ML102850605

Lr - Draft RAI Set 17 - TLAA and AMP RAIs

ML102850605
Person / Time
Site:	Diablo Canyon
Issue date:	07/19/2010
From:	Office of Nuclear Reactor Regulation
To:	Division of License Renewal
References
Download: ML102850605 (6)

Text

DiabloCanyonNPEm Resource From: Ferrer, Nathaniel Sent: Monday, July 19, 2010 3:33 PM To: Grebel, Terence; Soenen, Philippe R Cc: DiabloHearingFile Resource

Subject:

Draft RAI Set 17 - TLAA and AMP RAIs Attachments: Draft RAI Set 17 TLAA and AMP RAIs.doc Terry and Philippe, Attached is Draft RAI Set 17 containing draft RAIs, specifically on portions of the TLAA and aging management programs 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.

Nathaniel Ferrer Project Manager Division of License Renewal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission (301)4151045 1

Hearing Identifier: DiabloCanyon_LicenseRenewal_NonPublic Email Number: 1944 Mail Envelope Properties (26E42474DB238C408C94990815A02F090A76E5A915)

Subject:

Draft RAI Set 17 - TLAA and AMP RAIs Sent Date: 7/19/2010 3:32:48 PM Received Date: 7/19/2010 3:32:49 PM From: Ferrer, Nathaniel Created By: Nathaniel.Ferrer@nrc.gov Recipients:

"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: HQCLSTR01.nrc.gov Files Size Date & Time MESSAGE 582 7/19/2010 3:32:49 PM Draft RAI Set 17 TLAA and AMP RAIs.doc 58362 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: 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 17 TLAA/Aging Management Programs D-RAI 4.7.2-1 In LRA section 4.7.2, within the Pressurizer section, the applicant states that the fatigue crack growth analyses were projected to the end of the period of extended operation and are therefore valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

1. Discuss how the actual plant transient cycles are monitored to ensure that they are bounded by the number assumed in the fatigue crack growth analysis.

2. Discuss the transient cycles used in the crack growth analyses, including the number of cycles.

D-RAI 4.7.2-2 In LRA section 4.7.2, within the Pressurizer section, the applicant states that [n]o base-metal corrosion analyses exist for the pressurizers, since no half-nozzle or similar repairs have exposed the base metal to reactor coolant. The applicant also states that [t]he Unit 1 pressurizer and its nozzles and safe ends contain no Alloy 600 or Alloy 82/182 weld material.

The above statements are not clear regarding whether the half nozzle method was used in repairing heater sleeves in the pressurizer in both units.

1. For each unit, list all the pressurizer nozzles (e.g., pressurizer safety valve nozzle and heater sleeve nozzle). Identify the materials used to fabricate the nozzles. If a nozzle is welded to a safe end, identify the material of the safe end. If a nozzle is repaired or replaced, identify the repair method. Please provide the information in a table.

2. Discuss in detail the repair method. Discuss the material used in the repaired or replaced nozzle. Discuss whether fatigue crack growth calculation was performed for the remnant Alloy 82/182 welds. If so, discuss how the transient cycles used in the fatigue crack growth calculation are monitored to ensure they bound the actual plant cycles. If no fatigue crack growth calculation was performed, justify the structural integrity of the pressurizer shell.

3. Discuss any flaws that remained in service in the heater sleeves and in the attachment welds in both units. If so, discuss how these flaws are monitored and evaluated for the period of extended operation.

D-RAI 4.7.2-3 Discuss whether reactor vessel internals contain any nickel-based Alloy 600 components or nickel-based Alloy 82/182 welds. If so, discuss how these components are monitored for primary water stress corrosion cracking.

D-RAI 4.7.2-4 In LRA section 4.7.2, within the Steam Generators section, the applicant states that

[r]eplacement steam generators contain no Alloy 600 components or Alloy 82/182 welds.

1. Identify the material specification of the welds that join the replacement steam generator nozzles to the piping.

2. Identify the material specification of the safe ends that are welded to the steam generator nozzles.

D-RAI 4.7.2-5 In LRA section 4.7.2, within the Alloy 600 Program and Other Locations section, the applicant states that DCPP procedural guidance provides a comprehensive Alloy 600 control program for materials in the RCS.

Describe the DCPP procedure guidance in terms of the following items:

1. If and how it relates to the Nickel-Alloy Aging Management Program

2. Scope of the program (identify the components).

3. Describe how the components are monitored to ensure their structural integrity at the end of 60 years.

4. If degradation is detected, describe the corrective actions.

D-RAI 4.7.2-6 In LRA section 4.7.2, within the Alloy 600 Program and Other Locations section, the applicant states that other than the Unit 2 pressurizer none of the Alloy 600 locations have yet been subject to repairs.

