ML100350096
| ML100350096 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/01/2010 |
| From: | Bethay S Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.10.004, BWRVIP-108, PRR-20 | |
| Download: ML100350096 (10) | |
Text
N-,ýEnteW Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Stephen J. Bethay Director, Nuclear Assessment February 1,2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001
SUBJECT:
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Entergy Pilgrim Relief Request (PRR)-20, Alternative Examination Requirements for Pilgrim Nozzle-to-Vessel Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108
REFERENCE:
- 1. ASME Boiler and Pressure Vessel Code, Code case N-702, "Alternative Requirements for Boiling Water Reactor (BWR)
Nozzle Inner Radii and Nozzle-to-Vessel Shell Welds, Section Xl, Division 1," dated February 20, 2004
- 2. BWRVIP-108, "BWR Vessel Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,"
dated October 2002 LETTER NUMBER: 2.10.004
Dear Sir or Madam,
Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Pilgrim) requests NRC approval of Pilgrim Relief Request (PRR)-20, Alternative Examination Requirements for Pilgrim Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii using the requirements of American Society of Mechanical Engineers (ASME) Code Case N-702 and BWRVIP-108 provisions (References 1 and 2).
Specifically, ASME Code Section Xl, Table IWB-2500-1 requires 100% examination of Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii during the current In-service Inspection (ISI) interval. Pilgrim is in the 4 th ISI interval, which began on July 1, 2005 and will end on June 30, 2015. The Code of Record for Pilgrim is ASME Section Xl, 1998 Edition with 2000 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 inner radii sections, including at least one nozzle from each system and nominal pipe size. BWRVIP-108 provides the technical basis for the reduction of examination requirements for BWR nozzle-to-vessel shell welds and nozzle blend radii. AO 7I Kll4(L
Entergy Nuclear Operations, Inc. Letter Number: 2.10.004 Pilgrim Nuclear Power Station Page 2 The proposed alternative provides an acceptable level of quality and safety because it follows the requirements of ASME Code N-702 and the NRC approved technical basis (BWRVIP -108) for the alternative examination presented in the enclosed Pilgrim Relief Request (PRR)-20.
The enclosed PRR-20 provides the details of the proposed alternative examination requirements.
NRC has previously approved the proposed alternative examination requirements for the following plants:
- 1. Duane Arnold Energy Center, Docket No. 50-331, TAC No. MD8193/August 29, 2008
- 2. Perry Nuclear Power Plant, Docket No. 50-440, TAC No. MD8458/December 29, 2008
- 3. Columbia Generating Station, Docket No. 50-397, TAC No. MD9850/April 8, 2009
- 4. Clinton Power Station, Docket No. 50-461, TAC No. ME0218/August 24, 2009
- 5. 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 10 CFR 50.55a(a)(3)(i) by March 1, 2011, in support of Pilgrim Refueling Outage (RFO)-1 8, which is currently planned for April 15, 2011.
There are no commitments made in this submittal.
If you have any questions, please call Mr. Joseph Lynch, Pilgrim Licensing Manager at 508-830-8403.
Sincerely, Stephen J. Bethay Director, Nuclear Safety Assurance SJB/wgl
Enclosure:
Fourth Interval ISI Program Pilgrim Relief Request (PRR)-20 (7 pages) cc: Regional Administrator, Region 1 Mr. James S. Kim, Project Manager U.S. Nuclear Regulatory Commission Plant Licensing Branch I-1 475 Allendale Road Division of Operator Reactor Licensing King of Prussia, PA 19406 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North 0-8C2 Senior Resident Inspector 11555 Rockville Pike Pilgrim Nuclear Power Station Rockville, MD 20852
Enclosure to Entergy Letter Number 2.10.004 Fourth Interval ISI Program Pilgrim Relief Request (PRR)-20 (7 pages)
Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20 ASME Code Component(s) Affected Code Class:
Component Numbers: N1, N3, N6, N7, N8, N9 (See Attachment 1 for detailed list of components)
Code
References:
(1) ASME Section XI, 1998 Edition with 2000 Addenda (2) Code Case N-702 (3) BWRVIP-108, Technical Basis for the Reduction of Inspection Requirements for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner 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 Pilgrim (PNPS) /Fourth ( 4 th) 10-year interval Applicability: starting July 1,2005, ending June 30, 2015
2. 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 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 10 CFR 50.55a(b)(2)(xv). The subject components for this request for alternative are the Reactor Vessel Nozzle-to-Vessel Welds (Item B3.90) and the Reactor Vessel Nozzle Inside Radius Section (Item B3.100).
