ML24100A121

Submittal of Revised Pressure and Temperature Limits Report

ML24100A121
Person / Time
Site:	Prairie Island
Issue date:	04/09/2024
From:	Conboy T Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-Pl-24-013
Download: ML24100A121 (1)

Text

April 9, 2024 (l Xcel Energy ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 1717 Wakonade Drive Welch, MN 55089 L-Pl-24-013 TS 5.6.6.c Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and 2 Revised Pressure and Temperature Limits Report Pursuant to the requirements of Technical Specification 5.6.6.c, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"),

hereby submits the Prairie Island Nuclear Generating Plant Units 1 and 2 Pressure and Temperature Limits Report {PTLR) Revision 9.

Please contact Mr. Jeff Kivi at (612) 330-5788 or Jeffrey.L.Kivi@xcelenergy.com if there are any questions or if additional information is needed.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

T2s~?45 <:

Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

ENCLOSURE PRAIRIE ISLAND NUCLEAR GENERATING PLANT PRESSURE AND TEMPERATURE LIMITS REPORT UNITS 1 AND 2 REVISION 9 27 pages follow

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pressure and Temperature Limits Report Page 1 of 27 RECORD OF REVISION Revision No.

Approval Date Remarks 0

5/5/98 Original Pressure and Temperature Limits Report. Issued after May 4, 1998, approval of License Amendment Request dated March 6, 1998, as Amendment 135/127. Distribution with Technical Specification Revision 135.

1 4/6/00 Revised discussion of surveillance data credibility.

Revisions to References 5.6 and 5.7 identified which incorporate findings from Comprehensive Revised Response to GL 92-01.

Revised Table 6.5 data to reflect the data incorporated into the updated References 5.6 and 5.7.

Changed title for the operating limit Temperature for Disabling both Safety Injection Pumps to the terminology Safety Injection (SI) Pump Disable Temperature in preparation for ITS.

Changed titles of Table 6.1 and 6.2 to match Table of Contents.

Changed titles of Figures 6.1 and 6.2 in the Table of Contents to match the titles on the figures.

Distributed with Technical Specification Revision 153.

2 10/12/2002 This revision makes the PTLR consistent with the license amendments 158/149.

Details:

Revised Table of Contents to reflect changes in page numbering and the addition of 2 new subsections in section 3: Pressurizer Temperature Limits and Steam Generator Temperature/Pressure Limit.

Revised the wording in section 1.0 to be consistent with license amendments 158/149 5.6.6 as to the items contained in the PTLR document.

Revised the list of Technical Specifications LCO/SRs that reference items in the PTLR. These changes are due to the different license amendments 158/149 numbering.

Revised the Referenced in portions of the section 3.0 subparagraphs to reflect the different license amendments 158/149 numbering.

Added subsection Pressurizer Temperature Limits to section 3.0.

These limits were moved to the PTLR for license amendments 158/149 versus an LCO in the previous Technical Specification.

Added subsection Steam Generator Temperature/Pressure Limit to section 3.0. This limit was moved to the PTLR for license amendments 158/149 versus an LCO in the previous Technical Specification revisions.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pressure and Temperature Limits Report Page 2 of 27 RECORD OF REVISION Revision No.

Approval Date Remarks 3

10/21/2002 This revision corrects errors in references to the TRM, a transcription error for the Maximum Pressurizer Cooldown rate and adds Site Director of Engineering as approver.

Details:

On cover page, added Approval of Site Director of Engineering.

On page 1, changed TRM reference 3.10.1 to TRM references 3.4.4 and 3.4.5.

On page 4, subsection Pressurizer Temperature Limits, revised the Maximum Pressurizer Cooldown Rate to 200°F per hour versus 100°F per hour, to correct a transcription error in the last revision.

On page 4, subsection Pressurizer Temperature Limits, revised Referenced in specification to TRM 3.4.4 to make the PTLR consistent with the TRM.

On page 4, subsection Steam Generator Temperature/Pressure Limit, revised the Referenced in specification to TRM 3.4.5 to make the PTLR consistent with the TRM.

4 3/30/2015 Updated section 4.0 discussion of ART, RTPTS, fluence, and CF with values applicable to 54 EFPY. Throughout, changed the effective until from 35 EFPY to 54 EFPY.

Updated tables 6.4 and 6.5 with ART values at 54 EFPY.

Added references supporting change to effective until 54 EFPY.

5 8/5/2015 Remove reference to TS section 3.5.3 in section 3.0 of the PTLR which discusses SI pump disable temperature. Refer to License Amendment 213/201.

