check valves, control valves, hand valves, MOVs, relief *.alves, and tanks.
Group 3 includes wear for all valves sub,!cet to AMR, with the following exceptions: (1) check valves and MOVs in the pil System piping sections that are retired in placs, and capped off; and (2) all relief valves, i.e., WG decay tank relief valves.
Group 4 includes crevice corrosion, general corrosion, and pitting for the external bolting of the MOVs in the containment normal sump drain lines. These MOVs are the only components in the Cl Group with an aging management concern for the external surfaces because they are the only carbon steel subcomponents potentially exposed to boric acid from system leakage.
Application for License Renewal 5.57 ~'alvert Cliffs Nuclear Power Plant
ArrACllMENT (2)
APPENDIX A - TECllNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP TABLE 5.5-2 ,
1 POTENTIAL AND PLAUSIBLE ARDMn FOR Tile Cl GROUP Device Types"' Not Potential ARDMs ll H"' ilC"' CKY CV llY MOV RV TK Pinusible Cavitation Erosion x Corrosion Fatigue x
~
Crevice Corrosion <(1) <(2) <(i.2) <(2) <(1,2) <(1.2 A) <(2) <(2)
Dynamic Loading x Erosion Corrosion x Fatigue x Fouling x Galtsnic Corrosion x General Corrosion <(1) <( 1,2) <(2) ((1,2) <(l.2A) liydrogen Damage x Intergranular Attack x M1C <(I) <(1) <(I) <(I)
Paniculate Wear x Erosion l'itting <(1) ((2) <( 1.2) <(2) ((1,2) ((1.2A) ((2) ((2)
Radiation Damage x Rubber Degradation x Saline Water Attack x Selective Leaching x Stress Corrosion x Cracking Thermal Damage x !
Thermal x Embrittlement ~
Wear <(3) <(3) <O) ((3)
< . Indicates that the ARDM is plausible for component (s) within the device type
(#)- Indicates the Group (s)in which this ARDM/ device type combination is evaluated
. Notes:
l f.1) Not every component included within the device types listed here may be susceptible to a given ARDM. This is because components within a device type are not always fabricated from the same materials or snbjected to the same environments. Exceptions for each device type will be indicated Iri the aging management section for each ARDM discussed in this report.
(2) Class lill piping is carbon steel piping with well water as the internal Duid.
(3) Class llc piping is stainless steel piping with treated water as the internal Guid.
i l
Application for License Renewal 5.58 Calved Cliffs Nuclear Power Plant L
AIIArilMf%'T (2)
APPENDIX A TECilNICAL INFORMATION 5.5 CONTAINMENT ISOLATION GROUP
'!he 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 effectr, methods to manage aging, aging management program (s), and aging management demonstration.
Group 1 (crevlee rosion, general corrosion, MIC, and pitting for all components esposed to well water) . Materials and Environment Group 1 is comprised of carbon steel piping and components exposed to well water. Portions of the following systems are included: [ References 4,21, and 22]
- FP containment penetration ponion of the fire main heas.er supply to the containment hose stations;
- Pil . containment penetration portion of the supply and return lines for the containment unit heaters (currently retire in place and capped inside containment); and e PW containment penetration ponion of the supply line is the reactor head washdown area.
Check valves, hand valves, and MOVs in these lines are primarily constructed of carbon steel with some internal parts, i.e., wedge, seat, trim, hangers, and internal bolting, also constructed of alloy steels, stellited carbon ste:1, and stainless steel. In all cases, the valve disks / seats are also relied on for containment pressure boundary. The valves are normally in the closed position because the containment penettstion portions of these systems are not normally in use during plant power operation. The Pil System containment penetration components are currently retired in place. Ilowever, they are still considered susceptible to the subject ARDMs. [ Reference 1. Attachment 4s and 6s]
The Group I components are exposed to well water that is drawn from local wells. The water is nonnally pretreated by Sitering through an activated cubon Olter to remove suspended solids and chlorine from the Guid and to improve the taste, odor, and calor of the water, llowever, microbes are not removed or destroyed by the process. Water may occasionally Cow directly, i.e., bypass the carbon Otters, to the storage tanks from the wells if the standby pump is started due to a low level in the storage tanks. During nonnat power operation, the water in these lines is stagnant because the systems are not operating. [Refe.ence 1, Attachment 3s; References 4,21,22, and 28]
Group I (crevice corrosion, general cot rosion, MIC, and pitting for all components exposed to well water). Aging Mechanism Effects General corrosion is the thinning (wastage) of a metal by chenbal attack (dissolution) at the surface of the metal by an aggressive environment. General corrosion re ;uires an aggressive environment and materials susceptible to that environment. Wastage is not a co tern for austenitic stainless steel alloys and some high alloy steels. The consequences of the damage are loss ofload carrying cross sectional area. [ Reference 1, Attachment 6s and 7s]
Crevice corrosion is intense, localized corrosion within erevices 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, designed crevices for attaching thermal sleeves to safe-ends, and integral weld backing rings or back up bars. The crevice must be wide enough to permit liquid entry and narrow enough to Application for License Renewal 5.5 9 Calvert Cliffs Nuclear Power Plant
KiTACIIMENT {2)
APPENDIX A TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP ,
maintain stagnant conditions, typically a few thousandths of an inch or less. Crevice corrosion is closely related to pitting corrosion and can initiate pits in many cases, es well as leading to stress corrosion cracking. [ Reference 1, Attachment 6s and 75]
Pitting is another form of localized attack with greater corrosion rates at some locations than at others.
Pitting can be very insidious and destructive, with sudden failures in high pressure applications (especially in tubes) occurring by perforation. This form of corrosior essentially produces holes of varying depth to-diameter ratios in the steel. Deep pitting is more common with passive metals, such as sustenitic stainless steels, than with non-passive metals. Pits are generally elongated in the direction of gravity. [ Reference 1 Attachment 6s and 7s]
Microbiologically induced corrosion is accelerated corrosion of materleis resulting from surface microbiological activity. Sulfate reducing bacteria, sulfur oxidizers, and iron oxidizing bacteria arc most ,
commonly associated with these corrosion effects. Microbiologically induced corrosion most often tesults in pitting followed by excessive deposition of corrosion products. Stagnant or low How areas are most susceptible. Essentially all systems using untreated water and most commonly used materials are susceptible. Consequences range from leakage to excessive differential pressure and now blockage.
Microbiologically-induced corrosion is generally observed in systems utilizing raw untreated water.
[ Reference 1 Attachment 6s and 7s]
For Group I components, long term exposure to the well water environment may result in localized and/or general area material loss and, if unmanaged, could eventually result in loss of the pressure retaining capability under current licensing basis (CLD) design loading conditions. The areas where there are stagnant conditions, e.g., drain lines and crevices, are the locations most susceptible to these corrosion mechanisms. All of the ARDMs are plausible for carbon steel, alloy steel, and stellited carbon steel subcomponents. Subcomponents constructed of stainless steel are only susceptible to crevice corrosion, MIC, and pitting because the stainless steel material is resistant to general corrosion.
Since the valves in this group are required to maintain pressure boundary while in the closed position, degradation of the internal surfaces of all subcomponents required for the pressure-retaining function must be managed. [ Reference 1, Attachment 4s, $s, and 7s]
~
Group I (creviec corrosion, general corrosion, MIC, and pitting for all components exposed to well water). Methods to Manage Aging Mitigatiom For systems untaining well water and whose Dow is low or stagnant, water treatment and/or periodic Oushings will mitigate the corrosive processes described above. Ilowever, the occurrence of crevice corrosion, general corrorinn, MIC, and pitting is expected to be limited and not likely to affect ,
the intended cunction of the Group I components. The discovery techniques described below are deemed adequate for effectively managing aging of the subject components so that no mitigation techniques are required at this time. [ Reference 1, Attachment 6s]
Dinoiny; The effects of corrosion (crevice corrosion, general corrosion, MIC, and pitting) on CI Group I components can be discovered and monitered through non destructive examination techniques such as visual inspections. [ Reference 1, Attachment 8] These types of corrosion occur over a long period of time and can be discovered prior to wall thickness reaching an unacceptable value. Inspection results from representative samples of susceptible locations can be used to assess the need for additional Application for License Renewal 5.5-10 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (2)
APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP inspections at less susceptible locations. Based en piping / component geometry and fluid flow condPions, areas most likely to experience corrosion can be determined and evaluated, if corrosion i= occurring on valve seating surfaces, the degradation can be detected through pressure tests of the valves in the closed position. Corrosion would cause loss of material that can lead to valve leakage. Pressure testing for valve leakage would provide an early ladication of degradation of the valve seating surfaces so that corrective actions can be taken prior to the salves loosir.g their ability to satisfactorily perform their intended fundion. [ Reference 1, Attachment 8]
Group 1 (crevice corrosion, general corrosion, MIC and pitting for all components exposed to well water)- Aging Management Program (s)
Mitigation. Since there are no mitigation tectniques deemed necessary at this time, there are no mitigation programs credited for managing corrosion of Group I components.
Discoverv: For Gmup I componen's, crevice corrosion, general corrosion, MIC, and pitting can be readily detected through non-destructive examination techniques.110 wever, the occurrence of crevice corrosion, general corrosion, MIC, and pitting is expected to be limited and not likely to affect the intended function of the Group 1 components. To provide the additional assurance needed to conclude that the effects of corrosion are being effectively managed, the Group I components exposed to well water will be included in the scope of an ARDI Program. In addition, the CIVs will periodically be leak tested to provide an early indication of degradation of the valve seating surfaces. (Reference 1, 3 Attachment 8]
All Group 1 components will be included within a new plant program to accomplish the needed inspections for corrosion. This program is considered an Age-Related Degradation Inspection (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;
- Identification of inspection locations in the system / component based on plausible aging effects and consequences ofloss of component intended t~ unction;
- Determination of examination techniques (including acceptance criteria) that would be effective, considering the aging effects for whh:h the component is examined;
- 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.
Corrective actions iesulting from the ARDI will be taken in accordance with the CCNPP Corrective Action Program and will ensure that the components will remain capable of performing the pressure boundary integrity function under all CLB conditions.
Application for License Renewal 5.5-11 Calvert CI'.ffs Nuclear Power Plant
ATTACllh1ENT (2)
APPENDIX A - TECIINICAL INFORhlATION 5.5 - CONTAINMENT ISOLATION GROUP in addition to the ARDI Program, the C1 check valves for the FP System containment penetratlans and the manual valves for the PW System containment penetrations will bs subject to periodic pressure testing for valve leakage. Dese components are subject to local leak rate testing under the _CCNPP Surveillance Test Procedures in accordance with 10 CFR Part 50, Appendix J. [ Reference 1 Attachment 8: References 29 through 33) Continued local leak rate testing on a periodic basis will assure acceptable leak tightness at the seating surfaces of these valves and will also ensure that any leakage remains within the guidelines of the Technh al Speci0 cations.
The local leak rate test (LLRT) is part of the overall CCNPP Containment Leakage Rate Testing Program, which is implemented through Surveillance Test Procedures. The CCNPP Containment Leakage Rate Testing Program is discussed in detail below for Group 3. The corrective actions taken as part of the Containment Leakage Rate Testing Program will ensure that corrosion of the seating surfaces does not begin to affect the capability of the CIVs to perfonn their containment pressure boundary integrity function under all CLB conditions.
Group I (crevice corrosion, general corrosion, MIC and pitting for all components exposed to well water)- Demonstration of Aging Management Based on the factors presentcd above, the following conclusions can be reached with respect to crevice corrosion, general corrosion, MIC, and pitting of Cl Group components exposed to well water:
e The Group 1 components provide a system pressure-retaining boundary and their integrity must be maintained under all CLB conditions.
- Crevice corrosion, general corrosion, MIC, and pitting are plausible for the scoped components and result in material loss which, if len unmanaged, can lead to loss of pressure-retaining boundary integrity.
- The occurrence of crevice corrosion, general corrosion, MIC, and pitting is expected to be limited and not likely to affect the intended function of the Group I components.
- To provide the additional assurance needed to conclude that the effects of corrosion are being effectively managed, the Group I components exposed to well water will be included in the scope of an ARDI Program. Inspections will be performed and appropriate corrcMive action will be taken if signincant corrosior, is discovered.
e in addition to the ARDI Program, the C1 check valves for the FP System containment penetrations and the manual valves for the PW System containment penetrations will be subject to periodic pressure testing for valve leakage. Pressure testing for valve leakage would provide an early indication of degradation of the valve seating surfaces so that corrective actions can be taken prior to the valves loosing their ability to satisfactorily perform their intended function.
Therefore, there is reasonable assurance that the effects of crevice corrosion, general corrosion, MIC, and pitting on Group I components will be managed in such a way as to maintain the components' pressure boundary integrity, consistent with the CLB, during the period of extended operation.
i r
1 -
l Application for License Renewal 5.5-12 Calvert Cliffs Nuclear Power Plant l
ATTACIIMFNT {2)
APPENDIX A - TECHNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Group 2 (crevice corrosion, general corrosion, and pitting for all components exposed to treated water or gaseea: waste)- Materials and Environment Group 2 is comprised of stainless steel piping and components that are exposed to treated water or gaseous waste. Portions of the following systems are included: [ References 1,3,5 and 23 through 271
- WO - containment penetration portion of the reactor coolant drain tank vent header to the WG
- surge tank, and pressure boundary portion of the W3 decay tanks and associated piping and valves;
- PD - containment penetration portion of the drain line for the containment normal sump; e LW - containment penetration portion of the drain line for the RCS drain tank; and e DW - containment penetration portion of the line supplying DW to the quench tank.
The valve bodies of check valves, control valves, hand valves, MOVs, and relief valves in these systems are constructed of either stainless steel or carbon steel. Internal parts, i.e., wedge, seat, trim, hangers, and internal bolting, are constructed of a combination of steels including carbon steel, alloy steels, stellited carbon steel, and stainless steel. [ Reference 1, Attachment 4s]
The WG decay tanks are constructed of carbon steel and are internally clad with stainless steel. Flaages and couplings are stainless sieel and the bolting is carbon and low alloy steel. All of these subcomponents support the pressure boundary function and are subject to AMR. [ Reference 1, Attachment 4 for TK]
The Group 2 component internal surfaces are exposed to water and gas from a number of sources including reactor coolant drains, DW lines, containment normal sump drain, and gaseous waste lines.
The water and gas is from process systems that use treated water from the DW System as makeup water.
The DW System reduces the concentration of oxygen and removes mineral salts and ions thereby providing . source of pure water to minimize the corrosive environment of plant process streanis.
Further, chemistry controls are placed on several of the systen s, such as the RCS, which helps to minimize the corrosiveness of the environment. However, the containment normal sump drain lines and RCS drain tank drain lines may contain water from a number of sources, including the Reactor Coolant and Chemical and Volume Control Systems that contain boric acid, which is corrosive to some steels.
[ Reference 1, Attachment 3s; References 3,5, and 23 through 27]
Group 2 (crevice corrosion, general corrosion, and pitting for all components exposed to treated water or gaseous waste)- Aging Mechanism Effects Steels are susceptible to general and localized (crevice and pitting) corrosion mechanismr in a water environment. The aggressiveness of these corrosion mechanisms are particularly dependent on local water chemistry conditions including oxygen levels and on the component construction materials. The areas where there are stagnant conditions, e.g., drain lines and crevices, are the locations most i susceptible to these corrosion mechanisms. Refer to the discussion for Group 1 above for a detailed discussion of the effects of crevice corrosion, general corrosion, and pitting of steels. [ Reference 1, Attachment 6s and 7s]
Application for License Renewal 5.5-13 Calvert Cliffs Nuclear Power Plant I
i
ATTACilMFNT (2)
Ar"ENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Long-term exposure to these environments may result in localized pitting and/or general area material loss and, if unmanaged, could eventually result in loss of the pressure-retaining capability under CLB design loading conditions. Therefore, general corrosion, crevice corrosion, and pitting corrosion have been determined to be plausible ARDMs for which aging effects must be managed for the internal surfaces of Group 2 components. All of the ARDMs are considered plausible for carbon steel, alloy steel, and stellited carbon steel subcomponents. Subcomponents constructed of stainless steel are only susceptible to crevice corrosion and pitting. They are not subject to general corrosion because stainless steel is resistant to general corrosion. Corrosion due to exposure to boric acid is not plausible because all subcomponents exposed to water that contains boric acid are constructed of stainless steel or high alloy steel, [ Reference 1, Attachment 4s,5s, and 7s]
Degradation resulting from these ARDMs must be managed for all internal surfaces and subcomponena of the piping, tanks, and valves. This includes the disks / seats for the valves, with the exception of some hand valves in instrument lines, because the valves are required to maintain pressure boundary while in the closed position. The stainless steel valve trim for the control valves also supports the pressure boundary function and is subject to crevice corrosion and pitting. All other valves with trim do not rely on the trim to support the pressure boundary function and, therefore, the trim is not subject to AMR.
[ Reference 1, Attachment 4s,5s and 6s]
Group 2 (crevice corrosion, general corrosion, and pitting for all components exposed to treated water or gaseous waste) . Methods to Manage Aging %
Mitigation: For systems containing treated water or gaseous waste, crevice corrosion, general corrosion, and pitting are considered plausible, especially in areas where the flow is low or stagnant and in cracks and crevices. Ilowever, the occurrence of crevice corrosion, general corrosion, and pitting is expected to be limited and not likely to affect the intended funct on of the Group 2 components. The discovery technique described below is deemed adequate for effectively managing aging of the subject components so that no mitigation techniques are required. [Referenct 1, Attachment 6s]
Discoverv: The effects of corrosion (crevice corrosion, general corrosion, and pitting) on CI Gcoup 2 components can be discovered and monitored through non-destructive examination techniques such as visual inspuctions. [ Reference 1, Attachment 8] These types of corrosion occur over a long period of time and can be discovered prior to any threat of minimum wall thickness reaching an unacceptable value. Representative samples of susceptible locations can be used to assess the need for additional inspections at less susceptible locations. Based on piping / corr.ponent geometry and Guid Row conditions, areas most likely to experience corrosion can be determined and evaluated.
if corrosion is occurring on valve swating surfaces, any leakage caused by this corrosion can also be detected through pressure tests of the valves in the closed position. Pressure testing for valve leakage, such as through local leak rate testing of CIVs, would provide an early inoication of degradation of the valve seating surfaces so that corrective actions can be taken prior to the valves losing their ability to satisfactorily perform their intended function. Degradation of valve seating surfaces can be discovered by either the leakage testing or visual inspection methods. [ Reference 1, Attachment 8J Application for License Renewal 5.5-14 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (2)
APPENDIX A - TECIINICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Group 2 (erevlee corrosion, general corrosion, and pitting for all components exposed to treated water or gaseous waste)- Aging Managerient Program (s)
Mitigation: Since there are no mitigation techniques deemed necessary at this time, there are no mitigation programs credited for managing corrosion of Group 2 components.
Discoverv: For Group 2 components, crevice corrosion, general corrosion, and pitting can be readily detected through non-destructive examination techniques. Ilowever, the occurrence of crevice corrosion, general corrosion, MIC, and pitting is expected to be limited and not likely to affect the intended function of the Group I components. To provide the additional assurance needed to conclude that the effects of corrosion are being effectively managed, all of the components exposed to treated water or gaseous waste will be included in the scope of an ARDI Program. In addition, the CIVs will periodically be leak tested to provide an early indication of degradation of the valve seating c,urfaces. [ Reference 1, Attachment 8]
All Group 2 components will be included within a new plant program to accomplish the needed inspections for corrosion. This program is considered an ARDI Program as denned in the CCNPP IPA Methodology presented in Section 2.0. Refer to the Group 1 discussion on aging management programs for a detailed discussion of the ARDI Program.
Corrective actions will be taken in accordance with the CCNPP Corrective Action Program and will ensure that the components will remain capable of performing the system pressure boundary integrity
- function under all CLB conditions.
In addition to the ARDI Program, the CIVs will be subject to periodic pressure testing for valve leakage.
