ML023230354
| ML023230354 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/07/2002 |
| From: | Nazar M Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML023230354 (42) | |
Text
Committed to N Mctear Excel Nula xclec November 7, 2002 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50-282 License No. DPR-42 Docket No. 50-306 License No. DPR-60 Pressure and Temperature Limits Report (PTLR) Revisions 2 and 3 Attached are approved copies of Revisions 2 and 3 to the Pressure and Temperature Limits Report (PTLR) for the Prairie Island Nuclear Generating Plant (PINGP). These revisions are submitted in accordance with the provisions of Technical Specification 5.6.6.c.
Revision 2 was issued to make the PTLR consistent with Technical Specifications (TS) license amendments (LA) 158/149, conversion to Improved Technical Specifications.
The Revision 2 changes included revision of references to TS to conform to LA 158/149 and addition of subsections for pressurizer and steam generator pressure and temperature limits.
Revision 3 was issued to add approval of the Site Director of Engineering, correct the maximum pressurizer cooldown rate and conform references to the Technical Requirements Manual (TRM) to the latest revision of the TRM.
Mano K. Nazar Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Dr. East
- Welch MN 55089
USNRC NUCLEAR MANAGEMENT COMPANY November 7, 2002 Page 2 of 2 The Nuclear Management Company has not made new or revised existing Nuclear Regulatory Commission commitments in this letter or the attachments. Please contact Dale Vincent (651-388-1121) if you have any questions related to these reports.
Mano K. Nazar Site Vice Pres ent Prairie Island uclear Generating Plant c:
Regional Administrator - Region Ill, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC James Bernstein, State of Minnesota Attachments:
Pressure and Temperature Limits Report, Revision 2 Pressure and Temperature Limits Report, Revision 3
Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report Revision 2 (Effective until 35 EFPY)
Prepared by:
R'y Waterman Sr. Engineer Nuclear Engineering Reviewed by:
Gene Rckholt ile:z 9A Licensing Supervisor Approved by:
Ran W mack Programs Engineering Manager Date I o(.0oz.
Date 2,9-I De 4Z C
I I
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
Table of Contents Section 1.0 Purpose 2.0 Applicability 3.0 Operating Limits Over Pressure Protection System (OPPS) Enable Temperature Safety Injection (SI) Pump Disable Temperature RCS Pressure/Temperature (PIT) Limits Instrumentation Uncertainty for P/T Curves RCS Heatup/Cooldown Rate Limits Over Pressure Protection System (OPPS) PORV Setpoint RCS Minimum Temperature When Not Vented Minimum Boltup Temperature Pressurizer Temperature Limits Steam Generator Temperature/Pressure Limit 4.0 Discussion Adjusted Reference Temperature (ART)
End of Life Fluence Reference Temperature (RTpts)
Neutron Fluences (f)
Chemistry Factor (CF)
Reactor Vessel Material Surveillance Program Supplemental Data Tables Surveillance Data Credibility RCS Minimum Temperature When Not Vented Minimum Boltup Temperature 5.0 References 6.0 Tables and Figures Table 6.1 35 EFPY Heatup Data Points Table 6.2 35 EFPY Cooldown Data Points Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule Table 6.4 Prairie Island Unit I 1/4T and 3/4T ART Calculations at 35 EFPY Table 6.5 Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 35 EFPY Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY Page 1
1 2
2 2
3 3
3 3
4 4
4 4
5 5
6 6
7 7
7 7
8 8
9 10 11 12 13 14 15 16 i
Pressure and Temperature Limits Report Revision 2 (Effective untl 35 EFPY)
Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations Applicable to 35 EFPY 17 ii
Pressure and Temperature Umits Report Revision 2 (Effective until 35 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 I 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 35 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 TS 3.4.6 TS 3.4.7 TS3.4.10 TS 3.4.12 TS 3.4.13 TS 3.5.3 TRM 3.10.1 RCS Pressure and Temperature (P/T) Limits RCS Loops - MODE 4 RCS Loops - MODE 5, Loops Filled Pressurizer Safety Valves Low Temperature Overpressure Protection (LTOP) - Reactor "Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature Low Temperature Overpressure Protection (LTOP)- Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature ECCS - Shutdown Miscellaneous Specifications-Technical Requirements Manual 1
Pressure and Temperature Limits Report Revision 2 (Effective until 35 EFPY) 3.0 Operating Limits All limits are valid until 35 EFPY, which is projected to be beyond the expiration of the operating license for each of Prairie Island Units I and 2.
