NL-02-006, Response to Request for Additional Information Indian Point 2 License Amendment Request for Reactor Coolant System Heatup and Cooldown Limitation Curves

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Response to Request for Additional Information Indian Point 2 License Amendment Request for Reactor Coolant System Heatup and Cooldown Limitation Curves
ML020170071
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 01/11/2002
From: Dacimo F
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-02-006, TAC MB2419
Download: ML020170071 (66)
v • d • e


Text

"EntergyNuclear Northeast Entergy Nuclear Operations, Inc.

Indian Pont Energy Center

'--' 295 Broadway, Suite 1 P.O. Box 249 Buchanan, NY 10511-0249 January 11, 2002 Re: Indian Point Unit No. 2 Docket No. 50-247 NL-02-006 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

Response to Request for Additional Information Indian Point 2 License Amendment Request for Reactor Coolant System Heatup and Cooldown Limitation Curves (TAC No.: MB2419)

References:

1. Consolidated Edison letter (NL-01 -092) to NRC, "Indian Point 2 License Amendment Request for Reactor Coolant System Heatup and Cooldown Limitation Curves and Request for Exemption from the Requirements of 10CFR50.60(a) and Appendix G," dated July 16, 2001 By letter dated July 16, 2001 (Ref. 1), Consolidated Edison (the former licensee) submitted an application for an amendment to the Technical Specifications (TS) for Indian Point Unit No. 2 (IP2). The proposed amendment requested revised Reactor Coolant System Heatup and Cooldown Limitation Curves, as well as new Overpressure Protection System (OPS) limits. The U.S. Nuclear Regulatory Commission (NRC) staff reviewed this submittal, determined that additional information was required to complete its review, and requested that additional information in telephone conferences on November 14, 2001 and December 18, 2001. As a result of the telephone conferences, Entergy Nuclear Operations, Inc. (ENO - the current licensee) initiated a revision to the original Ref. 1 Attachment 4, "WCAP-15629, Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

Revision 1 to WCAP-15629 (December 2001) is included as Enclosure 1 of this submittal.

In addition, excerpts from an existing IP2 operating procedure for plant heatup are being submitted as an example to demonstrate how IP2 applies instrument uncertainty to the values in the TS curves. Although the 10CFR50, Appendix G pressure/temperature limitations included in the Indian Point 2 Technical Specifications do not include explicit margins to account for instrument uncertainties, the limits in the operating procedures are decreased to account for pressure and temperature uncertainties, as well as system hydraulic losses and elevation corrections. Attachment 1 of this submittal contains excerpts from an IP2 operating procedure.

This letter contains no new commitments.

NL 02-006 Page 2 of 4 Should you or your staff have any questions regarding this submittal, please contact Mr.

John F. McCann, Manager, Nuclear Safety and Licensing at (914) 734-5074.

Sincerely, Fred Dacimo Vice President - Operations Indian Point 2 Attachment Enclosure cc: See page 3

NL 02-006 Page 3 of 4 cc:

Mr. Hubert J. Miller Regional Administrator-Region I US Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1498 Mr. Patrick D. Milano, Senior Project Manager Project Directorate I-1 Division of Licensing Project Management US Nuclear Regulatory Commission Mail Stop 0-8-C2 Washington, DC 20555-0001 Senior Resident Inspector US Nuclear Regulatory Commission Indian Point Unit 2 PO Box 38 Buchanan, NY 10511 Mayor, Village of Buchanan 236 Tate Avenue Buchanan, NY 10511 Mr. Paul Eddy NYS Department of Public Service 3 Empire Plaza Albany, NY 12223 Mr. William F. Valentino, President NYS ERDA Corporate Plaza West 286 Washington Ave. Extension Albany, NY 12223-6399

NL 02-006 Page 4 of 4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

ENTERGY NUCLEAR OPERATIONS, INC. ) Docket No. 50-247 Indian Point Nuclear Generating Unit No. 2 )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Commission (NRC), Entergy Nuclear Operations, Inc., as holder of Facility Operating License No.

DPR-26, hereby submits additional information in support of the July 16, 2001 application for amendment of the Technical Specifications contained in Appendix A of this license. The specific additional information is set forth in Enclosure 1 and .

As required by 10CFR50.91 (b)(1), a copy of this submittal has been provided to the appropriate New York State official designated to receive such amendments.

BY: '- 2.

F redý Dacimo Vice President - Operations Indian Point 2 Subscribed and sworn to efore me this _z-I day 2002.

Notary Public EASILIA A.AMANNA NoWy Puft 8W9 of NmvYork No, 01AMS08*89 aueed InWestcuhster coun Commiwedo" fto Maroh 20, 2M20

ATTACHMENT 1 TO NL-02-006 3 pages from an Indian Point Unit 2 operating procedure

PLANT RESTORATION FROM COLD SHUTDOWN NL-02-006 TO HOT SHUTDOWN CONDITIONS {COMMITMENT: 6.2.141 Attachment 1 Page 2 of 4 2.3 The Reactor shall be maintained subcritical by at least 1 percent K/K UNTIL an ACTUAL water level of 33 - 40 percent is established in the Pressurizer (Technical Specification 3.1 .C.4).

2.4 RCS pressure increases should be limited to 100 psig per hour when above 1700 psig to limit the potential for safety valve leakage.

NOTE

" IF any heatup OR cooldown rate is violated, a safety evaluation SHALL be performed. (Reference 6.2.15)

" The heatup, and cooldown rates in Technical Specification Figure 3.1 .B-1, and 3.1 .B-2 do NOT make allowance for instrument error. Compensation for pressure, and temperature instrumentation error is as follows:

o During Steady-State, use Figure 1 (DM), for RCS Pressure, and Tempenotuie inrument i cumpIsidn. ....

o Durinn Heatup, use Figure 2,(D3), for RCS Pressure and Temperature instrument error compensation.

" The CCR instrumentation to be used is as follows:

o RCS Temperature, as indicated on RCS Cold Leg RTD TE-413 (TR-413J), TE-433 (TR-433J), or TE-443 (TR-443J).

o RCS pressure above 1500 psig, as indicated on Pressurizer Pressure (if on scale), OR Wide Range indicated pressure on PT-402, or PT-403.

o RCS pressure 0 - 1500 psig, as indicated on PT-413 (PI-413K), PT-433 (PI-433K), or PT-443 (PI-443K).

2.5 RCS Heatup Requirements:

2.5.1 RCS RCS temperature, AND Pressure SHALL be maintained within the limits of Technical Specification Figure 3.1 .B-2, as compensated for, per Step Note, 2 nd Bullet, as applicable, AND Graph RCS-12A, 50°F Subcooling and Saturation Curves

PLANT RESTORATION FROM COLD SHUTDOWN NL-02-006 TO HOT SHUTDOWN CONDITIONS {COMMITMENT: 6.2.141 Attachment 1 Page 3 of 4 FIGURE 1 (DM), RCS TEMPERATURE VS. PRESSURE - STEADY STATE (CORRECTED FOR INSTRUMENT ERROR) 2500 i I I I I I I PZR Press. Ind.

2000 ........ WR Press. Rec.

..... -OPS Press. Ind.

0) 1500 0 Deg/hr Cooldown limit 0.

