Text

Submittal of Pressure and Temperature Limits Report
	ML21104A371
Person / Time
Site:	Vogtle
Issue date:	04/14/2021
From:	Whitley B; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	ND-21-0366
	Download: ML21104A371 (40)
	v • d • e

A Southern Nuclear Brian H. Whitley Southern Nuclear Director, Regulatory Affairs Operating Company, Inc.

3535 Colonnade Parkway Birmingham , AL 35243 Tel 205.992.7079 April 14, 2021 Docket Nos. : 52-025 ND-21-0366 52-026 10 CFR 52.97(c)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3&4 Pressure and Temperature Limits Report {PTLR)

Ladies and Gentlemen:

In accordance with Technical Specifications (TS) 5.6.4.c for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Southern Nuclear Operating Company (SNC) submits the VEGP Unit 3 Pressure and Temperature Limits Report (PTLR), Revision 0, and the VEGP Unit 4 PTLR, Revision 0. provides the VEGP Unit 3 PTLR, Revision 0. provides the VEGP Unit 4 PTLR, Revision 0.

This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.

If you have any questions, please contact Amy Chamberlain at 205.992.6361.

Respectfully submitted, Brian H. Whitley Regulatory Affairs Director Southern Nuclear Operating Company

U.S. Nuclear Regulatory Commission ND-21-0366 Page 2 of 4

Enclosures:

1) Vogtle Electric Generating Plant (VEGP) Unit 3 Pressure and Temperature Limits Report (PTLR), Revision 0

2) Vogtle Electric Generating Plant (VEGP) Unit 4 Pressure and Temperature Limits Report (PTLR), Revision 0

U.S. Nuclear Regulatory Commission ND-21-0366 Page 3 of 4 cc:

Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures)

Mr. P. P. Sena III (w/o enclosures)

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Mr. B. H. Whitley Mr. W. Levis Ms. C. A. Gayheart Ms. M. Ronnlund Mr. J. M DeLano Mr. M. J. Yox Mr. C. T. Defnall Mr. J. Tupik Ms. A. C. Chamberlain Mr. S. Leighty Ms. K. Roberts Mr. J. Haswell Mr. D. T. Blythe Mr. K. Warren Mr. A. S. Parton Mr. A. Nix Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. M. King (w/o enclosures)

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U.S. Nuclear Regulatory Commission ND-21-0366 Page 4 of 4 Nuclear Regulatory Commission (contd)

Mr. J. Gaslevic Mr. O. Lopez-Santiago Mr. M. Webb Mr. B. Gleaves Mr. T. Fredette Mr. S. Rose Ms. K. McCurry Mr. B. Davis State of Georgia Mr. R. Dunn Oglethorpe Power Corporation Mr. M. W. Price Mr. B. Brinkman Mr. E. Rasmussen Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson Dalton Utilities Mr. T. Bundros Westinghouse Electric Company, LLC Mr. L. Oriani (w/o enclosures)

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Mr. M. Corletti Mr. D. Hawkins Mr. J. Coward Other Mr. S. W. Kline, Bechtel Power Corporation Ms. L. A. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc.

Mr. S. Roetger, Georgia Public Service Commission Mr. R.L. Trokey, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch Bingham

Southern Nuclear Operating Company ND-21-0366 Enclosure 1 Vogtle Electric Generating Plant (VEGP) Unit 3 Pressure and Temperature Limits Report (PTLR), Revision 0 (This Enclosure consists of 18 pages, including this cover page)

Vogtle Electric Generating Plant (VEGP) Unit 3 Pressure and Temperature Limits Report (PTLR)

Revision 0 April 2021

Pressure and Temperature Limits Report - Unit 3 Table of Contents List of Tables ........................................................................................................................... iii List of Figures ........................................................................................................................... iv Record of Revision ...................................................................................................................... v 1.0 RCS Pressure and Temperature Limits Report (PTLR) ................................................... 1 2.0 Operating Limits ............................................................................................................. 1 2.1 RCS Pressure and Temperature Limits ............................................................................. 1 2.2 Low Temperature Overpressure Protection (LTOP) System .............................................. 2 3.0 Reactor Vessel Material Surveillance Program ............................................................... 3 4.0 Supplemental Data Tables .............................................................................................. 4 5.0 References ....................................................................................................................12 ii

Pressure and Temperature Limits Report - Unit 3 List of Tables Table 1: RCS Heatup Limits at 54 EFPY ................................................................................... 7 Table 2: RCS Cooldown Limits at 54 EFPY ............................................................................... 8 Table 3: Maximum Composition Limits for Reactor Vessel Beltline Materials............................. 9 Table 4: Mechanical Properties for Reactor Vessel Materials .................................................... 9 Table 5: Peak Reactor Vessel Neutron Fluence Projections at 56 EFPY ................................... 9 Table 6: Adjusted Reference Temperature Calculations at 56 EFPY ........................................10 Table 7: RTNDT and RTPTS Calculations at 56 EFPY ................................................................ 11 Table 8: Beltline Reactor Vessel Materials USE Projection at 56 EFPY .................................... 11 iii

Pressure and Temperature Limits Report - Unit 3 List of Figures Figure 1: Reactor Coolant System Heatup Limitations (Maximum Heatup Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Errors) .......................... 5 Figure 2: Reactor Coolant System Cooldown Limitations (Maximum Cooldown Rate of 100°F/hr) Applicable to 54 EFPY (without Margins for Instrumentation Error)............. 6 iv

Pressure and Temperature Limits Report - Unit 3 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.4. The limiting condition for operation (LCO) 3.4.3, reactor coolant system (RCS) pressure and temperature (P/T) limits, also referred to as heatup and cooldown limit curves, are contained in this PTLR. Revisions to the PTLR shall be provided to the U.S. Nuclear Regulatory Commission (NRC) after issuance. Note that these PTLR limits are consistent with, and bounded by, the generic AP1000 PTLR limits

[18]. Reference 17 utilized site-specific material properties as inputs to validate that the P/T limit curves in Reference 18 remained bounding for VEGP Units 3 and 4.