Over Pressure Protection System (OPPS) Enable Temperature 310 OF*
- Analytical limit [225 OF] plus indicating instrument channel uncertainty [18 °F]
(Reference 5.11) plus additional margin for operational simplicity.
Safety Iniection (SIh Pump Disable Temperature 218 OF *
- Analytical limit [200 OF] plus indicating instrument channel uncertainty [18 IF]
(Reference 5.11).
2 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
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
RCS Pressure/Temperature (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: Tables 6.1 and 6.2 contain a tabulated version of the curves.
Instrumentation Uncertainty for P/T Curves 124 psig Pressure Uncertainty 18 OF Temperature Uncertainty Note: These values must be applied to the P/T limit curves in 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 Referenced in: TS 3.4.3, 1
1 SR 3.4.3.1 Over Pressure Protection System (OPPS) PORV Setpoint 500 psig*
Referenced in: TS 3.4.12, TS 3.4.13
- 8)
- This setpoint accounts for instrument channel uncertainty (Reference 5.8).
3
Pressure and Temperature Limits Report Revision 2 (Effective until 35 EFPY)
RCS Minimum Temperature When Not Vented 86 OF*
Referenced in: TS 3.4.3, TS 5.6.6
- Analytical limit [68 0FJ plus indicating instrument channel uncertainty [1 80F]
(Reference 5.11)
Minimum Boltup Temperature 60 OF**
Referenced in: ITS 5.6.6
- No instrument uncertainty included.
Pressurizer Temperature Limits 100 OF per hour Maximum Pressurizer Heatup Rate 100 OF per hour Maximnum Pressurizer Cooldown Rate 320 OF Maximum Temperature Difference Between the Pressurizer and the Spray Fluid for which the Pressurizer Spray can be used.
i Referenced in: TRM Miscellaneous I
I I Specification 3.10.1 Steam Generator Temperature/Pressure Limit 200 psig Maximum secondary side Pressure if the temperature of the steam generator is below 70 OF.
Referenced in: TRM Miscellaneous I Specification 3.10.1 4
I II
Pressure and Temperature Umits Report Revision 2 (Effechve until 35 EFPY) 4.0 Discussion This PTLR for Prairie Island Units I 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 beitline 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 I 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 I and 2 Reactor Vessel Radiation Surveillance Program Data. The surveillance program capsules were removed as indicated in Table 6.3.
Adiusted Reference Temperature (ART)
The adjusted reference temperature is the reference temperature (as defined in the ASME Boiler and Pressure Vessel Code, Section Xl, 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 I/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 is the Unit I nozzle to intermediate shell forging circumferential weld material (Table 6.4 and 6.5). The limiting ARTs are as follows:
114T = 154 OF 3/4T = 136 OF
References:
5.35 5
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
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 beitline materials in both Prairie Island Units I and 2. The projected end of life for both units is 35 Effective Full Power Years (35 EFPY). Comparison of RTpts for all materials indicates that the limiting material is the Unit I nozzle to intermediate shell forging circumferential weld material. The limiting RTpL, is as follows:
RTpts = 162 OF
References:
5.4 1
15.7 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.
The neutron fluences used in determining the 35 EFPY limiting ART for the reactor vessels are as follows:
Units are 1019 n/cm2, for energies > 1.0 MeV at 35 EFPY Clad/Base Metal Interface = 2.2 1/4T= 1.47 3/4T= 0.66
References:
5.2 1
15.5 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 Unit I nozzle to intermediate shell forging circumferential weld. The highest fluences were obtained at the unit 2 intermediate to lower shell forging circumferential weld. (Reference 5. 5).
6
Pressure and Temperature Umits Report Revision 2 (Effective until 35 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 are as follows.
1/4T = 79.5 OF 3/4T = 79.5 OF
References:
5.3 1
15.6 Reactor Vessel Material Surveillance Pro-ram 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 I and Unit 2 respectfully, including all the parameters.