CL 1000 cn 500 0

0 50 100 150 200 250 300 350 400 RCS Temperature (Deg. F)

PLANT RESTORATION FROM COLD SHUTDOWN NL-02-006 TO HOT SHUTDOWN CONDITIONS MCOMMITMENT: 6.2.141 Attachment 1 Page 4 of 4 FIGURE 2 (D3), RCS TEMPERATURE VS. PRESSURE - HEATUP (CORRECTED FOR INSTRUMENT ERROR)

RCS Heatup Limitations 2500 I4 -- 60 deg/hr

- 100 deg/hr II 2000

_s 1500

-1000

/

500 0

0 50 100 150 200 250 300 350 400 RCS Temperature (Deg. F)

ENCLOSURE 1 TO NL-02-006 WCAP-15629, Revision 1, "Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation" ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT UNIT NO. 2 DOCKET NO. 50-247

Westinghouse Non-Proprietary Class 3 Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation I

Westinghouse Electric Company LLC

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15629, Revision 1 Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation T. J. Laubham December 2001 Prepared by the Westinghouse Electric Company LLC for Entergy Approved ý!V &'-Xý C. H. Boyd, ManageH Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2001 Westinghouse Electric Company LLC All Rights Reserved

ii PREFACE This report has been technically reviewed and verified by:

Section 1 through 6 and Appendices A, C through G J.H. Ledger i .

S.L. Anderson Q~)JSJL Appendix B Record of Revision Revision 0: Original Issue Revision 1: The following was revised in this revision:

"* Updated text on pages 2, 22, C-I and G2 to address typos.

"* Added clarification to the plate chemistry values in Table 1 (Page 3), and revised the nickel value for the Lower Shell Plate B-2003-2. In turn the chemistry factor for the lower shell plate B-2003-2 was revised in Table 5 (Page 8). This chemistry factor changed resulted in changes to Tables 9 and 10 (Pages 16 & 17).

"* Clarified the references for the unirradiated USE in Table D- 1. This resulted in adding Reference 17.

"* Changed note on page E-I to read, 'Vithdrawal Schedule to be provided in PTLR only by Indian Point Unit 2".

TABLE OF CONTENTS LIST O F TA B L E S.................................................................................................................................. iv LIST O F FIGURES ................................................................................................................................. v EX EC UTIV E SUM M A RY ..................................................................................................................... vi 1 INTRODUCTION ........................ ...... .................................... . .. 1 2 FRACTURE TOUGHNESS PROPERTIES .......................................................................... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .............. 9 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .................................... 13 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 19 6 R EFE REN CE S ......................................................................................................................... 27 APPENDIX A: PRESSURIZED THERMAL SHOCK (PTS) RESULTS ........................................ A-0 APPENDIX B: CALCULATED FLUENCE DATA ............................................................................. B-0 APPENDIX C: UPDATED SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES ................ C-0 APPENDIX D: REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES ..................................................................... D-0 APPENDIX E: UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE ...................... E-0 APPENDIX F: ENABLE TEMPERATURE CALCULATIONS AND RESULTS ............................... F-0 APPENDIX G: PRESSURE TEMPERATURE LIMIT CURVES USING CODE CASE N-588 ........... G-0

iv LIST OF TABLES Table 1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTrDT Values for the Indian Point Unit 2 Reactor Vessel Materials .................................................. 3 Table 2 Inlet (Tcold) Operating Temperatures .................................................................... 4 Table 3 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Indian Point Unit 2, Indian Point Unit 3 and H.B. Robinson Unit 2 ......................................... 5 Table 4 Calculation of Chemistry Factors using Indian Point Unit 2 Surveillance Capsule Data .... 6 Table 5 Summary of the Indian Point Unit 2 Reactor Vessel Beltline Material Chemistry Factors .. 8 Table 6 Calculated Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10'9 n/cm2, E > 1.0 MeV) ................................................ 14 Table 7 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 25 EFPY Heatup/Cooldown Curves .............................................. 14 Table 8 Summary of the Calculated Fluence Factors used for the Generation of the 25 EFPY Heatup and Cooldown Curves .................................................................................. 15 Table 9 Calculation of the ART Values for the 1/4T Location @ 25 EFPY .............................. 16 Table 10 Calculation of the ART Values for the 3/4T Location @ 25 EFPY .............................. 17 Table 11 Summary of the Limiting ART Values Used in the Generation of the Indian Point Unit 2 Heatup/Cooldown Curves ............................................................................... 18 Table 12 25 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) ........................................................ 23 Table 13 25 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) ........................................................ 25

LIST OF FIGURES Figure 1 Indian Point Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 & 100°F/hr) Applicable for the First 25 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology ..... 21 Figure 2 Indian Point Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology ..... 22

vi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for normal operation of the Indian Point Unit 2 reactor vessel. In addition, Pressure Temperature Limits Report (PTLR) support information, such as Fluence, PTS , EOL USE and Withdrawal Schedule, are documented herein under the Appendices. The PT curves were generated based on the latest available reactor vessel information and updated fluences (Appendix B). The new Indian Point Unit 2 heatup and cooldown pressure-temperature limit curves were generated using ASME Code Case N-640t 33 (which allows the use of the K1, methodology) and the axial flaw methodology of the 1995 ASME Code,Section XI through the 1996 Addenda.

It should be noted that Indian Point was limited at the 1/4T location by the intermediate to lower shell circumferential weld and at the 3/4T location by the intermediate shell plate B-2002-3. The pressure temperature (PT) limit curves presented in Section 5 are those developed using the axial flaw methodology with the most limiting axial flaw adjusted reference temperatures (ARTs). Theses PT t 41 curves bound the PT curves that used the ASME Code Case N-588 (Circ. Flaw Methodology) with the most limiting Circ Flaw ARTs. The circ. flaw PT curves are presented in Appendix G herein.

1 INTRODUCTION Heatup and cooldown lnimt curves are calculated using the adjusted RTNDT (reference mil-ductility temperature) corresponding to the limiting beitline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RThryT is designated as the higher of either the drop weight nil ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 601F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in 51 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials.t Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 2[6],

"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values, not the best estimate fluence values (See Appendix B). 2) The Kir critical stress intensities are used in place of the KIa critical stress intensities. This methodology is taken from approved ASME Code Case N-640P3 ]. 3) The 1996 Version of Appendix G to Section XIf will be used rather than the 1989 version. 4) PT Curves were generated with the most limiting circumferential weld ART value in conjunction with Code Case N-588143. The curves, which are included in Appendix (4 are bounded by the curves using the standard "axial" flaw methodology from ASME Code 1996 App. G with the ART from the limiting plate material B-2002-3.

WCAP-15629

2 2 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 Plan3S]. The beltline material properties of the Indian Point Unit 2 reactor vessel is presented in Table 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 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. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Table 4. These CF values are summarized in Table 5. 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 to surveillance plate B-2002-2.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 2.

WCAP-15629

3 TABLE 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) 340F Intermediate Shell Plate B-2002-2(e) 0.17 (0.15) 0.46 (0.44) 21OF 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)(" d) 0.21 1.01 -56 0F Intermediate to Lower Shell Girth Weld (Heat # 0.19 1.01 -56 0F 34B009) (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)(" d)

H.B. Robinson Unit 2 Surveillance Weld (Heat 0.32 0.66

  1. W5214)0, d)

Notes:

(a) The Initial RTDrT 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 (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-1039, Rev. 2[151 and/or NPSD-1 119, Rev. 1[16]. The values from the CE Report NPSD-1119, Rev. 2 for the Indian Point 2 vessel axial and circ.

welds matches that 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 was considered, then the best estimate would match RVID2. The data above for the intermediate shell plates are conservative with exception to plate B-2002-1. The chemistry for plate B-2002-1 produces a Table chemistry factor of 156.2"F as compared to the chemistry factor calculated using credible surveillance data (114*F, See Tables 4 & 5). Thus, this non-conservative difference versus RVID2 is negligible. Values from WCAP-12796 will be used herein. RVID2 Values are in Parenthesis.