This PTLR follows the guidelines set forth in U.S. NRC Generic Letter 96-03 [1].

2.0 OPERATING LIMITS 2.1 RCS PRESSURE AND TEMPERATURE LIMITS The RCS P/T limits for LCO 3.4.3, presented in Figures 1 and 2 and tabulated in Tables 1 and 2, were developed using the NRC-approved methodology in WCAP-14040-A [2].

The boltup temperature shall be 60°F. The minimum (allowable) boltup temperature is established as the higher of 60°F or the highest material reference temperature (initial RTNDT) in the highly-stressed reactor pressure vessel (RPV) flange region. The heatup and cooldown curves utilize 60°F, since this value is limiting for the respective vessel materials. This allows the plant to perform the boltup operations at 60°F or higher.

The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°F in any one hour period.

b. A maximum cooldown rate of 100°F in any one hour period.

c. A maximum temperature change of 10°F in any one-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

Page 1 of 13

Pressure and Temperature Limits Report - Unit 3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 1 and 2. Data points for these figures are tabulated in Tables 1 and 2, respectively. The 54 effective full power years (EFPY) term of applicability identified for these limit curves is based on adjusted reference temperature (ART) calculations that utilize Regulatory Guide 1.99, Revision 2 [3] Position 1.1 chemistry factors and peak neutron fluence projections on the vessel forgings and beltline welds. The inputs for the ART calculations are contained in APP-RXS-M3C-026

[4] and APP-MV01-Z0-101 [5], and are summarized in Tables 3 through 5. Calculated ART values, used in the original development of the P/T limits, are documented in APP-RXS-M3C-012 [6]. These ART values are conservative for use as summarized in Table 6 and APP-RXS-M3C-033 [16]. The inputs and calculations are discussed in Section 4.0.

2.2 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM Low Temperature Overpressure Protection (LTOP) is required to provide overpressure protection when any RCS cold leg temperature is <275F with the reactor vessel head on as required by Technical Specifications and APP-RCS-M3-001 [13]. This temperature is above the minimum LTOP enable temperature calculated using ASME Code Case N-641 (See Reference [2]).

A method of LTOP of the reactor vessel is provided by two parallel relief valves in the Normal Residual Heat Removal System (RNS) suction line inside containment. The larger RNS pump suction line relief valve (RNS-PL-V021) has a lift setpoint of 500 psig with a full open pressure of 550 psig [14]. The smaller relief valve (RNS-PL-V020) has a lift setpoint of 470 psig with a full open pressure of 517 psig [15]. The set pressure of RNS-PL-V020 is lower than the set pressure for RNS-PL-V021 so that the RNS-PL-V020 valve lifts first, and it is intended to be the only relief valve to lift during pressure transients that can be relieved with its capacity.

The lift setpoints were determined using the methodology in WCAP-14040-A [2], and the validation of the adequate capacity and set pressures of the valves is in APP-RNS-M3C-002 [12].

The maximum allowable RNS pump suction line relief valve lift setpoints are derived by analysis of the LTOP design basis Mass Input (MI) and the Heat Input (HI) transients.

Operation with RNS pump suction line relief valves lift setpoints of less than or equal to the maximum allowable setpoint, ensures that the allowable steady state pressure temperature limit shown in Table 2 [6] will not be violated with consideration for: (1) relief valves set pressure tolerances and accumulations; and (2) effects of reactor coolant pump operation.

To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications and operating procedures place limitations on coolant input capabilities and RCP operation during the appropriate LTOP modes. The magnitude of the maximum mass injection rate is controlled by isolating accumulators Page 2 of 13

Pressure and Temperature Limits Report - Unit 3 and closing the Chemical and Volume Control System (CVS) makeup line containment isolation valve, CVS-PL-V091, to limit the mass injection rate to the relieving capacity of RNS-PL-V020. The magnitude of the maximum heat injection rate is also administratively controlled. To limit the heat injection rate, the reactor coolant pumps are started at less than 25% speed and the steam generator temperatures must be less than or equal to 50°F higher than each of the cold leg temperatures prior to pump start [12, 13]. The required relieving rate for this transient exceeds the capacity of RNS-PL-V020, and causes lifting of the higher capacity relief valve RNS-PL-V021. Both relief valves will lift and maintain the peak pressure below the limit in Table 2.

3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. Table 1 of ASTM E-185 [7]

identifies the requirement for four capsules to be withdrawn for a maximum projected transition temperature shift (RTNDT) of the beltline materials exceeding 100°F. The surveillance program withdrawal schedule for removal of the capsules for post-irradiation testing exceeds this requirement and identifies five capsules to be withdrawn.

The surveillance capsule withdrawal schedule and pressure vessel steel surveillance program is in compliance with Appendix H to 10CFR50, Reactor Vessel Material Surveillance Program Requirements [8]. Accordingly, the surveillance capsule withdrawal schedule meets the requirements of ASTM E185 [7], supplemented as needed for a 60-year design life. The results of these examinations shall be used to update the RCS P/T limits. The recommended surveillance capsule withdrawal schedule, shown below, is consistent with the AP1000 certified design.

Capsule Withdrawal Time 1st When the accumulated neutron fluence of the capsule is 5 x 1018 n/cm2.

2nd When the accumulated neutron fluence of the capsule corresponds to the approximate end of life fluence at the reactor vessel 1/4T location.

