ML14077A175

Pilgrim, Request for Alternative - Alternative Examination Requirements for Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241

ML14077A175
Person / Time
Site:	Pilgrim
Issue date:	03/12/2014
From:	Lynch J R Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.14.028
Download: ML14077A175 (10)

Text

{{#Wiki_filter:' En tergyEntergy Nuclear Operations, Inc.600 Rocky Hill RoadPlymouth, MA 02360Pilgrim Nuclear Power StationLETTER NUMBER 2.14.028March 12, 2014U.S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555-0001

SUBJECT:

Request for Alternative -Alternative Examination Requirements for PilgrimNuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner Radii UsingASME Code Case N-702 and BWRVIP-241 Pilgrim Nuclear Power StationDocket No. 50-293License No. DPR-35

REFERENCES:

1. ASME Boiler and Pressure Vessel Code, Code Case N-702,"Alternative Requirements for Boiling Water Reactor (BWR) NozzleInner Radii and Nozzle-to-Vessel Shell Welds, Section Xl, Division 1",dated February 20, 20042. BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for theBoiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii",dated October 20103. NRC Letter to Entergy, "Relief Request PRR-20, Alternative Examination Requirements for Nozzle-to-Shell and Inner Radii WeldsUsing ASME Code Case N-702 and BWRVIP-108

-Pilgrim NuclearPower Station (TAC NO. ME3290)", dated August 25, 2010 (1.10.035)

4. NRC Letter to BWRVIP, "Final Safety Evaluations of the Boiling WaterReactor Vessel Internals Project (BWRVIP)-241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii (TAC NO. ME6328)",

dated April19, 2013

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Entergy) requests NRCauthorization to implement alternative examination requirements based on the AmericanSociety of Mechanical Engineers (ASME) Code Case N-702 (Reference

1) and Boiling WaterReactor Vessel Inspection Program (BWRVIP)

-241 (Reference

2) as documented in theenclosed Pilgrim Nuclear Power Station (PNPS) Relief Request (PRR) -24, "Alternative Examination Requirements for PNPS Nozzle-to Vessel Welds and Nozzle Inner Radii UsingASME Code Case N-702 and BWRVIP-241".

This template format is similar to the submittal the Nuclear Regulatory Commission (NRC) Staff has approved in a Safety Evaluation (Reference

3) for other PNPS Nozzle-to-Vessel Welds and Nozzle Inner Radii in PRR-20.

LETTER NUMBER 2.14.028Page 2 of 2In accordance with 10 CFR 50.55a(a)(3)(i), the proposed alternative to the referenced requirements may be approved by the NRC provided an acceptable level of quality and safetyare maintained. Entergy believes the proposed alternative provides an acceptable level ofquality and safety as well as a significant reduction in worker radiation exposure. The NRCprovided a Safety Evaluation (Reference

4) approving the generic technical bases andacceptability criteria for application of Code Case N-702 and BWRVIP-241, which Entergyhas followed as detailed in the enclosure.

Entergy requests approval of the proposed alternative on or before April 1, 2015 toaccommodate application of this request during the next refueling outage. Entergy plans toimplement this alternative during the third period of the fourth ISI interval. Although thisreview is neither exigent nor emergency, your prompt review is requested. There are no new commitments made in this submittal. If you have any questions or requireadditional information, please contact me at (508) 830-8403. Sincerely, Joseph R. Lynch, Regulatory Assurance ManagerJRL/mew

Enclosure:

Pilgrim Relief Request (PRR) -24, "Alternative Examination Requirements forPilgrim Nuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner RadiiUsing ASME Code Case N-702 and BWRVIP-241 in Accordance with 10CFR50.55a (a)(3)(i)" cc: Mr. William M. DeanRegional Administrator, Region 1-U. S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100King of Prussia, PA 19406-1415 Ms. Nadiyah Morgan, Project ManagerOffice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission MS O-8C2AWashington, DC 20555NRC Senior Resident Inspector Pilgrim Nuclear Power Station Enclosure toLetter 2.14.028Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Sheet 1 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)"

1. ASME Code Component(s)

AffectedCode Class: 1Component Numbers: N2 (See Attachment 1 for detailed list of components) Applicable Code Edition: ASME Section Xl, 1998 Edition with 2000 AddendaExamination Category: B-D (Inspection Program B)Item Numbers: B3.90 and B3.100Description: Alternative to ASME Section Xl, Table IWB-2500-1 Unit/Inspection Interval Pilgrim (PNPS) /Fourth (4th)10-year interval startingApplicability: July 1, 2005, ending June 30, 20152. Applicable Code Requirement ASME Section Xl, 1998 Edition with 2000 Addenda Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles In Vessels -Inspection Program Brequires a volumetric examination of all nozzles with full penetration welds to the vesselshell (or head) and integrally cast nozzles each 10-year interval. Additionally, forultrasonic examinations, ASME Section Xl, Appendix VIII, "Performance Demonstration forUltrasonic Examination Systems," is implemented, as required and modified by 10 CFR50.55a(b)(2)(xv). The subject components for this request for alternative are the N2Recirculation Inlet Reactor Vessel Nozzle-to-Vessel Welds (Item B3.90) and the ReactorVessel Nozzle Inner Radius Sections (Item B3.100).3. Reason for RequestThe twenty-five percent sampling level stated in Code Case N-702 (Reference 3)provides a significant cost savings and reduction in worker dose exposure. PNPShas estimated that the proposed reduction of inspection requirements would result ina cost savings of approximately $200,000 and reduction in worker radiation exposure of approximately 3.7 Person-Rem over the remainder of the currentinterval while providing an acceptable level of quality and safety.Enclosure to Letter 2.14.028Sheet 2 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)"

