LR-N17-0066, Request for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations

Request for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations

ML17090A186
Person / Time
Site:	Hope Creek  (NPF-057)
Issue date:	03/31/2017
From:	David Mannai Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N17-0066, RR HC-14R-170
Download: ML17090A186 (15)

Text

l>SEG Nuclear Ll,C P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 LR-N 17-0066 MAR Sl 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354 PSlG NudearlLC 10 CFR 50.55a

Subject:

Request for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations In accordance with 10 CFR 50.55a, "Codes and standards, " PSEG Nuclear LLC (PSEG) hereby requests NRC approval of a proposed alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components, "

(ASME Section XI), for nozzle-to-vessel weld and nozzle inside radius examinations.

The details of the proposed alternative are provided in Attachment 1. The proposed alternative is consistent with ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds. " The NRC approved the generic technical basis for use of ASME Code Case N-702 in a safety evaluation on December 19, 2007 (ADAMS Accession No. ML073600374).

Applicability of the generic technical basis to Hope Creek is demonstrated in Appendix A to Attachment 1 of this letter.

PSEG requests approval of the proposed request by March 31, 2018, to support planning activities for Hope Creek refueling outage RF21 currently scheduled to begin in April 2018. Relief Request HC-14R-170 applies to the Hope Creek Fourth 1 0-year interval which is scheduled to begin on December 13, 2017 and is scheduled to end on December 31, 2026.

The Code of Record for the Hope Creek Fourth 1 0-year lnservice Inspection Interval is the ASME Section XI, 2007 Edition through the 2008 Addenda.

LR-N17-0066 Page 2 There are no regulatory commitments contained in this letter.

10 CFR 50.55a Should you have any questions concerning this matter, please contact Mr. Brian Thomas at 856-339-2022.

Sincerely, If!}

David J. Mannai Senior Director, Regulatory Operations PSEG Nuclear LLC :

10 CFR 50.55a Relief Request HC-14R-170 cc:

Administrator, Region I, NRC NRC Senior Resident Inspector, Hope Creek C. Parker, Project Manager, Hope Creek, USNRC P. Mulligan, Chief, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator T. MacEwen, Hope Creek Commitment Tracking Coordinator LR-N 17-0066 1 0 CFR 50.55a Relief Request HC-14R-170 Page 1 of 13 LR-N17-0066 ATTACHMENT 1 Hope Creek Nuclear Generating Station Renewed Facility Operating License No. NFP-57 NRC Docket No. 50-354 10 CFR 50.55a Request for Alternative HC-14R-170 Alternative with acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1)

1. ASME Code Component(s) Affected Code Class:

ASME Section XI Code Class 1 Component Numbers:

Code

References:

Various (see Table 1 for detailed list of components)

ASME Section XI, 2007 Edition with 2008 Addenda Examination Category:

Item Number(s):

Code Case N-702 B-D 83.90 and 83.100 The reactor pressure vessel (RPV) nozzle-to-vessel welds and inner radii subject to this request are listed below in Table 1:

TABLE 1 Identification Description Size Total Minimum Number Inches Number Number to be Examined N2 Recirculation Inlet 12 10 3

N3 Main Steam Outlet 26 4

1 N5 Core Spray 10 2

1 N6 Head Spray 6

2 1

N7 Head Vent 4

1 1

N8 Jet Pump Instrumentation 4

2 1

N17 Low Pressure Coolant Injection 12 4

1 (LP C I)

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME)Section XI, 2007 Edition through the 2008 Addenda, for the Fourth lnservice Inspection (lSI) 1 0-Year Interval scheduled to begin on December 13, 2017 and scheduled to end on December 31, 2026.

Page 2 of 13 LR-N17-0066

3. Applicable Code Requirement

ASME Section XI, 2007 Edition through the 2008 Addenda Table IWB-2500-1, "Examination Category 8-D, Full Penetration Welded Nozzles in Vessels, " requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 1 0-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems, " is implemented.

4. Reason for Request

NRC Regulatory Guide 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1, " conditionally accepts the use of Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1." This code case provides an alternative to performing examination of 100% of the Nozzle-to-Vessel Welds and Inner Radii for Examination Category 8-D nozzles with the exception of the feedwater and control rod drive return line nozzles. The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size.

5. Proposed Alternative and Basis for Use

Pursuant to 10 CFR 50.55a(z)(1 ), Hope Creek Generating Station (HCGS), requests approval to implement the alternative of Code Case N-702 in lieu of the code required 100% examination of all nozzles identified in Table 1. As an alternative, for the nozzle-to-shell welds and inner radii identified in Table 1, HCGS proposes to examine a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702.

