GNRO-2017/00009, Proposed Alternative with 10 CFR 50.55a(z)(1), Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines

Proposed Alternative with 10 CFR 50.55a(z)(1), Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines

ML17074A625
Person / Time
Site:	Grand Gulf
Issue date:	03/15/2017
From:	Nadeau J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2017/00009
Download: ML17074A625 (27)

Text

• Entergy

".~ Entergy-Operations, Inc.

P. O. Box 756 Port-Gibson, MS 39150 James Nadeau Regulatory Assurance Manager Tel. (601) 437-2103 GNRO-2017/00009 March 15,2017 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Proposed Alternative with 10 CFR 50.55a (z) (1), Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines at Grand Gulf Nuclear Station Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29

Dear Sir or Madam:

Pursuant to 10 Code of Federal Regulations (CFR) 50.55a (z) (1), Entergy hereby requests an alternative for Grand Gulf Nuclear Station (GGNS), specific portions of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for In-service Inspection of Nuclear Power Plant Components," on the basis that the proposed alternative provides an acceptable level of quality and safety.

Entergy is requesting an alternative to ASME Section XI, 2007 Edition through the 2008 Addenda, Table IWB-2500-1, Examination Categories B-N-1 and B-N-2, Item Nos. B13.10, which requires examination of accessible areas of the reactor vessel interior each in-service inspection ('-$1) interval by the VT-3 method (B-N-1), B13.20 examination of accessible interior attachment welds within the beltline region each interval by the VT-1 method (B-N-2), B13.30, examination of accessible interior attachment welds beyond the beltline region each interval by the VT-3 method (B-N-2), and B13.40, examination of accessible surfaces of the welded core support structure each interval by the VT-3 method (B-N-2). This relief request is associated with GGNS fourth 10-year lSI interval, which is currently scheduled to begin on June 2, 2017 and ends on June 1,2027. The lSI Code of Record for the fourth interval is ASME Section XI, 2007 edition through the 2008 Addenda.

Entergy request NRC Staff review and approval of this proposed GGNS alternative on or before March 23, 2018.

GNRO-2017/00009 Page 2 of 3 There are no regulatory commitments made in this submittal. If you have any questions or require additional information, please contact James Nadeau at 601-437-2103.

~::::e:ade~~4 Regulatory Assurance Manager IN:sgd

Attachment:

Proposed Alternative Request in Accordance with 10 CFR 50.55a (z) (1) : Comparison of ASME Code Section XI Table IWB-2500-1 and Vessel Attachment Welds : Comparison of ASME Section XI Code Examination Requirements to BWRVIP Examination Requirements : Grand Gulf Reactor Vessel Internal Inspection History cc: with Attachment and Enclosures Mr. John P. Boska, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2 Washington, DC 20555

GNRO-2017/00009 Page 3 of 3 cc: without Attachment and Enclosures Mr. Kriss Kennedy Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 U.S. Nuclear Regulatory Commission ATTN: Mr. Siva Lingam, NRR/DORL (w/2)

Mail Stop OWFN/8 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 Dr. Mary Currier, M.D., M.P.H State Health Officer Mississippi Department of Health P.O. Box 1700 Jackson, MS 39215-1700 Email: mary.currier@msdh.ms.gov

GNRO-2017/00009 Attachment Attachment to GNRO-2017/00009 Proposed Alternative Request in Accordance with 10 CFR 50.55a (z) (1)

GNRO-2017/00009 Attachment Entergy Nuclear Operations, Inc.

Grand Gulf Nuclear Station (GGNS), Unit 1 Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP)

Guidelines in Lieu of Specific ASME Code Requirements Fourth Interval In-service Inspection (lSI) Program Request No. GG-ISI-020 Proposed Alternative Request in Accordance with 10 CFR 50.55a (z) (1)

- Acceptable Level of Quality and Safety -

1. ASME Code Component(s) Affected Code Class: American Society of Mechanical Engineers (ASME)

Code Class 1 Examination B-N-1 (Interior of Reactor Vessel), B-N-2 (Welded Core Supports Category: Structures and Interior Attachments to Reactor Vessel)

Item Number(s): B13.10 (Vessel Interior), B13.20 (Interior Attachments within Beltline Region), B13.30 (Interior Attachments beyond Beltline Region), and B13.40 (Core Support Structure)

Unit / Inspection GGNS, Unit 1, 4 th 10-year lSI Interval starting June 2, 2017 and Interval Applicability: ending on June 1, 2027

2. Applicable Code Requirement(s)

ASME Section XI Code, 2007 Edition through the 2008 Addenda, requires the examination of components, within the Reactor Pressure Vessel. These examinations are included in Table IWB-2500-1, Examinations Categories B-N-1 and B-N-2 and identified with the following Item Numbers:

• Item No. B13.10 - Examine accessible areas of the reactor vessel interior each period by the VT-3 method (B-N-1)

• Item No. B13.20 - Examine accessible interior attachment welds within the beltline region each interval by the VT-1 method (B-N-2)

• Item No. B13.30 - Examine the accessible interior attachment welds beyond the beltline region each interval by the VT-3 method (B-N-2)

• Item No. B13.40 - Examine accessible surfaces of the welded core support structure each interval by the VT-3 method (B-N-2)

These examinations are performed to assess the structural integrity of the reactor vessel interior, its welded attachments and the welded core support structure within the boiling water reactor pressure vessel.

Page 1 of 7

GNRO-2017/00009 Attachment

3. Reason for Request

Entergy is requesting NRC authorization of this proposed alternative to the ASME Section XI Code requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVIP Inspection and Evaluation (I&E) guidelines have recommended aggressive specific inspections by BWR operators to completely identify material condition issues with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. The BWRVIP I&E guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanisms, and required reexamination at conservative intervals. In contrast, the ASME Section XI Code requirements were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience.

4. Proposed Alternative and Basis for Use Proposed Alternative Pursuant to 10 CFR 50.55a (z) (1), Entergy requests authorization to utilize the alternative requirements of BWRVIP Guidelines in lieu of the requirements of ASME Section XI. The proposed alternative is detailed in Enclosure 1, Table 1, "Comparison of ASME Code Section XI Table IWB-2500-1 Examination Category B-N-1 and B-N-2 Requirements to BWRVI P Guidance Requirements," that shows a comparison between the existing ASME Section XI Code and BWRVIP requirements that will be used under this alternative.

