JAFP-16-0136, Alternative Examination Requirements Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241
| ML16238A004 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/24/2016 |
| From: | Drews W Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-16-0136 | |
| Download: ML16238A004 (10) | |
Text
h21Yflf Entergy Nuclear Northeast I LLUIJ Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 William C. Drews Regulatory Assurance Manager
- JAF JAFP-1 6-0136 August 24, 2016 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555
Subject:
Alternative Examination Requirements for James A. FitzPatrick Nuclear Power Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 James A. FitzPatrick Nuclear Power Plant Docket No.
50-333 License No.
Dear Sir or Madam:
Pursuant to 10 CFR 50.55a(z)(1), Entergy Nuclear Operations, Inc. (Entergy) requests NRC authorization to implement alternative examination requirements based on the American Society of Mechanical Engineers (ASME) Code Case N-702 and Boiling Water Reactor Vessel Inspection Program (BWRVIP)-241 as documented in the enclosed James A.
FitzPatrick Nuclear Power Plant (JAF) Inservice Inspection Program Relief Request (RR)-18.
The NRC provided a Safety Evaluation approving the generic technical bases and acceptability criteria for application of Code Case N-702 and BWRVIP-241, which Entergy has followed as detailed in the enclosure. Entergy requests approval of the proposed alternative on or before January 7, 2017 to accommodate application of this request during the next refueling outage.
Entergy plans to implement this alternative for the remainder of the fourth ISI interval. Although this review is neither exigent nor emergency, your prompt review is requested.
There are no new regulatory commitments made in this letter. Should you have any questions, please contact the Regulatory Assurance Manager, Mr. William C. Drews, at (315) 349-6562.
Very trulyyours, William C. Drews Regulatory Assurance Manager WCD:ds :
James A. FitzPatrick Nuclear Power Plant Inservice Inspection Program RR-1 8 cc:
USNRC, Regional Administrator, Region I
USNRC, Project Directorate USNRC, Resident Inspector
JAFP-J 6-0136 Enclosure I James A. FitzPatrick Nuclear Power Plant Inservice Inspection Program RR-18 (8 pages)
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18 1.
ASME Code Component(s) Affected Code Class:
ASME Section XI Code Class 1 Component Numbers:
N2 Code
References:
ASME Section Xl, 2001 Edition with 2003 Addenda ASME Code Cases 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.
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.
Examination Category:
B-D (Inspection Program B)
Item Number(s):
B3.90 and 83.100 Unit/Inspection James A. FitzPatrick Nuclear Power Plant (JAF) / Fourth (4th) 10-Interval:
year interval starting March 1, 2007, ending February 3, 2017.
2.
Applicable ASME Code Requirements 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, Performed Demonstration for Ultrasonic Examination Systems, is implemented; as required and modified by OCFR5O.55a(b)(2)(xv).
The subject components for this request for alternative examination requirements are the N2 Recirculation Inlet Nozzle-to-Vessel Welds (Items 83.90) and the N2 Recirculation Inlet Nozzle Inner Radius Sections (Item 83.100).
===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 exposure. JAF has estimated that the proposed reduction of inspection requirements would result in an approximate cost savings Page 1 of 5
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 70 CFR 50.55a(z)(7)
Fourth Interval ISI Program Relief Request No. RR-18 of $750,000 and reduction in worker dose of 4.5 Rem over the remainder of the current interval while providing an acceptable level of quality and safety.
===4.
Proposed Alternative and Basis for Use===
Pursuant to 10 CFR 50.55a(z)(1), an alternative is requested from performing the required examinations on 100% of the N2 Recirculation Inlet nozzles (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 nominal pipe size. JAF has a total often N2 Recirculation Inlet nozzle assemblies, all of nominal 12 pipe size. Fulfillment of the Code Case N-702 requirement will be accomplished via inspection of three N2 Recirculation Inlet nozzle assemblies. Five of the N2 Recirculation Inlet nozzle assemblies or 50% have been inspected during the current interval, with no recordable indications identified. Therefore, the Code Case N-702 requirements have been met.
JAF received NRC approval to utilize ASME Code Case N-702 for the fourth 10-year inservice inspection interval by letter dated October 17, 2012 (Relief Request No. 8; Reference 10).
