ML18040A434

Submittal of Revised Limits Report (PTLR) Pressure and Temperature

ML18040A434
Person / Time
Site:	Watts Bar
Issue date:	02/09/2018
From:	Simmons P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18040A434 (17)

Text

Tennessee Valley Authority, Post Office Box 2000 Spring City, Tennessee 37381 February 9, 2018 10 cFR 50.36 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket Nos. 50-391

Subject:

Watts Bar Nuclear Plant Unat 2 - Revised Pressure and Temperature Limits Report (PTLR)

The purpose of this letter is to provide the enclosed copy of the Watts Bar Unit 2 Pressure and Temperature Limits Report (PTLR) Revision 5, in accordance with Technical specification section 5.9.6.c. Revision 5 of the PTLR updates the Surveillance Capsule Withdrawal Schedule for capsule "U" from the first refueling outage to the end of cycle 2 based on the revised lead factor and expected fluence values for this capsule.

There are no new regulatory commitments in this letter. Should you have questions regarding this submittal, please contact Kim Hulvey, Manager of watts Bar Site Licensing, at (423) 365-77 20.

Respectfully, Paul Simmons Site Vice President Watts Bar Nuclear Plant

U.S. Nuclear Regulatory Commission Page 2 February 9, 2018

Enclosure:

Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR),

Revision 5.

cc (Enclosure):

U. S. Nuclear Regulatory Commission Region ll Marquis One Tower 245 Peachtree CenterAve., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident lnspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

ENCLOSURE Watts Bar Nuclear Plant, Unat 2 Pressure and Temperature Limits Report, Revision 5

system I REACTOR COOLANT SYSTEM lSOo-ttg-084001 Description I Unitl/Unit2 lRev.0039 Document I OARecord lpaqe 251 ot2l0 Appendix B (Page 1 ot 141 Watts Bar Unit 2 - RCS Pressure and Temperature Limits Report (PTLR) - Revision 5 APPENDIX "B" rO REC SYSTEM DESCRIPTION N3.68-4001 WATTS BAR UNIT 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

REVISION 5 Prepared by: M. R. Smith Checked by: C. A. Boudreaux Approved by: D. S. Acselrod

System REACTOR COOLANT SYSTEM ISDD.N3.68.4OO1 Description I Unitl/Unit2 lRev.0039 Document I OA Record Paqe 252 of 27A Appendix B (Page 2 ot 141 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR for Watts Bar Unit 2 has been prepared in accordance with the requirements of Technical Specification 5.9.6. Revisions to the PTLR shall be provided to the NRC within 30 days of issuance.

The Technical Specifications affected by this report are listed below:

LCO 3.4.3, RCS Pressure and Temperature (Pff) Limits LCO 3.4.12, Cold Overpressure Mitigation System (COMS) 2.0 RCS PRESSURE AND TEMPERATURE LIMITS The limits for LCO 3.4.3 are presented in the subsection which follows. These limits have been developed (Ref. 1) using the NRC-approved methodologies specified in Technical Specification 5.9.6.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60oF.

2.1.2 The RCS temperature rate-of-change limits are:

A. A maximum heatup rate of 100'F per hour.

B. A maximum cooldown rate of 100"F per hour.

C. A maximum temperature change of ( 10"F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 RCS P/T Limits for Heatup, Cooldown, lnselvice Hydrostatic and Leak Testing, and Criticality The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2 (Ref. 1).

3.0 coLD oVERPRESSURE MITIGATION SYSTEM (LCO 3.4.12)

The lift setting limits for the pressurizer Power Operated Relief Valves (PORVS) are presented in the subsection that follows. These lift setting limits have been developed using the NRC-approved methodologies specified in Technical Specification 5.9.6.

3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setting limits are specified by Figure 3.1-1 and Table 3.1-1 (Ref.2).

System I REACTOR COOLANT SYSTEM ISOO-ttg-O84OOl Description I Unitl/Unit2 lRev.0039 Document I OARecord lPaqe 253ot2lD Appendix B (Page 3 of 141 3.1 Pressurizer PORV Lift Setting Limits (continued)

NOTE: These setpoints include allowance for pressure difference between the pressure transmitter and reactor midplane, and also includes a71.8 psig pressure channel uncertainty, and a 16.3"F temperature uncertainty.