3rd When the accumulated neutron fluence of the capsule corresponds to the approximate end of life fluence at the reactor vessel inner wall location.

4th When the accumulated neutron fluence of the capsule corresponds to a fluence not less than once or greater than twice the peak end of vessel life fluence.

5th End of plant design objective of 60 years.

6th Standby 7th Standby 8th Standby Page 3 of 13

Pressure and Temperature Limits Report - Unit 3 4.0 SUPPLEMENTAL DATA TABLES Reference 17, Appendix A, provides less restrictive VEGP Units 3 and 4 specific updated tables that reflect the plant-specific material properties; however, the generic AP1000 tables remain bounding and are utilized as the approved plant-specific data tables.

Table 3 contains the elemental chemistry limits for the reactor vessel beltline materials, along with the Regulatory Guide 1.99, Revision 2 [3] Position 1.1 chemistry factors based on these limits. The chemistry values listed are the maximum allowed values and are therefore conservative for use in determining the effects of irradiation embrittlement for the beltline materials.

Table 4 contains the mechanical properties of the reactor vessel materials utilized in calculating the ARTs for the P/T limit curves, the preliminary pressurized thermal shock (PTS) reference temperature (RTPTS) values, and projection of upper shelf energy (USE) for the reactor vessel materials.

Table 5 contains the maximum projected neutron fluence values for the beltline materials of the reactor vessel at end-of-life (EOL), which was assumed to be 56 EFPY, along with the calculated fluence values at the 1/4T position (vessel wall quarter thickness from the inside surface) and 3/4T position (vessel wall quarter thickness from the outside surface).

Table 6 contains the calculations of the ARTs at the 1/4T position and 3/4T position, which are bounded by the ART values used in the determination of the P/T limit curves for normal operation [9, 10]. Note that the 56 EFPY ART values in Table 6 are lower than those utilized in the original development of the P/T limit curves, which are listed on Figures 1 and 2. The Figure 1 and 2 54 EFPY ARTs remain acceptable for use, since they bound the values in Table 6 for 56 EFPY.

Table 7 contains the calculations of RTNDT and RTPTS. As noted in Section 3, values of RTNDT exceeding 100°F require four surveillance capsules to be withdrawn in accordance with ASTM E185 [7]. The maximum RTNDT value is calculated to be 109.0°F (based on Position 1.1 Chemistry Factors). The screening criteria for PTS are provided in 10CFR50.61 [11], which states that the values of RTPTS (using EOL neutron fluence projections) must be less than 270°F for plates, forgings, and axial weld materials, and less than 300°F for circumferential weld materials. The preliminary RTPTS values are calculated to be 78.4°F for the beltline forgings and 145.0°F or less for the beltline circumferential welds, and are in compliance with 10CFR50.61 [11].

Table 8 contains the conservative projections of USE for the reactor vessel beltline materials to show compliance with the requirements of 10CFR50, Appendix G [9].

Page 4 of 13

Pressure and Temperature Limits Report - Unit 3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: SHELL FORGING BOUNDING ART VALUES AT 54 EFPY: 1/4T, 90°F 3/4T, 82°F 1.....T,.._tiPJil r rP *..

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Figure 1: Reactor Coolant System Heatup Limitations (Maximum Heatup Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Errors)

(Plotted Data provided on Table 1)

Page 5 of 13

Pressure and Temperature Limits Report - Unit 3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: SHELL FORGING BOUNDING ART VALUES AT 54 EFPY: 1/4T, 90°F 3/4T, 82°F

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Figure 2: Reactor Coolant System Cooldown Limitations (Maximum Cooldown Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Error)

(Plotted Data provided on Table 2)

Page 6 of 13

Pressure and Temperature Limits Report - Unit 3 Table 1: RCS Heatup Limits at 54 EFPY (without uncertainties for instrumentation errors) 50°F/hr Heatup Critical Limit 100°F/hr Heatup Critical Limit Leak Test Limit T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 145 -14.7 60 -14.7 145 -14.7 127 2,000 60 621 145 621 60 621 145 621 127 2,000 65 621 145 621 65 621 145 621 145 2,485 70 621 145 621 70 621 145 621 145 2,485 75 621 145 621 75 621 145 621 80 621 145 621 80 621 145 621 85 621 145 621 85 621 145 621 90 621 145 621 90 621 145 621 95 621 145 621 95 621 145 621 100 621 145 621 100 621 145 621 105 621 150 621 105 621 150 621 110 621 155 621 110 621 155 621 115 621 160 621 115 621 160 621 120 621 165 621 120 621 165 621 125 621 170 621 125 621 170 621 130 621 170 1,268 130 621 170 1,004 130 621 175 1,337 130 621 175 1,042 130 1,268 180 1,413 130 1,004 180 1,085 135 1,337 185 1,499 135 1,042 185 1,134 140 1,413 190 1,593 140 1,085 190 1,188 145 1,499 195 1,697 145 1,134 195 1,249 150 1,593 200 1,812 150 1,188 200 1,318 155 1,697 205 1,940 155 1,249 205 1,394 160 1,812 210 2,081 160 1,318 210 1,478 165 1,940 215 2,236 165 1,394 215 1,572 170 2,081 220 2,408 170 1,478 220 1,675 175 2,236 175 1,572 225 1,790 180 2,408 180 1,675 230 1,917 185 1,790 235 2,058 190 1,917 240 2,213 195 2,058 245 2,384 200 2,213 205 2,384 Page 7 of 13

Pressure and Temperature Limits Report - Unit 3 Table 2: RCS Cooldown Limits at 54 EFPY (without uncertainties for instrumentation errors)