The N2 Recirculation Inlet nozzles were excluded from the alternative associated with Relief Request No. 8 because they did not meet the third criterion specified in Section 5.0 of the staffs safety evaluation for the BWRVIP-108 report for plant-specific application of ASME Code Case N-702. On April 19, 2013, the NRC issued a safety evaluation (Reference 9) approving the use of BWRVIP-241, which contains relaxed criteria. As demonstrated herein, JAF meets the BWRVIP-241 criteria. Therefore, JAF is requesting NRC approval to apply ASME Code Case N-702 to the N2 Recirculation Inlet nozzles for the remainder of the fourth 10-year inservice inspection interval (ending February 3, 2017).
5.
Basis for Proposed Alternative In August of 2014, Revision 17 to Regulatory Guide (RG) 1.147 (Reference 11) was issued by the NRC. This revision added Code Case N-702 to Table 2: Conditionally Acceptable Section Xl Code Cases with the following condition:
The technical basis supporting the implementation of this Code Case is addressed by 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 fML-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 (ML119AO4). The applicability of Page 2 of 5
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18 Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-lOS dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (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.
The applicability of the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP 241 to the recirculation inlet nozzles at JAF is demonstrated as shown below:
The general terms used in the SE Section 5 applicability evaluations are:
CIRpv = recirculation inlet nozzles (from BWRVIP-OSNP model) = 19332 psi C1 NOZZLE = recirculation inlet nozzles (from BWRVIP-OSNP model) = 1637 psi The JAF 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 = 110.375 t = RPV wall thickness, t = 6.875 rIN2 inner radius for Recirculation Inlet N2 nozzles, rIN2 = 6.19 roN2 = outer radius for Recirculation Inlet N2 nozzles, roN2 = 10.22 (1) Max RPV Heatup / Cooldown Rate First 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)
Limits, the maximum RPV heatup / cooldown rate is limited to 100°F when averaged over any one hour period. JAF meets this criterion.
(2) Recirculation Inlet (N2) Nozzles Second criterion Equation: (pr/t) / CjRpv < 1.15
[(1040)(110.375)/6.875]/19332 = 0.864 < 1.15 The JAF result is 0.864, which meets the requirement of this criterion.
(3) Recirculation Inlet (N2) Nozzles Third criterion Equation: [p(roN22+rjN22)/( rON2 -rN2 )1 / CNoZzLE < 1.47
[1040 (10.222 + 6.192) / (10.222 6.192)] /1637 =1.371 < 1.47 The JAF result is 1.371, which meets the requirement of this criterion.
Page 3 of 5
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18 Criteria 4 and 5 relate solely to recirculation outlet (Ni) nozzles which were granted relief to utilize Code Case N-702 on October 17, 2012 as part of Relief Request No. 8 (Reference 10).
The NRC Safety Evaluation Section Criteria are met for all nozzles listed in Attachment 1.
Therefore, the basis for using Code Case N-702 is demonstrated for the JAF N2 Recirculation Inlet nozzles.
===5.
Duration of Proposed Alternative===
Upon approval by the NRC staff, this alternative will be utilized through the remainder of JAFs fourth inspection interval (March 1, 2007 February 3, 2017) for the N2 Recirculation Inlet nozzle assemblies (Attachment 1).
===6.
Precedents===
The NRC Staff has approved similar Requests for Alternative for the following plants:
1)
COLUMBIA GENERATING STATION
- REQUEST FOR ALTERNATIVE 3151-14 TO THE REQUIREMENTS OF THE ASME CODE (TAC NO. MF3435) dated February 13, 2015, ML15036A220 2)
PILGRIM NUCLEAR POWER PLANT
- RELIEF REQUEST PRR-24 REGARDING NOZZLE-TO-VESSEL WELDS AND NOZZLE INNER RADII EXAMINATIONS (TAC NO. MF4187) dated April 21, 2015, ML15103A069 Page 4 of 5
Entergy Nuclear Operations, Inc lames A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18
- 7. References 1)
ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, 2001 Edition with the 2003 Addenda.
2)
ASME Boiler and Pressure Vessel Code, Code Case N-702, Alternate 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 letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document Control Desk, Project No. 704-BWRVIP-1OSNP: 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.
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)
EPRI Technical Report 1021005, BWRVIP-241, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, dated October 2010.
7)
NRC Safety Evaluation of Proprietary EPRI Report, BWR Vessels 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.
8)
BWRVIP-1O8NP: BWR Vessel and Internals Project: Technical Basis forthe Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI Technical Report 1016123, November 2007.
