NL-08-092, Official Exhibit - ENT000193-00-BD01 - Entergy Nuclear Operation Inc., Indian Point, Units 2 & 3, Amendment 5 to License Renewal Application (LRA)

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Official Exhibit - ENT000193-00-BD01 - Entergy Nuclear Operation Inc., Indian Point, Units 2 & 3, Amendment 5 to License Renewal Application (LRA)
ML12338A516
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 06/11/2008
From: Dacimo F
Entergy Nuclear Northeast
To:
Atomic Safety and Licensing Board Panel, Document Control Desk, Office of Nuclear Reactor Regulation
SECY RAS
References
RAS 22112, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, NL-08-092
Download: ML12338A516 (33)
v • d • e


Text

United States Nuclear Regulatory Commission Official Hearing Exhibit Entergy Nuclear Operations, Inc.

In the Matter of:

(Indian Point Nuclear Generating Units 2 and 3) c.\..t.P.RREG(J~)o ASLBP #: 07-858-03-LR-BD01

.!~~\

Docket #: 05000247 l 05000286

  • 0 Exhibit #: ENT000193-00-BD01 Identified: 10/15/2012

~ ~

~ Admitted: 10/15/2012 Withdrawn: ENT000193

~

~-s-:

i Rejected: Stricken:

~ ,-d' Submitted: March 29, 2012

.***. Other:

Entergy Nuclear Northeast Indian Point Energy Center

~Entergx 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal June 11, 2008 Re: Indian Point Units 2 & 3 Docket Nos. 50-247 & 50-286 NL-08-092 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations Inc.

Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 Amendment 5 to License Renewal Application (LRAl

REFERENCES:

1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)
2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings (NL-07-040)
3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References (NL-07-041)
4. Entergy Letter dated October 11, 2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07-124)
5. Entergy Letter November 14,2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA)

Environmental Report References" (NL-07-133)

Dear Sir or Madam:

In the referenced letters, Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Energy Center operating license.

This letter contains Amendment 5 of the License Renewal Application (LRA) which consists of three attachments. Attachment 1 consists of the annual update amendment to the LRA.

. NL-08-092 Docket Nos. 50-247 & 50-286 Page 2 of 2 consists of an amendment for (a)( 2) clarification. Attachment 3 cqnsists of an'.

amendment for reactor vessel clarification.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on

~/JlIQ:t

. , .

Attachments:

1. Annual Update Amendment
2. (A)(2) Clarification Amendment
3. Reactor Vessel Clarification Amendment cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. Kenneth Chang, NRC Branch Chief, Engineering Review Branch I Mr. Bo M. Pham, NRC Environmental Project Manager Mr. John Boska, NRR Senior Project Manager Mr. Paul Eddy, New York State Department of Public Service NRC Resident Inspector's Office Mr. Paul D. Tonko, President, New York State Energy, Research, & Development Authority

ATTACHMENT 1 TO NL-08-092 Annual Update Amendment ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-08-092 Attachment 1 Docket Nos. 50-247 & 50-286 Page 1 of 5 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION ANNUAL UPDATE AMENDMENT (Changes are shown as strikethroughs for eleletioRs and underlines for additions)

LRA Section 2.3.2.2, Containment Spray System, System Description, Unit 2, second and third paragraphs, are revisec:l as follows.

Long-term post-accident retention of iodine is assured by four trisoelil:lm ~Aes~Aate sodium tetra borate baskets located in the containment at an elevation (46') that will be flooded under accident conditions, allowing the trisoelil:lm ~Aos~Aate sodium tetra borate to dissolve into the fluid for pH control. The four trisoelil:lm ~Aos~Aate sodium tetraborate baskets are included in the containment structural evaluation (summarized in Section 2.4.1) but are not discussed further as they have no license renewal intended function and are therefore not subject to aging management review.

The containment spray system has the following intended functions for 10 CFR 54.4(a)(1).

  • Provide means for rapid reduction of containment pressure and temperature by providing borated water from the RWST following a design basis LOCA or a steam line break accident inside containment.
  • Distribute flow from the containment recirculation pumps or RHR pumps to the containment atmosphere during the recirculation phase of an accident.
  • Provide for chemical additives (trisoelil:lm ~Aos~Aate sodium tetraborate) to increase the pH of post-accident fluids in the recirculation and containment sumps.
  • Provide containment isolation capability for lines penetrating containment.

LRA Section 2.3.3.10, Control Room Heating, Ventilation and Cooling, System Description, Unit 2, second, third, and fifth paragraphs are revised as follows.

Unit 1 and Unit 2 share a central control room. The Unit 1 control room ventilation equipment for the central control room has been modified for recirculation mode only. +he URit 1 seRtrel room veRtilatieR eEll:li~meRt is Ret sreeliteel fer sooliRI or filtratioR.

The central control room (CCR) HVAC system has the following intended function for 10 CFR 54.4(a)(1).

  • Maintain a suitable environment in the main control room for operating personnel and safety-related equipment.
  • Proviele filtratioR of iRsomiRI air aREI maiRtaiR a ~ositive ~ressl:lre iR tAe sORtrol room.
  • Provide a means to isolate the CCR to prevent the infiltration of toxic gas, and remove airborne radioactivity from the outside air intake during high radiation or safety injection conditions to ensure protection of the operators and CCR habitability.
  • Provide a slight positive pressure in the CCR during accident conditions.

NL-08-092 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 5

  • Provide ventilation. supervisorv control panel emergency by-pass fan. to remove heat from the supervisory panel whenever the fan in the air conditioning unit is not available.
  • Provide cooling as required to maintain acceptable temperatures inside the CCR during accident conditions.

The CCR HVAC system has the following intended function for 10 CFR 54.4(a)(3).

  • Provide ventilation system isolation (via damper§. 9J3eraeility) as required during an Appendix R event (10 CFR 130.48).
  • Maintain the CCR in a safe. habitable environment during an Appendix R Event and Station Blackout.

LRA Table 2.4-1, Containment Building, Components Subject to Aging Management Review, is revised as follows.