WCAP-15629

4 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 fluence values used to determine the CFs in Table 4 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases.

The measured ARTNDT values for the weld data were adjusted for temperature difference between differing plants and for chemistry using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2. See Table 2 for the Tcold operating temperatures at Indian Point Units 2 and 3 and H.B.

Robinson Unit 2.

TABLE 2 Inlet (Tcold) Operating Temperatures Indian Point Unit 2(a) Indian Point Unit 3(b) H.B. Robinson Unit 2(c) 543 0F (Cycle 1) 540 0F (Capsule T) 5471F (Capsule S) 543 IF (Cycle 2) 5401F (Capsule Y) 547°F (Capsule T) 522.5.F (Cycle 3) . 540. (Capsule Z) --

522.5°F (Cycle 4) 522.8-F (Cycle 5) 522.8*F (Cycle 6) 522.8 0F (Cycle 7) 522.5*F (Cycle 8) ...

S28F ~

(veage 4~F(Avrag ~5470F (Average )7 Notes:

(a) Confirmed by Indian Point Unit 2. Average over eight matches E900 Database. Note that cycle 8 is when the last capsule was withdrawn, IP2 is currently in cycle 15.

(b) Per E900 Database. Confirmed by Indian Point Unit 3.

(c) Per E900 Database the value for all Capsules at H.B. Robinson Unit 2 was 546"F, however Ted Huminski at 0 0 Robinson indicated that the Inlet Operating Temperatures was documented as being between 546 F and 547 F.

0 Thus, for conservatism (i.e. larger delta versus IP2) 547 F will be assumed.

WCAP-15629

5 All calculated fluence values (capsule and projections) for Indian Point Unit 2 were updated and documented in Appendix B. These fluences were calculated using the ENDF/B-VI scattering cross section data set. In addition, capsule fluences from Indian Point Unit 3 and H.B. Robinson Unit 2 are included since they share the same surveillance weld material and can be used in the calculation of chemistry factor. The Indian Point Unit 3 fluences are taken from Letter INT-00-21 I[', and the H.B.

0 Robinson fluences were taken from WCAP-14044[' *. The Indian Point Unit 3 fluences are calculated fluences using ENDF/B-VI cross-sections. The best available fluence data for H.B. Robinson are the fluences from WCAP-14044. Calculated fluences exist in WCAP-14044, however they were determined using ENDF/B-IV & V cross-sections and would increase if ENDF/B-VI cross-sections were used. Thus, for conservatism the calculated fluences were increased 15% to account for going to ENDF/B-VI and used herein for the calculation of chemistry factor. It should be noted that the measured fluences would not increase under ENDF/B-VI. Table 3 is a summary of the capsule fluences from Indian Point Unit 2 and 3 and H.B Robinson.

TABLE 3 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Indian Point Unit 2, Indian Point Unit 3 and H.B. Robinson Unit 2 Capsule IFluence Indian Point Unit 20)

T 2.53 x 1018 n/cm2, (E > 1.0 MeV)

Y 4.55 x 10" n/cm2, (E > 1.0 MeV)

Z 1.02 x 10"9 n/cm 2, (E > 1.0 MeV)

V 4.92 x 1018 n/cm 2, (E > 1.0 MeV)

Indian Point Unit 3°)

T 2.88 x 10"8 n/cm2, (E > 1.0 MeV)

Y 7.52 x 101" n/cm 2, (E > 1.0 MeV)

Z 1.12 x 1019 n/cm 2, (E > 1.0 MeV)

H.B. Robinson Unit 2(c)

S 5.80 x 1018 n/cm2, (E > 1.0 MeV)

V 6.20 x 101 n/cm2 , (E > 1.0 MeV)

T 4.66 x 1019 n/cr 2 , (E > 1.0 MeV)

NOTES:

(a) Per Appendix B.

(b) The fluences are calculated fluences per Letter INT-00-2 11 using ENDF/B-VI.

(c) The fluences are Calculated values per WCAP-14044 plus 15%.

WCAP-15629

6 TABLE 4 Calculation of Chemistry Factors using Indian Point Unit 2 Surveillance Capsule Data Material Capsule Capsule fa) FF(b) ARTNDT(C) FF*ARTNDT FF 2 Intermediate Shell T 0.253 0.627 55.0 34.49 0.393 Plate B-2002-1 Z 1.02 1.006 125.0 125.75 1.012 SUM: 160.24 1.405 CFB-200 2-1 = X(FF

  • RTNDT) +( FF2) = (160.24) - (1.405) = 114.0°F Intermediate Shell T 0.253 0.627 95.0 59.57 0.393 Plate B-2002-2 Z 1.02 1.006 120.0 120.72 1.012 V 0.492 0.802 77.0 61.75 0.643 SUM: 242.04 2.048 CFB-2002-2 = X(FF
  • RTNDT) + FF 2) = (242.04) (2.048) = 118.2*F Intermediate Shell T 0.253 0.627 115.0 72.11 0.393 Plate B-2002-3 Y 0.455 0.781 145.0 113.25 0.610 Z 1.02 1.006 180.0 181.08 1.012 SUM: 366.44 2.015 CFs-2oo2-2 = Y(FF
  • RTNDT) Y-( FF 2) = (366.44) - (2.015) = 181.9°F Surveillance Weld Y (IP2) 0.455 0.781 208.65 (195) 162.96 0.610 Material(d) V (IP2) 0.492 0.802 218.28 (204) 175.06 0.643 T (IP3) 0.288 0.660 173.6(143) 114.58 0.436 Y (IP3) 0.752 0.920 215.04 (180) 197.84 0.846 Z (IP3) 1.12 1.03 259.84 (220) 267.64 1.061 V(HBR2) 0.620 0.866 248.87 (209.32) 215.52 0.750 T(HBR2) 4.66 1.39 334.72 (288.08) 465.26 1.932 SUM: 1598.86 6.278 CF Su,. Weld = YX(FF
  • RTNDT) - Y-( FF2) = (1598.86°F) (6.278) 254.7 0 F See Next Page for Notes WCAP-15629

7 Notes:

(a) f= fluence. See Table 3, (x 10'9 n/cm2, E > 1.0 MeV).

(b) FF = fluence factor = f(o.2 -o.-lof).

(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 (Which Refers back to the Original Southwest Research Institute Report for each Capsule.)

- Indian Point Unit 3 Weld...WCAP-1 1815[" 1.

2

- H.B.Robinson Unit 2...Letter Report CPL-96-2031' ]

(d) Per Table 2 Indian Point Unit 3 operates with an inlet temperature of approximately 540°F, H.B. Robinson 0

Unit 2 operates with an inlet temperature of approximately 547 F, and Indian Point Unit 2 operates with an inlet temperature of approximately 5281F. The measured ARTN 1 values from the Indian Point Unit 3 surveillance program were adjusted by adding 12°F to each measured ARTrDT and the H.B. Robinson Unit 2 surveillance program were adjusted by adding 190F to each measured ARTcDT value before applying the ratio procedure. The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of:

Ratio IP2 = 230.2 + 215.8 = 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 -BR2 = 230.2 + 210.7 = 1.09 for the H.B. Robinson Unit 2 data.

