NL-08-092, Amendment 5 to License Renewal Application (LRA)
| ML081760265 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/11/2008 |
| From: | Dacimo F Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-08-092 | |
| Download: ML081760265 (33) | |
Text
SNEnhtergy Enteravy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 1051 1-0249 Tel (914) 788-2055 Fred Dacimno 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 (LRA)
- 1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)
REFERENCES:
- 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.
A12-?
NL-08-092 Docket Nos. 50-247 & 50-286 Page 2 of 2 consists of an amendment for (a)( 2) clarification. Attachment 3 consists 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 Sý ere ed R. Dacimo Vice President License Renewal 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 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 deletiens and underlines for additions)
LRA Section 2.3.2.2, Containment Spray System, System Description, Unit 2, second and third paragraphs, are revised as follows.
Long-term post-accident retention of iodine is assured by four tr!cOdiu'm pho*phate sodium tetraborate baskets located in the containment at an elevation (46') that will be flooded under accident conditions, allowing the tricodium phosphate sodium tetraborate to dissolve into the fluid for pH control. The four tricodium phosphate 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 (tFsidium phosphate 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. T-he Unit 1 cont*rol room vontilation equipment i* not credited for cooling Or filtrtion.
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.
Provide filtrntion of incoming air and maintain a positive preScuro in the control 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 Docket Nos. 50-247 & 50-286 Page 2 of 5 Provide ventilation, supervisory control panel emer-gency 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 dampers eper-abi*ty) as required during an Appendix R event-(41-GF.R-.60-8.
° 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.
Sump linerc and penetrations Pressure boun~darFy Shelter Or protection Support for Crfiteio*n (a)(*) equipment Sump screens, strainer, barriers, annulus Shelter or protection trash racks, gratin-g, and flow barriers Support for Criterion (a)(1) equipment
NL-08-092 Docket Nos. 50-247 & 50-286 Page 3 of 5 LRA Table 3.5.2-1, Containment Building Structural Components and Commodities (IP2 and !P3), is revised as follows.
SUP Ii~eI-F, EN, 1-PBT QGabeR-steI Expoede4to-fkli Loss of mateilCJIE-peinetratiaens
&SR eGfeie~tCntainment Leak Rate SUMP Green*,
EN, SSR ab steel Air-ieei Losf mAteria StFuGtuFe 3l.A5-1 35.26
-G trainer and flow uncentFlled M,,it,,ig J4T-)--
Sump screens, EN, SSR Stainless Air - indoor None None Ill.B1.3-7 3.5.1-59 C
strainer, barriers, steel uncontrolled (TP-5) annulus trash racks, grating, and flow barriers Sumps PB, SSR Concrete Exposed to fluid None Structures environment Monitoring 501
NL-08-092 Docket Nos. 50-247 & 50-286 Page 4 of 5 LRA Section A.2.1.17, Inservice Inspection - Inservice Inspection (ISI) Program, third paragraph, is revised as follows.
On Ju*y March 1, 20071-994, theplant IP2 entered the third-fourth ISI interval. The ASME code edition and addenda used for the third-fourth interval is the 1-9892001 Edition with Re2003 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 Xl Subsection IWE and IWL requirements as modified by 10 CFR 50.55a.
The IP2 program uses the ASME Boiler and Pressure Vessel Code, Section Xl, 1-992 2001 Edition, 1-992 2003Addenda. The IP3 program uses the ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition, no Addenda. Every 10 years, each unit's program is updated to the latest ASME Section Xl 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 July March 1, 20071-994, IP2 entered the thkd-fourth ISI interval and on July 21, 2000, IP3 entered the third ISI 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 both units is the 1989 Edition with no addenda.
LRA Section B.1.18, Inservice Inspection, Element 4, eighth paragraph, is revised as follows.
For beth-I P2 and 1P3, Article IWF of ASME Section Xl, 1989 Edition 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 Xl, Division 1, IWF-1230.