6 8/2/2019 Update Table 6.3 for newly approved material surveillance removal schedule. Refer to NRC Safety Evaluation of 7/3/19. (ML19177A380) 7 9/17/2020 Update Table 6.3 for newly approved material surveillance removal schedule. Refer to NRC Safety Evaluation of 9/3/20. (ML20230A051) 8 2/16/2021 Revises PTLR Section 3.0 to fix reference errors and more clearly organize as a result of issues identified in CAP QIM 501000046600.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pressure and Temperature Limits Report Page 3 of 27 RECORD OF REVISION Revision No.

Approval Date Remarks 9

2/28/24 Updated analytical limit temperature and referenced in Section 3.4.1 Overpressure Protection Setpoints.

Updated end of life fluence reference temperature, neutron fluences, Chemistry Factors, limiting ARTS, number of surveillance capsules removed, and references in Section 4.0.

Updated references to delete references that are replaced with newer documentation and added those references in Section 5.0.

Revised Tables 6.1 and 6.2 to provide the newest heatup and cooldown data points.

Updated Table 6.3 with the newest data generated for lead factors, withdrawal EFPY, and fluence values. Updated notes as applicable.

Revised Tables 6.4 and 6.5 with the newest Chemistry Factors, 1/4, 3/4 delta RT calculations and associated references with the data.

Revised Figure 6.1 and Figure 6.2 with the newest the heatup and cooldown limit curves.

Added Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results. This Table was moved from Section 04 of the USAR.

References were alphabetized and renumbered, so reference numbers noted in Remarks for previous Revisions should be reviewed against that revision of the PTLR.

Page 4 of 27 Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Approval:

604000001090

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 5 of 27 Table of Contents Section Title Page 1.0 PURPOSE.................................................................................................................... 6 2.0 APPLICABILITY........................................................................................................... 6 3.0 OPERATING LIMITS.................................................................................................... 7 4.0 DISCUSSION............................................................................................................... 9

5.0 REFERENCES

........................................................................................................... 13 6.0 ATTACHMENTS......................................................................................................... 14 List of Tables Table 6.1 54 EFPY Heatup Data Points..................................................................................... 15 Table 6.2 54 EFPY Cooldown Data Points................................................................................ 17 Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule........................... 19 Table 6.4 Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 54 EFPY............................ 20 Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 54 EFPY............................ 21 Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results.......................................................................... 22 List of Figures Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 54 EFPY................................................................................................................. 26 Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 54 EFPY................................................................................................................. 27

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 6 of 27 1.0 PURPOSE The purpose of the Prairie Island Nuclear Generating Station Pressure and Temperature Limits Report (PTLR) is to present operating limits for Units 1 and 2 relating to; (1) RCS pressure and temperature during Heatup, Cooldown and low temperature operation; (2) RCS heatup and cooldown rates; (3) the Over Pressure Protection System (OPPS) arming temperature; (4) OPPS lift settings; (5) Safety Injection Pump disable temperature as well as (6) thermal stress related temperature limitations for the pressurizer and steam generators. This report has been prepared in accordance with the requirements with Technical Specification 5.6.6.

2.0 APPLICABILITY This report is applicable to both Units 1 and 2 until 54 Effective Full Power Years (EFPY) is reached on that particular units Reactor Pressure Vessel. The Technical Specifications that are affected by the information contained in this report are:

TS 3.4.3 RCS Pressure and Temperature (P/T) Limits TS 3.4.6 RCS Loops - MODE 4 TS 3.4.7 RCS Loops - MODE 5, Loops Filled TS 3.4.10 Pressurizer Safety Valves TS 3.4.12 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature TS 3.4.13 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) Safety Injection (SI) Pump Disable Temperature TRM 3.4.4 PTLR Compliance - Pressurizer TRM 3.4.5 PTLR Compliance - Steam Generator(s)

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 7 of 27 3.0 OPERATING LIMITS All limits are valid until 54 EFPY, which is projected to be beyond the expiration of the operating license for each of Prairie Island Units 1 and 2.

3.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 3.1.1 RCS P/T Limits Figure 6.1* RCS P/T limits for heatup Figure 6.2* RCS P/T limits for cooldown

• Figures are analytical limits and do not include instrumentation uncertainty.

NOTE:

Table 6.1 and Table 6.2 contain a tabulated version of the curves.

Instrumentation Uncertainty for P/T Curves 124 psig Pressure Uncertainty 18 °F Temperature Uncertainty NOTE:

These values must be applied to the P/T limit curves in operating procedures (Reference 5.2 and 5.3).