The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of the NRC Safety Evaluation regarding BWRVIP-1 08 dated December 19, 2007 (ADAMS Accession No. ML073600374) or Section 5.0 of the NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ADAMS Accession No. ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case.

As permitted by Code Case N-702, a VT-1 visual examination of Item No. 83.100 may be performed in lieu of a volumetric examination. In the event HCGS elects to apply the alternative VT-1 examination, HCGS will meet NRC conditions specified for Code Case N-648-1 in Regulatory Guide 1.147, Revision 17 Table 2, as noted below:

Page 3 of 13 LR-N 17-0066 In lieu of a UT examination, licensees may perform a VT-1 examination in accordance with the code of record for the lnservice Inspection Program utilizing the allowable flaw length criteria of Table IWB-3512-1 with limiting assumptions on flaw aspect ratio.

The Boiling Water Reactor (BWR) Vessel Internals Project (BWRVIP), in association with Electric Power Research Institute (EPRI), has issued two topical reports:

BWRVIP-1 08NP, "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 1016123, November 2007 (ADAMS Accession No. ML073300050) and BWRVIP-241 NP, "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 (ADAMS Accession No. ML11119A043)

BWRVIP-108NP contains the technical basis supporting ASME Code Case N-702 for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 1 0-year interval. The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a low temperature overpressure event are very low (i.e., <1 x 1 o-6 for 40 years) with or without inservice inspection.

BWRVIP-241 NP provides supplemental analyses for BWR RPV recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 NP was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation for BWRVIP-108NP.

The analyses in BWRVIP-108NP and BWRVIP-241NP are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation. Based on review of industry operating experience, the Hope Creek LPCI nozzle fluence was evaluated to the end of the extended license, and the evaluation confirmed that the nozzles are below the failure criteria and meet the acceptable probability considering elevated fluence level per ASME Code Case N-702.

The analyses in BWRVIP-108 and BWRVIP-241NP were based on predicted fatigue crack growth over the initial licensed operating period (40 years) and assumed additional fatigue cycles in evaluating fatigue crack growth. HCGS is within the initial 40 year license through inspections performed in the 4th lSI interval and therefore is within the predicted fatigue crack growth assumed in BWRVIP-1 08NP and BWRVIP-241 NP.

Page 4 of 13 LR-N17-0066 Regulatory Guide 1.147, Revision 17 conditionally accepts the use of Code Case N-702 with the following condition:

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-1 08 dated December 1 [9], 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated Apri1 19, 2013 (ML13071A240) are met.

Section 5.0 of the NRC Safety Evaluation for BWRVIP-241 states:

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-241 report as the technical basis for use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:

1: The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour; For Recirculation Inlet (N2) nozzles:

2: (pr/t)/CRPV :5 1.15 where p = RPV normal operating pressure (psi),

r = RPV inner radius (inch),

t = RPV wall thickness (inch), and CRPV = 19332; where p = RPV normal operating pressure (psi),

ro = nozzle outer radius (inch),

n = nozzle inner radius (inch),

CNOZZLE = 1637; For Recirculation Outlet (N1) nozzles:

4: (pr/t)/CRPV :::; 1.15 where p = RPV normal operating pressure (psi),

r = RPV inner radius (inch),

t = RPV wall thickness (inch), and CRPV = 1 61 71 ;

Page 5 of 13 LR-N 17-0066 where p = RPV normal operating pressure (psi),

ro = nozzle outer radius (inch),

ri = nozzle inner radius (inch),

CNOZZLE = 1977; All HCGS RPV nozzle-to-vessel shell penetration welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles, meet the general and nozzle-specific criteria in BWRVIP-241. Therefore, ASME Code Case N-702 is applicable. See Appendix A for details.

The results of the last examination performed on each nozzle-to-vessel weld and nozzle inner radii are included in Appendix B.

ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) for the RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections listed in Table 1.

6.

Duration of Proposed Alternative This request for alternative is applicable to the HCGS Fourth 1 0-year ASME Section XI lSI Interval which begins on December 13, 2017 and scheduled to end December 31, 2026; however, approval is only requested until the end of the original (40-year) license which ends on April 11, 2026 (all required fourth interval examinations related to this relief request will be complete by the conclusion of the fall 2025 refueling outage).

7.

Precedents

1.

A similar request (PRR-50) was approved for Pilgrim Nuclear Power Station for the fifth 1 0-year lSI interval (ADAMS Accession No. ML15338A309) dated January 5, 2016

2.

A similar request (RI-08) was approved for Cooper Nuclear Station for the fifth 1 0-year lSI interval (ADAMS Accession No. ML15134A242) dated May 20, 2015

3.

A similar request (31SI-14) was approved for Columbia Generating Station for the third 1 0-year lSI interval (ADAMS Accession No. ML15036A220) dated February 13, 2015

4.