Specifically, Entergy will satisfy the Examination Category B-N-1 and B-N-2 requirements at GGNS as described in Enclosure 1, Table 1 in accordance with BWRVIP guidelines in lieu of the associated ASME Section XI Code requirements, including examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting.

Not all of the components addressed by the BWRVIP guidelines are components that require ASME Section XI Code examinations, but the particular guidelines that are applicable to ASME Section XI Code components are listed below along with the BWRVIP-94 Administrative Guide that GGNS will use to implement this alternative:

BWRVIP-03 "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internal Examination Guidelines" BWRVIP-18-R1-A "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" BWRVIP-25 "BWR Core Plate Inspection and Flaw Evaluation Guidelines BWRVIP-26-A "BWR Top Guide Inspection and Flaw Evaluation Guidelines BWRVIP-27-A "BWR Standby Liquid Control System/Core Plate dP Inspection and Flaw Evaluation Guidelines" BWRVIP-38 "BWR Shroud Support Inspection and Flaw Evaluation Guidelines" Page 2 of 7

GNRO-2017/00009 Attachment BWRVIP-41-R3 "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines BWRVIP-47-A "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines BWRVIP-48-A(1) "VessellD Attachment Weld Inspection and Flaw Evaluation Guidelines" BWRVI P-76-R 1-A(2) "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" BWRVIP-94 "BWR Vessel and Internals Project "Program Implementation Guide" BWRVI P-1 OO-A (2) "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds" Notes:

(1) Currently, there are no existing BWRVIP guidelines or ASME Section XI Code requirements regarding the feedwater spargers except for BWRVIP-48-A which governs inspection of reactor vessel internal attachment welds, namely the feedwater sparger brackets. GGNS will continue to use NUREG-0619 on the feedwater sparger piping, spacer brackets, pins, end brackets, flow holes and sparger tee welds outside of this request.

(2) If flaw evaluations are required for BWRVIP-76-R1-A examinations, the fracture toughness values of BWRVIP-1 OO-A will be utilized.

BWRVIP-138 "BWRVIP Updated Jet Pump Beam Inspection and Flaw Evaluation" BWRVIP-139 "BWR Vessel Internals Project, Steam-Dryer Inspection and Flaw Evaluation Guidelines" BWRVIP-183 "BWR Vessel Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines" Note:

(1) Although these BWRVIP guidelines are not part of this request they are used at GGNS.

Enclosure 1, Table 1 compares current ASME Section XI, Table IWB-2500-1, Examination Category B-N-1 and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to GGNS.

In addition to the items in Enclosure 1, Table 1 a detailed Table 2, "Vessel Attachment Welds - Fabricated Either from E-308/E-309 (Furnace Sensitized) Austenitic Stainless Steel or Inconel 182 Material," lists specific vessel attachment welds, fabricated with these materials, that have an increased concern for cracking, with inspection results, and that are currently examined under Examination Category B-N-2 /

BWRVIP and will meet the inspection strategy found in Section 3 of BWRVIP-48-A and BWRVIP-38.

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GNRO-2017/00009 Attachment When implementing the guidance of BWRVIP-94 as requirements to use this request GGNS will meet the following paragraph:

'When BWRVIP guidelines are approved by the Executive Committee and are initially distributed, or subsequently revised, each utility shall modify their vessel and internals program documentation to reflect the new requirements and implement the guidance within two refueling outages, unless a different schedule is identified by the BWRVIP at the time of document distribution. Implementation means not only incorporating the requirements into the utility program, but also performing the initial or baseline inspection and evaluation requirements. As a minimum, the BWRVIP guidelines listed in Section 3.2.21 which have been reviewed and approved by the BWRVIP Executive Committee will be implemented within two refueling outages of the approval date of the Executive Committee, unless a different schedule is identified by the BWRVIP at the time of document distribution. Changes to NRC approved BWRVIP guidelines that are less conservative than those approved by the NRC shall be implemented only after NRC approves the changes. "NRC approved" generally means publication of a

"-A" document or equivalent."

Therefore, where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC authorized proposed alternative to the requirements of 10 CFR 50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for an alternative has been approved. Enclosure 1, Table 1 represents the most current comparison at the time of this request.

Any deviations from the referenced BWRVIP guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process.

In the event that conditions are identified that require repair or replacement and the component is within the jurisdiction of ASME Section XI Code (welded attachments to the RPV or Core Support Structure), the repair or replacement activities will be performed in accordance with ASME Section XI, Article IWA-4000.

Subsequent examinations will be in accordance with the applicable BWRVIP guideline.

Basis for Use BWRs now examine reactor internals in accordance with BWRVIP guidelines. These guidelines have been written to address the safety significant vessel internal components and to examine and evaluate the examination results for these components using appropriate methods and reexamination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVI P approach in principle and has issued Safety Evaluations for many of these guidelines (References 1 - 10).

In support of this request GGNS employs an Online NobleChem' (OLNC) for mitigating IGSCC in BWR Internals. The most current cycle to date Electrochemical Potential (ECP) is shown here as follows:

Cycle = 21: the average ECP mV(SHE) is -487.7 with a minimum of -499 mV(SHE) and a maximum of-422 mV(SHE).

GGNS does not measure molar ratio, but calculates the value using BWRVIP-202, BWRVIA, for Radiolysis and ECP Analysis, Version 3.1. Three values are developed at the beginning of the cycle; BOC (Beginning of Cycle), MOC (Middle of Cycle), and EOC (End of Cycle). The values selected are from the upper downcomer location which is considered the most conservative location by the BWRVIP and thus the values are:

Page 4 of 7

GNRO-2017/00009 Attachment BOC Molar Ratio = 6.2 GGNS does measure Molar Ratio and values are much higher (- 240)

MOC Molar Ratio = 5.9 EOC Molar Ratio = 4.1 Since GGNS uses OLNC, catalyst loading is not applicable and the availability of HWC/HWC+NMCA during normal operation is listed below:

Cycle 19 93.5 0/0 Cycle 20 98.4 0/0 Cycle 21 91.7  % thru 9/16 For specific activities that are required related to leakage assessment in BWRVIP-18-R1-A, Section 5.13, BWRVI P-41-R3, Section 5.1.3, and BWRVI P-76-R 1-A, a plant-specific integrated leakage assessment has been performed for assumed cracks as no known cracks related to leakage are identified in Enclosure 3, "Grand Gulf Reactor Vessel Internal Inspection History." Therefore, GGNS has assumed leakage for flaws in the Core Spray A,R, C,P, F,M, and H,K welds along with the Jet Pump RS-1, RS-2, RS-3, IN-1, IN-2, MX-2, DF-1, DF-2, DF-3, AD-2, RS-6/7, and RS-8/9 welds.