3.1.2 RCS Heatup/Cooldown Rate Limits 100 °F per hour Maximum RCS Heatup Rate 100 °F per hour Maximum RCS Cooldown Rate 3.1.3 Minimum Boltup Temperature 60 °F**

• • No instrument uncertainty included.

3.2 Pressurizer Temperature Limits (TRM 3.4.4) 100 °F per hour Maximum Pressurizer Heatup Rate 200 °F per hour Maximum Pressurizer Cooldown Rate 320 °F Maximum Temperature Difference Between the Pressurizer and the Spray Fluid for which the Pressurizer Spray can be used

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 8 of 27 3.3 Steam Generator Temperature Pressure Limit (TRM 3.4.5) 200 psig Maximum secondary side Pressure if the temperature of the steam generator is below 70°F.

3.4 Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12 and 3.4.13)

Referenced in:

TS 3.4.6 TS 3.4.7 TS 3.4.10 TS 3.4.12 TS 3.4.13 SR 3.4.12.4 SR 3.4.13.5 3.4.1 Over Pressure Protection System (OPPS) Enable Temperature (LCO 3.4.12) 290 °F*

• Analytical limit [206 °F] plus indicating instrument channel uncertainty [18 °F] (Reference 5.3) plus additional margin for operational simplicity.

3.4.2 Safety Injection (SI) Pump Disable Temperature (LCO 3.4.12 and 3.4.13) 218 °F *

• Analytical limit [200 °F] plus indicating instrument channel uncertainty [18 °F] (Reference 5.3).

3.4.3 Over Pressure Protection System (OPPS) PORV Setpoint (LCO 3.4.12 and 3.4.13) 500 psig*

• This setpoint accounts for instrument channel uncertainty (Reference 5.4).

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 9 of 27 4.0 DISCUSSION This PTLR for Prairie Island Units 1 and 2 has been prepared in accordance with the requirements contained in Technical Specification 5.6.6. Periodic adjustments to the curves, limits and setpoints based on new irradiation fluences of the reactor vessel or changes in instrument uncertainty can be made under the conditions of 10CFR50.59, with the updated PTLR submitted to the NRC upon issuance.

Changes to the curves, limits, setpoints or parameters in the PTLR resulting from new or additional analysis of either beltline or weld material properties (e.g.

additional capsule data) must be submitted to the NRC prior to issuance of an updated PTLR.

The results of the analysis of the Units 1 and 2 reactor vessel material surveillance capsule tests show that the limitations for Unit 1 are the most restrictive and conservative. For simplicity these results have been applied to both units.

The following parameters were used in the development of the curves, limits, and setpoints given in section 3.0 of this report. These values were obtained from Prairie Island Units 1 and 2 Reactor Vessel Radiation Surveillance Program Data. The surveillance program capsules were removed as indicated in Table 6.3.

Adjusted Reference Temperature (ART)

The adjusted reference temperature is the reference temperature (as defined in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G for Nil-ductility transition) that has been adjusted for radiation effects. This temperature was determined for all beltline materials for both Prairie Island Units 1 and 2 at the 1/4T and 3/4T thicknesses from the reactor vessel clad/base metal interface radius, where T is the reactor vessel thickness. Comparison of ARTs for all materials shows that the limiting material at 54 EFPY is the Unit 1 intermediate to lower shell forging circumferential weld material. Since this material is a circumferential flaw material, the membrane stress and stress intensity factors are lower than the most limiting material with an axially oriented flaw. Therefore the limiting material with an axially oriented flaw, which is Unit 1 intermediate shell forging C, was determined by analysis to be controlling and was used to develop the curves shown in Tables 6.1 and 6.2 and Figure 6.1 and Figure 6.2. The limiting ARTs at 54 EFPY for this material are 140.7°F for 1/4T, and 126.9°F for 3/4T. However, for added conservatism, the ARTs used to generate the heatup and cooldown limits were 170oF for 1/4T and 160oF for 3/4T.

The Heatup and Cooldown limitations and curves were recalculated in Reference 5.8.

Reference:

5.8

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 10 of 27 End of Life Fluence Reference Temperature (RTpts)

The RTpts reference temperature is the end of life reference temperature determined at the clad/base metal interface radius of the reactor vessel and adjusted for radiation effect to the projected end of plant life. The reference temperature has been obtained for all beltline materials in both Prairie Island Units 1 and 2. The projected end of life for both units is 54 Effective Full Power Years (54 EFPY).