3. Reason for Request
The twenty-five percent sampling level stated in Code Case N-702 (Reference 3) provides a significant cost savings and reduction in worker dose exposure. PNPS has estimated that the proposed reduction of inspection requirements would result in a cost savings of up to $197,000 and reduction in worker dose exposure of approximately 1.4 Person-Rem over the remainder of the current interval.
Page 1 of 7
Entergy Nuclear Operations,,Inc Pilgrim Nuclear Power Station Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20
- 4. Proposed Alternative Pursuant to 10 CFR 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 (Reference 3) would require examination at a minimum, 25% of the nozzle-to-vessel welds and 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 or will be examined during the 4 th interval.
Table 4-1 PNPS Summary - Affected Components Minimum Number Nozzle Total to be Group Description Number Examined Comments Recirculation 1 completed in,,
Outlet RFO17 3Main Steam 4 1 All 4 completed in RFO16 N6 Core Spray 2 1 1 completed in RFO16 N7 Spare 2 1 1 scheduled in Instrumentation RFO20 N8 Head Vent 1 1HRFO20 1 scheduled in N9 Jet Pump 2 1 1 completed in Instrumentation RFO17
- 5. Basis for Proposed Alternative Electric Power Research Institute (EPRI) Technical Report 1003557, "BWRVIP-108: 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" (References 1 and 2) provides the technical basis for use of Code Case N-702. The 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.r for 40 years) with or without inservice inspection.
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 by the BWRVIP via BWRVIP Letter 2002-323 on November 25, 2002 and supplemented by Page 2 of 7
Entergy Nuclear Operations, Inc Pilgrim Nuclear power Station Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20 Tennessee Valley Authority (TVA) letter dated November 15, 2004, and BWRVIP letters dated July 25, 2006, and September 13, 2007.
On December 19, 2007, the NRC issued a Safety Evaluation (SE) (Reference 4) approving the use of BWRVIP-108. Within Section 5 of the SE, it states that each licensee should demonstrate the plant-specific applicability of the BWRVIP-108 report to
' their units in the request for alternative by meeting the criteria discussed in Section 5 of the SE.
The applicability of the BWRVIP-108 report to PNPS is demonstrated by showing the criteria within Section 5 of the SE are met.