Steady State 20°F/hr 40°F/hr 60°F/hr 100°F/hr T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.7 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 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 1,557 130 1,557 130 1,557 130 1,557 130 1,557 135 1,653 135 1,653 135 1,653 135 1,653 135 1,653 140 1,758 140 1,758 140 1,758 140 1,758 140 1,758 145 1,874 145 1,874 145 1,874 145 1,874 145 1,874 150 2,003 150 2,003 150 2,003 150 2,003 150 2,003 155 2,145 155 2,145 155 2,145 155 2,145 155 2,145 160 2,302 160 2,302 160 2,302 160 2,302 160 2,302 165 2,476 165 2,476 165 2,476 165 2,476 165 2,476 Page 8 of 13

Pressure and Temperature Limits Report - Unit 3 Table 3: Maximum Composition Limits for Reactor Vessel Beltline Materials Elements Maximum Weight %

Reactor Vessel Beltline Materials CF (a)

Cu Ni P V S Beltline Forgings 0.06 0.85 0.01 0.05 0.01 37 Beltline Circumferential Welds 0.06 0.85 0.01 0.05 0.01 82 Notes:

(a) Chemistry factors (CFs) are based on the maximum allowed Cu and Ni weight percentage values for manufacture [5, 6, and 16]. These chemistry factors were determined in accordance with Position 1.1 of Regulatory Guide 1.99, Revision 2 [3] and have units of °F.

Table 4: Mechanical Properties for Reactor Vessel Materials Reactor Vessel Materials Initial RTNDT (a) Initial USE (b)

Beltline Forgings -10°F > 75 ft-lbs Beltline Circumferential Welds -20°F > 75 ft-lbs Reactor Vessel Closure Head 10°F N/A Reactor Vessel Flange 10°F N/A Notes:

(a) These values for initial RTNDT are the basis for the ART calculations used to determine the P/T limit curves.

These values are also used in the RTPTS calculations for EOL.

(b) The USE for the unirradiated materials that comprise regions of the reactor vessel that will be exposed to neutron fluences estimated to be over 1 x 1019 n/cm2 (E > 1.0 MeV) must be at least 75 ft-lbs [11].

Table 5: Peak Reactor Vessel Neutron Fluence Projections at 56 EFPY Reactor Vessel I Surface Fluence I I I (a) 1/4T Fluence(b) 3/4T Fluence(c)

Materials Beltline Forgings 7.32 x 1019 n/cm2 4.42 x 1019 n/cm2 1.61 x 1019 n/cm2 Upper Circumferential Weld 2.24 x 1019 n/cm2 1.35 x 1019 n/cm2 0.494 x 1019 n/cm2 Lower Circumferential Weld I 3.54 x 1019 n/cm2 I 2.14 x 1019 n/cm2 I 0.780 x 1019 n/cm2 Notes:

(a) The surface fluence values represent the maximum projected neutron fluence (E > 1.0 MeV) at the clad/base metal interface for each of these separate reactor vessel beltline materials.

(b) The 1/4T fluence values represent the attenuated neutron fluence values (E > 1.0 MeV) at the quarter-thickness position in the vessel wall from the inside surface, based on a wall thickness of 8.4 inches, calculated using Equation 3 in Regulatory Guide 1.99, Revision 2 [3].

(c) The 3/4T fluence values represent the attenuated neutron fluence values (E > 1.0 MeV) at the quarter-thickness position in the vessel wall from the outside surface, based on a wall thickness of 8.4 inches, calculated using Equation 3 in Regulatory Guide 1.99, Revision 2 [3].

Page 9 of 13

Pressure and Temperature Limits Report - Unit 3 Table 6: Adjusted Reference Temperature Calculations at 56 EFPY 1/4T Position ART Calculations Reactor Vessel Materials CF 1/4T IRTNDT(b) RTNDT(c) Margin(d) ART(e)

(°F) FF(a) (°F) (°F) (°F) (°F)

Beltline Forgings 37 1.378 -10 51.0 34.0 75.0 Upper Circumferential Weld 82 1.084 -20 88.9 56.0 124.9 Lower Circumferential Weld 82 1.207 -20 98.9 56.0 134.9 3/4T Position ART Calculations Reactor Vessel Materials CF 3/4T IRTNDT(b) RTNDT(c) Margin(d) ART(e)

(°F) FF(a) (°F) (°F) (°F) (°F)

Beltline Forgings 37 1.132 -10 41.9 34.0 65.9 Upper Circumferential Weld 82 0.803 -20 65.9 56.0 101.9 Lower Circumferential Weld 82 0.930 -20 76.3 56.0 112.3 Notes:

(a) The fluence factor (FF) is determined from the corresponding neutron fluence value by the equation provided in Regulatory Guide 1.99, Revision 2 [3], which states that FF = f0.28 - 0.10 log f , where f is the high energy neutron fluence value (1019 n/cm2, E > 1.0 MeV). 1/4T and 3/4T fluence values are provided in Table 5.

(b) Initial RTNDT values are the values taken from Table 4.

(c) RTNDT = CF

FF, as stated by Equation 2 in Regulatory Guide 1.99, Revision 2 [3].

(d) Margin (M) = 2 *(i2 + 2)1/2, as stated by Equation 4 in Regulatory Guide 1.99, Revision 2 [3], where i is the standard deviation for the initial RTNDT values and is the standard deviation for RTNDT I is set equal to 0°F for the calculations since the initial RTNDT values are to be measured values per [5]. The values were 28°F for the welds and 17°F for the base metal (forging), as specified in the last paragraph of Section 1.1 of Regulatory Guide 1.99, Revision 2 [3].