(The pre-adjusted values are in parenthesis.)

WCAP-15629

8 TABLE 5 Summary of the Indian Point Unit 2 Reactor Vessel Beltline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 CF's Intermediate Shell Plate B-2002-1 144 0F 114 Intermediate Shell Plate B-2002-2 115. I°F 118.2 Intermediate Shell Plate B-2002-3 176 0F 181.9 Lower Shell Plate B-2003-1 152 0F --

Lower Shell Plate B-2003-2 128.8 0F - -

Intermediate & Lower Shell 230.2 0F 254.7 Longitudinal Weld Seams (Heat # W5214)

Intermediate to Lower Shell 220.90F Girth Weld Seam (Heat # 34B009)

Indian Point Unit 2 Surveillance 214.30F Weld (Heat # W5214)

Indian Point Unit 3 Surveillance 206.2 0F --

Weld (Heat # W5214)

H.B. Robinson Unit 2 Surveillance 210.7 0F Weld (Heat # W5214)

WCAP-15629

9 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1 c, for the metal temperature at that time. Ki, is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section XI"13' &71 of the ASME Appendix G to Section XI. The Ki. curve is given by the following equation:

0 Ki,= 33.2 + 20.734 *e [° 2

(T-RTr)] (1)

where, Ki = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTND This KIv curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* K* + K1 t < K1. (2)

where, K = stress intensity factor caused by membrane (pressure) stress Kt = stress intensity factor caused by the thermal gradients K1 c function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-15629

10 For membrane tension, the corresponding KI for the postulated defect is:

Kim = M. x (p!l / t) (3) where, Mm for an inside surface flaw is given by:

Mm = 1.85 for t < 2, Mm = 0.926 r for 2J I< :!ý3.464, Mm = 3.21 for [tJ > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm = 1.77 for f < 2, Mm = 0.893 -ýt for 2_<7 ft < 3.464, Mm = 3.09 for ft- > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding K, for the postulated defect is:

Krm = Mb

  • Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kit = 0.953x10 3 x CR x t2-, where CR is the cooldown rate in OF/hr., or for a postulated outside surface defect, K1 t = 0.753x10 3 x HU x t2-, where HU is the heatup rate inmF/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1 .

(a) The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b) Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

Kt = (1.0359Co + 0.6322C+/- + 0.4753C 2 + 0.3855C 3)

  • r (4)

WCAP-15629

11 or similarly, Krr during heatup for a 1/4-thickness outside surface defect using the relationship:

Kit = (1.043Co + 0.630C, + 0.481C2 +0.401C3) *V= (5) where the coefficients Co, CI, C2 and C 3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

3 (6) ex(x) = Co + Ci(x / a) + C2(x / a) 2 + C3(xl a) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldwon Limit Curves"'t Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1 , is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Ki, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Kic exceeds KI, the calculated allowable pressure during cooldown will be greater than the steady-state value.

WCAP-15629

12 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Ki. for the 1/4T crack during heatup is lower than the K10 for the 1/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K10 values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix Gd133 addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Indian Point Unit 2. The limiting unirradiated RTNDT of 60°F occurs in both the closure head and vessel flanges of the Indian Point Unit 2 reactor vessel, so the minimum allowable temperature of this region is 180'F at pressures greater than 621 psig. This limit is shown in Figures 5-1 and 5-2 wherever applicable.

WCAP-1 5629

13 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + ARTNDT + Margin (7)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of 14 Section III of the ASME Boiler and Pressure Vessel Codei ]. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTNT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTNDT = CF

  • f(O.

28

- 0.10 log ) (8)

To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

fdpth) = fsrac* e (-0.24x) (9) where x inches (vessel beltline thickness is 8.625 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections in Appendix B and are also presented in a condensed version in Table 6 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP 14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"' 21 . Table 6 contains the calculated vessel surface fluences values at various azimuthal locations. Tables 7 and 8 contain the 1/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the Indian Point Unit 2 reactor vessel.

WCAP-15629

14 TABLE 6 Calculated Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (1019 n/cm2 , E > 1.0 MeV)

Azimuthal Location EFPY 00 150 300 450 8.62(a) 0.145 0.231 0.275 0.416 16.87(b) 0.256 0.415 0.498 0.744 25 0.350 0.553 0.677 1.016 32 0.446 0.690 0.855 1.283 48 0.666 1.004 1.263 1.894 Notes:

(a) Date of last capsule removal.

(b) Current EFPY.

TABLE 7 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 25 EFPY Heatup/Cooldown Curves Material Surface 1/4 T(a) 3/T12)

Intermediate Shell Plate B-2002-1 1.02 x 1019 6.08 x 1018 2.16 x 1018 Intermediate Shell Plate B-2002-2 1.02 x 1019 6.08 x 1018 2.16 x 10"8 Intermediate Shell Plate B-2002-3 1.02 x 1019 6.08 x 10I 2.16 x 10"8 Lower Shell Plate B-2003-1 1.02 x 10i 9 6.08 x 1018 2.16 x 1018 Lower Shell Plate B-2003-2 1.02 x 1019 6.08 x 1018 2.16 x 1018 Intermediate & Lower Shell Longitudinal 6.77 x 10"8 4.03 x 1018 1.43 x 1018 Welds (Heat # W5214) - 0° 150 & 300 Intermediate to Lower Shell Girth Weld 1.02 x 10'9 6.08 x 10"8 2.16 x 1018 (Heat # 34B009)

Note:

(a) 1/4T and 3/4T = F(s*,ý) *e(-' 24"x), where x is the depth into the vessel wall (i.e. 8.625*0.25 or 0.75)

WCAP-15629

15 TABLE 8 Summary of the Calculated Fluence Factors used for the Generation of the 25 EFPY Heatup and Cooldown Curves Material 1/4T F 1/4T FF 3/4T f 3/4T FF 2

(n/cm , > 1.0 MeV) (n/cm ,E > 1.0 MeV)

Intermediate Shell Plate B-2002-1 6.08 x 1018 0.861 2.16 x 1018 0.588 Intermediate Shell Plate B-2002-2 6.08 x 1018 0.861 2.16 x 1018 0.588 Intermediate Shell Plate B-2002-3 6.08 x 1018 0.861 2.16 x 1018 0.588 Lower Shell Plate B-2003-1 6.08 x 1018 0.861 2.16 x 1018 0.588 Lower Shell Plate B-2003-2 6.08 x 1018 0.861 2.16 x 1018 0.588 Intermediate & Lower Shell 4.03 x 1018 0.748 1.43 x 1018 0.492 Longitudinal Welds (Heat # W5214) - 0°, 15- & 300 Intermediate to Lower Shell Girth 6.08 x 1018 0.861 2.16 x 1018 0.588 Weld (Heat # 34B009) I Margin is calculated as, M = 2 _ i + ~2 The standard deviation for the initial RTNDT margin term, is ayi 0

00F when the initial RTNDT is a measured value, and 17 F when a generic value is available. The standard 0

deviation for the ARTNDT margin term, 5 A,is 17 F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, aA is equal to 28OF when surveillance capsule data is not used, and is 140F (half the value) when credible surveillance capsule data is used. aA need not exceed 0.5 times the mean value of ARTNDT.

