ML20195J618

Forwards Responses to 980903 RAI for Review of Ccnpp,Units 1 & 2 Integrated Plant Assessment Rept for Rcs,Per License Renewal Application

ML20195J618
Person / Time
Site:	Calvert Cliffs
Issue date:	11/19/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9811250033
Download: ML20195J618 (15)

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CHA01.E0 H. CRUSE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 November 19,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 1

Response to Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Report for the Reactor Coolant System l

REFERENCES:

(a)

Letter from Mr. C. II. Cruse (BGE) to NRC Document Control Desk, dated December 17,1997," Request for Review and Approval of System and Commodity Reports for License Renewal" j

(b)

Letter from Mr. D. L. Solorio (NRC) to Mr. C. IL Cruse (BGE),

September 3,1998," Request for AdditionalInformation for the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrate Plant Assessment Reports for the Reactor Coolant System System" (c)

Letter from Mr. D. L. Solorio (NRC) to Mr. C. IL Cruse BGE),

September 24,1998, " Renumbering of NRC Requests for Additional Information on Calvert Cliffs Nuclear Power Plant License Renewal Application Submitted by the Baltimore Gas and Electric Company" Reference (a) forwarded three Baltimore Gas and Electric Company (BGE) system and commodity reports for license renewal. Reference (b) forwarded questions from NRC staff on one of those three reports, the Integated Plant Assessment Report on the Reactor Coolant System.

Reference (c) j forwarded a numtsering system for tracking BGE's response to all of the BGE License Renewal Application requests for additional information and the resolution of the responses. Attachment (1) g provides our responses to the questions contained in Reference (b). The questions are renumbered in I

accordance with Reference (c).

r o 9811250033 981119 _

PDR ADOCK 05000317 p

PDR NRC Distribution Code A0360

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s Do.cument Control Desk l

November 19,1998 Page 2 Should you have further questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, W

STATE OF MARYLAND

TO WIT:

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

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Subscri ed and worn before me, a Notary Public in and for the State of M::ryland and County of

,this 19 dayof Aow.mlue,1998.

WITNESS my Hand and Notarial Seal:

Notary Public

!h My Commission Expires:

Mte CHC/KRE/ dim Attachments: (1) Response to Request for Additional Information; Integrated Plant Assessment Report for the Reactor Coolant System cc:

R. S. H.eishman, Esquire C. I. Grimes, NRC J. E. biiberg, Esquire D. L. Solorio, NRC S. S. Bajwa, NRC Resident Inspector, NRC A. W. Dromerick, NRC R. I. McLean, DNR H. J. Miller, NRC

1. H. Walter, PSC l

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' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT, REACTOR COOLANT SYSTEM i

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Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant November 19,1998 i

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T ATTACHMENT (1) l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR COOLANT SYSTEM NRC Onention No. 4.13 Table 4.1-2 of the application (Baltimore Gas and Electric Company's [BGE's] License Renewal l

Application [LRAJ) indicates that Reactor Coolant System (RCS) piping with " device codes" of"-CC,"

" GC,""-HB," and "-liC" are subject to aging management review (AMR). Please explain these " device codes" and describe components represented by them. Also, the description should identify wh::ther I

these components include cold-leg, hot-leg, pressurizer surge line, spray line, connected American Society of Mechanical Engineers (ASME) Class I branch lines, and nozzles and safe ends at the reactor l

vessel, steam generators, pressurizer, pumps, and valves.

BGE Response Baltimore Gas and Electric Company uses " device codes" for piping to designate different design parameters and materials. For the device codes listed in Section 4.1 of the LRA, the following piping

" device codes" were used:

-CC - This category represents the majority of piping in the RCS, including the cold-leg, hot-e leg, spray line, pressurizer surge line, nozzles and sa e ends. This pipe category is made from r

carbon steel with a stainless steel clad or Type 316 stainless steel, American Society for Testing and Materials (ASTM) A-312, A-376 or A-351 CF8M, and has a design pressure rating of 2485 psig.

-GC - This category represents pip:ng connected to the RCS that is made from Type 304 stainless steel, ASTM A 312 or A-376 r.nd has a design rating of 350 psig.