These components are subject to local leak rate testing under the CCNPP Surveillance Test Procedures in l
accordance with 10 CFR Part 50, Appendix J. Included are control valves for the vent header and drains of the reactor coolant drain tank, contr6i ulves for the DW supply to the quench tank, and MOVs for the
- containment normal sump drain lines. [ Reference 1, Attachment 8; References 29, 30, 31, and 34 l through 39] Continued local leak rate testing on a periodic basis will assure acceptable leak tightness at I the seating surfaces of these valves and will also ensure that any leakage remains within the guidelines of the Technical Speci6 cations.
The LLRT is part of the overall CCNPP Containment Leakage Rate Testing Program, which is t
implemented through Surveillance Test Procedures. The CCNPP Containment Leakage Rate Testing Program is discussed in detail below for Group 3. The corrective actions taken as part of the l
l Containment Leakage Rate Testing Program will ensure that corrosion of the seating surfaces does not l begin to affect the capability of the CIVs to perform their containment pressure boundary integrity j function under all CLB conditions.
l Application for License Renewal 5.5-15 Calvert Cliffs Nuclear Power Plant
ATTACllMrNT (2)
APPENDIX A - TECIINICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Group 2 (crevice corrosion, general corrosion, and pitting for all components exposed to treated water or gaseous waste)- 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 Cl Group 2 components exposed to treated water or gaseous waste:
- The Group 2 components provide a system pressure-retaining boundary and their integrity must be maintained under all CLB conditions.
- Crevice corrosion, general corrosion, and pitting are plausible for the scoped components and result in material loss which, ifleft unmanaged, can lead to loss of pressure-retaining boundary integrity.
- The occurrence of crevice corrosion, general corrosion, and pitting is expected to be limited and not likely to affect the intended function of the Group 2 components, e To provMe the additional assurance needed to conclude that the effects of corrosion are being effecSvely managed, all of the components exposed to treated water or gaseous waste will be Scluded in the scope of an ARDI Program. Inspections will be performed and appropriate corrective action will be taken if significant corrosion is discovered.
. in addition to the ARDI Program, the Group: CIVs will be subject to periodic pressure testing for valve leakage. Pressure testing for valve leakage would provide an early indication of degradation of the valve seating surfaces so that corrective actions can be taken prior to the valves loosing their ability to satisfactorily perfonn their intended function.
Therefore there is reasonable assurance that the effects of crevice corrosion, general corrosion, and pitting on Group 2 components will be managed in such a way as to maintain the components' pressure boundary integrity, consistent with the CLB, during the period of extended operation.
Group 3 (near of valves)- Materials and Environment All check valves, control valves, hand valves, and MOVs, with the exception of some hand valves in the instrument lines, in the CI Group have disks and seats that are relied on for system pressure boundary because the valves must be in the closed position to perform the intended function. Wear is considered plausible for the disks and seats of all the Cl Group valves, with the exception of those in the Pli System, that are retired in place and no longer operated. Wear is also not plausible for any relief valves, i.e., relief valves for the WG decay tanks, because they operate relatively infrequently. The valve bodies are constructed of either stainless steel or carbon steel. Internal parts, i.e., wedge, seat, trim, hangers, and internal botting, are constructed of a combination of steels including carbon steel, alloy steels, stellited carbon al, and stainless steel. (Reference 1, Attachment 4s]
The internal environment for the Group 3 valves is treated water or reactor coolant gaseous waste. The environmental conditions for these valves are discussed above in Group 2.
Application for License Renewal 5.5 16 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (2)
APPENDIX A - TECIINICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Group 3 (wcar of valves) - Aging Mechanism Effects Wear results from relative motion between two surfaces (adhesive wear), from the influence of hardi abrasive particles (abrasive wear) or fluid stream (erosion), and from small, vibratory or sliding motions under the influence of a corrosive environment (fretting). Motions may be linear, circular, or vibratory in inen or corrosive environments. In addition to material loss from the above wear mechanisms, impeded relative motion between two surfaces held in intimate contact for extended periods may result in galling /self welding. Wear rates may accelerate as expanded clearances result in higher contact stresses. [ Reference 1. Attachment 7 for Valves]
The disks and seats of check valves subject to AMR for wear are required to maintain containment pressure boundary integrity or system pressure boundary integrity. Wear is considered plausible for the disks and seats of Group 3 valves because they may experience cyclic relative motion at the tight fitting surfaces. Movement of the disk against the seat can result in a gradual loss of material, which could result in a small amount ofleakage. If left unmanaged, wear could eventually lead to a loss of pressure boundary integrity. Wear is not plausible for the relief valves on the WG decay ttnks because they change position, i.e., relieve pressure, relatively infrequently. Wear is also not plausible for the valves in the Pil System because they are retired in place. [ Reference 1, Attachment 4s,5s, and 6s]
Group 3 (wear of valves). Methods to Manage Aging Effects Mitigation: Since the wear of valve disk and seats is due to valve operation, decreased operation of the valves would slow the degradati< d the valves seating surfaces. This is not a feasible mitigation technique because it would plac6 mecessary restrictions on plant operation. The restrictions are unnecessary because limited leakage through the valves will not significantly impact the intended function. Furthermore, the discovery methods discussed below are deemed adequate for verifying that significant degradation is not occurring. It should be noted that galling /self welding occur when there is impeded relative motion between two surfaces held in intimate contact for extended periods. Periodic valve operation actually minimizes this phenomenon. [ Reference 1, Attachment 6s]
Discoverv: Wear for valve disks and seats can be detected by performing leak rate testing or including the valve component in an inspection program. [ Reference I, Attachment 8] Since wear occurs gradually over time, periodic leak testing can be used to discover leakage that may be caused by wear of the seating surfaces so that corrective actions can be taken prior to the loss of the intended ftmetion. In an inspection program, representative samples of susceptible locations can be used to assess the need for additional inspections at less susceptible locations.
Group 3 (wcar of valves)- Aging Management Program (s)
Mitigation: There are no feasible methods of mitigating wear of the valve disks and seats; therefore, there are no programs credited with mitigating the aging effects due to this ARDM.
Discovery: All of the Group 3 check valves, control valves, and MOVs that perform the containment pressure boundary function, and the hand valves for the PW System containment penetration, are subject to local leak rate testing under the CCNPP Containment Leakage Rate Testing Program, as required by 10 CFR Part 50 Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Application for License Renewal 5.5-17 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT_(2)
APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Reactors," Option B. This program is implemented in accordance with the plant Technical Speci0 cations. The check valve in the DW System and all hand valves, with the exception of the hand valves that serve as CIVs for the PW System, are not subject to local leak rate testing. Thesc components will be included within a new plant program, ARDI, to accomplish the needed inspections for wear. [ Reference 1 Attachment 8; References ?^ :hrough 41]
CCNPP Containment Leakage Rate Testing PrDgam The LLRT is performed under Surveillance Test Procedures, References 30,31, and 34 through 39, as part of the overall CCNPP Containment Leakage Rate Testing Program. The CCNPP Containment Leakage Rate Testing Program was established to implement the leuage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J. Appendix J specines containment leakage testing requirements, including the types of tests required, frequency of testing, test methods, test pressures, acceptance criteria, and reporting requirements. Containment leakage testing requirements include performance of ii.tegrated leakage rate tests, also known as Type A tesis, and LLRTs, also known as Type B and C tests. Type A tests measure the overall leakage rate of the containment. Type B tests are intended to detect leakage paths and measure leakage for certain containment penetrations (e.g., airlocks, ihnges, and electrical penetrations). Type C tests are intended to measure CIV leakage rates. [ References 29, Section 6.5.6; References 40 and 41]
The CCNPP Containment Leakage Rate Testing Program is based on 10 CFR Part 50, Appendix J, requirements and implements the requirements in CCNPP Technical Speci0 cations 3.6.1.2,4.6.1.2, and 6.5.6. The scope of the program Scludes Type B and C testing of containment penetrations.
[ References 29 through 39]
The LLRTs are performed at a frequency in accordance with 10 CFR Part 50, Appendix J, Option B. Per References 30 through 39, the LLRT currently includes the following procedural steps:
- Leak rate monitoring test equipment is connected to ti:e appropriate test point.
. Test volume is pressurized to at least 53 1 psig, which is conservative with respect to 10 CFR Part 50, Appendix J, test pressure requirements.
. Leak rate, pressure, and temperature are monitored at the frequency speciGed by the LLRT procedure and the results are recorded, e The maximum indicated leak rate is compared against administrative limits that are more restrictive than the maximum allowable leakage limits.
. "As found" leakage equal to or greater than the administrative limit, but less than the maximum allowable limit, is evaluated to determine if further testing is required and/or if corrective maintenance is to be perfonned.
- For "as found" leakage that exceeds the maximum allowable limit, the appropriate supervisory plant personnel determine if Technical Speci0 cation Limiting Condition for Operation (LCO) 3.6.1.2.b has been exceeded. Technical SpeciGcation 3.6.1.2.b contains the maximum allowable combined leakage for all penetrations and valves subject to the Type B and C tests.
Corrective action is taken as required to restore the leakage rates to within the appropriate acceptance criteria.
Application for License Renewal 5.5-18 Calvert Cliffs Nuclear Power Plant
ATTACIIMENT (2)
APPENDIX A - T.ECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP
- If any maintenance is required on a CIV that changes the closing characteristics of the valve, an "as left" test must be performed on the penetration to ensure leakage rates are acceptable.
The CCNPP Containment Leakage Rate Testing Program has been inspected by the NRC on numerous occasions through routine inspections and during reviews of technical specification amendment requests.
Routine inspections at the site included procedure reviews, leakage test witnessing, test reviews, and results evaluation of both integrated leakage rate tests and LLRTs. Inspectors noted when individual CIVs failed their leakage tests and reviewed the repair and resetting actions taken by BGE. With some specific exceptions, the inspections typically noted acceptable conditions. No aging related deficiencies were identified. [ References 12 through 15]
Baltimore Gas and Electric Company has requested, and received, Technical Specification amendments for revising the containment Type C testing schedule required under 10 CFR Part 50, Appendix J.1he requests were initiated to accommodate extending the fuel cycle to 24 months, and to recognize the added Option B under Appendix J. Currently, CCNPP follows the schedule of Option B, which is a performance-based scheduling process. During the reviews of these requests, significant analysis of past -
operating experience was performed for CCNPP and the industry as a whole. The NRC has indicated, based on their reviews of Type C performance history, that the wear-out portion of the component life has not been reached, and may not be reached provided good maintenance practices continue to be followed. Furthermore, reviews of site-specific data indicate that the leakage rate data at the end of the CCNPP Unit I operating cycles falls within a typical range. [ References 16 through 19]
These reviews demonstrate that CCNPP has normal and acceptable operating experience with respect to component aging of components relied on for Cl. The corrective actions taken as part of the Containment Leakage Rate Testing Program will ensure that the Cl check valves and MOVs remain capable of performing their containment pressure boundary integrity function under all CLB conditions.
CCNPP ARDI Program The check valve in the DW System and all tne Group 3 hand valves are not currently subject to local leak rate testing, with the exception of the hand valves that serve as CIVs for the PW System. Wear can be detected for these components through visual inspections or leakage testing. These components will be included in a new plant program to accomplish the needed inspections for wear. This program is considered an ARDI Program as defined in the CCNPP IPA Methodology presented in Section 2.0.
Refer to the Group 1 discussion on aging management programs for a detailed discussion of the ARDI Program, [ Reference 1, Attachment 8)
Corrective actions will be taken in accordance with the CCNPP Corrective Action Program and will ensure that the components will remain capable of performing the system pressure boundary integrity function under all CLB conditions.
i
~
Application for License Renewal 5.5-19 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (2)
APPENDIX A - TECHNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Group 3 (wcar of valves)- Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to wear of Group 3 valves for the Cl Group:
- The valve disks and seats maintain containment pressure boundary and their integrity must be maintained under all CLB conditions.
- Wear is plausible for valve disks and seats and results in material loss which, if left unmanaged, could eventually lead to leakage and loss of pressure boundary.
- The CIVs are subject to local leak rate testing in accordance with the CCNPP Containment Leakage Rate Testing Program.
. Leak testing will continue to be performed by these programs in accordance with the plant Technical Specifications, and appropriate corrective actions will be taken if signincant leakage due to wear of the seating surfaces is discovered.
- The check valve in the DW System and all the hand valves, with the exception of the PW System CIVs, are not subject to local leak ute testing and, therefore, will be included in the scope of an ARDI Program. Inspections will be performed, and appropriate corrective action will be taken if signi0 cant wear is discovered.
Therefore, there is reasonable assurance that the effects of wear for Group 3 valves will be managed in such a way as to maintain the components' pressure boundary integrity, consistent with the CLB, during the period of extended operation.
Group 4 (crevice corrosioa, general corrosion, and pitting for the external bolting of the MOVs)-
Materials and Environment This group includes crevice corrosion, general corrosion, and pitting for the external botting of the MOVs in the containment normal sump drain lines. The MOVs in the containment normal sump drain lines contain bolts external to the process Guid that support the system pressure boundary function.
These bolts are constructed of carbon and low alloy steel materials. The containment normal sump drain lines contain water from the Reactor Coolant and Chemical and Volume Control Systems, which contain boric acid, if the valves develop a leak, there is the potential for the carbon steel bolts to be exposed to corrosive boric acid. These MOVs are the only components in the Cl Group with an aging management concern of the external surfaces because they are the only carbon steel subcomponents potentially exposed to boric acid from system leakage. [ Reference 1, Attachments 4 and 6 for MOV)
Group 4 (crevice corrosion, general corrosion, and pitting for the external bolting of the MOVs)-
Aging Mechanism Effects General corrosion is the thinning (wastage) of a metal by chemical attack (dissolution) at the surface of j
the metal by an aggressive environment. The consequences of the damage are loss of load-carrying cross-sectional area. General corrosion requires an aggressive environment and materials susceptible to that environment. This ARDM is plausible for the external surfaces of the Group 4 MOVs because susceptible materials of construction of the bolts (e.g., low alloy steel, carbon steel) are potentially exposed to borated water leakage. Additionally, crevice corrosion and pitting can occur when crevices Application for License Renewal 5.5-20 Calvert Cliffs Nuclear Power Plant
4 A'ITAC11 MENT (2)
APPENDIX A - TECIINICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP (e.g., under nuts and/or bolt heads) are exposed to leakage. Refer to the discussion for Group 1 above for a detailed discussion of the efTects of crevice corrosion, general corrosion, and pitting of steels.
[ Reference I, Attachments 6 and 7 for MOVs]
Long term exposure to this environment may result in localized pitting and/or general area material loss and, if unmanaged, could eventually result in loss of the pressure-retaining capability under CLB design loading conditions. Therefore, general corrosion, crevice corrosion, and pitting corrosion have been determined to be plausible ARDMs for which aging effects must be managed for the external botting of the Group 4 MOVs. [ Reference 1, Attachment 6 for MOVs]
Group 4 (crevice corrosion, general corrosion, and pitting for the external bolting of the MOVs)-
Methods to Manage Aging-hiitigation: The effects of crevice corrosion, general corrosion, and pitting, which occur due to leakage of borated water, can be mitigated by minimizing leakage so that the components are not exposed to a corrosive environment. There is no mitigation technique required because the discovery programs discussed below are deemed adequate to manage aging. [ Reference 1, Attachment 8]
Dnwicry: The degradation of the Group 4 MOVs, that does occtr, can be discovered and monitored by performing visual inspections of the bolting that has been subject to boric acid leakage for signs of crevice corrosion, general corrosion, and pitting, and taking appropriate corrective action when appropriate. [ Reference 1, Attachmeat 8; Reference 42]
Group 4 (crevice corrosion, general corrosion, and pitting for the external bolting of the MOVs)-
Aging Management Program (s)
Mitigation: Since no mitigation techniques are required, there are no programs credited for mitigation.
Discovery: The CCNPP Boric Acid Corrosion Inspection (BACI) Program provides systematic requirements to ensure that boric acid corrosion does not degrade the reactor coolant pressure boundary and thereby increase the probability of abnormal leakage, rapidly propagating failure, or gross rupture.
The program controls examination and test methods and actions for minimizing the loss of structural and pressure-retaining integrity of components due to boric acid corrosion. It has been established in response to NRC Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressore Boundary Components in PWR [pressuri:cd water reactorf Plants." [ Reference 42]
The scope of the program is threefold and provides: (1) examination locations where leakage may cause degradation of the primary pressure boundary by boric acid corrosion; (2) examination requirements and methods for the detection ofleaks; and (3) the requirements for initiating engineering evaluations and the subsequent proposed corrective actions. [ Reference 42]
The program requires a containment walkdown, i.e., VT 2 visual exam (a type of visual examination described in ASME XI, IWA-2212), followirg each reactor shutdown (as soon as possible after attaining Ilot Shutdown condition) to identify and quantify any leakage found in speci6c areas of the Containment and Auxiliary Buildings. A second walkdown is performed during heatup, prior to plant startup (after attaining normal operating pressure and temperature), if leakage was identiGed and corrective actions Application for License Renewal 5.5-T Calvert Cliffs Nuclear Power Plant
A1TACllMENT (2)
APPENDIX A - TECilNICAL INFORMATION 5,5 - CONTAINMENT ISOLATION GROUP were taken. Only locations where the initial inspection identified leakage are included in this second walkdown. The walkdowns are performed in accordance with CCNPP Administrative Procedure MN.3110," Inservice Inspection of ASME Section XI Components," and Procedure MN-3 301, " Boric -
Acid Corrosion inspection Program." A containment walkdown is not required if a reactor shutdown occurs within 30 days of a previous shutdown unless the reason for the shutdown is excessive RCS leakage. [ Reference 42]
The program also requires examination of specific components for discovery of leakage during each refueling outage. Some of the components examined include carbon steel holting on Class i valves, valves in systems containing borated water, which could leak onto Class I carbon steel components, and components that are the subject of issue Reports where boric acid leakage has been identified.
[ Reference 42]
The procedures require that, if leakage or corrosion is discovered, corrective actions will be taken in accordance with the CCNPP Corrective Action Program to document and rerolve the deficiency. The corrective actions address the removal of boric acid residuc and inspection of the cornponents for corrosion. Follow-up actions also address the evaluation of the component for continued service and initiation of corrective actions to prevent recurrences. The combined visual inspection and corrective action programs currently in place will ensure that the MOVs will rema!n capable of performing the system pressure boundary integrity function under all CLB conditions. [ Reference 42]
The BACI Program has evolved over the years to account for operational experience in both CCNPP Units due to occurrences of boric acid leakage, in one case, the BACI Program was modified to include new inspection locations. In another case, the program was modified to include a requirement that all occurrences of boric acid leakage be formally evaluated within the program. Prior to that, it was possible for boric acid leakage to be corrected through other means and without a formal review within the BACI Program. [ Reference 42]
The llACI Program is subject to periodic internal assessment activities, including audits that serve to provide a comprehensive, independent verification and evaluation of the quality-related activities and procedures of the BACI Program. A Master Assessment Plan is created to provide a standardized plan for assessing performance of the BACI Program. The Master Assessment Plan identifies the critical program elements, expected results, and key attributes that should be visible during field operatioas or assessment activities. Long-term assessment results are provided to personnel responsible for maintaining the BACI Program. The site Quality Assurance Policy requires an assessment of the BACI Program every two years. [ References 43,44, and 45]
Group 4 (crevice corrosion, general corrosion, and pitting for the external bolting of the MOVs)-
Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to the Group 4 MOV external bolting subject to crevice corrosion, general corrosion, and pitting:
- The subject MOVs maintain containment pressure boundary and their integrity must be maintained under all Cl B conditions.
Application for License Renewal 5.5-22 Calvert Cliffs Nuclear Power Plant
ATTACilMENT (2)
APPENDIX- A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP e Crevice corrosion, general corrosion, and pitting are plausible for the external bolting of Group 4 MOVs due to potential leakage of borated water creating a corrosive environment around the carbon and low alloy steel bolts. If left unmanaged, corrosion could eventua!!y result in loss of material such that the components may not be able to perform their pressure boundary function.
