ML061080367

License Renewal Application - Annual Update Information Required by 10CFR54.21(b)

ML061080367
Person / Time
Site:	Nine Mile Point
Issue date:	12/20/2005
From:	O'Connor T Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMPIL 2009, TAC MC3272, TAC MC3273
Download: ML061080367 (25)

Text

,Constellation Energy-Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 1309$

December 20, 2005 NMP1L 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Nine Mile Point Units 1 and 2 Docket Nos. 50-220 and 50-410 Facility Operating License Nos. DPR-63 and NPF-69 License Renewal Application - Annual Update Information Required by 10CFR54.21(b) (TAC Nos. MC3272 and MC3273)

Gentlemen:

By letter dated May 26, 2004, Nine Mile Point Nuclear Station, LLC (NMPNS) submitted a License Renewal Application (LRA) for the operating licenses of Nine Mile Point Units 1 and 2.

In accordance with 10CFR 54.21(b), Nine Mile Point Nuclear Station (NMPNS) is required to submit a summary of the current licensing basis (CLB) changes that have occurred during the NRC review of the Application that materially affects the contents of the Application, including the Updated Final Safety Analysis Report (UFSAR) supplement (Unit 1) and Updated Safety Analysi s Report (USAR) supplement (Unit 2).

NMPN.S has completed a review of pertinent documents, including the UFSAR and USAR supplements, and identified changes that materially affect the Application. provides a description of each of the changes to the Application due to modifications. Attachment 2 provides Application changes due to new or revised analyses.

If you have any questions about this submittal, please contact David Del rio, NMPNS License Renewal Project Manager, at (315) 349-7141.

o

'Co or Plani neral Manager JAS/MSL/sac (f~ o Qq

Page 2 NMP1IL 2009 STATE OF NEW YORK

TO WIT:

COUNTY OF OSWEGO I, Timothy J. O'Connor, being duly sworn, state that I am Nine Mile Point Plant General Manager, and that I am duly authorized to execute and file this information on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this submittal are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Nine Mile Point employees and/or consultants. Such information has been reviewed in acc ance with company practice and I believe it to be reliable.

imothyln.

'Connor Plant G ral Manager Subscribed and sworn before me, a Notary Public in and for the State of New York and County of Oswego, this cRO44 day of, eneA J

_, 2005.

WITNESS my Hand and Notarial Seal:

r -

i 0f 0

Nay Pubnc te Sta N

York Ouwgotounty Rsto.

No. 01 Jp "54 My Conmsi6 Expres I CIto Not My Commission Expires:

date Attachments:

1: Application changes due to modifications.

2: Application changes due to new or revised analyses.

cc:

Mr. S. J. Collins, NRC Regional Administrator, Region I Mr. L. M. Cline, NRC Senior Resident Inspector Mr. T. G. Colburn, Senior Project Manager, NRR Mr. N. B. Le, License Renewal Project Manager, NRR Mr. J. P. Spath, NYSERDA

ATTACHMENT 1 to NMPIL 2009 The following are nine (9) modifications to NMP1 and NMP2, since the submittal of the initial LRA, that materially affect the Amended License renewal Application (ALRA).The first five (5) of these modifications are for NMPl and the last four (4) are for NMP2.

Revisions to the existing ALRA are shown with italics for additions and strilkethreughs for deletions.

MODIFICATION 1 -NMP1 A Zinc Injection System was installed at NMPL. The purpose is to inject zinc ions into the reactor coolant via the Feedwater System to reduce the corrosion of stainless steel surfaces in the reactor coolant system. This lowers radiation levels due to 60Co deposition.

The NSR system is not within scope of license renewal since it does not meet any of the three scoping criteria which would bring it into scope as follows:

54.4(a)(1) - The system is non-safety-related (NSR); therefore, this criterion does not apply.

54.4(a)(2) -

1. The system does not contain any NSR components credited in the current licensing basis to accomplish safety functions (i.e., protection from missiles, overhead cranes, flooding, or high energy line breaks).

2. The system does not contain NSR components that are within the boundary between a connection to safety-related components and the first seismic or equivalent anchor.

3. The system does not contain any NSR components that could spatially interact (via spray, leakage, jet impingement, or provision of any other harsh environment) with safety-related components.

54.4(a)(3) - The system does not contain any NSR components that are relied upon to provide compliance with the regulations for fire protection, environmental qualification, anticipated transients without scram, or station blackout.

The change to the ALRA as a result of the installation of this system is its addition to Table 2.2-1, NMP1 Plant Level Scoping Results, on ALRA p. 2.2-6 as shown below.

Mechanical Systems System or Commodity License Reneoal?

Comments Turbine Building HVAC System Yes (Section 2.3.3.A.26)

Yes Zinc Injection System No 1 of 14

MODIFICATION 2-NMP1 A NMP1 Containment Spray pump was replaced. The material of the original pump casing was gray cast iron that has been replaced with stainless steel. As a result of this modification, the changes to ALRA p. 3.2-38 are as shown below.