Sl::IFRP liAer aAE! peAetratieAS . Pressl::Ire gel::lAE!ary

. SRelter er pretestieA Sl::Ippert fer GriterieA (aH~) eEll::lipFReAt Sump screens, strainer, barriers. annulus Shelter or protection trash racks grating, and flow barriers Support for Criterion (a)(1) equipment

NL-08-092 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 5 LRA Table 3.5.2-1, Containment Building Structural Components and Commodities (IP2 and IP3), is revised as follows.

S~m~ liReF aRe eN, PS, GaFl:}eR steel e*~esee te fl~ie bess ef mateFial Gil IWe -G

~eAetFatieRs SSR eRviFeAmeRt GeAtaiRmeRt beak Rate 1~12 S~m~ SGFeeRS, eN,SSR GaFl:}eR steel .A.iF . iReeeF bess ef mateFial StF~Gt~FeS 3.5.1 25 -G stFaiReF aRe fie\'! ~RGeRtFeilee MeAiteFiRJ l:}aFFieFS Sump screens, IEN,SSR I Stainless I Air - indoor I None I None I III.B 1.3-7 3.5.1-59 C strainer, barriers, steel uncontrolled (TP-5) annulus trash racks, grating, and flow barriers Sumos I PB. SSR I Concrete I Exposed to fluid I None I Structures l....

environment Monitorina 501

NL-08-092 Attachment 1 Docket Nos. 50-247 & 50'-286 Page 4 of 5 LRA Section A.2.1.17, Inservice Inspection -Inservice Inspection (lSI) Program, third paragraph, is revised as follows.

On Jlliy March 1, 20074-994, tRe J:)lant IP2 entered the tRifG-fourth lSI interval. The ASME code edition and addenda used for the tRifG-fourth interval is the 4-9892001 Edition with AG2003 addenda.

LRA Section B.1.8, Containment Inservice Inspection, Program Description, is revised as follows.

The Containment Inservice Inspection (CII) Program is an existing program encompassing ASME Section XI Subsection IWE and IWL requirements as modified by 10 CFR 50.55a.

The IP2 program uses the ASME Boiler and Pressure Vessel Code,Section XI, ~ 2001 Edition, ~ 2003Addenda. The IP3 program uses the ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition, no Addenda. Every 10 years, each unit's program is updated to the latest ASME Section XI code edition and addenda approved by the Nuclear Regulatory Commission in 10 CFR 50.55a. '

LRA Section B.1.18, Inservice Inspection, Program Description, seventh paragraph, is revised as follows.

On Jlliy March 1, 20074-994, IP2 entered the tRifG-fourth lSI interval and on July 21, 2000, IP3 entered the third lSI interval. The ASME code edition and addenda used for the IP2 fourth interval is the 2001 Edition with 2003 addenda. The ASME code edition and addenda used for the IP3 third interval fer BetR l;Jnits is the 1989 Edition with no addenda.

LRA Section B.1.18, Inservice Inspection, Element 4, eighth paragraph, is revised as follows.

For GetA-IP2 anElIP3, Article IWF of ASME Section XI, 1989 eElitien 2001 Edition and 2003 Addenda, does not contain any specific exemption criteria for component supports. For IP3.

Gcomponents exempt from examination are in accordance with the criteria contained in Code Case N-491-2, Alternate Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1, IWF-1230.

Additional LRA Clarification LRASection 4.3.3, Effects of Reactor Water Environment on Fatigue Life, is revised to delete the tenth paragraph as follows.

Fer tRese leGatiens witR CUFs less tRan 1.0, tRe TLAA Ras Been J:)raj,eGteEl tRrel;J!1JR tRe J:)erieEl ef ex:tenEleEl eJ:)eratien J:)er 10CFR.§4 .21 (G)(1 )(ii).

LRA Section A.2.1.20, Nickel Alloy Inspection Program, last paragraph, is revised as follows.

(Refer to RAI 3.0.3.3.5-2 response in letter NL-08-051 dated March 12, 2008)

NL-08-092 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 5 The site will sontinl;le to iffiJ3leffient sommi!ments assosiatea with (1) ~JRG Oraers, Bl;llIetins ana Generis betters assosiatea '.vith niskel alloys ana (2) staff asseJ3tea inal;lstl)' Jl;liaelines.

The site commits to comply with future applicable NRC Orders. In addition. IPEC commits to implement applicable (1) Bulletins and Generic Letters associated with nickel alloys and (2) staff-accepted industry guidelines associated with nickel alloys.

LRA Section A.3.1.20, Nickel Alloy Inspection Program, last paragraph, is revised as follows.

(Refer to RAI 3.0.3.3.5-2 response in letter NL-08-051 dated March 12, 2008)

The site will sontinl;le to imJ3lement somffiitments assosiatea with (1) ~JRG Oraers, Bl;llIetins ana Generis betteFS assosiatea with nisl<el alloys ana (2) staff asseJ3tea inal;lstl)' Jl;liaelines.

The site commits to comply with future applicable NRC Orders. In addition. IPEC commits to implement applicable (1) Bulletins and Generic Letters associated with nickel alloys and (2) staff accepted industry guidelines associated with nickel alloys.

LRA Section B.1.21, Nickel Alloy Inspection, Program Description, last paragraph, is revised as follows. (Refer to RAI 3.0.3.3.5-2 response in letter NL-08-051 dated March 12,2008)

IPEG will sontinl;le to imJ3lement sOffiffiitments assosiatea with (1) ~JRG Oraers, Bl;llIetins ana Generis betteFS assesiatea with niskel alleys ana (2) staff asseJ3tea inablstl)' Jbliaelines.

IPEC commits to comply with future applicable NRC Orders. In addition. IPEC commits to implement applicable (1) Bulletins and Generic Letters associated with nickel alloys and (2) staff accepted industry guidelines associated with nickel alloys.

LRA Section B.1.12, Fatigue Monitoring, Exceptions to NUREG-1801, is revised as follows.

The Fatigue Monitoring Program is consistent with the program described in NUREG-1801,Section X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary, with the follmo'JinJ e>EseJ3tion.

