ML103420101

Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application (Lra), and LRA Amendment No. 27

ML103420101
Person / Time
Site:	Palo Verde
Issue date:	11/23/2010
From:	Hesser J Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06285-JHH/GAM
Download: ML103420101 (10)

Text

AMIP A subsidiaryof Pinnacle West CapitalCorporation John H. Hesser Mail Station 7605 Palo Verde Nuclear Vice President Tel: 623-393-5553 PO Box 52034 Generating Station Nuclear Engineering Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-06285-JHH/GAM November 23, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529 and 50-530 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application (LRA), and LRA Amendment No. 27 In a conference call on October 22, 2010, as documented in a conference call summary dated November 16, 2010 (NRC Agencywide Document Access and Management System [ADAMS] Accession No. ML102990530), the Nuclear Regulatory Commission (NRC) staff discussed with Arizona Public Service Company (APS) a draft request for additional information (RAI) regarding the steam generators related to the PVNGS license renewal application (LRA). Enclosure 1 contains APS's responses to the draft RAI. Enclosure 2 contains LRA Amendment No. 27 for new Commitment Nos. 61 and 62 in LRA Table A4-1.

Should you need further information regarding this submittal, please contact Glenn Michael, Licensing Engineer for License Renewal, at (623) 393-5750.

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

Executed on 11/z3// 0 (date)

Sincerely, JHH/RAS/GAM/gat A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde

• San Onofre

• South Texas

• Wolf Cre ek

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application (LRA), and LRA Amendment No. 27 Page 2

Enclosures:

1. Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application

2. Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 27 cc: E. E. Collins Jr. NRC Region IV Regional Administrator J. R. Hall NRC NRR Senior Project Manager L. K. Gibson NRC NRR Project Manager J. H, Bashore NRC Senior Resident Inspector (acting) for PVNGS L. M. Regner NRC License Renewal Project Manager G. A. Pick NRC Region IV (electronic)

ENCLOSURE 1 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application

Enclosure I Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application NRC Draft RAI Regarding Steam Generators

1. The Updated Final Safety Analysis Report (UFSAR), Section 1.2.3.3, states that a vertical divider plate separates the inlet and outlet plenums in the lower head of the steam generators (SGs), but no information about the materials of the divider plate assembly, nor its junction to the lower head and to the tubesheet was found in the UFSAR nor the LRA. Clarify how the divider plate is assembled to the lower head and to the tubesheet and what are the materials of this divider plate and potential associated welds. Ifthis information is in the UFSAR, please specify its location.

(The NRC Staff provided the following request subsequent to the draft RAI above.)

If the compositions of the SG divider plate divider bar welds (all areas) are susceptible to primary water stress corrosion cracking (PWSCC), thereby potentially compromising SG reactor coolant system pressure boundary, provide an appropriate aging management program, including commitment, to manage for the possible effects of PWSCC on these welds. An inspection method must be capable of detecting PWSCC.

2. Regarding LRA item 3.1.1-35, the applicant states that it's not applicable because they have recirculating SGs and the GALL Report item is for once-through SGs.

However, this aging effect is applicable also for recirculating SGs tube-to-tubesheet welds with tubesheet cladding of Alloy 600 and SG tubes of Alloy 690TT (which is the case for Palo Verde). In such a configuration, the weld may have insufficient chromium content to prevent PWSCC. If the tubesheet cladding is Alloy 600, describe how PWSCC is managed for aging in the tube-to-tubesheet welds.

3. LRA Table 3.1.2-4 does not include an item addressing GALL item IV.D1-17, which corresponds to ligament cracking due to corrosion in steel tube support plates.

However, in LRA Table 3.1.2-4, the applicant addresses GALL item IV.D1-19, which corresponds to SG tube denting due to corrosion of carbon steel tube support plate.

In LRA Section B2.1.8, the applicant also states that the tube support system is similar to the original design, and like the original design is fabricated from 409 ferritic stainless steel. Explain this apparent inconsistency.

