View source for BVY 07-047, License Renewal Application, Amendment 27, Submitted on January 25, 2006

{{Adams
| number = ML071900203
| issue date = 07/03/2007
| title = License Renewal Application, Amendment 27, Submitted on January 25, 2006
| author name = Sullivan T
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000271
| license number = DPR-028
| contact person = 
| case reference number = BVY 07-047, TAC MC9668
| document type = Letter type:BVY
| page count = 36
| project = TAC:MC9668
| stage = Other
}}

=Text=
{{#Wiki_filter:SwEntergy Entergy Nuclear Operations, Inc.Vermont Yankee P.O. Box 0500 185 Old Ferry Road Brattleboro.
VT 05302-0500 Tel 802 257 5271 July 3, 2007 Docket No. 50-271 BVY 07-047 TAC No. MC 9668 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==Reference:==

==Subject:==
: 1. Letter, Entergy to USNRC, "Vermont Yankee Nuclear Power Station, License No. DPR-28, License Renewal Application," BVY 06-009, dated January 25, 2006.2. Letter, USNRC to Entergy, "Safety Evaluation Report with Confirmatory Items Related to the License Renewal of Vermont Yankee Nuclear Power Station," NVY 07-036, dated March 30. 2007.Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)License Renewal Application, Amendment 27 On January 25, 2006, Entergy Nuclear Operations, Inc. and Entergy Nuclear Vermont Yankee, LLC (Entergy) submitted the License Renewal Application (LRA) for the Vermont Yankee Nuclear Power Station (VYNPS) as indicated by Reference
: 1. This letter contains attachments to address the Safety Evaluation Report (SER) Confirmatory Items as indicated by Reference 2, and other issues discussed with the NRC since the issuance of that report.An updated License Renewal Commitment List (Rev. 8) is provided that includes the following changes;S 0 0 Source document additions to commitments 4, 32, 34 and 43.New commitment number 50.Change to commitment 27.Should you have any questions concerning this letter, please contact Mr. Dave Mannai at (802) 258-5422.I declare under penalty of perjury that the foregoing is true and correct, executed on July 3, 2007.Sincerey St~ ice sident Ve rmont Yankee Nuclear Power Station cc: See next page enc: Attachments 1-5.7 4"K BVY 07-047 Docket No. 50-271 Page 2 of 3 cc: Mr. James Dyer, Director U.S. Nuclear Regulatory Commission Office O5E7 Washington, DC 20555-00001 Mr. Samuel J. Collins, Regional Administrator U.S. Nuclear Regulatory Commission, Region 1 475 Allendale Road King of Prussia, PA 19406-1415 Mr. Jack Strosnider, Director U.S. Nuclear Regulatory Commission Office T8A23 Washington, DC 20555-00001 Mr. Jonathan Rowley, Senior Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike MS-O-11 F1 Rockville, MD 20853 Mr. Mike Modes USNRC RI 475 Allendale Rd, King of Prussia, PA 19406 Mr. James S. Kim, Project Manager U.S. Nuclear Regulatory Commission Mail Stop 0 8 C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC P.O. Box 157 (for mail delivery)Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street -Drawer 20 Montpelier, Vermont 05620-2601 Diane Curran, Esq.Harmon, Curran, Spielberg 
& Eisenberg, LLP 1726 M Street, N.W., Suite 600 Washington, D.C. 20036 BVY 07-047 Docket No. 50-271 Page 3 of 3 List of Attachments" Attachment 1: VYNPS License Renewal Commitment List, Revision 8* Attachment 2: SER Confirmatory Items Response* Attachment 3: Vernon Hydro-Electric Station Information" Attachment 4: Vernon Hydro-Electric Station RAI 3.6.2.2-N-08-1 Clarification
* Attachment 5: AxialWeld Analysis Clarification BVY 07-047 Docket No. 50-271 Attachment I Vermont Yankee Nuclear Power Station License Renewal Application Amendment 27 VYNPS License Renewal Commitment List, Revision 8 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 During the development and review of the Vermont Yankee Nuclear Power Station License Renewal Application, Entergy made commitments to provide aging management programs to manage the effects of aging on structures and components during the extended period of operation.
The following table lists these license renewal commitments, along with the implementation schedule and the source of the commitment.
ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 1 Guidance for performing examinations of buried piping will be enhanced to March 21, 2012 BVY 06-009 B.1 .1 specify that coating degradation and corrosion are attributes to be Audit Items 5 &evaluated.
130 2 Fifteen (15) percent of the top guide locations will be inspected using As stated in the BVY 06-009 B.1.7 enhanced visual inspection technique, EVT-1, within the first 18 years of commitment Audit Item 14 the period of extended operation, with at least one-third of the inspections to be completed within the first 6 years and at least two-thirds within the first 12 years of the period of extended operation.
Locations selected for examination will be areas that have exceeded the neutron fluence threshold.
3 The Diesel Fuel Monitoring Program will be enhanced to ensure ultrasonic March 21, 2012 BVY 06-009 B.1.9 and thickness measurement of the fuel oil storage and fire pump diesel storage BVY 07-018 regional (day) tank bottom surfaces will be performed every 10 years during tank inspection cleaning and inspection.
4 The Diesel Fuel Monitoring Program will be enhanced to specify UT March 21, 2012 BVY 06-009 B.1.9 and measurements of the fuel oil storage and fire pump diesel storage (day) BVY 07-018 regional tank bottom surfaces will have acceptance criterion 
> 60% Tnom. inspection Page 1 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 5 The Fatigue Monitoring Program will be modified to require periodic update March 21, 2012 BVY 06-009 B.1.11 of cumulative fatigue usage factors (CUFs), or to require update of CUFs if the number of accumulated cycles approaches the number assumed in the design calculation.
6 A computerized monitoring program (e.g., FatiguePro) will be used to March 21, 2012 BVY 06-009 B.1.11 directly determine cumulative fatigue usage factors (CUFs) for locations of interest.7 The allowable number of effective transients will be established for March 21, 2012 BVY 06-009 B.1.11 monitored transients.
This will allow quantitative projection of future margin.8 Procedures will be enhanced to specify that fire damper frames in fire March 21, 2012 BVY 06-009 B.1.12.1 barriers will be inspected for corrosion.
Acceptance criteria will be Audit Items 35, enhanced to verify no significant corrosion.
