L-PI-15-034, Pressure and Temperature Limits Report (PTLR) Revision 4

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Pressure and Temperature Limits Report (PTLR) Revision 4
ML15134A153
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 05/14/2015
From: Davison K
Northern States Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-15-034
Download: ML15134A153 (23)
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Text

(l Xcel Energy@

MAY 1* 4 2015 L-PI-15-034 US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 Pressure and Temperature Limits Report (PTLR) Revision 4 Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), encloses an approved copy of Revision 4 to the Pressure and Temperature Limits Report (PTLR) in accordance with Technical Specification 5.6.6.c.

Revision 4 updates the Effective Full Power Years (EFPY) values throughout the report from 35 EFPY to 54 'EFPY, which is projected to be beyond the expiration of the renewed operating licenses for Prairie Island Nuclear Generating Plant Units 1 and Unit 2.

Summary of Commitments This letter contains no new commitments and no changes to existing commitments.

)/tu&'v-jJ_,~~

Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure (1) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota 1717 Wakonade Drive East

  • Welch, Minnesota 55089-9642 Telephone: 651.388.1121

ENCLOSURE 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) REVISION 4 21 Pages Follow

PRAIRIE ISLAND NUCLEA R GENERA TING 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 Amendmen t Request dated March 6, 1998, as Amendmen t 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 ofT echnical 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 NUCL EAR GENE RATIN G PLANT Pressu re and Tempe rature 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.

O,n 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 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, RT PTs. 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.

Page 2 of 21

Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY) c51[)~

Sr. Engineer Engineering Reviewed by: Dq I~.. /Vt \/1 V\ C. 1(: J**

Print.Name .

CQ.A R)t;L~

Acting Regulatory 3!3tJ/rt; Date Affairs Manager

.<"~).

Reviewed by: =-' '- -"'- "g;. ;. r~tA;. :_r'l Print Nam S."""':C-"F*-=a...;_r)"--, v Program rlnginee ring Manager sj_~~r 1

Date Approve d by: E..c~o-l'd .ldCln0\1) 'j~~ ing Print Name Site Director of Engineer

. Page 3 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

Table of Contents

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. ' .. ,... -..-. iiH~'* .. - ****- ....

.~

Page Section 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 11/4T 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 4 (Effective until 54 EFPY) 1.0 PURPOSE and The purpose of the Prairie Island Nuclear Generating Station Pressure 1 and 2 Temperature Limits Report (PTLR) is to present operating limits for Units Heatup , Cooldo wn and low relating to; (1) RCS pressure and temperature during

  • temperature* operatio-n;-(2)' RGS heatt:l'p *and cooldown rates; (3) the O~er Pressu re (5) Safety Protection System (OPPS) arming temperature; (4) OPPS lift settings; l stress related temperature Injection Pump disable temperature as well as (6) therma prepared limitations for the pressurizer and steam generators. This report has been in accordance with the requirements with Technical Specification 5.6.6.

2.0 . APPLI CABIL ITY Years This report is applicable to both Units 1 and 2 until 54 Effective Full Power Reacto r Pressu re Vessel . The Techni cal (EFPY) is reached on that particular units' in this report are:

Specifications that are affected by the information contained TS 3.4.3 RCS Pressure and Temperature (Pff) 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) >Safe ty Injection (SI) Pump Disable Temperature TS 3.4.13 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) :5 Safety Injection (SI) Pump Disable Temperature TS. 3.5.3 ECCS - Shutdown Miscellaneous Specif ication s- Technical Requirements Manual TRM 3.4.4 PTLR Comp liance - Pressurizer TRM 3.4.5 PTLR Comp liance - Steam Generator(s)

Page 5 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY) 3.0 OPERATING LIMITS of All limits are valid until 54 EFPY, which is projected to be beyond the expiration the operating license for each of Prairie Island Units 1 and 2.

Over Pressu re Protect ion System (OPPS) Enable Tempe rature 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 oF]

(Reference 5.11) plus additional margin for operational simplicity.

Safety Injectio n (SI) Pump Disable Tempe rature Referenced in: TS 3.4.12, TS 3.4.13, TS 3.5.3

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

(Reference 5.11 ).

Page 6 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

RCS Pressu re/Tem peratur e (PIT) 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.

