L-PI-16-054, Pressure and Temperature Limits Report, Revision 5

Pressure and Temperature Limits Report, Revision 5

ML16175A639
Person / Time
Site:	Prairie Island
Issue date:	06/22/2016
From:	Northard S Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-16-054, TS 5.6.6.c
Download: ML16175A639 (24)

Text

fl Xcel Energy* Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch. MN 55089 L-PI-16-054 JUN 2 2 2016 TS 5.6.6.c U S Nuclear Regulatory Commission ATTN: Document Control Desk 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 Pressure and Temperature Limits Report. Revision 5

Reference:

1. NRC letter to K. Davison, PINGP, Prairie Island Nuclear Generating Plant, Units 1 and 2- Issuance of License Amendments Regarding Revision to Technical Specification 3.5.3, "ECCS- Shutdown" (TAC Nos.

MF0727 and MF0728), dated 5/20/2015 (ADAMS Accession No. ML15062A013)

Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), provides the enclosed Revision 5 of the Prairie Island Nuclear Generating Plant (PINGP) Pressure and Temperature Limits Report (PTLR) in accordance with Technical Specification (TS) 5.6.6.c.

Revision 5 deletes reference to TS Section 3.5.3 from Section 3.0 of the PTLR and brings the PTLR into accord with PINGP License Amendment Nos. 213 and 201 for Units 1 and 2, respectively, (Reference 1). These amendments deleted reference to the PTLR from the TS Section 3.5.3 Applicability.

Summary of Commitments This letter contains no new commitment and no change to an existing commitment.

Scott Northard Acting Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure (1)

Document Control Desk Page 2 cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

ENCLOSURE 1 PRESSURE AND TEMPERATURE LIMITS REPORT. REVISION 5 (21 pages follow)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pressure and Temperature Limits Report RECORD OF REVISION Revision Approval Remarks No. Date 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.

Page 1 of 21

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pressure and Temperature Limits Report RECORD OF REVISION Revision Approval Remarks No. Date 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 3.1 0.1 to TRM references 3.4.4 and 3.4.5.

On page 4, subsection "Pressurizer Temperature Limits," revised the Maximum Pressurizer Goold own Rate to 200°F per hour versus 1OOoF 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, RT Prs, 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 Sl pump disable temperature. Refer to License Amendment 213/201.

Page 2 of 21

Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Prepared by: - /omDt,>jvh!0 '

. Print Name Sr. Engineer

• Engineering Reviewed by: . 1'1\MI.C P'd:..(JSor/ ~~

Print Name

• moo ..,~7 ~Ia tory Wr3{ 15Affairs Manager slrJ~-'

. Date Approved by: G;lrzfr ...Joh YJ~I'-~~

rint Name Page 3 of 21

• Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table of Contents Section Title Page 1.0 PURPOSE .................................................................................................................... 5 2.0 APPLICABILITY ........................................................................................................... 5 3.0 OPERATING LIMITS .................................................................................................... 6 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 ............................................................................. 16 Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule ........................ 17 Table 6.4 Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 54 EFPY ......................... 18 Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 54 EFPY ......................... 19 List of Figures Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 54 EFPY .......................................................*....................................................... 20 Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 54 EFPY .............................................................................................................. 21 Page 4 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY) 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 (PIT) Limits TS 3.4.6 RCS Loops - MODE 4 TS 3.4.7 RCS Loops- MODE 5, Loops Filled TS 3.4.1 0 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 TS 3.5.3 ECCS- Shutdown Miscellaneous Specifications- Technical Requirements Manual TRM 3.4.4 PTLR Compliance- Pressurizer TRM 3.4.5 PTLR Compliance- Steam Generator(s)

Page 5 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY) 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.

Over Pressure Protection System (OPPS) Enable Temperature 310 °F*

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

• Analytical limit [225 oF] plus indicating instrument channel uncertainty [18 °F]

(Reference 5.11) plus additional margin for operational simplicity.