151, 152, 153 and 159 9 Procedures will be enhanced to state that the diesel engine sub-systems March 21,2012 BVY 06-009 B.1.12.1 (including the fuel supply line) will be observed while the pump is running. Audit Items 33, Acceptance criteria will be enhanced to verify that the diesel engine did not 150 & 155 exhibit signs of degradation while it was running; such as fuel oil, lube oil, coolant, or exhaust gas leakage.Page 2 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 10 Fire Water SystemProgram procedures will be enhanced to specify that in March 21, 2012 BVY 06-009 B.1.12.2 accordance with NFPA 25 (2002 edition), Section 5.3.1.1.1, when sprinklers have been in place for 50 years a representative sample of sprinkler heads will be submitted to a recognized testing laboratory for field service testing.This sampling will be repeated every 10 years.11 The Fire Water System Program will be enhanced to specify that wall March 21, 2012 BVY 06-009 B.1.12.2 thickness evaluations of fire protection piping will be performed on system Audit Items 37 &components using non-intrusive techniques (e.g., volumetric testing) to 41 identify evidence of loss of material due to corrosion.
These inspections will be performed before the end of the current operating term and during the period of extended operation.
Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.12 Implement the Heat Exchanger Monitoring Program as described in LRA March 21, 2012 BVY 06-009 B.1.14 Section B.1.14.13 Implement the Non-EQ Inaccessible Medium-Voltage Cable Program as March 21, 2012 BVY 06-009 B.1.17 described in LRA Section B.1.17.14 Implement the Non-EQ Instrumentation Circuits Test Review Program as March 21, 2012 BVY 06-009 B.1.18 described in LRA Section B.1.18.15 Implement the Non-EQ Insulated Cables and Connections Program as March 21, 2012 BVY 06-009 B.1.19 described in LRA Section B.1.19.Page 3 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA Section No./SCHEDULE Comments 16 Implement the One-Time Inspection Program as described in LRA Section March 21, 2012 BVY 06-009 B.1.21 B.1.21. BVY 07-009 Audit Items 239, 240, 330, 331 17 Enhance the Periodic Surveillance and Preventive Maintenance Program to March 21, 2012 BVY 06-009 B.1.22 assure that the effects of aging will be managed as described in LRA Audit Item 377 Section B.1.22.18 Enhance the Reactor Vessel Surveillance Program to proceduralize the March 21, 2012 BVY 06-009 B.1.24 data analysis, acceptance criteria, and corrective actions described in the program description in LRA Section B.1.24.19 Implement the Selective Leaching Program as described in LRA Section March 21, 2012 BVY 06-009 B.1.25 B.1.25.20 Enhance the Structures Monitoring Program to specify that process facility March 21, 2012 BVY 06-009 B.1.27.2 crane rails and girders, condensate storage tank (CST) enclosure, C02 Audit Item 377 tank enclosure, N 2 tank enclosure and restraining wall, CST pipe trench, diesel generator cable trench, fuel oil pump house, service water pipe trench, man-way seals and gaskets, and hatch seals and gaskets are included in the program.21 Guidance for performing structural examinations of wood to identify loss of March 21, 2012 BVY 06-009 B.1.27.2 material, cracking, and change in material properties will be added to the Structures Monitoring Program.Page 4 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 22 Guidance for performing structural examinations of elastomers (seals and March 21, 2012 BVY 06-009 B.1.27.2 gaskets) to identify cracking and change in material properties (cracking when manually flexed) will be enhanced in the Structures Monitoring Program procedure.
23 Guidance for performing structural examinations of PVC cooling tower fill to March 21, 2012 BVY 06-009 B.1.27.2 identify cracking and change in material properties will be added to the Structures Monitoring Program procedure.
24 System walkdown guidance documents will be enhanced to perform March 21, 2012 BVY 06-009 B.1.28 periodic system engineer inspections of systems in scope and subject to Audit Items 187, aging management review for license renewal in accordance with 10 CFR 188 & 190 54.4 (a)(1) and (a)(3). Inspections shall include areas surrounding the subject systems to identify hazards to those systems. Inspections of nearby systems that could impact the subject system will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(2).25 Implement the Thermal Aging and Neutron Irradiation Embrittlement of Cast March 21, 2012 BVY 06-009 B.1.29 Austenitic Stainless Steel (CASS) Program as described in LRA Section B.1.29.26 Procedures will be enhanced to flush the John Deere Diesel Generator March 21, 2012 BVY 06-009 B.1.30.1 cooling water system and replace the coolant and coolant conditioner every Audit Items 84 &three years. 164 Page 5 of 11 BVY,07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 27 At least 2 years prior to entering the period of extended operation, for the locations identified in NUREG/CR-6260 for BWRs of the VY vintage, VY will refine our current fatigue analyses to include the effects of reactor water environment and verify that the cumulative usage factors (CUFs) are less than 1. This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:
: 1. For locations, including NUREG/CR-6260 locations, with existing fatigue analysis valid for the period of extended operation, use the existing CUF to determine the environmentally adjusted CUF.2. More limiting VY-specific locations with a valid CUF may be added in addition to the NUREG/CR-6260 locations.
: 3. Representative CUF values from other plants, adjusted to or enveloping the VY plant specific external loads may be used if demonstrated applicable to VY.4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.During the period of extended operation, VY may also use one of the following options for fatigue management if ongoing monitoring indicates a potential for a condition outside the analysis bounds noted above: 1) Update and/or refine the affected analyses described above.2) Implement an inspection program that has been reviewed and approved by the NRC (e.g., periodic nondestructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC).3) Repair or replace the affected locations before exceeding a CUF of 1.0.March 21, 2012 March 21, 2010 for performinga fatigue analysis that addresses the effects of reactor coolant environment on fatigue (in accordance with an NRC approved version of the ASME Code)BVY-06-058 4.3.3 Audit Items 29, 107 & 318 Page 6 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 28 Revise program procedures to indicate that the Instrument Air Program will March 21, 2012 BVY 06-009 B.1.16 maintain instrument air quality in accordance with ISA S7.3 Audit Item 47 29 VYNPS will perform one of the following:
March 21, 2012 BVY 06-009 B.1.7 1. Install core plate wedges, or, Audit Item 9 2. Complete a plant-specific analysis to determine acceptance criteria for continued inspection of core plate hold down bolting in accordance with BWRVIP-25 and submit the inspection plan and analysis to the NRC two years prior to the period of extended operation for NRC review and approval.30 Revise System Walkdown Program to specify C02 system inspections March 21, 2012 BVY 06-009 B.1.28 every 6 months. Audit Items 30, 141, 146 & 298 31 Revise Fire Water System Program to specify annual fire hydrant gasket March 21, 2012 BVY 06-009 B.1.12.2 inspections and flow tests. Audit Items 39 &40 32 Implement the Metal Enclosed Bus Program. March 21, 2012 BVY 06-058 Audit Item 97 Details are provided in a LRA Amendment 16, Attachment 3 and LRA BVY 07-003 Amendment 23, 7. BVY 06-091 Page 7 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 33 Include within the Structures Monitoring Program provisions that will ensure March 21, 2012 BVY 06-009 B.1.27 an engineering evaluation is made on a periodic basis (at least once every Audit Item 77 five years) of groundwater samples to assess aggressiveness of groundwater to concrete.