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

Instrum entatio n Uncerta inty for PIT Curves 124 psig Pressure Uncertainty 18 oF Temperature Uncertainty These values must be applied to the PIT limit curves in operatin g procedu res (Reference 5.10 and 5.11 ).

RCS Heatup /Cooido wn Rate limits 100 oF per hour Maximum RCS Heatup Rate 100 oF per hour Maximum RCS Cooldown Rate I Referenced in: ITS 3.4.3, SR 3.4.3.1 Over Pressu re Protect ion System (OPPS) PORV Setpoin t 500 psig*

I Referenced in: ITS 3.4.12, TS 3.4.13

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

Page 7 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

RCS Minimum Tempera ture When Not Vented 86 °F*

I Referenced in: TS 3.4.3,

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

(Reference 5.11)

Minimum Boltup Tempera ture 60 °F**

Referenced-in: ITS 5.5.6

    • No instrument uncertainty included.

Pressuri zer Tempera ture 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.

Referenced in: TRM Specification 3.4.4 Steam Generat or Tempera ture/Pre ssure Limit 200 psig Maximum secondary side Pressure if the temperature of the steam generator is below 70 °F.

Referenced in: I TRM Specification 3.4.5 Page 8 of 21

Pressure and Temperature Limits Report Revision 4 (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 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 (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 II Page 9 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

End of Life Fluence Reference Temperature <RTill§.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 oF

Reference:

15.12 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:

19 2 Units are 10 n/cm , 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 most limiting material. The highest fluences were obtained at the unit 2 intermediate shell forging.

(References 5.12 and 5.13).

Page 10 of 21

Pressure and Temper ature Limits Report Revision 4 (Effective until 54 EFPY)

Chemis try Factor (CF)

Chemis try Factors are parame ters used in the develop ment of the ARTs for the beltline materials and accoun t for the Copper and Nickel content in the reactor vessel beltline materials. The chemistry factors determi ned for the limiting ARTs, as"*'.. **.

I corresponding to the Unit 1 intermediate to lower shell circumferential weld, are follows.

1/4T =80.8 OF 3/4T =80.8 OF

References:

I 5.13 Reactor Vessel Material Surveillance Program The Reactor Vessel Material Surveillance Program is described* in the USAR (Reference 5.9). The schedul e 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 Supplem ental Data Tables Tables 6.4 and 6.5 contain the develop ment of all of the ARTs for the beltline materia ls for Unit 1 and Unit 2 respectfully, including all the parameters.

References:

I 5.13 II Page 11 of 21

Pressure and Temperature Limits Report Revision 4 (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 a8 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 4 (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-4508 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 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY) 6.0 ATTACHMENTS 6.1 Table 6.1-54 EFPY Heatup Data Points 6.2 Table 6.2- 54_ EF,P'(,qopldown.Q"C?.~a 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.

Page 14 of 21

Pressure and Temperature Limits Report Revision 4 (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 p T p T p T p T p T 60 0 .... 273 ..... o 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 r 315 2363 Page 15 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

Table 6.2 54 EFPY Cooldown Data Points I ou t Margms f (W'th orIns t rumen taf1on Unee rt't) amty]

Cooldown Curves 100 deg F 20 deg F 40 deg F 60 deg F Steady State p p T p p T p T T T 0 60 0 0 60 0 60 0 60 60 -

.. -. '590 **~*" ... 60 563. 60 - 537 60 510 60 455 40 60~--*--

568 65 542 65 515 65 460 65 594 65 465 573 70 547 70 520 70 70 599 70 471 579 75 552 75 526 75 75 605 75 478 585 80 558 80 532 80 80 611 80 85 485 85 591 85 565 85 539 85 617 90 493 90 598 90 572 90 546 90 621 95 502 95 605 95 580 95 554 95 621 100 511 100 613 100 588 100 563 100 621 105 520 105 621 105 597 105 572 105 621 110 531 110 621 110 607 110 582 110 621 115 543 115 621 115 617 115 592.