References:
5.3
'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, Revisi6n 2, Position 1.1, the surveillance data must be used. If the surveillance 7
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY) capsule data gives lower values, either may be used. In the case of the Prairie Island limiting material, the Unit I nozzle to intermediate shell forging circumferential weld, surveillance data was not available and Position 1.1 of Regulatory Guide 1.99, Revision 2, was applied. 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.
8
Pressure and Temperature Umits Report Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY) 5.0 References 5.1 WCAP-14040-NP-A, Methodologqy Used to Develop Cold Overpressure Mitigation, Revision 2, January 1996.
5.2 WCAP-14779, Analysis of Capsule S from the Northern States Power Company Prairie Island Unit I 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-1 4613, 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 I Reactor Cooiant Hot Leg Pressure Control Room Indication at I PR-420 (0-750 psig scale) with 2 RC Pumps Running," Revision 0.
5.11 NSP Calculation No. SPCRC003, 'Unit I Wide Range RCS Cold Leg Temperature Control Room Indication Loop 1T-450B Uncertainty with Streaming Effects," Revision 0.
9
Pressure and Temperature Umits Report RevisbDn 2 (Effective until 35 EFPY) 6.0 Tables and Figures Table 6.1 Table 6.2 Table 6.3 Table 6.4 Table 6.5 Figure 6.1 Figure 6.2 35 EFPY Heatup Data Points 35 EFPY Cooldown Data Points Reactor Vessel Material Surveillance Capsule Removal Schedule.
Prairie Island Unit 1 1/4T and 314T ART Calculations at 35 EFPY Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 35 EFPY Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY.
Prairie Island Reactor Coolant System Cooldown Limitations Applicable to 35 EFPY.
10
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
TABLE 6.1 35 EFPY Heatup Data Points (Without Instrumentation Uncertainty Margins)
Heatun Curves 60 Critical. Limit 100 Heatup Critical. Limit Leak Test Limit Heatup 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 11
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
TABLE 6.2 35 EFPY Cooldown Data Points (Without Margins for Instrumentation Uncertainty)
Cooldown Curves Steady State 20 deg F 40 degF 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 12
Pressure and Temperature Limits Report Revision 2 (Effective until 35 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 (degree)
Lead Factor(a)
EFPY(b)
(n/cm 2, E> 1.0 MeV)
V 77 2.94 1.34 5.630 x 101"c' P
247 1.72 4.60 1.318 x 10"1Yc)
R 257 2.99 8.56 4.478 x 10 u(c)
S 57 1.77 18.12 4.017 x 1OIc)
T 67 1.89 Standby N
237 1.77 Standby Recommended Surveillance Capsule Removal Schedule for Unit 2 Capsule Location Withdrawal Fluence(d)
Capsule (degree)
Lead Factor(d)
EFpyfb)
(n/cm2, E> 1.0 MeV)
V 77 2.95 1.39 6.206 x 1016c' T
67 1.75 4.00 1.199 x 10V1c)
R 257 2.99 8.81 4.376 x 101uc)
P 247 1.84 17.24 4.165 x 10I9(c)
N 237 1.72 Standby S
57 1.72 Standby I
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.
13
Pressure and Temperature Umits Report Revision 2 (Effectve until 35 EFPY)
Table 6.4 Prairie Island Unit I 114T and 314T ART Calculations at 35 EFPY Material CF If@35 11/4T f 14TTFF I, e) M IRTND ART EFPY(a) 3/4Tf 3/4TFF IOF) I (OF)
T T(OF)
(OF) 1/4T Calculations Nozzle (upper)Shell 51 2.20 1.47 1.11
-4 34 56.6 87 Forging B Nozzle to Inter. Shell 79.5 2.20 1.47 1.11 0 c) 66 88.2 154 Circ Weld (Heat 2269)
Intermediate Shell 44.0 3.95 2.64 1.26 14 34 55.4 103 Forging C Using S/C Data 54.7 3.95 2.64 1.26 14 3 4 (b) 68.9 117 Circumferential Weld 69.7 3.95 2.64 1.26
-13 56 87.8 131 Using S/C Data 80.8 3.95 2.64 1.26
-13 5 6(b) 101.8 145 Lower Shell Forging D 44.0 3.95 2.64 1.26
-4 34 55.4 85 3/4T Calculations Nozzle (upper) Shell 51 2.20 0.660 0.884
-4 34 45.1 75 Forging B Nozzle to Inter. Shell 79.5 2.20 0.660 0.884 OcF 66 70.3 136 Circ Weld (Heat 2269)
Intermediate Shell 44.0 3.95 1.18 1.05 14 34 46.2 94 Forging C Using S/C Data 54.7 3.95 1.18 1.05 14 3 4 (b) 57.4 105 Circumferential Weld 69.7 3.95 1.18 1.05
-13 56 73.2 116 Using SIC Data 80.8 3.95 1.18 1.05
-13 5 6 (b) 84.8 128 Lower Shell Forging D 44.0 3.95 1.18 1.05
-4 34 46.2 76 NOTE:
(a)
Fluence values (f) are x 1019 n/cm 2 (E > 1.0 MeV). In addition, the values used are the calculated values since they are higher than the best-estimate values (Ref. 5.3).