-HB - This category represents piping associated with the reactor coolant pump (RCP) lubrication oil system that is made from carbon steel, ASTM A-106 Grade B, and has a design rating of 150 psig.

-HC - This category represents other associated piping that is made from Type 304 stainless e

steel, ASTM A-312 or A-376, and has a design pressure rating of 150 psig.

NRC Ouestion No. 4.1.6 Provide a summary of the RCS piping sizes, piping material, and the corrosion allowances used in the design. Describe the basis upon which BGE concluded that the corrosion allowances are adequate for the period of extended operation.

BGE Response The RCS piping sizes and materials of construction are located in the Calvert Cliffs Nuclear Power -

Plant (CCNPP) Update Final Safety Analysis Report Chapter 4.1.3.

Baltimore Gas and Electric Company does not have a corrosion allowance for RCS piping. The RCS piping is either carbon steel, with a corrosion resistant stainless steel clad interior surface, or it is entirely stainless steel. The only corrosion expected to occur would be on the outside of the carbon steel piping due to either moisture or boric acid leakage. The Boric Acid Corrosion Inspection (BACI) Program, which is described in Section 4.1.2 of the LRA, is credited with finding corrosion on the outside surface of the I

carbon steel RCS piping.

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I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR Tile REACTOR COOLANT SYSTEM NRC Ouestion No. 4.1.7 The application does not apparently discuss several aging effects associated with certain RCS components. Summarize how the following aging effects have been addressed by BGE's AMR.

crack initiation and growth (stress corrosion cracking [ SCC]) for the pressurizer shell/ heads a.

I (including clad cracking), spray line nozzle, surge line nozzle, valve nozzle, manway, support

. skirt, integral attachments, and Unit 2 heater sleeve;

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b. corrosion and boric acid wastage for the pressurizer instrument nozzle and integral attachments;

c.. loss of preload for the pressurizer manway bolting;

d. crack initiation and growth (SCC) for the RCS carbon steel -- hot and cold leg piping, nozzles, safe ends, and integral support; SCC for stainless steel -- RCP nozzles, safety and relief valve bodies and body flanges, bonnet e.

and bonnet flanges, and nozzles; hot and cold leg, surge line, spray line, nozzles and safe ends;

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for stainless steel auxiliary piping of the decay heat removal system, core flood system and any other included Class 1 piping; fittings, nozzles, and safe ends of auxiliary piping; and component integral supports; cast austenitic stainless steel (CASS)-- RCP casing, cover, casing flange, cover flange; safety and relief valve bodies, bonnets, body and bonnet flanges; cold and hot legs; surge line, nozzles; fittings, nozzles, and safe ends of auxiliary piping; f.

SCC for nickel alloy -- auxiliary piping safe ends; g.

SCC for high strength low alloy steel-RCP closure bolting and safety valves closure bolting;

h. general corrosion (boric acid corrosion from leakage of primary coolant) for integral supports (carbon steel), safety and relief valve bodies, bonnets, body flange, bonnet flange (stainless steel and CASS); RCP casing, cover, casing flange, cover flange (CASS); and safety valve closure bolting;

i. thermal embrittlement for CASS components -- RCP casing and cover flanges; safety and relief valve body, bonnet, body and bonnet flange, hot and cold legs; surge lines; nozzles and safe ends; auxiliary piping fittings, nozzles, and safe ends;

j. loss of preloadhtress relaxation for RCP closure bolting and safety and relief valve closure bolting.

BGE Response Baltimore Gas and Electric Company reviewed the list of aging effects and components above and has answered these according to the format given in the request for additional information below:

a.

Crack initiation and growth (SCC) for the:

Pressurizer shell/ heads (including clad cracking)- SCC is not plausible.

l Spray line nozzle forging - SCC is not plausible.

l Spray line nozzle safe end - SCC is plausible and is managed by the Inservice Inspection (ISI) Program (CCNPP Administrative Procedure MN-3-110, " Inservice Inspection of l

ASME Section XI Components") as described in Group 7 in Section 4.1.2.

Surge line nozzle forging and safe end - SCC is not plausible.

Valve nozzle - The spray nozzle in the pressurizer is not within scope of license renewal.

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A'ITACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACI'OR COOLANT SYSTEM Manway - SCC is not plausible for the manway cover plate, but is plausible for the manway bolts, which is managed by the DACI Program (Group 7 in Section 4.1.2)

Support skirt - SCC is not plausible.