3.2 Arming Temperature COMS shall be armed when any RCS cold leg temperature is <225F for Unit 2.

4.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The results of these examinations shallbe used to update Figures 2.1-1,2.1-2, and 3.1-1.

The pressure vessel steel surveillance program (Ref. 3) is in compliance with Appendix H to 10 CFR 50 (Ref. 4), entitled "Reactor Vessel Material Surveillance Program Requirements."

The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTruor, which is determined in accordance with ASTM E208 (Ref. 5). The empirical relationship between RTNor and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Fracture Toughness Criteria for Protection Against Failure," to Section X! of the ASME Boiler and Pressure Vessel Code (Ref. 6). The surveillance capsule removalschedule meets the requirements of ASTM E185-82 (Ref. 7).

The removalschedule is provided in Table 4.0-1.

5.0 SUPPLEMENTAL DATA TABLES

. Table 5-1 contains a Summary of the Best Estimate Cu and Ni Weight Percent and lnitial RTxslValues for the Watts Bar Unit 2 Reactor Vessel Materials.

o Table 5-2 shows a Summary of the lnitial RTr.ror Values for the Watts Bar Unit 2 Closure Head and Vessel Flange.

o Table 5-3 provides the Summary of the Watts Bar Unit 2 ReactorVessel Beltline Material Ghemistry Factors.

o Table 5-4 provides Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline Materials.

o Table 5-5 shows Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline materials through 7 EFPY at the 1/4T Location.

o Table 5-6 contains Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 3/4T Location.

System I REACTOR COOLANT SYSTEM ISDD-N3-68.4001 Description I Unitl/Unit2 lRev.0039 Document I GIA Record I Paqe 254 of 270 Appendix B (Page 4 ot 141 5.0 SUPPLEMENTAL DATA TABLES (continued) o Table 5-7 provides a Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves.

o Table 5-8 shows RTprs Calculations for the Watts Bar Unit 2 Beltline Materials at 32 EFPY.

6.0 REFERENCES

1. WCAP-I 7035-NP, Revision 2, "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," December 2009.

2. Westinghouse Letter WBT-D-S147, dated December 10, 2014,'PORV Analyses."

3. WCAP-9455, Revision 3, "Tennessee Valley Authority Watts Bar Unit No. 2 Reactor Vessel Radiation Surveillance Program," September 2009.

4. Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Gommission, Federal Register, Volume 60, No. 243, December 19, 1995.

5. ASTM E208, "Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials.

6. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, "Fracture Toughness Criteria for Protection Against Failure."

7. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," E706 (!F), ASTM 1982.

L WCAP-13830, Revision 1, "Heat Up and Cool Down Limit Curves for Normal Operation forWatts Bar Unit 2," J.M. Chicots, et al, February 1995.

9. Westinghouse letter WAT-D-12366, Rev. 0, "Transmittal of Justification for the Surveillance Gapsule Withdrawal Schedule Update for Watts Bar Unit 2" which references WCAP-18191-NP, Rev. 0, "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel lntegrity Evaluations" for TPBAR project.

System I REACTOR COOLANT REAGTOR SYSTEM ISDD.N3.684OO1 SYSTEM ISDD-N3-G Description I Unitl/Unit2 lRev.0039 Document I CtA Record I Paqe 255 of 270 Appendix B (Page 5 of 141 7.4 FIGURES AND TABLES MATE RIAL PROPERTY BASIS LIMITING MATERIAL: lntermediate Shell Forging 05 INITIAL RTr,ror: 14"F LIMITING ART VALUES AT 7 EFPY: 1t4T, 3t4T, 2500 2250.

2000 I 750 Io A

o-Y I 500 o

r

=

vt o

o E 125o o-t,o

+,

g lt- { ooo J

Gl C'

750 500 250 o

o 50 loo {50 2@ 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2.1-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 6O"F/hr and 1OO"F/hr) Applicable lor 7 EFPY (without Margins for lnstrumentation Errors) Using 1998 through 2OOO Addenda App.G Methodology (w/Kr")

(Plotted data (Ref. 1) provided in Table 2.1-'l)

system I REACTOR COOLANT SYSTEM ISOO+a-e84001 Description I Unitl/Unit2 lRev.0039 Document I aARecord lPaEe 2i6ot270 Appendix B (Page 6 of 141 7.0 FIGURES AND TABLES (continued)