4. Proposed Alternative Pursuant to 10 CFR 50.55a (a)(3)(i),

an alternative is requested from performing therequired examinations on 100% of the identified nozzle assemblies listed inAttachment

1. As an alternative, incorporation of Code Case N-702 (Reference 3)would require examination of a minimum of 25% of the nozzle-to-vessel welds andinner radius sections, including at least one nozzle from each system and nominalpipe size as shown in Table 4-1. For three of the Reactor Recirculation nozzleassemblies shown in Table 4-1, both the inner radius region and the nozzle-to-shell weld have already been examined during the 4th interval.

Table 4-1PNPS Summary -Affected Components of PRR-24MinimumNumberNozzle Total to beGroup Description Number Examined CommentsRecirculation 3 completed inInlet RFO17 (2009)5. Basis for Proposed Alternative Electric Power Research Institute (EPRI) Technical Report 1021005, "BWRVIP-241: BWRVessel and Internals Project (BWRVIP), Probabilistic Fracture Mechanics Evaluation forthe Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii"(Reference

6) provides the technical basis for use of Code Case N-702. BWRVIP-241 was developed to propose a relaxation of the criteria in BWRVIP-108 allowing BWR's toobtain inspection relief for their Reactor Recirculation inlet and outlet nozzles.

Theevaluation found that failure probabilities due to a low temperature overpressure event atthe nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e. < lX 10.6for 40 years) with or without inservice inspection. The report concludes that inspection of25% of each nozzle type is technically justified. BWRVIP-108 was originally submitted to the NRC for review and approval by the BWRVIPvia BWRVIP Letter 2002-323 on November 25, 2002 (Reference

2) and supplemented byTennessee Valley Authority (TVA) letter dated November 15, 2004, and BWRVIP lettersdated July 25, 2006, and September 13, 2007. Reference 4 provided the NRC SafetyEvaluation approving use of ASME Code Case N-702 in accordance with BWRVIP-108 guidance.

On April 19, 2013, the NRC issued a Safety Evaluation (SE) (Reference

7) approving theuse of BWRVIP-241.

Within Section 5 of the SE, it states that each licensee shoulddemonstrate the plant-specific applicability of the BWRVIP-241 report to their units in therequest for alternative by meeting the criteria discussed in Section 5 of the SE.Enclosure to Letter 2.14.028Sheet 3 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" The applicability of the BWRVIP-241 report to PNPS is demonstrated by showing thecriteria within Section 5 of the SE are met for the recirculation inlet nozzles.The general terms used in the SE Section 5 applicability evaluations are:Ci.RPV = recirculation inlet nozzles (from BWRVIP-108 model) = 19332 psiCI-NOZZLE = recirculation inlet nozzles (from BWRVIP- 108 model) = 1637 psiThe PNPS nozzle-specific terms to be used in the SER Section 5 applicability evaluations are as follows:Heatup / Cooldown rate = 100°F/hrp= Reactor Pressure Vessel (RPV) normal operating

pressure, p = 1035 psir= RPV inner radius, r = 113.40625" t= RPV wall thickness, t = 6.5"riN2 = inner radius for Recirculation Inlet N2 nozzles, rjN2 = 5.75"roN2 -outer radius for Recirculation Inlet N2 nozzles, roN2 = 9.125"The results of the equations in Attachment 2 demonstrate the applicability of the BWRVIP-108 1 BWRVIP-241 reports to PNPS by showing the criteria within Section 5.0 of the NRCSE is met for all nozzles listed in Attachment 1, Table 1. Therefore, the basis for usingCode Case N-702 is demonstrated for the PNPS reactor recirculation inlet nozzles inTable 1.In addition, Code Case N-702 stipulates that a VT-1 examination may be used in lieu ofthe volumetric examination for the inner radii (Item No. B3.100).

Note that PNPS is notcurrently using Code Case N-648-1 and has no plans of using the code case in the future.All examinations on Item B3.100, inner radius section, will be volumetric examinations. The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on theassumption that fluence at the nozzles is negligible. The PNPS N2 nozzles are predicted to experience fluence in excess of 1 x 1017 by the end of the currently licensed period ofextended operation. Therefore, a plant specific probabilistic fracture mechanics evaluation (Reference

8) was performed to supplement the criteria of Code Case N-702 andBWRVIP-241 in order to demonstrate that the probability of failure remains acceptable over this period of extended operation (to 60 years). This analysis was performed usingthe same methods as were used in BWRVIP-241, with PNPS specific fracture mechanics analyses.