AttFibutes AtteGted &>EGeptiens

4. Detestion of !\§inJ Effests NlJREG 18(;)1 sJ3esifies J3erioais l;IJ3aates of fatiJl;le I;IsaJe salsl;llations. The IPEG J3roJram l;IJ3aates fatiJl;le I;IsaJe salsl;llations when the nl;lffiser of astl;lal sysles aJ3J3roash the analYi!ea nl;lffiser of sysles 4-E>EseJ3tlon Notes 4-- 61 ,aeiates sf fati!iill:le I:Isa!iile salsl:llatisRs aFe RstResessary I:IRless tRe Rl:lmeeF sf assl:lml:llateei fati!iill:le sysles a,a,aFsasRes tRe Rl:lmeeF sf aRalyzeei eieSi!iilR sysles: TRe IPeG ,aFs!iilFam ,aFsvieies feF ,aeFiseiis assessmeRt sf tRe Rl:lmeeF sfassl:lml:llateei sysles. If aRY tFaRsieRt a,a,aFsasRes its Rl:lmeeF sf aRalyzeei Gysles, sSFFestive astisR is takeR 'NRiSR may iRsll:leie l:I,aeiatesf tRe fati!iill:le I:Isa!iile salsl:llatisR.

ATTACHMENT 2 TO NL-08-092 (A)(2) Clarification Amendment ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

  • NL-OS-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 1 of 15 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION AMENDMENT FOR (a)(2) CLARIFICATION During the NRC regional inspection and audits, specific questions by the staff prompted further review of nonsafety-related systems, structures and components that could affect components that perform a safety function. This review resulted in the following changes to the LRA section 3.3.2-19 tables. These section 3.3.2-19 table changes only impacted the periodic surveillance and preventive maintenance program by adding one new inspection activity for the condensate pump suction system as described below. (underline - added, strikethrough - deleted)

LRA Table 3.3.2c19-4-IP2, Condensate System, is revised as follows.

Table 3.3.2-19-4-IP2: Condensate System Aging Effect Aging NUREG-Component Intended Table 1.

Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Flow element Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ~

boundar~ steel .tin!i fatigue fatigue (S-OS)

Flow element Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 .A..,.

boundar~ steel (int) material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Heat exchanger Pressure Carbon Treated water Loss of Water VII I. E-34 3.4.1-4 .A..,.

(shell) . boundar~ steel (int) material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

EiQlo.g Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ~

boundar~ steel .tin!i fatigue fatigue (S-08)

EiQlo.g Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 .A..,.

boundar~ steel > 140°F (int) Chemistr~ (SP-17) 314 Control-Primar~ and Secondar~

Piping Pressure Stainless Treated water Cracking - TLAA- metal VII.E1-16 3.3.1-2 ~

boundar~ steel > 140°F (int) fatigue fatigue (A-57) 302

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 2 of 15 Table 3.3.2-19-4-IP2: Condensate System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item ElQl.Q.g . Pressure Stainless Treated water Loss of Water VIILB1-4 3.4.1-16 .A..

boundar~ steel > 140°F (int) material Chemistr~ (SP-16) 314 Control-Primar~ and Secondar~

Strainer housing Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundar~ steel (inn fatigue fatigue (S-08)

Strainer housing Pressure Carbon Treated water Loss of Water VIILE-34 3.4.1-4 .A..

boundar~ steel f.i.o1l material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Strainer housing Pressure Stainless Treated water Cracking Water VIILE-30 3.4.1-14 .A..

boundar~ steel > 140°F (int) Chemistr~ (SP-17) 314 Control-Primar~ and Secondar~

Strainer housing Pressure Stainless Treated water Cracking - TLAA- metal VILE1-16 3.3.1-2 ...Q.....

boundar~ steel > 140°F (int) fatigue fatigue (A-57) 302 Strainer housing Pressure Stainless Treated water Loss of Water VIILB1-4 3.4.1-16 .A..

boundar~ steel > 140°F (int) material Chemistr~ (SP-16) 314 Control-Primar~ and Secondar~

Tank Pressure Carbon Treated water Loss of Water VIILE-34 3.4.1-4 .A..

boundar~ steel f.i.o1l material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Tank Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 .A..

boundar~ steel > 140°F (int) Chemistr~ (SP-17) 314 Control-Primar~ and Secondar~

,-, .... !

NL-08~092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 3 of 15 Table 3.3.2-19-4-IP2: Condensate System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Tank Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 b boundar~ steel > 140°F (int} material Chemistr~ (SP-16} 314 Control-Primar~ and Secondar~

Thermowell Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ~

boundar~ steel (int) fatigue fatigue (S-08)

Thermowell Pressure Carbon Treated water Loss of Water VII I. E-34 3.4.1-4 b boundar~ steel (int) material Chemistr~ (S-10} 314 Control-Primar~ and Secondar~

Thermowell Pressure Stainless Treated water Cracking Water VIII.E-30, 3.4.1-14 b boundar~ steel > 140°F (int} Chemistr~ (SP-17} 314 Control-Primar~ and Secondar~

Thermowell Pressure Stainless Treated water Cracking ...:. TLAA- metal VII.E1-16 3.3.1-2 ..Q.."

boundar~ steel > 140°F (int) fatigue fatigue (A-57) 302 Thermowell Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 b boundar~ steel > 140°F (int} material Chemistr~ (SP-16} 314 Control-Primar~ and 8econdar~

Valve bod~ -Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ~

boundar~ steel (int) fatigue fatigue (8-08)

Valve bod~ Pressure Carbon Treated water Loss of Water VII I. E-34 3.4.1-4 b boundar~ steel llo!l material Chemistr~ (8-10} 314 Control-Primar~ and Secondar~

Valve bod~ Pressure Gra~ cast Treated water Cracking - TLAA - metal VIII.B1-10 3.4.1-1 ~

boundar~ iron (int) fatigue fatigue (8-08)

'.1 . .

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 4 of 15 Table 3.3.2-19-4-IP2: Condensate System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Valve body Pressure Gray cast Treated water Loss of Water VIILE-34 3.4.1-4 .A..

boundary iron (int) material Chemistry (S-10) 314 Control-Primary and Secondary Valve body Pressure Stainless Treated water Cracking Water VIILE-30 3.4.1-14 .A....

boundary steel > 140°F (int} Chemistry (SP-17} 314 Control-Primary and Secondary Valve body Pressure Stainless Treated water Cracking - TLAA- metal VILE1-16 3.3.1-2 b boundary steel > 140°F (int) fatigue fatigue (A-57) 302 Valve body Pressure Stainless Treated water Loss of Water VIILB1-4 3.4.1-16 .A..

boundary steel > 140°F (int} material Chemistry (SP-16) 314 Control-Primary and Secondary LRA Table 3.3.2-19-5-IP2, Chemical and Volume Control System, is revised as follows.