Samples will be monitored for sulfates, pH and RAI 3.5-7 chlorides.
34 Implement the Bolting Integrity Program. March 21, 2012 BVY 06-058 Audit Items 198, Details are provided in a LRA Amendment 16, Attachment 2 and LRA BVY 07-003 216, 218, 237, Amendment 23, Attachment
: 5. 331 & 333 BVY 06-09 1 35 Provide within the System Walkdown Training Program a process to March 21, 2012 BVY 06-058 Audit Item 384 document biennial refresher training of Engineers to demonstrate inclusion of the methodology for aging management of plant equipment as described in EPRI Aging Assessment Field Guide or comparable instructional guide.36 If technology to inspect the hidden jet pump thermal sleeve and core spray March 21, 2010 BVY06-058 Audit Item 12 thermal sleeve welds has not been developed and approved by the NRC at least two years prior to the period of extended operation, VYNPS will initiate plant-specific action to resolve this issue. That plant specific action may be justification that the welds do not require inspection.
37 Continue inspections in accordance with the Steam Dryer Monitoring March 21, 2010 BVY 06-079 Audit Item 204 Program, Revision 3 in the event that the BWRVIP-139 is not approved prior to the period of extended operation.
Page 8 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 38 The BWRVIP-116 report which was approved by the Staff will be March 21, 2012 BVY 06-088 Response to implemented at VYNPS with the conditions documented in Sections 3 and RAI B.1.24-1 4 of the Staff's final SE dated March 1, 2006, for the BWRVIP-1 16 report.39 If the VYNPS standby capsule is removed form the reactor vessel without March 21, 2012 BVY 06-088 Response to the intent to test it, the capsule will be stored in a manner which maintains it RAI B.1.24-2 in a condition which would permit its future use, including during the period of extended operation, if necessary.
40 This Commitment has been deleted and replaced with Commitment
: 43. N/A BVY 07-018 N/A 41 This Commitment has been deleted and replaced with Commitment
: 43. N/A BVY 07-018 N/A 42 Implement the Bolted Cable Connections Program. March 21, 2012 BVY 07-003 Response to: RAI 3.6.2.2-N-01 Details are provided in LRA Amendment 23, attachment
: 7. BVY 07-018 LRA Sections: 3.6.2.1 A.2.1.39 B.1.33 Table 3.6.1 Table 3.6.2-1 43 Establish and implement a program that will require testing of the two 13.8 March 21, 2012 BVY 07-009 Am. 24 kV cables from the two Vernon Hydro Station 13.8 kV switchgear buses to BVY 07-018 Response to: the 13.8 kV / 69 kV step up transformers before the period of extended RAIs operation and at least once every 10 years after the initial test. 3.6.2.2-N-08-2 3.6.2.2-N-08-4 Page 9 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 44 Guidance for performing examinations of buried piping will be revised to March 21, 2012 BVY 07-018 Regional include the following. "A focused inspection will be performed within the inspection first 10 years of the period of extended operation, unless an opportunistic inspection (or an inspection via a method that allows an assessment of pipe condition without excavation) occurs within this ten-year period." 45 Enhance the Service Water Integrity Program to require a periodic visual March 21, 2012 BVY 07-018 Regional inspection of the RHRSW pump motor cooling coil internal surface for loss inspection of material.46 Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the March 21, 2012 BVY 07-018 Regional fire pump diesel storage (day) tank will be analyzed according to ASTM inspection D975-02 and for particulates per ASTM D2276. Also, fuel oil in the John Deere diesel storage tank will be analyzed for particulates per ASTM D2276.47 Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the March 21, 2012 BVY 07-018 Regional common portable fuel oil storage tank will be analyzed according to ASTM inspection D975-02, per ASTM D2276 for particulates, and ASTM D1796 for water and sediment.48 Perform an internal inspection of the underground Service Water piping March 21, 2012 BVY 07-018 Regional before entering the period of extended operation.
inspection 49 Revise station procedures to specify fire hydrant hose testing, inspection, March 21, 2012 BVY 07-009 Audit Item 38 and replacement, if necessary, in accordance with NFPA code specifications for fire hydrant hoses.Page 10 of 11 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST REVISION 8 ITEM COMMITMENT IMPLEMENTATION SOURCE Related LRA SCHEDULE Section No./Comments 50 During the period of extended operation, review the Vernon Dam owner March 21, 2012 BVY 06-009 RAI FERC required report(s) at a minimum of every five years to confirm that BVY 07-047 3.6.2.2.N-08-1 the Vernon Dam owner is performing the required FERC inspections.
Document deficiencies in the Entergy Corrective Actions Program and evaluate operability as described in BVY 96-043 and BVY 97-043 if it is determined that the required inspections are not being performed.
Page 11 of 11 BVY 07-047 Docket 50-271 BVY 07-047 Docket No. 50-271 Attachment 2 Vermont Yankee Nuclear Power Station License Renewal Application Supplement Amendment 27 SER Confirmatory Items Response VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 Supplemental response to RAI 2.1-2 (or to NRC telecom of 6/1/2007)1) For the following systems: Service Water Circulating Water Reactor Water Cleanup has the applicant identified the nonsafety-related portions of systems attached to safety-related SSCs required to be in the scope of license renewal in accordance with 10 CFR 54.4(a)(2)?
Have all component types been identified and included in the applicable tables in the LRA?2) Are there any nonsafety-related systems for which the applicant has not identified the nonsafety-related portions of systems which are attached to safety-related systems and required to be in the scope of license renewal in accordance with 10 CFR 54.4(a)(2)?
