GNRO-2017/00034, Proposed Alternative Request for Forth Interval Inservice Inspection Program

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Proposed Alternative Request for Forth Interval Inservice Inspection Program
ML17145A321
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 05/25/2017
From: Nadeau J
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2017/00034
Download: ML17145A321 (20)
v • d • e


Text

~Entergy Entergy Operations, Inc.

P. O. Box 756 Port Gibson, MS 39150 James Nadeau Regulatory Assurance Manager Tel. (601) 437-2103 GNRO-2017/00034 May 25,2017 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Proposed Alternative Request for Forth Interval Inservice Inspection Program 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 (a) (z) (1), Entergy hereby requests an alternative for Grand Gulf Nuclear Station (GGNS) Inservice Inspection Program. This request is applicable to the upcoming 4 th ten-year inservice inspection interval starting December 1, 2017 and ending on November 1, 2024. The details of the 10 CFR 50.55a request are provided in the attachment.

The proposed relief request alternative is for performing the ASME Code required examinations on 100 % of the nozzle-to-vessel welds and nozzle inner radius sections using ASME Code Case N-702. Entergy request NRC Staff review and approval of this proposed GGNS alternative on or before March 23, 2018.

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.

Sincerely, ~ft?~

James Nade~ -/ ec Regulatory Assurance Manager IN/sgd

GNRO-20 17/00034 Page 2 of 2 Attachments:

1. Proposed Alternative Request (GG-ISI-021) for Fourth Interval Inservice Inspection Program
2. Attachment 1 to Relief GG-ISI-021, Responses to BWRVIP - 108, Plant-Specific Applicability Criteria 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 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: marv.currier@msdh.ms.gov to GNRO-2017/00034 Relief Request Number GG-ISI-021

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

1. ASME Code Component(s) Affected Components/Numbers: 1. Reactor Pressure Vessel (RPV) Nozzle-to-Vessel Welds (see Table 1)
2. RPV Nozzle Inside Radius Sections (see Table 2)

Code Class: American Society of Mechanical Engineers (ASME)Section XI Code Class 1 Examination Category: 8-0, Full Penetration Welds of Nozzles in Vessels

==

Description:==

1. RPV Nozzle-to-Vessel Welds
2. RPV Nozzle Inside Radius Sections Item Number(s): 83.90 and 83.100 GGNS, 4th 10-year lSI Interval starting December 1, 2017 and Unit / Inspection Interval ending on November 1, 2024, which is the end of the initial GGNS Applicability: 40-year operating license (License No. NPF-29) that corresponds to the conditional failure probability for 40 years (Ref. 6).
2. Applicable Code Requirement(s)

ASME Section XI, 2007 Edition through the 2008 Addenda (Reference 1), Table IW8-2500-1 , Examination Category 8-0, Full Penetration Welded Nozzles in Vessels, (nozzle assemblies) requires a volumetric examination of all RPV nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles, but excludes manways and handholes either welded to or integrally cast in the vessel and examination of each nozzle's Inside radius section is required each 1O-year interval. The examination of these nozzle assemblies are detailed by the item numbers shown below.

  • Item No. 83.90 - Requires a volumetric examination of Reactor Vessel Nozzle-to-Vessel Welds.
  • Item No. 83.100 - Requires a volumetric examination of Reactor Vessel Nozzle Inside Radius Sections.

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (151) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR SO.SSa(z)(1)

- Acceptable Level of Quality and Safety -

Additionally, for the ultrasonic examinations performed on these nozzle assemblies, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is to be implemented; with the 2007 Edition through the 2008 Addenda of Section XI.

3. Reason for Request

rd For the 3 10-year lSI Interval a minimum of 25% of the Class 1 nozzle assemblies per system and nominal pipe size were examined [except for the Feedwater (N04) and Control Rod Drive (N1 0) nozzle assemblies that are excluded from the Case]. These examinations were performed in accordance with ASME Code Case N-702 (Reference 2), based on authorization of the 3 rd Interval request GG-ISI-013 (Reference 3) by NRC Safety Evaluation Report (Reference 4) to the end of the 3rd Interval, which is November 30, 2017.