Table 3.3.2-19-5-IP2: Chemical and Volume Control System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Filter housing Pressure Stainless Air- indoor None None VILJ-16 3.3.1- ..A boundary steel (ext) (AP-18) 99 Filter housing Pressure Stainless Treated Loss of Water VILE1-17 3.3.1- ..A boundary steel borated water material Chemistry (AP-79) ID.

(int) Control-Primary and Secondary Heat exchanger Pressure Stainless Air - indoor None None VILJ-16 3.3.1- ..A housing boundary steel (ext) (AP-18) 99

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 5 of 15 Table 3.3.2-19-5-IP2: Chemical and Volume Control System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Heat exchanger Pressure Stainless Treated Loss of Water VII.E1-17 3.3.1- .£ housing boundar~ steel borated water material Chemistr~ (AP-79) 91 (int) Control-Primar~ and Secondar~

LRA Table 3.3.2-19-17-IP2, Heating, Ventilation, and Air Conditioning System, is revised as follows.

Table 3.3.2-19-17-IP2: Heating, Ventilation, and Air Conditioning System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Strainer housing Pressure Carbon Air - indoor Loss of External VII.F1-2 3.3.1- .A boundar~ steel (ext) material Surfaces (A-1m 56 Monitoring Strainer housing Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ~

boundar~ steel (int) fatigue fatigue (S-08) 309 Strainer housing Pressure Carbon Treated water Loss of Water VIII.B1-11 3.4.1-4 ~

boundar~ steel !In!l material Chemistr~ (S-1m 314 Control-Primar~ and Secondar~

LRA Table 3.3.2-19-24-IP2, Miscellaneous System, is revised as follows.

Table 3.3.2-19-24-IP2: Miscellaneous System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item eiQing Pressure Carbon Air - indoor Loss of External V.B-1 3.2.1-32 .£ boundar~ steel (int) material Surfaces (E-2S)

Monitoring

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 6 of 15 Table 3.3.2-19-24-IP2: Miscellaneous System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Valve bod~ Pressure Carbon Air - indoor Loss of External V.B-1 3.2.1-32 ..£ boundar~ steel ill!!l material Surfaces (E-25)

Monitoring LRA Table 3.3.2-19-31-IP2, Radiation Monitoring System, is revised as follows.

Table 3.3.2-19-31-IP2: Radiation Monitoring System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs '2 Item Puml2 casing Pressure Stainless Air - indoor None None VII.J-15 3.3.1- .A boundar~ steel (ext) (AP-17) 94 Puml2 casing Pressure Stainless Treated water Cracking Water VIII.B1-5 3.4.1-14 ~

boundar~ steel > 140°F (int) Chemistr~ (SP-17) 314 Control-Primar~ and Secondar~

Puml2 casing Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 ~

boundar~ steel > 140°F (int) material Chemistr~ (SP-16) 314 Control-Primar~ and Secondar~

Puml2 casing Pressure Stainless Treated water Loss of Water VII.C2-10 3.3.1- ..§ boundar~ steel (int) material Chemistr~ (A-52) 50 Control-Closed Cooling Water Puml2 casing Pressure Carbon Air - indoor Loss of External VI 1.1-8 3.3.1- .A boundar~ steel (ext) material Surfaces (A-77) 58 Monitoring Puml2 casing Pressure Carbon Raw water Loss of Periodic VII.C1-19 3.3.1- ..£ boundar~ steel (int) material Surveillance (A-38) 76 and Preventive Maintenance

\...:'_t NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 7 of 15 Table 3.3.2-19-31-IP2: Radiation Monitoring System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Tank Pressure Stainless Air- indoor None None VII.J-15 3.3.1- ..8 boundar~ steel (ext) (AP-17) 94 Tank Pressure Stainless Treated water Loss of Water VII I. E-40 3.4.1-6 £.,.

boundar~ steel (int) material Chemistr~ (S-13) 314 Control-Primar~ and Secondar~

LRA Table 3.3.2-19-33-IP2, Station Air System, is revised as follows.

Table 3.3.2-19-33-IP2: Station Air System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item CemJ7)resser Pressl:lre Careen Air ineeer bess ef e*ternal ~ ~ -A hel:lsin§ eel:lneary steel ~ material SI:lRaSeS ~ §.7 Meniterin§ CemJ7)resser PreSSl:lr9 Careen Treatee water bess ef Watef VII.C214 ~ ..g hel:lsin§ eel:lneary steel fiAtt material Chem istry ~ 47 Centrel GleseG Ceelin§ Water

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 8 of 15 LRA Table 3.3.2-19-9-IP3, Condensate Pump Suction System, is revised as follows.

Table 3.3.2-19-9-IP3: Condensate Pump Suction System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item EXQansion joint Pressure Elastomer Air - indoor Change in Periodic VILF1-7 3.3.1- g boundar~ (ext) material Surveillance (A-17) 11 QroQerties and Preventive Maintenance EXQansion joint Pressure Elastomer Air - indoor Cracking Periodic VILF1-7 3.3.1- ~

boundar~ (ext) Surveillance (A-17l 11 and Preventive Maintenance EXQansion joint Pressure Elastomer Treated water Change in Periodic VILA4-1 3.3.1- .£ boundar~ (int) material Surveillance (A-16) 11 QroQerties and Preventive Maintenance

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 9 of 15 Table 3.3.2-19-9-IP3: Condensate Pump Suction System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Ex~ansion joint Pressure Elastomer Treated water Cracking Periodic VILA4-1 3.3.1- J; boundary Dn.U Surveillance J..g (A-16) and Preventive Maintenance LRA Table 3.3.2-19-16-IP3, Emergency Diesel Generator System, is revised as follows.

Table 3.3.2-19-16-IP3: Emergency Diesel Generator System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item E!Qin.g Pressure Carbon Treated water Loss of Water VILH2-23 3.3.1- l!

boundary steel Dn.U material Chemistry (A-25) 47 Control-Closed Cooling Water LRA Table 3.3.2-19-27-1 P3, Heater Drain / Moisture Separator Drains / Vents System, is revised as follows.