- The general terms used in the SE Section 5 applicability evaluations are:
CiRPV = recirculationinlet nozzles (from BWRVIP-108 model) = 19332 psi Ci-NOZZLE.= recirculationinlet nozzles (from BWRVIP-108 model) = 1637 psi Co-RPV= recirculation outlet nozzles (from BWRVIP-108 model) = 16171 psi Co-NOZZLE= recirculationoutlet nozzles (from BWRVIP- 108 model) = 1977 psi
- The Pilgrim nozzle-specific terms to be used in the SE Section 5 applicability evaluations are as follows:
Heatupi Cooldown rate = 100 0F/hr p= Reactor PressureVessel (RPV) normal operatingpressure,p = 1035 psi r= RPV inner radius, r = 113.40625" t= RPV wall thickness, t = 6.5" riN2- inner radiusfor Recirculation Inlet N2 nozzles, riN2= 5.75" roN2 = outer radius for Recirculation Inlet N2 nozzles, roN2 = 9.125" rjN1 - inner radius for Recirculation Outlet N1 nozzles, rNj = 13.031" roN, = outer radius for Recirculation Outlet N1 nozzles, roNl = 22.3125" The results of the equations in Attachment 2 demonstrate the applicability of the BWRVIP-108 report to Pilgrim by showing the criteria within Section 5.0 of the NRC SE is met for all nozzles listed in Table 1 with the exception of the PNPS Recirculation system inlet (N2) nozzles. The PNPS N2 nozzles did not meet the third criterion of the SE and therefore-Code Case N-702 would not be applied to the N2 nozzles. The assumption, in Section 5.0 of the SE transmitted by Reference 4, that only recirculation inlet and outlet nozzles need to be checked because the conditional probability of failure (P(FIE)) for other nozzles is an order of magnitude lower, remains valid. Therefore, the basis for using Code Case N-702 is demonstrated for Pilgrim for all nozzles listed in Table 1. It is noted that the Feedwater (N4) nozzles and CRD Return (N10) nozzle are outside the scope of Code Case N-702 and are excluded from this application.
In addition, Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (Item No. B3.100). Note that PNPS is not currently 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.
Page 3 of 7
Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISl Program Pilgrim Relief Request No. PRR-20 In conclusion, use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for all applicable RPV 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 the remainder of Pilgrim's fourth inspection interval (July 1, 2005 - June 30, 2015) for the nozzle assemblies listed in Attachment 1.
- 7. Precedents The NRC Staff has approved similar Requests for Alternative for the following plants:
(1) Duane Arnold Energy Center, Docket No. 50-331, TAC No. MD8193/August 29, 2008 (2) Perry Nuclear Power Plant, Docket No. 50-440, TAC No. MD8458/December 29, 2008 (3) Columbia Generating Station, Docket No. 50-397, TAC No. MD9850/April 8, 2009 (4) Clinton Power Station, Docket No. 50-461, TAC No. ME0218/August 24, 2009 (5) Dresden Nuclear Power Station, Docket Nos. 50-237 & 50-249, TAC Nos. ME0882
& ME0883/November 3, 2009
- 8. References (1) 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 Blend Radii," dated October 2002 (2) 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 (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) Letter from Matthew A. Mitchell (NRR), to Rick Libra, BWRVIP Chairman, Safety Evaluation of Proprietary EPRI Report, 'BWR Vesseland 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 Page 4 of 7