- The BACI Program inspections will detect signs ofleakage, boric acid residue, or the effects of corrosion on the external surfaces of the Group 4 MOVs, and will ensure corrective actions will be taken such that there is reasonable assurance that the passive intended function will be maintained, Therefore, there is reasonable assurance that the effects of crevice corrosion, general corrosion, and pitting will be managed for the external bolting of the Group 4 MOVs such that they will be capable of performing their passive intended function, consistent with tl.e CLB, during the period of extended operation.
5.5.3 Conclusion The programs discussed for the CI Group are listed in Table 5.5 3. These programs are (and will be for new programs) administratively controlled by a formal review and approval process. As has been demonstrated in the above section, these programs will manage the aging mechanisms and their effects such that the intended functions of the components of the Cl Group will be maintained, consistent with the CLB, 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.
Application for License Renewal 5.5-23 Calvert Cliffs Nuclear Power Plant f
NITACIIMENT m APPENDIX A - TECHNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP Table 5 5-3 1 IST OF AGING MANAGEMENT PROGRAMS FSR THE CI GROUP Pmaram Credited As Existing CCNPP _!-akage Rate Testing
- Discovery and management of leakage that Program: could be the result of the effects of crevice LLRT Procedures STP-M 571 A 1, c rr si n, general c rr si n, MIC, and pitting n the seating surfaces of the Cl Group CIVs STP M 571 A 2, STP M-571D-1, that are exposed to well water (Group I)
STP-M 571D-2, STP-M 57IE-1, STP M 571E-2, STP M 571G 1, a Discovery and management of leakage that STP-M 571G-2, STP-M 571M-1, could be the result of the efTects of crevice and STP-M 571M 2 corrosion, general corrosion, and pitting on the seating surfaces of the Cl Group CIVs that are exposed to treated water or gaseous waste (Group 2)
- Discovery and management of leakage that could be the result of the effects of seating surface wear on all of the Cl Group CIVs (Group 3)
Existing CCNPP BACI Program:
- Discovery and management of the effects of crevice corrosion, general corrosion, and CCNPP Administrative Procedure MN-3 301, " Boric Acid Corrosion pitting of the external bolting on Cl Gr up MOVs that are located in borated water inspection Program,,
systems (Group 4)
New ARDI Program .
Discovery and management of the effects of crevice corrosion, general corrosion, MIC, and pitting of the Cl Group components that are exposed to well water (Group 1)
- Discovery and management of the effects of crevice corrosion, general corrosion, and pitting of the CI Group components that are exposed to treated water or gaseous Waste (Group 2)
Discovery and management of the effects of wear of the Cl Group valves that are not CIVs, i.e., check valve in the DW System, the relief valves, and all hand valves, with the exception of the PW System CI hand valves. (Group 3)
Application for License Renewal 5.5-24 Calvert Cliffs Nuclear Power Plant
m ATTACllMENT Q) k9 APPENDIX A - TECilNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP ,
5.5.4 References
- 1. CCNPP Aging Management Review (AMR) Report for the Containment Isolation Group, Revision 1, October 1997
- 2. CCNPP Updated Final Safety Analysis Report, Units 1 and 2, Revision 20
- 3. CCNPP Drawing No. 60735Sl!0001, " Waste Gas and Miscellaneous Waste Processing Systems," Revision 38, October 20,1996
- 4. CCNPP Drawing No. 60746S110003, " Plant Water and Air Service System," Revision 16, May 6,1996
- 5. CCNPP Drawing No,60733Sil000)," Auxiliary Building, Waste Processing Equipment, and Area Drains," Revision 24, January 22,1997
- 6. CCNPP Drawing No. 92769Sil llB-5, "M 601 Piping Class Summary," Revision 28, April 29,1993
- 7. CCNPP Drawing No. 92769S11-118-6, "M-601 Piping Class Summary," Revision 20, April 29,1993
- 8. CCNPP Drawing No. 92769Sil-IIC-1, "M-601 Piping Class Summary," Revision 25, May 22,1995
- 9. CCNPP Drawing No. 92769Sli IIC-2, "M 601 Piping Class Summary," Revision 21, April 29,1993
- 10. CCNPP Drawing No. 92769Sil-llc-3, "M-601 Piping Class Summary," Revision 20, April 29,1993
- 11. CCNPP Drawing No. 92769S11-11C-4, "M-601 Piping Class Summary," Revision 20, April 29,1993
- 12. NRC Inspection No. 50 317/82-15, " Routine, Unannounced Inspection of the Containment Penetration Leakage Testing Program, the Containment Integrated Leakage Rate Test, Tours of Facility, and Follow-up on Previous inspection Findings," June 16,17,18,21,22,1982
- 13. Letter from Mr. T. T. Martin (NRC) to Mr. A. E. Lundvall, Jr. (BGE) dated January 20,1983,
" Inspection No. 50-318/82 26" (Routine, Unannounced inspection of Procedure Review, Witnessing and Results Eval nation of Local Leak Rate Test and Integrated Leak Rate Test, December 15 through 18,1982)
- 14. Letter from Mr. S. D. Ebneter (NRC) to Mr. A. E. Lundvall, Jr. (DGE), dated June 25,1985,
" Inspection No. 50-317/85-10"(Routine, Announced inspection of the Containment Leakage Testing Program including Procedure Review of Containment Integrated Leakage Rate Test (CILRT) and Local Leak Rate Test (LLRT) Procedures, CILRT and LLRT Witnessing, CILRT and LLRT Test Review, On-Line Primary Containment Leakage Monitoring, and General Tours of the Facility, April 29 - May 2, and May 17 - 21,1985)
- 15. Letter from Mr. S. D. Ebneter (NRC) to Mr. A. E. Lundvall, Jr. (BGE), dated December 24,1985, " Combined Inspection Nos. 50-317/85 33 and 50-318/85-33" (November 18 througn 25,1985)
Application for License Renewal 5.5-25 Calvert Cliffs Nuclear Power Plant
e NITACllMENT (2)
APPENDIX A - TECliNICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP
- 16. - Letter from Mr. S. A. McNeil (NRC) to Mr. O. C. Creel (DGE), dated March IS,1989,
" Issuance of Technical Specification Amendment and Temporary Exemption Concerning Retest Schedular Requirements of Appendix J to 10 CFR Part 50 for Types B and C Local Leak Rate Tests (TAC No. 71589)"[ Amendment No. I18, Unit 2]
- 17. Letter from Mr. D. G. Mcdonald, Jr. (NRC) Mr. Mr. G. C. Creel (DGE), dated February 19,1992," Issuance of Amendments for CCNPP Unit No.1 (TAC No. M82213) and Unit No. 2 (TAC No. M82212)" [ Amendment Nos. 168/147]
- 18. Letter from Mr. A.-W. Dromerick (NRC) Mr. C.11. Cruse (DGE), dated February 11,1997,
" Issuance of Amendments for CCNPP Unit No I (TAC No M97341) and Unit No. 2 (TAC No, M97342)" [ Amendment Nos. 219/196]
- 19. Letter from Mr. C. 11. Cruse (BGE) to NR. Document Control Desk, dated November 26,1996," License Amendment Request; Adoption of 10 CFR Part 50, Appendix J, Option B for Type B and C Testing"
- 20. CCNPP " Pre Evaluation Results for the Containment Isolation Group (#013, 029, 037,051, 069,071)," Revision 0, March 19,1997 21 CCNPP Drawing No. 60714S110002," Plant Fire Protection System Auxiliary and Containment Buildings," Revision 23, May 31,1996
- 22. CCNPP Drawing No. 60728S110002, " Plant Heating System Auxiliary Building and Containment," Revision 22, May 31,1996
- 23. CCNPP Drawing No. 60729S11000)," Reactor Coolant System," Revision 61, August 5,1996
- 24. CCNPP Drawing No. 62729S11000), " Reactor Coolant System," Revision 68, December 12,1996
- 25. CCNPP Drawing No. 60733S110002," Auxiliary Building, Waste Processing Equipment, and Area Drains," Revision 23, January 22,1997
- 26. CCNPP Drawing No. 60733S110004, " Miscellaneous Containment Drains, Sump Piping, and Reactor Coolant Pump Lube Oil Collection System," Revision 8, August 15,1996
- 27. CCNPP Drawing No. 60734S110001, " Reactor Coolant Waste Processing Systems,"
Revision 31 September 9,1996
- 28. CCNPP Drawing No. 60717S110002,"Well Water, Pretreated Water, Demineralized Water and Condensate Storage System," Revision 33, November 26.1996
- 29. CCNPP Unit 1(2) Technical Specifications, Amendment No. 217(194), December 10,1996
- 30. CCNPP Serveillance Test Procedure STP-M-571G 1, " Local Leak Rate Test, Penetrations 9 (Cont Spray),10 (Cont Spray),23 (RC Drain Tank),24 (PXR Quench Tank),37 (Plam Service Water,39 (SI Test),(Unit 1)," Revision 0, May 17,1991.
- 31. CCNPP Surveillance Test Procedure STP-M-571G-2, " Local Leak Rate Test, Penetrations 9 (Cont Spray),10 (Cont Spray),23 (RC Drain Tank),24 (PXR Quench Tank),37 (Plant Service Water),39 (SI Test),(Unit 2)," Revision 0, October 17,1991
- 32. CCNPP Surveillance Test Procedure STP-M 571M-1, " Local Leak Rate Test, 44 (Fire Protection)(Unit 1)," Revision 0, May 16,1991 Application for 1.icense Renewal 5.5 26 Calvert Cliffs Nuclear Power Plant
O ATTACllMENT m ,
APPENDIX A - TECIINICAL INFORMATION 5.5 - CONTAINMENT ISOLATION GROUP
- 33. CCNPP Surveillance Test Procedure STP-M 571M 2, " Local Leak Rate Test, 44 (Fire Protectivi., (Unit 2)," Revision 0,0:tober 17,1991
- 34. CCNPP Surveillance Test Procedure STP-M 571 A-1," Local Leak Rate Test, Penetrations l A (RC & PZR Sampling), IB (Cont Vent lleader), IC (RC Pump Seals), (Unit 1)," Revision 0, May 16,1991
- 35. CCNPP Surveillance Test Procedure STP M 571 A-2, " Local Leak Rate Test, Penetrations l A (RC & PZR Sampling),1B (Cont Vent lleader), IC (RC Pump Seals), (Unit 2)," Revision 0, October 17,1991
- 36. CCNPP Surveillance Test Procedure STP M 571D 1, " Local Leak Rate Test, Penetration 8 (Containment Sump),(Unit 1)," Revision 0, May 16,1991
- 37. CCNPP Surveillance Test Procedure STP M 571D-2, " Local Laak Rete Test, Penetration 8 (Containment Sump), (Unit 2)," Revision 0, October 17,1991
- 38. CCNPP Surveillance Test Procedure STP-M-571E-1," Local Leak Fate Test, Penetrations 15 (Purge Air Monitor),16,18 (Comp Cooling Water), 38 (Demin Water), 59, 61 (Refueling Pool), 60 (STM to RX head Cleanup), 62, 64 (Unit IIe ters;, (Unit 1)," Revision 0 May 17,1991
- 39. CCNPP Surveillance Test Procedure STP M 571E-2," Local Leak Rate Test, Penetrations 15 (Purge Air Monitor),16,18 (Comp Cooling Water), 38 (Demin Water), 59, 61 (Refueling Pool), 60 (STM to RX head Cleanup), 62, 64 (Unit IIcaters), (Unit 2)," Revision 0, October 17,1991.
- 40. 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."
- 41. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated January 16,1996,
" License Amendment Request: Adoption of 10 CFR Part 50, Appendix J, Option B for Type A Testing"
- 42. CCNPP Administrative Procedure MN-3 301, " Boric . id Corrosion Inspection Program,"
Revision 1/C'iange 0, December 15,1994
- 43. "BGE Quality Assurance Policy for CCNPP," Revision 48, March 28,1997
- 44. Nuclear Performance Assessment Department Guideline NPADG-5," Master Assessment Plan (M AP) Development and Control," Revision I, July 8,1997
- 45. CCNPP Administrative Procedure QL-3103," Assessments," Revision 0, July 1,1997 l
Application for License Renewal 5.5-27 Calvert Cliffs Nuclear Power Plant
e-ATTACHMENT (3)
APPENDIX A - TECHNICAL INFORMATION 5.13 - NUCLEAR STEAM SUPPLY SYSTEM SAMPLING SYSTEM Haltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant Novemher 14,1997
4 A*ITACIIMENT m
' APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SJ1PLING SYSTEM 5.13 Nuclear Steam Supply System Sampling System This is a section of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), addressing the Nuclear Steam Supply System (NSSS) Sampling System. The NSSS Sampling System was evaluated in accordance with the Calvert Cliffs Nuclear Power Plant (CCNPP) Integrated Plant Assessment (IPA) Methodology described in Section 2.0 of the BGE LRA. These sections are prepared independently and will, collectively, comprise the entire BGE LRA.
5.13.1 Scoping System level scoping describes conceptual boundaries for plant systems and structures, develops screening tools which capture the 10 CFR 54.4(a) scoping criteria, and then applies the tools to identify systems and structures within the scope of license renewal. Component level scoping describes the
' components within the boundaries of those systems and structures that contribute to the intended functions. Scoping to determine components subject to aging management review (AMR) begins with a listing of passive intended functions and then dispositions the component types as either only associated with active functions, subject to replacement, or subject to AMR either in this report or another report.
Representative historical operating experience pertinent to aging is included in appropriate areas to provide insight supporting the aging management demonstrations. This operating experience was obtained through key-word ser.rches of BGE's electronic database ofinformation on the CCNPP dockets and through documented discussions with currently assigned cognizant CCNPP personnel, Section 5.13.1.1 presents the results of the system cel scoping: 5.13.1.2 the results of the component level scoping: and 5.13.1.3 the results of scoping to octermine components subject to an AMR.
5.13.1.1 System Level Scoping This section begins with a description of the system which includes the boundaries of the system as it was scoped. The intended functions of the system are listed and are used to define what portions of the system are within the scope oflicense renewal.
l
! System Descrintion/ Conceptual Boundaries The NSSS Sampling System is designed to permit the sampling of liquids, steam, and gases for radioactive and chemical control of plant primary fluids. [ Reference 1, Table 1; Reference 2, Section 9.6.1] Five subsystems comprise the NSSS Sampling System: reactor coolant sampling, steam generator blowdown sampling, radioactive miscellaneous waste sampling, gas analyzing sampling, and post-accident sampling. [ Reference 1, Table 1]
. Reactor Coolant sampling: The purpose of the reactor coolant sampling subsystem is to sample liquids and gases for analysis and control of chemical and radiochemical concentrations.
[ Reference 3, Section 1.1.1] The reactor coolant sampling subsys'.em consists of a stainless steel sink cnclosed inside a hood. The hood is ventilated by an individual blower through a high-L efficiency filter. The hood contains piping, valves, coo'ers, instrumentation, and sample vessels necessary to take liquid and gaseous samples from various systems. One hood is installed for each unit inside the Sampling Room in the Auxiliary Building. Samples obtained from two locations in the pressurizer (liquid, vapor) and one from the Reactor Coolant System (RCS) hot leg can be controlled by three handswitches. Should any one of the remotely-operated sampling Application for License Renewal 5.13 1 Calvert Cliffs Nuclear Power Plant l
.4 .
ATTACHMENT (3) l APPENDIX A - TECIINICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM valves fall to close after a sample is taken, a second remotely-operated valve can be shut from the ,
Control Room. These valves are also closed by a Safety injection Actuation Signal. The l remotely-operated valves are backed up by manually oper ted valves at tiv reactor coolant r, ample hood, liigh pressure samples flow through metering valves in order to reduce their pressure, liigh-temperature samples can be cooled in a sample cooler supplied with water from the Component Cooling (rC) System. [ Reference 2, Sectio 9.5.2.1]
When 12 and 18-month fuel cycles were used at CCNPP, pressurizer samples were required at the beginning of each fuel cycle to assist in startup testing (* .. rod woith testirg). Since these samples are no longer being taken, the pressurizer surge lin. . d va;or space sample headers ar; normally isolate:1 by manually operatad valves. [ Reference 4]
The original control valves (CVs) in the RCS hot leg sampling lines exhibited excessive scat / disk leakage caused by insufficient actuator spJ.ng closure force. This caused full system pressure drop to be taken across the upstream isolation valve, resulting in steam flashing and erosion of the valve seating surfaces. Replacement of these CVs was coinpleted in 1993.
The reactor coolant sampling subsystem also provides a means for obtaining liquid samples from the RCS or the containment sump in the post-accident environment. [ Reference 5] To provide this capability, modifications included rerouting and installation of tubing in the reactor coolant sampling subsystem hoods to permit draining post-accident liquid samples to the reactor coolant drain tank (RCDT) in lieu of the normal return path to the volume control tank. [Refcrence 6]
This capability can be used to obtain samples from the RCS or the Safety Injection (SI) System.
[ Reference 2, Section 9.6.2.2]
. Steam Generator Blowdown Samnling: The purpose of the steam generator blowdown sampling subsystem is to provide a means for sampling of liquids from the steam generators to detect conditions that cause carryover, corresion and fouling of heat transfer surfaces, and to aid in detection of a possible reactor coolant-to-steam genciator leak. [Refereace 3, Section 1.1.1) This subsystem also provides a means for sampling reactor coolant makeup water. [ Reference 7] The steam generator blowdown sampling subsystem consists of one conditioning rack-panel unit and one ventilating hood installed for each unit; these are located inside the same Sampling Room as the reactor coolant sample hoods. [ Reference 2, Section 9.6.2.3]
The conditioning rack section of the steam generator blowdown subsystem contains isolation vahes, primary coolers, rod-in-tube devices, an isothermal bath, and chiller. Iligh pressure samples are passed through a pressure-reducing valve (rod in-tube type) located downstream of the primary coolers and upstream of the isothermal bath. High-temperature samples first pass through a primary cooler (supplied with water from the CC System) and then through the isothermal bath. All samples pass through the isothermal bath, a large tank of demineralized water where temperature is maintained by a chiller unit. The chiller is supplied with cooling water from the CC System. Sample outlets from the conditioning rack are connected to the ventilating hood. [ Reference 2, Section 9.6.2.3] The original sample chillers were replaced in 1992 when leakage could not be repaired.
Application for License Renewal 5.13 2 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (3)
APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM ,
'l The panel section of the steam generator blowdown subsystem contains conductivity and pH monitors, three hand sw9ches for pressurizer sample selection, chiller controls, and an annunciator. The pH and conductivity of water in the steam generator blowdown sample lines are continuously monitored and alarm on abnormal conditions. In addition, pH and conductivity are trended on the cor,ipuw-based display in the chemistry laboratory. High sample temperature (downstream cf the isothermal bath) actuates a common alarm point. Any point alarming on the local annunciator will acteate a master trouble alarm in the Control Room. [ Reference 2, Section 9.6.2.3)
The ventilating hood contains two stainless steel sinks and is ventilated by an individual blower through a high-efficiency filter. The radioactive miscellaneous sample subsystem, discussed below, is also located inside the ventilating hood for Unit 1. The steam generator blowdown part of the hood contairs all piping, grab sample valves, instrumentation including pH and conductivity cells, and other equipment necessary to obtain samples for determining the chemical and radiochemical content of the steam generator blowdown and reactor coolant makeup water.
[ Reference 2, Section 9.6.2.3]
- Radioactive Miscellaneous Waste Samnling: The purpose of the rr "cactive miscellaneous waste sampling subsystem is to provide a means for sampling ofliquids from various radioactive waste processing systems to determine the chemical and radiochemical content prior to discharge, and to aid in evaluation of waste system component performance. [ Reference 3, Section 1.1.1) The radioactive miscellaneous waste sampling subsystem is located inside the ventilating hood for the Unit i steam generator blowdown sampling subsystem, and is used to obtain samples from which the chemical and radiochemical content of miscellaneous waste is determined. This subsystem is common to both units. All samples are low pressure and are cooled, as necessary, in sample coolers supplied with water from the CC System. This part of the hood contains isolation valves, piping, valves, and instrumentation necessary for obtaining liquid samples from both units. The analyses of these samples are performed in the laboratory located in the Auxiliary Building.