Table 3.2.2.A-1 Engineered Safety Features Systems NMP1 Containment Spray System-Summary of Agin Management Evaluation n

Aging Effect NUREG-Component Type Fntended Material Environment Requiring Aging Vanagement o1801 Table I Notes Functionanagement Program Volume 2 Item ManaItemen Pumps (cont'd)

PB (cont'd)

Gray Cast Iron Raw Water Loss of Material Open-Cycle Cooling VII.Cl.5-a 3.3.1.A-29 A

Water System Progran Selective Leaching of Materials Program Treated Water, Loss of Material One-Time Inspection VII.C2.3-a 3.3.1.A-15 E

temperature Program

<F140wF, Low Water Chemist Flowv Control Program Selective Leaching of VII.C2.3-a 3.3.1.A-29 A

Materials Program Wrought Treated Water, Loss of Material One-Time Inspection VIII.E.5-b 3.4.1.A-02 D

Austenitic temperature Progra Stainless Steel

< 140'F, Low Water Chemis Flowv Control Program 2 of 14

I MODIFICATION 3-NMP1 The carbon steel pressure safety valve thit.sens.es as the air release valve for the NMMP1 diesl-drjven fire pump was replaced with a cast iron model. The resultant changes to the ALRA from this modification are on p. 3.3-144 as shown below.

Table 3.3.2.A-8 Auxiliary Systems NMPl Fire Detection and Protection System-Summary of Aging Management Evaluation IneddAging Effect NUREG-FupntTp nctionde Material Environment Requiring Aging Management 1801 TableI Noe Function lManaement Program Volume 2 Item M

anag mentItem Valves LBS Carbon or Low Air Loss of Material One-Time Inspection VII.H2.2-a 3.3..A-05 A

PB Alloy Steel (Yield Programr Strength < 100 Ksi)

Dried Air or Gas None None None Raw Water, Loss of Material Fire Water System VII.G.6-b 3.3..A-21 A

Low Flow Program Copper Alloys Raw Water, Loss of Material Fire Water System VII.G.6-b 3.3..A-21 A

(Zinc <15%)

Low Flow Program Copper Alloys Raw Water, Loss of Material Fire Water System VII.G.6-b 3.3. LA-21 A

(Zinc > 15%) and Low Flow Program Aluminum Bronze Selective Leaching of VII.C1.2-a 3.3.1.A-29 A

Materials Program Gray Cast Iron Raw Water, Loss of Material Fire Water System VII.G.6-b 3.3.1.A-21 A

Low Flow ProEgram Selective Leaching of VII.Cl.5-a 3.3.1.A-29 C

Materials Program I

3 of 14

MODIFICATION 4-NMP1 The carbon steel piping and valves associated with one of the main NMPl Feedwater Pump's lube oil system were replaced with stainless steel. The resultant ALRA changes from this modification apply to pp. 3.4-40 and 3.4-42 for Piping and Fittings and Valves, respectively, and are as shown below and on the following page.

Table 3.4.2.A-2 Steam and Power Conversion System NNIP1l Feedwater/High Pressure Coolant Injec ion System - Summ ry of Aging Management Evaluation Aging Effect NUREG-Component Type Intended Material Environment Requiring Agg Pogem 1801 Table 1 Notes F oManagement Program Volume 2 Item

____Item Piping and LBS Carbon Steel, Low Lubricating Oil None None None Fittings SIA Alloy Steel (Yield Strength < 100 Treated Water Cumulative TLAA. evaluated in VIII.D2.1-c 3.4.1.A-01 A

Ksi) or Steam, Fatigue Damage accordance with 10 temperature CRF 54.21(c)

>212°F, but Loss of Material Flow-Accelerated VIII.D2.1-a 3.4.l.A-06 A

< 4820 F Corrosion Program Loss of Material One-Time Inspection VIII.D2.1-b 3.4.1.A-02 B

(cont'd)

Program Water Chemistry Control Program Wrought Lubricating Oil None None None Austenitic Stainless Steel PB Carbon Steel, Low Treated Water Cumulative TLAA. evaluated in IV.Cl.l-d 3.1.1.A-01 A, 16 PH Alloy Steel (Yield or Steam, Fatigue Damage accordance with 10 Strength < 100 temperature CFR 54.21(c)

Ksi)

Ž212'F, but Loss of Material Flow-Accelerated IV.Cl.1-c 3.1.1.A-25 A, 16

< 4820 F Corrosion Program One-Time Inspection H, 16 Program Water Chemist

_nt-ml Pmr_1m 4 of 14

Table 3.4.2.A-2 Steam and Power Conversion System NAMP1 Feedwater/High Pressure Coolant Injection System - Summary of Aging Management Evaluation A

A l....1?t-.-&

NUREG-1 Copnn ye Intended Maeil Evrnet Rqiig Aging Management 1801 Table 1 oe Component Typl Function Environment Requiring l

Program Volume 2 Item Noe 1~¶anagement Program Item Valves (cont'd)

LBS Carbon Steel, Low Treated Water Cumulative TLAA. evaluated in VIII.D2.1-c 3.4.l.A-0I C_ 7 SIA (cont'd)

Alloy Steel (Yield or Steam, Fatigue Damage accordance with 10 Strength < 100 temperature CFR 54.21(c)

Ksi) (cont'd) 2 212 0F, but Loss of Material Flow-Accelerated V11I.D2.2-a 3.4.1.A-06 A

< 4820 F Corrosion Program One-Time Inspection V11I.D2.2-a 3.4.1.A-02 B

Program Water Chemistry Control Program Treated Water Cumulative TLAA, evaluated in VIII.D2.1-c 3.4.1.A-01 C, 7 or Steam, Fatigue Damage accordance with 10 temperature CFR 54.21(c)

Ž212'F, but Loss of Material One-Time Inspection VIII.D2.2-b 3.4.1.A-02 B

< 4820F, Low Prom Flow Water Chemistry Control Program Wrought Lubricating Oil None None None Austenitic Stainless Steel PB Carbon Steel, Low Treated Water Cumulative TLAA. evaluated in IV.Cl.3-d 3.1.1.A-01 A, 16 Alloy Steel (Yield or Steam, Fatigue Damage accordance with 10 Strength < 100 temperature CFR 54.21(c)

Ksi) 12'21F, but Loss of Material Flow-Accelerated IV.Cl.3-a 3.1. LA-25 A, 16

< 4820F Corrosion Program 5 of 14

MODIFICATION 5-NMP1 One of the carhon steel NMP1 Control Rod nrive Plmmps was reptla ed wth Anf e nnpvinft a stainle P

te r-asM T Thb reultahntf ^Unhni to ALRA p. 3.1-68 are as shown below.