Additional LRA Clarification LRA Section 4.3.3, Effects of Reactor Water Environment on Fatigue Life, is revised to delete the tenth paragraph as follows.
FGo these locations with CUEs less than 1.0, the TU.,. has been Pr.jp.ted through the poriod of oxton~ded operation per 1CRI21 (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 Docket Nos. 50-247 & 50-286 Page 5 of 5 The site will continue to implement cemmitments assoeiated with (1) NRC
- ioders, Bulletis and-Generic Letters associmatod with nfickel alloys and (2) staff accepted industry guidelines.
The site Commits to COMDIV with future aDDlicable 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 continue to implement commitments associated with (1) NRC Orders, Bulletins and G.eneric Letters associated with nickel alloys and (2) staff accrepted industr,' guidelines.
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)
I*
A IlI II I
IIEG Will l Intnue to impIement commitments aSSIiateI wlth1 (1) NIV U*rpes, Uuil1tins and Gen-r-I Letters assocf-iated with Rickel alloys and (2) staff accepted industy guidelines.
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-1 801, 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 following ex(Gepif.
Attributes Affcctcd Emeeptiens
- 4. Detection of Aging E~ffects N.UREG 1801 specifies periodic updates of fatigue usage calculations. The IPEC programn updates fatigue usage calculations when the number of actual cycles apprac the analyzed number of cycles E~eepon-Ntes 4-- Updateo*.f fatigue uage a*.culat**,.
a.e nt.necw
,y unless the numbe, r of a,.ccmulated fatigue cycles approaches tha numbr of analyzed de'ign c*c*l*.
The iPECG prrogram pr.idoS for peFr*odI assesMeAt ef thevnu,.
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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-08-092 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 3.3.2-19-4-1P2, Condensate System, is revised as follows.
Table 3.3.2-19-4-1P2: Condensate System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Flow element Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatiue fatigue (S-08)
Flow element Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and Secondary Heat exchanger Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A,
(shell) boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Piping Pressure Carbon Treated water Cracking-TLAA-metal VIII.B1-10 3.4:1-1 C
boundary steel int) fatigue fatigue (S-08)
Piping Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary PnPressure Stainless Treated water Cracking-TLAA-metal VII.E1-16 3.3.1-2 C
boundary steel
> 140°F (int) fatigue (A-57) 302
NL-08-092 Docket Nos. 50-247 & 50-286 Page 2 of 15 Table 3.3.2-19-4-1P2: Condensate System Component Intended Aging Effect Aging NUREG-Table 1 Notes Type Inten Material Environment Requiring Management 1801 Vol.
Item Management Programs 2 Item Pressure Stainless Treated water. Loss of Water VIII.B1-4 3.4.1-16 A,
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Strainer housing Pressure Carbon Treated water Cracking-TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int fatigue fatigue (S-08)
Strainer housing Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (intl material Chemistry (S-1) 314 Control -
Primary and Secondary Strainer housing Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Strainer housing Pressure Stainless Treated water Cracking -
TLAA - metal VII.El-16 3.3.1-2 C
boundary steel
> 140°F (int) fatigue fatigue (A-57) 302 Strainer housingq Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 A,
boundary steel
> 1409F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Tank Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and Secondary Tank Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
'boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary
NL-08-092 Docket Nos. 50-247 & 50-286 Page 3 of 15 Table 3.3.2-19-4-lP2: Condensate System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Tank Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 A
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Thermowell Pressure Carbon Treated water Crackingq-TLAA-metal VIII.B1-10 3.4.1-1 C
boundary steel (int fatigue fatigue S-08)
Thermowell Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (intl material Chemistry (S-1) 314 Control -
Primary and Secondary Thermowell Pressure Stainless Treated water Cracking Water VIII.E-30, 3.4.1-14 A,
boundary steel
> 140[F (nt)
Chemistry (SP-17) 314 Control -
Primary and Secondary Thermowell Pressure Stainless Treated water Crackinq -
TLAA - metal VII.El-16 3.3.1-2 C
boundary steel
> 140°F (int) fatigue fatigue (A-57) 302 Thermowell Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 A
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Valve body Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08)
Valve body Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and Secondary Valve body Pressure Gray cast Treated water Cracking-TLAA-metal VIII.B1-10 3.4.1-1 C
boundary iron (int) fatigue fatigue S-08)
NL-08-092 Docket Nos. 50-247 & 50-286 Page 4 of 15 Table 3.3.2-19-4-1P2: Condensate System Component Intended Aging Effect Aging NUREG-Table 1 Cmpen Fnctind Material Environment Requiring Management 1801 Vol.