WCAP-15629

16 Contained in Tables 9 and 10 are the calculations of the 25 EFPY ART values used for generation of the heatup and cooldown curves.

TABLE 9 Calculation of the ART Values for the I/4T Location @ 25 EFPY Material RG 1.99 CF FF IRTNDT(a) ARTNDTt') Margin(c) ART(d)

R2 Method (OF) (OF) (OF) (OF) (OF)

Position 1.1 144 0.861 34 124.0 34 192 Intermediate Shell Plate B-2002-1 Position 2.1 114.0 0.861 34 98.2 17(e) 149 Position 1.1 115.1 0.861 21 99.1 34 154 Intermediate Shell Plate B-2002-2 Position 2.1 118.2 0.861 21 101.8 34(e) 157 Position 1.1 176 0.861 21 151.5 34 207 Intermediate Shell Plate B-2002-3 Position 2.1 181.9 0.861 21 156.6 17(e) 195 Position 1.1 152 0.861 20 130.9 34 185 Lower Shell Plate B-2003-1 Position 1.1 128.8 0.861 -20 110.9 34 125 Lower Shell Plate B-2003-2 Position 1.1 230.2 0.748 -56 172.2 65.5 182 Intermediate & Lower Shell Long. Welds (Heat # W5214)(c) Position 2.1 254.7 0.748 -56 191.0 44.0(e) 179 Intermediate to Lower Shell Position 1.1 220.9 0.861 -56 190.2 65.5 200 Girth Weld (Heat # 34B009)

Notes:

(a) Initial RTNT values are measured values except for the welds.

(b) ARTNrT = CF

  • FF 2 2 (c) M = 2 *(ci2 + a, )&

(d) ART = Initial RTNDT + ARTNDT + Margin (OF)

(e) All surveillance data is credible except for the lower shell plate B-2002-2. For this case a full aA was used.

WCAP-15629

17 TABLE 10 Calculation of the ART Values for the 3/4T Location @ 25 EFPY FF IRTNDT*a) ARTNDTPb) Margin(c) ART(d)

Material RG 1.99 CF R2 Method (OF) (OF) (OF) (OF) (OF)

Position 1.1 144 0.588 34 84.7 34 153 Intermediate Shell Plate B-2002-1 Position 2.1 114.0 0.588 34 67.0 17(e) 118 Position 1.1 115.1 0.588 21 67.7 34 123 Intermediate Shell Plate B-2002-2 Position 2.1 118.2 0.588 21 69.5 34(e) 125 Position 1.1 176 0.588 21 103.5 34 159 Intermediate Shell Plate B-2002-3 Position 2.1 181.9 0.588 21 107.0 17(e) 145 Position 1.1 152 0.588 20 89.4 34 143 Lower Shell Plate B-2003-1 Position 1.1 128.8 0.588 -20 75.7 34 89 Lower Shell Plate B-2003-2 Position 1.1 230.2 0.492 -56 113.3 65.5 123 Intermediate & Lower Shell Long. Welds (Heat # W5214)(c) Position 2.1 254.7 0.492 -56 125.3 44.0&-) 113 Intermediate to Lower Shell Position 1.1 220.9 0.588 -56 130.0 65.5 140 Girth Weld (Heat # 34B009)

Notes:

(a) Initial RTiNDT values are measured values except for the welds..

(b) ARTNDT = CF

  • FF2 1 (c) M = 2 *((y 12 + aA ) /2 (d) ART = Initial RTNDT + ARTNDT + Margin (OF)

(e) All surveillance data is credible except for the lower shell plate B-2002-2. For this case a full aA was used.

WCAP-15629

18 The intermediate to lower shell girth weld is the limiting beltline material for the 1/4T location (See Table

9) and the intermediate shell plate B-2002-3 is the limiting beltline material for the 3/4T location (See Table 10). Contained in Table 11 is a summary of the limiting ARTs to be used in the generation of the Indian Point Unit 2 reactor vessel heatup and cooldown curves. Since there are different limiting materials and one of which is a circumferential weld, then two sets of curves will be generated. One set will use the methodology from ASME Code Case N-588 with the limiting circ weld ARTs, while the other will use the methodology from the 1996 ASME Code Section XI, Appendix G with the limiting plate ARTs. The most limiting curves will be presented in Section 5, while the other set will be documented in Appendix G.

TABLE 11 Summary of the Limiting ART Values Used in the Generation of the Indian Point Unit 2 Heatup/Cooldown Curves 1/4T Limiting ART %T Limiting ART Circ Weld ART 200 140 Intermediate Shell Plate B-2002-3 195 145 WCAP-15629

19 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3.0 and 4.0 of this report. This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section 1 of this report.

Figure 1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for the first 25 EFPY. This curve was generated using the1996 ASME Code Section XI, Appendix G with the limiting plate ARTs. It bounds the heatup curves (found in Appendix G) generated using ASME Code Case N-588 with the limiting circ weld ARTs. Figure 2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100'F/hr applicable for 25 EFPY. Again, this curve was generated using the 1996 ASME Code Section XI, Appendix G with the limiting plate ARTs. It bounds the cooldown curves (found in Appendix G) generated using ASME Code Case N-588 with the limiting circ weld ARTs.

Allowable combination of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 and 2. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 1. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR 3

Part 50. The governing equation for the hydrostatic test is defined in Code Case N-6401 1(approved in February 1999) as follows:

1.5 K* < K1c

where, Kim is the stress intensity factor covered by membrane (pressure) stress, Kiv = 33.2 + 20.734 e[oO 2 CTRTr)],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 13. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Indian Point Unit 2 reactor vessel at 25 EFPY is 255 OF. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure temperature limit curve, constitutes the limit for core operation for the reactor vessel.

WCAP-15629

20 Figures 1 and 2 define all of the above limits for ensuring prevention of nonductile failure for the Indian Point Unit 2 reactor vessel. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 and 2 are presented in Tables 12 and 13. By comparison to the curves and data points in Appendix (Qit can be seen that the curves in Figures 1 and 2 bound the curves using code case N 588 with a slightly higher 1/4T ART.

WCAP-15629

21 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE LIMITING ART VALUES AT 25 EFPY: 1/4T, 195-F 3/4T, 145-F 2500 2250 2000 1750 I 1500 I 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure I Indian Point Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60 &

100'F/hr) Applicable for the First 25 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology WCAP-15629

22 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE LIMITING ART VALUES AT 25 EFPY: 1/4T, 195-F 3/4T, 145°F 2500 2250 2000 1750 I 1500 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2 Indian Point Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 25 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology WCAP-15629

23 TABLE 12 25 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Heatup Curves 50 Heatup 60 Limit 100 Heatup 100 Limit Leak Test Limit Critical Critical T P T P T P T P T P 0 255 0 60 0 255 0 238 2000 23 60 255 621 60 581 255 581 255 2485 60 621 621 255 621 65 581 255 581 65 621 255 621 70 581 255 582 70 621 255 621 75 581 255 583 75 621 255 621 80 581 255 585 80 621 255 621 85 581 255 587 85 621 255 621 90 581 255 590 90 621 255 621 95 581 255 592 95 621 255 621 100 581 255 596 100 621 255 621 105 581 255 599 105 621 255 621 110 581 255 603 110 621 255 621 115 581 255 608 115 120 621 255 621 120 582 255 613 621 255 621 125 585 255 620 125 621 255 621 130 590 255 621 130 135 621 255 621 135 596 255 621 621 255 621 140 603 255 621 140 621 255 621 145 613 255 621 145 150 621 255 621 150 621 255 621 155 621 255 621 155 621 255 621 160 621 255 621 160 621 255 621 165 621 255 621 165 621 255 621 170 621 255 621 170 621 255 621 175 621 255 621 175 621 255 621 180 621 255 888 180 621 255 734 180 621 255 917 180 621 255 762 180 888 255 950 180 734 255 792 185 917 255 986 185 762 255 826 190 950 255 1026 190 792 255 864 195 986 255 1070 195 826 255 906 200 1026 255 1119 200 864 255 952 205 1070 255 1173 205 906 255 1003 WCAP-15629