Table 3.1.2.A-5 Reactor Vessel. Internals, and Reactor Coolant System NMP1 Control Rod Drive Svstem - Summarv of Aging Management Evaluation IneddAging Effect Agn aaeet NUREG-Component Type Intended Material Environment Requiring Agn Progem 1801 Table 1 Notes Funtin anagement Program Volume 2 Item

__________________Item_

Piping and PB (cont'd)

WVrought Treated Water, Cracking ASME Section XI IV.C1.l-i 3.1.1.A-07 B

Fittings (cont'd)

Austenitic temperature Inservice Inspection Stainless Steel 21400F, but (Subsections IWB.

< 212'F, Low IWC. IWD) Program Floow One-Time Inspection Program Water Chemistry Control Program One-Time Inspection IV.Cl.1-i 3.1.1.A-07

_, 24 Program Water Chemistry Control Program Pumps LBS Carbon or Low Treated Water, Loss of Material One-Time Inspection VIII.E.3-a 3.4.1.A-02 B

Alloy Steel temperature Program (Yield Strength

>140'F, but Water Chemis

< 100 Ksi)

< 2120 F Conatl Proerami Wrought Treated Water, Cracking One-Time Inspection H

Austenitic temperature Program Stainless Steel

>140'F, but Water Chemist

< 2120 F Control Proaram Tank SIA Carbon or Low Treated Water, Loss of Material One-Time Inspection VIII.E.5-a 3.4. LA-02 B

Alloy Steel temperature Program (Yield Strength

< 1400F, Low Water Chemistry

< 100 Ksi)

Flow aCntrol Pmrnh m_

6 of 14

MoniFIrATIAN 6-NMP2 A filtration skid for Reactor Recirculation Pump seal water was installed at NMP2. The components in this skid are within scope of license renewal since they meet the 54.4(a)(2) criterion with the intended function of Leakage Boundary (Spatial). The ALRA changes that follow are applicable with the installation of this equipment. The piping and fittings included with the skid are encompassed by the existing Piping and Fittings entries in the AMR table.

From p. 2.3-22 of the ALRA for scoping of the filters, the resultant table changes are as shown below.

Table 2.3.1.13.4-1 NMP2 Reactor Recirculation System Component Type Intended Functions Closure Bolting Pressure Boundary Filters Leakage Boundary (Spatial)

Pressure Boundary Piping and Fittings Leakage Boundary (Spatial)

Structural Integrity (Attached)

Pumps Pressure Boundary Radiation Collars Shielding Restriction Orifices Throttle, Pressure Boundary Seal Coolers Heat Transfer, Pressure Boundary Pressure Boundary Valves Leakage Boundary (Spatial)

Structural Integrity (Attached) 7 of 14

From ALRA p. 3.1-96, for the AMR of the Filters, the table changes are as shown below.

Table 3.1.2.1B-4 Reactor Vessel, Internals, and Reactor Coolant System NMrP2 Reactor Recirculation System - Summary of Aging Management Evaluation II I

I Component Type Intended Function Material Environment Aging Effect Requiring Management Agina Management Program NUREG-1801 Volume 2 Item Table 1 Item Notes I

I Closure Bolting PB Carbon or Low Alloy Steel (Yield Strength

>1 00 Ksi)

I I

Closure Bolting for Non-Borated Water Systems with operating temperatures l221°F Cumulative Fatigue Damage TLAA, evaluated in accordance with 10 CFR 54.21 (c)

V.C1 2-f 31.1.1.- 01 A

_.I IV.CI.3-g 3.1.11.B-01 A

Loss of Material r Bolting Integrity IV.CI.2-d 31.1.13-26 I Proeram IV.CI.3-e 3.1.1.B-26IA Loss of Preload Bolting Integrity Program IV.Cl.2-e 311.1 -26 A

IV.Cl.3-f 3.1.1.13-26 A

Wrought Austenitic Stainless Steel 4

I I

4-Closure Bolting for Non-Borated Water Systems with operating temperatures l221°F Cumulative Fatigue Damage TLAA. evaluated in accordance with 10 CFR 54.21 (c)

F I

I 1

I Loss of Preload Bolting Integrity Program IV.CI.2-e 3.1.1.13-26 H

Extemal Surfaces PB Cast Austenitic Air None None None Stainless Steel PB Nickel Based Air None None None Alloys LBS Wrought Air None None None PB Austenitic SIA Stainless Steel Filters LBS Wrought Treated Water, Loss of Material One Time Inspection VIII.E.5-b 3.4.1.13-02 D

Austenitic temperature Progra Stainless Steel

< 140 0F Water Chemistry Control Programr Piping and LBS Wrought Treated Water, Loss of Material One-Time Inspection VIII.E.5-b 3.4.1.13-02 D

Fittings PB Austenitic temperature Program SIA Stainless Steel

< 140OF Water Chemistry Control Proaram 8 of 14

From ALRA p. 3.1-102, for AMR of the associated cast austenitic stainless steel valves, the table changes are as shown below.