(Reference 2, Section 9.6.2.4]
. Gas Analyzing Samnling: The purpose of the gas analyzing sampling subsystem is to provide a means for sampling of gases to determine: (a) the hydrogen concentration of the containment atmosphere and the reactor coolant waste tanks; and (b) the oxygen concentration in the pressurizer quench tanks and various miscellaneous waste systems. [ Reference 2, Section 9.6.2.6; Reference 3, Section 1.1.1] The gas analyzing sampling subsystem is installed in the Cryogenics Room located in the Auxiliary Building at Elevation-10'-0" and consists of two hydrogen analyzer cabinets, two hydrogen sample select cabineta with separate manifolds for isolation valves and sample selection solenoid valves, and one oxygen analyzer cabinet with a manifold for isolation valves. The two analyzer cabinets used for hydrogen measurement each include a sample pump, cooler, tubing, valves, and analyzer elements. Separate control cabinets at Elevation 45'-0" in the Auxiliary Building include a hydrogen sequencer panel for selection of individual readouts, a sequencer for control of sample solenoid valves, local and remote analyzer indicators with a multipoint recorder in the Control Room, and alarm contacts for activation of a master alarm in the Comrol Room. The analyzer cabinet used for oxygen grab sample measurement includes a sample pump, cooler, piping, valves, and a sample syringe port. An exhaust system on the oxygen analyzer cabinet purges any hydrogen that may leak into the Application for License Renewal 5.13-3 Calvert Cliffs Nuclear Power Plant
ATTACHMENT W APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SAMPl>NG SYSTEM cabinet. Sample selection from the six containment locations for each hydrogen analyzer cabinet and the two quench tank locations for the oxygen analyzer cabinet can be controlled through remotely operated solenoid valves. [ Reference 2, Section 9.6.2.6]
The gas analyzing sampling subsystem also provides a means for obtaining grab samples of gases in the containment atmosphere in the post-accident environment. [ Reference 3, Section 1.1.1) To provide this capability, a sample vessel was placed into each of the sampling lines coming from the west side of the Unit I and 2 Containment Buildings at Elevation 135'-0". These sample vessels are located at Elevation 45'-0" in the Auxiliary Building and allow syringe samples to be taken and analyzed in the laboratory. [ Reference 2, Section 9.6.2.6]
The original hydrogen analyzing subsystem required modification because the sampling equipment did not meet the range (0% to 10%) or post accident accessibility requirements of NUREG-0737," Clarification of TMI Action Plan Requirements." In 1982, the control cabinets were relocated to Elevation 45' 0" in the Auxiliary Building, and two new cabinets with high-range hydrogen analyzers were installed at Elevation -10'-0" in the Auxiliary Building.
[ Reference 8]
- Post-Accident Samoling: The purpose of the skid-mounted Post-Accident Sampling System (PASS; no longer in service at CCNPP) was to provide a means for remote sampling of liquids and gases in the post accident environment. [ Reference 3, Section 1.1.1] The PASS, located at E'evation 45'-0" in the Auxiliary Building, contains piping, valves, coolers, and instrumentation n:cessary to sample either Unit 1 or Unit 2 RCS via either the normal RCS sampling line, or Unh 1 or Unit 2 Containment sumps via the low pressure SI System header. [ Reference 2, Section 9.6.2.2]
The PASS was installed initially in 1982 to meet the requirements of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations," and upgrades to conform to NUREG-0737 requirements were completed in 1987. [ References 6 and 9] In 1985, multiple subcomponent failures demonstrated the PASS to be a maintenance-intensive and unreliable system. [ Reference 10] For these reasons, BGE modified the reactor coolant sampling and gas analyzing subsystems to provide a post-accident capability that relies, with only one exception, on grab sample analyses to meet regulatory requirements for both the RCS and Containment atmosphere. [ Reference 5] The original PASS is no longer in service, and the original PASS sample lines from the RCS and SI System have been disconnected to eliminate the potential of cross-contamination of samples between units. [ Reference 11]
Aside from the component replacements noted above, operating experience relative to the NSSS Sampling System has shown that there have been no additional instances of significant degradation due to aging of passive, long-lived NSSS Sampling System components.
Application for License Renewal 5.13-4 Calvert Cliffs Nuclear Power Plant t
ATTACHMENT (h APPENDIX A - TECliNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM The following general categories of equipment and device comprise the five subsystems of the NSSS
. Sampling System: (Reference 3, Section 1.1.2]
Accumulators To store pressurized gases needed to perform sampling operations; Air dryers To remove moisture from air systems; Piping To convey Culds to perform system sample functions; Valves and valve operators CVs, check valves (CKVs), hand valves (IIVs), and motor-operated valves, to provide system alignment, isolation, and protection; Panels To provide support for components; Instruments To measure system parameters, provide control / actuation signals; Sample Vessels To store sample Guids as part of the sample collection and testing process; licat Exchangers (IIXs) To remove heat from the sample Guids; and Pumps To transfer sample fluids for testing purposes.
System Interfaces The NSSS Sampling System interfaces with the following systems and components: (Reference 3, Section 1.1.2; References 7 and 12 through 15]
. RCS (sample points);
- Demineralized Water and Condensate Storage System (water for make-up to the isothermal baths);
- CC System (water for cooling samples);
- SI System (sample points);
- Spent Fuel Pool Cooling System (sample points);
Liquid Waste System (sample points, sample drainage, sample return flow);
. Chemical and Volu ne Control System (sample points, sample return Dow);
a Containment (sample points, sample return Dow);
- Main plant vent (sample return Dow);
- Waste Gas System (sample points, sample return Dow); and
- Main Steam System (steam generator blowdown sample points).
System S:oning Results The NSSS Sampling System is in scope for license renewal based on 10 CFR 54.4(a). The fo!!owing intended functions of the NSSS Sampling System were determined based on the requirements of l54.4(a)(1) and (2) in accordance with the CCNPP IPA Methodology Section 4.1.1: [ Reference 3, Section 1.1.3]
+
To maintain the pressure boundary of the system (liquid and/or gas);
To provide containmem isolation of the NSSS Sampling System during a loss-of coolant accident; Application for License Renewal 5.13-5 Calvert ClitTs Nuclear Power Plant
A'ITACHMENT 0)
APPENDIX A - TECIINICAL INFORMATION
$.13 - NSSS SAMPLING SYSTEM To sample and analyze containment hydrogen gas concentration following an accident;
- To provide the capability to operate RCS hot leg sample valves to obtain samples fcilowing a loss-of coolant accident;
- To provide seismic integrity and/or protection of safety-related components; To provide closure of the RCS hot leg sample isolation valve on receipt of a Safety injection Actuation Signal;
- To maintain electrical continuity and/or provide protection of the electrical system; and
. To restrict flow to a specified value in support of a Design Basis Event response.
The following intended functions of the NSSS Sampling System were determined based on the requirements of Q54.4(a)(3): [ Reference 3, Section 1.1.3]
- To provide information used to assess the environs and plant conditions during and following an accident; and
- To maintain the functionality of electrical components as addressed by the Environmental Qualification (EQ) Program.
All components of the NSSS Sampling System that support intended functions based on the requirements of QS4 4(a)(3) are also safety-re'ated.
The NSSS Sampling System must safely perform intended functions during normal and Design Basis Event conditions. The gas analyzing subsystem and those portions ofIlXs that form part of the pressure bcundary for the CC System are designed to meet Seismic Class I requirements. [ References 7 and 12 through 16] Additionally, the following valves are designed to meet Seismic Class I requirements because they constitute the boundary between piping in interfacing systems, which complies with Section ill of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, and non Class piping in the cabinets: [ Reference 2, Section 9.6.3]
- Charging pump discharge header entry valve (l[2]IlVPS-172);
Low pressure Si sample header entry valve (l[2]IIVPS 193);
Spent fuel pool filter inlet entry valve (0llVPS-226);
Spent fuel pool demineralizer outlet entry valve (0llVPS-229);
- Steam generator bottom blowdown header entry valves (l[2]lIVPS-126,1[2]IIVPS-139);
- Steam generator surface blowdown header entry valves (l[2]IIVPS-128,1[2]IIVPS-137); and
- Steam generator header rod-in-tube valves (l[2]IlVPS-129,1[2]IIVPS-140).
The remaining components within the associated cabinets for the reactor coolant sampling, steam generator blowdown sampling, and miscellaneous waste sampling subsystems are designed to meet Seismic Class 11 requirements. [ Reference 2, Section 9.6.3]
The RCS sample piping within the NSSS Sampling System boundaries (i.e., piping in the pressurizer surge line and RCS hot leg sample headers between the sample points and the second manual isolation Applica. tion Ihr License Renewal 5.13-6 Calvert Cliffs Nuclear Power Plant
ATTACHMENT G)
APPENDIX A - TECilNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM valve) complies with the design requirements for Class I piping in American National Standards Institute (ANSI) Nuclear Power Piping Code B31,7.1969. [ Reference 7; Reference 17, Piping Class CC 16]
Other piping within the NSSS Sampling System boundaries (i.e., containment penetration portions of the reactor coolant sampling and gas analyzing sampling subsystems) complies with Class Il requirements of ANSI B31.7. [ References 7,13, and 14; Reference 18, Piping Class CC 8; Reference 19, Piping Class HC-43) Tubing meets American Society for Testing and Materials (ASTM) Standard A450-68, which requires an eddy-current test for new and replacement tubing. [ Reference 2, Section 9.6.3]
5.13.1.2 Con.ponent Level Scoping Based on the intended functions listed above, the portion of the NSSS Sampling System that is within the scope of license renewal includes all safety-related components (electrical, mechanical, and instrument) and their supports. [ Reference 20, Table 2; References 7 and 12 through 15] The safety-relsted portions of each subsystem are described below:
. The RCS sample header isolation CVs (speciGcally, those isolating piping from the pressurizer surge line, the pressurizer vapor space, and the RCS hot leg), the RCS sample isolatioa CV, all intervening piping, and the test / vent / drain root valves connected to the intervening piping are safety-related. The containment isolation solenoid-operated valves (SVs) in the sample retum lines from the reactor coolant sample hoods to the RCDT, and the piping between these valves and RCS tubing inside containment, are also safety-related. These components form part of the containment pressure boundary.
In addition, piping in the RCS sample headers 'etween e the RCS and the CVs, test valves connected to this piping, and isolation valves in the Dow path are safety-related because they con,titute the boundary from ASME Section 111 piping to non-Class piping.
In each of the reactor coolant sample hoods, the sample cooler is also safety-related; the shell body aw' tubes form part of the pressure boundary for the CC System, Additionally, the hand valves in the sample lines from the charging pump discharge and the low pressure SI pump discharge constitute boundaries from ASME Section til piping to non-Class piping and are safety-related. The remaining components in the reactor coolant sample hoods are non-safety-related.
For a simpliGed diagram of the reactor coolant sampling subsystem components described above, refer to Figure 5.13-1.
The steam generator blowdown sampling subsystem components, from the sample points in the steam generator blowdown piping (i.e., two lo.;ations for each steam generator (surface and bottom blowdown]), through the tubes in the sample coolers, up to and including the rod-in-tube pressure-reducing ilVs downstream of the sample coolers in the conditioning racks, are safety-related. Rese components form part of the pressure boundary for the Main Steam System. The shcIl body and tubes in these sample coolers also form part of the pressure boundary of the CC System. Additionally, the piping and tubing in the sample chillers in each of the conditioning rack panels are safety-related; they also form part of the pressure boundary for the CC System.
The radioactive miscelk.neous waste sampling subsystem IIVs in the spent fuel pool Glter and demineralizer sampling lines constitute boundaries from ASME Section 111 piping to non-Class piping and are safety-related. The four sample coolers in this subsystem are also safety-related; the shell body and tubes form part of the pressure boundary for the CC System.
Application for License Renewal 5.13-7 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (3) .
APPENDIX A - TECIINICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM WSLR - withm the scope of heense renewal BOLD LINES indicate components that From OTHER see WSLR for NSSS Samphng Sysh From LPSI PUMP OtSCHARGE SAMPLE SAMFLE POINT ...-- POINTS dotted LINES andcate Components that (WSLR - etMr to Secean 5.15) P (not WSLR) are part of interfacpJ systems. From CHARGING Pt d b h I
aes mea ! OtSCHARGE SAMPLE PO4NT I
" **' " " " acs = = = (WSLR - refer to Section 52) I E I From I 'd lk I RCS HOT LEG 4[OJ
^ O A
^ j j SAMPLE POtNT (WSLR - refer to Seebon 41)
[
ryg l
Y t
,_ . = = - FronVTo PRESSUR1ZER SURGE LINE -gjO*r4- 9, O,-
- COMPONENT e SAMPLE POINT COOLING WATER- ** SAMPLE (WSLR - refer to Seebon 4.1) SYSTEM 4 --- - COOLER U
(WSLR - refer to Sechon
- 53) _
"& " " ' ' " UQU D GRAB e ES F ANAL IS rm PRESSURtZF.R VAPOR SPACE 4[03 O SAM *LE pot *8T (WSLR - tv fer to Section 4.1)
" " " y REACTOR COOLANT O
Control Valve ,
(WSLR - refer to Sechon 6 2)
Hand Valve ,....... ; Q, e gy.
To wue .ew 'o ** To
~ ~ " ' " "
d So.enoiderated vaeve "$$"u yT l votuME CONTROL TANK (not WSLR)
?
FIGURE 5.13-1 REACTOR COOLANT SAMPLING SUBSYSTEM (SIMPLIFIED DIAGRAM - FOR INFORMATION ONLY)
Application for License Renewal 5.13-8 Calvert Cliffs Nuclear Power Plant
- _ - - - - - - - - _ _ - - - _ _ - - _ _ _ - - - - . - - - _ - - - _ - - - - - - - - - - _- --. , m
l .
AITAC? MENT m APPENDIX A - TECIINICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM
)
All gas analyzing sampling subsystem piring and components associated with sampling and analysis of gases for hydrogen concentration are safety-related. This comprises separate piping / tubing in each Unit that leads from six sample points in containment, through containment -
isolation SVs and sample selection SVs to the hydrogen analyzer cabinet, through the analysis equipment in the cabinet, and through the gas return lines leading: (a) through containment isolation SVs to the Containment atmosphere; or (b) through SVs and CKVs to the common Auxiliary Building lleating i Ventilating System. He additional piping / tubing and normally-closed ilVs that connect the sampling lines coming from the west side of the Unit t and 2 Containment Buildings at Elevation 135'-0" to each Unit's post-accident hydrogen gas sample vessel, as well as bottled gas cylinders and associated equipment used for instrument calibration, are also included. The containment isolation SVs, as well as the piping and components between them, form part of the containment pressure boundary. Additionally, all components in the flowpath described above are required to maintain the system pressure boundary when the hydrogen analyzers are placed in operation.
Additionally, in the lines provided for sampling oxygen concentration for each Unit's pressurizer quench tank, the containment isolation SVs and the piping between these valves and the quench ,
tank are safety-related. These components form part of the containment pressure boundary, The sample cooler in the oxygen analyzer cabinet is also =fety-related; the shell body and tubes form part of the pressure boundary for the CC System. The remaining components associated with sampling and analysis of gases for oxygen concentration are non-safety-related.
. Because it is no longer in service and has been isoiated from other safety-re'ated equipment, all components in the original PASS are non safety-related.
The following 33 device types in the NSSS Sampling System have been designated as within the scope oflicense renewal because they have at least one intended function: [ Reference 3, Section 2.2 Table 2-1; Reference 20, Table 2]
Class "CC" Piping (stainless steel, primary rating 1500 psig at i125 F)
Class "HC" Piping (stainless steel, primary rating 150 psig at 500 F)
Analyzer Alarm Control Valve Handswitch Pump / Driver Assembly Accumulator Control Valve Operator Hand Valve Relay Analyzer Element Air Dryer Heat Exchanger Solenoid Valve Analyzer Indicator Voltage / Current Device Power Lamp indicator Temperature Controller Analyzer Recorder Flow Indicator Pressure Control Valve Heater Analyzer Converter (relay) Flow Indicator Controller Pressure Indicator Position Indicating Lamp Circuit Breaker Flow Orifice Panel Position Switch Check Valve Fuse Pressure Switch Some components in the NSSS Sampling System are common to many other plant systems and have been included in separate sections of the BGE LRA that address those components as commoditie? for the entire plant. These components include the following: [ Reference 3, Section 3.2; Reference 21, Section 3.0]
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.
Applicatior icense Renewal 5.13-9 Calvert Cliffs Nuclear Power Plant
1 ATTACHMENT (3)
APPENDIX A - TECIINICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM
- Electrical control and power cabling are evaluated for the effects of aging in the Electrical Cables Commodity Evaluation in Section 6.1 of the BGE LRA. This commodity evaluation completely addresses the passive intended function entitled "to maintain electrical continuity and/or provide protection of the electrical system" for the NSSS Sampling System.
- Small-bore piping and tubing in the NSSS Sampling System and the associated tubing supports are evaluated for the effects of aging in the instrument Lines Commodity Evaluation in Section 6.4 of the BGE LRA. This commodity evaluation partially addresses the passive intended function entitled "to maintain the pressure boundary of the system (liquid and/or gas)" for the NSSS Sampling System.
5.13.1.3 Components Subject to AMR This section describes the components within the NSSS Sampling System that are subject to AMR. It begins with a listing of passive intended functions and then dispositions the device types as either only associated with active functions, subject to replacement, evaluated in other reports, evaluated in commodity reports, or remaining to be evaluated for aging manageraent in this section.
Passive Intended Functions in accordance with CCNPP IPA Methodology Section 5.1, the following NSSS Sampling System functions were determined to be passive: (Reference 3, Table 3 1]
- Ma!ntain the pressure boundary of the system (liquid and/or gas);
- Provide containment isolation of the system during a loss-of-coolant accident;
- Provide seismic integrity and/or protection af safety-related components;
- Maintain electrical continuity and/or provide protection of the electrical system; and
- Restrict flow to a specified value in support of a Design Basis Event response.
Device Types Subject to AMR Cf the 33 device types within the scope oflicense renewal for the NSSS Sampling System:
- Fourteen device types were as;ociated with only active functions: Analyzer Alann, Analyzer Indicator, Analyzer Recorder, Analyzer Converter (relay), Circuit Breaker, Voltage / Current Device, Fuse, llandswitch, Power Lamp indicator, Relay, Temperature Controller, Heater, Position Indicating Lamp, and Position SwitcN
- No device ty pes were identified as subject to replacement over the period of extended operation;
- No device types in this system were evaluated in the AMR for a system addressed in another section of the BGE LRA; and One device type, Panel, is associated with a separate commodity evaluation. Panels in the NSSS Sampling System are subject to an AMR and are evaluated separately in the Electrical Panels Commodity Evaluation in Section 6.2 of tne BGE LRA.
The remaining 18 device types, listed in Table 5.13-1, are subject to AMR and are included in the scope of this section. [ Reference 3, Table 3-2; Reference 20, Table 2]
Application for License Renewal 5.13-10 Calvert Cliffs Nuclear Power Plant
ATIACilMENT- (,1) ,
APPENDIX A TECHNICAL INFORMATION 5.13 . NSSS SAMPLING SYSTEM llaltimore Gas and Electric Company may elect to replace components for which the AMR identlSes that ,
further analysis or examination is needed. In accordance with the License Renewal Rule, components subject to replacement based on quali0ed life or specified time period would not be subject to AMR.