Integral attachments - Attachments to components that are not safety-related are not within the scope oflicense renewal. Integral attachments (e.g., supports) to safety-related components are within the scope of license renewal and are evaluated for aging management in Section 3.1, " Component Supports," and Section 3.l A, " Piping Segments that Provide Structural Support." The weld between the attachment and the RCS component was considered and evaluated for aging as part of the RCS component.

Unit 2

heater sleeve SCC / primary water stress corrosion cracking (PWSCC)/intergranular stress corrosion cracking (IGSCC) is not plausible (modified to Alloy 690). Primary water stress corrosion cracking, IGSCC, and SCC are plausible for

. Alloy 600 weld material as shown in Table 4.1-3.

b. Corrosion and boric acid wastage for the pressurizer instrument nonle - Corrosion is not plausible (stainless steel). The BACI Program and the ISI Program were credited with finding corrosion on integral attachments susceptible to such corrosion that are connected to safety-related RCS equipment.

Loss of preload for the pressurizer manway bolting - Loss of preload is covered under c.

corrosion (Group 5) and SCC (Group 7) of bolting in Section 4.1.2. The BACI Program and ISI Program will discover the effects of corrosion on bolting.

d. Crack initiation and growth (SCC) for the RCS carbon steel:

Hot and cold leg piping - SCC is not plausible.

Nonles forgings-SCC is not plausible.

Safe ends - SCC is not plausible.

Integral support - Attachments to components that are not safety-related are not l

within the scope of license renewal. Integral attachments (e.g., supports) to safety-related components are within the scope of license renewal and were evaluated for aging management in Sections 3.1 and 3.l A. The weld between the attachment and l

the RCS component was considered and evaluated for aging as part of the RCS component.

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SCC for stainless steel components:

RCP noules - RCPs have a casing that is not susceptible to SCC.

Safety and relief valve bodies and body flanges - SCC is not plausible.

l Bonnet and bonnet flanges, and nonles - SCC is not plausible.

Hot and cold leg - SCC is not plausible (made of CS with SS clad).

Surge line - SCC is not plausible.

Spray line - SCC is not plausible.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; j

INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR COOLANT SYSTEM l

Nonles and safe ends - SCC is not plausible for the surge line nonle and safe ends, and the spray line nonle. SCC is plausible for spray line nonle safe ends and is managed by the ISI Program as described in Group 7 in Section 4.1.2..

For stainless steel auxiliary piping; i.e., RCS piping connecting to the decay heat removal system:

Core flood system and any other included Class 1 piping - SCC is not plausible.

Fittings - Bolting is susceptible to SCC 'and is managed by the BACI Program (Group 7 in Section 4.1.2).

Nonles and safe ends - SCC is not plausible.

l Component integral supports - Attachments to components that are not safety-related are not within the scope of license renewal. Integral attachments (e.g., supports) to safety-related components are within the scope of license renewal and are evaluated for aging management in Sections 3.1 and 3.l A. The weld between the attachment and the RCS component was considered and evaluated for aging as part of the RCS component.

CASS:

RCP casing, cover, casing flange, cover flange - SCC is not plausible (thermal embrittlement is plausible and managed by the CASS Evaluation Program - Group 8).

Safety and relief valve bodies, bonnets, body and bonnet flanges - Relief valve's base is CASS and SCC is not plausible, electrically-operated relief valve cage is CASS and SCC is not plausible.

' Cold and hot legs - Are not made of CASS.

Surge line, nonles - SCC is not plausible.

Nonles and safe ends of auxiliary piping - Are not made of CASS.

f.

SCC for nickel alloy; auxiliary piping safe ends - There are none; therefore, not applicable.

g. - SCC for High strength low alloy steel:

RCP closure bolting and safety valves closure bolting - SCC is plausible and managed by the BACI Program (Group 7 in Section 4.1.2).