Table 3.1.2.B-4 Reactor Vessel. Internals. and R aetor (nnolnt SRytrni NMP2 Reactor Recirculation System - Summary of Aging Management Evaluation IneddAging Effect NUREG-Component Type ntended Material Environment Requiring Agin2 Management 1801 Table I Notes Function IaaeetProaram Volume 2 Item Management Item Seal Coolers HT Wrought Treated Water or Cracking One-Time Inspection H

PB Austenitic

Steam, Progra Stainless Steel temperature Water Chemisty

_A82 0F Control Program Cumulative TLAA. evaluated in IV.CI.2-a 3.1.1.B-0I C

Fatigue Damage accordance with 10 CFR 54.21 (c)

Valves LBS Cast Austenitic Treated Water, Loss of Material One-Time Inspection VIII.E.5-b 3.4.1.B-02 D

Stainless Steel temperature Program

<1400F Water Chemistry Control Program LBS Wrought Treated Water, Loss of Material One-Time Inspection VIII.E.2-b 3.4. 1.B-02 D

PB Austenitic temperature Program SIA Stainless Steel

< 140'F, Low Water Chemistr Flowv Control Program Treated Water, Cracking One-Time Inspection IV.C1.1-i 3.1.1.B-07 E, 24 temperature Program

>140'F, but Water Chemistry F 212'F, Low Control Program F low 9 of 14

MODIFICATION 7-NMP2 Severa1 NMP2 Tiirhinn RBilding SRimp Puimpn that had rhimininlwn casngs were replaced with new pumps having cast ;ren"c aings.

With the environment of Raw Water, the Selective Leaching of Materials Program will now be credited. The resultant changes to p.

3.3-58 to add this program are as shown below.

Aging Mana2ement Pro2rams The following aging management programs manage the aging effects for the NMP2 Floor and Equipment Drains System components:

10 CFR 50 Appendix J Program

• Bolting Integrity Program

• One-Time Inspection Program

• Preventive Maintenance Program

• Selective Leaching of Materials Program

• Systems Walkdown Program 10 of 14

The resultant AMR table changes on p. 3.3-241 are as shown below.

Table 3.3.2.B-14 Auxiliary Systems He A.

INM2 Floor and Egul ment Drains System - Summary of Aging Management Evaluation IneddAging Effect NUREG-Component Typeu Atinm ManaEnment 1801 Table I Notes Function Matrilnnarometmequrng rpr Volume 2 Item Management PItem Pumps LBS Aluminum Raw Water Loss of Material One-Time Inspection Pogram Gray Cast Iron Raw Water Loss of Material Preventive J

Maintenance Program Selective Leaching of Materials Program PB Wrought Austenitic Treated Water, Loss of Material One Time Inspection J

Stainless Steel temperature

<1400F Treated Water, Cracking One-Time Inspection J

temperature Program

Ž140'F, but

< 2120F Orifices FC Wrought Austenitic Air, Moisture or Loss of Material One-Time Inspection J

PB Stainless Steel

Wetting, Program temperature

__140OF Spray Nozzle PB Wrought Austenitic Air, Moisture or Loss of Material One-Time Inspection J

SPR Stainless Steel

Wetting, Program temperature

<140 0F__

11 of 14

MODIFICATION 8-NMP2 Several NMP2 Feedwater System check valves that were originally fabricated of carbon steel have been replaced by stainless steel check valves. The resultant changes to ALRA pp. 3.4-15 and 3.4-65 are as shown below.

3.4.2.B.3 NMP2 FEEDWVATER SYSTEM Materials The materials of construction for the NMP2 Feedwater System components are:

• Carbon Steel, Low Alloy Steel (Yield Strength < 100 Ksi)

• Wrought Austenitic Stainless Steel Table 3.4.2.B-3 Steam and Power Conversion System NMP2 Feedwater System-Summary of Agin Management Evaluation IneddAging Effect NUREG-Component Type et 11n Material Environagement 1801 Table I Notes Function Matralnnvroeetmequrng rar Volume 2 Item Management rItem Valves (cont'd)

PB Carbon Steel, Low Treated Water Cumulative TLAA. evaluated in IV.C1.3-d 3.1.1.B-01 Al 16 PH Alloy Steel (Yield or Stearn, Fatigue Damage accordance with 10 Strength < 100 temperature CFR 54.21(c)

Ksi) t2l20F, but Loss of Material Flow-Accelerated IV.C1.3-a 3.1.11.B-25 A_ 16

< 4820F Corrosion Program Treated Water Cumulative TLAA, evaluated in IV.C1.3-d 3.1.1.B-01 A, 16 or Steam, Fatigue Damage accordance with 10 temperature CFR 54.21(c)

>212'F, but Loss of Material Flow-Accelerated IV.C1.3-a 3.1.1.13-25 A, 16

< 4821F, Low Corrosion Program Flow Wrought Treated Water, Loss of Material One Time Inspection VIII.E.5-b 3.4.1.13-02 D

Austenitic temperature Program Stainless Steel

< 1401F Water Chemis ControlProgram r

jIi 12 of 14

MODIFICATION 9-NMP2 Several NMP2 Turbine Buildingt Closed L.nn Cooling Water relief valves that had been fabrrcated ofcrbohnn stPe1 were replaced with new valves fabricated of cast austenitic stainless steel. The resultant changes to ALRA p. 3.3-83 are as shown below and the changes to p. 3.3-307 are as shown on the next page.