Comparison of RTpts for all materials indicates that the limiting material is the Unit 1 intermediate shell to lower shell circumferential weld material. The limiting RTpts is as follows:

RTpts = 183.9 °F

Reference:

5.8 Neutron Fluences (f)

The ARTs are determined, in part, based on neutron fluence that is determined by using analytical techniques and passive neutron flux monitoring devices included within the Reactor Vessel Material Surveillance Program. Neutron fluence is determined for the present and future condition of the reactor vessel. Unit 1 intermediate shell forging neutron fluences used in determining the 54 EFPY limiting axial flaw material ART for the reactor vessels are as follows:

Units are 1019 n/cm2, for energies > 1.0 MeV at 54 EFPY Clad/Base Metal Interface = 5.63 1/4T = 3.77 3/4T = 1.69

Reference:

5.8 NOTE:

These values are not the highest fluences that were obtained in the reactor vessels, but are the values determined for the most limiting material. The highest fluences were obtained at the unit 2 intermediate shell forging.

(Reference 5.8).

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 11 of 27 Chemistry Factor (CF)

Chemistry Factors are parameters used in the development of the ARTs for the beltline materials and account for the Copper and Nickel content in the reactor vessel beltline materials. The chemistry factors determined for the limiting ARTs, corresponding to the Unit 1 intermediate to lower shell circumferential weld, are as follows.

1/4T = 99.1 °F 3/4T = 99.1 °F The chemistry factors for the limiting material for which an axially oriented flaw is considered, and which is bounding for determining the pressure-temperature limits corresponding to the Unit 1 Intermediate Shell Forging C are:

1/4T = 69.0 °F 3/4T = 69.0 °F

Reference:

5.8 Reactor Vessel Material Surveillance Program The Reactor Vessel Material Surveillance Program is described in the USAR (Reference 5.5). The schedule for removal of the Units 1 and 2 capsules is contained in Table 6.3 of this report.

References:

5.5 5.7, 5.9, Supplemental Data Tables Table 6.4 and Table 6.5 contain the development of all of the ARTs for the beltline materials for Unit 1 and Unit 2 respectfully, including all the parameters.

References:

5.8

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 12 of 27 Surveillance Data Credibility The credibility of surveillance capsule data is determined as specified in Regulatory Guide 1.99, Revision 2, Section B. Five radiation surveillance capsules have been removed from each of the Prairie Island Reactor Vessels, as shown in Table 6.3, and the credibility of these capsule data is analyzed in references 5.7 and 5.9. The credibility of the surveillance data effects how it is applied in the development of the materials ARTs.

When two or more credible surveillance data sets become available, the data sets may be used to determine the ART values as described in Regulatory Guide 1.99, Revision 2, Position 2.1. If the ART values based on surveillance capsule data are larger than those calculated per Regulatory Guide 1.99, Revision 2, Position 1.1, the surveillance data must be used. If the surveillance capsule data gives lower values, either may be used. In the case of the Prairie Island limiting axial flaw material, the Unit 1 intermediate shell forging surveillance data was available but considered non-credible. This resulted in the use of the full margin of 17F. The ART calculated using surveillance capsule data is larger than that calculated using position 1.1. For comparison Table 6.4 and Table 6.5 contains the ARTs for all those materials in the surveillance programs using both Regulatory Guide 1.99, Revision 2, development methods: Position 1.1 and Position 2.1.

Minimum Boltup Temperature The Minimum Boltup Temperature is the minimum temperature of the reactor vessel flange metal required any time reactor vessel flange is under tensioning stress.

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 13 of 27

5.0 REFERENCES

5.1 "Pressure Temperature Limits," Chapter 5.3.2 of the Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

5.2 NSP Calculation No. SPCRC002, Unit 1 Reactor Coolant Hot Leg Pressure Control Room Indication at 1PR-420 (0-750 psig scale) with 2 RC Pumps Running, Revision 0A.

5.3 NSP Calculation No. SPCRC003, Unit 1 Wide Range RCS Cold Leg Temperature Control Room Indication Loop 1T-450B Uncertainty with Streaming Effects, Revision 0A.

5.4 NSPM-LTP-TR-AA-000001-P, "Prairie Island Units 1 and 2 Low Temperature Overpressure Protection System (LTOPS) Analysis Revision 1, August 2, 2022.

5.5 USAR Section 4.7.2, Reactor Vessel Material Surveillance Program 5.6 WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 4, May 2004.