APS Response to Draft RAI Regarding Steam Generators Response (1)

The steam generator (SG) primary side divider plates are attached to the channel head, stay cylinder, and tubesheet via a tongue-in-groove connection. The divider plate edges are grooved to slide onto the tongue portion of divider plate bars that are welded to the channel head, stay cylinder, and tubesheet cladding. Set screws at the divider plate groove to divider plate bar tongue connection are utilized to minimize movement of the floating divider plate. Small patch plates are installed at the two upper corners on 1

Enclosure 1 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application both sides of each divider plate via a bolted connection. All screws and bolts are welded over. All components (two divider plates, divider plate bars, patch plates, bolts, and screws) are manufactured from Alloy 690 material. The SG specifications show the divider plate bars welded to the channel head, stay cylinder, and tubesheet cladding using Alloy 52, 82, 152, and 182 filler materials. This level of detailed information is not included in the UFSAR.

There is no routine inspection requirement for the divider bar welds because of the following:

• These welds do not provide a reactor coolant system pressure boundary.

• These welds do not provide structural support to the SGs.

• The divider plate "floats" in the tongue and groove, and the force on the divider plate that would get transferred to the divider plate bar welds would be the relatively low (compared with RCS pressure) differential pressure between the SG inlet (hot leg) and outlet (cold leg).

• A crack in the divider bar weld due to PWSCC would need to propagate from the divider bar weld through the channel head cladding to get to the base metal.

In response to the NRC staff concern regarding potential failure of the SG reactor coolant system pressure boundary due to possible PWSCC of SG divider plate bar welds, APS commits to perform one of the following three resolution options:

1. Perform an inspection of each Palo Verde Unit 1, 2, and 3 steam generator to assess the condition of the divider plate bar welds. The examination technique(s) will be capable of detecting PWSCC in the divider plate bar welds.

OR

2. Perform an analytical evaluation of the steam generator divider plate bar welds in order to establish a technical basis which concludes that the SG reactor coolant system pressure boundary is adequately maintained with the presence of steam generator divider plate bar weld cracking.

OR

3. If results of industry and NRC studies and operating experience document that potential failure of the SG reactor coolant system pressure boundary due to PWSCC cracking of SG divider plate bar welds is not a credible concern, this commitment will be revised to reflect that conclusion.

The original SGs were replaced in Units 1, 2, and 3, during the fall of 2005, 2003, and 2007, respectively. If Option (1) is selected, it will be completed for each SG in each unit during an SG tube eddy-current inspection outage between 20 and 25 calendar years of SG operation. For Units 1 and 3, this would approximately correspond to the 2

Enclosure 1 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application first five years after entering the period of extended operation (PEO). For Unit 2, this would approximately correspond to between three years prior to and two years after entering the PEO.

Unit 1, Option (1): Perform on each SG during an SG tube eddy-current inspection outage between 9/1/25 and 12/1/30. (PEO would begin 6/1/25.)

Unit 2, Option (1): Perform on each SG during an SG tube eddy-current inspection outage between 9/1/23 and 12/1/28. (PEO would begin 4/24/26.)

Unit 3, Option (1): Perform on each SG during an SG tube eddy-current inspection outage between 9/1/27 and 12/1/32. (PEO would begin 11/25/27.)

If Option (2) or Option (3) is selected, it will be completed prior to September 1, 2023.

Response (2)

The tubes in the PVNGS SGs are manufactured from Alloy 690TT, and the tubesheet cladding is composed of Alloy 82. A review of industry operating experience has found no instances of Alloy 82 primary water stress corrosion cracking (PWSCC) in the industry when it is not used with Alloy 182. The tube-to-tubesheet weld is an autogenous weld, which is created by melting the corner of the tubesheet clad to the tube end without adding filler metal. The weld is thus a combination of both materials.

Section 2.2 of EPRI Technical Report 1006696, Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), states that Alloy 82, which has a chromium content of between 18-20%, took 3 to 4 times longer to crack than Alloy 182, which has a chromium content of approximately 14.5%. For comparison purposes, the chromium content of Alloy 600 is between 14-17%. The EPRI document also states that weld metals with 30% chromium were resistant to cracking, with a threshold for PWSCC resistance being between 22 and 30% chromium. Since the chromium content of Alloy 690TT is 30%, the tube-to-tubesheet weld in the PVNGS SGs would be expected to have a chromium content somewhere between 20% and 30%.

Alloy 82 weld metal, with its higher chromium content, has proven to be less susceptible to PWSCC and to have lower crack growth rates than Alloy 182 weld metal. Since the early 1990s, Palo Verde has used Alloy 82 for repairs on several Alloy 600 components (high temperature locations). Those repairs were on instrumentation nozzles on the hot leg and pressurizer, and pressurizer heater sleeves. The repairs consisted of half nozzle replacements using Alloy 690 nozzles welded with Alloy 82. The nozzles and welds are inspected visually every refueling outage and have had no leakage.