3.3.2.B.40 NMIP2 TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM Materials The materials of construction for the NMP2 Turbine Building Closed Loop Cooling Water System components are:

Carbon or Low Alloy Steel (Yield Strength 2100 Ksi)

Carbon or Low Alloy Steel (Yield Strength < 100 Ksi)

Cast Austenitic Stainless Steel 13 of 14

Table 3.3.2.B-40 Auxiliary Systems NMP2 Turbine Building Closed Loop Cooling Water System-Summary of Aging Management Evaluation l

.n I

[

I Aging Effect [

Aging NUREG-1801 1 1

TypeFuncti Material Environment Requiring Management Volume 2 Table 1 Item Notes Type_

Function Management Program Item Bolting LBS Carbon or Low Air Loss of Bolting VIT.T.1-b 3.3.1.B-05 A

Alloy Steel Material Integrity (Yield Strength Program 2l00 Ksi External LBS Carbon or Low Air Loss of Systems VII.I.l-b 3.3.1.B-05 A

Surfaces Alloy Steel Material Walkdown (Yield Strength Program

< 100 Ksi Heat LBS Carbon or Low Demineralized Loss of Closed-Cycle VII.C2.4-a 3.3.1.B-15 C

Exchangers Alloy Steel Untreated Material Cooling Water (Yield Strength WVater Syste

< 100 Ksi Program Piping and LBS Carbon or Low Demineralized Loss of Closed-Cycle VII.C2.1-a 3.3.1.B-15 A

Fittings Alloy Steel Untreated Material Cooling Water (Yield Strength Water System

< 100 Ksi

_gram l

Valves LBS Carbon or Low Demineralized Loss of Closed-Cycle VII.C2.2-a 3.3.1.B-15 A

Alloy Steel Untreated Material Cooling Water (Yield Strength Water Syste

< 100 Ksi Program Cast Austenitic Demineralized Loss of Closed-Cycle VII.C2.2-a 3.3.1.B-15 A

Stainless Steel Untreated Material Cooling Water Water Syste Program 14 of 14

ATTACHMENT 2 to NMP1L 2009 A review of NMPNS new and revised analyses completed since submittal of the LRA resulted in changes to ALRA Time-Limited Aging Analysis (TLAA) sections. Of the following four (4) analyses, three (3) are associated with NMPl and the fourth is associated with NMP2.

Revisions to the existing ALRA are shown with italics for additions and stri4kethreughs for deletions.

ANALYSIS 1 -NMP1 A revised analysis related to ALRA Section 4.6.1, Torus Shell and Vent System Fatigue Analysis (NMP1 Only), was performed that resulted in revised information in that section of the ALRA.. The revisions on p. 4.6-2 of the ALRA are as shown below.

Disposition:

§54.21(c)(1)(i) - The analyses remain valid for the period of extended operation; AND

§54.21(c)(1)(ii) - The analyses have been projected to the end of the period of extended operation.

The design basis accident (DBA) was identified as the major load contributing to the fatigue evaluation for all high stress locations in the vent header system. The controlling usage factor was 0.86 at the vent header/vent pipe spherical intersection. Provided that a DBA (the major contributor to fatigue) does not occur during the original 40-year license period, this usage factor will not be exceeded during the period of extended operation; therefore, the NMP1 vent header fatigue usage analyses remain valid in accordance 'with

§54.21(c)(1)(i).

1 of9

ANALYSIS 2-NMPI A revised analysis related to ALRA Section 4.6.5, Fatigue of Primary Containment PenetratiDns, was performed that resulted in revised information in that section of the ALRA. Changes to ALRA Table 4.1-1 on p. 4.1-3 are as shown on page 6 of this attachment. The changes to ALRA Section 4.6.5 on p. 4.6-10 of the ALRA are as shown below.

4.6.5 FATIGUE OF PRIMARY CONTAINMENT PENETRATIONS NMPI - Summary Description The NMP1 drywell was designed as a Class B Vessel in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1965 Edition (ASME Section III, 1965). The 1965 Edition of the ASME Section III B&PV Code did not require fatigue analysis of Class B vessels. The dryw ell penetrations were considered an extension of the drywell and thus did not require fatigue analysis. For NMP1, fatigue of torus penetrations was addressed in the same analysis as the torus attached piping, the "Plant Unique Analysis Report of the Torus Attached Piping for Nine Mile Point Unit 1 Nuclear Generating Station," which was transmitted to the NRC in a letter dated May 22, 1984 (Reference 4.8-61). Additional fatigue analyses were performed for a number of specific penetrations subsequent to the submittal of Reference 4.8-61. These analyses were performed in accordance with ASME Section. III, 1977 Edition, through the Summer 1977 Addenda. Fatigue analyses were performed for the safety/relief valve (SRV) penetration (where the SRV line penetrates the vent header spherical intersection) and torus attached piping penetrations.

Changes to p. 4.6-11 are as shown below starting with the insertion of a new paragraph after the first full paragraph on the page.

In addition to the bounding penetration fatigue evaluations reported in the Plant Unique Analysis Report (Reference 4.8-61), specific stress and fatigue analyses were performed for 26 torus piping penetrations. These analyses were performed in accordance with the 1977 ASME Code Section III, Subsection NC through the 1977 Summer Addenda. T he loading conditions considered in the calculations were developed from the 27 possible cases listed in Table 2-5 of the Mark I Containment Program Structural Acceptance Criteria Plant Unique Application Guide, NEDO-24583-1 (Reference 4.8-47). For the fatigue evaluation, the load case having the maximum alternating stress was assumed for each fatigue cycle. This load case included deadweight, thermal, safe shutdown earthquake, and condensation oscillation loads. The number of cycles for all load events was estimated to be 10,000, including accident, earthquake, and normal operating conditions. The fatigue usage determined by the analyses is considered to be conservative because the stress used for each cycle includes stresses from accident and earthquake conditions which do not normally occur. By contrast, the only types of cyclic loads that occur during normal operation are safety/relief valve discharges. The alternating stresses resulting from SRV discharges alone is a fraction of the stress for :he load case assumed in the analyses; therefore, SRV discharge will result in negligible fatigue usage per cycle. Also, only 520 SRV discharges are projected for NMP1 which 2of9

will result in 4,500 effective stress cycles compared to the 10,000 assumed in the analyses. The subject calculations were dispositioned in 2003 for a hypothetical 5 degree increase in torus water temperature, resulting in a minimal change to the calculated fatigue usages. Therefore, since the actual fatigue usage for these penetrations through the period of extended operation will be low compared to the calculated values, the analyses remain valid for the period of extended operation in accordance with

§54.21(c)(l)(i).