115 621 120 555 116 621 120 628 120 604 116 621 125 568 116 644 125 640 125 616 116 668 130 583 120 652 130 653 130 630 120 676 135 599 125 664 135 667 135 644 125 687 140 615 130 676 140 682 140 660 130 699 145 634 135 690 145 698 145 676 135 712 150 653 140 704 150 715 150 695 140 726 155 674 145 720 155 734 155 714 145 741 160 697 150 736 160 754 160 735 150 757 165 722 155 754 165 776 165 757 155 774 170 748 160 773 170 799 170 782 160 793 808 175 777 813 165 794 175 824 175 165 180 836 180 808 170 834 170 816 180 851 841 185 880 185 866 185 841 175 857 175 876 866 190 911 190 899 190 180 882 180 915 894 195 945 195 934 195 185 909 185 956 924 200 981 200 972 200 190 937 190 1001 195 956 205 1019 205 1012 205 195 968 1056 210 1048 1001 200 990 210 1061 210 200 215 1102 215 1100 205 1036 205 1027 215 1106 1067 220 1154 220 1153 220 1155 210 1075 210 215 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 4 (Effective until 54 EFPY)

Table 6.3 Reacto r Vessel Material Surveil lance Capsul e Removal Schedu le Recom mende d Surveil lance Capsul e Remov al Schedu le for Unit 1 Withdraw al Fluence(a)

Capsule l,.ocation Lead FaGtor(a) *

'.,".Caps ule EFPY(b) 2 (degree) (n/cm , E> 1.0 MeV) v 77 2.94. 1.34 5.630 X 1018(c) p 247 1.72 4.60. 1.318x1 019 (c)

R 257 . 2.99 8.56 . 4.478x1 0 19(c) s 57 1.77 18.12 4.01.7 X 10 19(c)"

T 67 1.89 Standby ---

N 237 1.77 Standby ---

Recom mende d Surveil lance Capsul e Remov al Schedu le for Unit 2 Withdraw al Fluence(d)

Capsule Location Lead Factor(d)

Capsule (degree) EFPY(b) 2 (n/cm , E> 1.0 MeV) v* 77 2.95 1.39 6.206 X 1018 (c)

T 67 1.75 4.00 1.199 X 1019(c)

R 257 2.99 8.81 4.376 X 1019(c) p 247 1.84 ° 17.24 4.165 X 1019 (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 ana.lysis.

Page 17 of 21

Pressure and Temperature Limits Report Revision 4 (Effective until 54 EFPY)

Table 6.4 Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 54 EFPY 1/4T f 1/4T FF(d) l(e) M(g) ART NDT ART f@54 Material CF EFPY(a) (oF) (oF) 3/4Tf 3/4T FF CF) CF) 1/4T Calculations Upper Shell 51 1.770 1.185 1.047 -4 34 53.4 84 Forging B o(c) 83.3 149 Upper to Inter. Shell 79.5 1.770 1.185 1.047 66 Circ WeldW2 Intermediate Shell 44.0 5.162 3.455 1.324 14 34 58.2. 107 Forging C Using S/C 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 80.8 4.969 3.326 1.315 -13 56(b) 106.2 150 Using SIC Data 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 o(c)

Upper to Inter. Shell 79.5 1.770 0.5307 0.823 66 65.4 131 CircWeldW2 Intermediate Shell 44.0 5.162 1.548 1.121 14 34 49.3 98 Forging C Using S/C 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 S/C 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 Ol:!,. margin of 1?"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)

Page 18 of 21

Pressure and Tempera ture Limits Report Revision 4 (Effective until 54 EFPY)

Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART C.alculations at 54 EFPY 1/4T FF(c) l(d) M(e) ~RTNDT ART f@54 1/4T f Material CF 3/4T fa) 3/4T FF CF) (oF) - n~} * . CF). *' *...:..:-** ... -*,.,.,... *.

EFPY 1/4T Calculat ions 44.0 1.743 1.167 1.043 -13 34 45.9 67 Upper S.hell Forging B 1.743 1.167 1.043 ~13 56 . 72.7 116 Upper to Inter. Shell Weld 70.0 W2 80.8 1.743 1.167 1.043 -13 56 84.3 128 Using Unit 1 S/C Data

44.0 5.196 3.478 1.325 14 34 58.3 107 Intermediate Shell Forging c

52.0 5.043 3.375 1.318 -31 56 68 93 Inter. to Lower Shell Weld W3 80.0 5.043 3.375 1.318 -31 28 105.7 103 Using S/C Data 51.0 5.112 3.421 1.321 -4 34 67.4 98 Lower Shell Forging D 34(1) 60.0 5.112 3.421 1.321 -4 78.8 109 Using S/C Data 3/4T Calculat ions 44.0 1.743 0.5226 0.8187 -13 34 36.0 57 Upper Shell Forging B 70.0 1.743 0.5226 0.8187 -13 56 57.1 101 Upper to Inter. Shell Weld W2 ..