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h. General corrosion (boric acid corrosion from leakage of primary coolant) for integral supports (carbon steel), safety and relief valve bodies, bonnets, body flange, bonnet flange (stainless steel and CASS); RCP casing, cover, casing flange, cover flange (CASS); and safety valve 3

closure bolting; General corrosion is plausible for carbon steel and alloy steel (not CASS) components that are subject to boric acid leakage. The components subject to general corrosion in the Section 4.1.2, Group 5, were:

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; I

INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR COOLANT SYSTEM l

l General Corrosion - External Device Type -CC - Pipe, elbows, and nozzle forging (carbon steel), bolting studs (alloy steel),

bolting hex nuts (carbon steel);

Device Type -GC - Bolting studs and bolting hex nuts (carbon steel);

e Device Type CKV - Some of the check valves' bolting (carbon steel);

Device Type ERV - Bracket stud (alloy steel) and nut (carbon steel);

e Device Type HX S/G - Primary manway (alloy steel, clad - Alloy 600), manway cover plate e

(carbon or alloy steel), primary head torus (carbon steel with stainless steel clad), spherical j

head (carbon steel); secondary manway studs (alloy steel) and hex nuts (carbon steel),

secondary manway yoke (carbon steel), handhole studs and nuts (alloy steel); primary manway studs and nuts (alloy steel); lower support sliding base and cap plate (carbon steel),

lower support flange bolts (alloy steel), and flange nuts (carbon steel);

Device Type MOV - Bonnet stud and nut (carbon steel);

e Device Type PUMP - Closure studs and nuts (carbon steel);

Pressurizer - Alloy steel shell, top head and bottom head (alloy steel); safety / relief valves, spray and surge nozzle forgings (forged alloy steel); manway forging (alloy steel), manway I

cover plate (carbon steel), manway bolting studs and bolts (alloy steel); carbon steel welds; support ring assembly and base ring assembly (carbon steel), support skirt forging (alloy steel), and lifting lugs (carbon steel); and Device Type RV - Bonnet / spring / bonnet studs (carbon or alloy steel).

e The programs credited for the discovery of general corrosion were the BACI Program, ISI Program and CCNPP Procedure SG-20,"SG Primary h1anway Cover Removal and Installation."

Integral supports attached to safety-related components are within the scope of license renewal and are evaluated for aging management in Sections 3.1 and 3.l A of the BGE LRA. The weld between the support / attachment and the RCS component was considered and evaluated for aging as part of the RCS component.

i. Thermal embrittlement for CASS components - RCP casing and cover flanges; safety and relief valve body, bonnet, body and bonnet flange, hot and cold legs; surge lines; nozzles and safe ends; auxiliary piping fittings, nozzles, and safe ends; The RCS components subject to thermal embrittlement in Section 4.1.2, Group 8, of the LRA are those listed below. These components will be managed by the new CASS Evaluation Program.

Other CASS components in the RCS that were not subject to sufficiently high temperatures were not considered to be subject to thermal embrittlement.

Device Type-CC - Surge pipe, surge elbows; surge nozzle safe end, shutdown cooling i

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nozzle safe end, safety injection nozzle safe end (CASS);

l Device Type PUMP -(RCP) case and pump cover (CASS); and Device Type PZV - Surge nozzle safe end (CASS).

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR COOLANT SYSTEM i

J. Loss of preload/ stress relaxation for RCP closure bolting and safety and relief valve closure t

bolting.

Loss of preload is covered under corrosion (Group 5) and SCC (Group 7) of RCP closure bolting and safety / relief valve bolting in Section 4.1.2. The BACI Program and ISI Program will discover the effects of corrosion on bolting which would lead to loss of preload.

l Baltimore Gas and Electric Company determined that stress relaxation was not plausible.

NRC Ouestion No. 4.L8 l

The application does not apparently contain an AMR of the following pressurizer components: heater i

belt forgings; heater sheaths and end caps; heater bundles; and bundle cover plates. If these components j

are applicable to the Calvert Cliffs units, describe where these components are addressed in the LRA, or justify why these components have been excluded.

BGE. Response Baltimore Gas and Electric Company evaluated all the pressurizer components that provided a license renewal intended function. The specific components related to the pressurizer heaters were evaluated and the heater support plates were not in scope for license renewal because they did not contribute to a license renewal intended function.