A CUF was not reported for the torus attached piping penetrations. However, considering the major loads listed above to be significant contributors to fatigue usage, each causing a load cycle equal to the maximum load, there are 951 significant loading events during the 40-year design life. Section 3.4.7 of the PUAR indicates that other loads such as normal SRV actuation, intermediate break accident condensation oscillation (IBA.CO), and chugging can cause up to 10,450 cycles, but only at greatly reduced stress levels; therefore, these loads are assumed to produce a negligible contribution to the overall fatigue usage of the penetrations. Assuming 10,000 cycles of the maximum load equates to a CUF of 1.0, the fatigue CUF resulting from these loads is 0.0951. Projecting this to 60 years by multiplying by 1.5, the 60-year fatigue usage is 0.143, which is well below the code allowable of 1.0.

Disposition:

§54.21(c)(1)(i) - The analyses remain valid for the periDd of extended operation; OR §54.21(c)(1)(ii) - The analyses have been projected to the end of the period of extended operation.

The paragraph that starts on the bottom of p. 4.6-11 and carries over to p. 4.6-12 is revised as shown below.

For the fatigue analyses reported in the Plant-Unique Analysis Report (Reference 4.8-61),

the number of anticipated significant transient cycles for a 40-year life divided by the maximum number of allowable cycles for the transient producing the maximum stress was used to estimate the 40-year design CUF. Linear projection of this CUF to 60 years results in a CUF far below the allowable. Therefore, the fatigue analysis of the torus attached piping penetrations has been projected to the end of the period of extended operation in accordance with §54.21(c)(1)(ii). For the additional fatigue analyses of specific penetrations, the actual fatigue usage based on transients that occur during normal operation is predicted to be negligible. Therefore, these analyses remain valid for the period of extended operation in accordance with §54.21 (c)(1)(i).

Additionally, on p. 4.8-6, Reference 4.8-47 is changed as shown below.

4.8-47 Mark 1 Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide, Task 3.1.3, NEDO 24583-1 79 NED 125, Class 1, October 1979 3 of9

ANALYSIS 3-NMP1 A new fatigue flaw growth analysis was performed for an NMP1 Reactor Water Cleanup System weld overlay repair that generated a new TLAA. The new TLAA comprises new ALRA Section 4.7.5. The resultant changes to ALRA p. 4.1-3 are as shown on page 6 of this attachment. The new Section 4.7.5 that would start on new ALRA p. 4.7-8 is as shown below.

4.7.5 REACTOR WATER CLEANUP SYSTEM WELD OVERLAY FATIGUE FLAW GROWTH EVALUATIONS (NMP1 ONLY)

Fatigue crack growth analyses have been performed for two weld overlays in the reactor water cleanup system. The weld overlay design is in accordance with Code Case N-504 (Weld 33-FW-22) and N-504-2 (Weld 33-FW-23A). The repaired welds are located at the inlet nozzle of the regenerative heat exchanger and the transition pipe between the upper and lower shells of the regenerative heat exchanger, respectively. The weld overlays consist of IGSCC-resistant austenitic stainless steel material and, thus, are nDt susceptible to continued IGSCC crack propagation. However, the first 1/16" thick layer of weld metal deposited is not assumed to be IGSCC-resistant due to weld dilution; thus, it is assumed to be cracked. A fatigue crack growth analysis was performed in accordance with ASME Section XI, Appendix C, with the crack propagating into the overlay from the hypothetical 1/16" deep crack. The acceptance criteria for fatigue crack growth analyses are based on the depth to thickness ratios from Tables IWB-3641-5 and IWB-3641-6.

Disposition:

§54.21(c)(1)(i) - The analyses remain valid for the period of extended operation.

The analysis for weld 33-FW-22 assumed 154 startup/shutdown cycles. The flaw depth to weld overlay thickness ratio remained acceptable at 0.22 versus an allowable of 0.29.

The number of cycles assumed was based on an estimate of 7 cycles/year for 22 years (12 remaining years of the original license period plus 10 additional years). There have been 34 startup/shutdown cycles since the beginning of 1997; this overlay was installed in May, 1997. Therefore, the weld overlay could experience 120 more cycles before exceeding the assumptions of the analysis. During the last ten years of operation, NIVMPl has been experiencing startups and shutdown cycles at a rate of approximately 4/year.

Therefore, it is expected that the actual number of startups/shutdowns will remain below 120 cycles in the next 24 years (4 years original license plus 20 year period of extended operation). Therefore, the fatigue crack growth analysis for the weld 33-FW-22 overlay is expected to remain valid for the period of extended operation.

The analysis for weld 33-FW-23A was performed in 2005 and assumed 168 startup/shutdown cycles based on 7 cycles/year for 24 years (the 4 remaining years of original license plus the 20 year period of extended operation). The final flaw depth to thickness ratio was 0.17 compared to an allowable of 0.42; thus, the weld was acceptable for the cycles analyzed. Therefore, since this analysis was performed considering the period of extended operation, the analysis remains valid for the period of extended operation.