9)
NRC SE of the Boiling Water Reactor Vessel Internals Project (BWRVIP) 241 Report, Probabilistic Fracture Mechanics for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii (TAC NO. ME623$) dated April 19, 2013.
- 10) James A. FitzPatrick Nuclear Power Plant Issuance of Relief From the Requirements of The American Society of Mechanical Engineers Boiler And Pressure Vessel Code (TAC No. ME7243) dated October 17, 2012.
- 11) Regulatory Guide 1.147, Revision 17: Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 dated August 2014.
Page 5 of 5
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with JO CFR 50.55a(z)(7)
Fourth Interval ISI Program Relief Request No. RR-78 Table of ASME Code Components Affected at JAF Component Description Code Code Item Inspections Indications ID Category (note 6)
N-2A-IR 12 Recirc B-D B3.100 1990, 1998 NRI Inlet Nozzle to Inner Radius N-2A 12 Recirc B-D B3.90 1977, 1990, 1990 (note 2)
Inlet Nozzle 1998 1998 (note 4) to Vessel Weld N-2B-IR 12 Recirc B-D B3.100 1995, 1998, NRI Inlet Nozzle 2008 to Inner Radius N-2B 12 Recirc B-D 83.90 1995, 1998, NRI Inlet Nozzle 2008 to Vessel Weld N-2C-IR 12 Recirc B-D 83.100 1988, 2006 NRI Inlet Nozzle to Inner Radius N-2C 12 Recirc B-D 83.90 1978, 1989, NRI Inlet Nozzle 2006 to Vessel Weld N-2D-IR 12 Recirc B-D 83.100 1988, 2006 NRI Inlet Nozzle to Inner Radius N-2D 12 Recirc B-D B3.90 1929, 2006 1989 (note 1)
Inlet Nozzle to Vessel Weld N-2E-IR 12 Recirc B-D 83.100 1995, 1998, NRI Inlet Nozzle 2008 to Inner Radius Page 1 of 3
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18 Component Description Code Code Item Inspections Indications ID Category (note 6)
N-2E 12 Recirc B-D 83.90 1977, 1995, NRI Inlet Nozzle 1998, 2008 to Vessel Weld N-2F-IR 12 Recirc B-D 83.100 1988, 2002, NRI Inlet Nozzle 2008 to Inner Radius N-2F 12 Recirc B-D 83.90 1989, 2002, NRI Inlet Nozzle 2008 to Vessel Weld N-2G-IR 12 Recirc B-D B3.100 1988, 2006 NRI Inlet Nozzle to Inner Radius N-2G 12 Recirc B-D 83.90 1989, 2006 NRI Inlet Nozzle to Vessel Weld N-2H-IR 12 Recirc B-D 83.100 1995, 1998, NRI Inlet Nozzle 2012 to Inner Radius N-2H 12 Recirc B-D 83.90 1977, 1995, NRI Inlet Nozzle 1998, 2012 to Vessel Weld N-2J-lR 12 Recirc B-D 83.100 1988, 1992, NRI Inlet Nozzle 2006 to Inner Radius N-2J 12 Recirc B-D 83.90 1981, 1992, 1992 (note 3)
Inlet Nozzle 2006 to Vessel Weld Page 2 of 3
Entergy Nuclear Operations, Inc James A. FitzPatrick Nuclear Power Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
Fourth Interval ISI Program Relief Request No. RR-18 Component Description Code Code Item Inspections Indications ID Category (note 6)
N-2K-IR 12 Recirc B-D 83.100 1995, 1998, NRI Inlet Nozzle 2012 to Inner Radius N-2K 12 Recirc B-D 83.90 1995, 1998, 199$ (note 5)
Inlet Nozzle 2012 to Vessel Weld Notes:
1.
N-2D
- 1989, Laminar Reflector was evaluated in the base metal and was found acceptable under ASME XI.
2.
N-2A 1990, Laminar indication noted approximately 1 x 1 and evaluated as acceptable under ASME XI.
3.
N-2J 1992, Fabrication indications/reflectors identified in the pre-service inspection were also identified in the 1992 inspection.
4.
N-2A 1998, Typical of plate segregates and required no evaluation for acceptance
acceptable.
5.
N-2K
- 1998, Typical of plate segregates and required no evaluation for acceptance
acceptable.
6.
No Reportable Indications (NRI)
Page 3 of 3