In the context of the integrated leakage assessment over one 24-month cycle, the leakage through the Core Spray welds is conservatively calculated to be 0.0 gallons per minute (gpm) with an allowable limit of 323.3 gpm. For the Jet Pump DF-2 welds, the leakage was calculated to be 0.0 gpm with an allowable limit of 174.7 gpm. For the LPCI coupling, there is a leakage value of 4.194 gpm due to an identified flaw.

The GGNS SAFER/GESTR-LOCA analysis assumed 115 gpm leakage in Core Spray welds, so adding the leakage of the Jet Pump welds DF-2 of 0.0 gpm along with the postulated flaws in the Core Shroud of 22.37 gpm (GIN 2017-00028) and the LPCI coupling of 4.194 gpm, the combined leakage would increase the Peak Cladding Temperature (PCT) approximately 0.0 degrees Fahrenheit (OF). This is because GGNS currently analyzes with relaxed parameters of 115 gpm per core spray line and 860 gpm per LPCI injection line.

The plant-specific integrated leakage assessment concluded that postulated leakage through the Core Shroud at all H3, H4, H6A and H7 combined with leakage in the Jet Pump DF-2 weld and Core Spray piping would increase the PCT analyzed in the GGNS SAFER/GESTR-LOCA analysis by approximately 0.0 of, or from 1712°F to 1712°F but still below the 10 CFR 50.46(b) regulatory limit of 2200 of. The assessment, using conservative assumptions and BWRVIP methodology, demonstrated that leakage from these reactor pressure vessel internals over a 24 month cycle results in a 0 % increase PCT. This is because the Appendix K analysis currently uses relaxed parameters of 7000' gpm versus 7115 gpm per core spray line and 6660 gpm versus 7450 gpm per LPCI line.

As additional justification, Enclosure 2, "Comparison of ASME Section XI Code Examination Requirements to BWRVIP Examination Requirements," provides specific examples that compare the inspection requirements of ASME Section XI Code Item Numbers B13.1 0, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are Page 5 of 7

GNRO-2017/00009 Attachment provided as examples. This comparison also includes a discussion of the inspection methods and where they are applied.

Therefore, based on the Safety Evaluations of many of the guidelines and the comparisons performed demonstrating the use of these guidelines above, Entergy concludes that this alternative request to the subject ASME Section XI Code requirements will avoid unnecessary inspections, while in some cases conserving radiological dose, because the inspections will now be focused on the most recent and actual BWR experience available. Thus, this request when authorized will provide an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

5. Duration of Proposed Alternative Upon authorization by NRC, this request for an alternative to use the BWRVIP Guidelines in lieu of ASME Section XI Code requirements will be implemented during the 4th 10-year lSI Interval beginning on June 2, 2017 and ending on, June 1, 2027, (which will include a portion of the period of extended operation when approved by NRC from November 2, 2024 to the end of the 4th Interval).

6. Precedents The NRC Staff has authorized similar requests for the following licensees and also for the GGNS 3 rd 10-year lSI Interval.

• Cooper Nuclear Station, SER Docket No. 50-298, CAC No. ME6366, February 16, 2016 (ADAMS Accession No. ML16034A479)

• River Bend Nuclear Station, SER Docket No. 50-458, TAC No. MF1867, May 30,2014, Supplemented with SER Correction on November 13, 2014 (ADAMS Accession No. ML14183A086)

• Grand Gulf Nuclear Station, Unit 1, SER Docket No. 50-416, TAC No. MF3678, August 1, 2014 (ADAMS Accession No. ML14184A782)

7. References

1. Letter USNRC to BWRVIP, dated January 30, 2012 (ML113620684), "Final Safety Evaluation for Electric Power Research Institute Boiling Water Reactor Vessel and Internals Project Technical Report 1016568, BWRVIP-18, Revision 1: BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" (TAC No. ME2189)"

2. Letter USNRC to BWRVIP, dated December 19, 1999, "Safety Evaluation of BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25),"

(ADAMS Accession Nos. ML993620267 and ML993620274)

3. Letter USNRC to BWRVIP, dated August 29,2005, "NRC Approval Letter of BWRVIP-26-A, "BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines", (ADAMS Accession Nos. ML043290158 and ML052490550).

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GNRO-20 17/00009 Attachment

4. Letter USNRC to BWRVIP, dated June 10,2004, Proprietary Version of NRC Staff Review of BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate DP Inspection and Flaw Evaluation Guidelines", (ADAMS Accession No. ML041700446).

5. Letter USNRC to BWRVIP, dated July 24,2000, "Final Safety Evaluation of the "BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38),"

EPRI Report TR-1 08823 (TAC NO. M99638)"

6. Letter USNRC to BWRVIP, dated February 4,2001, "Final Safety Evaluation of the "BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines (BWRVIP-41)," (TAC NO. M99870)"

7. Letter USNRC to BWRVIP, dated September 1,2005, "NRC Approval Letter of BWRVIP-47-A, "BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines". (ADAMS Accession Nos. ML043290026 and ML052490537).

8. Letter USNRC to BWRVIP, dated July 25,2005, "NRC Approval Letter of BWRVIP- 48-A, "BWR Vessel and Internals Project Vessel 10 Attachment Weld Inspection and Flaw Evaluation Guideline".

(ADAMS Accession Nos. ML12214A522 and ML052130284).

9. Letter USNRC to BWRVIP, dated July 28,2006, "Safety Evaluation of the "BWR Vessel and Internals Project, BWR Core Shroud Weld Inspection and Evaluation Guidelines (BWRVIP-76)"

1o. Letter USNRC to BWRVIP, dated November 1, 2007, "NRC Approval Letter with Comment for BWRVIP-100-A, BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds," (ADAMS Accession No. ML073050135)

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GNRO-2017/00009 Enclosure 1 to Attachment Enclosure 1 Comparison of ASME Code Section XI Table IWB-2500-1 and Vessel Attachment Welds TABLE 1 - Comparison of ASME Code Section XI Table IWB-2500-1 Examination Category B-N-1 and B-N-2 Requirements to BWRVIP Guidance Requirements (1)

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B13.10 Reactor Vessel Interior Areas of the RPV VT-3 Each Period None While there is not a specific BWRVIP Guideline that above and below addresses the scope of B-N-1 , the examinations performed the core made by BWRVIP-18, 25, 26, 27, 41, 47,138 provide a general accessible during a overview of the reactor interior which may be considered normal refuel. representative of the B-N-1 scope ..