The Federal Register Notice (FRN) published November 5, 2014, contains the rulemaking that amends 10 CFR 50.55a to incorporate by reference Regulatory Guide (RG) 1.147, Revision 17, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1." In order to continue the same 25% sampling provisions of Code Case N-702 for the 4th Interval until the end of the current 40-year operating license this request is required based on the NRC condition to use Code Case N-702 in NRC Regulatory Guide 1.147, Revision 17, (Reference 5). See Section 5 below for the condition required to use Code Case N-702.

4. Proposed Alternative Pursuant to 10 CFR 50.55a(a)(z)(1), Entergy requests an alternative from performing the ASME Code required examinations on 100% of the nozzle-to-vessel welds and nozzle inner radius sections identified in Tables 1 and 2, respectively. Specifically, Entergy proposes to adopt ASME Code Case N-702, which allows examination of a minimum of 25% of the nozzle-to-vessel welds and nozzle inside radius sections, including at least one nozzle from each system and nominal pipe size during the 4th inspection interval. For each of the identified nozzle assemblies in Table 3, both the nozzle-to-vessel weld and the inside radius section for that scheduled nozzle assembly will be examined. During this time frame, between the start of the 4th Interval and the end of the 40-year initial 40 year operating license, examinations will be distributed and scheduled as required by IWB-2400 and Table IWB-2411 to the extent practical. Thus, this will include only scheduling and performing examinations for the applicable nozzle assemblies listed in Table 3 for the first and second inspection periods of the 4th Interval up to the end of the initial GGNS operating license on November 1, 2024.

To support this proposed alternative, results of previous examinations are shown in Tables 1 and 2 based on the 2nd Interval examinations performed in accordance with ASME Section XI and the 3rd Interval examinations of Alternative Request GG-ISI-013, where ASME Code Case N-702 was authorized for use in Reference 4. All the indications noted were recorded and evaluated in accordance with the ASME Section XI Code of Record for the 2nd and 3rd intervals and determined to be within the allowable limits of IWB-3000 and as such were NOT reportable. The actual UT examinations were performed in accordance with Appendix VIII. Based on sizing and evaluation all recordable indications listed were believed to be geometric and/or fabrication related.

In addition, none of the recorded indications required repair. Refer to Tables 1 and 2 below in the Exam Results column for the disposition of indications for each applicable nozzle assembly examined at GGNS.

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

ASME Section XI requires that all thirty five (35) nozzle assemblies listed in Tables 1 and 2 be examined during each lSI inspection interval. For the 2nd Interval ASME Section XI applied and 100% of the nozzle assemblies were examined in accordance with Table IW8-2500-1 Examination Category 8-D and for the 3 rd Interval a minimum of 25% of the nozzle assemblies for each system and pipe size were examined using ASME Code Case N-702 as authorized by Reference 4. Table 3 identifies those nozzle assemblies to be scheduled and examined during the 4th 10-year lSI Interval using this alternative request.

However, because ASME Code Case N-702 excludes the six Feedwater (N04) and one Control Rod Drive Return (N1 0) nozzle assemblies, only twenty eight (28) total nozzle assemblies are included in the scope of this request. The excluded nozzle assemblies are listed in Tables 1 and 2 and identified as N/A and will continue to meet the requirements of ASME Section XI, by performing 100% of these nozzle assemblies during the 4th Interval.

ASME Code Case N-702 also includes a provision that stipulates that a VT-1 visual examination method may be used in lieu of the volumetric examination method for the inside radius sections of (Item No. 83.100) listed in Table 2, but as of now all inside radius section examinations within the scope of this alternative request have been performed previously at GGNS using an ultrasonic examination method. With this option available in ASME Code Case N-702, GGNS may perform examinations on the inside radius sections listed in Table 2 under Code Case N-702 with either the VT-1 or the volumetric examination method for the 4th Interval.