5.7 WCAP-18660-NP, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, Revision 0, November 2021.

5.8 WCAP-18746-NP, "Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, Revision 2, January 2023.

5.9 WCAP-18795-NP, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, Revision 0, December 2022.

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 14 of 27 6.0 ATTACHMENTS 6.1 Table 6.1 - 54 EFPY Heatup Data Points 6.2 Table 6.2 - 54 EFPY Cooldown Data Points 6.3 Table 6.3 - Reactor Vessel Material Surveillance Capsule Removal Schedule.

6.4 Table 6.4 - Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 54 EFPY 6.5 Table 6.5 - Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 54 EFPY 6.6 Figure 6.1 - Prairie Island Reactor Coolant System Heatup Limitations Applicable to 54 EFPY.

6.7 Figure 6.2 - Prairie Island Reactor Coolant System Cooldown Limitations Applicable to 54 EFPY.

6.8 Table 6.6 - Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 15 of 27 Table 6.1 54 EFPY Heatup Data Points (Without Instrumentation Uncertainty Margins) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 60

-14.7 219

-14.7 60

-14.7 219

-14.7 60 621 219 1101 60 621 219 954 65 621 220 1111 65 621 220 961 70 621 225 1159 70 621 225 998 75 621 230 1212 75 621 230 1039 80 621 235 1270 80 621 235 1084 85 621 240 1335 85 621 240 1134 90 621 245 1406 90 621 245 1189 95 621 250 1485 95 621 250 1250 100 621 255 1572 100 621 255 1317 105 621 260 1668 105 621 260 1391 110 621 265 1774 110 621 265 1472 115 621 270 1892 115 621 270 1563 116 621 275 2021 116 621 275 1662 116 783 280 2164 116 721 280 1772 120 793 285 2322 120 727 285 1894 125 807 125 736 290 2028 130 823 130 747 295 2175 135 840 135 759 300 2338 140 859 140 772 145 881 145 788 150 905 150 805 155 931 155 825 160 960 160 847 165 992 165 871 170 1028 170 898 175 1067 175 928 180 1111 180 961 185 1159 185 998 190 1212 190 1039 195 1270 195 1084 200 1335 200 1134 205 1406 205 1189

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 16 of 27 Table 6.1 54 EFPY Heatup Data Points (Without Instrumentation Uncertainty Margins) (CONTD) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 210 1485 210 1250 215 1572 215 1317 220 1668 220 1391 225 1774 225 1472 230 1892 230 1563 235 2021 235 1662 240 2164 240 1772 245 2322 245 1894 250 2028 255 2175 260 2338 Leak Test Limit T (°F)

P (psig) 200 2000 219 2485

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 17 of 27 Table 6.2 54 EFPY Cooldown Data Points (Without Margins for Instrumentation Uncertainty)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown 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 60 601 65 621 65 621 65 621 65 621 65 607 70 621 70 621 70 621 70 621 70 614 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 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 116 621 116 621 116 621 116 621 116 621 116 849 116 825 116 800 116 777 116 729 120 861 120 838 120 814 120 791 120 745 125 878 125 856 125 833 125 811 125 768 130 897 130 875 130 854 130 833 130 793 135 918 135 897 135 877 135 857 135 820 140 941 140 921 140 903 140 885 140 851 145 966 145 948 145 931 145 915 145 885 150 994 150 978 150 962 150 948 150 922 155 1025 155 1010 155 997 155 985 155 964 160 1059 160 1046 160 1035 160 1026 160 1010 165 1096 165 1086 165 1078 165 1071 165 1061 170 1138 170 1131 170 1125 170 1121 170 1118 175 1184 175 1179 175 1177 175 1176 175 1176 180 1235 180 1233 180 1233 180 1233 180 1233 185 1291 185 1291 185 1291 185 1291 185 1291 190 1353 190 1353 190 1353 190 1353 190 1353 195 1422 195 1422 195 1422 195 1422 195 1422

~-~-------r-----------.---------r-------t I

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 18 of 27 Table 6.2 54 EFPY Cooldown Data Points (Without Margins for Instrumentation Uncertainty) (CONTD)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 200 1498 200 1498 200 1498 200 1498 200 1498 205 1581 205 1581 205 1581 205 1581 205 1581 210 1674 210 1674 210 1674 210 1674 210 1674 215 1777 215 1777 215 1777 215 1777 215 1777 220 1890 220 1890 220 1890 220 1890 220 1890 225 2015 225 2015 225 2015 225 2015 225 2015 230 2153 230 2153 230 2153 230 2153 230 2153 235 2306 235 2306 235 2306 235 2306 235 2306 240 2475 240 2475 240 2475 240 2475 240 2475