Table 3.3.2-19-27-IP3: Heater Drain I Moisture Separator Drains I Vents System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Ex~ansion joint Pressure Stainless Treated water Cracking Water VIILE-30 3.4.1-14 b boundary steel > 140°F (int) Chemistry (SP-17) 314 Control-Primary and Secondary EXl2ansion joint Pressure Stainless Treated water Loss of Water VIILE-29 3.4.1-16 b boundary steel > 140°F (int) material Chemistry (SP-16) 314 Control-Primary and Secondary

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 10 of 15 Table 3.3.2-19-27-IP3: Heater Drain I Moisture Separator Drains I Vents System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Heat exchanger Pressure Carbon Treated water Loss of Water VIILE-34 3.4.1-4 ~

(shell) boundary steel Qn!l material Chemistry (S-10) 314 Control-Primary and Secondary Orifice Pressure Stainless Treated water Cracking Water VIILE-30 3.4.1-14 ~

boundary steel > 140°F (int) Chemistry (SP-17) 314 Control ~

Primary and Secondary Orifice Pressure Stainless Treated water Cracking - TLAA- metal VILE1-16 3.3.1-2 ~

boundary steel > 140°F (int) fatigue fatigue (A-57) 302 Orifice Pressure Stainless Treated water Loss of Water VIILE-29 3.4.1-16 ~

boundary steel > 140°F (int) material Chemistry (SP-16) 314 Control-Primary and Secondary

.E.iQlo..g Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundary steel (int) fatigue fatigue . (S-08)

.E.iQlo..g Pressure Carbon Treated water Loss of Water VIILE-34 3.4.1-4 ~

boundary steel (int) material Chemistry (S-10) 314 Control-Primary and Secondary

.E.iQlo..g Pressure Stainless Treated water Cracking Water VIILE-30 3.4.1-14 ~

boundary steel > 140°F (int) Chemistry (SP-17) 314 Control-Primary and Secondary

.E.iQlo..g Pressure Stainless Treated water Cracking - TLAA- metal VILE1-16 3.3.1-2 ~

boundary steel > 140°F (int) fatigue fatigue (A-57) 302

.E.iQlo..g Pressure Stainless Treated water Loss of Water VIILE-29 3.4.1-16 ~

boundary steel > 140°F (int) material Chemistry (SP-16) 314 Control-Primary and

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 11 of 15 Table 3.3.2-19-27-IP3: Heater Drain I Moisture Separator Drains I Vents System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Secondar~

Pump casing Pressure Carbon SteaFfl (iAt) Loss of Water VilLA H~ a.4.~ ~ G-,

boundary steel Treated water material Chemistry ~ 3.4.1-4 aM Dn!l Control- &

Primary and VIILE-34 314 Secondary (S-10)

Sight glass Pressure Carbon Treated water Cracking - Periodic VIILB1-10 3.4.1-1 J; boundar~ steel Dn!l fatigue surveillance (S-08) and I2reventive maintenance Sight glass Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A....

boundar~ steel (int) material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Sight glass Pressure Glass Treated water None None VIILI-8 3.4.1-40 A boundar~ (int) (SP-35)

Strainer housing Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundar~ steel Dn!l fatigue fatigue (S-08)

Strainer housing Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A....

boundar~ steel Dn!l material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Tank Pressure Carbon Treated water Loss of Water VIILE-34 3.4.1-4 A....

boundar~ steel (int) material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Thermowell Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundar~ steel .(int) fatigue fatigue (S-08)

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 12 of 15 Table 3.3.2-19-27-IP3: Heater Drain I Moisture Separator Drains I Vents System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Thermowell Pressure .Carbon Treated water Loss of Water VII I. E-34 3.4.1-4 .A..,.

boundar~ steel .on.u material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

Tubing Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 .A..,.

boundar~ steel > 140°F (int} Chemistr~ (SP-17} 314 Control-Primar~ and Secondar~

Tubing Pressure Stainless Treated water Cracking - TLAA- metal VII.E1-16 3.3.1-2 ..Q...

boundar~ steel > 140°F (int} fatigue fatigue (A-57) 302 Tubing Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 .A..,.

boundar~ steel > 140°F (int} material Chemistr~ (SP-16} 314 Control-Primar~ and Secondar~

Valve bod~ Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ..Q boundar~ steel (inn fatigue fatigue (S-08l Valve bod~ Pressure Carbon Treated water Loss of Water VII I. E-34 3.4.1-4 .A..,.

boundar~ steel (inn material Chemistr~ (S-10} 314 Control-Primar~ and Secondar~

Valve bod~ Pressure Stainless Treated water Cracking Water VII I. E-30 3.4.1-14 .A..,.

boundar~ steel > 140°F (int} Chemistr~ (SP-17} 314 Control-Primar~ and Secondar~

Valve bod~ Pressure Stainless Treated water Cracking - TLAA- metal VII.E1-16 3.3.1-2 ..Q...

boundar~ steel > 140°F (int} fatigue fatigue (A-57} 302 Valve bod~ Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 .A..,.

boundar~ steel > 140°F (int} material Chemistr~ (SP-16} 314 Control-Primar~ and Secondar~

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 13 of 15 LRA Table 3.3.2-19-45-IP3, Reheat Steam System, is revised as follows.

Table 3.3.2-19~45-IP3: Reheat Steam System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item*

Heat exchanger Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 b (shell) boundary steel (int) material Chemistry (S-10) 314 Control-Primary and Secondary ElQlng Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundary steel (int) fatigue fatigue (S-08)

ElQlng Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 b boundary steel (int) material Chemistry (S-10) 314 Control-Primary and Secondary Steam tral2 Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundary steel (int) fatigue fatigue (S-08)

Steam tral2 Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 b boundary steel (int) material Chemistry (S-10) 314 Control-Primary and Secondary Strainer Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q Housing boundary steel (int) fatigue fatigue (S-08)

Strainer Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 b Housing boundary steel illl!l material Chemistry (S-10) 314 Control-Primary and Secondary Thermowell Pressure Carbon Treated water Cracking - TLAA- metal VIII.B1-10 3.4.1-1 ..Q boundary steel illl!l fatigue fatigue (S-08)

Thermowell Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 b boundary steel illl!l material Chemistry (S-10) 314 Control-Primary and