Notes Type Function Management Programs 2 Item Item Valve body Pressure Gray cast Treated water Loss of Water VIII.E-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 VIII.E-30 3.4.1-14 A,
boundary steel
> 140OF (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Valve body Pressure Stainless Treated water Crackingq-TLAA - metal VII.El-16 3.3.1-2 C,
boundary steel
> 140°F (int) fatiue fatigue (A-57) 302 Valve body Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 A
boundary steel
> 140OF (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary LRA Table 3.3.2-19-5-1P2, Chemical and Volume Control System, is revised as follows.
Table 3.3.2-19-5-1P2: Chemical and Volume Control System Aging Effect Aging NUREG-Component Intended Material Environment Requiring Management 1801 Vol.
Table 1 Notes Type Function Management Programs 2 Item Item Filter housing Pressure Stainless Air-indoor None None VII.J-16 3.3.1-A boundary steel (ext)
(AP-18) 99 Filter housing Pressure Stainless Treated Loss of Water VII.El-17 3.3.1-A boundary steel borated water material Chemistry (AP-79) 91 (int)
Control -
Primary and Secondary Heat exchanger Pressure Stainless Air - indoor None None VII.J-16 3.3.1-A housing boundary steel (ext)
(AP-18) 99
NL-08-092 Docket Nos. 50-247 & 50-286 Page 5 of 15 Table 3.3.2-19-5-1P2: Chemical and Volume Control System Component Intended Aging Effect Aging NUREG-Table 1 Type Fntin Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Heat exchanger Pressure Stainless Treated Loss of Water VII.E1-17 3.3.1-E housinq boundary steel borated water material Chemistry (AP-79) 91 (int)
Control -
Primary and Secondary LRA Table 3.3.2-19-17-1P2, Heating, Ventilation, and Air Conditioning System, is revised as follows.
Table 3.3.2-19-17-1P2: Heating, Ventilation, and Air Conditioning System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
te Notes Type Function Management Programs 2 Item Item Strainer housinq Pressure Carbon Air - indoor Loss of External VII.F1-2 3.3.1-A boundary steel (ext) material Surfaces (A-10) 56 Monitoring Strainer housing Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08) 309 Strainer housing Pressure Carbon Treated water Loss of Water VIII.B1 -11 3.4.1-4 C,
boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and Secondary LRA Table 3.3.2-19-24-1P2, Miscellaneous System, is revised as follows.
Table 3.3.2-19-24-1P2: Miscellaneous System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Pii Pressure Carbon Air - indoor Loss of External V.B-1 3.2.1-32 E
boundary steel (int) material Surfaces (E-25)
Monitorinq
NL-08-092 Docket Nos. 50-247 & 50-286 Page 6 of 15 Table 3.3.2-19-24-IP2: Miscellaneous System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Valve body Pressure Carbon Air - indoor Loss of External V.B-1 3.2.1-32 E
boundary steel (int) material Surfaces (E-25)
Monitorincq LRA Table 3.3.2-19-31-1P2, Radiation Monitoring System, is revised as follows.
