ML102530517

Lr - Draft RAI Set 14 - AMR and TLAA

ML102530517
Person / Time
Site:	Diablo Canyon
Issue date:	07/01/2010
From:	Office of Nuclear Reactor Regulation
To:	Division of License Renewal
References
Download: ML102530517 (5)

Text

1 DiabloCanyonNPEm Resource From:

Ferrer, Nathaniel Sent:

Thursday, July 01, 2010 4:36 PM To:

Grebel, Terence; Soenen, Philippe R Cc:

Green, Kimberly; DiabloHearingFile Resource

Subject:

Draft RAI Set 14 - AMR and TLAA Attachments:

Draft RAI Set 14 AMR and TLAA.doc Terry and Philippe, Attached is Draft RAI Set 14 containing draft RAIs, specifically on portions of the aging management and TLAA review. Please review the attached draft RAIs and let me know if and when you would like to have a teleconference call. The purpose of the call will be to obtain clarification on the staff's request.

Please let me know if you have any questions.

NathanielFerrer ProjectManager DivisionofLicenseRenewal OfficeofNuclearReactorRegulation U.S.NuclearRegulatoryCommission (301)4151045

Hearing Identifier:

DiabloCanyon_LicenseRenewal_NonPublic Email Number:

1180 Mail Envelope Properties (Nathaniel.Ferrer@nrc.gov20100701163500)

Subject:

Draft RAI Set 14 - AMR and TLAA Sent Date:

7/1/2010 4:35:30 PM Received Date:

7/1/2010 4:35:00 PM From:

Ferrer, Nathaniel Created By:

Nathaniel.Ferrer@nrc.gov Recipients:

"Green, Kimberly" <Kimberly.Green@nrc.gov>

Tracking Status: None "DiabloHearingFile Resource" <DiabloHearingFile.Resource@nrc.gov>

Tracking Status: None "Grebel, Terence" <TLG1@PGE.COM>

Tracking Status: None "Soenen, Philippe R" <PNS3@PGE.COM>

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 573 7/1/2010 4:35:00 PM Draft RAI Set 14 AMR and TLAA.doc 47610 Options Priority:

Standard Return Notification:

No Reply Requested:

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Normal Expiration Date:

Recipients Received:

Diablo Canyon Nuclear Power Plant, Units 1 and 2 (DCPP)

License Renewal Application (LRA)

Draft Request for Additional Information Set 14 Aging Management Review/TLAA D-RAI 3.1.2.2.7.2-1 DCPP LRA Section 3.1.2.2.7.2 with LRA Table 3.3.1, item 3.1.1.24 addresses stainless steel Class 1 PWR cast austenitic stainless steel (CASS) piping and components exposed to reactor coolant. The LRA section states that for managing the aging of cracking due to stress corrosion cracking for the CASS components the Water Chemistry Program will be augmented by the ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD Program. LRA Section 3.1.2.2.7.2 also states that the susceptibility to thermal aging embrittlement will be evaluated in the Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program (B2.1.39). LRA Section B2.1.39 indicates that the applicants Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program is a new program that will be consistent with GALL AMP XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) with no exception or enhancement.

The staff noted that although LRA Section 3.1.2.2.7.2 addresses the material evaluation criteria used to manage the thermal embrittlement of CASS, the LRA section does not address the material screening criteria used to further evaluate and manage stress corrosion cracking of the CASS components.

In order to manage the stress corrosion cracking of the CASS components, the GALL Report, under item IV.C2-3, recommends further evaluation for CASS that has carbon content greater than 0.035% or ferrite content less than 7.5%.

1. Clarify how the applicants material screening criteria used to further evaluate and manage the stress corrosion cracking of CASS are consistent with GALL Report item IV.C2-3 which recommends that stress corrosion cracking of CASS with carbon content greater than 0.035% or ferrite content less than 7.5% be further evaluated and adequately managed.

2. Clarify whether the stress corrosion cracking in the CASS components under GALL Report item IV.C2-3 is managed by the inspections, flaw evaluations, and repairs and replacements in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD Program and the material screening criteria that the GALL Report recommends for the further evaluation. If the ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD Program or the material screening criteria recommend for the further evaluation is not used to manage the stress corrosion cracking, justify why the applicants aging management approach is adequate to manage the aging effect.

D-RAI 3.5.2.3.10-1 LRA Table 3.5.2-10 indicates that for aluminum components encased in concrete (external),

there are no aging effects requiring management. The AMR line item cites generic note J, indicating that neither the component nor the material and environment combination is evaluated in the GALL Report.

Corrosion of aluminum due to alkaline reaction could occur when it is used in contact with concrete. No Justification for why there are no aging effects requiring management for the line item referenced above is provided.

Provide justification for why there are no aging effects requiring management for the identified aluminum components exposed to a concrete environment.

D-RAI 4.6.2-1 In LRA Section 4.6.2, Design Cycles for Containment Penetrations, the applicant states:

1. The 14,000 additional thermal cycles used in the original analysis for the steam generator blowdown lines is greater than the maximum of 7000 cycles which are expected in 60 years. Therefore, the fatigue analysis for the main steam generator blowdown line flued heads is valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

2. The original number of transients used in the containment airlocks, hatches, penetration sleeves, end plates, and flued head analyses (not including the steam generator blowdown lines flued heads) will be monitored by the DCPP Metal Fatigue of the Reactor Coolant Pressure Boundary program, described in Sections 4.3.1 and B3.1, to ensure that fatigue will be adequately managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii). Action limits will permit completion of corrective actions before the design basis number of events is exceeded.

The staff reviewed LRA Sections 4.6.2, 4.3.1, and B3.1 and was unable to find the following information:

1. The total number of transients (thermal cycles, OBE events) used in the original analysis for the steam generator blowdown line flued heads.

2. Total number of transients used to determine that requirements of a fatigue waiver per Subparagraph N-415.1, Vessels Not Requiring for Cyclic Operation, and Figure N-415(A) were met for airlocks, equipment hatches, containment penetration sleeves, and end plates.

3. Total number of transients assumed in the current design basis for airlocks, equipment hatches, containment penetration sleeves, and end plates.

The staff needs the below information to confirm that an evaluation the fatigue analysis for the steam generator blowdown line flued heads is valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1):

1. The total number of cycles used for the original analysis for the main steam generator blowdown lines flued heads for 40 years of operation.

2. The projected number of cycles for the main steam generator blowdown line flued heads during 60 years of operation.

3. Total number of transients used to determine that requirements of a fatigue waiver per Subparagraph N-415.1, Vessels Not Requiring for Cyclic Operation, and Figure N-415(A) were met for airlocks, equipment hatches, containment penetration sleeves, and end plates.

4. Total number of transients assumed in the current design basis for airlocks, equipment hatches, containment penetration sleeves, and end plates.