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This TLAA results in the addition of new ALRA Section A1.2.5.2 starting on p. A1-34 as shown below.

A1.2.5.2 REACTOR WATER CLEANUP SYSTEM WELD OVERLAY FATIGUE FILAW GROWTH EVALUATIONS Fatigue crack growth analyses have been performed for two weld overlays in the reactor water cleanup system. The repaired welds are located at the inlet nozzle of the regenerative heat exchanger and the transition pipe between the upper and lower shells of the regenerative heat exchanger, respectively. The weld overlays consist of IGSCC-resistant austenitic stainless steel material and, thus, are not susceptible to continued IGSCC crack propagation. However, the first 1/16" thick layer of weld metal deposited is not assumed to be IGSCC-resistant due to weld dilution; thus, it is assumed to be cracked. A fatigue crack growth analysis was performed in accorcance with ASME Section XI, Appendix C, with the crack propagating into the overlay from the hypothetical 1/16" deep crack. The fatigue crack growth analyses for the weld overlays are expected to remain valid for the period of extended operation.

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ANALYSIS 4-NMP2 There was a new fatigue analysis performed for the NMP2 Downcomer and Safety/Relief Discharge Line that generated a new TLAA. The new TLAA comprises new ALRA Section 4.6.6. The resultant changes to ALRA Table 4.1-1 on p. 4.1-3 are shown below.

Table 4.1-1 Time-Limited Apinp Analvqpq Annlicable to NMPN.R TLAt Description Category Section

1.

Reactor Vessel Neutron Embrittlement Analysis

__.2 Upper-shelf Energy

§54.21(c)(l)(ii) 4.2.1 Pressure-Temperature (P-T) Limits

§54.21(c)(l)(ii) 4.2.2 Elimination of Circumferential Weld Inspection (NMP 1 only)

§54.21(c)(1)(ii) 4.2.3 Axial Weld Failure Probability

§54.21(c)(1)(ii) 4.2.4

2.

Metal Fatigue Analysis Z..3 Reactor Vessel Fatigue Analysis

§54.21(c)(1)(iii) 4.3.1 ASME Section III Class I Piping and Components Fatigue

§54.21(c)(1)(iii) 4.3.2 Analysis (NMP2 only)

Feedwater (FWS) Nozzle and Control Rod Drive Return Line

§54.21(c)(1)(iii) 4.3.3 (CRDRL) Nozzle Fatigue and Cracking Analyses Non-ASME Section III Class 1 Piping and Components Fatigue

§54.21(c)(l)(iii) 4.3.4 Analysis Reactor Vessel Internals Fatigue Analysis

§54.21(c)(1)(iii) 4.3.5 Environmentally Assisted Fatigue

§54.21(c)(l)(iii) 4.3.6 Fatigue of the Emergency Condenser (NMP1 only)

§54.21(c)(1)(iii) 4.3.7

3.

Environmental Qualification (EQ) 4.4 Electrical Equipment EQ

§54.21(c)(1)(iii) 4.4.1 Mechanical Equipment EQ (NMP2 only)

§54.21(c)(1)(iii) 4.4.2

4.

Concrete Containment Tendon Prestress Analysis Not Applicable 4.5

5.

Containment Liner Plate, Metal Containments, and 4.6 Penetrations Fatigue Analysis Torus Shell and Vent System Fatigue Analysis (NMP1 only)

§54.21(c)(1)(i) 4.15.1 and

§54.21(c)(1)(ii)

Torus Attached Piping Analysis (NMPI only)

§54.21(c)(1)(iii) 4.6.2 Torus Wall Thickness (NMP I only)

§54.21(c)(1)(iii) 4.6.3 Containment Liner Analysis (NMP2 only)

§54.21(c)(1)(ii) 4.6.4 Fatigue of Primary Containment Penetrations

§54.21(c)(1)(i),

4.6.5

§54.21(c)(1)(ii) and

§54.21(c)(1)(iii)

Downcomer and Safety/Relief Valve Discharge Line Fatigue

§54.21(c)(1)(ii) 4.6.6 Evaluation (NMP2 Only) and

§54.21(c)(1)(iii)

6.

Other Plant-specific TLAAs

_7 RPV Biological Shield (NMP2 only)

§54.21(c)(1)(ii) 43.1 Main Steam Isolation Valve Corrosion Allowance (NMP2 only)

§54.21(c)(l)(iii) 4..2 Stress Relaxation of Core Plate Hold-Down Bolts (NMP2 only)

§54.21(c)(1)(iii) 42.3 Reactor Vessel and Reactor Vessel Closure Head Weld Flaw

§54.21(c)(1)(i) 4.7.4 Evaluations (NMPI only) and

§54.21(c)(1)(iii)

Reactor Water Cleanup System Weld Overlay Fatigue Flaw

§54.21(c)(l)(i) 4.7.5 Growth Evaluation (NMP 1 Only) 6of9

New ALRA Section 4.6.6 that would start on p. 4.6-14 is as shown below.

4.6.6 DOWNCOMER AND SAFETY/RELIEF VALVE DISCHARGE LINE FATIGUE EVALUATION (NMP2 ONLY)

Summary Description - Downcomer Fatigue Evaluation The NMP2 downcomers consist of 121 pipes open to the drywell and submerged 9.5 ft below the low water level (operating minimum) of the suppression pool, providing a flow path for uncondensed steam into the pool.

Although the downcomers are a structural component of the primary containment structure, a fatigue analysis using ASME Section III Class 1 rules was performed for the downcomers. Since the analysis includes an assumption of the numbers of cycles that will occur over the 40-year life of the plant, the analysis is a time-limited aging analysis.