Table 5.131 NSSS SAMPLING SYSTEM DEVICE TYPES SUILIECT TO AMR Class CC piping Flow Indicator Controller Clasr4 IIC Piping Flow Orince Accumulator lland Valve Analyrer Element lleat Exchanger Check Valve Pressure Control Valve Control Valve Pressure Indicator Control Valve Operator Pressure Switch Air Dryer Pump / Driver Assembly Flow Indicator Solenoid Valve 5.13.2 Aging Management The list of potential Al e Related Degradation Mechanisms (ARDMs) identified for the NSSS Sampling System components is ghen in Table 5.13 2, with plausible ARDMs identined by a check mark (/) in the appropriate device type column. [ Reference 3, Attachment $s and Attachment 6s] A check mark indicates that the ARDM applies to at least one component for the device type listed. For efficiency in presenting the results of these evaluations in this report, ARDM/ device type combinations are grouped together where there are similar characteristics and the discussion is applicable to all component 1.
Table 5.13 2 also identifies the group to which each ARDM/ device type combination belongs.
Exceptions are noted where appropriate. The following groups have beer selected for the NSSS Sampling System: ;
Group 1: gcatralsstrosion of utcrual surfacer due to leak ;ge of Lorated water; Group 2: citylce corrosion and nitting ofinternal sutfacts exposed to chemically treated water; Group 3: genctalcottmlon of internal surfaccs for control valve operators (CVOPsmposed to air from the Compressed Air System; Group 4: fatisus for piping and valves associated with sampling the RCS hot leg; Group 5 clastomer destadation for valve internals; and Group 6: wcar for CVs associated with sampling the RCS hot leg.
Application for License Renewal 5.13 11 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (3)
APPENDIX A - TECHNICAL INFORMATION 5.I3 - NSSS SAMPLING SYSTEM Table 5.I3-2 POTENTIAL AND PLAUSIBLE ARDMs FOR THE NSSS SAMPLING SYSTEM Device Types Not PotestinI ARDMs _i . A A C C C D F F F H H P P P P S M C HjC E K V V R I I O V X C I S U y k C C'C y O P
Y C i
V M P
'W Cavitsion Erosion ! ! i x Corrosion Fatigue i i i x Crevice Corrosion i /(2)i/(2)! -
/(2)
Erosion Corrosion i i i
- x Fatigue /(4)! /(4) /(4)i i Fouling i i x i Galva. tic Corrosion ! i i x !
General Corrosion ! ! /(1) /(3) /(I)!/(1)! c liydrogen Damage i i i i x Intergranular Attack i i i x ,
Microbiologically-induced Coriosion , ! x Particulate Wear Erosion i ! i x Pitting ! /(2) :/(2)! /(2) i Radiation Damage i i x Elastomer Degradation i /(5) !
Saline Water Attack i
! i x Selective Leaching i i i x Stress Corrosion Cracking 5 i i x Stress Relaxation i + +
! x Thermal Damage i i i x Thermal Embrittlement 4 ! x Wear i /(6) i
/ mdicates plausible ARDM determination for this device type Note- Not every component within the device types listed here may be (number) indicates the group m.which this ARDM/ device type susceptible to a given ARDM. This is because (uugv.wh (and sub > - - 6 a device y a M dw faW fm h combmation is evaluated.
me a m ed m h - - h e fw each device type will be indic.ned in the aging m.u%ugni subsection for each ARDM discussed in this report.
Application for License Renewal 5.13-12 Caivert Clifts Nuclear Power Plant
AITACllMENT (3) ~
APPENDIX A TECHNICAL INFOILMATION ,
5.13 . NSSS SAMPLING SYSTEM Sample coolers and chillers a.e included as components in the NSSS Sampling System because they perform a cooling function by reducing the temperature of the samples obtained. These llXs are within the scope of license renewal because they provide pressure boundary integrity for the interfacing systems, which is a passive intended function of both the CC System (which supplies cooling water to i sample coolers in Groups 1 and 2) and the Main Steam System (which includes sample points for the steam generator blowdown sampling subsystem, with IlXs evaluated in Group 2). [ Reference 3, Attachment 3s for llXs)
Control valve opere1ers are evaluated as components in the NSSS Sampliag System because they
. perform positioning functions that are within the < cope oflicense renewal for CVs in the RCS sample header (i.e., to operate following a loss-of coolant accident, to close on applicable Safety injection Actuation Signal, and ta close following loss of AC power). These functions are active and do not require AMR for the CV0Ps. Ilowever, these componend also provide pressure boundary integrity, which is a passive intended function of the Compressed Air lystem (which supplies instrument air [lA] ,
l to the CVOPs, evaluated in Group 3). { Reference 3, Attachn ent 3s for CVOPs]
The following is a discussion of the aging management demonstration process for each group identified above. It is presented by group and includes a discussion of materials and environment, aging mechanism effects, methods of managing aging, aging management program (s), and eging management demonstration.
Group 1 -(general corrosion of esternal surfaces) . Materials and Environment Group I comprises the various NSSS Sampling System components that are exposed to climate controlled air in the Auxiliary Dullding or the Containment and whose external surfaces are subject to general corrosion. The components in this group are included in the llX, CV, s.nd ilV device types. All of these components provide the passive intended function of maintaining the system pressure boundary. [ Reference 3, Attachment 1] The applicable subcomponents in these device types are constructed of the following materials: [ Reference 3, Attachments 4 and 5 for llXs, CVs, ilVs]
IlXs . carbon steel end plates / piping for the miscellaneous waste evaporator concentrate pump discharge sample cooler (part of the radioactive miscellaneous waste sampling subsystem) and for sample coolers in the reactor coolant sample hoods;
- CVs . carbon steel nuts, alloy steel studs for CVs in the reactor coolant sampling subsystem; and a llVs carbon steci nds, alloy steel studs for normally closed isolation / vent / drain /t est IIVs in the RCS sample header, and for noimally open root, backup IIVs in charging pump discharge sampling lines.
The external surfaces evaluated in Group 1 are not normally exposed to a corrosive cuvironment, but may be exposed to boric neid as a result of L*akage from the associated components. The possible efTect of such leakage is general corrosion of sus:cptible external surfaces. The sources of potential leakage from components in Group I are listed b110w:
For the miscell meous waste evaporator concentrate pump discharge sample cooler, borated water inside the tubes with a design pressure of 150 psig, and normal operating temperature of 130'F;
[ Reference 3, Attachment 3s for llXs: Reference 22, Piping Class llc-2]
Application for 1.icense Renewal 5.13 13 Calvert Cliffs Nuclear Power Plant
~ ,. . . , . . . . . . . _ _ _ _ _ , . _ _ . _,__
..__ _ _j
A"ITACJIMENT L1)
APPENI)IX A . TECilNICAL INFORMATION ,
5.13 - NSSS SAMPLING SYSTEM
- For llXs, CVs, and ilVs in the reactor coolant sampling subsystem, and for liVs in sampling lines from the charging pump, borated water with maximum operating pressures as high as 2485 psig (charging pump discharge), and normal operating temperatures as high as 653'F (saturation conditions in the pressurizer). [ Reference 3, Attachment 6s for llXs, CVs, llVs; Reference 173 Piping Class CC 16; Reference 18, Piping Classes CC 7 and CC 8]
For all components evaluated in Group 1. the external surfaces are exposed to an environment of climate controlled air in the Auxiliary llullding or the Containment. [ Reference 3, Attachment 3s] The containment atmosphere is applicable only to some of the CVs and ilVs in the RCS sample header.
During normal operation, temperature and humidity in the Auxiliary Ilullding do not exceed 160'F and 70%, respectively, [ Reference 23, page 54] for the general areas inside Containment, the maximum normal temperature and humidity values are 120'F and 70%, respectively. [ Reference 23, pages 29,30,62, and 63]
Group 1 -(general corrosion of external s[rfaces)- Aging Mechanism Effects General corrosion is thinning of a rnetal by the chemical attack of an aggressive environment at its surface. An 'mportant concern for pressurized water reactors is boric acid attack upon carbon steels and low alloy stects. General corrosion is not a concern for austenitic stainless steels. [ Reference 3, Attachment 7s for llX, valve]
General corrosion is plausible for all carbon steel and alloy steel subcomponents in this group.
Mechanicaljoints in pressure boundary subcomponents provide the opportunity for leakage of borated water onto exteraal carbon steel and alloy steel component surfaces (i.e., end plate / piping ferrous materials for llXs, carbon steel and alloy steel body / bonnet bolting for valves). These components are particularly susceptible to significant acceleration of corrosion when exposed to boric acid in the concentrations present in the reactor coolant sampling and radioactive miscellaneous waste sampling subsystems. [ Reference 3, Attachment 6s for llXs, CVs,llVs]
The result of this corrosion mechanism is reduction in the integrity of the corroded parts and a resulting increase in the likelihood of mechanical failure. If unmanaged, long term expos' ire to general corrosion could eventually result in loss of the pressure-retaining capability under current licensing basis (CLI3) design loading conditions.
Group 1 -(general corrosion of external surfaces)- Methods to Manage Aging Mitigation: 11oric acid corrosion can be mitigated by minimizing leakage. The susceptible areas of the NSSS Sampling System (i.e., bolted joints) can be routinely observed for signs of borated water leakage, and appropriate corrective action can be initiated as necessary to eliminate leakage, clean spill areas, and assess any corrosion. [ Reference 3, Attachment 6s for ilXs,ilVs]
Discosery: The efTects of corrosion are generally detectable by visual techniques. Visual inspections would need to be performed to detect corrosion associated with leakage of fluids onto the external surfaces of NSSS Sampling System components. [ Reference 3, Attachment 6s for ilXs,ilVs]
Application for License Renewal 5.13 14 Calvert Cliffs Nuclear Power Plant
ATTACitMENT m :
I APPENDIX A TECHNICAL INFORMATION ,
5.13 NSSS SAMPLING SYSTEM !
Group 1 -(general corrosion of esternal surfaces) . Aging Management Program (s)
Mitigation: ne CCNPP Iloric Acid Corrosion inspection (BACI) Program (MN 3 301) is credited with mitigating the effects of boric acid corrosion through timely discovery of leakage of borated water and t removal of any boric acid residue that is found. [ Reference 3, Attachment 8] This program requires visual inspection of the components containing boric acid for evidence of leaks, quanti 0 cation of any leakage indications, and removal of any leakage residue from component surfaces. [ Reference 24]
Further details on the BACI Program are detailed in the Discovery subsection below.
f Dhany: Discovery of boric acid leakage is ensured by the BACI Program. [ Reference 3 '
Attachment 8] His program also requires investigation of any leakage that is found. A visual examination of extemal surfaces is performed for components containing boric acid. [ Reference 24]
ne Inservice Inspection Program required the establishment of the BACI Program to systematically ensure that boric acid corrosion does not degrade the primary system boundary. [ Reference 25, page 23, Section 5.8.A.l.] ne program also applies to " valves in systems containing borated water which could leak onto Class I carbon steel components," and it identines other plant areas to be examined.
[ Reference 24, Section 5.111] The program controls examination, test methods, and actions to minimize the loss of htructural and pressure-retaining integrity of components due to boric acid corrosion.
[ Reference 24, page 7. Section 3.0.C] ne basis for the establishment of the program is Generic Letter 88 05,"floric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." [ Reference 24, Section 1.1) ne scope of the program is threefold in that it: (a) identines locations to be examined; (b) provides examination requirements and methods for the detection ofleaks; and (c) provides the responsibilities for initiating enginnring evaluations and necessary corrective actions. [ Reference 24, Section 1.2]
During each refuelirig outage, inservice inspection personnel perform a walkdown inspection to identify and quantify any leakage found at specine locations inside the Containment and in the Auxiliary llullding. The inservice inspection ensures that all components where boric acid leakage has been previously documented are also examined in accordance with the requirements of this program. A second inspection of these components is performed prior to plant startup (at normal operating pressure and temperature) if leakage was identified previously and corrective actions were taken. [ Reference 24, Sections 5.1 and 5.2]
Under the PACI Program, the walkdown inspections applicable to NSSS Sampling System components are type VT 2 (a type of visual examination described in ASME Section XI, IWA 2212). The VT 2 visual examinations include the accessible external exposed surfaces of pressure retaining, non insulated components: Goor areas or equipment surfaces located underneath non insulated components; vertical surfaces of insulation at the lowest elevation where leakage may be detected, and horizontal surfaces at each insulation joint for insulated components; Door areas and equipment surfaces beneath components and other areas where water may be channeled for insulated components whose external insulation surfaces are inaccessible for direct exarination; and for discoloration or residue on any surface for evidence of boric acid accumulation. [ Reference 24, Section 5.2)
Application for License Renewal 5.13 15 Calvert Cliffs Nuclear Power Plant
ATTACitMENT 0)
APPENDIX A TECilNICAL INFORMATION !
5.13 - NSSS SAMPLING SYSTEM If either leakage or corrosion is discovered, issue reports (irs) are generated in accordance with CCNPP procedure QL-2100, ' Issue Reporting and hsessment," to document and resolve the denciency.
Corrective actions address the removal of boric acid residue and inspection of the affected components for general corrosion, if general corrosion is found on a component, the IR provides for evaluation of the component for continued service and corrective actions to prevent recurrence. [ Reference 24, Section 5.3]
%e UACI Program has evolved with regard to boric acid leaks discovered during other types of walkdowns and inspections. He program specifies the minimum qualincation level for inspectors evaluating boric acid leaks. Apparent leaks that are discovered during these other walkdowns/ inspections are documented in irs by the individual discovering the leak. These irs are then routed to the inservice inspection group for closer inspection and evaluation by a qualified inspector.
This approach provides for more boric acid leakage inspection coverage while still ensuring that appropriately-qualified individuals assess and quantify any resultant damage.
The corrective actions taken as a result of irs under this program will ensure that NSSS Sampling System components containing borated water remain capable of performing their intended function under all CLil conditions during the period of extended operation.
Since the ventilating hood for the Unit I steam generator blowdown sampling subsystem (which contains the radioactive miscellaneous waste sampling subsystem) is not within the scope of the 11ACI Program, CCNPP currently plans to include the miscellaneous waste evaporator concentrate pump discharge sampic cooler in an Age Related Degradation inspection (ARDI) Prograrn to verify that degradation of the end plates / piping due to general corrosion is not occurring. [ Reference 3. Attachment 8]
The ARDI Program is 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 examinat!an sample size based on plausible aging effects;
- Identification of inspection locations in the system / component based on plausible aging effects and consequtnces ofloss of component intended function;
- Determination of examination techniques (including acceptance criteria) that would be effective, considering the aging efTects for which the component is examined;
=
Methods for interpretation of examination results;
- Methods for resolution of adverse examination findings, including consideration of all design loading conditions required by the CLD and specification of required corrective actions; and
+ Evaluation of the need for follow up examinations to monitor the progression of any age-related degradation, Any corrective actions that are required will be taken in accordar.cc with the CCNPP Corrective Actions i Program and will ensure that the components will remain capabic of performing their intended function under all CLD conditions.
- _ i App &ation for License Renewal 5,13 16 Calvert Cliffs Nuclear Power Plant '
ATTACllMENT m APPENDIX A - TECilNICAL INFORMATION ;
5.13 - NSSS SAMPLING SYSTEM l Group 1 -(general corrosion of enternal surfaces) . Demonstration of Aging Management llased on the information presented above, the following conclusions can be reached with respect to general corrosioa of external surfaces for NSSS Sampling System components:
- The llXs, CVs, and ilVs in Group 1 contribute to maintaining the pressure boundary of interfacing systems. Their integrity must be maintained under all CLB design conditions.
- The materials of construction for subcomponents in this group are carbon steel or alloy steel, l
- General cormslon is a plausible ARDM for this group because susceptible external surfaces are i exposed to p wential boric acid leakage from mechanicaljoints. If unmanaged, this ARDM could eventually result in the loss of pressure retaining capability under CLB design loading conditions. ;
- The corrosive effects of boric acid leakage will be managed by means of the BACI Program.
When bc,ric acid leakage is identified, either through required program inspections or through irs resulting from other types of walkdowns and inspections, this program will ensure that corrosion induced by boric acid is discovered and that appropriate corrective action is taken.
. The miscellaneous waste evaporator concentrate pump discharge sample cooler will be subjected to a new ARDI Program. This program will examine a representative sample of the components for degradation, and ensure that appropriate corre$ve actions are initiated on the basis of the findings.
Therefore, there is a reasonable assurance that the effects of general corrosion will be adequately managed for external surfaces of NSSS Sampling System components such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under all design loading conditions.
i Group 2 -(cretice corros.on and pitting ofinternal surfaces esposed to chem!cally treated water)-
Materials and Environment Group 2 comprises the various NSSS Sampling System components that are exposed to chemically.
treated water and whose internal surfaces are subject to crevice corrosion and pitting. The components in this group are included in the llX, IIV, and SV device types. The SVs in this group provide a containment isolation function for the sample return lines from the reactor coolant sample hoods to the RCDT. [ Reference 20, Table 2) The remaining components provide the passive intended function of maintaining the system pressure boundary. (Reference 3. Attachment 1) The applicable subcomponents in these device types are constructed of the following materials: [ Reference 3, Attachments 4 and 5 for llXs, llVs, SVs]
IlXs carbon steel end plates / piping and stainless steel shell, tubes, welds, capscrews, and fittings for sample coolers in the steam generator blowdown conditioning racks, the radioactive miscellaneous waste sampling subsystem, and the oxygen analyzer cabinet; a llXs carbon steel piping and copper tubing for sample chillers in steam generator blowdown conditioning racks;
- IIVs - stainless steel body / bonnet, gland nut, stem, disk, and seat for normally closed isolation llVs in low pressure SI pump, spent fuel pool filter and demineralizer sampling lines; Application for License Renewal 5.13 17 Calvert Cliffs Nuclear Power Plant
- . . -- - _ _ - . .. . - . - - . ~ = - --
ATTACllMENT 0) i APPENDIX A TECilNICAL INFORMATION 5.13 NSSS SAMPLING SYSTEM
+
llVs carbon steel body / bonnet and stem for normally open root, backup llVs in steam generator blowdown sampling lines;
- IIVs stainless steel body / bonnet, stem, union nut, and packing bolt for normally closed isolation llVs in steam generator blowdown sampling conditioning racks; e llVs stainless steel body / tubes, rod assembly, and nuts for rod in-tube pressure reducing IIVs in steam generator blowdown sampling conditioning racks; and
- SVs stainless steel body, bonnet, spring, and disc assembly for containment isolation SVs in the sample return lines from the reactor coolant sample hoods to the RCDT.
Piping and valves in the RCS sample header and in the charging pump sampling lines are excluded from Group 2. %ese portions of the NSSS Sampling System are subject to the hydrogen overpressure utilized as a corrosion inhibitor for the RCS. Due to the extremely low oxygen concentrations in the process Duld, the minimal impurity content that results, and stainless steel materials of construction, crevice corrosion and pitting are not considered plausible for the internal surfaces of these components.
[ Reference 3, Attachment 6s for CVs, IIVs; Reference 17, Piping Class CC 16; Reference 18, Piping Class CC 8]
For all components evaluated in Group 2, the internal surfaces are exposed to an environment of chemically treated water from the system being sampled, or from the CC System, or both. (Reference 3, Attachment 3s) ,
For the llXs in this group, stagnant Dow conditions may exist due to the physical geometry of the components and due to idle operation for portions of the system. Stagnant How may allow impurities in the process Huids to concentrate. [ Reference 3, Attachment 6s for ilXs) %e applicable Guids in the llXs are identined below:
- For sample coolers in the NSSS Sampling System, the internal environment is chemically treated water from the CC System between the inside of the shell and the outside of the tubes that contain the sample Dulds being cooled. The CC System has a design pressure of 150 psig and mrximum operational temperature of 167'F. { Reference 3, Attachment 3s for ilXs; Reference 22, Piping Class 110 23] For these sample coolers, the Gulds being cooled now inside the tubes. The cooled
'luids for sample coolers in the reactor coc iant sample hoods and radioactive miscellaneous waste sampling subsystem are described in subsection Group 1 - Materials and Environment, above.
