ML17313A128

Revised Pressure and Temperature Limits Report (PTLR)

ML17313A128
Person / Time
Site:	Watts Bar
Issue date:	11/09/2017
From:	Simmons P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML17313A128 (17)

Text

Tennessee Valley Authority , Post Office Box 2000 Spring City, Tennessee 37381 November 9, 2017 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 Unit 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 4, in accordance with Technical Specification Section 5.9.6.c.

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) 365-7720.

Respectfully ,

Paul Simmons Site Vice President Watts Bar Nuclear Plant

U.S. Nuclear Regulatory Commission Page 2 November 9, 2017

Enclosure:

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

Revision 4.

cc (Enclosure):

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

ENCLOSURE Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report, Revision 4 E-1 of 15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 250 of 269 APPENDIX "B" TO REC SYSTEM DESCRIPTION N3-68-4001 WATTS BAR UNIT 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

REVISION 4 Prepared by: C. S. Kerlin Checked by: M. R. Smith Approved by: R. E. Cox E-2of15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 251 of 269 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 (PIT) 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 60°F.

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, lnservice Hydrostatic and Leak Testing, and Criticality The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1and2.1-2 (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).

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System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev. 0038 Document QA Record Page 252 of 269 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 a 71 .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 ~ 225°F 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 shall be 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, RT NOT. which is determined in accordance with ASTM E208 (Ref. 5) . The empirical relationship between RT NOT 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 XI of the ASME Boiler and Pressure Vessel Code (Ref. 6). The surveillance capsule removal schedule meets the requirements of ASTM E185-82 (Ref. 7).

The removal schedule 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 Initial RT NOT Values for the Watts Bar Unit 2 Reactor Vessel Materials.

• Table 5-2 shows a Summary of the Initial RT NOT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange.

• Table 5-3 provides the Summary of the Watts Bar Unit 2 Reactor Vessel Beltline Material Chemistry Factors.

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

• 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.

• 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 .

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System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 253 of 269 5.0 SUPPLEMENTAL DATA TABLES (continued)

• Table 5-7 provides a Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves.

• Table 5-8 shows RT PTs Calculations for the Watts Bar Unit 2 Beltline Materials at 32 EFPY.

6.0 REFERENCES

1. WCAP-17035-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-5147, 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 Commission, 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 XI, 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 (IF), ASTM 1982.

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

E-5of15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 254 of 269 7.0 FIGURES AND TABLES MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 INITIAL RTNoT: 14°F LIMITING ART VALUES AT 7 EFPY: 1/4T, 61 ° F 3/4T, 45° F 2500 -------------

Leak Test Limit 2250 Critical Limit 100 Deg. F/Hr 2000 1750 6'

~-U) 1500 -+-------.

Unacceptable Acceptable Operation

~ Operation

l VI VI

~ 1250 D..

"Cl

~

"B 1000 cu 0

750 500 Criticality Lim it based on

• ___ inservice hydrostatic test 1

temperature (122°F) for the service period up to 7 EFPY 250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2.1-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°F/hr and 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K1c)

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

E-6of15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 255 of 269 7.0 FIGURES AND TABLES (continued)

TABLE 2.1-1 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 (oF) (psig) (oF) (psig) (oF) (psig) (oF) (psig) (oF) (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 1160 100 1256 150 1381 100 1128 150 1199 105 1314 155 1458 105 1160 155 1245 110 1381 160 1544 110 1199 160 1298 115 1458 165 1640 115 1245 165 1358 120 1544 170 1748 120 1298 170 1426 125 1640 175 1868 125 1358 175 1503 130 1748 180 2001 130 1426 180 1590 135 1868 185 2149 135 1503 185 1687 140 2001 190 2312 140 1590 190 1795 145 2149 145 1687 195 1915 150 2312 150 1795 200 2048 155 1915 205 2196 160 2048 210 2360 165 2196 170 2360 E-7of15

System REACTOR COOLANT SYSTEM SDD-NJ-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 256 of 269 7.0 FIGURES AND TABLES (continued)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 INITIAL RT NDT: 14°F LIMITING ART VALUES AT 7 EFPY: 1/4T, 61°F 3/4T, 45°F 2500 .-.==============================;-*

Operlim Version :5 .2 Run :30881 Operlim .xls Version : 5.2 2250 Unacceptable 2000 Operation 1750 s

~ 1500 Acceptable

~ Operation

s Cl)

Cl)

<V 1250

~ Cool down "C Rates

~

"F/Hr steady-state "B 1000 *20

-40 (ij 0 *60

-100 750 500 250 0 -+-i-.................++-.--..,....,-+-;-........-...-+-.......-..-.-+-r-..-....-r-+-r-r-.....,...,l....,-,,....,-,-+-t-,--,-,-+-,-..,-,_,.-+-,-...,....,.....-r...,.....,....,.....,..-l 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2.1-2 Watts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K1c)

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

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System REACTOR COOLANT SYSTEM SDD-NJ-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 257 of 269 7.0 FIGURES AND TABLES {continued)

TABLE 2.1-2 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°F/HR 40°F/HR 60°F/HR 100°F/HR T p T p T p T p T p (oF) (psig) (oF) (psig) (oF) (psig) (oF) (psig) (oF) (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 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 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 1508 105 1508 105 1508 105 1508 105 1508 110 1603 110 1603 110 1603 110 1603 110 1603 115 1709 115 1709 115 1709 115 1709 115 1709 120 1825 120 1825 120 1825 120 1825 120 1825 125 1954 125 1954 125 1954 125 1954 125 1954 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 E-9of15

System REACTOR COOLANT SYSTEM SDD-NJ-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 258 of 269 7.0 FIGURES AND TABLES {continued)

Setpoint Window 2700 2200 llQ

': 1700 -

c

• sQ.

a>

V\

>~ 1200 c..