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 14 of 15 Table 3.3.2-19-45-IP3: Reheat Steam System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Secondar~

Tubing Pressure Stainless Treated water Cracking Water VIILC-2 3.4.1-14 .A....

boundar~ steel > 140°F (int) Chemistr~ (SP-17) 314 Control-Primar~ and Secondar~

Tubing Pressure Stainless Treated water Cracking - TLAA- metal VILE1-16 3.3.1-2 ..Q...

boundar~ steel > 140°F (int) fatigue fatigue (A-57) 302 Tubing Pressure Stainless Treated water Loss of Water VIILC-1 3.4.1-16 .A....

boundar~ steel > 140°F (int) material Chemistr~ (SP-16) 314 Control-Primar~ and Secondar~

Valve bod~ Pressure Carbon Treated water Cracking - TLAA- metal VIILB1-10 3.4.1-1 ..Q boundar~ steel Dn!l fatigue fatigue (S-08)

Valve bod~ Pressure Carbon Treated water Loss of Water VIILC-7 3.4.1-4 .A....

boundar~ steel Dn!l material Chemistr~ (S-10) 314 Control-Primar~ and Secondar~

NL-08-092 Attachment 2 Docket Nos. 50-247 & 50-286 Page 15 of 15 LRA Table 3.3.2-19-48-IP3, Station Air System, is revised as follows.

Table 3.3.2-19-48-IP3: Station Air System Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item CeFR(3resser Pressl:lre CarbeA Air iAgeer bess et exteFAal ~ ~ -A hel:lsiAI bSI:IA9ary steel ~ FRatsrial SI:IRaGSS ~ &7-MeAitsriAI CeFR(3rssser Prsssl:lrs CarbsA Trsats9 water bess st Watef VII.C2 1 ~ ~ -9 hel:lsiAI bSl:lA9ary steel fffitt FRaterial CheFRistry ~ 47-CSAtrel GteseG CssliAI \llJatsr Valve bod~ Pressure Carbon Air- indoor Loss of External VII.D-3 3.3.1- ..h.

boundar~ steel (ext) material Surfaces (A-80) 57 Monitoring Valve bod~ Pressure Carbon Condensation Loss of Periodic VII.D-2 3.3.1- ..£ boundar~ steel (int) material Surveillance (A-26) 53 and Preventive Maintenance LRA Section A.1.28, Periodic Surveillance and Preventive Maintenance Program, second paragraph, twenty-ninth bullet is revised as follows.

  • chlorination, circulating water, city water makeup, condensate pump suction. emergency diesel generator, floor drain, gaseous waste disposal, instrument air, liquid waste disposal, nuclear equipment drain, river water, station air piping, steam generator sampling, and secondary plant sampling piping components, and piping elements LRA Section B.1.29, Periodic Surveillance and Preventive Maintenance, Program Description, Nonsafety-related systems affecting IP3 safety-related systems, is revised to add the following task.

Use visual or other NDE techniques to inspect inside and outside surfaces of a representative sample of condensate pump suction system elastomer components to manage loss of material and cracking and change in material properties.

ATTACHMENT 3 TO NL-08-092 Reactor Vessel Clarification Amendment ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 1 of 8 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION AMENDMENT FOR REACTOR VESSEL CLARIFICATIONS Based on response to RAI 4.2.1-1 documented in letter NL-07-140 dated November 28,2007 and clarified in letter NL-08-014 dated January 17, 2008, the following LRA amendments are required .. (strikethroughs - deleted, underlines - added)

LRA Section 4.2, Reactor Vessel Neutron Embrittlement, second paragraph, is revised as follows.

The IPEC current licensing basis analyses evaluating reduction of fracture toughness of the reactor vessel for 40 years are TLAA. The reactor vessel neutron embrittlement TLAA for each unit is summarized below. Forty.;.eight effective full-power years (EFPY) are conservatively projected for the end of the period of extended operation (60 years) based on actual capacity factors from the start of commercial operation until 2005 and em IP2 and IP3 average capacity factor§. of 9~&% and >100% respectively from 2005 untill the end of the period of extended operation.

LRA Section A.2.2.1, Reactor Vessel Neutron Embrittlement, is revised as follows The current licensing basis analyses evaluating reduction of fracture toughness of the reactor vessel for 40 years are TLAA. The reactor vessel neutron embrittlement TLAA is summarized below. Forty-eight effective full-power years (EFPY) are conservatively projected for the end of the period of extended operation (60 years) based on actual capacity factors from the start of commercial operation until 2005 and an average capacity factor of 9~&% from 2005 to the end of the period of extended operation.

LRA Section A.3.2.1, Reactor Vessel Neutron Embrittlement, is revised as follows The current licensing basis analyses evaluating reduction of fracture toughness of the reactor vessel for 40 years are TLAA. The reactor vessel neutron embrittlement TLAA is summarized below. Forty-eight effective full-power years (EFPY) are conservatively projected for the end of the period of extended operation (60 years) based on actual capacity factors from the start of commercial operation until 2005 and an average capacity factor of 9&> 100% from 2005 to the end of the period of extended operation.

Consistent with LRA Section 4.2.4, Appendix A of the LRA will bejrevised to include discussion of PORV (LTOP) setpoint updates with the discussion of P-T limits. .

LRA Section A.2.2.1.2, Pressure-Temperature Limits, third paragraph is revised as follows.

The site IP2 will submit additional P-T curves as 10 CFR 50, Appendix G requires prior to the period of extended operation as part of the Reactor Vessel Surveillance Program.

LTOP (PORVl setpoints will be re-evaluated when pressure/temperature curves are submitted.

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 2 of 8 LRA Section A.3.2.1.2 , Pressure-Temperature Limits, third paragraph is revised as follows.

The site IP3 will submit additional P-T curves as 10 CFR 50, Appendix G requires prior to the period of extended operation as part of the Reactor Vessel Surveillance Program.

LTOP (PORVl setpoints will be re-evaluated when pressure/temperature curves are submitted.

The following provides additional reactor vessel information.

RAI 4.2.2-2 in letter NL-07-140 dated November 28,2007 requested the following IP2 information.

