DCL-10-107, Response to NRC Letter Dated July 22, 2010 Request for Additional Information Re License Renewal Application

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Response to NRC Letter Dated July 22, 2010 Request for Additional Information Re License Renewal Application
ML102310035
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 08/18/2010
From: Becker J
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-10-107
Download: ML102310035 (7)
v • d • e


Text

Pacific Gas and ElectricCompany James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P 0. Box 56 August 18, 2010 Avila Beach, CA 93424 805.545.3462 PG&E Letter DCL-10-107 Internal: 691.3462 Fax: 805.545.6445 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Letter dated July 22, 2010, Request for Additional Information (Set 14) for the Diablo Canyon License Renewal Application

Dear Commissioners and Staff:

By letter dated November 23, 2009, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA), and Applicant's Environmental Report -

Operating License Renewal Stage.

By letter dated July 22, 2010, the NRC staff requested additional information needed to continue their review of the DCPP LRA.

PG&E's response to the request for additional information is included in Enclosure 1.

PG&E makes no regulatory commitments in this letter.

If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4160.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on August Sincerýlv, -

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon e Palo Verde
  • San Onofre
  • Wolf Creek

Document Control Desk PG&E Letter DCL-10-107 August 18, 2010 Page 2 pns/50329699 Enclosure cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, Office of Nuclear Reactor Regulation A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde e San Onofre o South Texas Project
  • Wolf Creek

Enclosure 1 PG&E Letter DCL-10-107 Sheet 1 of 5 PG&E Response to NRC Letter dated July 22, 2010 Request for Additional Information (Set 14) for the Diablo Canyon License Renewal Application RAI 3.1.2.2.7.2-1 Diablo Canyon Power Plant (DCPP)license renewal application (LRA) Section 3.1.2.2.7.2 with LRA Table 3.3.1, item 3.1.1.24 addresses stainless steel Class I pressurized water reactor(PWR) cast austenitic stainless steel (CASS) piping and components exposed to reactorcoolant. 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 American Society of Mechanical Engineers (ASME) Section X1 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 applicant's Thermal Aging Embrittlement of CASS Programis a new program that will be consistent with Generic Aging Lessons Learned (GALL) aging management program (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 materialevaluation criteria used to manage the thermal embrittlement of CASS, the LRA section does not address the materialscreening 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 greaterthan 0. 035% or ferrite content less than 7.5%.

1. Clarify the materialscreening 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 greaterthan 0.035% or ferrite content less than 7.5% be further 6evaluated and adequatelymanaged.
2. Clarify whether stress corrosion cracking in the CASS components under GALL Report item IV. C2-3 is managed by inspections, flaw evaluations, and repairsand replacements in accordance with the ASME Section X1 Inservice Inspection, Subsections IWB, IWC and IWD Programand the materialscreening criteriathat 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 applicant'saging management approach is adequate to manage the aging effect.

Enclosure 1 PG&E Letter DCL-1 0-107 Sheet 2 of 5 PG&E Response to RAI 3.1.2.2.7.2-1

1. According to Generic Aging Lessons Learned Report, Item IV.C2-3, the recommended Aging Management Program (AMP) for aging effect of stress corrosion cracking (SCC) is the Water Chemistry Program if the material selected meets the NUREG-0313 guideline of carbon content less than 0.035 percent and ferrite content greater than 7.5 percent. Certified material test reference (CMTR) of cast austentic stainless steel (CASS) reactor coolant loops components were reviewed and found to have carbon content greater than 0.035 percent or ferrite content less than 7.5 percent. As described in License Renewal Application (LRA) Section 3.1.2.2.7.2, the Water Chemistry AMP (LRA Section B2.1.2) will be augmented by ASME Section XI, Inservice Inspection, Subsections IWB, IWC, and IWD (LRA Section B2.1.1) to ensure adequate inspection methods for detection of cracks. CASS components that are determined to be susceptible to thermal aging embrittlement, will be managed through either enhanced volumetric examinations or component-specific flaw tolerance evaluations as described in the DCPP Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program (LRA Section B2.1.39).
2. As described in LRA Section 3.1.2.2.7.2, the Water Chemistry Program (LRA Section B2.1.2), will be augmented by ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (LRA Section B2.1.1) to ensure that adequate inspection methods ensure detection of cracks.