: 3) For all nonsafety-related systems, are all scoping activities complete, and have all component types been identified and included in the applicable tables of the LRA?4) Is the applicant taking any credit for, or committing to perform, any future license renewal scoping activities?
The scoping and screening methodology described in VYNPS LRA Section 2.1 was completed in the integrated plant assessment for the VYNPS LRA. This included the complete identification of nonsafety-related portions of systems, including service water, circulating water, and reactor water cleanup systems, attached to safety-related SSCs required to be in the scope of license renewal in accordance with 10 CFR 54.4(a)(2).
The identified component types are included in the applicable tables in the LRA. There are no nonsafety-related systems for which Entergy has not identified the nonsafety-related portions of systems which are attached to safety-related systems and required to be in the scope of license renewal in accordance with 10 CFR 54.4(a)(2)
As a result of discussions with the staff during the Region I inspection (February 2007), Entergy determined that some safety-related SSCs in the VY turbine building required consideration for potential spatial impacts from nonsafety-related SSCs per 10 CFR 54.4(a)(2).
Therefore, an expanded review for SSCs in the turbine building determined that additional, components required aging management review (AMR). Those additional component types are added to the applicable LRA tables. This includes the addition of new component types in the service water and circulating water systems discussed in Confirmatory Items 2.3.3.2a-1, 2.3.3.2a-2, and 2.3.3.13e-1, among other systems. No new component types are added for the reactor water clean up system (RWCU)discussed in Confirmatory Item 2.3.3.13m-185.
For all nonsafety-related systems, all scoping activities are complete, and all component types been identified and included in the applicable tables of the LRA. VYNPS is taking no credit for any future license renewal scoping activities.
Based on "lessons learned" during the review of LRAs for other Entergy nuclear power plants, it was determined that information obtained while performing aging management reviews of nonsafety-related SSCs attached to safety-related SSCs could be added to the applicable AMR document to assist the site in. the implementation of aging management activities during the period of extended operation.
This information for VYNPS is summarized in the following table.Page 1 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-33600-A217 J-4 and K-4 None; the boundaries shown are for an (a)(2)intended function (MSIV leakage____________00870 pathway).LRA-5920-00870 (no boundaries on this drawing)LRA-5920-4147 at expansion tank and heat Not required -exchangers Components already included for spatial interaction LRA-5920-4150 F-2 Drain line off lube Not required -oil day tank Components already included for spatial interaction LRA-G-191156 None; the boundaries shown are for an (a)(2)intended function (MSIV leakage pathway).LRA-G-1 91159 C-5 and H-5 3" SW-37B end of backwash line at sh.1 river discharge LRA-G-191159 D/E/F and G-4 line to intake end of screen wash sh.1 screens lines at intake structure LRA-G-1 91159 D-6 2" line to P-123-1A and 1B, P-sh.1 chlorinator 122-1A and 1 B, P-1 25A chemical treatment and 1B, P-126-1A and system 1 B, S-70-13 and S-70-14 in intake structure LRA-G-191159 E-6 line to CW pumps end of overboard sh.1 discharge lines at intake structure LRA-G-191159 E-10 line to RRU-1 0, 11, Not required -sh.1 12 and 1" SW-564 Components already included for spatial interaction LRA-G-1 91159 H-11 20" SW-9 Not required -sh.1 Components already included for spatial interaction LRA-G-191159 H-11 4" SW-567 Not required -sh.1 Components already included for spatial interaction Page 2 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191159 F-13 and F-14 4" SW-12C and 4" Not required -sh. 1 SW-12D Components already included for spatial interaction LRA-G-1 91159 D-1 1 line from refuel Not required -sh. 1 B-7 floor air cond. at Components already valve 23C included for spatial interaction LRA-G-191159 L-8 lines from P-8 Not required -sh. 1 pumps Components already included for spatial interaction LRA-G-1 91159 C-3 SW-565 Not required -sh.2 Components already included for spatial interaction LRA-G-191159 J-3 2 1/2" SW-493 Not r'equired 
-sh.2 Components already included for spatial interaction LRA-G-1 91159 J-6 and H-7 Manual fill from P-63-1 A and -1 B in sh.2 demin water to A turbine building, TK-6-and B EDG cooling 1A in yard LRA-G-1 91159 E-7 20" SW-12 Not required -sh.2 Components already included for spatial interaction LRA-G-191159 E-7 line to SW rad Not required -sh.2 monitor Components already included for spatial interaction LRA-G-1 91159 Q-2 and Q-6 small lines to Not-required 
-sh.3 coolers Components already included for spatial interaction LRA-G-1 91159 N-2 and 0-2 and L-7 3" RCW-8 Not required -sh.3 Components already included for spatial interaction LRA-G-1 91159 M-4 and N-4 Not required -sh.3 Components already included for spatial interaction LRA-G-1 91159 L/M/O/P-1 1 3" RCW-9 and 8" Not required -sh.3 RCW-IA Components already included for spatial interaction LRA-G-191159 G-3 Lines to P-1 8-1A Not required -sh.5 Components already included for spatial interaction Page 3 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191160 F-17 2" line to C-2-1A C-2-1A, E-30-1A, E-31-sh.3 1A, TK-55-1A in reactor building LRA-G-191160 K-14 1 1/2" line to D-2-1A, TK-55-1A, TK-sh.3, system 154-1 A in reactor LRA-G-191160 K-2 building sh.4 LRA-G-1 91160 I-1 1" line at X-22 end of air lines in sh.4, drywell LRA-G-191167 B-9 LRA-G-1 91160 L-8 2" line at X-21 end of air lines in sh.6 drywell LRA-G-191160 D-6 and D-14 2" line to system TK-5-1A and TK-5-1B sh.7, in turbine building LRA-G-191160 C-13 sh.5 LRA-G-191160 F-8, G-9 and F-11, G-1 2 3/4" SA C-3-1 A and C-3-1 B in sh.7 turbine building LRA-G-191160 E-4 TK-55-1A, TK-154-1A sh.8 in reactor building LRA-G-191162 E-2 and E-6, 4" FO-1 Not required -sh.2 D-4, D-5, and E-5 Components already included for spatial interaction LRA-G-1 91163 None; the boundaries sh.1 shown are for an (a)(3)intended function (fire protection).
LRA-G-191163 None; the boundaries sh.2 shown are for an (a)(3)intended function (fire protection).