The cooled Huid for sample coolers in the steam generator blowdown sampling subsystem is a saturated mixture of steam and feedwater with normal operating parameters of approximately 750 psig and $20 F. [ Reference 26, Piping Class EB 14) %e cooled fluid for the sample cooler in the oxygen analyzer cabinet is a radioactive mixture of gases with maximum operating pressures and temperatures as high as 150 psig and 303'F, irspectively, from the RCS quench tank, waste gas decay tanks, or waste gas system piping. [ Reference 19, Piping Class llc-43; Reference 27, Piping Class llc 29; Reference 28, Piping Class llc 60)
For the sample chillers in the steam generator tiowdown conditioning racks, the internal environment is also water from the CC System that is contained inside the piping connected to the chiller condenser and the chiller tubing. A liquid / vapor refrigerant mixture is contained in the chiller condenser shell surrounding the tubing. The normal operating pressure for the CC System is greater than the refrigerant condenser pressure. [ Reference 3, Attachment 6 for sample chiller)
Application for License Renewal 5.13 18 Calvert Cliffs Nuclear Power Plant
ATTACHMFNT LM APPENDIX A TECHNICAL INFORMATION 5.13 . NSSS SAMPLING SYSTEM The internal environments for the valves in Group 2 are listed below:
- For ilVs in sampling lines from the low pressure Si pump discharge headers: borated water with maximum operating pressures and temperatures as high as 450 psig and 300'F, respectively; !
[ Reference 29 Piping Class GC 1)
- For llVs in sampling lines from the spent fuel pool filter and demineralizer: borated water with maximum operating pressures and temperatures as high as 148 psig and 155'F, respectively; and
[ Reference 22, Piping Class llc-4]
+ For ilVs in the steam generator blowdown sampling subsystem: a saturated mixture of steam and feedwater with normal operating parameters of approximately 750 psig and 520*F; and
[ Reference 26, Piping Class 111114]
. For SVs in this group: borated water from the reactor coolant sample hoods with a maximum operating pressure and temperature of 55 psig and 303'F, respectively. [ Reference 19, Piping Class llc-43]
Group 2 -(crevice corrosion and pitting ofinternal surfaces esposed to chemically treated water) .
Aging Mechanism Effects Crevice corrosion and pitting are related forms of intensive, localized corrosion. Crevice corrosion occurs in crevices that are wide enough to permit liquid entry and narrow enough to maintain stagnant conditions. Such locations may include holes, gasket surfaces, lap joints, spaces under bolt heads, surface deposits, designed crevices for attaching thermal sleeves to safe-ends, and integral weld backing rings or back up bars. Pitting occurs when corrosion proceeds at one small location at a rate greater than the corrosion rate of the surrounding area. Pitting is an autocatalytic process that produces conditions that stimulate the continuing activity of the pit. In either case, the stagnant fluid within the pit or crevice tends to s.ccumulate corrosive chemicals, and thereby to accelerate the local corrosion process. Crevice corrosion can initiate pitting in many cases. Pitting can result h complete perforation of the material.
[ Reference 3, Attachment 7s for ilX, valve]
Crevice corrosion and pitting are plausible at mechanical joints (i.e., areas of IlXs that are idled or not exposed to the general flow stream, and body / bonnet joints for valves) since the mechanica' joint presents a crevice geometry at the scaling surfaces that may allow process fluids to stagnate and can concentrate environmentally produced impurities. [ Reference 3, Attachment 6s for ilXs, llVs, SVs]
Similar stagnation and impurity deposits are possible at other component interior crevices that are formed by close-fitting interface points at interior subcomponents (i.e., tubing for the sample chillers in steam generator blowdown conditioning racks, seating surfaces for containment isolation SVs, and various intemal subcomponents for valves). [ Reference 3, Attachment 6s for llXs, ilVs, SVs]
iherefore, crevice corrosion and pitting are plausible for all components in this group.
Application for License Renewal 5.13 19 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (3) ,
APPENDIX A TECIINICAL INFORMATION i 5.13 NSSS SAMPLING SYSTEM f
Group 2 - (crevice corrosion and pitting oflaternal surfaces espesed to chemically-treated water)- l Methods to Manage Agleg .
Mitigation: Control of fluid chemistry in systems that interface with the NSSS Sampling System can !
significantly limit the effects of crevice corros!on and pitting. [ Reference 3, Attachment 6s for llXs, ilVs, SVs) 1he chemistry control program should monitor peninent chemical parameters on a frequency that would allow for corrective actions to minimize creation of an environment conducive to corrosion. ,
Dhtomy: The effects of corrosion are generally detectable by visual techniques. Obsenation of corrosion products in chemistry samples can indicate unexpected system corrosion. Seating surface degradation can be discovered by testing the components that are susceptible to this ARDM. Pressure testing cf the containment isolation SVs can provide for detection ofleakage that could be the result of crevice corrosion and pitting of the valve seating surfaces. Internal surfaces of components that are not routinely inspected can be subjected to inspection to determine the extent of general and/or localized degradation that may be occurring. [ Reference 3, Attachment 6s for llXs,llVs, SVs)
Group 2 -(crevice corrosion and pitting ofInternal surfaces exposed to chemically treated water)-
Aging Management Program (s)
Mitigation: Maintenance of proper fluid chernistry in systems that interface with the NSSS Sampling System will limit the effects of crevice corrosion and pitting on internal subcomponents for Group 2 IlXs and valves. [ Reference 3, Attachment 8)
The CCNPP Chemistry Program has been established to: minimize impurity ingress to plant systems; reduce corrosion product generation, transport, and deposition; reduce collective radiation exposure through chemistry; improve integrity and availability of plant systems; and extend component and plant life. [ Reference 30, Section 6.1.A] The program is based on Technical SpeciFcations, DGE's interpretation of industry standards, and recommendations made by Combustion Engineering.
[ Reference 31, Section 2.0; Reference 32, Section 2.0)
Calvert Clifts Technical Procedure CP 204," Specification and Surveillance Primary Systems," provides for monitoring and maintaining chemistry in the RCS and associated systems. [ Reference 31, Section 2.0, Attachments I through 15] Control of primary water chemistry is credited with limiting the clTects of crevice corrosion and pitting in NSSS Sampling System components that interface with prima.y systems. [ Reference 3, Attachment 8)
Calven ClitTs Technical Procedure CP 206, " Specifications and Surveillance Component Cooling / Service Water Systems," provides for monitoring of CC System chemistry to control the concentrations of oxygen, chlorides, and other chemicals and contaminants. [ Reference 33, Section 2.0, Attachment 1) Control of the water chemistry prevents a corrosive environment for NSSS Sampling System IlXs that are cooled by water from the CC System. [ Reference 3, Attachment 8) l Calvert Cliffs Technical Procedure CP 217, " Specification and Surveillance Secondary Chemistry,"
provides for control of fluid chemistiy in the steam generators in order to minimize the concentration of l corrosive impurities (chlorides, sulfates, oxygen) and to optimize fluid pil. [ Reference 32, Section 2.C, Attachments I through 13] It has been demonstrated that this chemistry control program, with its mrJ.a l
Application for License Renewal 5.13 20 Calvert Cliffs Nuclear Power Plant
ATTACllMENT {3)
APPENDIX A - TECilNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM focus on steam generator preservation, is successful at controlling corrosion rates in the balance of the i secondary cycle. Control of secondary fluid chemistry mi:.imites the corrosive environment for components in the steam generator blowdown sampling otisystem, and limits the rate and effects of corrosion. [ Reference 3, Attachment 8]
Each of the program procedures describes the f.urveillance and specifications for monitoring fluid chemistry for the applicable systems. Hey list the parameters to be monitored, the frequency for monitoring of each parameter, and the acceptable value or range of values for each parameter.
[ Reference 31, Attachments I through 15; Reference 32, Attachments I through 13; Reference 33, Attachment 1] Each parameter is measured at a procedurally specified frequency (e.g., daily, weekly, monthly) and compared against a target value that represents a goal or predetermined warning limit.
(Reference 31, Section 3.0; Reference 32, Section 3.0; Reference 33, Section 3.0] If a measured value is outside of its required range, corrective actions are taken (e.g., power reduction, plant shutdown) as prescribed by the procedure, thereby ensuring timely response to chemical excursions. [ Reference 31, Section 6.0.C; Reference 32, Section 6.0.C; Reference 33, Section 6.0.C] He procedures provide for raplJ assessment of off normal chemistry parameters so that steps can be taken to return them to normal levels. [ Reference 31, Section 2.0; Reference 32, Section 2.0]
He CCNPP Chemistry Program has been subject to periodic internal assessment activities, internal audits are performed to ensure that activities and procedures established to implement the requirements of 10 CFR Part 50, Appendix B, comply with BGE's overall Quality Assurance Program. These audits provide a comprehensive independent verification and evaluation of qui.lity related activities and procedures. Audits of selected aspects of operational phase activities are pirformed with a frequency commensurate with their strength of performance and safety significance ani in such a manner as to assure that an audit of all safety related functions is completed within i. period of two years.
[ Reference 34, Section 1B.18] These activities, as well as other external assessments, help to maintain highly effective chemistry control, and facilitate continuous improvement through monitoring industry initiatives and trends in the area of corrosion control.
A review of operating experience identified no site specific problems or events related to crevice corrosion or pitting that required significant changes or adjustments to the CCNPP Chemistry Program, it has been efTective in its function of mitigating corrosion, and thereby preventing corrosion related failures and problems. Calvert Cliffs has been proactive in making programmatic changes to the secondary chemistry program over its history, largely in response to developments within the industry (e.g., successful experimentation with a new alternate amine). In 1996, CP 206 was revised to include monitoring of dissolved iron as a method for discovering any unusual corrosion of carbon steel components. An internal BGE chemistry summary report for 1996 included recommendations to: (a) determine outage evolutions that can affect the CC/ Service Water chemical parameters; and (b) take action to prevent chemistry parameters from being exceeded.
Discoverv: Calvert Cliffs procedures STP M 57111(2)," Local Leak Rate Test, Penetrations ID,47A, 478,47C,47D,48A,480,49A,498,49C." which cover local leak rate testing (LLRT) for the sample return lines from the reactor coolant sample hoods to the RCDT, are part of the overall CCNPP Containment Leakage Rate Testing Program. [ References 35 and 36] The CCNPP Containment Leakage Rate Testing Program was established to implement the leakage testing of the containment as required by 10 CFR 50,54(o) and 10 CFR Part 50, Appendix J, Option B, " Primary Reactor Containment
~
Application for License Renewal 5.13 21 Calvert Clifts Nuclear Power Plant
NITACllMINT 0)
APPENDIX A TECilNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM Leakage Testing for Water Cooled Power Reactors." Appendix J specines containment leakage testing requirements, including the types of tests required, frequency of testing, test methods, test pressures, acceptance criteria, and reporting requirements. Containment leakage testing requirements include performance ofintegrated Leakage Rate Tests, also known as Type A tests, and LLRTs, also known as Type B and C tests. Type A tests measure the overall leakage rate of the containment. Type B tests are intended to detect leakage paths and measure leakage for certain containment penetrations (e.g., airlocks, Danges, and electrical penetrations). Type C tests are intended to measure containment isolation valve leakage rates. [ Reference 37, Section 6.5.6; References 38 and 39]
The CCNPP LLP.T Program is based on the requirements of CCNPP Technical Specifications 3.6.l.2,4.6.1.2, and 6.5.6. The scope of the program includes Type B and C testing of containment penetrations. The valves that isolate the containmc..t penetration piping for the sample return lines from the reactor coolant sample hoods to the RCDT are included in the scope of this program as part of the leakage testing for the associated containment penetrations. (Reference 37]
%e LLRT is done on a performance-based testing chedule in accordance with Option B of 10 CFR Part 50, Appendix J, as implemented by CCNPP Technical Speel0 cations. [ References 37,38, and 39). Local leak rate testing presently includes the following procedural steps:
. Leak rate monitoring test equipment is connected to the appropriate test point.
- The test volume is pressurued to the LLRT Program test pressure, which is conservative with respect to the 10 CFR Part 50, Appendix J, test pressure requirements. Appendix J requires testing at a pressure "P.," which is the peak caiculated containment internal pressure related to the design basis accident.
Leak rate, pressure, and temperature are monitored at the frequency specified by the LLRT procedure and the results are recorded.
De maximum indicated leak rate is compared against administrative limits that are more restrictive than the maximum allowable leakage limits.
"As found" leakage equal to or greater than the administrative limit, but less than the maximum allowable limit, is evaluated to determine if further testing is required and/or if corrective maintenance is to be performed.
- For "as found" leakage that exceeds the maximum allowable limit, plant personnel determine if Technical Specification Limiting Condition for Operation 3.6.1.2.b has been exceeded. Technical Specification 3.6.1.2.b contains the maximum allowable combined leakage for all penetrations and valves subject to the Type B and C tests. Corrective action is taken as required to restore the leakage rates to within the appropriate acceptance criteria.
- If any maintenance is required on a containment isolation valve that changes the closing characteristic of the valve, an "as len" test must be performed on the penetration to ensure leakage rates are acceptable.
He corrective actions taken as part of the LLRT Program will ensure that the containment isolation SVs in the sample return lines from the reactor coolant sample hoods to the RCDT remain capable of performing their intended functions under all CLB conditions during the period of extended operation.
Application for License Renewal 5.13 22 Calvert Cliffs Nuclear Power Plant I
ATTACHMENT (3)
APPENDIX A - TECilNICAL INFOl(MATION 5.13 NSSS SAMPLING SYSTEM l l
Haltimore Oas arid Electric Company currently plans to include all Group 2 components in an ARDI l Program to verify that unacceptable degradation ofinternal surfaces by crevice corrosion or pitting L not i occurring. [ Reference 3, Attachment 8] For a discussion of the elements of the ARDI Program, refer to subsection Group 1 Aging Management Programs, above.
Group 2 -(crevice carrosion and pitting ofinternal surfaces exposed to chemically treated water)-
Demonstration of Aging Management Ilased on the information presented above, the following conclusions can be reached with respect to crevice corrosion and pitting of internal surfaces for NSSS Sampling System components that are exposed to chemically treated water:
- The Group 2 IlXs and ilVs contribute to maintaining the pressure boundary of interfacing systems, and the SVs provide a containment isolation function. The integrity of these components must be maintained under all CLB design conditions.
a The materials of construction for subcomponents in this group are carbon steel, stainless steel, or copper.
- Crevice corrosion and pitting are plausible ARDMs for this group and, if unmanaged, these ARDMs could eventually result in the loss of pressure retaining capability under CLB design loading conditions.
- Maintenance of proper Guid chemistry in primary systems (in accordance with CP 204) rad secondary systems (in s.ccordance with CP 217) that are sampled through the NSSS Sampling System will limit the effects of crevice corrosion and pitting on susceptible pressure boundary subcomponents for Group 2 !!Xs and valves. Chemistry control in accordance with CP 206 will ensure that the cooling water supplied to NSSS Sampling ilXs is of an appropriate chemistry to minimize corrosion.
- He CCNPP LLRT Program performs leakage testing that could detect the efTects of crevice corrosion and pitting on the : ating surfaces of the containment isolation SVs in the sample retum lines from the reactor coolant sample hoods to the RCDT (i.e., degraded leak tightness). This program ensures that appropriate corrective actions will be taken if signincant leakage is discovered.
- All Group 2 components will be subjected to a new ARDI Program. This program will examine a representative sample of the componente for degradation, and ensure that appropriate corrective actions are initiated on the basis of the Undings.
Therefore, there is a reasonable assurance that the effects of crevice corrosion and pitting will be adequately managed for internal surfaces of NSSS Sampling System components exposed to chemically-treated water such that they will be capable of performing their intended functions consistent with the CLB during the period of extended operation under all design loading conditions.
Application for License Renewal 5.13 23 Calvert Cliffs Nuclear Power Plant
- .- - . - ~ -__ _ -- -_ - -_ - - - - -.-_=- - -. .. -
ATTACllMENT Q)
APPENI)lX A - TECilNICAL INFORMATION :
5.13 - NSSS SAMPLING SYSTEM Group 3 - (general corro.lon of internal surfaces esposed to compressed air) - Materials and Environment Group 3 comprises the operators associated with CVs in the reactor coolant sampling lines whose internal surfaces are subject to general corrosion. The applicable subcomponents are constructed of the following materials: [ Reference 3. Attachments 4 and 5 for CVOPs]
- For CVOPs associated with the CVs isolating sample piping from the pressurizer surge line and the pressuriter vapor space, carbon steel actuator cases, cast iron / steel yokes, and carbon steel bolting; and
. For CVOPs associated with the RCS sample isolation CVs and the CVs isolating sample piping from the RCS hot leg, ductile iron yoke' and 7.inc plated steel adjusting screws.
The internal environment for the CVOPs is nonnally compressed air supplied by the IA compressors in the Compressed Air System. The 1A is very dry, filtered, and oil. free air, with nonnat dew point maintained at -40'F at the design pessure of 100 psig. [ Reference 2, Section 9.10, Table 9 21]
Occasionally, air that does net meet the same air quality standards may enter the IA supply due to operation of the plant air compressors (with minimal drying capacity) or the saltwater air compressors (with no air dryer), which serve as backups to the IA compressors. [ References 40 through 43]
1herefore, there is a possibility that moisture may enter the lA supply, although its effect is expected to be minimal due to the limited operation of these backup sources. [ Reference 3, Attachment 8]
Group 3 - (general corrosion of internal surfaces exposed to compressed air) - Aging Mechanism Effects The effects of general corrosion are discussed in subsection Group 1 Aging Mechanism EfTects, above 4
General corrosion of all ferrous materials internal to the CVOPs is plausible because the materials of construcuon may occasionally be exposed tc slightly moist air; however, this ARDM is also considered unlikely. At the normal dew point of -40'F, there is insufficient moisture to cause signincant occurrences of this mechanism. 'the expected efTects would be super 0cial rust speckles and a slight dusting of loose passive surface rust. He consequence of general corrosion damage to the affected components would be a loss of load. carrying cross-sectional area. l Reference 44, Attachment 6s and Attachment 8]
Group 3 -(general corrosion of internal surfaces exposed to compressed air)- Methods to Manage Aging Mitigatiom Maintaining lA quality within industry standards for dew point can ensurc minimal degradation resulting from moisture for carbon steet/ iron subcomponents exposed to compressed air,
[ Reference 3, Attachment 6s for CVOPs] An inspection performed on the piping immediately dowmtream of the saltwater sir compressors, where the worst case of general corrosion in the IA supply l
is expected, revealed only very light surface rust on the inside of each piece. Aher more than 20 years in operation, approximately 60% of the pipe interior contained no rust and appeared similar to the inside of new pipe. Thickness measurements showed that the wall thickness averaged only 0.001 inch less than
! the neminal thickness of 0,179 inch. [ Reference 44, Attachment 8] In order to assure that the Application for License Renewal 5.13 24 Calvert Cliffs Nuclear Power Plant
. t AIIACilMENTE)
APPENDIX A TECHNICAL INFORMATION 5.13 . NSSS SAMPLING SYSTEM compressed air quality remains within acceptable limits, the air quality should be periodically checked and compared against the industry standards. [ Reference 44, Attachment 8] If testing shows a reduction ,
in air quality, corrective actbns can be initiated to return the air quality to normal.
Discorny: Since I A in the Compressed Air System is normally very dry, and since minimal corrosion is evident aller more than 20 years of operation, continued maintenance of IA quality is deemed an adequate aging management technique for general corrosion control in components connected to the IA !
supply. [ Reference 44, Attachment 8]
Group 3. (general corrosion ofInternal surfaces esposed to compressed air) Aging Management Program (s)
Mitigation: The CCNPP Preventive Maintenance Program has been established to maintain plant equipment, structures, systems, and components in a reliable conditicn for normal operation and emergency use, minimize equipment failure, and extend equipment and plant life. He program covers all preventive maintenance activities for nuclear power plant structures and equipment within thr, plant, including those applicable to NSSS Sampling System components within the scope of license renewal.