(b)
The full oA margin of 170F 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*iogf) (Ref. 5.3)
(e)
I is the unirradiated material reference temperature. (Ref. 5.3)
(g)
M is a margin term required for conservative results. (Ref. 5.3) 14
Pressure and Temperature Umits Report Revision 2 (Effective until 35 EFPY)
Table 6.5 Prairie Island Unit.2 114T and 314T ART Calculations at 35 EFPY Material CF f@35 1/4Tf 1/4T F
,c)
I 1ARTNDIART EFPY 3/4Tfa) 314T FF (OF)
(OF)
T (OF)
(OF) 1/4T Calculations Upper.Shell Forging B 44.0 2.379 1.59 1.13
-13 34 49.7 71 Upper to Inter. Shell Weld 70.0 2.379 1.59 1.13
-13 56 79.1 122 W2 Using Unit 1 SIC Data(b) 80.8 2.379 1.59 1.13
-13 56 91.3 134 Intermediate Shell Forging 44.0 4.183 2.80 1.27 14 34 55.9 104 C
Inter. to Lower Shell Weld 52.0 4.183 2.80 1.27
-31 56 66.0 91 W3 Using S/C Data 80.0 4.183 2.80 1.27
-31 28 101.6 99 Lower Shell Forging D 51.0 4.183 2.80 1.27
-4 34 64.8 95 Using S/C Data 60.0 4.183 2.80 1.27
-4 34 76.2 106 3/4T Calculations Upper Shell Forging B 44.0 2.379 0.71 0.90
-13 34 39.7 61 Upper to Inter. Shell Weld 70.0 2.379 0.71 0.90
-13 56 63.0 106 W2 Using Unit 1 SIC Data ()
80.8 2.379 0.71 0.90
-13 56 72.7 116 Intermediate Shell Forging 44.0 4.183 1.25 1.06 14 34 46.6 95 C
I I
Inter. to Lower Shell Weld 52.0 4.183 1.25 1.06
-31 55.1 55.1 79 W3 Using S/C Data 80.0 4.183 1.25 1.06
-31 28 84.8 82 Lower Shell Forging D 51.0 4.183 1.25 1.06
-4 34 54.1 84 Using SIC Data 60.0 4.183 1.25 1.06 1 -4 34 63.6 94 NOTE:
(a)
Fluence values (f) are x 1019 n/cm2 (E > 1.0 MeV). In addition, the values used are the Best Estimate values since they are higher than the calculated values.
(b)
This calculation is using the chemistry factor based on the surveillance capsule data for the Prairie Island Unit I 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(O.28-O.1*logf). (Ref. 5.3)
(d)
I is the unirradiated material reference temperature. (Ref. 5.3)
(e)
M is a margin term required for conservative results. (Ref. 5.3) 15
-I Pressure and Temperature Llmi.
sort Revision 2 (Effective until 35 EFPY)
Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY (w/o Margins for Instrument Uncertainty) 200 250 300 350 400 Indicated Temperature (Deg. F)
- For each curve, acceptable operation Is to the right and below the curve.
16 2500 2000 1500 VI
- 0.