1. Identify flaws, indications, or defects that have been detected in the components under the Alloy 600 Program.

2. Discuss the projection of the flaws to the end of 60 years and discuss the TLAA of the flaw evaluations.

D-RAI 4.7.5-1 In LRA section 4.7.5, within the Unit 2 RHR Piping Weld RB-119-11 section, the applicant states that [t]he service life for Weld RB-119-11 is based on operating for 40 years from the date the flaw was identified, i.e. until 2046, during which the flaw would experience 500 startup-shutdown cycles. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2. The above statements do not provide a clear reasoning as to how the flaw evaluation for 40 years encompasses 60 years of plant life. Clarify how the flaw evaluation encompassed a 60 year plant life in terms of cycle counting (e.g., are the 500 startup and shutdown cycles bound the actual plant cycles at the end of 60 years?).

D-RAI 4.7.5-2 In LRA section 4.7.5, within the Unit 2 RHR Piping Weld RB-119-11 section, the applicant states that [t]he DCPP licensing basis assumes 250 heatups and 250 cooldowns for a 50 year plant life.

1. Discuss why 50 years, rather than the conventional 40 years are used for the design basis.

2. Discuss why only heatup and shutdown cycles are discussed but not other transient cycles such as seismic, temperature, and pressure for the flaw evaluation.

3. Discuss how you ensure that transient cycles used in the flaw evaluation for the Unit 2 RHR piping weld RB-119-11 do not exceed the actual operating cycles.

D-RAI 4.7.5-3 In LRA section 4.7.5, within the Validation - Flaw Evaluation of Unit 1 RHR Weld WIC-95 section, the applicant states that [t]here have been no occurrences of a DE, DDE, or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events is sufficient to the end of the period of extended operation. Absence of earthquakes in the past 20 years does not imply that the earthquakes will not occur in the future and may not be appropriate to conclude that the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis is sufficient to bound the seismic cycles to the end of the period of extended operation.

1. Provide detailed technical basis to demonstrate why seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation bound the end of the period of extended operation.

2. Explain why other transient cycles (such as thermal) are not considered and assessed in this section.

3. Discuss why DE, DDE, or Hosgri seismic cycles were discussed for the flaw evaluation of Unit 1 RHR Weld WIC-95, but they were not discussed for the flaw evaluations of Unit 2 RHR Weld RB-119-11 and Unit 2 auxiliary feedwater line 567.

D-RAI 4.7.5-4 Discuss how you ensure that transient cycles used in the flaw evaluation for the RHR piping weld WIC-95 do not exceed the actual operating cycles.

D-RAI 4.7.5-5 Discuss whether 3 successive examinations of the flaws in Unit 2 RHR piping weld RB-119-11, Unit 2 auxiliary feedwater piping line 567, and Unit 1 RHR piping weld WIC-95 have been conducted per IWA-2000 of the ASME Code,Section XI. If yes, provide inspection result of the 3 successive examinations for each of the flaws. If no, justify why successive examinations were not performed.

D-RAI B2.1.39-1 Identify the scope of the Thermal Aging Embrittlement of CASS program by listing all components in the piping systems, including valves and pumps that are fabricated with CASS.

D-RAI B2.1.39-2 In LRA section B2.1.39, the applicant stated that the CASS aging management of potentially susceptible components is accomplished through an enhanced volumetric examination or a component-specific flaw tolerance evaluation. (1) Specify which of the above two methods will be used to manage CASS during the period of extended operation. (2) If the flaw tolerance evaluation is used, describe the details of the flaw tolerance evaluation.

D-RAI B2.1.39-3 In LRA section B2.1.39, the applicant stated that the Thermal Aging Embrittlement of CASS program will be implemented as part of the ASME Code,Section XI ISI program and will be completed within the 10-year inspection interval before the period of extended operation.

1. Describe in detail the CASS aging management program in terms of examination requirements (e.g., nondestructive examination methods, inspection frequency, and examined components).

2. Discuss exactly how the ASME Section XI ISI program is augmented and enhanced as a result of implementing the CASS AMP (e.g., discuss any changes to the ASME ISI program as a result of the CASS AMP in terms of inspection frequency and inspection methods).

3. The NRC staff notes that ultrasonic testing is not demonstrated via the ASME Code,Section XI, Appendix VIII, and therefore, not acceptable, to examine CASS components.

In light of this limitation, discuss how volumetric examination of CASS component will be accomplished.

D-RAI B2.1.39-4 (1) Discuss whether Diablo Canyon units 1 and 2 have implemented the risk-informed ISI program. (2) If yes, discuss how the CASS components will be inspected under the risk-informed ISI program considering the requirements of the CASS aging management program (e.g., whether the CASS AMP will increase the inspection frequency of the CASS components in the risk-informed ISI program and whether thermal aging embrittlement will be a degradation mechanism considered in the risk-informed ISI program).