Safety Injection (51) Pump Disable Temperature 218 Of*

I Referenced in: TS 3.4.12, II

. TS 3.4.13 .

• Analytical limit [200 °F] plus indicating instrument channel uncertainty [18 °F]

(Reference 5.11 ).

Page 6 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

RCS Pressure/Temperature (P/Tl Limits Figure 6.1* RCS PIT limits for heatup Figure 6.2* RCS PIT limits for cooldown Referenced in: TS 3.4.3, TS 3.4.12, TS 3.4.13, SR 3.4.3.1

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

NOTE: Tables 6.1 and 6.2 contain a tabulated version of the curves.

Instrumentation Uncertainty for P/T Curves Pressure Uncertainty Temperature Uncertainty These values must be applied to the P/T limit curves in NOTE: operating procedures (Reference 5.10 and 5.11 ).

RCS Heatup/Cooldown Rate Limits 100 oF per hour Maximum RCS Heatup Rate 100 oF per hour Maximum RCS Cooldown Rate I Referenced in: I TS 3.4.3, SR 3.4.3.1 Over Pressure Protection System COPPS) PORV Setpoint 500 psig*

I Referenced in: TS 3.4.12, TS 3.4.13

• This setpoint accounts for instrument channel uncertainty (Reference 5.8).

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Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

RCS Minimum Temperature When Not Vented 86 Of*

I Referenced in: TS 3.4.3,

. TS 5.5.6

• Analytical limit [68°F] plus indicating instrument channel uncertainty [18°F]

(Reference 5.11)

Minimum Boltup Temperature 60 Of**

Referenced in: ITS 5.5.6

• • No instrument uncertainty included.

Pressurizer Temperature Limits 100 of per hour Maximum Pressurizer Heatup Rate 200 of per hour Maximum Pressurizer Cooldown Rate Maximum Temperature Difference Between the Pressurizer and the Spray Fluid for which the Pressurizer Spray can be used.

I Referenced in: I3.4.4 TRM Specification Steam Generator Temperature/Pressure Limit 200 psig Maximum secondary side Pressure if the temperature of the steam generator is below 70 of.

Referenced in: I TRM Specification 3.4.5 Page 8 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY) 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 Tis 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 (Table 6.4 and 6.5). The limiting ARTs at 54 EFPY for this material are 150°F for 1/4T, and 133°F for 3/4T.

The Heatup and Cooldown limitations and curves remain unchanged from those developed for 35 EFPY. Because the ART values at 54 EFPY are lower than the values calculated for 35 EFPY in Reference 5.3, the ART values from Reference 5.3 remain as bounding values for development of the Pressure/Temperature limits.

The limiting ARTs used to develop the Pressure/Temperature limits are as follows:

1/4T =154 °F 3/4T =136 °F

References:

I 5.3, 5.6 Page 9 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

End of Life Fluence Reference Temperature (RTm!l 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 upper to intermediate shell forging circumferential weld material. The limiting RTpts is as follows:

RTpts = 157 °F

Reference:

I 5.12 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 to lower shell weld neutron fluences used in determining the 54 EFPY limiting ART for the reactor vessels are as follows:

Units are 1019 n/cm 2 , for energies> 1.0 MeV at 54 EFPY Clad/Base Metal Interface = 4.97 1/4T = 3.33 3/4T = 1.49

References:

15.12, 5.13 These values are not the highest fluences that were obtained in the reactor vessels, but are the values determined for the NOTE: most limiting material. The highest fluences were obtained at the unit 2 intermediate shell forging.

(References 5.12 and 5.13).

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Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

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 =80.8 OF 3/4T =80.8 °F

References:

I 5.13 Reactor Vessel Material Surveillance Program The Reactor Vessel Material Surveillance Program is described in the USAR (Reference 5.9). The schedule for removal of the Units 1 and 2 capsules is contained in Table 6.3 of this report.