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 19 of 27 Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule Surveillance Capsule Removal Schedule for Unit 1 Capsule Capsule Location (degree)

Lead Factor(a)

Withdrawal EFPY(b)

Fluence(a)

(n/cm2, E> 1.0 MeV)

V 77 2.98 1.40 6.09 x 1018 P

247 1.76 4.91 1.31 x 1019 R

257 3.08 8.63 4.02 x 1019 S

57 1.86 18.12 4.39 x 1019 N

237 1.97 40.06 8.45 x 1019 T

67 2.05 Standby Surveillance Capsule Removal Schedule for Unit 2 Capsule Capsule Location (degree)

Lead Factor(d)

Withdrawal EFPY(b)

Fluence(d)

(n/cm2, E> 1.0 MeV)

V 77 3.03 1.39 5.98 x 1018 T

67 1.80 4.13 1.10 x 1019 R

257 3.08 8.80 4.11 x 1019 P

247 1.88 17.24 4.27 x 1019 N

237 1.94 40.64 8.41 x 1019 S

57 1.95 Standby Notes:

(a) Updated in Capsule N dosimetry analysis.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Not used.

(d) Updated in Capsule N dosimetry analysis.

(e) Not used.

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 20 of 27 Table 6.4 Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 54 EFPY Material CF f @ 54 EFPY(a) 1/4T f 3/4T f 1/4T FF(d) 3/4T FF I(e)

(°F)

M(g)

(°F)

RTNDT

(°F)

ART

(°F) 1/4T Calculations Upper Shell Forging B 51 3.37 2.26 1.220

-4 34 62.2 92.2 Upper to Inter. Shell Circ Weld W2 79.5 3.63 2.43 1.239 0(c) 65.5 98.5 164.0 Intermediate Shell Forging C Using S/C Data 44 69.0 5.63 5.63 3.77 3.77 1.343 1.343 14 14 34 34(b) 59.1 92.7 107.1 140.7 Inter. to Lower Shell Weld W3 Using S/C Data 69.7 99.1 5.53 5.53 3.70 3.70 1.339 1.339

-13

-13 56 56(b) 93.3 132.7 136.3 175.7 Lower Shell Forging D 44.0 5.53 3.70 1.339

-4 34 58.9 88.9 3/4T Calculations Upper Shell Forging B 51 3.37 1.01 1.003

-4 34 51.1 81.1 Upper to Inter. Shell Circ Weld W2 79.5 3.63 1.09 1.024 0(c) 65.5 81.4 146.9 Intermediate Shell Forging C Using S/C Data 44.0 69.0 5.63 5.63 1.69 1.69 1.144 1.144 14 14 34 34(b) 50.3 78.9 98.3 126.9 Inter. to Lower Shell Weld W3 Using S/C Data 69.7 99.1 5.53 5.53 1.66 1.66 1.139 1.139

-13

-13 56 56(b) 79.4 112.9 122.4 155.9 Lower Shell Forging D 44.0 5.53 1.66 1.139

-4 34 50.1 80.1 NOTE:

(a)

Fluence values (f) are x 1019 n/cm2 (E > 1.0 MeV). (Ref. 5.8)

(b)

The full margin of 17°F for the forging and 28°F for the weld was used since the surv. data was deemed not credible (Ref. 5.8).

(c)

Estimated per Standard Review Plan Section 5.3.2 (Ref. 5.1).

(d)

FF, Fluence Factor = f(0.28-0.1*logf) (Ref. 5.8)

(e)

I is the unirradiated material reference temperature. (Ref. 5.8)

(g)

M is a margin term required for conservative results. (Ref. 5.8)

I I

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 21 of 27 Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 54 EFPY Material CF f @ 54 EFPY 1/4T f 3/4T f(a) 1/4T FF(c) 3/4T FF I(d)

(°F)

M(e)

(°F)

RTNDT

(°F)

ART

(°F) 1/4T Calculations Upper Shell Forging B 44.0 3.37 2.26 1.220

-13 34 53.7 74.7 Upper to Inter. Shell Weld W2 Using Unit 1 S/C Data(b) 69.7 101.4 3.61 3.61 2.42 2.42 1.238 1.238

-13

-13 56 56 86.3 125.5 129.3 168.5 Intermediate Shell Forging C

44.0 5.66 3.79 1.344 14 34 59.2 107.2 Inter. to Lower Shell Weld W3 Using S/C Data 51.6 87.5 5.58 5.58 3.73 3.73 1.341 1.341