Table 1: Nozzle-to-Vessel Weld Examinations Item Comp.ID Item Description Exam Results Number 2 nd Interval (1) 3 rd Interval (2) 83.90 N01A-KA 24" RPV Outlet Nozzle to NRI (3) Not Required (5)

Vessel 83.90 N018-KA 24" RPV Outlet Nozzle to NRI (3) NRI (3)

Vessel 83.90 N02A-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N028-KA 12" RCS Inlet Nozzle to Vessel NRI (3) NRI (3) 83.90 N02C-KA 12" RCS Inlet Nozzle to Vessel NRI (3) NRI (3) 83.90 N02D-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02E-KA 12" RCS Inlet Nozzle to Vessel Spot Indications NRI (3) in the base metal recorded. No indication of growth from previously recorded examinations.

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR SO.SSa(z)(1)

- Acceptable Level of Quality and Safety-Table 1: Nozzle-to-Vessel Weld Examinations Item Comp.ID Item Description Exam Results Number 2 nd Interval (1) 3rd Interval (2) 83.90 N02F-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02G-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02H-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02J-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02K-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02M-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N02N-KA 12" RCS Inlet Nozzle to Vessel NRI (3) Not Required (5) 83.90 N03A-KA 24" Main Steam Nozzle to Spot Indications NRI (3)

Vessel in the base metal recorded. No indication of growth from previously recorded examinations.

83.90 N038-KA 24" Main Steam Nozzle to 8ase Metal NRI (3)

Vessel Indications Examined, but recorded. No not counted for indication of the required growth from 25%.

previously recorded examinations.

83.90 N03C-KA 24" Main Steam Nozzle to NRI (3) Not Required (5)

Vessel 83.90 N03D-KA 24" Main Steam Nozzle to Nozzle Cladding Not Required (5)

Vessel Cut Back Geometry Indication.

83.90 N04A-KA 14" Feedwater Nozzle to Vessel N/A (4) N/A (4) 83.90 N048-KA 14" Feedwater Nozzle to Vessel N/A (4) N/A (4) 83.90 N04C-KA 14" Feedwater Nozzle to Vessel N/A (4) N/A (4)

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Table 1: Nozzle-to-Vessel Weld Examinations Item Comp.ID Item Description Exam Results Number 2 nd Interval (1) 3 rd Interval (2)

B3.90 N04D-KA 11 14 Feedwater Nozzle to Vessel N/A (4) N/A (4)

B3.90 N04E-KA 11 14 Feedwater Nozzle to Vessel N/A (4) N/A (4)

B3.90 N04F-KA 11 14 Feedwater Nozzle to Vessel N/A (4) N/A (4)

B3.90 N05A-KA 11 12 Core Spray Nozzle to Geometric NRI (3)

Vessel Indication due to Core Spray Bracket Pad Buildup and Nozzle Cladding Strip Backing.

No indication of growth from previously recorded examinations.

B3.90 N05B-KA 11 12 Core Spray Nozzle to Geometric Not Required (5)

Vessel Indication due to Core Spray Bracket Pad Buildup and Nozzle Cladding Strip Backing.

No indication of growth from previously recorded examinations.

B3.90 N06A-KA 11 12 RHR/LPCI Inlet Nozzle to NRI (3) NRI (3)

Vessel B3.90 N06B-KA 11 12 RHR/LPCI Inlet Nozzle to NRI (3) Not Required (5)

Vessel B3.90 N06C-KA 11 12 RHR/LPCI Inlet Nozzle to NRI (3) Not Required (5)

Vessel B3.90 N07-KA 11 15.5 RCIC Top Head Spray NRI (3) NRI (3)

Inlet Nozzle

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Table 1: Nozzle-to-Vessel Weld Examinations Item Comp.ID Item Description Exam Results Number 2 nd Interval (1) 3rd Interval (2) 83.90 N08-KA 15.5" RCIC Top Head Spray NRI (3) NRI (3)

Spare Inlet Nozzle 83.90 N09A-KA 4" Jet Pump Instrument Nozzle Inner Diameter Not Required (5) to Vessel (ID) cladding cut-back and nozzle bore geometry indications.