LRA-G-191163 (no boundaries on this sh.3 drawing)LRA-G-1 91163 (no boundaries on this sh.4 drawing)LRA-G-191164 (no boundaries on this drawing)LRA-G-1 91165 K/L -14/15 flush and sampling Not required -H/J -18 lines to post Components already G -16/19 accident sampling included for spatial panel and radiation interaction monitor Page 4 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-1 91167 B-5 2" line to drywell Not required -equipment sump Components already included for spatial interaction LRA-G-191167 F-2 16" FDW-14 Not required -Components already included for spatial interaction LRA-G-191167 H-2 16" FDW-15 Not required -Components already included for spatial interaction LRA-G-191167 K-3 4" CUW-18 Not required -LRA-G-191178 D-3 Components already sh. 1 included for spatial interaction LRA-G-1 91167 L-3 3/4" PSA Not required -LRA-G-191165 F-14 Components already included for spatial interaction LRA-G-191167 B-13 1" line from TK-55-1A, TK-154-1A accumulator in reactor building LRA-G-1 91168 K-1 2 and L-1 1 4" CS-7C and 7D Not required -Components already included for spatial interaction LRA-G-191168 C-12 3" CS-5B Not required-Components already included for spatial interaction LRA-G-191168 G-1 1 3" CS-5A Not required -LRA-G-191177 C-7 Components already sh.1 included for spatial LRA-G-1 91172 L-3 interaction LRA-G-191168 B-10 2" CST (flushing Not required -line) Components already included for spatial interaction LRA-G-191168 H-8 2" CST (flushing Not required -line) Components already included for spatial interaction LRA-G-1 91168 H-1 1 and H-14 1" pressurizing line Not required -Components already included for spatial interaction LRA-G-1 91169 J-5 purge line from Not required -sh.1 PASS Components already LRA-G-191165 L-16 included for spatial interaction Page 5 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191169 1-6 16" HPCI-14B end of vent line sh.1 LRA-G-191169 H-5 16" FDW-14 Not required -sh.1 Components already included for spatial interaction LRA-G-1 91169 F-7 1" pressurizing line Not required -sh.1 Components already included for spatial interaction LRA-G-191169 E-7 10" HPCI-5 Not required -sh.1 Components already included for spatial interaction LRA-G-191169 H-13/14 1" MSD Not required -sh.1 Components already included for spatial interaction LRA-G-191169 (no boundaries on this sh.2 drawing)LRA-G-1 91170 A-1 and A-9 vent to HVAC end of vent line in exhaust plenum reactor building, LRA-G-191170 E-12 3/4" line Not required -LRA-G-1 91267 C-3 Components already sh.2 included for spatial LRA-G-1 91178 A-5 interaction sh.1 LRA-G-191173 E-11 sh.1 LRA-G-191170 K-13 and K-15 1" line Not required -Components already included for spatial interaction LRA-G-191170 B/D -16/17/1.8 1/2" lines Not required -Components already included for spatial interaction LRA-G-1 91171 H-4 and 1-4 1" line Not required -Components already included for spatial interaction LRA-G-191171 E-5 1 1/2" line Not required -Components already included for spatial interaction Page 6 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-1 91171 G-8, K-8 2" line Not required -Components already included for spatial interaction LRA-G-191171 H-11 2-1/2" SLC-3 Not required -Components already included for spatial interaction LRA-G-191171, H-10 and E-14 1" line Not required -LRA-G-191159 D-9 Components already sh.3 included for spatial interaction LRA-G-1 91171, F-14 3/4" line D-2-1A in reactor LRA-G-191160 B-18 building sh.3 _ _ _'LRA-G-191172 H-4/5, H-10, C-12, C-5, G-9 4" RHR-38A, 4" Not required -RHR-37B, 4" RHR- Components already 36B, 4" RHR-35B included for spatial interaction LRA-G-1 91172 J-3 and J-14 1" pressurizing line Not required -Components already included for spatial interaction LRA-G-1 91172 J-7, K-4, L-4, L-7 4" RHR-26A, 4" Not required -RHR-27A, 4" RHR- Components already 26C, 4" RHR-27C included for spatial interaction LRA-G-1 91172 L-1 1, L-1 3, K-1 0, K-1 3 4" RHR-27D, 4" Not required -RHR-27B, 4" RHR- Components already 26B, 4" RHR-26D included for spatial interaction LRA-G-191172 H-13 4" RHR-4 Not required -Components already included for spatial interaction LRA-G-191172 F-16 line to process Not required -LRA-G-191165 G-17 sampling system Components already included for spatial interaction LRA-G-191173 F-5, F-2, H-4, H-5, G-7, 3" FPC-29B, 8" Not required -sh.1 K-6, 1-13 FPC-4, 3" FPC- Components already 27A, 4" FPC-7B, 8" included for spatial FPC-1 B, 4" FPC- interaction 24, 8" FPC-24, 6" FPC-24, 4" FPC-17, 6" FPC-19, 6" FPC-21 LRA-G-191173 F-7, G-7 2" FPC-39B, 1" Not required -sh.1 FPC-28B Components already included for spatial interaction Page 7 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191173 G-4 4" FPC-40 Not required -sh.1 H-2 Components already LRA-G-191177 included for spatial sh.1 interaction LRA-G-1 91173 D-1 1 2" RWCU bypass Not required -sh.1, Components already LRA-G-1 91178 B-7 included for spatial sh. 1 interaction LRA-G-191173 C-7, C-9, F-7, F-9 piping to sample Not required -sh.2 sink Components already included for spatial interaction LRA-G-1 91173 C-7, C-8, F-7, F-8 drain lines end of drain lines in sh.2 reactor building LRA-G-191174 E-10 4" RCIC-2 Not required -sh.1 Components already included for spatial interaction LRA-G-1 91174 F-1 1 1" pressurizing line Not required -sh.1 Components already included for spatial interaction LRA-G-191174 1-17 1" MSD Not required-sh.1 Components already included for spatial interaction LRA-G-1 91174 B-7, E-11, 1-8, M-10 vent and drain end of vent and drain sh.2 lines lines in reactor building LRA-G-191175, D-1 18" SGT-11 RTF-5 housing in sh.1, reactor building LRA-G-191238 F-7 LRA-G-1 91175 J-8 drain line end of drain line in sh.1, reactor building LRA-G-191177 G-6 sh.1 LRA-G-191175 G-12 1" AC-17 shown on VYI-AC-sh.1 PART 3&4 sh.1 to end in reactor building LRA-G-1 91175 G-15 6" line E-1000-2 in yard and sh.1 E-1000-3 in reactor LRA-G-191175 D-12 building sh.2 LRA-G-191175 1-16 air purge supply end of line to purge sh.1 connection in reactor building LRA-G-191175 C-13 TIP line ends at containment sh.2 penetration X-35E Page 8 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191175 C-10 3/8" line TK-55-1A and E-1000-sh.2. 