TABLE2.1.1 Watts Bar Unit 2 Heatup Limits 7 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology Data (Ref. 1) plotted on Figure 2.1-1 LEAK HEATUP CRITICALITY HEATUP CRITICALITY TEST RATE LIMITS RATE LIMITS LIMITS (60'F/HR) (60.F/HR) (100.F/HR) (100.F/HR)

T P T P T P T P T P

("F) (psig) ('F) (psig) ('F) (psig) ('F) (psig) ('F) (psig) 105 2000 60 0 122 0 60 0 122 0 105 2000 60 621 122 621 60 621 122 621 122 2485 65 621 122 621 65 621 122 621 122 2485 70 621 122 621 70 621 122 621 75 621 122 621 75 621 122 621 80 621 125 621 80 621 125 621 85 621 130 621 85 621 130 621 90 621 135 621 90 621 135 621 95 621 140 621 95 621 140 621 100 621 140 1256 100 621 140 1128 100 621 145 1314 100 621 145 1 160 100 1256 150 1 381 100 1128 150 1 199 10s 1314 155 1 458 105 1 160 155 1245 110 1381 160 1544 110 1 199 160 1298 115 1 458 165 1 640 115 1245 165 1 358 120 1544 170 1748 120 1298 170 1426 125 1 640 175 1 868 125 1 358 175 1 503 130 1748 180 2001 130 1426 180 1 590 135 1 868 185 2149 135 1 503 185 1687 140 2001 190 2312 140 1 590 190 1795 145 2149 145 1687 195 1915 150 2312 150 1795 200 2048 155 191s 205 2196 160 2048 210 2360 165 2196 170 2360

System REACTOR COOLANT SYSTEM ISDD.N3.684OO1 Description I Unitl/Unit2 lRev.0039 Document I CtA Record IPaqe 257 of 270 Appendix B (Page 7 ot 141 7.4 FIGURES AND TABLES (continued)

MATERIAL P ROPERTY BASIS LlMlTl NG MATERIAL: lntermediate Shell Forging O5 lNlTlAL RTr'ror: 14"F LIMITING ART VALUES AT 7 EFPY: 114T, 61 "F 314T, 45"F 2000 1750 o

a'-r

-o-o I 500 Y

o

-o L

J Itt o

L 125o^

o-E' c,

rlrt E

ll- I t OOO atr C'

75q^

500 25,o^

o o 50 {oo {50 2o,g^ 25,o 300 350 400 45,q^

Moderator Temperature (Deg. F)

Figure 2.1-2

\Ahtts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1OO"F/hr) Applicable lor 7 EFPY (without Margins for lnslrumentation Errors) Using 1998 through 2OOO Addenda App. G Methodology (w/K")

(Plotted data (Ref. 1) provided in Table 2.1-2)

System I REACTOR COOLANT SYSTEM ISDD-N3-68400{

Description I Unitl/Unit2 lRev.0039 Document I CIA Record lPaqe 258 of 27O Appendix B (Page I of 141 7.0 FIGURES AND TABLES (continued)

TABLE2.1-2 Watts Bar Unit 2 Cooldown Limits 7 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (Data (Ref. 1) plotted on Figure 2.1-2)

Steady State 20.t:/HR 40'F r/HR 60't :/HR 1000F/HR T P T P T P T P T P

("F) (psig) ('F) (psig) ("F) (psig) ('F) (psis) ("F) (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 BO 621 BO 621 80 621 85 621 85 621 85 621 85 621 B5 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 1422 100 1422 100 1422 100 1422 100 1422 105 1 508 105 1 508 105 1 508 105 1 508 105 1 508 110 1 603 110 1 603 110 1 603 110 1 603 110 1 603 115 1709 115 1 709 115 1 709 115 1709 115 1 709 120 1825 120 1825 120 1825 fia 1825 120 1825 125 1954 125 1 954 125 1954 125 1 954 125 1 954 130 2096 130 2096 130 2096 130 2096 130 2096 135 2253 135 2253 135 2253 135 2253 135 2253 140 2427 140 2427 140 2427 140 2427 140 2427

