ML101460077

Request for Additional Information for Salem Nuclear Generating Station Units 1 and 2 License Renewal Application (TAC ME1832 and ME1836)

ML101460077
Person / Time
Site:	Salem
Issue date:	06/10/2010
From:	Bennett Brady License Renewal Projects Branch 1
To:	Joyce T Public Service Enterprise Group
CuadradoDeJesus S, NRR/DLR, 415-2946
References
TAC ME1834, TAC ME1836
Download: ML101460077 (8)

Text

June 10, 2010 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear LLC P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR SALEM NUCLEAR GENERATING STATION UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (TAC NOS. ME1834 AND ME1836)

Dear Mr. Joyce:

By letter dated August 18, 2009, as supplemented by letter dated January 23, 2009, Public Service Enterprise Group Nuclear, LLC, submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 for renewal of Operating License Nos. DPR-70 and DPR-75 for Salem Nuclear Generating Station Units 1 and 2, respectively. The staff of the U.S.

Nuclear Regulatory Commission (NRC or the staff) is reviewing this application in accordance with the guidance in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants. During its review, the staff has identified areas where additional information is needed to complete the review. The staffs request for additional information is included in the Enclosure. Further requests for additional information may be issued in the future.

Items in the enclosure were provided to John Hufnagel and other members of your staff, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me by telephone at 301-415-2981 or by e-mail at bennett.brady@nrc.gov.

Sincerely,

/RA/

Bennett M. Brady, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosure:

As stated cc w/encl: See next page

ML101460077 OFFICE PM:DLR:RPB1 LA:DLR BC:DLR:RPB1 PM:DLR:RPB1 NAME B. Brady S. Figueroa B. Pham B. Brady DATE 6/9/10 6/1/10 6/9/10 6/10/10

Letter to T. Joyce from B. Brady dated June 10, 2010

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR SALEM NUCLEAR GENERATING STATION UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (TAC NOS. ME1834 AND ME1836)

DISTRIBUTION:

HARD COPY:

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Salem Nuclear Generating Station, Units 1 and 2 cc:

Mr. Robert Braun Mr. Michael Gaffney Senior Vice President Nuclear Manager - Hope Creek Regulatory PSEG Nuclear LLC Assurance One Alloway Creek Neck Road PSEG Nuclear LLC Hancocks Bridge, NJ 08038 One Alloway Creek Neck Road Hancocks Bridge, NJ 08038 Mr. Brian Booth Director Nuclear Oversight Mr. Ali Fakhar PSEG Nuclear Manager, License Renewal P.O. Box 236 PSEG Nuclear LLC Hancocks Bridge, NJ 08038 One Alloway Creek Neck Road Hancocks Bridge, NJ 08038 Mr. John Perry Station Vice President - Hope Creek Township Clerk PSEG Nuclear Lower Alloways Creek Township P.O. Box 236 Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. Larry Wagner Mr. Paul Bauldauf, P.E., Asst. Director Plant Manager - Hope Creek Radiation Protection Programs PSEG Nuclear LLC NJ Department of Environmental One Alloway Creek Neck Road Protection and Energy, CN 415 Hancocks Bridge, NJ 08038 Trenton, NJ 08625-0415 Mr. Michael Gallagher Mr. Brian Beam Vice President - License Renewal Projects Board of Public Utilities Exelon Nuclear LLC 2 Gateway Center, Tenth Floor 200 Exelon Way Newark, NJ 07102 Kennett Square, PA 19348 Regional Administrator, Region I Mr. Jeffrie J. Keenan, Esquire U.S. Nuclear Regulatory Commission Manager - Licensing 475 Allendale Road PSEG Nuclear LLC King of Prussia, PA 19406 One Alloway Creek Neck Road Hancocks Bridge, NJ 08038 Mr. Paul Davison Vice President, Operations Support Mr. Gregory Sosson PSEG Nuclear LLC Director Corporate Engineering One Alloway Creek Neck Road PSEG Nuclear LLC Hancocks Bridge, NJ 08038 One Alloway Creek Neck Road Hancocks Bridge, NJ 08038

Salem Nuclear Generating Station, Units 1 and 2 cc:

Ms. Christine Neely Director - Regulator Affairs PSEG Nuclear LLC One Alloway Creek Neck Road Hancocks Bridge, NJ 08038 Senior Resident Inspector Hope Creek Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038 Mr. Earl R. Gage Salem County Administrator Administration Building 94 Market Street Salem, NJ 08079

REQUEST FOR ADDITIONAL INFORMATION FOR SALEM NUCLEAR GENERATING STATION (SNGS) UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (LRA)

(TAC NOS. ME1834 AND ME1836)

RAI B.2.1.7-01

Background:

Based on a review of GALL Report Table 1, line item 80, for cast austenitic stainless steel (CASS) reactor vessel internal components, the GALL recommends use of the GALL AMP XI.M13, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program to manage the aging effect of loss of fracture toughness/thermal aging embrittlement. The GALL Report does not have an aging effect of cracking, stress-corrosion cracking (SCC), irradiation-assisted stress-corrosion cracking (IASCC), void swelling, or changes in dimension line item with CASS as the material.