Table 3-1 in WCAP-13587, Revision 1 indicates the Westinghouse 4-loop plant J-R applied values are applicable for reactor vessels with a thickness of 8.5 inches and an inner radius of 86.5 inches and subject to Level A, B, C, and D conditions specified in Section 3.0 of the WCAP. Compare the wall thickness and inlier radius of the IP2 reactor vessel at its beltline to the values used in the Westinghouse 4-loop plant analysis.

Compare the Level A, B, C, and D conditions specified in Section 3.0 of the WCAP to the Level A, B, C and D conditions for IP2. Explain why the Westinghouse 4-loop plant analysis in WCAP-13587, Revision 1 is applicable to IP2.

A similar comparison is provided below for IP3.

Compare the wall thickness and inner radius of the IP3 reactor vessel at its beltline to the values used in the Westinghouse 4-loop plant analysis The IP3 reactor vessel is 86.5" inner radius with an 8.625" nominal wall thickness.

Therefore the 86.5" inner radius and the 8.5" minimal wall thickness dimensions used in WCAP-13587 bound the IP3 reactor vessel dimensions since these values result in conservative pressure stresses and have no significant impact on the through wall thermal stresses.

Compare the Level A. B, C, and D conditions specified in Section 3.0 of the WCAP to the Level A. B. C and D conditions for IP3 The Level A and B condition used in WCAP-13587 was a cool down rate of 100 degrees F per hour. This cool down rate is the same as the cool down rate for IP3.

The Level C conditions used in WCAP-13587 correspond to a small steam line break with the pressure and temperature time histories provided in Figure 3-1. These conditions were reviewed against the analyses provided in Chapter 14 of the IP3 FSAR. The IP3 FSAR conditions were bounded by the conditions analyzed in the WCAP.

. The Level D conditions used in the WCAP analysis were associated with the large steam line break and are provided in figure 3-2. These conditions were compared to the conditions provided in Chapter 14 of the IP3 FSAR. The conditions provided in the FSAR are bounded by the conditions used in the WCAP analysis.

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 3 of 8

. Based on the above, the evaluations provided in WCAP-13587 are applicable to the IP3 reactor vessel.

Since the equivalent margin analysis performed in WCAP-13587 was based on draft Reg Guide DG-1023 and ASME Code Case N-512, IPEC compared the methodology used in WCAP-13587 to Reg Guide 1.161, Evaluation of Reactor Pressure Vessels with Charpy Upper Shelf Energy Less Than 50 ft-Ibs and ASME Code Section XI, Appendix K.

IPEC concluded that WCAP-13587 did not deviate from the methods and formulas cited in the regulatory guide and ASME Code.

The IP2 chemistry factor used to determine the axial weld RTpts values shown in LRA Table 4.2-3 (updated per letter NL-08-014, dated January 17, 2008) was derived using surveillance data from IP2, IP3, and HB Robinson Unit 2. The surveillance welds operate at different temperatures and contain different amounts of copper and nickel. The methods used to determine an accurate chemistry factor from the surveillance data are shown in the attached excerpt from a site-approved calculation for updated IP2 pressure-temperature curves.

WCAP-16752, IP2 Heatup and Cooldown Limit curves for Normal Operation Excerpt from WCAP-16752, IP2 Heatup and C~oldown Limit Curves for Normal Operation Fracture toughness properties The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan [Reference 4].

The beltline material properties of the Indian Point Unit 2 reactor vessel are presented in Table 2-1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 2-1.

Additionally, surveillance capsule data is available for four capsules (Capsules V, Z, Y and T) already removed from the Indian Point Unit 2 reactor vessel. The fluence data for the surveillance capsules is presented in Table 2-2 and is used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2. It should be noted that in addition to Indian Point Unit 2, surveillance weld data from Indian Point Unit 3 and H.B. Robinson Unit 2 was used in the determination of CF. In addition, all the surveillance data has been determined to be credible, with exception of surveillance plate B-2002-2 ..

The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1. Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date, including those capsules from Indian Point Unit 3 and H.B. Robinson Unit 2.

The measured ~RT NOT values for the weld data were adjusted for the temperature difference between differing plants and for chemistry using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2. Table 2-3 contains the TCOld operating temperatures at Indian Point Units 2 and 3 and H.B. Robinson Unit 2. Table 2-4 details the calculation of the surveillance material chemistry factors. A summary of the resulting CF values for all of the vessel and surveillance materials is presented in Table 2-5.

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 4 of 8 TABLE 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NOT Values for the Indian Point Unit 2 Reactor Vessel Materials 5

(}::,~t}~~;/;~~~(~~r!~.ij~~~~j{~if~?*i~\ifg t':::~~~H~{~~f '::~~ i":r:~~;]~~~:I1~: ~~~~~!~~~~!~~t~~]

Closure Head Flange --- - -- 60°F Vessel Flange --- -- - 60°F Intermediate Shell Plate 8-2002-1(e) 0.19(0.21) . 0.65 (0.62) 34°F Intermediate Shell Plate 8-2002-2(e) 0.17 (0.15) 0.46 (0.44) 21°F Intermediate Shell Plate 8-2002-3(e) 0.25(0.20) 0.60 (0.59) 21°F Lower Shell Plate 8-2003-1 0.20 0.66 20°F Lower Shell Plate 8-2003-2 0.19 0.48 ~20°F Intermediate & Lower Shell Longitudinal Weld Seams (Heat # W5214)(b, d) 0.21 1.01 -56°F Intermediate to Lower Shell Girth Weld .0.19 1.01 -56°F (Heat # 348009) (c, d)

Indian Point Unit 2 Surveillance Weld 0.20 0.94 ---

(Heat # W5214)(b, d)

Indian Point Unit 3 Surveillance Weld 0.16 1.12 ---

(Heat # W5214)(b, d)

H.8. Robinson Unit 2 Surveillance Weld 0.32 0.66 ---

(Heat # W5214)(b, d)

Notes:

(a) The Initial RTNOT values are measured values, with exception to the weld materials.

(b) The weld material in the Indian Point Unit 2 surveillance program was made of the same wire and flux as the reactor vessel intermediate shell longitudinal weld seams 0JVjre Heat No. W5214 RAC03 + Ni200, Flux Type Linde 1092, Flux Lot No. 3600). The lower shell longitudinal weld seam also had the same heat and flux type but different flux lot. Indian Pt. Unit 3 and H.B. Robinson Unit 2 also contain surveillance material of this heat.