References:

5.2, 5.5, 5.9 Supplemental Data Tables Tables 6.4 and 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:

I 5.13 Page 11 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Surveillance Data Credibility The credibility of surveillance capsule data is determined as specified in Regulatory Guide 1.99, Revision 2, Section B. Four 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 5.2 and 5.5. 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 material, the Unit 1 intermediate to lower shell forging circumferential weld, surveillance data was available but considered non-credible. This resulted in the use of the full crl!. margin of 28°F. The ART calculated using surveillance capsule data is larger than that calculated using position 1.1. For comparison Tables 6.4 and 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.

RCS Minimum Temperature When Not Vented This is the RCS lower temperature limit until the system is vented with at least a 3 square inch vent.

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.

Page 12 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

5.0 REFERENCES

5.1 WCAP-14040-NP-A, Methodology Used to Develop Cold Overpressure Mitigation, Revision 2, January 1996.

5.2 WCAP-14779, Analysis of CapsuleS from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, Revision 2, February 1998.

5.3 WCAP-14780, Prairie Island Unit 1 Heatup and Cooldown Limit Curves Normal Operation, Revision 3, February 1998.

5.4 WCAP-14781, Evaluation of Pressurized Thermal Shock for Prairie Island Unit 1, Revision 3, February 1998.

5.5 WCAP-14613, Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, Revision 2, February 1998.

5.6 WCAP-14637, Prairie Island Unit 2 Heatup and Cooldown Limit Curves Normal Operation, Revision 3, December 1999.

5.7 WCAP-14638, Evaluation of Pressurized Thermal Shock for Prairie Island Unit 2, Revision 3, December 1999.

5.8 Westinghouse Letter NSP-98-0120, "Prairie Island Units 1 and 2 COMS Setpoint Analysis," Revision 2, February 1998.

5.9 USAR Section 4.7.2, "Reactor Vessel Material Surveillance Program" 5.10 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 0.

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

5.12 Calculation CN-MRCDA-07-59, "Prairie Island Units 1 and 2 Measurement Uncertainty Recapture: Reactor Vessel Integrity Evaluation".

5.13 Calculation ENG-ME-819, "Adjusted Reference Temperature for Unit 1 and Unit 2 Reactor Vessel Materials at 54 EFPY".

Page 13 of21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY) 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.

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Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table 6.1 54 EFPY Heatup Data Points (Without Instrumentation Uncertainty Margins)

Heatup Curves 60 Heatup Critical. Limit 100 Heatup Critical. Limit Leak Test Limit T p T p T p T p T p 60 0 273 0 60 0 273 0 251 2000 60 584 273 594 60 560 273 560 273 2485 65 584 273 587 65 560 273 560 85 584 273 584 85 560 273 560 90 584 273 584 90 560 273 560 95 584 273 586 95 560 273 560 100 586 273 591 100 560 273 560 105 591 273 597 105 560 273 560 110 597 273 604 110 560 273 562 115 604 273 613 115 562 273 566 120 613 273 622 120 566 273 571 125 622 273 633 125 571 273 577 130 633 273 645 130 577 273 585 135 645 273 658 135 585 273 594 140 658 273 672 140 594 273 604 145 672 273 687 145 604 273 615 150 687 273 704 150 615 273 627 155 704 273 722 155 627 273 641 160 722 273 741 160 641 273 656 165 741 273 761 165 656 273 672 170 761 273 784 170 672 273 690 175 784 273 808 175 690 273 709 180 808 273 833 180 709 273 730 185 833 273 861 185 730 273 752 190 861 273 891 190 752 273 777 195 891 273 923 195 777 273 802 200 923 273 957 200 802 273 831 205 957 273 994 205 831 273 861 210 994 273 1033 210 861 273 893 215 1033 273 1076 215 893 273 928 220 1076 273 1121 220 928 273 966 225 1121 273 1170 225 966 273 1006 230 1170 275 1223 230 1006 275 1049 235 1223 280 1279 235 1049 280 1096 240 1279 285 1339 240 1096 285 1149 245 1339 290 1404 245 1146 290 1199 250 1404 295 1473 250 ' 1199 295 1257 255 1473 300 1548 255 1257 300 1318 260 1548 305 1628 260 1318 305 1384 265 1628 310 1713 265 1384 310 1455 270 1713 315 1805 270 1455 315 1531 275 1805 320 1903 275 1531 320 1612 280 1903 325 2007 280 1612 325 1699 285 2007 330 2119 285 1699 330 1792 290 2119 335 2231 290 1792 335 1892 295 2231 340 2347 295 1892 340 1998 300 2347 345 2471 300 1998 345 2112 305 2471 305 2112 350 2233 310 2233 355 2363 315 2363 Page 15 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table 6.2 54 EFPY Cooldown Data Points (Without Margins for Instrumentation Uncertainty)