-31

-31 56 28 69.2 117.3 94.2 114.3 Lower Shell Forging D Using S/C Data 51.0 74.4 5.58 5.58 3.73 3.73 1.341 1.341

-4

-4 34 34(f) 68.4 99.8 98.4 129.8 3/4T Calculations Upper Shell Forging B 44.0 3.37 1.01 1.003

-13 34 44.1 65.1 Upper to Inter. Shell Weld W2 Using Unit 1 S/C Data(b) 69.7 101.4 3.61 3.61 1.08 1.08 1.022 1.022

-13

-13 56 56 71.2 103.6 114.2 146.6 Intermediate Shell Forging C

44.0 5.66 1.70 1.146 14 34 50.4 98.4 Inter. to Lower Shell Weld W3 Using S/C Data 51.6 87.5 5.58 5.58 1.67 1.67 1.142 1.142

-31

-31 56 28 58.9 99.9 83.9 96.9 Lower Shell Forging D Using S/C Data 51.0 74.4 5.58 5.58 1.67 1.67 1.142 1.142

-4

-4 34 34(f) 58.2 85.0 88.2 115.0 NOTE:

(a) Fluence values (f) are x 1019 n/cm2 (E > 1.0 MeV). (Ref. 5.8)

(b) This calculation is using the chemistry factor based on the surveillance capsule data for the Prairie Island Unit 1 surveillance program. Per WCAP-14779 Rev. 1, the surveillance weld data is not credible, therefore, a full of 28°F was used in the margin term.

(c)

FF, Fluence Factor = f(0.28-0.1*logf). (Ref. 5.9)

(d) I is the unirradiated material reference temperature. (Ref. 5.8)

(e) M is a margin term required for conservative results. (Ref. 5.8)

(f)

The full margin of 17°F for the forging was used since the surveillance data was deemed not credible (Ref. 5.8).

I I I

I

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 22 of 27 Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results UNIT 1 - 54 EFPY Component Initial RTNDT (*F)

Source of RTNDT M (*F)

Wt % Cu WT % Ni CF (*F)

Fluence [a]

(n/cm2)

RTPTS (*F)

Nozzle Shell Forging

-4 Measured 34 0.08 0.68 51 3.37E+19 97.2 Nozzle to Intermediate Shell Weld (W2) 0 Generic 65.5 0.15 0.15 79.5 3.63E+19 171.6 Intermediate Shell Forging 14 Measured 34 0.07 0.80 44 5.63E+19 110.7 Intermediate Shell Forging Using Surveillance Data 14 Measured 34 NA[b]

NA[b]

69.0 5.63E+19 146.3 Intermediate to Lower Shell Weld (W3)

-13 Measured 56 0.13 0.13 69.7 5.53E+19 142.1 Intermediate to Lower Shell Weld (W3) Using Surveillance Data

-13 Measured 56 NA[b]

NA[b]

99.1 5.53E+19 183.9 Lower Shell Forging

-4 Measured 34 0.07 0.66 44 5.53E+19 92.6 Note: The data in this table comes from References 5.8 with fluence determined by the methodology of Reference 5.6.

[a]

Peak clad/base metal interface fluence.

[b]

Value is derived from the material specific chemistry factors - elemental percentages not used.

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 23 of 27 Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results (CONTD)

UNIT 2 - 54 EFPY Component Initial RTNDT (*F)

Source of RTNDT M (*F)

Wt % Cu WT % Ni CF (*F)

Fluence [a]

(n/cm2)

RTPTS (*F)

Nozzle Shell Forging {B}

-13 Measured 34 0.07 0.73 44 3.37E+19 79.0 Nozzle to Intermediate Shell Weld (W2)

-13 Measured 56 0.13 0.13 69.7 3.61E+19 136.0 Nozzle to Intermediate[c]

Shell Weld (W2) Using Surveillance Data

-13 Measured 56 NA[b]

NA[b]

101.4 3.61E+19 178.2 Intermediate Shell Forging 14 Measured 34 0.07 0.75 44 5.66E+19 110.7 Intermediate to Lower Shell Weld (W3)

-31 Measured 56 0.09 0.11 51.6 5.58E+19 98.4 Intermediate to Lower Shell Weld (W3) Using Surveillance Data

-31 Measured 28 NA[b]

NA[b]

87.5 5.58E+19 121.5 Lower Shell Forging

-4 Measured 34 0.08 0.67 51 5.58E+19 102.6 Lower Shell Forging Using Surveillance Data

-4 Measured 34 NA[b]

NA[b]

74.4 5.58E+19 135.9 Note: The data in this table comes from References 5.8 with fluence determined by the methodology of Reference 5.6.