Table 3.3.2-19-31-1P2: Radiation Monitoring System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Pump casing Pressure Stainless Air-indoor None None VII.J-15 3.3.1-A boundary steel (ext)
(AP-17) 94 Pump casingq Pressure Stainless Treated water Cracking Water VIII.B1-5 3.4.1-14 C
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Pump casing Pressure Stainless Treated water Loss of Water VIII.B1-4 3.4.1-16 C,
boundary steel
> 140OF (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Pump casing Pressure Stainless Treated water Loss of Water VII.C2-10 3.3.1-B boundary steel (intq material Chemistry (A-52) 50 Control -
Closed Cooling Water Pump casing Pressure Carbon Air - indoor Loss of External VII.1-8 3.3.1-A boundary steel (ext) material Surfaces (A-77) 58 Monitoring Pump casing Pressure Carbon Raw water Loss of Periodic VII.C1-19 3.3.1-E boundary steel (intl material Surveillance (A-38) 76 and Preventive Maintenance
NL-08-092 Docket Nos. 50-247 & 50-286 Page 7 of 15 Table 3.3.2-19-31-1P2: Radiation Monitoring System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Tank Pressure Stainless Air.- indoor None None VII.J-15 3.3.1-A boundary steel (ext)
(AP-17) 94 Tank Pressure Stainless Treated water Loss of Water VIII.E-40 3.4.1-6 C
boundary steel (int) material Chemistry
(
314 Control -
Primary and Secondary LRA Table 3.3.2-19-33-1P2, Station Air System, is revised as follows.
Table 3.3.2-19-33-1P2: Station Air System Component Intended Aging Effect Aging NUREG-Table 1 Tmpoen nctndd Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item CGeR;PF9660F Prpe&,- e CGarben A.6 ne-n es-e F
tr xtemnal W4lP--3 33.A-A, housin Y steel (e"4t nater-iel su4eee (A. 80)
GeRpressei P-reese Ga*ben Treated 4wae Less-ef Water V.CG2!4 3.31
--4 hquei4ng bhe'U4ary Steel k44 Fate~ial Ghe*nist&
(A-a64 47 CGRtrol Glesed
____~i~
Ge§Waet&
NL-08-092 Docket Nos. 50-247 & 50-286 Page 8 of 15 LRA Table 3.3.2-19-9-1P3, Condensate Pump Suction System, is revised as follows.
Table 3.3.2-19-9-1P3: Condensate Pump Suction System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Expansion *oint Pressure Elastomer Air-indoor Chanqe in Periodic VII.F1-7 3.3.1-E boundary (ext) material Surveillance (A-17) 11 properties and Preventive Maintenance Expansion ioint Pressure Elastomer Air-indoor Cracking Periodic VII.F1-7 3.3.1-E boundary (ext)
Surveillance A
11 and Preventive Maintenance Expansion ioint Pressure Elastomer Treated water Change in Periodic VII.A4-1 3.3.1-E boundary (int) material Surveillance (A-16) 12 properties and Preventive Maintenance
NL-08-092 Docket Nos. 50-247 & 50-286 Page 9 of 15 Table 3.3.2-19-9-1P3: Condensate Pump Suction System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Expansion joint Pressure Elastomer Treated water Cracking Periodic VII.A4-1 3.3.1-E boundary (int)
Surveillance 12 and (A-16)
Preventive Maintenance LRA Table 3.3.2-19-16-1P3, Emergency Diesel Generator System, is revised as follows.
Table 3.3.2-19-16-1P3: Emergency Diesel Generator System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
te Notes Type Function Management Programs 2 Item Item Pressure Carbon Treated water Loss of Water VII.H2-23 3.3.1-B boundary steel (int) material Chemistry (A-25) 47 Control -
Closed Cooling Water LRA Table 3.3.2-19-27-1P3, Heater Drain / Moisture Separator Drains / Vents System, is revised as follows.