83.90 N098-KA 4" Jet Pump Instrument Nozzle One Laminar NRI (3) to Vessel reflector in the base metal recorded. 10 cladding cut-back and nozzle bore geometry indications recorded.

83.90 N10-KA 4" CRD Return Nozzle to Vessel N/A (4) N/A (4) 83.90 N16-KA 8" Vibration Instrument Nozzle NRI (3) NRI (3) to Vessel Notes:

(1) Examinations were performed on 100% of these Nozzle-to-Vessel Welds for the 2 nd Interval.

(2) Examinations were performed on a minimum of 25% of these Nozzle-to-Vessel Welds by system and nominal pipe size for the 3 rd Interval.

(3) No Recordable Indications.

(4) Not Applicable to this request, ASME Code Case N-702 excludes the Feed water (N04) and Control Rod Drive Return (N10) nozzle assemblies. Therefore, these are not within the scope of this request and will continue to meet the requirements of ASME Section XI Code Table IW8-2500-1 by performing 100% examination of these nozzle assemblies during the 4th Interval.

(1) Not required to be selected for examination to meet the 25% sample per Code Case N-702.

Table 2: Nozzle Inside Radius Section Examinations

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR SO.SSa(z)(l)

- Acceptable Level of auality and Safety -

Item Comp.ID Item Description Exam Results Number 2nd Interval (1) 3rd Interval (2) 83.100 N01A-KA 24" RPV Outlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N018-KA 24" RPV Outlet Nozzle Inside NRI (3~ NRI (3)

Radius 83.100 N02A-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N028-KA 12" RCS Inlet Nozzle Inside NRI (3) NRI (3)

Radius 83.100 N02C-KA 12" RCS Inlet Nozzle Inside NRI (3) NRI (3)

Radius 83.100 N02D-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02E-KA 12" RCS Inlet Nozzle Inside NRI (3) NRI (3)

Radius 83.100 N02F-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02G-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02H-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02J-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02K-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02M-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N02N-KA 12" RCS Inlet Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N03A-KA 24" Main Steam Nozzle Inside NRI (3) NRI (3)

Radius 83.100 N038-KA 24" Main Steam Nozzle Inside NRI (3) NRI (3)

Radius Examined, but not counted for

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR SO.SSa(z)(1)

- Acceptable Level of Quality and Safety -

Table 2: Nozzle Inside Radius Section Examinations Item Comp.ID Item Description Exam Results Number 2 nd Interval (1) 3 rd Interval (2) the required 25%.

83.100 N03C-KA 11 24 Main Steam Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N03D-KA 11 24 Main Steam Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N04A-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N048-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N04C-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N04D-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N04E-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N04F-KA 11 14 Feedwater Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N05A-KA 11 12 Core Spray Nozzle Inside Geometric NRI (3)

Radius reflectors recorded.

83.100 N058-KA 11 12 Core Spray Nozzle Inside NRI (3) Not Required (5)

Radius 83.100 N06A-KA 11 12 RHR/LPCI Inlet Nozzle NRI (3) NRI (3)

Inside Radius 83.100 N068-KA 12 RHR/LPCI Inlet Nozzle 11 NRI (3) Not Required (5)

Inside Radius 83.100 N06C-KA 11 12 RHR/LPCI Inlet Nozzle NRI (3) Not Required (5)

Inside Radius 83.100 N07-KA 11 15.5 RCIC Top Head Spray NRI (3) NRI (3)

Inlet Nozzle Inside Radius 83.100 N08-KA 11 15.5 RCIC Top Head Spray NRI (3) NRI (3)

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (151) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Table 2: Nozzle Inside Radius Section Examinations Item Comp.ID Item Description Exam Results Number 2nd Interval (1) 3rd Interval (2)

