JAFP-11-0112, Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP

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Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP
ML112770093
Person / Time
Site: Dresden, Perry, Pilgrim, Columbia, Duane Arnold, Clinton, FitzPatrick  Constellation icon.png
Issue date: 10/03/2011
From: Joseph Pechacek
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BWRVIP-108NP, JAFP-11-0112, RR-8, TAC MD8193, TAC MD8458, TAC MD9850, TAC ME0218, TAC ME0882, TAC ME0883, TAC ME3290
Download: ML112770093 (11)
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Text

JAFP-11-0112 October 3, 2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Relief Request (RR-8), Alternative Examination Requirements for Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Using American Society of Mechanical Engineers Code Case N-702 and BWRVIP-108NP.

James A. FitzPatrick Nuclear Power Plant (JAF)

Docket No. 50-333 License No. DPR-59

REFERENCES:

1)

ASME 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.

2)

BWRVIP-108NP: 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.

Dear Sir or Madam,

Pursuant to 10CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Entergy) requests NRC approval of Relief Request (RR-8), Alternative Examination Requirements for JAF Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections using the requirements of ASME Code Case N-702 and BWRVIP-108NP provisions (References 1 and 2).

ASME Section XI, Table IWB-2500-1 requires 100% examination of nozzle-to-vessel shell welds and nozzle inside-radius sections during the current Inservice Inspection (ISI) interval. JAF is in the 4th ISI interval, which began on March 1, 2007, and will end December 31, 2016. The Code of Record for JAF is ASME Section XI, 2001 Edition with the 2003 Addenda. The proposed alternative follows ASME Code Case N-702, which requires an examination of a minimum of 25% of the nozzle-to-vessel welds and nozzle inside radius sections. In addition, the code case requires at least one nozzle from each system and nominal pipe size. BWRVIP-108NP provides the technical basis for the reduction of examination requirements for BWR Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii.

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Joseph Pechacek Licensing Manager - JAF

JAFP-11-0112 Page 2 of 2 The proposed alternative provides an acceptable level of quality and safety because it follows the requirements of ASME Code Case N-702 and the NRC approved technical basis (BWRVIP-108NP) for the alternative examinations presented in the enclosed Relief Request (RR-8).

The NRC has previously approved the proposed alternative examination requirements based on the above justification for the following plants:

1)

Pilgrim Nuclear Power Station, Docket No. 50-293, TAC No. ME3290 / August 25,2010 2)

Duane Arnold Energy Center, Docket No. 50-331, TAC No. MD8193 / August 29, 2008 3)

Perry Nuclear Power Plant, Docket No. 50-440, TAC No. MD8458 / December 29, 2008 4)

Columbia Generating Station, Docket No. 50-397, TAC No. MD9850 / April 8,2009 5)

Clinton Power Station, Docket No. 50-461, TAC No. ME0218 / August 24,2009 6)

Dresden Nuclear Power Station, Docket Nos. 50-237 & 50-249, TAC Nos ME0882 &

ME0883/ November 3,2009 Entergy requests NRC approval of the proposed alternative pursuant to 10CFR 50.55a(a)(3)(i) by October 2012 with a 30 day implementation period.

There are no commitments made in this submittal.

If you have any questions or require additional information, please contact Mr. Joseph Pechac k, Licensing Manager, at 315-349-6766.

In't Joseph techace Licensin1 Manager

Enclosure:

Fourth Interval lSI Program JAF Relief Request No. RR-8 cc:

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of NRC Resident Inspector James A. Fitzpatrick Nuclear Power Plant P.O. Box 136 Lycoming, New York 13093 Mr. Francis J. Murray Jr., President New York State Energy Research and Development Authority 17 Columbia Circle Albany, New York 12203-6399 Mr. Bhalchandra Vaidya, Project Manager Plant Licensing Branch U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2A Washington, DC 20555 Mr. Paul Eddy New York State Department of Public Services 3 Empire State Plaza Albany, New York 12223-1350

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Enclosure to Entergy Letter Number JAFP-11-0112 James A. FitzPatrick Fourth Interval lSI Program, Relief Request No. RR-8 (9 pages including cover page)

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Page 1 of 8

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

1 Component Numbers:

N1, N3, N5, N8, CH Inst., CH Vent (See Attachment 1 for detailed list of components)

Code

References:

ASME Section XI, 2001 Edition with 2003 Addenda ASME Code Case N-702: Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1.