System I REACTOR COOLANT REAGTOR SYSTEM SYSTEM ISDD.N3.68-4001 ISDD-N3-6 Description I Unitl/Unit2 lRev.0039 Document I CIA Record lPaqe 259 of 270 Appendix B (Page 9 of 141 7.0 FIGURES AND TABLES (continued)

Setpoint Window h0

,I rt E

v\

tr

'6 fE

(,

6 4

ro 2m 250 300 350 lndlcated RCs Tempereture, "F Figure 3.1-1 PORV Setpoint vs RCS Temperature (Plotted data (Ref. 2) provided in Table 3 .1-1)

system I REACTOR GOOLANT SYSTEM ISOO-ttS+84001 Description I Unitl/Unit2 lRev.0039 Document I OARecord lPaqe 260ot210 Appendix B (Page 10 of 14) 7.4 FIGURES AND TABLES (continued)

TABLE 3.1.1 Watts Bar Unit 2 PORV Setpoints vs Temperature (Data (Ref. 2) Plotted on Figure 3.1-1)

Temperature PCV-334 Setpoint PCV-340A Setpoint

('F) (psig) (psig) 60 425 425 120 425 425 130 495 495 170 495 495 195 720 640 250 720 640 300 724 640 350 720 640 450 2335 2335

system I REACTOR GOOLANT SYSTEM lsoo-N9o84001 Description I Unitl/Unit2 lRev.0039 Document I OARecord lpaqe 261 ot2t0 Appendix B (Page 11 ot 141 7.0 FIGURES AND TABLES (continued)

TABLE 4.0.1 Watts Bar Unat 2 Surveillance Gapsule Removal Schedule (a)

Gapsule Orientation of Lead Removal Expected Capsule Capsule Factor Time Fluence (n/cm',E> L0 MeV)

U Dual 34" 4.9 2.61 EFPY 0.77 x 101s (EOC 2)

(b)

W Single 34' 5.18 6.1 EFPY 3.17 x 101e X Dual 34o 5.1 3 6.2 EFPY to 3.17 x 1O1e to 12.5 EFPY (') 6.34 x 1o1e (c)

Z Single 34o 5.1 8 Standby ---r---

V Dual 31.5" 4.40 Standby Y Dual 31.5o 4.40 Standby -------

Notes:

(a) This information is taken from the withdrawal schedule contained in WCAP-9455, Revision 3 (Ref. 3) and WCAP-18191-Np, Rev. 0 / tetter WAT-D-12366 Rev.0 (Ref. 9).

(b) Approximate Fluence at vessel inner wall at End-of-Life (32 EFpy).

(c) Capsule X should be withdrawn between 6.2 EFpy and 12.5 EFpy, which corresponds to a capsule fluence of not less than once (3.17 x 1O1e n/cm2 (E > 1 .o MeV)) or greater than twice (6.34 x 1O1e nicm2 1E >

1.0 MeV)) the peak End-of-Life vessel fluence. This is consistent with the recommendations of ASTM E185-82.

System I REACTOR COOLANT SYSTEM SYSTEM ISDD-N36 ISDD.N3.68.4OO1 Description I unitl/Unit2 lRev.0039 Document I OARecord lPaEe 262o1270 Appendix B (Page 12 of 141 7.0 FIGURES AND TABLES (continued)

Table 5-1 Summary of the Best Estimate Cu and NiWeight Percent and lnitial RTNor Values for the Watts Bar Unit 2 Reactor Vessel Materials Materia! Description Chemical Gomposition !nitial Reactor Vessel RTNDT (a)

Cu Ni Beltline Region Location Wto/o wto/o lntermediate Shell Forging 05 0.05 0.78 14"F Lower Shell Forging 04 0.05 0.81 50F lntermediate to Lower Shell Circumferential 0.05 0.70 -50'F Weld Seam W05 Note:

(a) The initial RTxel values are measured values, taken from WCAP-13830, Revision 1

[Reference 8]

Table 5-2 Summary of the lnitial RTltor Values for the Watts Bar Unit 2 Closure Head and Vessel Flange Material ldentification lnitial RTNDT (a)

Closure Head Flange -40"F Vessel Flange -22"F Note:

(a) The initial RTrrror values are measured values, taken from WCAP-13830, Revision 1

[Reference 8]