Table 3.3.2-19-27-1P3: Heater Drain I Moisture Separator Drains / Vents System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Expansion ioint Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Expansion ioint Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 A,
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary
NL-08-092 Docket Nos. 50-247 & 50-286 Page 10 of 15 Table 3.3.2-19-27-1P3: Heater Drain / Moisture Separator Drains / Vents System Component Intended Aging Effect Aging NUREG-Table 1 Cmpen Fnctind Material Environment Requiring Management 1801 Vol Notes Type Function Management Programs 2 Item Item Heat exchanger Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 C
(shell) boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and Secondary Orifice Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Orifice Pressure Stainless Treated water Cracking -
TLAA - metal VII.E1-16 3.3.1-2 CQ, boundary steel
> 140°F (int) fatigue fatigue (A-57) 302 Orifice Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 A
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Piping Pressure Carbon Treated water Cracking -
TLAA-metal VIIIB1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08)
Pining Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A,
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Pinin Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Piping Pressure Stainless Treated water Cracking -
TLAA - metal VII.El-16 3.3.1-2 C
boundary steel
> 140°F (int) fatigue fatigue (A-57) 302 Piig Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 A,
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and
NL-08-092 Docket Nos. 50-247 & 50-286 Page 11 of 15 Table 3.3.2-19-27-1P3: Heater Drain / Moisture Separator Drains / Vents System Component Intended Aging Effect Aging NUREG-Table 1 Cmpen Fnctind Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Secondary Pump casing Pressure Carbon Steam-("+vt, Loss of Water VIltA46 3.4-.- 2 4--
boundary steel Treated water material Chemistry
(&-W6) 3.4.1-4 344 (int)
Control -
A, Primary and VIII.E-34 314 Secondary (S-10)
Sight glass Pressure Carbon Treated water Cracking-Periodic VIII.B1-10 3.4.1-1 E
boundary steel (int) fatigue surveillance (S-08) and preventive maintenance Sight glass Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (5-10 314 Control -
Primary and Secondary Sight glass Pressure Glass Treated water None None VIII.1-8 3.4.1-40 A
boundary (int)
(SP-35)
Strainer housing Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08)
Strainer housinq Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A,
boundary steel (intl material Chemistry (S-10) 314 Control -
Primary and Secondary Tank Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Thermowell Pressure Carbon Treated water Cracking -
TLAA-metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue
(
M
NL-08-092 Docket Nos. 50-247 & 50-286 Page 12 of 15 Table 3.3.2-19-27-1P3: Heater Drain / Moisture Separator Drains / Vents System C e eAging Effect Aging NUREG-Table I Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Thermowell Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Tubing Pressure Stainless Treated water Cracking Water VIII.E-30 3.4.1-14 A
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Tubing Pressure Stainless Treated water Cracking -
TLAA - metal VII.E1-16 3.3.1-2 C
boundary 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
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Valve body Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel int) fatigue fatigue
(-08)
Valve body Pressure Carbon Treated water Loss of Water VIII.E-34 3.4.1-4 A
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Valve body Pressure Stainless Treated water Cracking Water VIII.E-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 VII.E1-16 3.3.1-2 C
boundary steel
> 140OF (int) fatigue fatigue (A-57) 302 Valve body Pressure Stainless Treated water Loss of Water VIII.E-29 3.4.1-16 A
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary
NL-08-092 Docket Nos. 50-247 & 50-286 Page 13 of 15 LRA Table 3.3.2-19-45-1P3, Reheat Steam System, is revised as follows.