BWRVIP-108NP: 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.

Examination Category:

B-D (Inspection Program B)

Item Numbers:

B3.90 and B3.100

==

Description:==

Alternative to ASME Section XI, Table IWB-2500-1 Unit/Inspection Interval: James A. FitzPatrick (JAF) / Fourth (4th) 10-year interval starting March 1, 2007, ending December 31, 2016

2. Applicable Code Requirement

ASME Section XI, 2001 Edition with the 2003 Addenda (Reference 1), Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles In Vessels - Inspection Program B requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented; as required and modified by 10CFR 50.55a(b)(2)(xv). The subject components for this request for alternative examination requirements are the Reactor Pressure Vessel (RPV) nozzle-to-vessel welds (Item B3.90) and the RPV nozzle inside radius section (Item B3.100).

3. Reason for Request

The twenty-five percent sampling level stated in Code Case N-702 (Reference 2) provides a significant cost savings and reduction in worker dose. JAF has estimated that the proposed reduction of inspection requirements would result in an approximate cost savings of $22,444 and reduction in worker dose of 6.558 Rem over the remainder of the current interval.

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Page 2 of 8

4. Proposed Alternative Pursuant to 10CFR 50.55a(a)(3)(i), an alternative is requested from performing the required examinations on 100% of the identified nozzle assemblies listed in Attachment 1. As an alternative, incorporation of Code Case N-702 would require examination of a minimum, 25% of the nozzle-to-vessel welds and nozzle inner radius sections, including at least one nozzle from each system and nominal pipe size as shown in Table 4-1. For each of the identified nozzle assemblies in Table 4-1, both the inner radius region and the nozzle-to-shell weld either have already been examined or will be examined during the 4th interval.

Table 4-1 JAF Summary - Affected Components Group Nozzle Description Total Number Minimum Number to be Examined Comments N1 Recirculation Outlet 2

1 1 scheduled 2nd period N2 Recirculation Inlet 10 10 applicability criteria not met.

N3 Main Steam 4

1 1 completed 1st period N5 Core Spray 2

1 1 completed 1st period N8 Jet Pump Instrumentation 2

1 1 completed 1st period CH Inst.

Closure Head Instrumentation Nozzle 2

1 1 completed 1st period CH Vent Closure Head Vent Nozzle 1

1 1 scheduled 3rd period Note: Feedwater (N4) nozzles and CRD Return (N9) nozzle are outside the scope of Code Case N-702 and are excluded from this application.

Code Case N-702 stipulates that the VT-1 examination method may be used in lieu of the volumetric examination method for the inner radius sections (Item No. B3.100). JAF has adopted ASME Code Case N-648-1 Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles (Reference 3), with the provisions stipulated in Regulatory Guide 1.147, in the JAF Fourth Interval ISI Program Plan. Therefore, JAF may perform examinations on inner radius sections with either the VT-1 or the volumetric examination method.

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Page 3 of 8

5. Basis for Proposed Alternative Electric Power Research Institute (EPRI) Technical Report 1016123, "BWRVIP-108NP: 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 4) provides the technical basis for use of Code Case N-702. The BWRVIP evaluation found that failure probabilities due to a low temperature overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e. < 1X 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25% of each nozzle type is technically justified.

BWRVIP-108 (Reference 5) was originally submitted to the NRC for review and approval via BWRVIP Letter 2002-323 (Reference 6) on November 25, 2002. This report was supplemented by Tennessee Valley Authority (TVA) letter dated November 15, 2004, and BWRVIP letters dated July 25, 2006, and September 13, 2007.

BWRVIP-108NP was originally submitted to the NRC for review and approval via BWRVIP Letter 2007-349 (Reference 7) on November 21, 2007. The non-proprietary report BWRVIP-108NP superseded the proprietary version previously transmitted to the NRC as BWRVIP-108.

The content of the BWRVIP-108NP report is identical to that of the proprietary version. The BWRVIP is providing this non-proprietary version to facilitate use of the information in the report in public forums by the NRC and other organizations such as ASME.

On December 19, 2007, the NRC issued a Safety Evaluation (SE) (Reference 8) approving the use of BWRVIP-108. Section 5.0 of the SE identifies that each licensee should demonstrate the plant-specific applicability of the BWRVIP-108 report in their relief request.

The applicability of the BWRVIP-108NP report to JAF is demonstrated by showing that all the general and nozzle-specific criteria within Section 5.0 of the NRC Safety Evaluation are met.