3 in reactor building LRA-G-191175 D-12 sh.1 LRA-G-191160 J-14 sh.4 LRA-G-191160 H-14 sh.3 LRA-G-191176 E-12 10" CST-2 shown on VYI-CST-sh.1 PART 5 and VYI-HPCI-PART 4 to end in reactor building LRA-G-1 91176 E-12 4" CST-5 P-21-1A, P-21 -1B in sh.1 radwaste building LRA-G-1 91176 E-1 2 6" CST-6 shown on LRA-G-sh.1, 191173 sh 1 to end in LRA-G-191173 J-12 reactor building sh.1 LRA-G-191176 C-12 8" CST-31 end of overflow drain sh.1 line in yard LRA-G-1 91176 C-12 2" CST P4-1A, P4-1B in turbine shl building LRA-G-191176 E-13 4" DW-5 P-63-1A, P-63-1B in sh.1, turbine building LRA-G-191176 F-13 sh.2 LRA-G-191176 E-13 12" CST-7 P4-1A, P4-1B, E-6-1A, sh.1 E-6-1B, E-5-1A, E-5-1B, and condensate demineralizers in turbine building, and P38-1A and P38-1B in Reactor Building LRA-G-191176 F-13 2 1/2" HS-1 and 2" TK-11-1A and TK-6-1A sh.1 HSCR in the yard, E-1 7-1A and E-17-1B in turbine building LRA-G-191177 B-3 and G-3 1 1/2" CS Not required -sh.1 Components already included for spatial interaction LRA-G-191177 B-5, D-6, and C-8 3" CS Not required -sh.1 Components already included for spatial interaction Page 9 of 10 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 2 LRA Drawing Drawing Coordinates of Piping Line Structural Boundary Boundary Flag(s) Description LRA-G-191177 G-5 3" CS Not required -sh.1 Components already LRA-1 91159 sh.2 D-9 included for spatial interaction LRA-G-1 91178 B-5 4" CUW-55 Not required -sh.1 Components already included for spatial interaction LRA-G-1 91178 D-4 4" CUW-1 Not required -sh,1 Components already included for spatial interaction LRA-G-1 91237 (no boundaries on this sh.1 drawing)LRA-G-191237 C-4 and F-3 EXP-TK-SAC-2, SAC-sh.2 2, SCH-2, SP-2, SCC-1 G-191176 sh.2 J-2 in turbine building LRA-G-1 91237 K-3,4, and 5 control room HVAC SRHC-2 and humidifier sh.2 in turbine building G-1 91246 sh.2 LRA-G-1 91238 E-6, G-8 SGT-13 RTF-5 housing in reactor building LRA-G-191238 H-8 supply RSF-1A and 1B housings and plenum duct in turbine building LRA-G-191238, B-12/13 loop seal LG-1 -125-1A and 1B in LRA-G-191159 C-4 Reactor Building sh.2 LRA-G-1 91248 (no boundaries on this drawing)LRA-G-191267 (no boundaries on this sh.1 drawing)LRA-G-1 91267 C-7, D-7, F-7, H-7 Not required -sh.2 Components already LRA-G-191170 E-9 included for spatial interaction LRA-VY-E-75-002 L-15, L-18 1" lines end of vent lines LRA-VY-E-75-002 1-18 1" lines end of bottle connection lines Page 10 of 10 BVY 07-047 Docket 50-271 BVY 07-047 Docket No. 50-271 Attachment 3 Vermont Yankee Nuclear Power Station License Renewal Application Amendment 27 Vernon Hydro-Electric Station Information VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 3 Federal regulations in 18 CFR 12.4(b) define the following responsibilities for the FERC,"Any water power project and the construction, operation, maintenance, use, repair, or modification of any project works are subject to the inspection and the supervision of the Regional Engineer or any other authorized Commission representative for the purpose of: (i) Achieving or protecting the safety, stability, and integrity of the project works or the ability of any project work to function safely for its intended purposes, including navigation, water power development, or other beneficial public uses; or (ii) Otherwise protecting life, health, or property." A letter to Douglas J. Walters, Nuclear Energy Institute, from Christopher I. Grimes, NRC, dated May 5, 1999, states, in part, "It is the staff's opinion that dam inspection and maintenance programs under the jurisdiction of FERC or the Army Corps of Engineers, continued through the period of the license renewal, will be adequate for the purpose of aging management." As a water power project, the Vernon Dam is subject to the Federal Energy Regulatory Commission (FERC) inspection program.This program consists of visual inspections in accordance with FERC guidelines and complies with Title 18 of the Code of Federal Regulations, Conservation of Power and Water Resources, Part 12 (Safety of Water Power Projects and Project Works) and Division of Dam Safety and Inspections Operating Manual. The operational inspection frequency for licensed and exempt low hazard potential dams is biennial.
Reports of operational inspections are filed with the FERC.During the period of extended operation, at least once every five years, VYNPS will confirm that the Vernon Dam owner is performing the required FERC inspections based on a review of the Vernon Dam owner's reports to FERC. VY will document the condition in the Entergy Corrective Actions Program and evaluate operability as described in BVY 96-043 and BVY 97-025 if it is determined that the required inspections are not being performed.