(e) ART = Initial RTNDT + RTNDT + Margin, as stated by Equation 2 in Regulatory Guide 1.99, Revision 2 [3]. Note that the limiting ART values used in the development of the P/T limit curves conservatively bound the values shown in this table. The ART values utilized for P/T limit curve development are conservative values initially calculated for the forgings (90°F for 1/4T and 82°F for 3/4T) and consider the more restrictive axial flaw orientation in comparison to the relaxed Circ-Flaw methodology allowed for the circumferential girth welds

[9,10].

Page 10 of 13

Pressure and Temperature Limits Report - Unit 3 Table 7: RTNDT and RTPTS Calculations at 56 EFPY Reactor Vessel Materials CF IRTNDT(b) RTNDT(c) Margin(d) RTPTS(e)

FF(a)

(°F) (°F) (°F) (°F) (°F)

Beltline Forgings 37 1.470 -10 54.4 34.0 78.4 Upper Circumferential Weld 82 1.218 -20 99.9 56.0 135.9 Lower Circumferential Weld 82 1.329 -20 109.0 56.0 145.0 Notes:

(a) The fluence factor (FF) is determined from the corresponding neutron fluence value by the equation provided in Regulatory Guide 1.99, Revision 2 [3], which states that FF = f0.28 - 0.10 log f , where f is the high energy neutron fluence value (1019 n/cm2, E > 1.0 MeV).

(b) Initial RTNDT values are the values taken from Table 4.

(c) RTNDT = CF

FF, as stated by Equation 2 in Regulatory Guide 1.99, Revision 2 [3].

(d) Margin (M) = 2 *(i2 + 2)1/2, as stated by Equation 4 in Regulatory Guide 1.99, Revision 2 [3], where i is the standard deviation for the initial RTNDT values and is the standard deviation for RTNDT. I is set equal to 0°F for the calculations since the initial RTNDT values are to be measured values per [5]. The values were 28°F for the welds and 17°F for the base metal (forging), as specified in the last paragraph of Section 1.1 of Regulatory Guide 1.99, Revision 2 [3].

(e) RTPTS = Initial RTNDT + RTPTS + Margin, as stated by Equation 4 in 10CFR50.61 [11], where RTPTS RTNDT.

Table 8: Beltline Reactor Vessel Materials USE Projection at 56 EFPY Position 1.2 Projected EOL Reactor Vessel Materials Initial USE (a) 1/4T Fluence(b)

I I I USE Decrease(c) I USE(d)

Beltline Forgings > 75 ft-lbs 4.42 x 1019 n/cm2 27% 54.8 ft-lbs Upper Circumferential Weld > 75 ft-lbs 1.35 x 1019 n/cm2 26% 55.5 ft-lbs I Lower Circumferential Weld I

> 75 ft-lbs I 2.14 x 1019 n/cm2 I 29% I 53.3 ft-lbs I Notes:

(a) The upper shelf energy for the unirradiated materials that comprise regions of the reactor vessel that will be exposed to neutron fluences estimated to be over 1 x 1019 n/cm2 (E > 1.0 MeV) must be at least 75 ft-lbs [9].

Modern advances in manufacturing and steel production have generally provided reactor vessel plates, forgings, and welds with USE values that exceed 100 ft-lbs.

(b) USE projections are based on the 1/4T neutron fluence values in Table 5.

(c) The upper shelf energy is predicted to decrease (in accordance with Position 1.2 of Regulatory Guide 1.99, Revision 2 [3]) as a function of neutron fluence and Cu content as illustrated in Figure 2 of Regulatory Guide 1.99, Revision 2 [3].

(d) The projected EOL (56 EFPY) USE values are based on the limiting copper chemistry allowed (see Table 3) and an assumed 75 ft-lb USE for the unirradiated material (the minimum allowed). The predicted decrease in USE for the reactor vessel materials conservatively uses the 0.10 wt% Cu content line for base and weld metal (from Figure 2 of Regulatory Guide 1.99, Revision 2 [3]) in the projection, even though the maximum Cu content permitted is 0.06 weight percentage.

Page 11 of 13

Pressure and Temperature Limits Report - Unit 3

5.0 REFERENCES

1. U.S. Nuclear Regulatory Commission Generic Letter, 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, January 31, 1996.

2. Westinghouse Report, WCAP-14040-A, Rev. 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves WOG Programs: MUHP-5073 MUHP-3073, May 30, 2004.

3. U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

4. Westinghouse Calculation Note, APP-RXS-M3C-026, Revision 1, AP1000 Neutron Fast Fluence, DPA and Heating Rate Evaluation Long Term Power Distribution, April 13, 2012.

5. Westinghouse Design Specification, APP-MV01-Z0-101, Revision 15, Design Specification for AP1000 Reactor Vessel for System: Reactor Coolant System (RCS),

May 12, 2017.

6. Westinghouse Calculation Note, APP-RXS-M3C-012, Revision 1, AP1000 Pressure and Temperature Limit Curve Development, October 31, 2016.

7. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

8. Code of Federal Regulations, 10CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

9. Code of Federal Regulations, 10CFR50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C. Federal Register, Volume 60, No. 243, dated December 19, 1995.

10. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, 1998 Edition through 2000 Addenda.

11. Code of Federal Regulations, 10CFR50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

12. Westinghouse Calculation Note, APP-RNS-M3C-002, Revision 5 AP1000 LTOPS Analyses / Normal RNR Relief Valve Sizing Evaluation, June 13, 2017.

Page 12 of 13

Pressure and Temperature Limits Report - Unit 3

13. Westinghouse Specification, APP-RCS-M3-001, Revision 11, Reactor Coolant System, System Specification Document, September 5, 2017.