B13.20 Interior Attachments within Accessible Welds VT-1 Each 10-year I BWRVIP-48-A, Riser Brace EVT-1 100% in first 12 Beltline - Riser Braces Interval Table 3-2 Attachment years, 25% during each subsequent 6 years Lower Surveillance BWRVIP-48-A, I Bracket Attachment VT-1 Each 10-Year Specimen Holder Brackets Table 3-2 Interval I

B13.30 I Interior Attachments Accessible Welds VT-3 Each 10-year BWRVIP-48-A, I Bracket Attachment VT-3 I Each 10-Year beyond Beltline - Steam interval Table 3-2 Interval Dryer Hold-down Brackets Guide Rod Brackets I I I BW RVIP-48-A, I~et Attachment VT-3 Each 10-Year Table 3-2 Interval Steam Dryer Support BWRVIP-48-A, I Bracket Attachment EVT-1 I Each 10-Year Brackets Table 3-2 Interval Feedwater Sparger BWRVIP-48-A, Bracket Attachment EVT-1 Each 10-Year Brackets Table 3-2 Interval Page 1 of 3

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Core Spray Piping Brackets I I I BW RVIP-48-A, I Bracket Attachment EVT-1 Every 4 Refueling Table 3-2 Cycles Upper Surveillance BWRVIP-48-A, Bracket Attachment VT-3 Each 10-Year Specimen Holder Brackets Table 3-2 Interval Shroud Support BWRVIP-38, 3.1.3.2, Weld H-9 EVT-1 or UT I forMaximum of 6 years EVT-1, Maximum (Weld H9) Figures 3-2 of 10 years for UT and 3-5 B13.40 I Integrally Welded Core Accessible VT-3 Each 10-year BWRVIP-38, Shroud support EVT-1 or UT Based on as-found Support Structure Surfaces I interval 3.1.3.2, welds H8 and H9 (2) conditions, to a Figures 3-2 including gussets maximum 6 years for and 3-5 one side EVT-1, 10 years for UT where accessible Core Shroud Horizontal BWRVIP Welds H1-H7 as UT or EVT-1 I Based on as-found Welds R1-A,2.2 applicable conditions, to a maximum of 10 years for UT when inspected from both sides of the welds Core Shroud Vertical Welds BWRVIP Vertical Welds as EVT-1 or UT I Maximum 10 years R1-A,2.3 applicable for UT based on inspection of horizontal welds Core Shroud Repairs (3) BWRVIP Tie-Rod Repair VT-3 I In accordance with R1-A,3.5 designer recommendations per BWRVIP-76 R1 Notes:

(1) This table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1 and the appropriate BWRVI P Document.

Page 2 of 3

GNRO-2017/00009 to Attachment (2) In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined.

(3) No repairs have been performed on the core shroud.

TABLE 2 - Vessel Attachment Welds - Fabricated Either from E-308/E-309 (Furnace Sensitized) Austenitic Stainless Steel or Inconel 182 Material

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<<VY~I~~(i~9~"~)< < ill Jet Pump Riser Brace I (Furnace Sensitized) I BWRVIP-48-A I No indications have been identified; no repairs have been (B13D001-JPB01A1B to 12A1B) Austenitic Stainless Steal required Steam Dryer Hold Down (Furnace Sensitized) I BWRVIP-48-A No indications have been identified; no repairs have been Bracket Austenitic Stainless Steal required (B13D001-SHD01 to 06)

Core Spray Piping Bracket Inconel182 BWRVIP-48-A No indications have been identified; no repairs have been (B13D001-CS1 U to 6U and required B13D001-CS1L to 6L)

Steam Dryer Support Bracket Inconel182 BWRVIP-48-A No indications have been identified; no repairs have been (B13D001-SL 1 to 6) required Guide Rod Bracket Inconel182 BWRVIP-48-A No indications have been identified; no repairs have been (B13D001-GB1 and 2) required Feedwater Sparger Bracket Inconel182 BWRVIP-48-A No indications have been identified; no repairs have been (B13D001-FW01 to 12) required Shroud Support Welds (H9) Inconel182 BWRVIP-38 No indications have been detected; no repairs have been required Shroud Support Leg (H12) Inconel182 BWRVIP-38 Although not required by the BWRVIP-38, in their final safety evaluation of the BWRVIP-38 the NRC requires inspection of the shroud support leg welds (H12) when appropriate inspection tooling and methodologies are developed. At Grand Gulf welds H8 and H9 are structurally adequate; therefore, support leg inspection is not required.

Page 3 of 3

GNRO-2017/00009 to Attachment Enclosure 2 Comparison of ASME Section XI Code Examination Requirements to BWRVIP Examination Requirements The following discussion provides a comparison of the examination requirements provided in ASME Section XI Code Item Numbers B1*3.1 0, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. Code Requirement - B13.10 - Reactor Vessel Interior Accessible Areas (B-N-1)

ASME Section XI Code requires a VT-3 examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately 3 years, during the first inspection interval, and each period during each successive 1O-year Inspection Interval.

Typically, these examinations are performed every other refueling outage of the Inspection Interval. This examination requirement is a non-specific requirement that is a departure from the traditional Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts, loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products, wear, and structural degradation.

Portions of the various examinations required by the applicable BWRVIP Guidelines require access to accessible areas of the reactor vessel during each refueling outage. Examination of Core Spray Piping and Spargers (BWRVIP-18-R1-A~, Top Guide (BWRVIP-26-A), Jet Pump Welds and Components (BWRVIP-41-R3), Interior Attachments (BWRVIP-48-A), Core Shroud Welds (BWRVIP-76-R1-A), Shroud Support (BWRVIP-38) and Lower Plenum Components (BWRVIP-47-A) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible),

and surrounding the core (annulus area) entails access by remote camera systems that essentially performs equivalent VT-3 examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by the ASME Section XI. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements.