24 TABLE 12 - (Continued) 25 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Heatup Curves 60 Heatup 60 Limit 100 Heatup 100 Critical Limit Critical T P T P T P T P 210 1119 260 1232 210 952 260 1060 215 1173 265 1291 215 1003 265 1123 220 1232 270 1345 220 1060 270 1192 225 1291 275 1405 225 1123 275 1269 230 1345 280 1470 230 1192 280 1353 235 1405 285 1543 235 1269 285 1447 240 1470 290 1622 240 1353 290 1550 245 1543 295 1711 245 1447 295 1657 250 1622 300 1808 250 1550 300 1737 255 1711 305 1915 255 1657 305 1826 260 1808 310 2033 260 1737 310 1923 265 1915 315 2163 265 1826 315 2030 270 2033 320 2307 270 1923 320 2148 275 2163 325 2466 275 2030 325 2278 280 2307 280 2148 330 2422 285 2466 285 2278 290 2422 WCAP-15629

25 TABLE 13 25 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

UU,)IUU WiL ,*mt V*

Steady State 20F 40F 60F 10OF T T P T P T p PT1 60 0 60 0 60 0 60 0 60 0 532 60 480 60 373 60 621 60 583 60 535 65 483 65 376 65 621 65 586 65 538 70 486 70 380 70 621 70 589 70 542 75 490 75 384 75 621 75 592 75 546 80 494 80 389 80 621 80 596 80 550 85 499 85 394 85 621 85 600 85 555 90 504 90 400 90 621 90 605 90 560 95 510 95 407 95 621 95 610 95 566 100 517 100 415 100 621 100 615 100 573 105 524 105 424 105 621 105 621 105 581 110 532 110 434 110 621 110 621 110 589 115 541 115 445 115 621 115 621 115 120 599 120 552 120 457 120 621 120 621 125 609 125 563 125 471 125 621 125 621 130 621 130 576 130 486 130 621 130 621 621 135 590 135 504 135 621 135 621 135 621 140 606 140 523 140 621 140 621 140 621 145 621 145 544 145 621 145 621 145 150 621 150 621 150 568 150 621 150 621 155 621 155 621 155 595 155 621 155 621 621 160 621 160 621 160 621 160 621 160 165 621 165 621 165 621 165 621 165 621 170 621 170 621 170 621 170 621 170 621 175 621 175 621 175 621 175 621 175 621 180 621 180 621 180 621 180 621 180 621 180 621 180 621 180 621 180 621 180 621 180 835 180 812 180 779 180 888 180 860 893 185 871 185 852 185 828 185 917 185 929 190 911 190 897 190 884 190 950 190 969 195 955 195 946 195 945 195 986 195 1026 200 1013 200 1004 200 1000 200 205 1070 205 1062 205 1058 210 1119 210 1115

_____________ L ______________ ______________

WCAP-15629

26 TABLE 13 - (Continued) 25 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Cooldown Curves Steady State 20F 40F 60F 10OF T P T P T P T P T P 215 1173 220 1232 225 1298 230 1370 235 1451 240 1540 245 1638 250 1746 255 1866 260 1998 265 2144 270 2306 275 2485 WCAP-15629

27 6 REFERENCES

1. Southwest Research Final Report, SwRI Project 17-2108, "Reactor Vessel Material Surveillance Program for Indian Point Unit 2: Analysis of Capsule V", March 1990.
2. WCAP-12796, "Heatup and Cooldown Limit Curves for the Consolidated Edison Company Indian Point Unit 2 Reactor Vessel", N. K. Ray, January 1991.
3. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.
4. ASME Boiler and Pressure Vessel Code, Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels",Section XI, Division 1, Approved December 12, 1997.
5. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

6. WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.
7.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix (Q"Fracture Toughness Criteria for Protection Against Failure." Dated December 1995, through 1996 Addendum.
8. "Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
9. INT-00-2 11, "Evaluation of Reactor Vessel Flux and Fluence Calculations", R.R. Laubham, April 25, 2000.
10. WCAP-14044, "Westinghouse surveillance Capsule Neutron Fluence Re-evaluation", E.P Lippencott, April 1994.
11. WCAP- 11815, "Analysis of Capsule Z from the New York Power Authority Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program", S. E. Yanichko, et. al., March, 1988.
12. CPL-96-203, "Robinson Unit 2 Surveillance Capsule Charpy Test Results", P. A. Grendys, March 6, 1996.
13. Code of Federal Regulations, 10 CFR Part 50, Appendix (4 "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

14. 1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."
15. CE Report NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds", CEOG Task 902, By the CE Owners Group. June 1997.

WCAP-15629

28

16. CE Report NPSD-1 119, Revision 1, "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content", CEOG Task 1054, By the CE Owners Group. July 1998.
17. WOG CalcNote 92-016 (Westinghouse File # WOG-108/4-18), 'VOG USE Program - Onset of Upper Shelf Energy Calculations", J.M. Chicots, 3/8/93. [Note: This calcnote used the original Combustion Engineering CMTRs]

WCAP-15629

A-O APPENDIX A PRESSURIZED THERMAL SHOCK (PTS) RESULTS WCAP-15629

A-1 PTS Calculations:

The PTS Rule requires that for each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTPTS , accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material. This assessment must specify the basis for the projected value of RTpTs for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation. This assessment must be updated whenever there is a significant change in projected values of RTPTS, or upon request for a change in the expiration date for operation of the facility. (Changes to RTpTs values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.

To verify that RTrDT , for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. (Surveillance program results mean any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part 50, Appendix H.)

Calculations:

Table A-1 contains the results of the calculations for each of the beltline region materials in the Indian Point Unit 2 Reactor Vessel. Per ConEd, the actual EOL is less than 32 EFPY, however for conservatism EOL will be assumed to be 32 EFPY WCAP-15629

A-2 TABLE A-1 RTpTs Calculations for Indian Point Unit 2 Beltline Region Materials at 32 EFPY FF CF ARTpns(c) Margin RTNDT(U() RTpTs(b)

Material Fluence (nlcm2, E>1 .0 (OF) (OF) (OF) (OF) (OF)

MeV)

Inter. Shell Plate B-2002-1 1.28 x 10'9 1.07 144 154.1 34 34 222

- Using S/C Data 1.28 x 10'9 1.07 114 122.0 17 34 173 Inter. Shell...Plate ..B-2002-2 1.28 x 101" 1.07 115.1 123.2 34 21 178

- Using S/C Data 1.28 x 1019 1.07 118.2 126.5 34 21 182 Inter. Shell Plate B-2002-3 1.28 x 1019 1.07 176 188.3 34 21 243

- Using S/C Data 1.28 x 10'9 1.07 181.9 194.6 17 21 233 Lower Shell Plate B-2003-1 1.28 x 1019 1.07 152 162.6 34 20 217 Lower Shell Plate B-2003-2 1.28 x 10'9 1.07 142 151.9 34 -20 166 Intermediate & Lower Shell 8.55 x 1018 0.956 230.2 220.1 65.5 -56 230 Long. Welds (Heat # W5214) - - ------------------------- ---------

- Using S/C Data 8.55 x 1018 0.956 254.7 243.5 44.0 -56 232 Intermediate to Lower Shell Girth 1.28 x 1019 1.07 220.9 236.4 65.5 -56 246 Weld (Heat # 34B009) I I I Notes:

(a) Initial RTND values are measured values (b) RTprs = RTNDrM + ARTpTs + Margin (OF)

(c) ARTpTs = CF

  • FF All of the beltline materials in the Indian Point Unit 2 reactor vessel are below the screening criteria values of 270'F and 300'F at 32 EFPY.