Those pressurizer heater related components that were evaluated for AMR were:

Heater sheaths; Heater sleeves and bushings; e

Heater end plugs; e

Heater seal rings; and e

Heater !ocking collar assemblies.

e The pressurizer components above were discussed in Section 4.1.2, Group 4 - Fatigue, since this was the only plausible age-related degradation mechanism (ARDM) for these components. Because all pressurizer components were susceptible to fatigue, they were not individually listed. The AMR report for the RCS lists all of the specific pressurizer components subject to eging management.

NRC Ouestion No. 4.1.9 For the following aging effects and components, summarize the extent to which BGE relies upon the associated programs for aging management, and provide examples of any operating experience that l

demonstrates the effectiveness of the programs that are relied upon to manage these aging effects:

a.

boric acid corrosion - Technical Specifications (TS) leakage limits, and ASME Section XI, Subsection lWB, examination categories B-P;

b. cracking oflarge bore piping -- ASME Section XI, Subsection IWB, examination categories B-J

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and B-F, and flaw evaluation criteria IWB-3000; cracking of small bore piping (less than 4-inches but greater than 1-inch diameter) -- augmented c.

volumetric ISI; and, because some safe ends and welds on small bore piping are of Inconel, t

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ATTACHMENT (1) l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACI'OR COOLANT SYSTEM information resulting from the assessment of NRC Information Notice 90-10 " Primary Water Stress Corrosion Cracking (PWSCC) ofinconel 600;"

d. cracking of bolting -- programs consistent with ASME Section XI, Subsection IWB, examination categories B-G-1 and B-G-2, and NRC Bulletin 82 02 " Degradation of Threaded Fasteners in the RCPB of PWR;"

pressurizer shell, heads, heater belt forgings -- ASME Section XI, Subsection IWB, examination e.

categories B-B and B-P, and primary water chemistry; f.

pressurizer nozzles -- ASME Section XI, Subsection IWB, examination categories B-D, B-E, B-F, and B P, TS leakage limits, primary water chemistry, augmented inspection of small bore piping; and ifInconel is used, information resulting from Information Notice 90-10; g.

integral attachments - ASME Section XI, SubsectionIWB, examination category B-H, and primary water chemistry;

h. heater sheaths and end caps -- ASME Section XI, Subsection IWB, examination category B-P, and TS leakage limits;

i. loss of preload in bolting -- ASME Section XI, Subsection IWB, examination categories B-G-1, B-G-2, and B-P, response to NRC Bulletin 82-02 and Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants", and TS leakage limits.

BGE Response

' Table 4.1-4 is a compilation of all the programs relied upon for RCS aging management in LRA Section 4.1. Where applicable, operating experience for programs is given in Section 4.1.2 of the

-LRA.

boric acid corrosion -- The CCNPP " Reactor Coolant System Leakage Evaluation" a.

(Surveillance Test Procedures STP-O-27-1/2) is discussed on page 4.1-24 and credited only for managing aging effects caused by wear (Group 2 in Section 4.1.2) on certain RCS valve disks and seats. The BACI Program is the primary program used to discover corrosion caused by boric acid. The ISI Program is another method by which boric acid corrosion can be discovered. The BACI and ISI Programs and any applicable operating experience are discussed in detail on pages 4.1-21 through 4.1-23 of LRA Section 4.1.2 The BACI and ISI Programs are also credited with managing aging in Groups 3,5, and 7 of Section 4.1.2. In addition to these programs, Calvert Cliffs Procedures RCS-10, " Pressurizer Manway Cover Removal and Installation," and SG-20 are credited with discovering wear (Group 2 of Section 4.1.2) and the presence of boric acid on the manways studs. Procedure RCS-10 is described on page 4.1-23, and SG-20 is described on page 4.1-24 of the BGE LRA.

b. cracking of large bore piping -- The only plausible aging mechanisms for RCS piping are general corrosion (Group 5 of Section 4.1.2) for external surfaces of carbon steel piping, wear (Group 2 of Section 4.1.2) of pipe flanges, SCC /IGSCC (Group 7 of Section 4.1.2), fatigue (Group 4 of Section 4.1.2), and thermal embrittlement (Group 8 of Section 4.1.2). The programs relied upon for these aging mechanisms are the BACI Program, the Fatigue l

Monito' ring Program (FMP), the ISI Program and the CASS Evaluation Program. The ISI and L