[a]

Peak clad/base metal interface fluence.

[b]

Value is derived from the material specific chemistry factors - elemental percentages not used.

[c]

Data from Unit 1 W3 surveillance weld (same weld wire heat #1752) with the exception of fluence.

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 24 of 27 Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results (CONTD)

NOTES:

( )

PTS u

NDT PTS RT M

RT RT

+

+

=

and 2

2 u

2 M

+

=

and

(

)

Logf 10 0

28 0

PTS f

CF RT

=

Where:

( )

u NDT RT

= The initial reference temperature (RTNDT) of the unirradiated material.

M

= The margin to be added to cover uncertainties in the values of RTNDT, copper and nickel contents, fluence and the calculational procedures. M is evaluated per the equation above.

u

= The standard deviation for RTNDT(u) and is 0F for measured values and 17F for generic values.

= The standard deviation for RTPTS. The value for forgings is 17F when surveillance data is not used or is used but is not credible and 8.5F when surveillance data is used and is credible. The value for weld material is 28F when surveillance data is not used or is used but is not credible and is 14F when surveillance data is used and is credible. (Where Credibility is defined in 50CFR50.61.)

Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Page 25 of 27 Table 6.6 Pressurized Thermal Shock Reference Temperature Input Parameters and Calculation Results (CONTD)

NOTES:

PTS RT

= The mean value of the adjustment in the Pressurized Thermal Shock reference temperature caused by irradiation.

CF

= The chemistry factor, a function of copper and nickel content of the metal defined in Tables 1 and 2 of 10CFR50.61. Alternately, with credible surveillance data, the chemistry factor is defined by the following formula:

(

)

=

i i

0.20logf 56 0

i 0.10logf 0.28 i

i f

f A

CF f

= The best estimate neutron fluence in units 1019 n/cm2 (E greater than 1 MeV). The fluence values utilized in the RTPTS calculations were conversely based on the calculated maximum fast neutron flux at the reactor vessel clad-base metal interface. The calculated maximum fast neutron flux was obtained from References 5.7 and 5.9.

, Where Ai is the measured value of RTPTS for each surveillance data point

Figure 6.1 Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Prairie Island Reactor Coolant System Heatup Limitations Applicable to 54 EFPY (Without Margins for Instrument Uncertainty)

M.. *\\IERIALPROPERTYBASIS LIMITING 11.>ITERIAL: Unit 1, Intermediate Sh.ell Forging C LIMITING.ART VALUES AT 54 EFPY:

l/4T, 170"F (A."Wll Flaw)

314T, 160"F (Axial Flaw) 2500 2250 2000 1750 S' 1500 iii

! 1250

D..

"ti 1000

.s

.!!1 =

0 B 750 500 250

~~;

,_ *- -"11.

1 Unacceprabfe O ration Cpe,fmAnalysis Veraion:5.4 Film:273&3 OperJTI.xlsmVer.aion: !i.4.1 Criticality Limit based on inservice hydrootatlc test temperature (21 S°F) for ttle service period up to 54 EFPY I

J

' ---t---+~~ ~ower Jmit for ~CS p~ssure is 0 psia*I i

I i

i

-250 +-.-~.-t-.~..-1-~-.-+~.--r-1~~+-,-,~-+-,-~r+-,~..-+-m-.-+~.--r-1r-r-r.,..,..;

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

Page 26 of 27

Figure 6.2 Pressure and Temperature Limits Report Revision 9 (Effective until 54 EFPY)

Prairie Island Reactor Coolant System Cooldown Limitations to 54 EFPY (Without Margins for Instrument Uncertainty)

M-TIERIAL PROPER.IT BASIS LIMITING MATERIAL: *crut 1, Intermediate Shell Forging C LIMITING AfIT VALUES AI' 54 EFPY:

114T.

170°F (Axial Flaw) 3/4T.

160°F (A.--tia!Flaw) 2500 2250 2000 1750 S 1500

[

f 1250

I lll lll

~

0..

"C 1000 GO...

.!ll

I

.2.,

750 u

500 250 0

Unacceptable 0

ation ooldown Dpe'lltn.e.lla),151& Ves.'on:5.4Run:27388

~m.xlEmVersbr..: 5.-4.1 Acceptable 0 eration

', Lower Limit for RCS pres.sure is O psia I

-250 +-r-m+.-m+-r-m+-r-m+.-m-1-r-m-l-r-~-+-r~-+-r~-+-r~-+-r,--,-,-/

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

Page 27 of 27