Table 3.3.2-19-45-1P3: Reheat Steam System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item Heat exchanger Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A,
(shell' boundary steel tint) material Chemistry (S-10) 314 Control -
Primary and Secondary Piping Pressure Carbon Treated water Cracking-TLAA-metal VIII.B1-10 3.4.1-1 C
boundary steel int) fatigue fatigue (S-08)
Piping Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A,
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Steam trap Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B13-10 3.4.1-1 C
boundary steel (int fatigue fatigue (S-08)
Steam trap Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A,
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Strainer Pressure Carbon Treated water Cracking -
TLAA - metal VIII.B1-10 3.4.1-1 C
Housing boundary steel int) fatigue fatigue
(-08)
Strainer Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A
Housing boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary Thermowell Pressure Carbon Treated water Crackingq-TLAA - metal VIII.B1-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08)
Thermowell Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A
boundary steel (int) material Chemistry (S-1) 314 Control -
Primary and
NL-08-092 Docket Nos. 50-247 & 50-286 Page 14 of 15 Table 3.3.2-19-45-1P3: Reheat Steam System Component Intended Aging Effect Aging NUREG-Table 1 Type Function Material Environment Requiring Management 1801 Vol.
Item Notes Management Programs 2 Item Secondary Tubing Pressure Stainless Treated water Cracking Water VIII.C-2 3.4.1-14 A,
boundary steel
> 140°F (int)
Chemistry (SP-17) 314 Control -
Primary and Secondary Tubinq Pressure Stainless Treated water Cracking-TLAA-metal VII.E1-16 3.3.1-2 C,
boundary steel
> 140°F (int) fatiue (A-57) 302 Tubing Pressure Stainless Treated water Loss of Water VIII.C-1 3.4.1-16 A
boundary steel
> 140°F (int) material Chemistry (SP-16) 314 Control -
Primary and Secondary Valve body Pressure Carbon Treated water Cracking-TLAA-metal VIII.B13-10 3.4.1-1 C
boundary steel (int) fatigue fatigue (S-08)
Valve body Pressure Carbon Treated water Loss of Water VIII.C-7 3.4.1-4 A
boundary steel (int) material Chemistry (S-10) 314 Control -
Primary and Secondary
NL-08-092 Docket Nos. 50-247 & 50-286 Page 15 of 15 LRA Table 3.3.2-19-48-lP3, Station Air System, is revised as follows.
Table 3.3.2-19-48-1P3: Station Air System Aging Effect Aging NUREG-Table 1 Component Intended Material Environment Requiring Management 1801 Vol.
Item Notes Type Function Management Programs 2 Item GenmpeeeeF P~ees-rue Ga~bGR AiF
- Rdeer Less-el N111a VD42
-3.3.1
-A housing boundAWY steel
(-ext mater-l (A80) 67 GGRempSrsG P-ressFw GaFbGR Treated-water Iess-of We V.G214
-3.3.
-Q ho..
gb steel (4tM" rFatetel CheRostr (A-2E4 47 Clesed GGeeRgWateF Valve body Pressure Carbon Air-indoor Loss of External VII.D-3 3.3.1-A boundary steel (ext) material Surfaces (A-80) 57 Monitoring Valve body Pressure Carbon Condensation Loss of Periodic VII.D-2 3.3.1-E boundary steel (int) material Surveillance (A-26) 53 and Preventive Maintenance LRA Section A.1.28, Periodic Surveillance and Preventive paragraph, twenty-ninth bullet is revised as follows.
Maintenance Program, second 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 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 an IP2 and IP3 average capacity factors of 995% 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 995% 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 95>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 be'revised 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 (PORV) setpoints will be re-evaluated when pressure/temperature curves are submitted.
NL-08-092 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 (PORV) setpoints will be re-evaluated when pressure/temDerature 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-1 3587, 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 inner 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-1 3587 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-1 3587 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 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-1 023 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-lbs and ASME Code Section Xl, Appendix K.