Disposition:

§54.21(c)(1)(ii) -The analysis has been projected to the end ofthe period of extended operation.

Analysis The load combinations considered in the fatigue analysis of the downcomers include normal operating load conditions and accident conditions, including a small break accident (SBA), intermediate break accident (IBA), or design basis accident (DBA). An SBA, IBA, or DBA was assumed to occur one time during the life of the plant, combined with the fatigue usage resulting from upset conditions plus one Residual Heat Removal (RHR) System blowdown. Three different usage factors were determined: (1) upset +

RHR + SBA; (2) upset + IBA + DBA; and (3) upset + RHR + DBA. The highest usage factor was for the upset + RHR +DBA, with a 40-year usage factor of 0.182. Upset loads include safety/relief valve discharge (SRV), operating basis earthquake, and thermal loads. A total of 5,154 SRV discharges were included in the analysis (including first and subsequent actuations). The largest contributor to the usage factor was from the upset loads, amounting to 0.0966. Since the upset loads were assumed to occur more than once over the 40-year life of the plant, it is appropriate to project the usage resulting form these loads for an additional 20 years of operation by multiplying the usage by 1.5.

Therefore, the projected upset load usage for 60 years is 0.1449. Adding this to the contribution from DBA and RHR which is 0.0854, the total 60-year projected fatigue usage is 0.230 which remains below the ASME Section III allowable of 1.0. Therefore, the downcomer fatigue analysis has been projected in accordance with §54.21(c)(1)(ii).

Summary Description - SRV Penetration Fatigue Analysis There are eighteen (18) twelve-inch diameter SRV lines that penetrate the drywell floor via flued head type penetrations at NMP2. For these penetrations, a fatigue analysis using ASME Section III Class 1 rules was performed for the SRV piping penetrations through the drywell floor. Although the SRV piping system is Safety Class 3, these additional requirements are considered for additional assurance that steam bypass across the drywell floor will not occur. Additionally, since the highest loads occur at the penetration, thi:s serves to increase the confidence in the overall system.

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Disposition:

§54.21(c)(1)(iii) - The effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Analysis For the SRV penetration fatigue analysis, the analysis considered operating basis earthquake, SRV actuations, condensation oscillation, chugging, and thermal loads occurring during SRV discharge. A total of 5,154 SRV discharges (including first and subsequent actuations) were included in the analysis. The SRV penetration fatigue analysis was reevaluated as a result of power uprate. The revised 40-year CUF was 0.667. The contribution of the non-emergency load cases to the overall usage is 0.61.5. If the non-emergency usage is projected by multiplying by a 1.5 factor, the resulting non-emergency usage is 0.923. Adding the emergency usage of 0.053 to the non-emergency contribution yields a 60-year CUF of 0.976. Because this CUF approaches 1.0, NMPNS will revise the SRV penetration fatigue analysis to remove excessive conservatism, or will monitor fatigue usage of the SRV penetrations via the Fatigue Monitoring Program, with the aid of the FatiguePro software. The revision to the analysis or implementation of monitoring will be completed in conjunction with the analyses described in ALRA Section 4.3 prior to the start of the period of extended operation.

Since the SRV penetration analysis will be revised, or monitoring of the SRV penetrations will be initiated, prior to entry into the period of extended operation, the effects of aging will be adequately managed in accordance with §54.21(c)(1)(iii).

This T]AA results in the addition of new ALRA Section A2.2.4.3 starting on p. A2-29 as shown below.

A2.2.4.3 DOWNCOMER AND SAFETY/RELIEF VALVE DISCHARGE LINE FATIGUE EVALUATION The downcomers consist of 121 pipes open to the drywell and submerged 9.5 ft below the low water level (operating minimum) of the suppression pool, providing a flow path for uncondensed steam into the pool. The load combinations considered in the fatigue analysis of the downcomers include normal operating load conditions and accident conditions, including a small break accident (SBA), intermediate break accident (IBA),

or design basis accident (DBA). An SBA, IBA, or DBA was assumed to occur one time during the life of the plant, combined with the fatigue usage resulting from upset conditions plus one Residual Heat Removal (RHR) System blowdown. The largest contributor to the usage factor was from the upset loads. Since the upset loads were assumed to occur more than once over the 40-year life of the plant, it is appropriate to project the usage resulting form these loads for an additional 20 years of operation by multiplying the usage by 1.5. Adding this to the contribution from DBA and RHR the total projected fatigue usage is below the ASME Section III allowable of 1.0.

Therefore, the downcomer fatigue CUF has been projected for the period of extended operation.

There are eighteen (18) twelve-inch diameter SRV lines that penetrate the drywell floor via flued head type penetrations. For these penetrations, a fatigue analysis using ASME 8 of9

Section III Class 1 rules was performed for the SRV piping penetrations through the drywell floor. Although the SRV piping system is Safety Class 3, these additional requirements are considered for additional assurance that steam bypass across the drywell floor will not occur. Additionally, since the highest loads occur at the penetration, this serves to increase the confidence in the overall system. For the SRV penetration fatigue analysis, the analysis considered operating basis earthquake, SRV actuations, condensation oscillation, chugging, and thermal loads occurring during SRV discharge. The SRV penetration fatigue analysis was reevaluated as a result of power uprate utilizing significant conservatism. The projected CUF for the balance of plant life including the period of extended operation was close to the ASME Section III allowable of 1.0; therefore, prior to entry into the period of extended operation, the SRV Penetration fatigue analysis will be revised to remove the excessive conservatism or the fatigue usage of the SRV penetrations will be monitored by the Fatigue Monitoring Program.

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