80.8 1.743 0.5226 0.8187 -13 56 66.2 110 Using Unit 1 S/C Data

44.0 5.196 1.558 1.123 14 34 49.4 98 Intermediate Shell Forging c

52.0 5.043 1.512 1.114 -31 56 57.5 83 Inter. to Lower Shell Weld W3 80.0 5.043 1.512 1.114 -31 28 89.4 87 Using S/C Data 51.0 5.112 1.533 1.118 -4 34 57.0 87 Lower Shell Forging D 34(1)

Using S/C Data 60.0 5.112 1.533 1.118 -4 66.6 97 NOTE:

19 2 (a) Fluence values (f) are x 10 n/cm (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 a;:.. 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 a;:.. margin of 17°F 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 4 (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 r-'--1--'--.l..-.i--'---'-""'---'--,

+-+-l--i-t-H lnservice,l-l;ydrostatic .

Ptessiire Test. H2=+++-t---+--.r+--17-+--t-c;'t--l 111

-J-I'_J..JJ=~-J-_[lll 1 _u LJ_JrT____ *r-lLLLI:IV ~1 _v1 _]_

2ooo , I l I j 1J

_J UNACCEPTABLE l

-r-j rl7

-i V * ;

I;.;.L IT Heatup Rat~

Up tp_~? ~Hr 1

. i 1_1 OPERATfON [J 1 * -f*1 , ._Tl -Y Cntical 1 ...

01

]

1 l

j 1

I l l 1---+--i-l I . 1 '

v' J V . B~atup,Rate

./' I 1l 7, 1500 *  : * . l l I -J- I i, /1 1- 71it Up to tOO'F!Hr -*

1- I i l . 1:. t  ;; .

  • Critical~'

I -ll-+-_j .

~-

.1* I .* I I I I .*j vI 1.* vI l -H-i-==t==f-1 J I I.  :

  • v l .*f v I ACCEPTABLE l

_j_l1* 1 . H.ea.tup.~ateUpto~Ofl~r

_ ....:.~---. +/-E~V . 1

  • OPERATlON-1---"-*

,.,. Vi 1

  • 1 1* l . ., *s *--

__, _,__ L_L __ __ Subcntical . **1 V 1 l. .* --+--1 I I 1

, j I ~UJ_ I I ~r *t:tr~-- - . .r!-H-l---1-H-!-1

', I-r-+-+ I I j _ ~ 'v

  • .... 1

-~ -!+J-:-

y!A!- -_ !I -. J : Ltm1t crrnc.:t~ty . _Ba~ed on

  • C]- h-b -
  • _J.-1--
  • Heatup Rate 7

~ lnservtceHydrostatJc Test ouu I* EH .* t

!. I Bqltup

  • _j1_:_j.-

l . -iL UpSubcrltitar..

1_ toJOO F/Hr J -!+/-'* -.Servrce Te~per:atur . e:_(23& R_l For the.

Penod J.Jj:Ho 54-EFPY Temp.- II , I IJ I I l_j ] I ., 1---j-f-j I I I I

-i 0 I I I I I J _I l I I i I I I l 0 ;50 '100 150. 200 250 ;30Q 350 400

  • hidicate(J Tem pe_~~ture (DeQ. F)

~ For:erujlnurve, ;!~Cpe!)tab.te::~fion is toble rig!l~ IIM,-l>et""!*1h~,eurV-e~

Page 20 of21 ~)

Pressure and Temperature Limits Report Revision 4 (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

-Ol' '1200 OPERATION /

1/

"i.i.i Q.

....w

=;:

~"

'1000 A'Y

  • IJI 7

al* *19 W*

1- Cooldawn IA:J 'l"'

ACCEPTABLE 0.. 1- A W/

u Ql*

l l$

(;)*

0 8100 I--

1-1-

1-Rates" D~_F/Hr 0 --

./. ;,z-

/. ~ '/

~
;...-
.-
1/

OPERATION 5 - 20 ~

600 -

- 40 f.--

- 60

- - 100 f.--

400 200 Bolfup

.-1Temp_l I

0 -

0 50 100 '150 200 250 300 350 Indicated Temperature {Deg. F}

  • For each cmve. accepinble operation is to the right and below the curve.

Page 21 of 21

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