Cooldown Curves Steady State 20 deg F 40 deg F 60 deg F 100 deg F T p T p T p T p T p 60 0 60 0 60 0 60 0 60 0 60 590 60 563 60 537 60 510 60 455 65 594 65 568 65 542 65 515 65 460 70 599 70 573 70 547 70 520 70 465 75 605 75 579 75 552 75 526 75 471 80 611 80 585 80 558 80 532 80 478 85 617 85 591 85 565 85 539 85 485 90 621 90 598 90 572 90 546 90 493 95 621 95 605 95 580 95 554 95 502 100 621 100 613 100 588 100 563 100 511 105 621 105 621 105 597 105 572 105 520 110 621 110 621 110 607 110 582 110 531 115 621 115 621 115 617 115 592 115 543 116 621 116 621 120 628 120 604 120 555 116 668 116 644 125 640 125 616 125 568 120 676 120 652 130 653 130 630 130 583 125 687 125 664 135 667 135 644 135 599 130 699 130 676 140 682 140 660 140 615 135 712 135 690 145 698 145 676 145 634 140 726 140 704 150 715 150 695 150 653 145 741 145 720 155 734 155 714 155 674 150 757 150 736 160 754 160 735 160 697 155 774 155 754 165 776 165 757 165 722 160 793 160 773 170 799 170 782 170 748 165 813 165 794 175 824 175 808 175 777 170 834 170 816 180 851 180 836 180 808 175 857 175 841 185 880 185 866 185 841 180 882 180 866 190 911 190 899 190 876 185 909 185 894 195 945 195 934 195 915 190 937 190 924 200 981 200 972 200 956 195 968 195 956 205 1019 205 1012 205 1001 200 1001 200 990 210 1061 210 1056 210 1048 205 1036 205 1027 215 1106 215 1102 215 1100 210 1075 210 1067 220 1154 220 1153 220 1155 215 1115 215 1110 225 1205 220 1159 220 1156 225 1206 225 1205 230 1257 235 1311 240 1370 245 1432 250 1500 255 1572 260 1649 265 1732 270 1820 275 1915 280 2017 285 2126 290 2243 295 2367 Page 16 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule Recommended Surveillance Capsule Removal Schedule for Unit 1 Capsule Location Withdrawal Fluence(a)

Capsule Lead Factor(a) EFPY(b) 2 (degree) (n/cm , E> 1.0 MeV) v 77 2.94 1.34 5.630 X 10 18(c) p 247 1.72 4.60 1.318 X 10 19(c)

R 257 2.99 8.56 4.478 X 10 19(c) s 57 1.77 18.12 4.017 X 10 19(c)

T 67 1.89 Standby ---

N 237 1.77 Standby ---

Recommended Surveillance Capsule Removal Schedule for Unit 2 Capsule Location Withdrawal Fluence(d)

Capsule Lead Factor(d) EFPY(b) 2 (degree) (n/cm , E> 1.0 MeV) v 77 2.95 1.39 6.206 X 10 18(c)

T 67 1.75 4.00 1.199 X 10 19(c)

R 257 2.99 8.81 4.376 X 10 19(c) p 247 1.84 17.24 4.165 X 10 19(c)

N 237 1.72 Standby ---

s 57 1.72 Standby ---

Notes:

(a) Updated in Capsule S dosimetry analysis.