Spare Inlet Nozzle Inside Radius 83.100 N09A-KA 4" Jet Pump Instrument Nozzle NRI (3) Not Required (5)

Inside Radius 83.100 N098-KA 4" Jet Pump Instrument Nozzle NRI (3) NRI (3)

Inside Radius 83.100 N10-KA 4" CRD Return Nozzle Inside N/A (4) N/A (4)

Radius 83.100 N16-KA 8" Vibration Instrument Nozzle NRI (3) NRI (3)

Inside Radius Notes:

(1) Examinations were performed on 100 % of these Nozzle Inside Radius Sections for the nd 2 Interval.

(2) Examinations were performed on a minimum of 25% of these Nozzle Inside Radius rd Sections by system and nominal pipe size for the 3 Interval.

(3) No Recordable Indications.

(4) Not Applicable to this request, ASME Code Case N-702 excludes the Feed water (N04) and Control Rod Drive Return (N1 0) nozzle assemblies. Therefore, these are not within the scope of this request and will continue to meet the requirements of ASME Section XI by performing 100 % examination of these nozzle assemblies during the 4th Interval.

(5) Not required to be selected for examination to meet the 25% sample per Code Case N-702.

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Table 3: Nozzle Assemblies to be Examined for the 4th Interval Group Description of Nozzle Assemblies Quantity Minimum Examination N01 24" Recirculation Outlet 2 1 N02 12" Recirculation Inlet 12 3 N03 24" Main Steam Line 4 4 1 N05 12" Core Spray 2 1 N06 12" RHR/LPCllnlet 3 1 N07 15.5" RCIC Top Head Spray Inlet Nozzle 1 1 N08 15.5" RCIC Top Head Spray Inlet Spare Nozzle 1 1 N09 4" Jet Pump Instrumentation 2 1 N16 8" Vibration Instrumentation 1 1 Note: ASME Code Case N-702 excludes the Feed water (N04) and Control Rod Drive Return (N10) nozzle assemblies. Therefore, these are not listed here because they are not within the scope of this request. These nozzle assemblies will continue to meet the requirements of ASME Section XI Code Table IWB-2500-1, by performing 100% examination of these nozzle assemblies during the 4th Interval.

5. Basis for Use Currently, ASME Code Case N-702 is listed in NRC Regulatory Guide 1.147, Revision 17, (Reference 5) as conditionally acceptable for use. The condition reads as follows:

"The technical basis supporting the implementation of this Code Case is addressed by BWRVIP-1 08: BWR Vessel and Internals Project, "Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, October 2002 (ML-023330203) and BWRVIP-241: BWR Vessel and Internals Project, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, October 2010 (ML11119A041). The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 1 dated Apri/19, 2013 (ML13071A240) are met. The evaluation 1 GGNS meets the demonstration criteria in Section 5.0 of the NRC Safety Evaluation of BWRVIP-1 08 and does not need to use BWRVIP-241.

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of auality and Safety -

demonstrating the applicabilitv of the Code Case shall be reviewed and approved bv the NRC prior to the application of the Code Case."

GGNS has chosen to use the criteria in Section 5.0 of the BWRVI P-108 as allowed above for demonstrating the applicability of ASME Code Case N-702 with the background provided below.

Electric Power Research Institute (EPRI) Technical Report 1003557, BWR Vessel and Internals Project (BWRVIP), Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, (Reference 6) provides the basis for ASME Code Case N-702. The evaluation found that the failure probability due to a Low Temperature Overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld is very low (i.e., < 1 X 10-6 for 40 years) with or without lSI and this evaluation is only applicable to the initial 40 years of operation for GGNS. The report concludes that inspection of 25% of each nozzle type is technically justified.

BWRVIP-108 was originally submitted to the NRC for review and approval via BWRVIP letter 2002-323 on November 25,2002 (Reference 7), supplemented by Tennessee Valley Authority (TVA) letter dated November 15,2004 (Reference 8), and BWRVIP letters dated July 25,2006, and September 13, 2007 (References 9 and 10).