(c) The intermediate to lower shell circ. weld material was made of Wire Heat No. 34B009 RAC03 + Ni200, Flux Type Linde 1092, Flux Lot No. 3708.

(d) The weld best estimate copper and nickel weight percents were obtained from C~ Reports NPSD-1039, Rev. 2 and/or NPSD-1119, Rev. 1. The values from the CE Report NPSD-1119, Rev. 1 for the Indian Point 2 vessel axial and circ. welds match those in the NRC database RVID2. The values were rounded to two decimal points.

(e) Copper and Nickel Values were obtained from WCAP-12796, which in turn used Southwest Research Report 17-2108 (Capsule V Analysis) . This report calculated a best estimate Copper/Nickel weight percent excluding values that appeared to be outliers. If all data were considered, then the best estimate would match the RVID2 values shown in parentheses. The data above for the intermediate shell plates are conservative with exception to plate

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 5 of 8 data (See Tables 2-4 & 2-5) is used to provide a Position 2.1 chemistry factor of 114°F. Intermediate shell plate 8-2002-3 is more limiting than 8-2002-1 even if the highest CF were used for 8-2002-1. Values from WCAP-12796 will be used herein .

TABLE 2-2 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Indian Point Unit 2, Indian Point Unit 3 and H.B. Robinson Unit 2 18 2 T 2.53 X 10 n/cm , (E > 1.0 MeV) y 18 2 4.55 X 10 n/cm , (E > 1.0 MeV) z 1.02 X 10 19 n/cm 2

, (E > 1.0 MeV) 18 2 V 4.92 X 10 n/cm , (E > 1.0 MeV)

T 2.63 X10 18 n/cm 2 , (E > 1.0 MeV) (a) y 6.92 X 1018 n/cm 2 , (E > 1.0 MeV) (a) z 1.04 X 1019 n/cm 2 , (E > 1.0 MeV) (a) x 8.74 X 1018 n/cm 2 , (E > 1.0 MeV) (a) s 4.79 X 1018 n/cm 2 , (E > 1.0 MeV) (a)

V 5.30 X 1018 n/cm 2 , (E;' 1.0 MeV) (a)

T 3.87 X 1019 n/cm 2 , (E > 1.0 MeV) (a) x 4.49 X 1018 n/cm 2 , (E > 1.0 MeV) (a)

(a) Fluence values have been adjusted to be consistent with the methodology of Regulatory Guide 1.190

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 6 of 8 TABLE 2-3 Inlet (TCOld) Operating Temperatures 543°F (Cycle 1) 540°F (Capsule T) 54rF (Capsule S) 543°F (Cycle 2) 540°F (Capsule Y) 547°F (Capsule T) 522SF (Cycle 3) 540°F (Capsule Z) 547°F (Capsule X) 522.5°F (Cycle 4) 540°F (Capsule X) 522.8°F (Cycle 5) 522.8°F (Cycle 6) 522.8°F (Cycle 7) 522SF (Cycle 8)

(a) The temperatures listed above are consistent with historical treatment in previous Pressure-Temperature WCAP's, but are slightly conservative compared to measured operating history.

.. :,

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 7 of 8 TABLE 2-4 Calculation of Chemistry Factors using Indian Point Unit 2 Surveillance Capsule Data Intermediate Shell Plate B-2002-1 .

Intermediate Shell Plate' B-2002-2 Intermediate Shell .

Plate B-2002-3 Surveillance Weld Material(d)

Notes:

2 (a) f = fluence. See Table 2-3, W 19 10 n/cm , E> 1.0MeV).

(b) FF =fluence factor =f(0.28 . O. log f).

(c) ~RT NOT values are the measured 30 ft-Ib shift values taken from the following documents:

- Indian Point Unit 2 Plate and Weld ... WCAP-12796 (Refers back to the original Southwest Research Institute Report for each capsule.)

- Indian Point Unit 3 Weld ... WCAP-11815

- H.B.Robinson Unit 2 ... Letter Report CPL-96-203 (d) Per Table 2 Indian Point Unit 3 operates with an inlet temperature of approximately 540°F, H.B.

Robinson Unit 2 operates with an inlet temperature of approximately 547°F, and Indian Point Unit 2 operates with an inlet temperature of approximately 528°F. The measured ~RT NOT values from the Indian Point Unit 3 surveillance program were adjusted by adding 12°F to each measured ~RT NOT and the H.B. Robinson Unit 2 surveillance program values were adjusted by adding 19°F to each measured ~RT NOT value before applying the ratio procedure. The surveillance weld metal ~RT NOT values have been adjusted by a ratio factor of:

Ratio IP2 = 230.2 of- 214.3 = 1.07 for the Indian Point Unit 2 data.

Ratio IP3 = 230.2 of- 206.2 = 1.12 for the Indian Point Unit 3 data.

Ratio HBR2 = 230.2 of- 210.7 = 1.09 for the H.B. Robinson Unit 2 data.

Refer to Table 2-5 for the longitudinal weld seam CF of 230.2°F.

(The pre-adjusted values are in parenthesis.)

NL-08-092 Attachment 3 Docket Nos. 50-247 & 50-286 Page 8 of 8

. TABLE 2-5 Summary of the Indian Point Unit 2 Reactor Vessel Beltline Material Chemistry Factors Intermediate Shell Plate 8-2002-1 114 Intermediate Shell Plate 8-2002-2 118.2 Intermediate Shell Plate 8-2002-3 181.9 Lower Shell Plate 8-2003-1 Lower Shell Plate 8-2003-2 (a)

Intermediate & Lower Shell Longitudinal Weld Seams 251.8 (Heat # W5214)

Intermediate to Lower Shell Girth Weld Seam (Heat # 348009)

Indian Point Unit 2 Surveillance Weld (Heat # W5214)

Indian Point Unit 3 Surveillance Weld (Heat # W5214)

H.8. Robinson Unit 2 Surveillance 210.rF Weld (Heat # W5214)

(a) The 128.8°F CF listed here differs from previous analyses that used an excessively conservative CF of 142°F for this material. If the 142°F CF had been used in this analysis, this material would still not be.

limiting.

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