[ Reference 45]
Calvert Cliffs Preventive Maintenance Checklists are executed by preventive maintenance tasks that are automatically scheduled and implemented in accordance with safety related Preventive Maintenance Program procedures. [ Reference 45]
Calvert Cliffs initiated a Preventive Maintenance Checklist following a review of recommendations resulting from industry operating experience. Preventive Maintenance Checklist IPM 10000 (10001),
" Check Unit 1(2) Instrument Air Quality," checks IA quality at three locations in the Compressed Air System: at the dryer outlet, at the furthest point from the dryer, and at the approximate mid point between the other two. Measurerients of dew point are taken every 12 weeks at these locations.
[ Reference 3, Attachment 8; References 46 and 47] Dew point data are reviewed, trended, and evaluated in accordance with approved procedures, if the air quality is detennined to be abnormal, corrective action is initiated to return the air quality to normal and to investigate the condition of the dependent load internals (e.g . carbon steel piessure boundary subcomponents for CVOPs in reactor coolant sampling lines), as appropriate. [ References 46 and 47) This process ensures IA quality is maintained in accordance with industry standards for moisture (dew point). Operating experience relative to IA quality control has shown that the air normally provided is very dry and contains little particulate matter.
[ Reference 44, Attachment 8]
The Preventive Maintenance Program has been evaluated by the NRC as part of its routine licensee essessment activities. The plant Maintenance Program also has had numerous levels of management review, all the way down to the specific implen.entation procedures. Specific responsibilities are assigned to HGE personnel for evaluating and upgrading the Preventive Maintenance Program, and for initiating changes to the Preventive Maintenance Program based on system performance. These assessments and controls provide reasonable assuraice that the Preventive Maintenance Program will continue to be an effective method of mitigating the effects of general corrosion on internal surfaces of the CVOPs exposed to compressed air. [ Reference d]
Application for 1,icense Renewal 5.13 25 Calvert Cliffs Nuclear Power Plant
ATTACHMENT (3)
APPENDIX A . TECHNICAL INFORMATION 5.13 NSSS SAMPLING SYSTEM Dinomy: Continued implementation of the mitigation technique discussed above should ensure that .
exposure of the CVOP internal surfaces to moisture will continue to be minimal. Since the saltwater air i compressors only provide a backup to essential loads normally supplied by the IA compressors, introduction of moisture from this source will also be minimal. Corrosion of the CVOP internal surfaces is not expected to result in significant levels of degradation. It is deemed that the mitigation techniques described above are adequate aging management practices for corrosion, and no discovery techniques are necessary for Group 3 components. [ Reference 44, Attachment 8]
Group 3. (general corrosion of Internal surfaces esposed to compressed alr) . Demonstration of l Aging Management '
llased on the information presented above, the following conclusions can be reached with respect to ;
general corrosion of internal surfaces for NSSS Sampling System components that are exposed to I
compressed air;
+ 'the CVOPs associated with CVs in the reactor coolant sampling lines contribute to maintaining the pressure boundary of the Compressed Air System. Their integrity must be maintained under all CLil design condition',
- The materials of construction for subcomponents in this group are carbon steel, zinc plated steel, >
or iron.
+ General corrosion is a plausible ARDM for this group because susceptible materials are exposed to potentially moisture laden air from the Compressed Air System. If unmanaged, these ARDMs ,
could eventually result in the loss of pressure retaining capability under CLB design loading conditions.
I
+ Periodic monitoring and mainter.t.nce ofI A dryness (in accordance with IPM 10000 [10001]) will enitigate the efTects of corrosion on ferrous pressure boundary subcomponents for CVOPs in reactor coolant sampling lines.
Therefore, there is a reasonable assurance that the effects of general corrosion will be acequately managed for internal surfaces of NSSS Sampling System components exposed to compressed air such that they will be capable of performing their intended functions consistent with the CLD during the period of extended operation under all design loading conditions.
Group 4. (fatigue for piping and vaives associated with sampling the HCS hot leg) . Materials and Environment Group 4 comprises components in the reactor coolant sampling subsystem associated with sampling the RCS hot leg for which fatigue is a plausible ARDM. Specifically, this group consists of the following NSSS Sampling System components that art depicted in Figure 5.131:
+
RCS sample isolation CV (l[2]CVPS 5464);
+
RCS sample header test connection root val' c (ll2]IIVPS 196);
+
RCS sample header telltale root valve (l[2]IlVPS 200);
Application for License Renewal 5.13 26 Calvert Cliffs Nuclear Power Plant
l AIIACitMirr (3) 1 APPENDIX A TECilNICAL INFORMATION 5.13 NSSS SAMPLING SYSTEM
+ RCS sample header vent valve (l[2]IIVPS-401); :
+ intervening pipe segments (1#CC8 1001, 1002, .1003 (2#CC8 2001, .2002, .2003)).
All of these components provide the passive intended function of maintaining the system pressure boundary. [ Reference 3, Attachment 1) The applicable subcomponents in these device types are constructed of the following materials: [ Reference 3. Attachments 4 and 5 for pipe, CVs,IIVs]
+
Piping . stainless steel pipe, Ottings, and w cids;
+ CVs . carbon steel nuts, alloy steel studs, and stainless steel body / bonnet; and
+ llVs . carbon steel nuts, alloy steel studs, and stainless steel body / bonnet.
'Ihe original design code for the piping in this group is ANSI 1131.7 ClassII. [ Reference 18, Piping Class CC 8) At CCNPP, the original system design assumed a stress range reduction factor of one, which corresponds to 7000 full range thermal cycles during the anticipated life of the plant.
Replacement of the original CVs in the RCS hot leg sampling lines was completed in 1993. 'Ihc new valves were designed in accordance with Class I requirements for valves one inch and under in ASME Section ill(1977 Edition, including the Summer 1978 Addenda). [ Reference 48] <
Similar NSSS Sampling System piping and valves, from the pressurizer surge line and vapor space sample points up to and including the associated sample header isolation CVs, are excluded from Group 4. It has been conscivatively estimated that components in the pressurizer surge line and vapor space sample headers have experienced no more than 450 full range thermal cycles since installation. ,
Since samples from the presserizer are no longer being taken, no additional thennal fatigue cycles are being experienced by these components.
The internal environment for pressure boundary subcomponents in this group is borated water at an operating pressure of approximately 2250 psia. [ Reference 2, Section 4.1.1, Table 41]. Normal sampling operations cause rapid temperature transients in the sampling line from ambient (about 100 F) to the nonnal RCS hot leg operating temperature and back to ambient (i.e., difTerential temperature of up to 500*F). [ Reference 3. Attachment 6s for pipe, CVs, llVs]
The external environment is climate-controlled air in the Auxiliary 13uilding and the Containment.
[ Reference 3, Attachment 3s] Refer to subsection Group 1 Materials and Environment, above, for additional discussion.
Group 4 - (fatigue for piping and valves assoelated with sampling the RCS hot leg) . Aging Mechanism Effects i
l Fatigue is the process of progressive localized structural change occurring in a material subjected to conditions that produce Duetuating stresses and strains at some point or points in the material. This process may culminate in cracks or complete fracture aner a sufficient number of Ductuations. The fatigue life of a component is the number of cycles of stress or strain that it experiences before fatigue Application for License Renewal 51327 Calvert Cliffs Nuc! car Power Plant
ATTACHMENT 0)
APPENDIX A TECHNICAL INFORMATION 5.13 - N5SS SAMPLING SYSTEM failure occurs. Failures may occur at either a high or low number of cycles in response to various kinds of loads (e.g., mechanical or vibrational loads, thermal cycles, or pressure cycles). Low-cycle fatigue involves stressing of materials, often into the plastic range, with the number of cycles usually being less than 10'. 'this mechanism is typically associated with thermal gradients created in restrained members during rapid heatup or cooldown. A component subjected to sufficient cycling with significant strain <
ntes accumulates fatigue damage, which potentially can lead to crack initiation and crack growth. The '
cracks may then propagate under continuing cycIlc strains. [ Reference 3, Attachment 7s for pipe, CVs, ilVs; Reference 49]
low cycle thermal fatigue is plausible for the devices in this group since they experience severe thennal cycling during routine RCS sampling operations Nhich occur several times each week in accordance with CP-204). [ Reference 3 Attachment 6s for pipe, CVs,llVs; Reference 31 Attachments 1 through 3]
Italtimore Gas and Electric Company has not discovered any low-cycle fatigue related failures in the NSSS Sampling System. Ilowever, there have been occurrences in similar applicatior,s at other facilities, including cracks found in the inlet opening and around the base of the valve bore area in RCS hot leg sampling line CVs at Arizona Public Service's Palo Verde Nuclear Generating Station.
These aging mechanisms, if unmanaged, could eventually result in crack initiation and growth such that the Group 4 components may not be able to perform their pressure boundary function under CLB design loading conditions.
Group 4. (fatigue for piping and valves associated with sampling the RCS hot leg) Methods to Manage Aging hiltigation: The effects oflow-cycle fatigue can be mitigated by proper system design and material selection, and by operational practices that reduce the number and severity of thermal transients on the RCS hot leg sampling piping. (Reference 3, Attachment 65 for CVs,ilVs]
Discoverv: As discussed above, low cycle fatigue was addressed in the original design for the RCS hot leg sampling lines. Additional RCS sampling (e.g., that required for a 60 year period of ope ation) could .
result in more than the number of full range thermal cycles estimated in the original system design. The accumulation of fatigne effects on components in the RCS hot leg sampling line., can be monitored by counting the number of the thennal transients and by performing analysis to predict the remaining life of the alTected components.
Group 4 - (latigue for piping and valves associated with sampling the RCS hot leg) - Aging Management Program (s)
Mitigation: The number and severity of thermal transients experienced by components in the RCS sampling ilnes is dependent on RCS sampling requirements established by the CCNpP Chemistry Program. Maintaining the established sampling frequencies precludes modification of plant operating practices to reduce thermal cycles on the NSSS Sampling System. Therefore, there are no practicable means available (beyond proper system design and material selection) to mitigate the effects of thermal fatigue.
Application for License Renewal 5.13 28 Calvert Cliffs Nuclear Power Plant
AIIACilMFET d)
APPENDIX A - TECilNICAL INFOl(MATION 5.13 - NSSS SAMPLING SYSTEM Discovery: The CCNPP Fatigue Monitoring Program (FMP) ha; been established to monitor and track fatigue usage oflimiting components of the NSSS and the steam generators. Reference 50 was used in the development of this prograrr.. Eleven fatigue critical locations in these systems have been selected for monitoring of fatigue usage. These represent the most bounding locations for critical thermal transients. [Referev:e 51, Sections 1.1,1.2.A. 2.1.E,6.0]
1he FMP utilizes two methods to track fatigue usaget
+ One method is to track the number of critical thermal and pressure test transients (i.e., cycle counting) and compare them to the number allowed in the system design analysis. The system design analpis is performed assuming a particular number and severity of varlaus transients. In accordance with either ASME Section til or ANSI 1331.7, the analysis demonstrates that the "
component has an acceptable design as long as the assumptions remain valid. Therefore, if the actual number and sever!ty of transients everienced by the component remains below the number assumed in the analysis, the component remains within its design basis.
- The other method is to determine the fatigue life of a com;cnent using a calculated cumulative usage factor (CUF), which is denned as a normalized measure of total fatigue damage accumulated by a component as a result of all stress cycles that the component has experienced during its service life. The CUF can be calculated and tracked through plant life using thermal cycle counting or stress-based analysis techniques, in accordance with the ASME Boiler and Pressu'c Vessel Code, the component remains within its design basis for allowable fatigue life if the CUF remains less than or equal to one. [ Reference $1, Sections 1.2.A,3.0.11,3.0.F]
Iloth methods use actual plant operating data. At CCNPP, the usage factor for several locations is calculated through stress-based analysis, which is the more rigorous of the two methods. Since the FMP monitors actual fatigue usage, a more realistic CUF is calculated. The data for thermal translents is collected, recorded, and analyzed using a computer program that evaluates input data from plant instrumentation. The computer sonware is used to analyze plant data associated with real transients and to predict the number of thermal cycle tra.sients for 40 and 60 years of plant operation based on the historical records. [Refersace 51, Section 3.0.F]
Plant perameter data is collected on a periodic basis and reviewed to ensure that the data represents actual transients. Valid data is entered into the computer program that counts the critical transient cycles and calculates the CUFs. The data is tracked in accordance with procedures that are governed by a quality assurance program that meets 10 CFR Part 50, Appendix D, criteria. The transient data is evabiated and the CUFr are calculated on a semi annual basis, which provides a readily predictable approach to the alert value. Acceptable conditions exist, since no crack initiation would be predicted, when the calculated CUF for any given component is less than one, or w hen the design allowable number of cycles for the co.nponent has not been exceeded, in order to stay within the design basis, corrective action is initiated well in advance of the CUF approaching one or the number of cycles approaching the design allowable, so that appropriate corrective actions can be taken in a timely and coordinated manner.
[ Reference 51, Sections 1.2.A. 5.0)
Tracking the usage for the limiting components ensures that all remaining components will also remain below their fatigue limits. The FMP will perform an engineering evaluation to determine if the low-cycle fatigue usage for the piping and valves in the RCS hot leg sampling line is bounded by the existing Application for License Renewal 5.13 29 Calvert Clifts Nuclear Power Plant
A*ITACHMFNT (3)
APPENI)lX A TECilNICAL INFORMATION 5.13 . NSSS SAMPLING SYSTEM bounding components. if these components are not bounded, they will be added to the FMP. Inclusion of these devices in the FMP may require a suitable alternative to thermal cycle counting. His ahernative may consist of verification of the fatigue reduction factor for these components with full consideration given to the magaitude and frequency of the thermal cycles imposed by RCS sampling evolutions. t
[ Reference 3, Attachment 10]
Since the FMP has been initiated, no locations have reached the limit on fatigue usage and no cracking due to low cycle fatigue has been discovered. He FMP has undergone several mod 10 cations since its inception. Stress based analysis was added to the computer software to calculate the CUFs for several >
locations due to unique thermal transients experienced and the unique geometries involved. Other modi 0 cations have been made to the FMP to reflect plant operating conditions more accurately. The plant design change process has also been modified to require notincation to the Life Cycle Management Unit of any proposed changes to the critical locations being monitored.
He CCNPP FMP has been inspected by the NRC, which noted that the program has been developed toward providing assurance that fatigue life usage of primary system components has not exceeded limits provided for in ASME Section !!!. In addition, the NRC noted that the FMP can be used tu identify components where fatigue usage may challenge the remaining and extended life of the components and can provide a basis for corrective action where necessary. [ Reference 52]
To further address fatigue for lleense renewal, CCNPP participated in a task, sponsored by the Electric Power Research Institute, to demonstrate the industry fatigue position. The task applied industry. developed methodologies to identify fatigue-sensitive component locations that may require further evaluation or inspection for license renewal and evaluate environmental effects, as necessary.
He program objective included the development and justification of aging management practices for fatigue at various component locations for the renewal period. The demonstration systems were the Feedwater Systern, the pressurlier surge line in the RCS, and the letdown and charging subsystem in the Chemical and Volume Control System. [ Reference 53. Page 3]
Evaluation of *lhermal Fatiguchts to Addren.Gcacric Safety issue 166:
Generic Safety issue 166, Adequacy of Fatigue Life of Metal Components, presents concerns identified by the NRC that must be evaluated as part of the license renewal process. He NRC stafTecncerns about fatigue for license renewal fall into the following five categories: [ Reference 53, Page 2; Reference $4]
- He first category, adequacy of the fatigue design basis when environmental effects are considered, does not apply to the RCS hot leg sampling line because of stringent RCS water chemistry controls, exceptionally low oxygen concentrations (less than Ove parts per billion), and stainless steel materials used in fabrication of the affected piping and valve subcomponents.
- The second category concerns tne adequacy of both the number and severity of design basis transients. He engineering evaluation addressing fatigue in the RCS hot leg sampling line, >
discussed above, will consider the magnitude and frequency of the thermal cycles imposed by RCS sampling evolutions. [ Reference 3 Attachment 10]
He third category, adequacy of inservice inspection requirements and procedures to detect fatigue indications, does not apply because CCNPP does not rely on inservice inspection as the sole means for detection of fatigue.
Application for License Renewal 5.13 30 Calvert Cliffs Nuclear Power Plant i
AIIACILMENTJ)
APPENDIX A TECHNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM
+ The fourth eategory, adequacy of the fatigue design basis for Class I piping components designed in accordance with ANSI D31.1, does not apply because the intervening pipe segments in this group are designed in accordance with ANSI H31,7, Class il
- 1he final category, adequacy of actions to be taken when the fatigue design basis is potentially compromised, as discussed above, is adequately addressed by the CCNPP FMP, l Group 4 (fatigue for piping and valves associated with sampling the RCS hot leg)- Demonstration of Aging Management Based on the information presented above, the followihG conclusions can be reached with respect to fatigue for NSSS Sampling System piping and valves in the RCS hot leg sampling lines:
- Piping and valves in the RCS het leg sampling linc contribute to maintaining the system pressure boundary, Their integrity must be maintained under all CLD design conditions.
- The mt.tmials of construct 6n for subcomponents in this group are carbon steel, stainless steel, or ,
alloy steel. i
+ Fatigue is a plausible ARDM for this group because the components are subject to severe thermal cycling several times each week during routine RCS sampling operations. If unmanaged, this ARDM could eventually result in crack initiation and giowth such that the components may not be able to perform their pressure boundary function under CLil conditions.
- The FMP monitors fatigue usage at bounding locations to ensure that NSSS components and the steam generators remain within their design basis. The FMP will be modified to include an agineering evaluation of the piping and valves in the RCS hot leg sampling line, therefore, there is a reasc,nabic assurance that the effects of fatigue will be adequately managed for componerts in the RCS hot leg r:mphng lines such that they will be capable of perL; ming their intended functions consistent with the CLB during the period of extended operation under all design !
loading conditions.
Group 5 -(elastomer degradation for valve internals)- Materials ind Environment Group 5 conshts of the rKVs in the 7as return line to containment from the PASS cabinet whose internals ere subject to cintomer degradation.1hese CKVs provide the passive intended function of maintaining the sysum pressere boundary. [ Reference 3, Attachment 1) The internals for these CKVs are constructed of clastomers and stainless steel. (Reference 3, Attachments 4 and 5 for CKVs] The CKV internals are exposed to environments containing potentially radioactive gases from the c containment atmosphere (described in subsection Group 1 - Materials and Environment, above).
[ Reference 3, Attachment 3sj Group 5. (elastomer degradation for valve internals)- Agin;g Mechanism Effects When an clastomer ages, the primary mechanisms involved are scission, crosslinking, and changes associated with the compounding ingredients. S~ission is the process of breaking molecular bonds, '
typically due to inone attack, tdtraviolet light, or radiation. Crosslinking is the process of creating molecular bonds between adjacent lang-chain molecules, typically due to oxyg:n attack, heat, or curing.
Application for License Renewal 5.13 31 Calvert Cliffs Nuclear Powei Plant
ATTACllMENT 0)
APPENDIX A TECilNICAL INFORMATION ,
5.13 - NSSS SAMPl.ING SYSTEM Scission results in increased clongation, decreased tensile strength, and decreased modulus; crosslinking [
has the opposite effects (i.e., decreased clongation, increased tensile strength, and increased modulus).
The compouredinF ingredients used in an elastomer / rubber may be affected by evaporation, leaching, or mutation over their service life. [ Reference 3, Attachment 7s for valves)
The seating surfaces for components in this group are constructed from elastomers, which are subject to l the degradation described above. Degradation of wy of these subcomponents would result in process Guld leakage past the seal, and eventual failure of the pressure boundary function.
Group 5. (elastomer degradation for valve internals) . Methods to Manage Aidng Midgadon: The degradation of clastomers is related 1o time, material r.clection, and the environment.
Apart from removal of the items from their environment, either as part of a permanent modincation or under a periodic replacement program, there are no reasonable methods of mitigating the effects of these ARDMs for the subject subcomponents.
Discovery: Scating surface degradation for the internals of these CKVs can be detected through visual inspection.
Group 5 -(elastomer degradation for valve internals) . Aging Management Program (s)
Millption: As part of the in place retirement of the PASS, the tubing between the CKVs in the gas return line to containment and the PASS cabinet will be capped; as a result, the CKV internals will continue to serve as the system pressure boundary when the hydrogen analyzers are placed in operation.