Therefore, the specified BWRVIP Guideline requirements meet or exceed the subject Code requirements for examination method and frequency of the interior of the reactor vessel. Accordingly, these BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Section XI Code requirements.

2. Code Requirement - B13.20 - Interior Attachments Within the Beltline (B-N-2)

The ASME Section XI Code requires a VT-1 examination of accessible reactor interior surface attachment welds within the beltline each 10-year interval. In the boiling water reactor, this includes the jet pump riser brace welds-to-vessel wall and the lower surveillance specimen support bracket welds-to-vessel wall. In comparison, the BWRVIPrequires the same examination method and frequency for the lower surveillance Page 1 of 4

GNRO-20 17/00009 to Attachment specimen support bracket welds, and requires an EVT-1 examination on the remaining attachment welds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years.

The jet pump riser brace examination requirements are provided below to show a comparison between the ASME Section XI Code and the BWRVIP examination requirements.

Comparison to BWRVIP Requirements - Jet Pump Riser Braces (BWRVIP-41-R-3 and BWRVIP-48-A)

• The ASME Section XI Code requires a 100 % VT-1 examination of the jet pump riser brace-to-reactor vessel wall pad welds each 1O-year interval.

• The BWRVIP requires an EVT-1 examination of the jet pump riser brace-to-reactor vessel wall pad welds the first 12 years and then 25% during each subsequent 6 years.

• BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined.

The ASME Section XI Code VT-1 examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVI P Enhanced VT-1 (EVT-1) is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and inter-granular stress corrosion cracking (IGSCC), the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT- 1 examination.

The ASME Section XI Code 2007 Edition through the 2008 Addenda, VT-1 visual examination method requires that a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires the same 0.044 inch resolution on the examination surface and additionally the performance of a cleaning assessment and cleaning as necessary. While the jet pump riser brace configuration varies depending on the vessel manufacturer, BWRVIP-48-A includes diagrams for each configuration and prescribes examination for each configuration including GGNS.

The calibration standards used for BWRVIP EVT-1 exams utilize the same ASME Section XI Code characters, thus assuring at least equivalent resolution compared to the Code. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced flaw detection capability of an EVT-1, with a less frequent examination schedule provides an acceptable level of quality and safety to that provided by the ASME Section XI Code.

3. Code Requirement - B13.30 - Interior Attachment Beyond the Beltline Region (B-N-2)

The ASME Section XI Code requires a VT-3 examination of accessible reactor interior surface attachment welds beyond the beltline each 1O-year interval. In the boiling water reactor, this includes the core spray piping primary and supplemental support bracket welds-to-vessel wall, the upper surveillance specimen support bracket welds-to-vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support and hold-down bracket welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, the shroud support plate-to-vessel wall, and shroud support gussets.

BWRVIP-48-A requires as a minimum the same VT-3 examination method as the Code for some of the interior attachment welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of examination, the same scope of examination (accessible welds), the same examination Page 2 of 4

GNRO-2017/00009 to Attachment frequency (each 10 year interval) and ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provide by the ASME Section XI Code.

For the Core Spray support bracket attachment welds, the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-vessel welds, as applicable, the BWRVIP Guidelines require an EVT-1 examination at the same frequency as the ASME Section XI Code, or at a more frequent rate. Therefore, the BWRVIP requirements provide the same level of quality and safety to that provided by the ASME Section XI Code.

The Core Spray piping bracket-to-vessel attachment weld is used as an example for comparison between the Code and BWRVIP examination requirements as discussed below.

Comparison to BWRVIP Requirements - Core Spray piping Bracket Welds (BWRVIP-48-A)

• The ASME Section XI Code examination requirement is a VT-3 examination of each weld every 10 years.

• The BWRVIP examination requirement is an EVT-1 for the core spray piping bracket attachment welds with each weld examined every four cycles (8 years for units with a two year fuel cycle). The BWRVIP examination method EVT-1 has superior flaw detection and sizing capability, the examination frequency is greater than the Code requirements, and the same flaw evaluation criteria are used.

• The Code VT-3 examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An enhanced EVT-1 is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, the relevant degradation mechanisms for BWR internal attachments.

Therefore, with the EVT-1 examination method, the same examination scope (accessible welds), an increased examination frequency (8 years instead of 10 years) in some cases, the same flaw evaluation criteria (Section XI), the level of quality and safety provided by the BWRVIP criteria is superior than that provided by the Code.

4. Code Requirement - B13.40 - Integrally Welded Core Support Structures (B-N-2)

The ASME Section XI Code requires a VT-3 examination of accessible surfaces of the welded core support structure each 1a-year interval. In the boiling water reactor, the welded core support structure has primarily been considered the shroud support structure, including the shroud support plate (annulus floor) the shroud support ring, the shroud support welds, the shroud support gussets. In later designs, the shroud itself is considered part of the welded core support structure. Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examination replaces this ASME Section XI Code requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

Comparison to BWRVIP Requirements - BWR Shroud Support (BWRVIP-38)

• The ASME Section XI Code requires a VT-3 of accessible surfaces each 1a-year interval.

Page 3 of 4

GNRO-2017/00009 to Attachment

• The BWRVIP requires as a minimum the same examination method (VT-3) as the ASME Section XI Code for integrally welded Core Support Structures, and for specific areas, requires either an enhanced visual examination technique (EVT-1) or volumetric examination (UT).

BWRVIP recommended examinations of integrally welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones.

As a minimum, the same or superior visual examination technique is required for examination at the same frequency as the ASME Section XI Code examination requirements. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at a frequency identical to the ASME Section XI Code requirement.

For other integrally welded core support structure components, the BWRVIP requires an EVT-1 or UT of core support structures. The core shroud is used as an example for comparison between the ASME Section XI Code and BWRVI P examination requirements as shown below.

Comparison to BWRVIP Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline (BWRVIP-76-R1-A)

• The ASME Section XI Code requires a VT-3 examination of accessible surfaces every 10 years.

• The BWRVIP requires an EVT-1 examination from the inside and outside surface where accessible or ultrasonic examination of each core shroud circumferential weld that has not been structurally replaced with a shroud repair at a calculated "end of interval" (EOI) that will vary depending upon the amount of flaws present, but not to exceed ten years.