The results demonstrate that for the N2 nozzles at PNPS, the probability offailure was less than the limiting probability as defined in 10 CFR 50 Appendix H.Therefore, the probabilistic fracture mechanics criteria of BWRVIP-241 remain applicable to the PNPS N2 nozzles.Enclosure to Letter 2.14.028Sheet 4 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crackgrowth over the initial licensed operating period and assumed additional fatigue cycles inevaluating fatigue crack growth. PNPS is projected to exceed the total number of thermalcycles used in the BWRVIP analysis over its remaining life. However, the usage factor forthe N2 nozzles remains below 1.0. Previous BWRVIP documents have demonstrated thatfatigue crack growth is not a significant contributor to probability of failure for the N2nozzles. Consideration of fatigue crack growth for the PNPS N2 nozzles is included in theplant specific probabilistic fracture mechanics analysis. The plant specific analysisconfirms that the additional fatigue cycles will have an insignificant effect on probability offailure for the PNPS N2 nozzles.In conclusion, use of ASME Code Case N-702 provides an acceptable level of quality andsafety in accordance with 10 CFR 50.55a(a)(3)(i) for Reactor Recirculation Inlet nozzle-to vessel shell full penetration welds and nozzle inner radii sections identified in Attachment 1, Table 1.6. Duration of Proposed Alternative Upon approval by the NRC staff, this request for alternative will be utilized through theremainder of Pilgrim's fourth inspection interval (July 1, 2005 -June 30, 2015) for thenozzle assemblies listed in Attachment 1, Table 1.7. Precedents None. However, since the NRC issued the SER for BWRVIP-241 in April of 2013, anotherplant has initiated a Relief Request to use the alternate criteria of Code Case N-702.8. References (1) EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and Internals ProjectTechnical Basis for the Reduction of Inspection Requirements for the Boiling WaterReactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2002(2) BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document ControlDesk, "Project No. 704- BWRVIP-108: BWR Vessel and Internals

Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water ReactorNozzle-to-Vessel Shell Welds and Nozzle Blend Radii," November 25, 2002(3) 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 Xl, Division 1," dated February 20, 2004(4) NRC Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals

Project, Technical Basis for the Reduction of Inspection Requirements for the BoilingWater Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated December 19, 2007(5) NRC SE for Pilgrim Relief Request PRR-20, Alternative Examination Requirements For Nozzle to Shell and Inner Radii Welds Using ASME Code Case N-702 andBWRVIP-108 (TAC NO. ME3290) dated August 25, 2010 (Letter No: 1.10.035)

Enclosure to Letter 2.14.028Sheet 5 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" (6) EPRI Technical Report 1021005, "BWRVIP-241, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and NozzleBlend Radii," dated October 2010(7) NRC SE 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 (TAC NO. ME6328) dated April 19, 2013"(8) Structural Integrity Associates, "Evaluation of the Probability of Failure forRecirculation Inlet (N2) in the Nozzle-to-Shell-Welds and Nozzle Blend Radii Regionsat Pilgrim Nuclear Station", 1400071.301 Revision 0, February 2014Enclosure to Letter 2.14.028Sheet 6 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" Attachment 1Table ITable of ASME Code Components Affected at PNPSCodeComponent ID Description Category Code ItemRPV-N2A-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2A-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2B-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2B-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2C-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2C-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2D-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2D-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2E-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2E-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2F-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2F-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2G-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2G-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2H-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2H-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2J-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2J-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100RPV-N2K-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90RPV-N2K-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100Enclosure to Letter 2.14.028Sheet 7 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Attachment 2Given the general and plant-specific terms, Pilgrim's conformance with the three (3)criteria applicable to the recirculation inlet nozzles is demonstrated as follows:(1) Max RPV Heatup/Cooldown RateCriterion -the maximum RPV heatup / cooldown rate is limited to < 1150°F/hr In accordance with Technical Specification 3.6.A.2, Reactor Coolant System heatupand cooldown rates are procedurally limited to a maximum of 100'F when averagedover any one hour period and thus meets the requirement of criterion 1.(2) Recirculation Inlet (N2) NozzlesEquation to meet criterion: (prlt)/CRpv < 1.15[(1035)(113.40625)16.51/19332 = 0.93 < 1.15The PNPS result is 0.93 and thus meets the requirement of criterion 2 to be < 1.15.(3) Recirculation Inlet (N2) NozzlesEquation to meet criterion: (p(ro2 + ri2)l (ro1 -ri2)/CN=ozzLE < 1,47[1035(9.1252+5.752)/(9.1252-5.752)]11637 = 1.465 < 1.47The PNPS result is 1.465 and thus meets the requirement of criterion 3 to be < 1.47.Enclosure to Letter 2.14.028Sheet 8 of 8}}