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

(c) Plant specific evaluation.

(d) Updated in Capsule P dosimetry analysis.

Page 17 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table 6.4 Prairie Island Unit 11/4T and 3/4T ART Calculations at 54 EFPY 114T f 114T FF(d) l(e) M(g) .1RT NDT ART f@54 Material CF EFPY(a) (oF) (oF) 314Tf 314T FF CF) CF) 1/4T Calculations Upper Shell 51 1.770 1.185 1.047 -4 34 53.4 84 Forging B Upper to Inter. Shell 79.5 1.770 1.185 1.047 o<c) 66 83.3 149 CircWeldW2 Intermediate Shell 44.0 5.162 3.455 1.324 14 34 58.2 107 Forging C Using SIC Data 54.7 5.162 3.455 1.324 14 34(b) 72.4 121 Inter. to Lower 69.7 4.969 3.326 1.315 -13 56 91.6 135 SheiiWeldW3 Using SIC Data 80.8 4.969 3.326 1.315 -13 56(b) 106.2 150 Lower Shell Forging D 44.0 5.026 3.364 1.318 -4 34 58.0 88 3/4T Calculations Upper Shell 51 1.770 0.5307 0.823 -4 34 42.0 72 Forging B Upper to Inter. Shell 79.5 1.770 0.5307 0.823 o<c) 66 65.4 131 CircWeldW2 Intermediate Shell 44.0 5.162 1.548 1.121 14 34 49.3 98 Forging C Using SIC Data 54.7 5.162 1.548 1.121 14 34(b) 61.3 110 Inter. to Lower 69.7 4.969 1.490 1.110 -13 56 77.4 121 SheiiWeldW3 Using SIC Data 80.8 4.969 1.490 1.110 -13 56(b) 89.7 133 Lower Shell Forging D 44.0 5.026 1.507 1.114 -4 34 49.0 79 NOTE:

19 2 (a) Fluence values (f) are x 10 n/cm (E > 1.0 MeV). (Ref. 5.13)

(b) The full cr!J. margin of 17"F for the forging and 28°F for the weld was used since the surv. data was deemed not credible (Ref. 5.3).

(c) Estimated per Standard Review Plan Section 5.3.2 (Ref. 5.3).

(d) FF, Fluence Factor= f(0.28-0.1*1ogf) (Ref. 5.13)

(e) I is the unirradiated material reference temperature. (Ref. 5.3)

(g) M is a margin term required for conservative results. (Ref. 5.3)

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Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 54 EFPY f@54 1/4T f 1/4T FF(c) l(d) M(e) ~RTNDT ART Material CF 3/4T f(a) (oF) (oF) (oF)

EFPY 3/4T FF CF) 1/4T Calculations Upper Shell Forging B 44.0 1.743 1.167 1.043 -13 34 45.9 67 Upper to Inter. Shell Weld 70.0 1.743 1.167 1.043 -13 56 72.7 116 W2 Using Unit 1 SIC Data 80.8 1.743 1.167 1.043 -13 56 84.3 128 Intermediate Shell Forging 44.0 5.196 3.478 1.325 14 34 58.3 107 c

Inter. to Lower Shell Weld 52.0 5.043 3.375 1.318 -31 56 68 93 W3 Using SIC Data 80.0 5.043 3.375 1.318 -31 28 105.7 103 Lower Shell Forging D 51.0 5.112 3.421 1.321 -4 34 67.4 98 Using SIC Data 60.0 5.112 3.421 1.321 -4 34(1) 78.8 109 3/4T Calculations Upper Shell Forging B 44.0 1.743 0.5226 0.8187 -13 34 36.0 57 Upper to Inter. Shell Weld 70.0 1.743 0.5226 0.8187 -13 56 57.1 101 W2 Using Unit 1 SIC Data 80.8. 1.743 0.5226 0.8187 -13 56 66.2 110 Intermediate Shell Forging 44.0 5.196 1.558 1.123 14 34 49.4 98 c