Attachment 1 PRR-20 Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Request for Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20 Table 1 Table of ASME Code Components Affected at PNPS Code Component ID Description Category Code Item RPV-N1A-NV 28" Recirc Outlet Nozzle to Vessel Weld B-D B3.90 RPV-N1A-NIR 28" Recirc Outlet Nozzle to Inner Radius B-D B3.100 RPV-N1B-NV 28" Recirc Outlet Nozzle to Vessel Weld B-D B3.90 RPV-N1B-NIR 28" Recirc Outlet Nozzle to Inner Radius B-D B3.100 RPV-N3A -NV 20" Main Steam Nozzle to Vessel Weld B-D B3.90 RPV-N3A-NIR 20" Main Steam Nozzle to Inner Radius B-D B3.100 RPV-N3B-NV 20" Main Steam Nozzle to Vessel Weld B-D B3.90 RPV-N3B-NIR 20" Main Steam Nozzle to Inner Radius B-D B3.100 RPV-N3C-NV 20" Main Steam Nozzle to Vessel Weld B-D B3.90 RPV-N3C-NIR 20" Main Steam Nozzle to Inner Radius B-D B3.100 RPV-N3D-NV 20" Main Steam Nozzle to Vessel Weld B-D B3.90 RPV-N3D-NIR 20" Main, Steam Nozzle to Inner Radius B-D B3.100 RPVN6A-NV 12" Core Spray Nozzle to Vessel Weld B-D B3.90 RPV-N6A-NIR 12" Core Spray Nozzle to Inner Radius B-D B3.100 RPV-N6B-NV 12" Core Spray Nozzle to Vessel Weld B-D B3.90 RPV-N6B-NIR 12" Core Spray Nozzle to Inner Radius B-D B3.100 RPV-N7A-NV 9" Spare Instrumentation Nozzle to Vessel Weld B-D B3.90 RPV-N7A-NIR 9" Spare Instrumentation Nozzle to Inner Radius B-D B3.100 RPV-N7B-NV 9" Spare Instrumentation Nozzle to Vessel Weld B-D B3.90 RPV-N7B-NIR 9" Spare Instrumentation Nozzle to Inner Radius B-D B3.100 RPV-N8-NV 6" Head Vent Nozzle to Vessel Weld B-D B3.90 RPV-N8-NIR 6" Head Vent Nozzle to Inner Radius B-D B3.100 RPV-N9A-NV 5" Jet Pump Instrumentation Nozzle to Vessel B-D B3.90 Weld RPV-N9A-NIR 5" Jet Pump Instrumentation Nozzle to Inner B-D B3.100 Radius RPV-N9B-NV 5" Jet Pump Instrumentation Nozzle to Vessel B-D B3.90 I Weld RPV-N9B-NIR 5" Jet Pump Instrumentation Nozzle to Inner B-D B3.100 Radius Page 5 of 7
Attachment 2 PRR-20 Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Request for Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20 Response to NRC Plant-Specific Applicability of BWRVIP 108 Criteria Given the general and plant-specific terms, Pilgrim's conformance with the five (5) criteria is demonstrated as follows:
(1) Max RPV Heatup/Cooldown Rate Criterion - the maximum RPV heatup/cooldown rate is limited to < 11501F/hr In accordance with Technical Specification 3.6.A.2 Reactor Coolant System heatup and cooldown rates are limited to a maximum of 100OF when averaged over any one hour period and thus meets the requirement of Criterion 1.
(2) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (Pr/t)/ Ci.RPV < 1.15
[(1035 )(113.40625)/6.5]/19332 = 0.93 < 1.15 The PNPS result is 0.93 and thus meets the requirement of Criterion 2 to be < 1.15.
(3) Recirculation Inlet (N2) Nozzles Equation to meet criterion:
[p(roN2 2 + riN22) + (roN2 2 - riN22)] + C-NOZZLE < 1. 15
[1035(9.1252+5.752)/(9.1252-5.752)]/1637 = 1.465 > 1.15 The PNPS result is 1.465 and thus does not meet the requirement of Criterion 3 to be < 1.15.
(4) Recirculation Outlet (Ni) Nozzles Equation to meet criterion: (prWt) Co-Rpv < 1 .15
[(1035)(11 3.40625)/6.5]/1 6171 = 1.117 < 1.15 The PNPS result is 1.117and thus meets the requirement of Criterion 4 to be < 1.15 (5) Recirculation Outlet (Ni) Nozzles Equation to meet criterion:
[p(roN1 2 + riN1) (rNN 2 - riN1)] + Co-NOZZLE < 1.15
[1035(22.31252 + 13.0312) / (22.31252 - 13.0312)] / 1977 = 1.0655 < 1.15 The PNPS result is 1.065 and thus meets the requirement of Criterion 5 to be < 1.15 Page 6 of 7
Attachment 3 PRR-20 Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Request for Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
Fourth Interval ISI Program Pilgrim Relief Request No. PRR-20 PNPS 4 th INTERVAL OUTAGE SCHEDULE 4 th Inspection Interval started July 1, 2005 - ends June 30, 2015 PERIOD OUTAGE OUTAGE DATE 1 RFQ16 April 2007 1 RFO17 April 2009 2 RFO18 April 2011 3 RFO19 April 2013 3 RFO20 April 2015 Page 7 of 7