[ Reference 3, Attachment 10; Reference $5] Since removal of the CKV internals from their environment is not planned, there are no programs credited with mitigating the effects of clastomer degradation for the components in this group.
Discovery: Baltimore Gas and IIIectric Company currently plans to include the CKVs in the gas return line to containment from PASS in an ARDI Program to verify that unacceptable elastomer degradation of the valve internals is not occurring. For a discussion of the elements of the ARDI Program, refer to subsection Group I - Aging Management Programs, above.
Group 5-(elastomer degradation for valve Internals) Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to elastomer degradation for the CKVs in the gas return line to containment from PASS:
+ 1he CKVs in the gas return line to containment from PASS contribute to maintaining the pressure boundary of the NSSS Sampling System. The integrity of these components must be maintained under all CLB design conditions.
'Ihe materials of construction for seating surfaces of these components include clastomers.
Elastomer degradation is a plausible ARDM for this group because the valve internals are affected by scission, crosslinking, and changes associated with campounding ingredients in their installed locations. If unmanaged, degradation of these subcomponents could eventually result in the loss of pressure-retaining capability under CLD design loading conditions.
Application for License Renewal 5.13 32 Calvert Cliffs Nuclear Power Plant t
NITACHMENT d)
APPENDIX A - TECilNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM a
1hc CKVs in the gas return line to containment from PASS will be subjected to a new ARDI Program. 'lhis program will examine a representative sample of the components for degradation, and ensure that appropriate corrective actions are initiated on the basis of the findings.
1herefore, there is a reasonable assurance that the effects of elastomer degradation will be adequately managed for CKVs in the gas return line to containment from PASS such thra they will be capable of perfomiing their intended functions consistent with the CLB during the period of exteaded operation under all design loading conditions.
Group 6. (wcar for CVs associated with sampling the RCS hot leg)- Materials and Environment Group 6 consists of the CVs in the reactor coolant sampling subsystem associated with sampling the RCS hot leg whose seating surfaces are subject to wear. Specifically, this group consists of the RCS hot leg sample header isolation CV (l[2]CVPS $467) and the RCS sample isolation CV (l[2]CVPS 5464) as
, depicted in Figure 5.131. These CVs provide the passive intended function of maintaining the system pressure boundary, [ Reference 3. Attachment 1] Additionally, the seating surfees of these CVs provide a containment isolation boundary, lhe valve disk and cage are constructed of stainless steel.
[ Reference 3, Attachments 4 and 5 for CVs]
As described in subsection Group 4 Materials and Environment, above, the N.crnal environment affecting these valves is borated water in the reactor coolant sampling subsystem. As described in subsection Group 1 Materials and Environment, above, the external environment is climate-controlled air in the Auxiliary fluilding and the Containment. [ Reference 3, Attachment 3]
Over a 60 year service life, these valves can be expected to be cycled between 9,000 and 10,000 times to support routine hot leg sample operations. Similar NSSS Sampling System CVs associated with the pressuriter surge line and vapor space sample heaoers are excbided from Group 6 since samples from the pressurizer are no longer being taken.
Group 6 -(wear for CVs associated with sampling the RCS hot leg) . Aging Mechanism Effects Wear for the Group 6 valve internals results from relative motion between two surfaces (adhesive wear).
Motions may be linear, circular, or vibratory in inert or corrosive environments, in addition to material loss from this mechanism, impeded relative motion between two surfaces held in intimate contact for extended periods may result in galling /scif welding. Wear rates may accelerate as expanded clearances result in higher contact stresses. [ Reference 3, Attachment 7 for Valves]
The Group 6 valve internals are required to maintain the containnient pressure boundary, Wear is considered plausible for the seating surfaces of the CVs in the RCS hot leg sampling lines because they may experience cyclic relative motion at tight fitting surfaces. Mosement between the internal subcomptment parts during normal valve operation can result in a gradual loss of material, which could result in a small amount of leakage. If left unmanaged, wear could eventually result in a loss of leak tightness such that the Group 6 components may r.ot be able to perform their pressure boundary and containment isolation functions under CLU conutions.
Application for License Renewal 5.13-33 Calvert Cliffs Nuclear Power Plant
~ _ __. . . . . - . _ _ _ _ _ . __ _ _ _ ._. _ _ _ _ . _ .
s- .
t ATTACliMENT (3) l APPENDIX A - TECHNICAL INFOlWATION ;
5.13 - NSSS SAMPLIhG SYSTEM Group 6 -(wear for CVs associated with sampling the MCS hot leg)- Methods to Manage Agleg E Mitlation: Since wear of the valve internal subcomponent parts is dependent on the frequency of valve operation, decreased operation of the valves would slow the degradation of the valve seating surfaces.
Proper material selection for the valve internal parts can also slow the clTects of wear, it should also be noted that periodic valve operation can actually reduce the likelihood of the galling /self welding phenomenon.
Dhtscry: Wear can be discovered by inspecting the CVs that are susceptible to this ARDM. In addition, the effects of wear can be detected by performing leak rate testing. Since wear occurs gradually over time, periodic testing can be used to discover minor leakage so that corrective actions ccm be taken priir to the loss of an intended function, Group 6 - (wcar for CYs associated with sampling the RCS hot leg) Aging Management Program (s)
Mitigation: Operation of the CVs in Group 6 is dependent on RCS sampling requirements established by the CCNPP Chemistry Program. Maintaining the established sampling frequencies precludes modifkation of plant operating practices to reduce wear on the NSSS Sampling System. Therefore, there are f 3 practicable means available (beyond proper material selection) to mitigate the etfnM of wear.
Dhcarry: Calvert Cliffs procedures STP M 571 A 1(2), " Local Leak Rate Test, Penet ations l A, IB, IC," v hich cover local leak rate testing (LLRT) for the RCS hot leg sampling containment penetrations, ore part of the overall CCNPP Containment Leakage Rate Testing Program. (References 56 and 57] The valves that isolate the containment penetration piping for the RCS hot leg sample header s e included in .
the scope of this program as part of the leakage testing for the associated containment penetrations.
[ Reference 37] The CCNPP Containment Leaktge Rate Testing Program is dis:ussed further in subsection Group 2,- Aging Management Programs, above.
The corrective actions taken as part of the LLRT Program wil ensure that the CVs in the RCS hot leg sampling lines remain capable of performin8 their intended functions under all CLB conditions during the period of extended operation.
Group 6 - (wear for CVs associated with sampling the RCS hot leg) Demonstration of Aging Management Based on the information presented above, the following conclusions can be reached with respect to wear for the Group 6 components in the NSSS Sampling System:
The CVs in the RCS hot leg sampling lines contribute to maintaining the system pressure boundary and provide contain'nent isolation. Their integrity must be maintained under all CLB design conditions.
The material for the seating surfaces of these valves is stainless steel.
Wear is a plausible ARDM for the seating surfaces of these valves because of the repeatu relative motion resulting from routine sampling operations. If unmanaged, wear could eventually result in the loss of piessure-retaining capability under CLB design loading conditions.
Application for License Renewal 5.13 34 Calvert Cliffs Nuclear Power Plant
ATTAClIMENT (3)
APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM !
- The CCNPP LLRT Program performs leakage testing which could detect the effects of wear on the seating surfaces of the CVs in the RCS hot leg sampling linen (i.e., degraded leak tightness).
This program ensures that appropriate corrective actions will be taken if significant leakage is discovered. ;
Therefore, there is a reasonable assurance that the effects of wear will be adequately managed for the CVs in the RCS hot leg sampling lines 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 condit!ons.
5.13.3 Conclusion The aging management programs <liscussed for the NSSS Sampling System are listed in Table 5.13 3.
These programs are administratively controlled by a formal review and approval process. As demonstrated abo.'e, these programs will manage the aging mechanisms and their effects in such a way that the intended functions of the components of the NSSS Sampling Fystem 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." Qt 2 is pursuant to 10 CPR Part 50, A.ppendix II, and covers all structures and components subject to AMR.
Application for License Renewal 5.13-35 Calvert Clifts Nuclear Power Plant
0 ATTACHMENT (3) !
APPENDIX A TECilNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM 3 Table 5.13 3 AGING MANAGEMENT PROGRAMS FOR Tile NSSS SAMPLING SYSTEM Program Credited As fixisting CCNPP Administrative Procedure Program for tr.Neation and discovery of general MN 3 301,"floric Acid Corrosion corrosion for external surfaces of sample coolers, inspection Program" CVs, and ilVs (included in Group 1) that are exposed to borated water (due to leakage) by perfctming visual inspections.
lixisting CCNPP Technical Procedure CP 204, Progran for mitigation of crevice corrosion and "Speclucation and Surveillance pitting for internal surfaces of sample coolers, ilVs, Primary Systems" and SVs (included in Group 2) that are exposed to borated water (as process Guid) by controlling chemistn. conditions.
lixisting CCNPP Technical Procedure CP 206, Program for mitigation of erevice corrosion and
" Specification and Surveillance pitting for internal surfaces of IlXs (included in Component Cooling / Service Water Group 2) that are exposed to chemically treated System" water from the CC System by controlling chemistry conditions in the CC System.
lixisting CCNPP Technical Procedure CP 217, Piogram for mitigation of crevice corrosion snd "Speci0 cations and Surveillance: pitting for internal surfaces of sample coolers and Secondary Chemistry" IIVs (included in Grcup 2) that are exposed to steam and feedwater in the steam generator blowdown sampling subsystem (as process Huld) by controlling chemistry conditions.
lixisting CCNP" Surveillance Test Program for discovery and management of lec'. age M 5711 1(2),"lecal leak Rate Test, that could be the result of crevice corrosion and Penetrations ID,47A,471),47C, pitting for seating surfaces of the containment 47D, 48A, 4811,49A, 4911,49C" isolation SVs in the sample return lines from the reactor coolant ample hoods to the RCDT (included in Group 2).
lixisting CCNPP Preventive Maintenance Program for mitigation of general corrosion for Checklists IPM 10000 (10001), CVOPs (included in Group 3) by controlling lA
" Check Unit 1(2) Instrument Air quality.
Quality" lixisting CCNPP Surveillance Test Program for discovery and management of leakage M 571 A 1(2)," Local Leak Rate Test, that could be the result of wear for seating surfaces Penetrations l A, IB, IC" of the CVs in the RCS hot leg sampling lines (included in Group 6),
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Application for License Renewal 5.13 36 Calvert Cliffs Nuclear Power Plant
ATTACllMENT (3)
APPENDIX A - TECl!NICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM Program Credited As !
Modified CCNPP FMP Program for discovery and management of thermal f r piping and valves in the RCS hot leg
. Evaluate piping and valves in the I*IIE"?
samplmg line (Group 4) by evaluatmg low-cycle RCS hot leg sampling line under fatigue usage.
the FMP New ARDI Program Progrk n for discovery and management of general corrosion for external surfaces of the miscellaneous -
waste evapor stor concentrate pump discharge sample cooler (included in Group 1) by identifying and correcting degraded conditionr Program for discovery and mtnagement of crevice corrosion and pitting for internal surfaces of IlXs, llVs, and SVs (included in Group 2) by identifying and correcting degraded conditions.
Program for discovery and management of clustomer degradation for internals of the CKVs in the gas return line to containment from PASS (included in Group 5) by identifying and correcting degraded conditions.
Application for License Renewal 5.13 37 Calvert Cliffs Nuclear Power Plant
AlIACHMENT (3)
APPENDIX A TECilNICAL INFORMATION 5.13 NSSS SAMPLING SYSTEM ,
3 5.13.4 References
- 1. CCNPP Life Cycle Management System and Structure Screening Results, Revision 4
- 2. CCNPP Updated Final Safety Analysis Report (UFSAR), Units I and 2, Revision 20
- 3. CCNPP Aging Management Review Report,"NSSS Sampling System," Revision 2
.t CCNPP Operating instruction. 01310. "NSSS Sampling ,ystem," Revision 10
- 5. Letter from Mr. D.11. Jaffe (NRC) to Mr. J. A. Tiernan (B' C - ned May 6,1986, " Safety Evaluation, Pest Accident Sampling System, NUREG 0737,Ittm 11.B.3"
- 6. Letter from Mr. J. A. Tiernan (BGE) to NRC Document Contro Desk, dated March 11,1988, "Calven Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50 317 & 50 318 Repon of Changes, Tests and Experiments"
- 7. DGE Drawing 60724S11000), " Reactor Coolant and Waste Psocess Sample System, Post Accident Sampling System," Revision 45
- 8. Letter from Mr. R. W. Starostecki (NRC) to Mr. A. E. Lundvall, Jr. (BGE) dated March 25,1983,"RI inspection 50 317/8tCS,50 31W83 05"
- 9. Letter from Mr. A. E. Lundvali, Jr. (BGE) to Mr. R. A. Clark (NRC), dated November 30,1982,"Calven Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50 317
& $0 318 TMI Action Plan item li.D.3"
- 10. Letter tro n Mr. L. D. Russell (BGE) to T. 3. Murley (NRC), dated October 25,1985,"Calven Cliffs Nuclear Power Plant Un! s 1 and 2,Inoperability of Post Accident Sampling System"
- 11. Letter from Mr. T. T. Mertin (NRC) to Mr. J. A. Tieman (DGE) dated April 1,1987,
" Combined Inspection Nos,50 317/87 03 and 50-318/87-03"
- 12. IlGE Drawing 60724Sil0002, " Reactor Coolant and Waste Process Sample System, Post Accident Sampling System," Revision 12
- 13. DGE Draiwing 60724S110003, " Reactor Coolant and Waste Process Sample System, Post Accident Sampling System," Revision 24
- 14. BGE Drawing 60744S110001 " Gas Analyzing System," Revision 14 15 UGE Drawing 60744S110002," Gas Analyzing System," Revision 13
- 16. CCNPP Engineering Standard ESe.1, "Syste. t, Structure, and Component (SSC) Evaluation,"
Revision 2
- 17. BGE Drawing 92769Sil-CC 2,"M 601 Piping Class Summary," Revision 20
- 18. DGE Drawing 92769Sil CC 1,"M 601 Piping Class Sumiaary," Revision 23
- 19. DGE Drawing 92769Sil llc-4,"M 601 Piping Class Summary," Revisian 21
- 20. CCNPP Sampling (NSSS) System Component Level ITLR Screening Results, Revision 1
- 21. CCNPP Aging Management Review Report," Instrument Line Commodity," Revision 1
' 22. BGE Drawing 92769Sil-IIC 1,"M 601 Piping Class Summary," Revision 26 Application for License Renewal 5.13 38 Calvert Cliffs Nuclear Power Plant l
ATTACllMENT (3)
APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM
- 23. CCNPP Engineering Standard ES-014, " Summary of Ambient Environmental Service Conditions," Revision 0
- 24. CCNPP Ad.ninistrative Procedure MN-3 301, " Boric Acid Corrosion Inspection Program,"
Revision i
- 25. CCNPP Administrative Procedure MN 3 il0, " Inservice Inspection of ASME Section XI Components," Revision 2
- 26. BGE Drawing 92769Sil EB 2,"M 601 Piping Class Summary," Revision 19
- 27. BGE Drawing 92769Sil.IIC-3,"M 601 Piping Class Summary," Revision 20
- 28. BGE Drawing 92769Sil-flC-6. "M.601 Piping Class Summary," Revisica 27
- 29. BGE Drawing 92769Sil-GC-i, M-601 Piping Class Summary," Revision 23
- 30. CCNPP Nuclear Program Directive Cil 1," Chemistry Program," Revision 1
- 31. CCNPP Technical Procedure CP 204, " Specification and Surveillance Primary Systems,"
Revision 8
- 32. CCNPP Technical Procedure CP 217," Specifications and Surveillance: Secondary Chemistry,"
Revist,n 5
- 33. CCNPP Technical Procedure CP-206, " Specifications and Surveillance Compenent Cooling / Service Water System," Revision 3
- 34. BGE " Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant," Revision 48
- 35. CCNPP Surveillance Test Procedure M 5?ll 1," Local Leak Rate Test, Penetrations ID,47A, 47B,47C,47D,48A,48B,49A,49B,49C"(Unit 1), Revision 0
- 36. CCNPP Surveillance Test Procedure i,157112," Local Leak Rate Test, Penetrations ID,47A, 47B,47C,47D,48A,488,49A,498,49C"(Unit 2), Revision 1
- 37. Letter from Mr. A. W. Drcmerick (NRC) Mr. C.11. Cruse (BGE), dated February 11,1997,
" Issuance of Amendments for LCNPP Unit No.1 (TAC No, M97341) and Unit No. 2 (TAC No, M97342)"[ Amendment Nc: 2 W 196)
- 38. 10 CFR Pac. 50 Appendix J, " Primary Reactor Containment Leakage Testirig for Water-Ccsled Power Reactors"
- 39. Letter from Mr. C. :L Oruse (BGE) to NRC Document Control Desk, dated November 26,1996, T,h CJ7s Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 A 50-318 License Amenament Request: Adoption of 16 CFR Part 50, Appendix J, Option B for Types B and C Testing"
- 40. BGE Drawing 60712S110001, " Compressed Air Sptem, Instrument Air and Plant Air,"
Revision 46
- 41. BGE Drawing 60712S110003, " Compressed Air System, Instrument Air and Plant Air,"
Revision 75 -
- 42. BGE Drawing 62712ShJ001, " Compressed Air System, Instrument Air and Plant Air,"
Revision 37
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Application for License Renewal 5.13-39 Calver* Cliffs Nuclear Power Plant 1
NITACHMENT G)
APPENDIX A - TECHNICAL INFORMATION 5.13 - NSSS SAMPLING SYSTEM
- 43. BGE Drewing 62712S110003, " Compressed Air System, !mtrument Air and Plant Air,"
Revision 80
- 44. CCNPP Aging Management Review Report," Compressed Au System," Revision 4
- 45. CCNPP Administrative Procedure MN l.102," Preventive Maintenance Prgrara," Revision 5
- 46. CCNPP NUCLEIS Database Repetitive Tasks 10191024 (2019:012), "Caeck Unit I (2)
Instrument Air Quality at Selected System Low Points"
- 47. CCNPP Preventive Maintenance Checklists IPM 10000 (10001), "Chec : Unit 1(2) Instrument Air Quality," Revision 2
- 48. CCNPP Specification No. SP-248D," Nuclear Class Control Valves," Fevision 2
- 49. " Metal Fatigue in Engineering,"11. O. Fuchs and R. I. Stephens, John Wiley & Sons, Copyright 1980
- 50. Combustion Engineering Owners Group Task 571, Report No. CE-NPSD-634 P, " Fatigue Monitoring Program for Calvert Cliffs Nuclear Power Plants Units I and 2," April 1992
- 51. CCNPP Administrative Procedure EN-l 300, " Implementation of Fatigue Monitoring,"
Revision 0
- 52. Letter from Mr. J. P, Durr (NRC) to Mr. C. Stoiber (sic) (BGE) dated February 11, 1993,
" Inspection port Nos. 50-317/92 32 and 50 318/92-32"
- 53. BGE Procurement Specification 6422284S, " Technical Services to Evaluate Thermal Fatigue Effects on Calvert Cliffs Nuclear Power Plant Systems Requiring Aging Management Review for License Renewal," Revision 0
- 54. NUREG-0933, Generic Safety issue 166, " Adequacy of Fatigue Life of Metal Components,"
Revision 1, June 30,1995
- 55. (MNPP Engineering Service Package ES199601279," Isolate the Unused PASS System from
'e Remainder of the Plant," June 5,1996
- 16. CCNPP Surveillance Test Procedure M 571 A-1, " Local Leak Rate Test, Penetrations I A, IB, W"(Unit 1), Revision 0
- 57. CCNPP Surveillance Test Procedure M 571 A-2, " Local Leak Rate Test, Penetrations I A, IB, IC"(Unit 2), Revision 0 l
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Application for License Renewal 5.13-40 Calvert Cliffs Nuclear Power Plant
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