BACI Programs were previously discussed above in response to part (a) of this response and are also credited with managing aging in Group 3 of LRA Section 4.1.2. The FMP and applicable operating experience are described on pages 4.1-30 through 4.1-32 of LRA 7

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L Section 4.1.2. The CASS Evaluation Program is a new program credited with the discovery of thermal embrittlement, which is discussed on page 5.1-51 of Section 4.1.2.

j cracking of small bore piping (less than 4 inch-but greater than 1-inch diameter)-- Except for c.

thermal embrittlement (CASS Evaluation Program), the aging mechanisms and programs for small bore piping safe ends and welds are the same as those for the large bore piping discussed in the response to part (b) of this RAl.

d. cracking of bolting - The only plausible aging mechanisms for bolting are wear,(Group 2 of Section 4.1.2), general corrosion (Group 5 of Section 4.1.2), fatigue (Group 4 of i

Section 4.1.2) and SCC (Group 7 of Section 4.1.2). The programs credited for discovering these aging mechanisms are the BACI Program, the ISI Program, and the FMP previously discussed in response to parts (a) and (b) above.

pressurizer shell, heads, heater belt forgings -- The pressurizer shell and heads are susceptible e.

to external general corrosion and fatigue. The BACI, ISI, and Fatigue Monitoring Programs are credited with discovering these aging mechanisms. These programs were previously discussed in response to parts (a) and (b) above. Those pressurizer heater-related components that were evaluated for AMR were the heater sheaths; heater sleeves and bushings; and heater end plugs, seal rings and locking collar assemblies. Fatigue is the only plausible ARDM for these components and it is managed by the FMP previously discussed in response to part (b) above.

f.

pressurizer nozzles -- The pressure, level, and temperature nozzle forgings (except the Unit 2 pressure and level forgings-made ofInconel 690) are susceptible to SCC /lGSCC/PWSCC and fatigue. The pressure, level, temperature, safety / relief valve and spray nozzle safe ends are

- susceptible to SCC /IGSCC/PWSCC and fatigue. The Unit 2 pressurizer pressure and level nozzle forgings (Inconel 690) are only susceptible to fatigue. The pressurizer safety / relief valve, spray and surge nozzle forgings are susceptible to general corrosion and fatigue. The pressurizer spray and surge nozzle thermal sleeve (Inconel 600) are susceptible to SCC /lGSCC/PWSCC and fatigue.

The pressurizer surge nozzle safe end (CASS) is susceptible to fatigue and thermal embrittlement. The programs credited with managing these aging effects are the BACl, ISI, FMP, and CASS Evaluation Program, which were previously discussed in response to parts (a) and (b) above.

g.

integral attachments -- Attachments to components that are not safety-related are not within the scope of license renewal.

Integral attachments (e.g., supports) to safety-related components are within the scope of license renewal and are evaluated for aging management in Sections 3.1 and 3.l A.

The weld between ar attachment and an RCS component is considered and evaluated for aging as part of the ES component.

h. heater sheaths and end caps -- Fatigue is the only pleible ARDM for these components and it is managed by the FMP previously discussed in respoe to part (b) above.

i. loss of preload in bolting -- The only plausible aging mechanisms for bolting are wear, (Group 2 of Section 4.1.2), general corrosion (Group 5 of Section 4.1.2), fatigue (Group 4 of Section 4.1.2) and SCC (Group 7 of Section 4.1.2). The programs credited for discovering these aging mechanisms are the BACl, ISI, and the FMP previously discussed in response to parts (a) and (b) above.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR COOLANT SYSTEM NRC Ouestion No. 4.1.10 Describe the. manner by which Procedure STP-M-574-1/2, "EC Examination of CCNPP 1/2 Steam Generators," manages aging effects.

BGE Response The procedures STP-M-574-1/2, " Eddy Current Examination of CCNPP Unit 1 Steam Generators,"

and " Eddy Current Examination of CCNPP Unit 2 Steam Generators," respectively, are fully described in Section 4.1.2, Groups 1, 2, 5, and 7 of the BGE LRA. These procedures are the implementation for the requirements of ASME Section XI and provide acceptance criteria for steam generator tube degradation. If SG tubes are found with unacceptable degradation an Issue Report is written to address the steam generator tube degradation.