Inter. to Lower Shell Weld 52.0 5.043 1.512 1.114 -31 56 57.5 83 W3 Using SIC Data 80.0 5.043 1.512 1.114 -31 28 89.4 87 Lower Shell Forging D 51.0 5.112 1.533 1.118 -4 34 57.0 87 Using SIC Data 60.0 5.112 1.533 1.118 -4 34(1) 66.6 97 NOTE:

19 (a) Fluence values (f) are x 10 nlcm 2 (E > 1.0 MeV). (Ref. 5.13)

(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'> of 28°F was used in the margin term.

(c) FF, Fluence Factor= f(0.28-0.1 *logf). (Ref. 5.13)

(d) I is the unirradiated material reference temperature. (Ref. 5.6)

(e) M is a margin term required for conservative results. (Ref. 5.6)

(f) The full Of'> margin of 1rF for the forging was used since the surveillance data was deemed not credible (Ref. 5.6).

Page 19 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 54 EFPY (w/o Margins for Instrument Uncertainty) 2500 I i i I I I I I I 17 1 I j lnservice Hydrostatic tt:JiitJH-+-+'++-+-JH----t-_j_j_J_J I

1 i I I I I I Pressure Test t-1~/ I I/ I1 i

I l  !

i : I

I I

I 1/

• v I I

I i [ l I I I I I I I 2000 I I I i

i I I I

1/

J 1/ I *

[*<1111111-t v Heatup Rate I

~ ~~ UNACCEPTABLE ~0 1 1 1

• 1

• 1 ' 1*

1 1 ' 1 I I . 17 ' 17 ;- Up to_ F!Hr

, I '1 i 1I OPERATION '

• Cnt1cal" I I I I I II I I

• VI V) I I I I 1 Heatup Rate Ci l 1500 I i 1 . . J )~ Up to 100 F/Hr I-1 i :  : Critical* I-CD I

1 Ul I i  ! I .**I/ .* v II I II I Ul

... V

• V ACCEPTABLE '

CD Q.

1 J

1 1

I/  : I/ OPERATION Jt "C

CD

<II u

1000 l I

l I Heatup Rate '

Up to 60 F/Hr SubcriticaJ*

. .* V

• V I I I

1

[ l I

II 1i

-= I i

~ v v~

I  !

• A' i I I I

l I'

1 , ~ 1_ *

• • 1 Criticality Limit Based on

• 1 - - - - - * *  !.-~---' Heatup Rate ~II- lnseJVice Hydrostatic Test

• 1 Up to 100 F/Hr 1 Te":lperature (238 F) For the 500 1-+---1-- Boltup1 I SubcriticaJ* - 1 SeJVJce Period up to 54 EFPY Temp.l 1 1 1

l i l~ I I I !I I - I I : I I I I I I I I I I I I II l ,' j 0

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

• For each cur.<e, acceptable operation is to the right and below the cur.<e.

Page 20 of 21

Pressure and Temperature Limits Report Revision 5 (Effective until 54 EFPY)

Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 54 EFPY (w/o Margins for Instrument Uncertainty) 1800 1600 I I

/

1400 I

/

UNACCEPTABLE

/

1200 OPERATION 17 m

• J

-~

0..

1000 II)'

IJ

..41~

en Ill ~

Gl ll.

f-- Cooldown Ai;:Jr'l A W/ ACCEPTABLE f--

Rates* ~ ~ OPERATION

"'C 800 f-- _/. ~ 'l

.....as Gl Peg_ F I Hr I--

u I--

I-- 0 1-:/ 1:/:: ~ ~---

"'C 1/

f--

.E 600 '-----

20  ::;... v f-- 40  ::.-

~

f--

1- 60 1- 100 400 1Boltup 200 1Temp_

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

• For each curve, acceptable operation is to the right and below the curve.

Page 21 of 21