14. Westinghouse Datasheet, APP-PV16-Z0D-105, Revision 4, PV16 Datasheet 105, October 24, 2011.

15. Westinghouse Datasheet, APP-PV16-Z0D-211, Revision 1, PV16 Datasheet 211, October 6, 2016.

16. Westinghouse Calculation Note, APP-RXS-M3C-033, Revision 0, AP1000 Reactor Vessel Integrity Evaluations, April, 2018

17. Westinghouse Document SV0-RXS-GLR-001, Revision 0, Validation of the AP1000 Pressure-Temperature Limits Report (PTLR) for Use at Vogtle Units 3 and 4, May 2020.

18. Westinghouse Document APP-RXS-Z0R-001, Revision 3, AP1000 Generic Pressure Temperature Limits Report, December 2017.

Page 13 of 13

Southern Nuclear Operating Company ND-21-0366 Enclosure 2 Vogtle Electric Generating Plant (VEGP) Unit 4 Pressure and Temperature Limits Report (PTLR), Revision 0 (This Enclosure consists of 18 pages, including this cover page)

Vogtle Electric Generating Plant (VEGP) Unit 4 Pressure and Temperature Limits Report (PTLR)

Revision 0 April 2021

Pressure and Temperature Limits Report - Unit 4 Table of Contents List of Tables ........................................................................................................................... iii List of Figures ........................................................................................................................... iv Record of Revision ...................................................................................................................... v 1.0 RCS Pressure and Temperature Limits Report (PTLR) ................................................... 1 2.0 Operating Limits ............................................................................................................. 1 2.1 RCS Pressure and Temperature Limits ............................................................................. 1 2.2 Low Temperature Overpressure Protection (LTOP) System .............................................. 2 3.0 Reactor Vessel Material Surveillance Program ............................................................... 3 4.0 Supplemental Data Tables .............................................................................................. 4 5.0 References ....................................................................................................................12 ii

Pressure and Temperature Limits Report - Unit 4 List of Tables Table 1: RCS Heatup Limits at 54 EFPY ................................................................................... 7 Table 2: RCS Cooldown Limits at 54 EFPY ............................................................................... 8 Table 3: Maximum Composition Limits for Reactor Vessel Beltline Materials............................. 9 Table 4: Mechanical Properties for Reactor Vessel Materials .................................................... 9 Table 5: Peak Reactor Vessel Neutron Fluence Projections at 56 EFPY ................................... 9 Table 6: Adjusted Reference Temperature Calculations at 56 EFPY ........................................10 Table 7: RTNDT and RTPTS Calculations at 56 EFPY ................................................................ 11 Table 8: Beltline Reactor Vessel Materials USE Projection at 56 EFPY .................................... 11 iii

Pressure and Temperature Limits Report - Unit 4 List of Figures Figure 1: Reactor Coolant System Heatup Limitations (Maximum Heatup Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Errors) .......................... 5 Figure 2: Reactor Coolant System Cooldown Limitations (Maximum Cooldown Rate of 100°F/hr) Applicable to 54 EFPY (without Margins for Instrumentation Error)............. 6 iv

Pressure and Temperature Limits Report - Unit 4 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.4. The limiting condition for operation (LCO) 3.4.3, reactor coolant system (RCS) pressure and temperature (P/T) limits, also referred to as heatup and cooldown limit curves, are contained in this PTLR. Revisions to the PTLR shall be provided to the U.S. Nuclear Regulatory Commission (NRC) after issuance. Note that these PTLR limits are consistent with, and bounded by, the generic AP1000 PTLR limits

[18]. Reference 17 utilized site-specific material properties as inputs to validate that the P/T limit curves in Reference 18 remained bounding for VEGP Units 3 and 4.

This PTLR follows the guidelines set forth in U.S. NRC Generic Letter 96-03 [1].

2.0 OPERATING LIMITS 2.1 RCS PRESSURE AND TEMPERATURE LIMITS The RCS P/T limits for LCO 3.4.3, presented in Figures 1 and 2 and tabulated in Tables 1 and 2, were developed using the NRC-approved methodology in WCAP-14040-A [2].

The boltup temperature shall be 60°F. The minimum (allowable) boltup temperature is established as the higher of 60°F or the highest material reference temperature (initial RTNDT) in the highly-stressed reactor pressure vessel (RPV) flange region. The heatup and cooldown curves utilize 60°F, since this value is limiting for the respective vessel materials. This allows the plant to perform the boltup operations at 60°F or higher.

The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°F in any one hour period.

b. A maximum cooldown rate of 100°F in any one hour period.

c. A maximum temperature change of 10°F in any one-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

Page 1 of 13

Pressure and Temperature Limits Report - Unit 4 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 1 and 2. Data points for these figures are tabulated in Tables 1 and 2, respectively. The 54 effective full power years (EFPY) term of applicability identified for these limit curves is based on adjusted reference temperature (ART) calculations that utilize Regulatory Guide 1.99, Revision 2 [3] Position 1.1 chemistry factors and peak neutron fluence projections on the vessel forgings and beltline welds. The inputs for the ART calculations are contained in APP-RXS-M3C-026

[4] and APP-MV01-Z0-101 [5], and are summarized in Tables 3 through 5. Calculated ART values, used in the original development of the P/T limits, are documented in APP-RXS-M3C-012 [6]. These ART values are conservative for use as summarized in Table 6 and APP-RXS-M3C-033 [16]. The inputs and calculations are discussed in Section 4.0.

2.2 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM Low Temperature Overpressure Protection (LTOP) is required to provide overpressure protection when any RCS cold leg temperature is <275F with the reactor vessel head on as required by Technical Specifications and APP-RCS-M3-001 [13]. This temperature is above the minimum LTOP enable temperature calculated using ASME Code Case N-641 (See Reference [2]).

