W3F1-2018-0043, Amendment 2 to License Renewal Application

Amendment 2 to License Renewal Application

ML18211A593
Person / Time
Site:	Waterford
Issue date:	07/30/2018
From:	Dinelli J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2018-0043
Download: ML18211A593 (7)

Text

Entergy Operations Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6660 Fax 504-739-6678 jdinell@entergy.com John C. Dinelli Site Vice President Waterford 3 W3F1-2018-0043 10 CFR 54.21 July 30, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Amendment 2 to License Renewal Application (LRA)

Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy Letter W3F1-2016-0012, License Renewal Application, Waterford Steam Electric Station, Unit 3 dated March 23, 2016.

2. Entergy Letter W3F1-2016-0016, License Renewal Application Drawings, Waterford Steam Electric Station, Unit 3 dated March 23, 2016

3. Entergy Letter W3F1-2017-0081, Amendment 1 to License Renewal Application (LRA), Waterford Steam Electric Station, Unit 3 dated November 15, 2017

Dear Sir or Madam:

In accordance with 10CFR54.21(b), each year following submittal of the license renewal application (LRA) and at least 3 months before scheduled completion of the NRC review an amendment to the renewal application must be submitted that identifies any changes to the current licensing basis that materially affects the content of the LRA including the Updated Final Safety Analysis Report (UFSAR) supplement. In References 1 and 2, Entergy Operations, Inc (Entergy) applied for renewal of the Waterford 3 operating license. Reference 3 provided Amendment 1 to the Waterford 3 license renewal application. This letter provides the 2018 annual update which is Amendment 2 of the Waterford 3 license renewal application.

No new commitments have been identified in this letter.

W3F1-2018-0043 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on July 26, 2018.

If you have any questions or require additional information, please contact John Jarrell, Regulatory Assurance Manager, at 504-739-6685.

Attachments:

I . Annual Update Amendment 2 cc: Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 7601 1-4511 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Dr. April Pulvirenti Washington, DC 20555-0001 U. S. Nuclear Regulatory Commission Attn: Phyllis Clark Washington, DC 20555-0001 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P.O. Box 4312 Baton Rouge, LA 70821-4312

Attachment 1 to W3F1-2018-0043 Annual Update Amendment 2 Waterford 3 License Renewal Application to W3F1-2018-0043 Page 1 of 4 Annual Update Amendment 2 Waterford 3 License Renewal Application 1 INTRODUCTION In accordance with 10CFR54.21(b), each year following submittal of the license renewal application (LRA) and at least 3 months before scheduled completion of the NRC review an amendment to the renewal application must be submitted that identifies any changes to the current licensing basis (CLB) that materially affects the content of the license renewal application, including the Updated Final Safety Analysis Report (UFSAR) supplement.

Amendment 2 is based on review of documents potentially affecting the CLB during the period of July 1, 2017 through April 30, 2018.

2

SUMMARY

OF CHANGES The review concluded that certain sections of the LRA are affected by changes to the CLB documents and other related LRA reviews. The table below summarizes the changes listing the affected systems (if applicable), a description of the change (including effect on the LRA),

and the affected LRA section.

Affected LRA Sections Change LRA Section Affected Section A.1.6 W3F1-2017-0082 Section A.1.15 The applicable ASME Code Edition for the fourth 10-year Section A.1.32 interval was changed. The Inservice Inspection Plan and Section A.1.39 Containment Inservice Inspection Plan will be implemented Section B.1.6 in accordance with ASME Code Section XI, 2007 Edition, Section B.1.15 2008 Addenda. Applicable LRA sections are revised Section B.1.16 accordingly. Section B.1.32 Section B.1.39 to W3F1-2018-0043 Page 2 of 4 3 LRA CHANGES WF3 LRA changes are shown below. Additions are shown with underline and deletions with strikethrough.

A.1.6 Containment Inservice Inspection - IWE Program The code of record for the examination of the WF3 Class MC and Class CC components is ASME Code Section XI, Subsections IWE and IWL, 20012007 Edition with the 2003 2008 Addenda, as mandated and modified by 10 CFR 50.55a.

A.1.15 Inservice Inspection Program The Inservice Inspection Program manages cracking, loss of material, and reduction in fracture toughness for ASME Class 1, 2, and 3 pressure-retaining components including welds, pump casings, valve bodies, integral attachments, and pressure-retaining bolting using periodic volumetric, surface, and visual examination and leakage testing of ASME Class 1, 2 and 3 components as specified in ASME Section XI code, 20012007 Edition, 2003 addendum2008 Addenda. Additional limitations, modifications and augmentations described in 10 CFR 50.55a are included as a part of this program. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in 10 CFR 50.55a. Repair and replacement activities for these components are covered in Subsection IWA of the ASME code edition of record.