The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than "all accessible surfaces". The BWRVIP examination methods (EVT-1 or UT) are superior to the ASME Section XI Code required VT-3 examination for flaw detection and characterization. The BWRVIP examination frequency is equivalent to or more frequent than the examination frequency required by the ASME Section XI Code. The superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency and the comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that provided by the Code requirements.

Page 4 of 4

GNRO-20 17/00009 to Attachment Enclosure 3 Grand Gulf Reactor Vessel Internal Inspection History Core Shroud Spring 1995 UT Baseline per BWRVIP-01. All accessible areas of H3, H4, H6A and H7. No indications.

Spring 1998 UT All accessible areas of H3, H4, H6A, H7. No indications.

Spring 2004 UT 15.1% of H3 Lower Side and 34.6% of H4. Due to equipment failures this examination was deferred to next outage.

Fall 2005 UT 44% of H3 Lower Side, 56.6% H4 Both Sides, 17.3% H6A Both Sides and @ 20% H7 Both Sides.

One indication with characteristics associated with IGSCC/IASCC was detected on the lower side of the H4 weld. Indication is 1.11" in length. Due to disassembly of the JP11 mixer, a VT-3 examination was performed on accessible areas of H10, H11 and H12. No indications.

Spring 2014 UT 17.8% Upper Side and 16.5% Lower Side of H6A 18.8% Upper Side and 21.6% Lower Side of H7 No indications identified Spring 2016 UT 81.3% Upper Side and 80.0% Lower side of H3 73.4% Upper Side and 73.4% Lower Side of H4 Two indications were detected on the lower side of the H3 weld. Indications were a total of 4.33" in length. One additional indication was detected on the upper side of the H4 weld with a length of 1.19".

Each indication had characteristics associated with EGSCC/IASCC.

Shroud Support Spring 1992 VT-1 Shroud shell weld. No indications. (SIL 572)

Spring 1995 VT-3 SSHAC @ 180°. No indications.

Fall 1996 VT-1 Sect XI. Period 3 of 1O-yr interval. RF05/6 Attachment welds to vessel and shroud plate to shroud weld. No indications.

Spring 1998 UT 10.7% of total circumference of H8 (shroud support plate to shroud weld) and 15.4% of H9 (shroud support plate to vessel weld). No indications.

Fall 2002 VT-1 SSAHC @ 0°. No indications.

Page 1 of 9

GNRO-2017/00009 to Attachment

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Fall 2005 VT-1 SSAHC @ 0°. No indications.

Spring 2007 EVT-1 15% of the top of H8 and 18.5% of the top of H9.

No indications were noted.

Fall 2008 VT-1 SSAHC @ 0°. No indications.

Spring 2012 EVT-1 More than 10% of the top of both H8 and H9 was inspected. No indications were noted.

Spring 2014 EVT-1 SSAHC @ 0°. No indications noted.

Core Spray Spring 1998 EVT-1 All accessible piping locations. No indications.

Piping Fall 1999 EVT-1 All accessible P2, P2a, P3a, P5. 25% of remaining piping locations. No indications.

Spring 2001 EVT-1 All accessible P2, P2a, P3a, P5. 25% of remaining piping locations. No indications.

Fall 2002 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations. No indications.

Spring 2004 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations. No indications.

Fall 2005 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations. No indications.

Spring 2007 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations. No indications.

Fall 2008 EVT-1 All target welds (P3a and P5) with 25% of remaining piping .locations. No indications.

Spring 2010 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations with the exception of P3a(L)-270°,

P3a(R)-90° and P3a(L)-90° (Ref. CR-GGN-2013-06541). No indications.

Spring 2012 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations with the exception of P3a(L)-270° I P3a(R)-270° and P3a(R)-90° (Ref. CR-GGN-2013-06541). EVT-1 Indication discovered on P8A as documented on INR GGNS-IVVI-12-02 Spring 2014 VT-1 / VT-3 / All target welds (P3a and P5) with 25% of remaining EVT-1 piping locations. Indication discovered on P8A was inspected with no change in indication. No other indications discovered.

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GNRO-2017/00009 to Attachment

'InspectiorfResults Spring 2016 EVT-1 All target welds (P3a and P5) with 25% of remaining piping locations. No other indications discovered.

Core Spray Fall 1996 VT-3 Augmented exam per IE 80-13. No indications noted.

Sparger Spring 1998 EVT-1 / CS- Accessible areas of spargers, tee boxes, brackets and VT-1 supports. Broken tack welds @ Cap Screw 7A.

Fall 1999 VT-1 / VT-3 Upper Sparger - Accessible areas of spargers, tee boxes, brackets and supports. No indications noted.

Fall 2002 VT-1 / VT-3 All core spray sparger target welds and all accessible areas of the lower sparger welds. No indications noted.

All accessible areas of Core Spray Brackets (SB).

Broken tack welds @ Cap Screw 7A previously reported.

Fall 2005 VT-1 / VT-3 All core spray sparger target welds and all accessible areas of the upper sparger welds. Accessible areas of Core Spray Sparger Brackets (SB). No indications noted. Broken tack welds @ Cap Screw 7A previously reported. Additional broken tack weld identified at Cap Screw 15C.

Fall 2008 EVT-1 / VT-1 All core spray sparger target welds and all accessible areas of the lower sparger welds. Accessible areas of Core Spray Sparger Brackets (SB). No indications noted. Broken tack welds @ Cap Screw 7A and 15C previously reported.

Spring 2010 VT-1 Performed examinations of previous indications at Cap Screw 7A and 15C. No changes noted.

Spring 2012 EVT-1 / VT-1 All core spray sparger target welds and all accessible areas of the upper sparger welds.

Accessible areas of Core Spray Sparger Brackets (SB) were inspected with no indications. Performed examinations of previous indications at Cap Screw 7A and 15C. No changes noted. Tack weld, indications on alignment sleeve documented on INR GGNS-IVVI-12-03 and INR GGNS-IVVI-12-05.

Spring 2014 EVT-1 / VT-1 Performed examination of previous indications at Cap Screw 7A and 15C and tack weld indications on alignment sleeve. No discernible changes noted on the indications.

Page 3 of 9

GNRO-2017/00009 to Attachment Component Irlspectiorr Results Spring 2016 EVT-1/ VT-1 Performed VT-1 examinations of previous indications at Cap Screw 7A and 15C. No discernable changes noted on the indications.