Enclosure 1 PG&E Letter DCL-10-107 Sheet 3 of 5 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 aging management review (AMR) line item cites generic Note J, indicatingthat neitherthe component nor the materialand 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 ilne item referenced above is provided.

Providejustification for why there are no aging effects requiringmanagement for the identified aluminum components exposed to a concrete environment.

PG&E-Response to RAI 3.5.2.3.10-1 At the intake structure, the concrete hatch covers and hatch openings are constructed with aluminum angles forming the corners and edges. The function of these angles is to prevent damage to the edges of the concrete during maintenance activities in order to maintain the structural configuration of the hatches and openings.

In License Renewal Application Table 3.5.2-10, the aluminum component encased in concrete represents the embedded surface of these members, and the aluminum component in atmosphere/weather represents the exposed portions of these same angles. The Diablo Canyon Power Plant Structures Monitoring Program will manage the aging of these components by visually examining the accessible surfaces for loss of material.

Enclosure 1 PG&E Letter DCL-10-107 Sheet 4 of 5 RAI 4.6.2-1 In LRA Section 4.6.2, "Design Cycles for Containment Penetrations,"the applicant states:

1. The 14,000 additionalthermal cycles used in the original analysis for the steam generatorblowdown lines is greaterthan the maximum of 7000 cycles which are expected in 60 years. Therefore, the fatigue analysis for the main steam generatorblowdown, line flued heads is valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
2. The originalnumber of transientsused in the containment airlocks, hatches, penetration,sleeves, end plates, and flued head analyses (not including the steam generatorblowdown lines flued heads) will be monitored by the DCPP Metal Fatigue of the Reactor Coolant PressureBoundary Program, described in LRA Sections 4.3.1 and B3. 1, to ensure that fatigue will be adequatelymanaged for the period of extended operationin 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 (thermalcycles, OBE events) used in the original analysis for the steam generatorblowdown line flued heads.
2. Total number of transients used to determine that requirements of a fatigue waiver per SubparagraphN-415.1, Vessels Not Requiring for Cyclic Operation, and Figure N-415(A) were met for airlocks, equipment hatches, containment penetrationsleeves, and end plates.
3. Total number of transients assumed in the currentdesign basis for airlocks, equipment hatches, containmentpenetrationsleeves, and end plates.

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

1. The total number of cycles used for the originalanalysis for the main steam generatorblowdown lines flued heads for 40 years of operation.
2. The projected number of cycles for the main steam generatorblowdown line flued heads during 60 years of operation.

Enclosure 1 PG&E Letter DCL-1 0-1 07 Sheet 5 of 5

3. Total number of transients used to determine that requirementsof a fatigue waiver per SubparagraphN-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, containmentpenetrationsleeves, and end plates.

PG&E Response to RAI 4.6.2-1

1. The total number of cycles used in the original fatigue analysis for the steam generator (SG) blowdown line flued heads is 15,000 analyzed cycles for 40 years of operation.
2. As shown in License Renewal Application (LRA) Table 4.3-2, the Unit 1 main SG blowdown line flued heads expect to experience no more than 85 heatups, 87 cooldowns, and 1 seismic event (at 20 cycles per event) in 60 years of operation.

As shown in LRA Table 4.3-2, the Unit 2 main SG blowdown line flued heads expect to experience no more than 65 heatups, 63 cooldowns, and 1 seismic event at (20 cycles per event) in 60 years of operation.

3. The total number of transients used to determine that requirements of a fatigue waiver per subparagraph N-415.1 were met for airlocks, equipment hatches, containment penetration sleeves, and end plates is 500 cycles (250 heatups and 250 cooldowns). This is consistent with the Diablo Canyon Power Plant Final Safety Analysis Report Update.
4. As stated in LRA Section 4.6.2, the current design basis for the airlocks, equipment hatches, containment penetration sleeves, and end plates does not incorporate a limiting number of transients.
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