Page 1 of 1 BVY 07-047 Docket 50-271 BVY 07-047 Docket No. 50-271 Attachment 4 Vermont Yankee Nuclear Power Station License Renewal Application Amendment 27 Vernon Hydro-Electric Station RAI 3.6.2.2-N-08-1 Clarification VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 4 RAI 3.6.2.2-N-08-1 Response Clarification Entergy Letter to the NRC, BVY 07-003, LRA Amendment 23, Attachment 3, provided an RAI 3.6.2.2-N-08-1 Response Clarification for Amendment 17, consistent with the Grimes letter and the Peach Bottom license renewal precedent by crediting the FERC dam inspection program to manage the effects of aging on civil and structural elements of the Vernon Hydroelectric Station (VHS). This letter stated;"Since the inspections are not part of a VYNPS aging management program but are conducted by the owner of the dam under FERC oversight, the license renewal application is clarified as follows." After NRC re-evaluation, it was determined that items changed and deleted in Amendment 17 should be re-inserted into the VY LRA as follows: Section 3.5.2.1.5 Insert, "Vernon Dam FERC Inspection" under Aging Management Programs.Table 3.5.1, Item 3.5.1-43 Insert, 'The Vernon Dam FERC Inspection Program manages masonry wall cracking at the Vernon Dam facility." into the Discussion column.Table 3.5.1, Items 3.5.1-45 and 3.5.1-47 Insert, "Loss of material of structural steel components of the Vernon Dam are managed by the Vernon Dam FERC Inspection Program." from the discussion column.Notes for Table 3.5.2-1 through 3.5.2-6 Delete plant-specific note, 505.Delete paragraph; The VeoRno dam is subject to the Federal Regulatory (FERC) inspction program. This program consiMSt Of inspections in accorFdane wiFth FERC gUidolines and complios with Title 18 of the Codo of Federal Regulations, Consorvation o-f Poe4rF and Water Resources, Part 12 (Safety of Watr, PoW.r Projects and Poject VWVork-,,)
and Divisin of, Dam-..Safety and Inspections Operating Manual. accordanc with FERC regulations, the owner has booRn oantod an oemeFOtion fro Part 12. Suboart 1) A6 indicated in NIJREG- 1801 for INAtor control structures, NRC h-as foun'd- that FERO / US Armny Corp of EnRgineers daminpcos and mnaintenance programs6 are acceptable for agin manaRgemet.
In addition, VeFrnon dam personnel conducGt a daily visual inspA-etion of all1 the project facvilities6.
An operations cro attends the plant daily. Vernon Dam egnring staff pe~forms an annual inspection of all the project structures and diVers mnake a hrugnpection once every five years on both upstream and doWnsream sides.Page 1 of 2 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 4 Table 3.5.2-5 Revise line items as shown below (strike-outs 
= deleted, underlined 
= text added).Vernon Dam SRE Carbon Exposed to Loss of Vernon Dam II1.A6-11 3.5.1-47 E structural steel steel weather material FERC (T-21) A, 55, Inspection Vernon Dam SRE Carbon Protected Loss of Vernon Dam II1.A6-11 3.5.1-47 E structural steel steel from weather material FERC (T-21) A, ,O5 Inspection Vernon Dam SRE Carbon Exposed to Loss of Vernon Dam II1.A6-11 3.5.1-47 E structural steel steel fluid material FERC (T-21) A, 50&environment Inspection Vernon Dam SRE Concrete Exposed to Loss of Vernon Dam III.A6-7 3.5.1-45 E external walls fluid material FERC (T-20) A, 505 above/below environment Inspection grade Vernon Dam SRE Concrete Protected None Vernon Dam I, 501 external walls, from weather FERC 5O0 floor slabs and Inspection interior walls Vernon Dam SRE Concrete Exposed to Cracking Vernon Dam III.A6-10 3.5.1-43 E, 505 masonry walls block weather FERC (T-12)Inspection Section A.2.1.31 Replace title; "Section intentionally blank" with "Structures Monitoring 
-Vernon Program." Dam FERC Insert paragraph.
The Vernon dam is subject to the Federal Energy Regulatory Commission (FERC) inspection program. This-program consists of inspections in accordance with FERC guidelines and complies with Title 18 of the Code of Federal Regulations, Conservation of Power and Water, Resources, Part 12 (Safety of Water Power Projects and Project Works) and Division of Dam Safety and Inspections Operating Manual. In accordance with FERC regulations, the owner has been granted an exemption from Part 12, Subpart D. As indicated in NUREG-1 801 for water control structures, NRC has found that FERC / US Army Corp of Engineers dam inspections and maintenance programs are acceptable for aging management.
In addition, Vernon dam personnel conduct a daily visual inspection of all the project facilities.
An operations crew attends the plant daily. Vernon Dam engineering staff performs an annual inspection of all the project structures and divers make a thorough inspection once every five years on both upstream and downstream sides.Tables B-1, B-2 and B-3 Re-insert line items for Structures Monitoring 
-Vernon Dam FERC Inspection.
Section B.1.27 Re-insert subsection B.1.27.3, Vernon Dam FERC Inspection.
Page 2 of 2 BVY 07-047 Docket 50-271 BVY 07-047 Docket No. 50-271 Attachment 5 Vermont Yankee Nuclear Power Station License Renewal Application Amendment 27 Axial Weld Analysis Clarification VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 5 (additions 
= underlined, deletions 
= strikethrough)
Axial Weld Analysis Clarification 4.7 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES .............
4.7-1 4.7.1 Reflood Thermal Shock of the Reactor Vessel Internals 
................
4.7-1 4.7.2 TLAA in BW RVIP Documents 
....................................
4.7-1 4.7.2.1 BWRVIP-05, Reactor Vessel Axial Welds .......................
4.7-1 4.7.2.2 BW RVIP-25, Core Plate .....................................
4.7-1 4.7.2.3 BWRVIP-38, Shroud Support ..............................
4.7-2 4.7.2.4 BWRVIP-47, Lower Plenum Fatigue Analysis ....................
4.7-2 4.7.2.5 BWRVIP-48, Vessel ID Attachment Welds Fatigue Analysis .........
4.7-2 4.7.2.6 BWRVIP-49, Instrument Penetrations Fatigue Analysis ............
4.7-2 4.7.2.7 BWRVIP-74, Reactor Pressure Vessel .........................
4.7-2 4.7.2.8 BW RVIP-76, Core Shroud ...................................
4.7-3 4.7.3 References 
...................................................
4.7-4 4.2.6 Reactor Vessel Axial Weld Failure Probability Applicants must evaluate axially oriented RPV welds to show that their failure frequency remains below the 5x10-6 calculated in the BWRVIP-74 SER. The SER states that an acceptable way to do this is to show that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in Table 1 of that SER. (Note that the data in Table 1 of the BWRVIP-74 SER for license renewal is the same data that is in Table 3 of the supplemental SER for BWRVIP-05.)