Performed VT-1 examinations of accessible areas of the lower sparger welds, and accessible areas of Core Spray Sparger Brackets (SB) were inspected with no indications.

Indication of rolled metal located on 1720 Lower Boss Pad. Bolt threading protruding from backside of Bracket 12 Bolt "0".

Top Guide (Rim, Fall 1996 VT-3 Accessible surfaces and fasteners. No indications etc.) noted.

Spring 2001 VT-3 Accessible surfaces and fasteners. No indications noted.

Spring 2007 VT-3 Accessible surfaces and fasteners. No indications noted.

Spring 2012 EVT-1 Accessible surfaces and fasteners. Indication documented under INR GGNS-IVVI-12-06.

Core Plate (Rim, Fall 1996 VT-3 Section XI, under core plate. Where access was etc.) provided in RF08, camera work was performed. No indications noted.

Spring 2007 VT-3 Accessible surfaces of the shroud support structure.

No indications noted.

SLC N/A N/A N/A Jet Pump Fall 1996 UT UT performed on JP beams. Two beams cracked in Assembly RF06 and all were replaced with Unit 2 spares. No UT exams were done in RF07. RF08 changed out all beams with the new GE design.

Spring 1998 EVT-1 / VT-3 Accessible areas of RS-3 weld on JP 0102, JP 0304 and JP 0506. VT-3 on flow restriction on JP 09,10, 11 and 24. No indications noted.

Fall 1999 EVT-1 Accessible areas of RS-3 weld at JP07/08, JP09/10 and JP11/12. No indications noted.

Spring 2001 EVT-1 Accessible areas of RS-1 and RS-2 welds on JP01/02. No indications noted.

Fall 2002 EVT-1 All required locations for JP 0304 and JP 0910.

Examination exceptions are RB-1b, RB-1d, RB2a-d for JP0304; welds OF-1 for JP03 and JP04; OF-3 for JP03 and JP1 0; IN-1 and IN-2 for JP04; IN-2 for JP1 O. No indications noted.

Page 4 of 9

GNRO-2017/00009 to Attachment Spring 2004 EVT-1 / VT-1 Completed remaining examinations on JP 0304 and 0910. Completed baseline on 50% of low and medium priority locations and 100% of high priority (RS-3) locations. Identified and inspected an additional RS-1 weld at JP 0910 and inspected additional weld at the DF-3 location. The additional weld at the DF-3 location was identified in the Fall 2002 outage (DF-3a). No indications noted.

Jet Pump Fall 2005 EVT-1 Wedge examinations were completed on 12 jet Assembly pumps. Wedge exams have been completed on all (continued) jet pumps with no indications. Examined one IN-1 and IN-2 location with no indications noted.

Spring 2007 EVT-1 / UT Wedge examination performed on 4 wedges due to disassembly of Jet Pumps in previous outages.

EVT-1 was performed on the Riser Brace to vessel weld (5 locations). UT performed on 21 of 24 Jet Pump beams. Three beams have been replaced with new beams and do not require UT at this time.

Fall 2008 EVT-1 Performed examinations on Jet Pump wedges 1 thru

12. No wear was identified; however slight wear was noted on wedge rods JP 01, JP 02, JP 05, JP 06, JP 07 and JP 09. No additional exams were performed.

Spring 2010 EVT-1 Completed baseline examinations (148 locations).

Performed additional inspections of Jet Pump Wedges (12) and Riser Braces (12) due to Laguna Verde DE. No indications were noted.

Spring 2012 UT UT performed on 21 of 24 Jet Pump beams. Three beams have been replaced with new beams and do not require UT at this time. No indications noted.

Spring 2014 EVT-1 / VT-1 EVT-1NT-1 performed on 39 locations Jet Pump

/ VT-3 Riser and Diffuser Welds with no indications. VT-3 performed on eight (8) sensing lines with no indications. VT-1 was performed on all 24 jet pump wedges with 2 indications found on the Stellite Cladding of Jet Pumps #5 and #16. Indications documented under INR GGNS-IVVI-14-01 R1.

Spring 2016 EVT-1/ VT-1 VT-1 was performed on 2 jet pump wedges that had indication in previous outage. No changes were noted. EVT-1 performed on 50 locations Jet Pump Riser and Diffuser Welds with no indications.

CRD Guide Fall 1996 VT-3 8 guide tubes. When accessibility permits. No Tubes indications noted.

Page 5 of 9

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Spring 1998 VT-3 34 CRGT-1 exams completed with no indications noted.

Spring 2001 VT-3 12 guide tubes. 12 FS/GT-ARPIN-1 and CRGT-1.

Accessible portions of CRGT-2 (2 places). No indications noted.

Fall 2002 EVT-1 CRGT-2 & 3 (10 places). FS/GT-ARPIN-1 (2 places). No indications noted.

CRD Guide Spring 2008 EVT-1 Completed baseline exams on 10 CRD Guide Tubes Tubes. No indications were noted.

(continued)

Dry Tubes Spring 1998 VT-3 11 guide tubes. No indications noted.

Fall 2002 VT-1 Accessible areas of 6 LPRM dry tubes. No indications noted.

VT-1 Spring 2007 Accessible areas of 14 SRM/IRM and 7 LPRMS. No indications noted.

VT-1 Fall 2008 Performed inspections on 24 LPRM dry tubes. No indications noted VT-1 Performed exams on 14 SRM/IRM dry tubes. Four dry Spring 2010 tubes had indications.

Spring 2012 VT-1 Performed inspections on 5 LPRM dry tubes.

Replaced four dry tubes that had indications in Spring 2010. No indications noted.

Spring 2014 VT-1 Performed exams on 10 SRM/IRM dry tubes and identified two dry tubes with indications. SRM F was replaced during this outage and was one of the dry tubes with an indication. Indications documented under INR GGNS-IVVI-14-03.

Spring 2016 VT-1 Performed exams on 8 SRM/IRM and 5 LPRM dry tubes. One IRM and four LRPM dry tubes were replaced. No recordable indications were noted.

Instrument Fall 1996 VT-3 No indications.

Penetrations VessellD Fall 1996 VT-1 / VT-3 Section XI every 10 years on Attachment welds.

Brackets Other parts of brackets on general VT-3 exam. No indications.

Spring 2004 VT-1 Section XI Jet Pump attachment welds at two locations was inspected. No indications.