PCV 334 PCV 340A 200 -t----~~-t-~~-+-~~--r~~~+--~~-t---~~-t-~~-1-~~--+~~-

so 100 150 200 250 300 350 400 450 500 Indicated RCS Temperature, °F Figure 3.1-1 PORV Setpoint vs RCS Temperature (Plotted data (Ref. 2) provided in Table 3.1-1)

E-10of15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 259 of 269 7.0 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 (oF) (psig) (psig) 60 425 425 120 425 425 130 495 495 170 495 495 195 720 640 250 720 640 300 720 640 350 720 640 450 2335 2335 E-11 of 15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 260 of 269 7.0 FIGURES AND TABLES (continued)

TABLE 4.0-1 Watts Bar Unit 2 Surveillance Capsule Removal Schedule (a)

Capsule Orientation of Lead Removal Expected Capsule Capsule Factor Time Fluence (n/cm 2 ,E > 1.0 MeV) u Dual 34° 5.13 1st Refuel 0.50 x 10 19 Outage w Single 34° 5.18 6.1 EFPY 3 .17 x 10 19 (b )

x Dual 34° 5.13 6.2 EFPY to 3. 17 x 1019 to 12.5 EFPY (c) 6.34 X 10 19 (c) z Single 34° 5.18 Standby -------

v Dual 31.5° 4.40 Standby -------

y Dual 31.5° 4.40 Standby -------

Notes:

(a) This information is taken from the withdrawal schedule contained in WCAP-9455, Revision 3 (Ref. 3).

(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 19 2 10 19 n/cm 2 (E > 1.0 MeV)) or greater than twice (6.34 x 10 n/cm (E >

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

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System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 261 of 269 7.0 FIGURES AND TABLES (continued)

Table 5-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NOT Values for the Watts Bar Unit 2 Reactor Vessel Materials Material Description Chemical Composition Initial Reactor Vessel Cu RTNDT (a)

Ni Beltline Region Location wt% wt%

Intermediate Shell Forging 05 0.05 0.78 14°F Lower Shell Forging 04 0.05 0.81 5°F Intermediate to Lower Shell Circumferential 0.05 0.70 -50°F Weld Seam W05 Note:

(a) The initial RT NOT values are measured values, taken from WCAP-13830, Revision 1

[Reference 8]

Table 5-2 Summary of the Initial RT NOT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange Material Identification Initial RTNDT (a)

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

(a) The initial RT NOT 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 Chemistry Factor Intermediate Shell Forging 05 31 °F Lower Shell Forging 04 31 °F Intermediate to Lower Shell 68°F Circumferential Weld Seam W05 E-13 of 15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 262 of 269 7.0 FIGURES AND TABLES {continued)

Table 5-4 Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline Materials 7 EFPY Fluence 2

(n/cm , E>1.0 MeV)

Beltline Materials Inner 1/4T Location 3/4T Location Wetted (x=2.116 in.) (x=6.349 in.)

Surface Intermediate Shell Forging 05 6.93E+18 4.17E+18 1.51E+18 Lower Shell Forging 04 6.93E+18 4.17E+18 1.51E+18 Intermediate to Lower Shell 6.93E+18 4.17E+18 1.51E+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 1/4T f 1/4T .::\RTNOT IRTNOT(a) 01(a) a!!. M ART 2

Reactor Vessel Location (oF) (n/cm , FF (oF) (oF) (oF) (oF) (oF) (oF)

E>1.0 MeV)

Intermediate 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 0.757 23.5 5 0 11.7 23.5 52 Intermediate 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 RT NOT values are measured values; therefore, cri=0°F 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 3/4T f 3/4T .::\RT NOT IRTNoT(a) 01(a) a!!. M ART 2

Reactor Vessel Location (oF) (n/cm , FF (oF) (oF) (oF) (oF) (oF) (oF)

E>1.0 MeV)

Intermediate 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 Intermediate 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 RT NDT values are measured values; therefore, cri=0°F E-14 of 15

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 I Unit 2 Rev.0038 Document QA Record Page 263 of 269 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 1/4T 3/4T 7 61 45 Table 5-8 RT PTs Calculations for the Watts Bar Unit 2 Beltline Materials at 32 EFPY 32 EFPY Material CF Fluence FF(aJ IRTNOT LiRT NDT(b) Ou(c) 01i(d) M(eJ RTPTS(f) 2 (oF) (n/cm , (oF) (oF) (oF) (oF) (oF)

(oF)

E>1.0 MeV)

Intermediate Shell 31 3.17E+19 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 Intermediate to Lower 68 3.17E+19 1.30 -50 88.7 0 28 56 95 Shell Circumferential Weld Seam W05 Note:

(a) FF = fluence factor= t< 0 0 *1109 (t))

(b) ~RT NOT= ~RT PTS =CF* FF (c) As indicated in Table 5-1 of this report, the IRT NOT values are measured; hence, according to 10 CFR 50.61, Ou= 0°F (d) Per the guidance of 10 CFR 50.61, the base metal 01i = 17°F and the weld metal 01i = 28°F when surveillance data is not utilized. However, a~ need not exceed 0.5*~RT NOT (e) M =Margin= 2 *(cru 2 + cr~ 2 ) 112 (f) RT PTs - IRT NOT+ ~RT PTs + Margin E-15 of 15