A similar request was approved for the Peach Bottom Atomic Power Station, Units 1 and 2 (ADAMS Accession No. ML112770217) January 24, 2012 Page 6 of 13 LR-N17-0066

5.

A similar request (ISI-23) was approved for Browns Ferry Nuclear Units 1, 2 and 3 (ADAMS Accession No. ML102440565) dated October 28, 2010

8. References

1.

ASME Section XI 2007 Edition, 2008 Addenda,Section XI of the ASME Boiler and Pressure Vessel Code

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 WeldsSection XI, Division 1,"February 20, 2004

3.

BWRVIP-1 08NP, "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, November 21, 2007 (ADAMS Accession No. ML073300050)

4.

BWRVIP-241 NP, "Probabilistic Fracture Mechanics Evaluation for the Boling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, October 2010 (ADAMS Accession No. ML11119A043)

5.

BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman to USNRC Document Control Desk, "Project No. 704-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,"

November 25, 2002 (ADAMS Accession No. ML023330203)

6.

Matthew A Mitchell, Office of Nuclear Reactor Regulation, to Rick Libra, BWRVIP Chairman, "Safety Evaluation of Proprietary EPRI Report, "BWRVIP 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-1 08)," December 19, 2007 (ADAMS Accession No. ML073600374)

7.

Hope Creek Technical Specifications 3/4.4.6 "Pressure Temperature Limits "

8.

Sher Bahadur, Office of Nuclear Reactor Regulation, to Dennis Madison, BWRVIP Chairman, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)-241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," April 19, 2013 (ADAMS Accession No. ML13071A240)

Page 7 of 13 LR-N 1 7-0066 Appendix A Plant Specific Applicability Hope Creek Nozzle-specific criteria:

Nozzle Operating RPV Inner RPVWall Pressure Radius Thickness (psig)

(inches)

(inches)

Recirc Inlet 1005.3 126.5 6.102 (N2)

Recirc Outlet 1005.3 126.5

6. 102 (N1)

Nozzle Inner Nozzle Outer Radius Radius (inches)

(inches) 5.75 12.40 12.78 22.64

1. The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour.

HCGS Technical Specification Limiting Condition for Operations 3.4.6.1, limits the maximum heatup and cooldown rates to less than or equal to 100°F in any one hour period and thus meets the requirement of Criterion 1.

For Recirculation Inlet (N2) nozzles:

2: (pr/t)/CRPV =1.08 < 1.15 where p = RPV normal operating pressure, 1005.3 psig r = RPV inner radius, 126.5 inches t = RPV wall thickness, 6.102 inches CRPV = 19332; 3: [p(ro2

+ r?) I (ro2 - r?)]/CNozzLE = 0.95 < 1.47 where p = RPV normal operating pressure, 1005.3 psig ro = nozzle outer radius, 12.40 inches n = nozzle inner radius, 5. 75 inches CNOZZLE = 1637; For Recirculation Outlet (N1) nozzles:

4: (pr/t)/CRPV =1.29 > 1.15 where p = RPV normal operating pressure, 1005.3 psig r = RPV inner radius, 126.5 inches Page 8 of 13 LR-N17-0066 t = RPV wall thickness, 6.102 inches CRPV = 16171; where p = RPV normal operating pressure, 1005.3 psig ro = nozzle outer radius, 22.64 inches n = nozzle inner radius, 12.78 inches CNOZZLE = 1977; Therefore, the Recirculation Outlet (N1) nozzles are not included in this relief request as they do not meet Criterion 4 under the NRC Safety Evaluation for BWRVIP-241.

Page 9 of 13 LR-N17-0066 Appendix B Results of Previous Examinations Page 10 of 13 LR-N 17-0066 Nozzle ID Nozzle-to-Vessel (NV)

Inner Radius (IR)

N2A (NV)

N2A (IR)

N28 (NV)

N28 (IR)

N2C (NV)

N2C (IR)

N2D (NV)

N2D (IR)

N2E (NV)

N2E(IR)

N2F (NV)

N2F(IR)

N2G (NV)

N2G (IR)

N2H (NV)

N2H (IR)

N2J (NV)

N2J (IR)

N2K(NV)

N2K (IR)

N3A (NV)

N3A (IR)

N38 (NV)

N38 (IR)

N3C (NV)

N3C (IR)

N3D (NV)

N3D (IR)

Categor:y Number 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D B-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D Examination Results Item System Number 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100

_Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100.

Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Recirculation (Inlet) 83.100 Recirculation (Inlet) 83.90 Main Steam 83.100 Main Steam 83.90 Main Steam 83.100 Main Steam 83.90 Main Steam 83.100 Main Steam 83.90 Main Steam 83.100 Main Steam Page 11 of 13 Nominal PiQe Last Size (Inches}

Examination 12 HCRF0-16 12 HCRF0-9l1>

12 HCRF0-9 12 HCRF0-9l1>

12 HCRF0-9 12 HCRF0-9 12 HCRF0-14 12 HCRF0-7tzJ 12 HCRF0-16 12 HCRF0-7tzJ 12 HCRF0-20 12 HCRF0-7tLJ 12 HCRF0-14 12 HCRF0-7tLJ 12 HCRF0-14 12 HCRF0-7tLJ 12 HCRF0-9 12 HCRF0-9 12 HCRF0-9 12 HCRF0-9t¸J 26 HCRF0-11 26 HCRF0-11 26 HCRF0-11 26 HCRF0-11l6>

26 HCRF0-11 26 HCRF0-11 26 HCRF0-19 26 HCRF0-11 Result

§_

NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI NRI I

AQQendix VIII I Exam 1

YES NA NA NA NA NA YES NA YES NA YES NA YES NA YES NA NA NA NA NA NA NA NA NA NA NA YES NA LR-N17-0066 N5A (NV)

N5A (IR)

N58 (NV)

N58 (IR)

N6A (NV)

N6A (IR)

N68 (NV)

N68 (IR)

N7 (NV)

N7 (IR)

N8A (NV)

N8A (IR)

N88 (NV)

N88 (IR)

N17A (NV)

N17A (IR)

N178 (NV)

N178 (IR)

N17C (NV)

N17C (IR)

N17D (NV)

N17D (IR) 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 8-D 83.90 Core Spray 83.100 Core Spray 83.90 Core Spray 83.100 Core Spray 83.90 Head Spray 83.100 Head Spray 83.90 Head Spray 83.100 Head Spray 83.90 Head Vent 83.100 Head Vent 83.90 Jet Pump Instrumentation 83.100 Jet Pump Instrumentation 83.90 Jet Pump Instrumentation 83.100 Jet Pump Instrumentation 83.90 Low Pressure Coolant Injection (LPCI) 83.100 Low Pressure Coolant Injection (LPCI) 83.90 Low Pressure Coolant Injection (LPCI)

'83.100 Low Pressure Coolant Injection (LPCI) 83.90 Low Pressure Coolant Injection (LPCI) 83.100 Low Pressure Coolant Injection (LPCI) 83.90 Low Pressure Coolant Injection (LPCI) 83.100 Low Pressure Coolant Injection (LPCI)

P?ge 12 of 13 10 HCRF0-9 NRI NA 10 HCRF0-9 NRI NA 10 HCRF0-16 NRI YES 10 HCRF0-9l1J NRI NA 6

HCRF0-16 NRI YES 6

HCRF0-19 NRI YES 6

HCRF0-12 NRI YES 6

HCRF0-12 NRI YES 4

HCRF0-19 NRI YES 4

HCRF0-19 NRI YES 4

HCRF0-19 NRI YES 4

HCRF0-19 NRI YES 4

HCRF0-12 NRI YES 4

HCRF0-12 NRI YES 12 HCRF0-12 NRI YES 12 HCRF0-4(5)

NRI NA 12 HCRF0-12 NRI YES 12 HCRF0-4(4)

NRI NA 12 HCRF0-20 NRI YES 12 HCRF0-4(4)

NRI NA 12 HCRF0-12 NRI YES 12 HCRF0-4(4)

NRI NA LR-N 17-0066 Notes:

(1) Nozzle Inner Radii last performed with UT in HCRF0-9 (2000), a Visual Examination was performed in HCRF0-17 using Code Case N-648-1 (2) Nozzle Inner Radii last performed with UT in HCRF0-7 (1997), a Visual Examination was performed in HCRF0-14 using Code Case N-648-1 (3) Nozzle Inner Radii last examined with UT in HCRF0-11 (2003), a Visual Examination was performed in HCRF0-18 using Code Case N-648-1.

(4 Nozzle Inner Radii last performed with UT in HCRF0-4 (1992), a Visual Examination was performed in HCRF0-14 using Code Case N-648-1.

(5) Nozzle Inner Radii last performed with UT in HCRF0-4 (1992), a Visual Examination was performed in HCRF0-20 using Code Case N-648-1.

N R I = No Recordable Indications H CRF0-4 (1992)

H CRF0-7 (1997)

H C RF0-9 (2000)

HCRF0-11 (2003)

H CRF0-12 (2004)

H CRF0-14 (2007)

H CRF0-16 (2010)

H CRF0-18 (2013)

H CRF0-19 (2015)

H CRF0-20 (2016)

Page 13 of 13