The criteria are evaluated in attachment 2.

The general terms used in the SE Section 5.0 applicability evaluations are:

Ci RPV = recirculation inlet nozzles (from BWRVIP-108NP model) = 19332 psi Ci-NOZZLE = recirculation inlet nozzles (from BWRVIP-108NP model) = 1637 psi Co-RPV = recirculation outlet nozzles (from BWRVIP-108NP model) = 16171 psi Co-NOZZLE = recirculation outlet nozzles (from BWRVIP-108NP model) = 1977 psi The James A. FitzPatrick nozzle-specific terms to be used in the SE Section 5 applicability evaluations are as follows:

Heatup / Cooldown rate < 100° F/hr p = Reactor Pressure Vessel (RPV) normal operating pressure, p = 1040 psig r = RPV inner radius, r = 110.25" t = RPV wall thickness, t = 6.875" riN2 = inner radius for Recirculation Inlet N2 nozzles, riN2 = 6.0" roN2 = outer radius for Recirculation Inlet N2 nozzles, roN2 = 10.22" riNI = inner radius for Recirculation Outlet N1 nozzles, riN1 = 12.69" roNI = outer radius for Recirculation Outlet N1 nozzles, roNI =21.66"

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Page 4 of 8 The results of the equations in Attachment 2 demonstrate the applicability of the BWRVIP-108NP report to James A. FitzPatrick by showing the NRC Safety Evaluation Section 5.0 criteria are met for all nozzles listed in Table 4.1 with the exception of the JAF Recirculation system inlet (N2) nozzles.

Safety Evaluation Section 5.0 indicates that only the recirculation inlet and outlet nozzles criteria needs to be checked because the conditional probability of failure P(FlE), for other nozzles is an order of magnitude lower. The JAF N2 nozzles did not meet the SE third criterion and therefore Code Case N-702 would not be applied to the N2 nozzles.

Safety Evaluation Section 5.0 criteria and basis for using Code Case N-702 is demonstrated for James A. FitzPatrick for all nozzles listed in Attachment 1.

6. Duration of Proposed Alternative

Upon approval by the NRC staff, this relief request will be utilized through the remainder of James A. FitzPatrick's fourth inspection interval (March 1, 2007 - December 31, 2016) for the nozzle assemblies listed in Attachment 1.

7. Precedents

The NRC Staff has approved similar Requests for Alternative for the following plants:

1) Pilgrim Nuclear Power Station, Docket No. 50-293, TAC No. ME3290 / August 25, 2010
2) Duane Arnold Energy Center, Docket No. 50-331, TAC No. MD8193 / August 29, 2008
3) Perry Nuclear Power Plant, Docket No. 50-440, TAC No. MD8458 / December 29, 2008
4) Columbia Generating Station, Docket No. 50-397, TAC No. MD9850 / April 8, 2009
5) Clinton Power Station, Docket No. 50-461, TAC No. ME0218 / August 24, 2009
6) Dresden Nuclear Power Station, Docket Nos. 50-237 & 50-249, TAC Nos. ME0882 &

ME0883 / November 3, 2009

8. References
1) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, 2001 Edition with the 2003 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) ASME Boiler and Pressure Vessel Code, Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles, September 7, 2001.
4) BWRVIP-108NP: 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.

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program, Relief Request No. RR-8 Page 5 of 8

5) 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 Blend Radii, EPRI Technical Report 1003557, October 2002.
6) BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document Control Desk, "Project No. 704 - 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 Blend Radii," November 25, 2002.
7) BWRVIP letter 2007-349, Rick Libra, BWRVIP Chairman, to NRC Document Control Desk, "Project No. 704-BWRVIP-108NP: 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 21, 2007.
1) Letter from Matthew A. Mitchell (NRR), to Rick Libra, BWRVIP Chairman, 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.

Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program Relief Request No. RR-8 Page 6 of 8 Table of ASME Code Components Affected at JAF Component ID Description Code Category Code Item N-1A 28" Recirc Outlet Nozzle to Vessel Weld B-D B3.90 N-1A-IR 28" Recirc Outlet Nozzle Inner Radius B-D B3.100 N-1B 28" Recirc Outlet Nozzle to Vessel Weld B-D B3.90 N-1B-IR 28" Recirc Outlet Nozzle Inner Radius B-D B3.100 N-3A 24" Main Steam Nozzle to Vessel Weld B-D B3.90 N-3A-IR 24" Main Steam Nozzle Inner Radius B-D B3.100 N-3B 24" Main Steam Nozzle to Vessel Weld B-D B3.90 N-3B-IR 24" Main Steam Nozzle Inner Radius B-D B3.100 N-3C 24" Main Steam Nozzle to Vessel Weld B-D B3.90 N-3C-IR 24" Main Steam Nozzle Inner Radius B-D B3.100 N-3D 24" Main Steam Nozzle to Vessel Weld B-D B3.90 N-3D-IR 24" Main Steam Nozzle Inner Radius B-D B3.100 N-5A 10" Core Spray Nozzle to Vessel Weld B-D B3.90 N-5A-IR 10" Core Spray Nozzle Inner Radius B-D B3.100 N-5B 10" Core Spray Nozzle to Vessel Weld B-D B3.90 N-5B-IR 10" Core Spray Nozzle Inner Radius B-D B3.100 N-8A 4 Jet Pump Instrumentation Nozzle to Vessel Weld B-D B3.90 N-8A-IR 4 Jet Pump Instrumentation Nozzle Inner Radius B-D B3.100 N-8B 4 Jet Pump Instrumentation Nozzle to Vessel Weld B-D B3.90 N-8B-IR 4 Jet Pump Instrumentation Nozzle Inner Radius B-D B3.100 N-TH-A 5.75" CH Instrument Nozzle to Vessel Weld B-D B3.90 N-TH-A-IR 5.75" CH Instrument Nozzle Inner Radius B-D B3.100 N-TH-B 3.812" CH Vent Nozzle to Vessel Weld B-D B3.90 N-TH-B-IR 3.812" CH Vent Nozzle to Vessel Inner Radius B-D B3.100 N-TH-C 5.75" CH Instrument Nozzle to Vessel Weld B-D B3.90 N-TH-C-IR 5.75" CH Instrument Nozzle Inner Radius B-D B3.100 Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program Relief Request No. RR-8 Page 7 of 8 Response to NRC Plant-Specific Applicability of BWRVIP-108 Criteria Given the general and plant-specific terms, James A. FitzPatrick's conformance with the five (5) criteria is demonstrated as follows:

(1) Max RPV Heatup / Cooldown Rate 1st Criterion - the maximum RPV heatup / cooldown rate is limited to < 115°F/hr In accordance with Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) is limited to 100°F when averaged over any one hour period. JAF meets the requirement of Criterion 1.

(2) Recirculation Inlet (N2) Nozzles 2nd Criterion - Equation: (pr/t) / Ci-RPV < 1.15

[(1040)(110.25)/6.825]/19332 = 0.87 < 1.15 The JAF result is 0.87 and thus meets the requirement of Criterion 2 to be less than 1.15.

(3) Recirculation Inlet (N2) Nozzles 3rd Criterion - Equation: [p(roN2 2 + riN2

2) / (roN2 2-riN2 2)] / Ci-NOZZLE < 1.15

[1040 (10.222 + 62) / (10.222 - 62)] / 1637 = 1.303 < 1.15 The JAF result is 1.303 and thus does not meet the requirement of Criterion 3 to be less than 1.15.

(4) Recirculation Outlet (N1) Nozzles 4th Criterion - Equation: (pr/t) / Co-RPV < 1.15

[(1040)(110.25) / 6.825] / 16171 = 1.04 < 1.15 The JAF result is 1.04 and thus meets the requirement of Criterion 4 to be less than 1.15 (5) Recirculation Outlet (N1) Nozzles 5th Criterion - Equation: [p(r0N1 2 + riN1

2) + (r0N1 2 - riN1 2)] / Co-NOZZLE < 1.15

[1040(21.662 + 12.692) / (21.662 - 12.692)] / 1977 = 1.08 < 1.15 The JAF result is 1.08 and thus meets the requirement of Criterion 5 to be less than 1.15 Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative In Accordance with 10CFR 50.55a(a)(3)(i)

Fourth Interval ISI Program Relief Request No. RR-8 Page 8 of 8 JAF 4th INTERVAL OUTAGE SCHEDULE 4th Inspection Interval started March 1, 2007 and ends December 31, 2016.

PERIOD OUTAGE OUTAGE DATE 1

RFO18 September 2008 1

RFO19 September 2010 2

RFO20 September 2012 3

RFO21 September 2014 3

RFO22 September 2016

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