NRC Ouestion No. 4.1.11 How is erosion / corrosion managed for the secondary manway and cover plate, hand hole and cover plate 7 BGE Response Baltimore Gas and Electric Company has determined that erosion / corrosion is not a plausible aging mechanism for the secondary manway and cover plate, hand hole, and cover plate, but that it was an operational concern that would be corrected soon after discovery. Any damage caused by leakage at those locations would be repaired prior to bringing the plant back to power operation.

NRC Ouestion No. 4.1.12 It appears that BGE used ferrite criteria to screen components subject to thermal embrittlement.

However, the NRC regards ferrite content as inadequate criterion for screening as stated in NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal."

Therefore, justify using ferrite content as screening criteria.

BGE Respanie The industry position regarding thermal embrittlement of CASS was developed and first documented in an industry report in 1994 (Section 4.2. of Reference 1). More recently, the industry position was sent via Reference (2). Baltimore Gas and Electric Company and the Electric Power Research Institute will be providing a technical presentation on this subject to NRC staff, tentatively scheduled for December 3,1998. Baltimore Gas and Electric Company' response to NRC's Question No. 4.3.14

' in Reference (3) is also related to this issue.

NRC Ouestion No. 4.1.13 Steam generator tubes have experienced intergranular attack (IGA). The application does not identify IGA as an aging issue. Provide basis for this determination.

BGE.Responic Baltimore Gas and Electric Company has elected to include IGA of the steam generator heat exchanger tubes under IGSCC, which is discussed in Section 4.1.2 - Group 7 of the LRA.

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ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REALTOR COOLANT SYSTEM Inteigranular attack of the steam generator heat exchanger tubes was considered as a subset of IGSCC. The procedures STP-M-574-1/2 are credited with the discovery of SCC and PWSCC of the steam generator heat exchanger tubes. Ilowever, Section 4.1.2 of the BGE LRA does not credit these procedures for IGSCC of the steam generator heat exchanger tubes. The decision to integrate IGA with IGSCC has been incorporated in the annual update of the RCS AMR report. To show this in the current BGE LRA, IGSCC should be listed as plausible (with a /) under device type HX S/G in Table 4.1-3. Additionally, the device type HX S/G tubes should be listed under the SCC /IGSCC heading of the Materials and Environment portion of Group 7 (SCC /lGSCC/PWSCC) in Section 4.1.2. Calvert Cliffs Procedure STP M-574-1/2 should be credited with the discovery of IGSCC of the steam generator heat exchanger tubes in Group 7 of Section 4.1.2. Table 4.1-4 of Section 4.1.2 should also credit STP-M-574-1/2 with discovery oflGSCC on the steam generator heat exchanger tubes for Group 7.

NRC Ouestion No. 4.1.14 Discuss how BGE will manage SCC of the CASS surge nozzle safe end.

BGE Response Baltimore Gas and Electric Company has determined that SCC of the surge nozzle safe end (CASS) is not plausible. The BGE AMR report for the RCS concludes that those RCS components that are fabricated of stainless steel and are not sensitizci (heat treated) are not susceptible to SCC and IGSCC. The surge nozzle safe end (CASS) is managed for the effects of thermal embrittlement as described in Group 8 of Section 4.1.2.

NRC Ouestion No. 4.1.15 What are the acceptance criteria in Procedure RV-78,"RV Flange Protection Ring Removal and Closure Head Installation?"

BGE Responic As currently stated in Calvert Cliffs Technical Procedure RV-78, the acceptance criteria for this procedure is to blow the reactor pressure vessel head flange leak-off line clear with compressed air for approximately 30 seconds. Blowing the line clear with compressed air removes potential contaminants that might contribute to corrosion of this line.

NRC Ouestion No. 4.1.16 Describe how denting and pitting of the steam generator tubes will be managed.