Table 5-3 Summary of the Watts Bar Unit 2 Reactor Vessel Beltline Material Chemistry Factors Beltline Materials Ghemistry Factor lntermediate Shell Forging 05 31'F Lower Shell Forging 04 31"F lntermediate to Lower Shell 68'F Circumferential Weld Seam W05

System I REACTOR COOLANT SYSTEM ISDD-N3684001 Description I Unitl/Unit2 lRev.0039 Document I AA Record I Paqe 263o1270 Appendix B (Page l3 of 14) 7.0 FIGURES AND TABLES (continued)

Table 54 Fluence Values forthe Watts Bar Unit 2 Reactor Vessel Beltline Materials 7 EFPY Fluence (n/cm2, E>1.0 MeV)

Beltline Materials lnner 1l4T Location 3l4T Location Wetted (x=2.1 16 in.) (x=6.349 in.)

Surface lntermediate Shell Forging 05 6.93E+18 4.17E+18 1 .51 E+18 Lower Shell Forging 04 6.93E+18 4.17E+18 1 .51 E+18 lntermediate to Lower Shell 6.93E+18 4.17E+18 1 .51 E+18 Circumferential Weld Seam W05 Table 5-5 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 1/4T Location CF 1t4T t 1t4T ARTr.ror lRTr'ror(') or(") O6 M ART Reactor Vessel Location ('F) (n/cm2, FF ("F) ('F) ("F) ('F) ("F) ("F)

E>1.0 MeV) lntermediate Shell Forging 05 31 4.17E+18 0.757 23.5 14 0 11.7 23.5 61 Lower Shell Forging 04 31 4.17E+18 4.757 23.5 5 0 11.7 23.5 52 lntermediate to Lower Shell 68 4.17E+18 0.757 51 .5 -50 0 25.7 51.5 53 Circumferential Weld Seam W05 Note:

(a) The initial RTruor values are measured values; therefore, o;=Qof Table 5-6 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 3/4T Location CF 3t4T f 314T ARTruor lRTuor(") o,,(") O6 M ART Reactor Vessel Location ("F) (n/cm2, FF ('F) ('F) ("F) ("F) ("F) ('F)

E>1.0 MeV) lntermediate Shell Forging 05 31 1 .51E+18 0.504 15.6 14 0 7.8 15.6 45 Lower Shell Forging 04 31 1.51E+18 0.504 15.6 5 0 7.8 15.6 36

!ntermediate to Lower Shell 68 1 .51E+18 0.504 34.3 -50 0 17.1 34.3 19 Circumferential Weld Seam w05 Note:

(a) The initial RTr.ror values are measured values; therefore, oi=0oF

System REACTOR COOLANT SYSTEM sDD-N3-694001 Description Unit 1 lUnit2 Rev. 0039 Document OA Record Pase 264 of 270 Appendix B (Page 14 ot 141 7.0 FIGURES AND TABLES (continued)

Table 5-7 summary of the Limiting ART Values Used in the Generation of the watts Bar Unit 2 Heatup/Cooldown Curves Limiting ART ('F)

EFPY 1tAT 314T 7 61 45 Table 5-8 RTprs calculations for the watts Bar Unit 2 Belfline Materials at 32 EFpy 32 EFPY Material CF Fluence FF(") lRTlror ARTltor@) or(c) oo(d) ;y1(e) RTrrr(t)

('F) (n/cm2, ('F) ('F) ("F) ("F) ('F)

("F)

E>1.0 MeV) lntermediate Shell 31 3.1 7E+1 9 1.30 14 40.4 0 17 34 88 Forging 05 Lower Shell Forging 04 31 3.17E+19 1.30 5 40.4 0 17 34 79 lntermediate to Lower 68 3.1 7E+1 9 1.30 -50 88.7 0 28 56 95 Shell Circumferential Weld Seam W05 Note:

(a) FF = fluence factor = f{0 28 - 0'1 los (0)

(b) ARTxpl= ARTpls= CF . FF (c) As indicated in Table 5-1 of this report, the lRTHor values are measured; hence, according to 10 CFR 50.61, O, = OoF (d) Per the guidance of 10 CFR 50.61, the base metal o6 = 17"F and the weld metal oa = 2g"F when surveillance data is not utilized. However, o6 need not exceed 0.S*ARTxpl (e) Y = Margin = 2*(ou2, oi)'o (0 RTprs - lRTxpl + ARTpls + Margin