3

Enclosure 1 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application In response to the NRC staff concern regarding potential failure of the steam generator primary-to-secondary pressure boundary due to PWSCC cracking of tube-to-tubesheet welds, APS commits to perform one of the following two resolution options:

1. Perform a one-time inspection of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. If weld cracking is identified:

a. The condition will be resolved through repair or engineering evaluation to

. justify continued service, as appropriate.

AND

b. An ongoing monitoring program will be established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

OR

2. Perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to:

a. Establish a technical basis which concludes that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintained with the presence of tube-to-tubesheet weld cracking.

AND

b. Establish a technical basis which concludes that the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function.

If Option (1) is elected, it will be completed on the same schedule as Option (1) described in Response (1) above. If Option (2) is selected, it will be on the same schedule as Option (2) described in Response (1) above.

Response (3)

GALL line IV.D1-19 for steam generator tube denting due to corrosion of carbon steel support plates was applied in error and was deleted in LRA Amendment 26 submitted with APS letter no. 102-06279, dated November 10,/2010. The Palo Verde steam generator tube support system is fabricated with 409 ferritic stainless steel; therefore, the GALL line IV.D1-19 for denting due to corrosion of carbon steel support plates is not applicable.

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ENCLOSURE 2 Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 27 I

LRA Section Page No.

Table A4-1, Item 61 A-60 Table A4-1, Item 62 A-60

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 27 LRA Table A4-1, License Renewal Commitment No. 61:

Item Commim LRA Section Implementation NO.' Schedul el 61 In response to the NRC staff concern regarding potential failure of the SG Response to Draft RAI If Option (1) is selected, reactor coolant system pressure boundary due to possible PWSCC of SG divider Regarding Steam it will be completed for plate bar welds, APS commits to perform one of the following three resolution Generators in APS letter each SG in each unit options: No. 102-06285, dated during an SG tube

1. Perform an inspection of each Palo Verde Unit 1, 2, and 3 steam November 23, 2010. eddy-current inspection generator to assess the condition of the divider plate bar welds. The outage between 20 and examination technique(s) will be capable of detecting PWSCC in the 25 calendar years of SG divider plate bar welds. operation.

OR

2. Perform an analytical evaluation of the steam generator divider plate bar If Option (2) or Option welds in order to establish a technical basis which concludes that the SG (3) is selected, it will be reactor coolant system pressure boundary is adequately maintained with completed prior to the presence of steam generator divider plate bar weld cracking. September 1, 2023.

OR

3. If results of industry and NRC studies and operating experience document that potential failure of the SG reactor coolant system pressure boundary due to PWSCC cracking of SG divider plate bar welds is not a credible concern, this commitment will be revised to reflect that conclusion.

(RCTSAIs 3561775 [Ul], 3561777 [U2], 3561779 [U3])

Enclosure 1 Response to Draft Request for Additional Information for the Review of the PVNGS License Renewal Application LRA Table A4-1, License Renewal Commitment No. 62:

Iteml . Commitment , LRA Section Implementation 62 In response to the NRC staff concern regarding potential failure of the steam Response to Draft RAI If Option (1) is selected, generator primary-to-secondary pressure boundary due to PWSCC cracking of Regarding Steam it will be completed for tube-to-tubesheet welds, APS commits to perform one of the following two Generators in APS letter each SG in each unit resolution options: No. 102-06285, dated during an SG tube

1. Perform a one-time inspection of a representative number of tube-to- November 23, 2010. eddy-current inspection tubesheet welds in each steam generator to determine if PWSCC outage between 20 and cracking is present. If weld cracking is identified: 25 calendar years of SG

a. The condition will be resolved through repair or engineering operation.

evaluation to justify continued service, as appropriate.

AND If Option (2) is selected,

b. An ongoing monitoring program will be established to perform it will be completed prior routine tube-to-tubesheet weld inspections for the remaining life of to September 1, 2023.

the steam generators.

OR

2. Perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to:

a. Establish a technical basis which concludes that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintained with the presence of tube-to-tubesheet weld cracking.

AND

b. Establish a technical basis which concludes that the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function.

(RCTSAIs 3561782 [U1-, 3561784 [U2], 3561788 [U3])