A method of LTOP of the reactor vessel is provided by two parallel relief valves in the Normal Residual Heat Removal System (RNS) suction line inside containment. The larger RNS pump suction line relief valve (RNS-PL-V021) has a lift setpoint of 500 psig with a full open pressure of 550 psig [14]. The smaller relief valve (RNS-PL-V020) has a lift setpoint of 470 psig with a full open pressure of 517 psig [15]. The set pressure of RNS-PL-V020 is lower than the set pressure for RNS-PL-V021 so that the RNS-PL-V020 valve lifts first, and it is intended to be the only relief valve to lift during pressure transients that can be relieved with its capacity.

The lift setpoints were determined using the methodology in WCAP-14040-A [2], and the validation of the adequate capacity and set pressures of the valves is in APP-RNS-M3C-002 [12].

The maximum allowable RNS pump suction line relief valve lift setpoints are derived by analysis of the LTOP design basis Mass Input (MI) and the Heat Input (HI) transients.

Operation with RNS pump suction line relief valves lift setpoints of less than or equal to the maximum allowable setpoint, ensures that the allowable steady state pressure temperature limit shown in Table 2 [6] will not be violated with consideration for: (1) relief valves set pressure tolerances and accumulations; and (2) effects of reactor coolant pump operation.

To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications and operating procedures place limitations on coolant input capabilities and RCP operation during the appropriate LTOP modes. The magnitude of the maximum mass injection rate is controlled by isolating accumulators Page 2 of 13

Pressure and Temperature Limits Report - Unit 4 and closing the Chemical and Volume Control System (CVS) makeup line containment isolation valve, CVS-PL-V091, to limit the mass injection rate to the relieving capacity of RNS-PL-V020. The magnitude of the maximum heat injection rate is also administratively controlled. To limit the heat injection rate, the reactor coolant pumps are started at less than 25% speed and the steam generator temperatures must be less than or equal to 50°F higher than each of the cold leg temperatures prior to pump start [12, 13]. The required relieving rate for this transient exceeds the capacity of RNS-PL-V020, and causes lifting of the higher capacity relief valve RNS-PL-V021. Both relief valves will lift and maintain the peak pressure below the limit in Table 2.

3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. Table 1 of ASTM E-185 [7]

identifies the requirement for four capsules to be withdrawn for a maximum projected transition temperature shift (RTNDT) of the beltline materials exceeding 100°F. The surveillance program withdrawal schedule for removal of the capsules for post-irradiation testing exceeds this requirement and identifies five capsules to be withdrawn.

The surveillance capsule withdrawal schedule and pressure vessel steel surveillance program is in compliance with Appendix H to 10CFR50, Reactor Vessel Material Surveillance Program Requirements [8]. Accordingly, the surveillance capsule withdrawal schedule meets the requirements of ASTM E185 [7], supplemented as needed for a 60-year design life. The results of these examinations shall be used to update the RCS P/T limits. The recommended surveillance capsule withdrawal schedule, shown below, is consistent with the AP1000 certified design.

Capsule Withdrawal Time 1st When the accumulated neutron fluence of the capsule is 5 x 1018 n/cm2.

2nd When the accumulated neutron fluence of the capsule corresponds to the approximate end of life fluence at the reactor vessel 1/4T location.

3rd When the accumulated neutron fluence of the capsule corresponds to the approximate end of life fluence at the reactor vessel inner wall location.

4th When the accumulated neutron fluence of the capsule corresponds to a fluence not less than once or greater than twice the peak end of vessel life fluence.

5th End of plant design objective of 60 years.

6th Standby 7th Standby 8th Standby Page 3 of 13

Pressure and Temperature Limits Report - Unit 4 4.0 SUPPLEMENTAL DATA TABLES Reference 17, Appendix A, provides less restrictive VEGP Units 3 and 4 specific updated tables that reflect the plant-specific material properties; however, the generic AP1000 tables remain bounding and are utilized as the approved plant-specific data tables.

Table 3 contains the elemental chemistry limits for the reactor vessel beltline materials, along with the Regulatory Guide 1.99, Revision 2 [3] Position 1.1 chemistry factors based on these limits. The chemistry values listed are the maximum allowed values and are therefore conservative for use in determining the effects of irradiation embrittlement for the beltline materials.

Table 4 contains the mechanical properties of the reactor vessel materials utilized in calculating the ARTs for the P/T limit curves, the preliminary pressurized thermal shock (PTS) reference temperature (RTPTS) values, and projection of upper shelf energy (USE) for the reactor vessel materials.

Table 5 contains the maximum projected neutron fluence values for the beltline materials of the reactor vessel at end-of-life (EOL), which was assumed to be 56 EFPY, along with the calculated fluence values at the 1/4T position (vessel wall quarter thickness from the inside surface) and 3/4T position (vessel wall quarter thickness from the outside surface).

Table 6 contains the calculations of the ARTs at the 1/4T position and 3/4T position, which are bounded by the ART values used in the determination of the P/T limit curves for normal operation [9, 10]. Note that the 56 EFPY ART values in Table 6 are lower than those utilized in the original development of the P/T limit curves, which are listed on Figures 1 and 2. The Figure 1 and 2 54 EFPY ARTs remain acceptable for use, since they bound the values in Table 6 for 56 EFPY.

Table 7 contains the calculations of RTNDT and RTPTS. As noted in Section 3, values of RTNDT exceeding 100°F require four surveillance capsules to be withdrawn in accordance with ASTM E185 [7]. The maximum RTNDT value is calculated to be 109.0°F (based on Position 1.1 Chemistry Factors). The screening criteria for PTS are provided in 10CFR50.61 [11], which states that the values of RTPTS (using EOL neutron fluence projections) must be less than 270°F for plates, forgings, and axial weld materials, and less than 300°F for circumferential weld materials. The preliminary RTPTS values are calculated to be 78.4°F for the beltline forgings and 145.0°F or less for the beltline circumferential welds, and are in compliance with 10CFR50.61 [11].