BGE Response Baltimore Gas and Electric Company manages denting and pitting of the steam generator heat l

exchanger tubes with Procedure STP-M-574-1/2. These procedures were credited with discovery of denting / pitting and were described in Section 4.1.2 of the BGE LRA under Group 1, Denting, and Group 5, Galvanic / General Corrosion and Pitting.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATIONi INTEGRATED PLANT ASSESSMENT REPORT FOR Tile REACTOR COOLANT SYSTEM NRC Ouestion No. 4.1.17 Please provide a summary description for the following procedures regarding how their implementation will address the following elements for their related aging management program (s): (a) The scope of structures and components managed by the program; (b) Preventive actions designed to mitigate or prevent aging degradation; (c) Parameters monitored or inspected relative to degradation of specific structure and component intended functions; (d) Detection of aging effects before loss of structure and component intended functions; (e) Monitoring, trending, inspection, testing frequency, and sample size to ensure, timely detection of aging etTects and corrective actions; (f) Acceptance criteria to ensure structure and component intended functions; and (g) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.

a.

Technical Procedure SG-20;

b. Administrative Procedure MN-3110; Technical Procedure FASTENER-01," Torquing and Fastener Applications;"

c.

d. Procedure STP-M-574-1/2; e.

CASS Evaluation program;

' f.

Alloy 600 Program; g.

STP-O-27-1/2;

h. Administrative MN-3-301,"BACI Program;" and

i. Administrative EN-1-300, " Implementation of Fatigue Monitoring."

BGE Response Baltimore Gas and Electric Company has requested clarification from NRC on this item and has agreed to work toward clarification through a future teleconference. Baltimore Gas and Electric Company will supplement this response, based on the outcome of that interaction.

NRC Ouestion No. 4.1.18 Clarify whether crevice corrosion for the RCS is a plausible aging effect and, if so, provide a reference to where aging management is addressed in the LRA. If crevice corrosion is not a plausible aging effect for the RCS, describe the basis for that conclusion.

BGE Response Baltimore Gas and Electric Company has determined that crevice corrosion is not a plausible ARDM for the RCS. Crevice corrosion is not plausible because stainless steel is not susceptible to the ARDM in a demineralized water environment.

NRC Ouestion No. 4.1.19 The application discusses prior degradation of the RCP suction deflector at Calvert Cliffs. What was the

- cause of the suction deflector bolting failures? What was the material of the bolts that failed, and how are the bolts being managed for aging?

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; i

INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR COOLANT SYSTEM l

BGE Response The full discussion of the RCP suction deflector bolt failure is located in the Operating Experience write-up of Section 4.2.1,

" Reactor Pressure Vessel and Control Element Drive l

Mechanisms / Electrical System." The discussion was included in Section 4.2.1 because of the potential impact of the failed suction deflector bolts on the reactor pressure vessel cladding. The RCP l

suction deflectors bolts were not within the scope for license renewal since they did not have a license l

renewal intended function (e.g., they do not contribute to pressure boundary function),

i NRC Question No. 4.1.20

[

Are there any parts of the systems, structures and components within the RCS that are inaccessible for i

inspection? If so, describe what aging management program will be relied upon to maintain the integrity i

of the inaccessible areas. If the aging management program for the inaccessible areas is an evaluation of l

the acceptability ofinaccessible areas based on conditions found in surrounding accessible areas, please provide information to show that conditions would exist in accessible areas that would indicate the i

presence of, or result in degradation to, such inaccessible areas. If different aging effects or aging management technique; are needed for the inaccessible areas, please provide a summary to address the following elements for the inaccessible areas: (a) Preventive actions that will mitigate or prevent aging L

degradatica; (b) Parameters monitored or inspected relative to degradation of specific structure and i

component intended functions; (c) Detection of aging efTects before loss of structure and component l

intended functions; (d) Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging effects and corrective actions; (e) Acceptance criteria to ensure structure and component intended functions; and (f) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.

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i BGE Response l

Baltimore Gas and Electric Company can access all RCS components if required.

i Reference 1.

Electric Power Research Institute Report TR-103844, "PWR Reactor Coolant System License Renewal Industry Report," Revision 1 i

l 2.

Letter from Mr. D. J. Walters (NEI) to Mr. C.1. Grimes, (NRC), dated May 1,1998, enclosing Electric Power Research Institute Report TR-106092, " Evaluation of Thermal Aging Embrittlement for Cast Austenitic Stainless Steel Components in LWR Reactor Coolant l

Systems" 3.

Letter from Mr.

C.

H. Cruse (BGE) to NRC Document Control Desk, dated November 19,1998, " Response to Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Report for the Reactor Vessel Internals System" l

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