Issue:

Several line items in LRA Table 3.1.2-3, assigned generic note C, have an aging effect of cracking, SCC, IASCC, changes in dimension and void swelling with CASS material components. These line items credit the LRA pressurized-water reactor (PWR) Vessel Internals AMP. These line items reference GALL Report NUREG-1801 Volume 2 line items which list stainless steel instead of CASS as the material. Additionally, one LRA Table 3.1.2-3 line item associated with a loss of fracture toughness/neutron irradiation embrittlement, and void swelling credits the PWR Vessel Internals Program and assigns this line item generic note E. Although there is no specific GALL line item for an aging effect of cracking, SCC, IASCC, changes in dimension and void swelling with CASS material components, the staff questions why the LRA does not credit GALL AMP XI.M13, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program to manage these aging effects.

Request:

Provide a justification for using the PWR Vessels Internals AMP instead of GALL AMP XI.M13, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS).

RAI B.2.1.23-01

Background:

GALL AMP XI.M35, Element 4 detection of aging effects, states that inspections of ASME Code Class 1 small-bore piping should be volumetric.

Issue:

The LRA Sections B.2.1.23 and A.2.1.23 state socket welds that fall within the weld examination sample will be examined using visual examination (VT-2). The staff notes that the proposed inspection methodology is not consistent with GALL.

ENCLOSURE

Request:

Please provide technical basis how VT-2 can adequately manage aging in ASME Code Class 1 socket welds.

RAI B.2.2.2-03

Background:

Section A1.2.3 of NUREG-1800, Appendix A states that the acceptance criteria of the program and its basis should be described.

Issue:

In element 6 of the LRA AMP it states that acceptance criteria for loss of material are based on the original equipment design wall thickness and any corrosion allowance requirements. It is not clear to the staff what the acceptance criteria for determining effects of aging on aluminum components are.

Request:

Clarify what the acceptance criteria for determining effects of aging on aluminum components are.

RAI 3.1.1-01

Background:

Standard Review Plan-License Renewal (SRP-LR) Section 3.1.2.2.13 identifies that cracking due to primary water stress-corrosion cracking (PWSCC) could occur in PWR components made of nickel alloy and steel with nickel alloy cladding, including reactor coolant pressure boundary components and penetrations inside the reactor coolant system (RCS) such as pressurizer heater sheathes and sleeves, nozzles, and other internal components. GALL Report Volume 2 Item IV.D1-06 recommends Chapter XI.M2, Water Chemistry, for PWR primary water for managing the aging effect of cracking in the nickel alloy steam generator (SG) divider plate exposed to reactor coolant.

In LRA Table 3.1.1, Item 81, the applicant credits its Water Chemistry Program to manage cracking due to stress corrosion cracking in nickel alloy steam generator (SG) primary channel head divider plate exposed to reactor coolant in the steam generators.

In UFSAR Section 5.5.2.2.1, the applicant describes that the divider plate is made with Inconel 690 for Unit 2 replacement steam generators. The staff notes that the use of this nickel alloy 690 should prevent the aging effect of PWSCC.

However, there is no information about the construction material of the divider plate for Unit 1 steam generators.

Issue:

From recent operating experience in steam generators with a similar design to that of Salem Unit 1 steam generators, extensive cracking due to PWSCC has been identified in SG divider plates made with nickel alloy 600, even with proper primary water chemistry. Specifically, cracks have been detected in the stub runner, very close to the tubesheet/stub runner weld and with depths of almost a third of the divider plate thickness. Therefore, the staff notes that the Water Chemistry Program alone may not be effective in managing the aging effect of cracking due to PWSCC in SG divider plates.

Such cracks could impact adjacent items, such as the tubesheet and the channel head, if they propagate to the boundary with these items. For the tubesheet, PWSCC cracks in the divider plate could propagate to the tubesheet cladding with possible consequences to the integrity of the tube/tubesheet welds. For the channel head, the PWSCC cracks in the divider plate could propagate to the SG triple point and potentially affect the pressure boundary of the SG channel head.

Request:

Please discuss the materials of construction of your Unit 1 SG divider plate assembly. If these materials are susceptible to cracking (e.g., nickel alloy 600 or the associated nickel alloy 600 weld materials), please discuss the potential for cracking in the divider plate to propagate into other components (e.g., tubesheet cladding).

If propagation into these other components cannot be ruled out, please describe an inspection program (examination technique and frequency) to ensure that there are no cracks propagating into other items (e.g., tubesheet and channel head) that could challenge the integrity of other adjacent items.