Table 8 contains the conservative projections of USE for the reactor vessel beltline materials to show compliance with the requirements of 10CFR50, Appendix G [9].

Page 4 of 13

Pressure and Temperature Limits Report - Unit 4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: SHELL FORGING BOUNDING ART VALUES AT 54 EFPY: 1/4T, 90°F 3/4T, 82°F 2500 I I I 2250 ILeak Test Limit I

' II 2000 J Acceptable ,-

Operation

~I I- J j J ff Unacceptable 1750 I

Operation I

YI I ,'

B' 1500 J en I Hea~up Rate

"--.. Critical Limit 50 Deg. F/Hr

...a, 50 Deg. F/Hr I

l 1250 I I

II) -..........___

II) a, ~ I Critical Limit.I Heatup Rater 100 Deg. F/Hr 0.. 100 Deg. F/Hr "C

a, 1000

l

.!::?

750

(.)

500 Criticality Limit based on "1Boltup inservice hydrostatic test 250 I Temp temperature (145 F) for the service period up to 54 EFP 1 0 !

no 400 , 4!0 51 ,o I

t 150 200 250 300 350 5 0

-250 51° I I I I I I I Moderator Temperature (Deg. F)

Figure 1: Reactor Coolant System Heatup Limitations (Maximum Heatup Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Errors)

(Plotted Data provided on Table 1)

Page 5 of 13

Pressure and Temperature Limits Report - Unit 4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: SHELL FORGING BOUNDING ART VALUES AT 54 EFPY: 1/4T, 90°F 3/4T, 82°F 2500 I

2250 2000 17 50 J

I c, 1500 u;

!=.

= 1250 (I)

-IUnacceptable Operation (I) 0.

1000 I

'O a,

=

Cooldo"'1 Acceptable Rates Operation

(.) F/Hr

(.)

750 steady-state

-20

-40 I

-60

-100 500 Boltup Temo 250 0 ' .-, ,-.,---, I 1 ,--r-51,0 I fl I 1- 1 .- , ,-----,- 1 I t I I - 1 1 I I 1------,---r , 1- 1 11 - I I I 11)0 1h0 200 21,0 310 31,0 410 450 51° 51°

-250 I .

Mod erator Temperature (Deg. F)

Figure 2: Reactor Coolant System Cooldown Limitations (Maximum Cooldown Rate of 100°F/hr)

Applicable to 54 EFPY (without Margins for Instrumentation Error)

(Plotted Data provided on Table 2)

Page 6 of 13

Pressure and Temperature Limits Report - Unit 4 Table 1: RCS Heatup Limits at 54 EFPY (without uncertainties for instrumentation errors) 50°F/hr Heatup Critical Limit 100°F/hr Heatup Critical Limit Leak Test Limit T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 145 -14.7 60 -14.7 145 -14.7 127 2,000 60 621 145 621 60 621 145 621 127 2,000 65 621 145 621 65 621 145 621 145 2,485 70 621 145 621 70 621 145 621 145 2,485 75 621 145 621 75 621 145 621 80 621 145 621 80 621 145 621 85 621 145 621 85 621 145 621 90 621 145 621 90 621 145 621 95 621 145 621 95 621 145 621 100 621 145 621 100 621 145 621 105 621 150 621 105 621 150 621 110 621 155 621 110 621 155 621 115 621 160 621 115 621 160 621 120 621 165 621 120 621 165 621 125 621 170 621 125 621 170 621 130 621 170 1,268 130 621 170 1,004 130 621 175 1,337 130 621 175 1,042 130 1,268 180 1,413 130 1,004 180 1,085 135 1,337 185 1,499 135 1,042 185 1,134 140 1,413 190 1,593 140 1,085 190 1,188 145 1,499 195 1,697 145 1,134 195 1,249 150 1,593 200 1,812 150 1,188 200 1,318 155 1,697 205 1,940 155 1,249 205 1,394 160 1,812 210 2,081 160 1,318 210 1,478 165 1,940 215 2,236 165 1,394 215 1,572 170 2,081 220 2,408 170 1,478 220 1,675 175 2,236 175 1,572 225 1,790 180 2,408 180 1,675 230 1,917 185 1,790 235 2,058 190 1,917 240 2,213 195 2,058 245 2,384 200 2,213 205 2,384 Page 7 of 13

Pressure and Temperature Limits Report - Unit 4 Table 2: RCS Cooldown Limits at 54 EFPY (without uncertainties for instrumentation errors)

Steady State 20°F/hr 40°F/hr 60°F/hr 100°F/hr T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.7 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 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 621 130 1,557 130 1,557 130 1,557 130 1,557 130 1,557 135 1,653 135 1,653 135 1,653 135 1,653 135 1,653 140 1,758 140 1,758 140 1,758 140 1,758 140 1,758 145 1,874 145 1,874 145 1,874 145 1,874 145 1,874 150 2,003 150 2,003 150 2,003 150 2,003 150 2,003 155 2,145 155 2,145 155 2,145 155 2,145 155 2,145 160 2,302 160 2,302 160 2,302 160 2,302 160 2,302 165 2,476 165 2,476 165 2,476 165 2,476 165 2,476 Page 8 of 13

Pressure and Temperature Limits Report - Unit 4 Table 3: Maximum Composition Limits for Reactor Vessel Beltline Materials Elements Maximum Weight %