On December 19, 2007, the NRC issued a Safety Evaluation Report (SER) approving the use of BWRVIP-1 08 (Reference 11). Section 5.0 of the SER states that each licensee should, within the request for alternative, demonstrate the plant meets the specified criteria and the demonstration is contained in Attachment 1.

6. Conclusion This proposed alternative to use ASME Code Case N-702 in lieu of the ASME Code,Section XI requirements provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) for all applicable RPV nozzle-to-vessel welds and nozzle inside radius sections excluding the N04 and N10 nozzles listed in Tables 1 and 2, and for those nozzle assemblies to be examined and scheduled in Table 3.
7. Duration of Proposed Alternative Upon authorization by NRC, this request for an alternative to use ASME Code Case N-702 will be implemented during the 4th 10-year lSI Interval beginning on December 1, 2017 and ending on November 1, 2024, which corresponds to the 40 years of initial operation that the BWRVI P-108 (Ref. 6) evaluation is applicable to for GGNS.
8. Precedents The NRC Staff has authorized similar requests based on meeting the demonstration criteria of BWRVIP-1 08 as required for the use of Code Case N-702 including the last authorized request GG-ISI-013 to use this Code Case at GGNS for the 3 rd 10-year lSI Interval, which is Reference 4, and the precedents listed below:

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (lSI) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of auality and Safety -

  • Cooper Nuclear Station, Docket No. 50-298, "Cooper Nuclear Station -Request For Relief No. Ri-04 for The Fourth 10-Year Inservice Inspection Interval Regarding Inspection of Reactor Vessel Nozzle-To-Vessel Shell Welds, NRC SER (TAC No. ME3319)", October 8,2010 (ADAMS Accession No. ML100470703)
9. References
1. ASME Section XI, 2007 Edition through the 2008 Addenda
2. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1,"

February 20, 2004

3. Request for Alternative GG-ISI-013, Proposed Alternative to 10 CFR 50,55a Examination Requirements for Reactor Pressure Vessel Weld Inspections, Docket No. 50-416, License No, NPF-29, Dated: April 6, 2011, (ADAMS Accession No. ML110960459)
4. Grand Gulf Nuclear Station, Unit 1 - Relief Request GG-ISI-013 Based On Code Case N-702, NRC SER (TAC No. ME5990), Dated: November 4,2011, (ADAMS Accession No. ML112710328)
5. NRC Regulatory Guide 1.147, Revision 17, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Dated: August 2014, (ADAMS Accession No. ML13339A689)
6. EPRI Technical Report 1003557, BWRVIP-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Bend Radii, Dated: October 2002
7. Letter from the BWRVIP to the NRC (2002-323), Project No. 704 - BWRVIP-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water

GNRO-2017/00034 Entergy Nuclear Operations, Inc Grand Gulf Nuclear Station (GGNS), Unit 1 Fourth Interval Inservice Inspection (151) Program Request No. GG-ISI-021 Proposed Alternative Request in Accordance with 10 CFR SO.SSa(z)(1)

- Acceptable Level of Quality and Safety -

Reactor Nozzle-to- Vessel Shell Welds and Nozzle Bend Radii, Dated: November 25, 2002, (ADAMS Accession No. ML023330203)

8. Letter from TVA to the NRC, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 American Society of Mechanical Engineers (ASME)Section XI, Inservice Inspection (151) Program - Requests for Relief 2-15/-22, and 3-/5/-18 for Examination of Reactor Pressure Vessel (RPV) Nozzle-to- Vessel Shell Welds and Nozzle Inner Radius Sections - Response to NRC request for Additional Information (RAI), (TAC Nos. MC0167 and MC0168), Dated: November 15, 2004, (ADAMS Accession No. ML043380321)
9. Letter from the BWRVIP to the NRC (2006-349), Project 704 - BWRVIP Response to NRC Request for Additional Information on BWRVIP-108, Dated: July 25,2006, (ADAMS Accession No. ML062080159)
10. Letter from the BWRVIP to the NRC (2007-268), Project 704 - Supplemental Analyses Supporting BWRVIP-108, Dated: September 13, 2007, (ADAMS Accession No. ML072600167)
11. Letter from/the NRC to the BWRVIP, Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP- 108)," Dated: December 19, 2007, (ADAMS Accession No. ML073600374)
12. GGNS Calculation MC-Q1 B13-10002, Reduction of Inspection Requirements on Recirculation Inlet and Outlet Nozzles, Rev. o.