A.1.32 Reactor Head Closure Studs Program The Reactor Head Closure Studs Program manages cracking and loss of material due to wear or corrosion for reactor head closure studs bolting (studs, washers, and nuts) using inservice inspection (ASME Section XI 20012007 Edition, 2003 Addendum2008 Addenda Table IWB-2500-1) and preventive measures to mitigate cracking. Preventive actions include avoiding the use of metal-plated stud bolting, use of an acceptable surface treatment, use of stable lubricants, and use of bolting material that has actual yield strength of less than 150 ksi for all studs. The program detects cracks, loss of material and leakage using visual, surface and volumetric examinations as required by ASME Section XI. The program also relies on recommendations to address reactor head closure studs degradation listed in NUREG-1339 and NRC RG 1.65.

A.1.39 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program The Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program manages reduction in fracture toughness and cracking. The program consists of a determination of the susceptibility of CASS piping, piping components, and piping elements to thermal aging embrittlement based on casting method, molybdenum content, and percent ferrite. For potentially susceptible components aging management is accomplished through qualified visual inspections, such as enhanced visual examination, qualified ultrasonic testing methodology, or component-specific flaw tolerance evaluation in accordance with ASME Section XI code, to W3F1-2018-0043 Page 3 of 4 20012007 Edition, 2003 addendum2008 Addenda. Applicable industry standards and guidance documents are used to delineate the program.

B.1.6 Containment Inservice Inspection - IWE Program Description The Containment Inservice Inspection (CII) - IWE Program implements the requirements of 10 CFR 50.55a. The regulations in 10 CFR 50.55a impose the inservice inspection (ISI) requirements of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Subsection IWE, for steel containments (Class MC) and steel liners for concrete containments (Class CC).

The WF3 containment is a low-leakage, free-standing steel containment vessel (SCV) consisting of a vertical upright cylinder with a hemispherical dome and an ellipsoidal bottom.

The SCVs ellipsoidal bottom is encased in concrete and founded on the common concrete foundation with the shield building. The common concrete foundation with the shield building is classified as Class CC equivalent. The steel ellipsoidal bottom plate of the SCV was erected on top of the common concrete foundation slab with a concrete slab poured on top of the bottom plate. Since the Class CC equivalent concrete foundation slab and the bottom steel plate are inaccessible, they are exempted from examination in accordance with IWL-1220(b) and IWE-1220(b). There are no tendons associated with the WF3 SCV. The code of record for the examination of the WF3 Class MC and Class CC components is ASME Code Section XI, Subsections IWE and IWL, 20012007 Edition, with the 20032008 Addenda, as mandated and modified by 10 CFR 50.55a.

B.1.15 Inservice Inspection Program Description The Inservice Inspection (ISI) Program manages cracking, loss of material, and reduction in fracture toughness for ASME Class 1, 2, and 3 pressure-retaining components including welds, pump casings, valve bodies, integral attachments, and pressure-retaining bolting using periodic volumetric, surface, and visual examination and leakage testing of ASME Class 1, 2 and 3 components as specified in ASME Section XI code, 20012007 Edition, 2003 addendum2008 Addenda. Additional limitations, modifications and augmentations described in 10 CFR 50.55a are included as a part of this program. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in 10 CFR 50.55a. Repair and replacement activities for these components are covered in Subsection IWA of the ASME code edition of record.

B.1.16 Inservice Inspection - IWF Program Description The ISI-IWF Program is in its third 10-year ISI inspection interval. The ISI-IWF Program was developedis implemented in accordance with ASME Section XI, 20012007 Edition, through the 20032008 Addenda as approved by 10 CFR 50.55a. In accordance with 10 CFR 50.55a(g)(4)(ii), the WF3 ISI-IWF Program is updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified twelve months before the start of the inspection interval.

to W3F1-2018-0043 Page 4 of 4 B.1.32 Reactor Head Closure Studs Program Description The Reactor Head Closure Studs Program manages cracking and loss of material due to wear or corrosion for reactor head closure stud bolting (studs, washers, and nuts) using inservice inspection (ASME Section XI 20012007 Edition, 2003 Addendum2008 Addenda Table IWB-2500-1) and preventive measures to mitigate cracking. Preventive actions include avoiding the use of metal-plated stud bolting, use of an acceptable surface treatment, use of stable lubricants, and use of bolting material that has actual yield strength of less than 150 ksi for all studs. The program detects cracks, loss of material and leakage using visual, surface and volumetric examinations as required by ASME Section XI. The program also relies on recommendations to address reactor head closure studs degradation listed in NUREG-1339 and NRC RG 1.65.

B.1.39 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Program Description The Thermal Aging Embrittlement of CASS Program is a new program that will manage reduction in fracture toughness and cracking. The program consists of a determination of the susceptibility of CASS piping, piping components, and piping elements to thermal aging embrittlement based on casting method, molybdenum content, and percent ferrite. For potentially susceptible components aging management is accomplished through qualified visual inspections, such as enhanced visual examination, qualified ultrasonic testing methodology, or component-specific flaw tolerance evaluation in accordance with ASME Section XI code, 20012007 Edition, 2003 addendum2008 Addenda. Applicable industry standards and guidance documents are used to delineate the program.