WCAP-15629

B-O APPENDIX B CALCULATED FLUENCE DATA WCAP-15629

B-1 Neutron Fluence Calculations Discrete ordinates transport calculations were performed on a fuel cycle specific basis to determine the neutron environment within the reactor geometry of Indian Point Unit 2. The specific calculational methods applied are consistent with those described in WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology"'1 ] and in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

January 1996.[2]

In the application of this methodology to the fast neutron exposure evaluations for the Indian Point Unit 2 surveillance capsules and reactor vessel, plant specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

  • (r,0,z) = [4(r,0)] * [ý(rz)]/[ý(r)]

where cb(rO,z) is the synthesized three-dimensional neutron flux distribution, 4(rO) is the transport solution in r,9 geometry, ý(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ý(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the rO two-dimensional calculation.

For this analysis, all of the transport calculations were carried out using the DORT discrete ordinates code Version 3. P] and the BUGLE-96 cross-section library[41 . The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor application.

In these analyses, anisotropic scattering was treated with a P 5 legendre expansion and the angular discretization was modeled with an S16 order of angular quadrature. Energy and space dependent core power distributions as well as system operating temperatures were treated on a fuel cycle specific basis.

Results of the discrete ordinates calculations performed for Indian Point Unit 2 are provided in Tables 1 through 3. In Table 1, the calculated neutron exposures for the four surveillance capsules withdrawn to date are given in terms of both fast neutron (E > 1.0 MeV) fluence and iron atom displacements (dpa). The maximum neutron exposure of the pressure vessel at the clad/base metal interface is provided for several azimuthal angles in Table 2. Again, calculated exposure data are listed for both fluence (E > 1.0 MeV) and dpa. Calculated lead factors associated with each of the Indian Point Unit 2 surveillance capsules are listed in Table 3.

Following the completion of the plant specific transport analyses, the calculated results were compared with available measurements in order to demonstrate that the differences between calculations and measurements support the 20% (lo) uncertainty required by Draft Regulatory Guide DG-1053, 5

"Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence".J ] Two levels of comparison of calculation with measurement were made to demonstrate compliance with the requirements of DG-1053. In the first instance, ratios of measured and calculated sensor reaction rates (M/C) were compared for all fast neutron sensors contained in the surveillance capsules withdrawn to date. In the second case, comparisons of calculated and least squares adjusted best estimate values of neutron fluence (E > 1.0 MeV) and dpa were examined.

WCAP-15629

B-2 all The M/C comparisons of individual sensor reaction rates showed consistent behavior for all reactions at results.

capsule locations within the constraint of an allowable 20% (1a) uncertainty in the final calculated The overall average M/C ratio for the entire 13 sample data set was 1.07 with an associated standard deviation of 9.2%. The observed M/C ratios for twelve of the 13 samples ranged from 0.87 to 1.16 with the remaining sample [ 63Cu(n,a)6°Co reaction] exhibiting an M/C ratio of 1.22. This data set of M/C ratios from the Indian Point Unit 2 surveillance capsules indicates that the +/- 20% acceptance criterion specified in DG-105331' has been met by the current neutron transport calculations.

The corresponding best estimate to calculation (BE/C) comparisons for neutron fluence (E > 1.0 MeV) spanned a range of 0.948 to 1.056 with an average BE/C ratio of 1.017 +/- 1.4% (la). Likewise, in the case of iron atom displacements, the BE/C ratios spanned a range of 0.947 to 1.043 with an average BE/C of 1.008 +/- 4.2% (la). These comparisons also fall well within the +/- 20% criterion specified in DG-1053, thus supporting the validation of the current calculations for applicability for the Indian Point Unit 2 reactor.

Appendix B

References:

1. S. L. Anderson, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology," WCAP-15557-RO, August 2000.
2. J. D. Andrachek, et al., "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," WCAP-14040-NP-A, Revision 2, January 1996.
3. RSICC Computer Code Collection CCC-650, "DOORS3.1 One-, Two-, and Three- Dimensional Discrete Ordinates Neutron/Photon Transport Code System," Radiation Shielding Information Center, Oak Ridge National Laboratory, August 1996.
4. RSIC Data Library Collection DLC-185, "BUGLE-96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Radiation Shielding Information Center, Oak Ridge National Laboratory, March 1996.
5. Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, September 1999.

WCAP-15629

B-3 Table B-1 Summary of Calculated Surveillance Capsule Exposure Evaluations Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule IefpyI In/cm 2] [dpa]

T 1.42 2.53e+18 4.26e-03 y 2.34 4.55e+18 7.68e-03 Z 5.17 1.02e+19 1.72e-02 V 8.62 4.92e+18 7.91 e-03 WCAP-15629

B-4 Table 2 Summary of Calculated Maximum Pressure Vessel Exposure Clad/Base Metal Interface 2

Irradiation Neutron Fluence (E > 1.0 MeV) [n/cm ]

Time

[efpy] 0.0 Degrees 15.0 Degrees 30.0 degrees 45.0 Degrees 16.87 (EOC 14) 2.556e+18 4.152e+18 4.975e+18 7.443e+18 18.66 (EOC 15) 2.764e+18 4.453e+18 5.368e+18 8.038e+18 25.00 3.505e+18 5.526e+18 6.766e+18 1.016e+19 32.00 4.464e+ 18 6.900e+18 8.551e+18 1.283e+19 48.00 6.657e+18 1.004e+19 1.263e+19 1.894e+19 Irradiation Iron Atom Displacements [dpa]

Time

[efpy] 0.0 Degrees 15.0 Degrees 30.0 degrees 45.0 Degrees 16.87 (EOC 14) 4.140e-03 6.635e-03 8.01le-03 1.200e-02 18.66 (EOC 15) 4.476e-03 7.117e-03 8.643e-03 1.295e-02 25.00 5.884e-03 9.115e-03 1.125e-02 1.687e-02 32.00 7.438e-03 1. 132e-02 1.413e-02 2.118e-02 48.00 1.099e-02 1.636e-02 2.070e-02 3.105e-02 WCAP-15629

B-5 Table 3 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor T(40 0) Withdrawn EOC 1 3.43 Y(40-) Withdrawn EOC 2 3.48 Z(40°) Withdrawn EOC 5 3.53 V(40) Withdrawn EOC 8 1.18 S(40°) In Reactor 3.5 U(40 ) In Reactor 1.2 W(40 ) In Reactor 1.2 X(40 ) In Reactor 1.2 WCAP-15629