Attachment 2 to GNRO-2017/00034 Attachment 1 to Relief GG-ISI-021, Responses to BWRVIP - 108, Plant-Specific Applicability Criteria

GNRO-2017/00034 Attachment 1 to Relief GG-ISI-021, Responses to BWRVIP - 108, Plant-Specific Applicability Criteria Section 5.0 of the NRC's SER that approved the use of BWRVIP-108 (Reference 11) states in part:

"Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-1 08 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific application of the BWRVIP-1 08 report to their units in the relief request by showing that all the following general and nozzle-specific criteria are satisfied:"

GGNS applicability was evaluated and demonstrated as documented in GGNS Calculation MC-Q1 B13-10002, Reduction of Inspection Requirements on Recirculation Inlet and Outlet Nozzles, Rev. 0 (Reference 12). This demonstration is summarized below.

Applicable GGNS Terms and Parameters

  • RPV constant for inlet nozzle, CNozzle3 =1,637 psi
  • RPV constant for outlet nozzle, CRPV4 =16,171 psi
  • RPV constant for outlet nozzle, CNozzle5 = 1,977 psi
  • Reactor core pressure = 1,056 psig or 1,070.696 psia
  • Maximum RPV radius = 126.75 inches
  • RPV wall thickness at Nozzle N2, t2 = 7.4479 [(AvgLong + AvgTrans) / 2J
  • RPV wall thickness at Nozzle N1, t4 = 7.4948 inches [(AvgLong + AvgTrans) 12J
  • Nozzle inner radius for outlet nozzles, rj = 10.84375 inches
  • Nozzle outer radius for outlet nozzles, ro=20.03125 inches
  • Nozzle inner radius for inlet nozzles, rj= 5.8125 inches
  • Nozzle outer radius for inlet nozzles, ro= 13.125 inches

GNRO-2017/00034 Attachment 1 to Relief GG-ISI-021, Responses to BWRVIP - 108, Plant-Specific Applicability Criteria General Criterion Criterion 1 The maximum RPV Heat/Cool down rate is limited to less than 115°F/hour.

GGNS Compliance Maximum heat up and cool down rates are limited to :5 100°F in anyone-hour period, in accordance with GGNS Technical Specification Surveillance Requirement 3.4.11.1.

Criteria for the Recirculation Inlet Nozzles Criterion 2 GGNS Compliance (pxr) = (1056 psix126.75 in) = 0.9296 < Ll S CRPv i 2 19332(7.4479 in)

Criterion 3 GGNS Compliance Criteria for the Recirculation Outlet Nozzles Criterion 4

GNRO-2017/00034 Attachment 1 to Relief GG-ISI-021, Responses to BWRVIP - 108, Plant-Specific Applicability Criteria (pxr) <1.15 CRPV4/4 GGNS Compliance (p x r) = (1056 psi x 126.75 in) = 1.104 < 1.15 CRflV4/4 16171(7.4948)

Criterion 5 GGNS Compliance p x (r} + r;2) _ ---r--~----'--------';'----~--L.

1056 psi x (20.03125)2 + (10.84375)2

= 0.97698 < 1.15 Cllozzle5 (r 2 o - r/) - 1977 (20.03125)2 - (10.84375)2 The results of the above equations demonstrate the applicability of BWRVIP-1 08 to GGNS by showing the five criteria specified in Section 5.0 of the NRC SER are met.

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