Table 4.2-6 reproduces the 32 EFPY and 64 EFPY data from the SERs, and adds the VYNPS data for 32-afid 54 EFPY. The table shows that the VYNPS mean RTNDT is well below that in the SERs, and thus the VYNPS axial weld failure frequency is well below the acceptable limit of 5x1 0.6 Therefore, this analysis has been projected for the period of extended operation per 10 CFR 54.21 (c)(1)(ii).
Page 1 of 4 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 5 (additions 
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= strikethrough)
Table 4.2-6 VYNPS RPV Axial Shell Welds Parameter Description NRC VYNPS Limiting Data for Plant Axial Speolfie Weld Data (Pilgrim)EFPY 54 Initial (un-irradiated)  2' 0 reference temperature (RTNDT), OF Neutron Fluence !.3E !941 5.39E+17 1.48E+18 3 FF = Fluence Factor 0.500 0.305 Weld Copper content, % 0.219%3 0.04%Weld Nickel Content, % 0.996%3 1.00%11.08% _CF = Chemistry Factor 232.04 54.00 Increase in reference 116; 16.5 temperature (ARTNDT),&deg;F 1471= FF x CF Mean adjusted reference 114' 16.5 temperature (ART), 'F -74= RTNDT + ARTNDT 1 Taken from Table 1 of the Safety Evaluation Report for BWRVIP-74. (Also in Table 3 of the Supplemental SER for BWRVIP-05.)
2 The mean adiusted reference temperature minus the initial reference temperature. 
.3 Taken from the supplemental Safety Evaluation Report for BWRVIP-05.
4 Determined using formula and tables from Regulatory Guide 1.99.4.7.2.1 BWRVIP-05, Reactor Vessel Axial Welds BWRVIP-05 justified elimination of reactor vessel circumferential welds from examination.
BWRVIP-74 extended this justification to cover the period of license renewal. This justification involves both circumferential welds and axial welds as it shows examination of the circumferential welds is not required if the axial welds are beingq inspected and the Probability of failure of the axial welds is acceptable.
See Sections 4.2.5 and 4.2.6 for evaluation of the TLAA associated with this issue.4.2.7 References Page 2 of 4 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 5 (additions 
= underlined, deletions 
= strikethrough) 4.2-1 Richards, S. A. (NRC) to J. F. Klapproth, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)," MFN 01 -050, letter dated September 14, 2001.4.2-2 US NRC, Reactor Vessel Integrity Database (RVID), Version 2.0.1, July 2000.4.2-3 Thayer, J.K. (Entergy) to US NRC Document Control Desk, 'Technical Specification Proposed Change No.,263, Extended Power Uprate," BVY 03-80, letter dated September 10, 2003.4.2-4 Wiggins, J. T. (NRC) to L. A. England, "Acceptance for Referencing of Topical Report NEDO-32205, Revision 1, 10 CFR 50 Appendix G Equivalent Margin Analysis for Low Upper-Shelf Energy in BWR/2 through BWR/6 Vessels," letter dated December 8, 1993.4.2-5 Grimes, C.1. (NRC), to C. Terry (BWRVIP Chairman), "Acceptance for Referencing of EPRI Proprietary Report TR-1 13596, BWR Vessel and Internals Project, BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74) and Appendix A, Demonstration of Compliance with the Technical Information requirements of the License Renewal Rule (10CRF54.21)," letter dated October 18, 2001.4.2-6 Tremblay, Jr., L. A. (Entergy) to NRC Document Control Desk, "Response to Request for Additional Information, GL 92-01 -Reactor Vessel Structural*
Integrity," BVY 93-107, letter dated September 24, 1993.4.2-7 Ennis, R. B. (NRC) to M. Kansler (Entergy), "Issuance of Amendment re: Reactor Pressure Vessel Fracture Toughness and Material Surveillance Requirements (TAC NOS. MB8119 and MB8379)," NVY 04-027, letter dated March 29, 2004.4.2-8 DeVincentis, J. M. (Entergy), to USNRC Document Control Desk, "Supplement to Fourth-Interval Inservice Inspection (ISI) Program Plan -Submittal of Relief Request ISI-06," BVY 03-83, letter dated September 25, 2003.4.2-9 Lainas, G. C. (NRC), to C. Terry (Niagara Mohawk Power Company, BWRVIP Chairman), Final Safety Evaluation of the BWRVIP Vessel and Internals Project BWRVIP-05 Report, (TAC No. M93925), July 28,1998.4.2-10 Strosnider, J. R. (NRC) to C. Terry, (Niagara Mohawk Power Company, BWRVIP Chairman), Supplement to Final Safety Evaluation of the BWR Vessel and Internals Proiect BWRVIP-05 Report (TAC NO. MA3395), March 7, 2000 A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability The BWRVIP recommendations for inspection of reactor vessel shell welds (BWRVIP-05, Reference A.2-10) are based on generic analyses supporting an NRC SER (References A.2-11 and A.2-12) conclusion that the generic-plant axial weld failure rate is no more than 5 x 10-6 per reactor year as calculated in the BWRVIP-74 SER (Reference A.2-13). BWRVIP-05 showed that this axial weld failure rate is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described above.Page 3 of 4 BVY 07-047 Docket 50-271 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION ATTACHMENT 5 (additions 
= underlined, deletions 
= strikethrough)
The basis for this relief requost was, a plant spocifi analysis that Showo-fd thelmting" conditioa failureprbability for the VYNPS circumnferential welds at the end ofth oignl operating teerm. wiere less;1 than the- value-s;cluae in the R ARVIP 05 SER.The BWARVIP 05 SER concluded that the reactor vessel failure frequency du 4ofalr Of the limitinl axial welds i the BVVR fleet at the end of 4 , years of operation is le th-=pe r1x1 eactor year. This failure frequencGy is dependent upon give assumptions of flaw density, distIbution, ad location.
T-he failur~e frequency also assum-e~s thlat essentially 1100-9 of the reactor vessel axial welds wel! be inspected.
The BWRVIP-74 SER states it is acceptable for license renewal to show that the mean RTNDT Of the limiting beltline axial weld at the end of the period of extended operation is less than the limiting value given in thate SERs for BWRVIP-74 and BWR.VIP 05. The projected 54 EFPY mean RTNDT values for VYNPS (16.5 OF) is aFe-less than the limiting 64 EoPY RTNDTivalue (1114 OF) in the analysis peko"med by the NRCotaff (Table 12-.6-6 of the BWRVIP-0674 SER). As such, this TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).
Page 4 of 4 BVY 07-047 Docket 50-271}}