IPEC concluded that WCAP-1 3587 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 Cooldown 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 ARTNDT 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 Trcd 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 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 RTNDT Values for the Indian Point Unit 2 Reactor Vessel Materials Material, Description
-Cu(%)
Ni(%)
Initial RTNDT(a)
Closure Head Flange 60OF Vessel Flange 60OF Intermediate Shell Plate B-2002-1(e) 0.19 (0.21) 0.65 (0.62) 340 F Intermediate Shell Plate B-2002-2(e) 0.17 (0.15) 0.46 (0.44) 21 OF Intermediate Shell Plate B-2002-3(e) 0.25(0.20) 0.60 (0.59) 21OF Lower Shell Plate B-2003-1 0.20 0.66 20OF Lower Shell Plate B-2003-2 0.19 0.48
-20OF Intermediate & Lower Shell Longitudinal Weld Seams (Heat # W5214)(b' d) 0.21 1.01
-56 0 F Intermediate to Lower Shell Girth Weld 0.19 1.01
-560 F (Heat # 34B009) (c, d)
Indian Point Unit 2 Surveillance Weld 0.20 0.94 (Heat # W5 214 )(b1 d)
Indian Point Unit 3 Surveillance Weld 0.16 1.12 (Heat # W521 4 )(b. d)
H.B. Robinson Unit 2 Surveillance Weld 0.32 0.66 (Heat # W5214)(b, d)
Notes:
(a) The Initial RTNDT values are measured values, with exception to the weld materials.
(b) The weld material in theIndian Point Unit 2 surveillance program was made of the same wire and flux as the reactor vessel intermediate shell longitudinal weld seams (Wire Heat No. W5214 RACO3 + 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 RACO3 + Ni200, Flux Type Linde 1092, Flux Lot No. 3708.
(d) The weld best estimate copper and nickel weight percents were obtained from CE Reports NPSD-1 039, Rev. 2 and/or NPSD-1 119, Rev. 1. The values from the CE Report NPSD-1 119, 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 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 1 14°F. Intermediate shell plate B-2002-3 is more limiting than B-2002-1 even if the highest CF were used for B-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 T
2.53 x 10'" n/cm', (E > 1.0 MeV)
Y 4.55 x 101' n/cm 2, (E > 1.0 MeV)
Z 1.02 x 1019 n/cm2, (E > 1.0 MeV)
V 4.92 x 1018 n/cm2, (E > 1.0 MeV)
T 2.63 x.10 8 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/cM2, (E > 1.0 MeV)(a)
X 8.74 x 1018 n/cm 2, (E > 1.0 MeV)(a)
H.B.RobisonUnit2 S
4.79 x 10'8 n/cm 2, (E > 1.0 MeV) (a)
V 5.30 x 1018 n/cm2, (E > 1.0 MeV)(a)
T 3.87 x 1019 n/cm2, (E > 1.0 MeV)(a)
X 4.49 x 1018 n/cm2, (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 Docket Nos. 50-247 & 50-286 Page 6 of 8 TABLE 2-3 Inlet (Tcold) Operating Temperatures Indian Point Unit 2 Indian* Point Unit 3!,
H.B. Robinson Unit 2 5430F (Cycle 1) 540°F (Capsule T) 5470F (Capsule S) 5430F (Cycle 2) 540°F (Capsule Y) 5470F (Capsule T) 522.50F (Cycle 3) 540°F (Capsule Z) 5470F (Capsule X) 522.5°F (Cycle 4) 540°F (Capsule X) 522.8°F (Cycle 5) 522.80 F (Cycle 6) 522.80F (Cycle 7) 5 2 2.5 0 F ( C y c l e 8 )
7,528OF (Average) 5406F (Average) 5470 (Average)
(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 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 Material Capsule C~apsl FFb ARTNDT~c) ~
FF*ARTN DT FF2 Intermediate T
0.253 0.627 55.0 34.49 0.393 Shell Z
1.02 1.006 125.0 125.75 1.012 She SUM:
160.24 1.405 Plate B-2002-1 "
CFB-2002-1_ f XO),(FF *RTNT (FFr 2)=-(160.24)
(1.405) =111 4.0-F T
0.253 0.627 95.0 59.57 0.393 Intermediate Z
1.02 1.006 120.0 120.72 1.012 Shell V
0.492 0.802 77.0 61.75 0.643 Plate B-2002-2 SUM:
242.04 2.048 CFB2002-L..FE RTNDT F
2)= :(242.04).