Page 6 of 9

GNRO-2017/00009 to Attachment Fall 2005 VT-1 / VT-3 Section XI GS Piping Brackets, FW Sparger End Brackets, Guide Rod Brackets (upper), Steam Dryer Brackets, Surveillance Sample Brackets and attachment welds at JP1112. Due to disassembly of the JP11 mixer an examination was performed at one Shroud Support Stub weld. No indications.

Spring 2007 VT-1 / VT-3 Section XI Jet Pump attachment welds at 5 locations. VT-3 of accessible areas of H9. No indications noted.

Spring 20.14 EVT-1 / VT-3 Inspections were performed four (4) Steam Dryer Support Bracket due to indications on the Steam Dryer Seismic Blocks. No discernible changes.

VessellD Spring 2016 VT-1/ VT-3/ Inspections were performed on Steam Dryer Support Brackets EVT-1 Brackets, Steam Dryer Hold-down Brackets, (continued) Feedwater Sparger Brackets and Surveillance Specimen Holder Brackets. No discernible changes.

LPCI Coupling Spring 1996 VT-1 VT-1 on LPGI @ Az. 141° due to a previous loose parts impact concern. No indications.

Spring 1998 EVT-1 All chosen welds on LPGI couplings @ Az. 39° and 141 0. No indications.

Fall 1999 VT-1 All accessible areas @ 219°. VT-1 on LPGI @ Az.

141° due to a previous loose parts impact concern.

No indications.

Spring 2001 VT-1 VT-1 on LPGI @ Az. 141° due to a previous loose parts impact concern. No indications.

Fall 2002 EVT-1 All accessible areas @ Az. 39°. No indications. VT-1 on LPGI @ Az. 141° due to a previous loose parts impact concern. No indications.

Fall 2005 VT-1 VT-1 on LPGI @ Az. 141° due to a previous loose parts impact concern. No indications.

Spring 2007 EVT-1 VT-1 on all accessible areas of LPGI @ 141°.

Extra weld was located on the strut assembly at all LPGI locations. No indications noted.

Fall 2008 EVT-1 EVT-1 performed on the extra welds (6-4a) that were noted during RF15 at each LPGI strut. No indications were noted.

Spring 2010 EVT-1 Exams were performed on LPGI @ 219°. No indications were noted.

Page 7 of 9

GNRO-2017/00009 to Attachment Spring 2016 EVT-1 Exams were performed on LPGI @ 39°. No indications were noted.

Steam Dryer Spring 2007 VT-1 Completed BWRVIP-139 examination. Cracked tack welds were noted on all (4) lifting lugs. No movement was noted. Eleven indications (IGSCC) were identified on the dryer upper support ring. No indications were longer than 3 Y2 ".

Fall 2008 VT-1 Examined areas identified during RF15. Additional crack was noted on a lifting lug and addition linear indication (1" Ig.) was identified on the Upper Support Ring.

Spring 2010 VT-1 Examined previous indications (cracked tack welds at lifting lugs and IGSCC cracking on the upper support ring). No changes were noted.

Steam Dryer Spring 2012 Various Steam Dryer replaced due to Extended Power (continued) Uprate Spring 2014 VT-1 Performed inspections on all accessible welds on the interior and exterior of the dryer. Indications were identified on Seismic Blocks, Tie Rod Bolting and Lower Guide Bracket. Indications documented under INR GGNS-IVVI-14-06 and INR GGNS-IVVI-14-07.

Spring 2016 VT-1 Performed inspections on all accessible welds on the interior and exterior of the dryer. Indications were identified on Seismic Support Blocks.

Indications documented under INR GGNS-IVVI 02.

Dissimilar Metal Spring 2007 UT N1A-KB Nozzle to Safe End Weld Welds on N2A-KB Nozzle to Safe End Weld Reactor Nozzles N2K-KB Nozzle to Safe End Weld K2M-KB Nozzle to Safe End Weld K2N-KB Nozzle to Safe End Weld K9A-KB Nozzle to Safe End Weld N9A-KG Safe End to Safe End Ext.

No recordable indications noted.

Fall 2008 UT N5B-KB Nozzle to Safe End Weld N5B-KG Safe End to Safe End Ext.

N4A-KB Nozzle to Safe End Weld N4F-KB Nozzle to Safe End Weld N4B-KB Nozzle to Safe End Weld No recordable indications.

Page 8 of 9

GNRO-2017/00009 to Attachment Spring 2010 UT N02B-KB Nozzle to Safe End Weld N02C-KB Nozzle to Safe End Weld N02D-KB Nozzle to Safe End Weld N02E-KB Nozzle to Safe End Weld N06A-KB Nozzle to Safe End Weld N06A-KC Safe End to Extension N09B-KB Nozzle to Safe End Weld No recordable indications.

Dissimilar Metal Spring 2012 UT N01B-KB Nozzle to Safe End Weld Welds on N02F-KB Nozzle to Safe End Weld Reactor Nozzles N02G-KB Nozzle to Safe End Weld (continued)

N02H-KB Nozzle to Safe End Weld N02J-KB Nozzle to Safe End Weld N04C-KB Nozzle to Safe End Weld N04D-KB Nozzle to Safe End Weld N05A-KB Nozzle to Safe End Weld N05A-KC Safe End to Extension N06B-KB Nozzle to Safe End Weld N06B-KC Safe End to Extension N06C-KB Nozzle to Safe End Weld N06C-KC Safe End to Extension N09A-KB Nozzle to Safe End Weld Crack was discovered in N06B-KB weld and weld overlay was completed satisfactorily. Ref. CR-GGN-2012-06386.

Spring 2016 UT N02M-KB Nozzle to Safe End Weld N02N-KB Nozzle to Safe End Weld N05B-KB Nozzle to Safe End Weld N05B-KC Safe End to Extension N06B-KB Weld Overlay N10-KC Safe End to Pipe Cap Weld No changes in original recordable indication for N06-KB Weld No other recordable indications Reactor Bottom Spring 2012 UT Two 2" drain lines were inspected per the Flow Drain Line Accelerated Corrosion Program. Degradation documented in Calculation MC-Q1111-12004. Next scheduled inspections are RF23 and RF29, res ectivel .

Lower Plenum Spring 2016 VT-3 Lower plenum inspections performed with Cell 32-33 and 28-29 removed. No recordable indications.

Page 9 of 9