C-0 APPENDIX C UPDATED SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WCAP-15629

C-1 TABLE C- I Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured

2) (CF)(a) (OF) (%) (2) (%)(c)

(x 1019 n/cm (b)

T 2.53 x 1018 90.29 55.0 21 16 Intermediate Shell Z 1.02 x 1019 144.86 125.0 29 21 Plate B-2002-1 T 2.53 x 1018 72.17 95.0 19 17 Intermediate Shell Z 1.02 x 10'9 115.79 120.0 26 23 Plate B-2002-2 V 4.92 x 10" 92.31 77.0 22 4 T 2.53 x 1018 110.35 115.0 25 20 Intermediate Shell Y 4.55 x 10"8 137.46 145.0 28 28 Plate B-2002-3 Z 1.02 x 1019 177.06 180.0 34 28 Y 4.55 x 10is 167.37 195.0 28 45 Surv. Program Weld Metal V 4.92 x 1018 171.87 204.0 29 38 Y 4.55 x 101" -- 165 --- 13 Heat Affected Zone Material V 4.92 x 1018 150 --- 0 Correlation Monitor T 2.53 x 1018 75 --- 0 Material Y 4.55 x 10's 70 -- - 6 Z 1.02 x 101 9 -- 102 --- 15 V 4.92 x 101 8 -- 100 --- 0 Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b) Calculated using measured Charpy data.

(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.

WCAP-15629

D-0 APPENDIX D REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES WCAP-15629

D-1 TABLE D-1 Predicted End-of-License (32 EFPY) USE Calculations for all the Beltline Region Materials Material Weight % 1/4T EOL Unirradiated Projected Projected of Cu Fluence USE(a) USE EOL USE (1019 n/cm2) (ft-lb) Decrease (%) (ft-lb)

Intermediate Shell Plate B-2002-1 0.19 0.763 70 20 56 0.17 0.763 73 21 58 Intermediate Shell Plate B-2002-2 0.25 0.763 74 32 50.3 Intermediate Shell Plate B-2002-3 0.20 0.763 71 27 52 Lower Shell Plate B-2003-1 0.19 0.763 88 27 64 Lower Shell Plate B-2003-2 0.21 0.510 121 43 69 Intermediate & Lower Shell Longitudinal Welds (Heat # W5214)

Intermediate to Lower Shell Girth Weld 0.19 0.763 82(b) 32 56 (Heat # 34B009)

Notes:

(a) These values were obtained from Reference 17. Values reported in the NRC Database RVID2 are identical with exception to Intermediate Shell Plates B-2002-1, 2. RVIID2 reported the initial USE as 76 and 75. This evaluation conservatively used the lower values of 70 and 73.

(b) Value was obtained from the average of three impacts tests (71, 84, 90) at 10'F performed for the original material certification.

WCAP-15629

E-O APPENDIX E UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE WCAP-15629

E-1 Withdrawal Schedule To Be Provided in PTLR Only by Indian Point Unit 2 WCAP-15629

F-0 APPENDIX F ENABLE TEMPERATURE CALCULATIONS AND RESULTS WCAP-15629

F-1 Enable Temperature Calculation:

system to be ASME Section XI, Appendix G requires the low temperature overpressure (LTOP or COMS) in operation at coolant temperatures less than 200'F or at coolant temperatures less than a temperature RTNDT is corresponding to a reactor vessel metal temperature less than RTNDT + 5 0°F, whichever is greater.

one fourth the highest adjusted reference temperature (ART) for the limiting beltline material at a distance determined of the vessel section thickness from the vessel inside surface (ie. clad/base metal interface), as by Regulatory Guide 1.99, Revision 2.

32 EFPY is The highest calculated 1/4T ART for the Indian Point Unit 2 reactor vessel beltline region at 25 EFPY 0 F.

200 From the OPERLIM computer code output for the Indian Point Unit 2 25 EFPY P-T limit curves without margins (Configuration # 14146 & 22915) the maximum ATme* is:

Cooldown Rate (Steady-State Cooldown):

max (ATmeti) at 1/4T = 0°F Heatup Rate of 100°F/Hr:

max (ATmw) at 1/4T = 30.084°F Enable Temperature (ENBT) = RTNDT + 50 + max (ATmebt), OF

= (200 + 50 + 30.084) OF

= 280.0840 F The minimum required enable temperature for the Indian Point Unit 2 Reactor Vessel is 280OF at 25 EFPY of operation.

WCAP-15629

G-O APPENDIX G PRESSURE TEMPERATURE LIMIT CURVES USING CODE CASE N-588 WCAP-15629

G-1 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE TO LOWER SHELL GIRTH WELD LIMITING ART VALUES AT 25 EFPY: 1/4T, 200-F 3/4T, 140-F 2500 2250 2000 1750 ci,,

0~

1500 U) 1250 I

a 0

1000 0

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure G-1 Indian Point Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 601F/hr) Applicable for the First 25 EFPY (Without Margins for Instrumentation Errors) Using Code Case N-588 WCAP- 15629

G-2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE TO LOWER SHELL GIRTH WELD LIMITING ART VALUES AT 25 EFPY: 1/4T, 200°F 3/4T, 140°F 2500 2250 2000 1750 I

1500 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure G-2 Indian Point Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 0F/hr) Applicable for the First 25 EFPY (Without Margins for Instrumentation Errors) Using Code Case N-588 WCAP-15629

G-3 TABLE G- 1 25 EFPY Heatup Curve Data Points Using Code Case N-5 88 (without Uncertainties for Instrumentation Errors)

Heatup Curves 60 Heatup 60 Limit 100 Heatup 100 Limit Leak Test Limit Critical Critical T P T P T P T P T P 60 0 186 0 60 0 186 0 138 2000 621 186 620 60 621 186 620 186 2485 60 65 621 186 620 65 621 186 620 70 621 186 620 70 621 186 620 75 621 186 620 75 621 186 620 80 621 186 620 80 621 186 620 85 621 186 620 85 621 186 620 621 186 620 90 621 186 620 90 621 186 620 95 621 186 620 95 621 186 620 100 621 186 620 100 621 186 620 105 621 186 620 105 621 186 620 110 621 186 620 110 621 186 620 115 621 186 620 115 120 621 186 620 120 621 186 620 621 186 620 125 621 186 620 125 621 186 620 130 621 186 620 130 621 186 620 135 621 186 620 135 621 186 620 140 621 186 620 140 621 190 620 145 621 190 620 145 150 621 195 620 150 621 195 620 155 621 200 620 155 621 200 620 160 621 205 620 160 621 205 620 165 621 210 620 165 621 210 620 621 215 620 170 621 215 620 170 175 621 220 620 175 621 220 620 180 621 220 1800 180 621 220 1545 180 621 225 1856 180 621 225 1606 180 1800 230 1918 180 1545 230 1675 185 1856 235 1986 185 1606 235 1751 190 1918 240 2061 190 1675 240 1835 195 1986 245 2145 195 1751 245 1929 200 2061 250 2237 200 1835 250 2032 205 2145 255 2339 205 1929 255 2147 210 2237 260 2451 210 2032 260 2274 215 2339 215 2147 265 2414 220 2451 220 2274 225 2414

.1 __________________ - -

WCAP-15629 WCAP-15629

G-4 TABLE G-2 25 EFPY Cooldown Curve Data Points Using Code Case N-588 (without Uncertainties for Instrumentation Errors)

WCAP-15629

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