(20i
= 18.2 0F T
0.253 0.627 115.0 72.11 0.393 Intermediate Y
0.455 0.781 145.0 113.25 0.610 Shell.
Z 1.02 1.006 180.0 181.08 1.012 Plate B-2002-3 SUM:
366.44 2.015 CF-022= L(ff RTN
- X FE2) = (366.44) -(2.015)
= 181.9'F Y (IP2) 0.455 0.781 208.7(195) 162.9 0.610 V (IP2) 0.492 0.802 218.3 (204) 175.1 0.643 T (IP3) 0.263 0.637 183.2(151.6) 116.7 0.405 Y (IP3) 0.692 0.897 206.1(172.0) 184.8 0.804 Surveillance Z (IP3) 1.04 1.011 270.1 (229.2) 273.1 1.022 Weld X(IP3)
.874
.962 229.8 (193.2) 221.1 0.926 Material(d)
V(HBR2) 0.530 0.823 248.9(209.3) 204.7 0.677 T(HBR2) 3.87 1.349 334.8 (288.2) 451.6 1.820 X(HBR2) 4.49 1.381 310.6 (265.9) 428.8 1.906 SUM:
2218.9 8.813 Notes:
CF sur, Weid =,(FIF F RTNo4A*):
FF2 E
(221 8.96F)*+ (8.813) = 251.8°*
Notes:
(a) f = fluence. See Table 2-3, (x 1019 n/cm2, E > 1.0 MeV).
(b)
FF = fluence factor = f(O 28 -0. log (c)
ARTNDT values are the measured 30 ft-lb 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 5400 F, H.B.
Robinson Unit 2 operates with an inlet temperature of approximately 5470 F, and Indian Point Unit 2 operates with an inlet temperature of approximately 5280F. The measured ARTNDT values from the Indian Point Unit 3 surveillance program were adjusted by adding 120 F to each measured ARTNDT and the H.B. Robinson Unit 2 surveillance program values were adjusted by adding 190F to each measured ARTNDT value before applying the ratio procedure. The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of:
Ratio IP2 = 230.2 - 214.3 = 1.07 for the Indian Point Unit 2 data.
Ratio IP3 = 230.2 ÷206.2 = 1.12 for the Indian Point Unit 3 data.
Ratio HBR2 = 230.2 + 210.7 = 1.09 for the H.B. Robinson Unit 2 data.
Refer to Table 2-5 for the longitudinal weld seam CF of 230.20F.
(The pre-adjusted values are in parenthesis.)
NL-08-092 Docket Nos. 50-247 & 50-286 Page 8 of 8 TABLE 2-5 Summary of the Indian Point Unit 2 Reactor Vessel Beitline Material Chemistry Factors Intermediate Shell Plate B-2002-1 1440F 114 Intermediate Shell Plate B-2002-2 115.10 F 118.2 Intermediate Shell Plate B-2002-3 1760 F 181.9 Lower Shell Plate B-2003-1 1520F Lower Shell Plate B-2003-2 (a) 128.8 0F Intermediate & Lower Shell Longitudinal Weld Seams 230.20F 251.8 (Heat # W5214)
Intermediate to Lower Shell Girth Weld Seam (Heat # 34B009) 220.9OF Indian Point Unit 2 Surveillance Weld (Heat # W5214)
Indian Point Unit 3 Surveillance Weld (Heat # W5214)
H.B. Robinson Unit 2 Surveillance Weld (Heat # W5214)
(a) The 128.80F CF listed here differs from previous analyses that used an excessively conservative CF of 1420F for this material. If the 142 0F CF had been used in this analysis, this material would still not be limiting.