( 1000
.2 500 0
Inservice Hydrostatic Pressure Test 1
.V [
Heatup Rate
-Upto 60 F/Hr JUNACCEPTABLE CT 1
OPERATION Heatup Rate
- 7 up to 100 F/Hr S--
Critical*
"__/
-ACCEPTABLE SHeatup Rate
-7 OPERATION 0
Upto60F/Hr Subcritlcal*
Criticality Limit Based on Inservice Hydrostatic Test Heatup Rate Temperature (238 F) For the
-Up to100F/Hr Service Perlod up to 35 EFPY Subrltcal
- T-.
- p.
0 50 100 150
,4 Pressure and Temperature Lmits...,port Revision 2 (Effective until 35 EFPY)
Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 35 EFPY (wlo Margins for Instrument Uncertainty)
UNACCEPTABLE OPERATION 150
/
200 1800 1600 1400
-. 1200
"- 1000 800 U
0 rn 250 300 350 Indicated Temperature (Deg. F)
- For each curve, acceptable operation Is to the right and below the curve.
17 ACCEPTABLE OPERATION 400 200 0-Cooldown Rates
- D., F/ Hr 0
20 40 60 100 I-100 1
-. 4--
ý I oltup T
Temp.I 0
50 m
I I
Prairie Island Nuclear Generating Plant Units One and Two Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
Prepared by:
Reviewed by:
Reviewed by:
Approved by:
Ro Waterman Sr. Engineer Nuclear Engineering Gene E-kholt Licensing Supervisor Randy ý ack Programs Engineering Manager Scott Northard Site Director of Engineering Date Date iohe~lzco7-I Date 10 -- 1-02.
Date I
Table of Contents Purpose 2.0 Applicability 3.0 Operating Limits Over Pressure Protection System (OPPS) Enable Temperature Safety Injection (SI) Pump Disable Temperature RCS Pressure/Temperature (PIT) Limits Instrumentation Uncertainty for P/T Curves RCS Heatup/Cooldown Rate Limits Over Pressure Protection System (OPPS) PORV Setpoint RCS Minimum Temperature When Not Vented Minimum Boltup Temperature Pressurizer Temperature Limits Steam Generator Temperature/Pressure Limit Discussion Adjusted Reference Temperature (ART)
End of Life Fluence Reference Temperature (RTpts)
Neutron Fluences (f)
Chemistry Factor (CF)
Reactor Vessel Material Surveillance Program Supplemental Data Tables Surveillance Data Credibility RCS Minimum Temperature When Not Vented
- Minimum Boltup Temperature References Tables and Figures Table 6.1 Table 6.2 Table 6.3 Table 6.4 Table 6.5 Figure 6.1 35 EFPY Heatup Data Points 35 EFPY Cooldown Data Points Reactor Vessel Material Surveillance Capsule Removal Schedule Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 35 EFPY Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 35 EFPY Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY Section 1.0 Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
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4.0 5.0 6.0 1
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10 11 12 13 14 15 16 i
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations Applicable to 35 EFPY 17 ii
Pressure and Temperature Limits Report Revision 3 (Effective until 35 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 heafup 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 35 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 TS 3.4.6 TS 3.4.7 TS 3.4.10 TS 3.4.12 TS 3.4.13 TS 3.5.3 TRM 3.4.4 TRM 3.4.5 RCS Pressure and Temperature (P/T) Limits RCS Loops - MODE 4 RCS Loops - MODE 5, Loops Filled Pressurizer Safety Valves Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) - Safety Injection (SI) Pump Disable Temperature ECCS - Shutdown Miscellaneous Specifications - Technical Requirements Manual PTLR Compliance - Pressurizer PTLR Compliance - Steam Generator(s) 1 I
I I
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY) 3.0 Operating Limits All limits are valid until 35 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 OF*
- Analytical limit [225 OF] plus indicating instrument channel uncertainty [18 OF]
(Reference 5.11) plus additional margin for operational simplicity.
Safety Iniection (SI) Pump Disable Temperature 218 OF
- Analytical limit [200 OF] plus indicating instrument channel uncertainty [18 OF]
(Reference 5.11).
2 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
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
RCS Pressure/Temperature (PIT) Limits Figure 6.1*
RCS P/T limits for heatup Figure 6.2*
RCS P/T 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 124 psig Pressure Uncertainty 18 OF Temperature Uncertainty Note: These values must be applied to the P/T limit curves in operating procedures (Reference 5.10 and 5.11).
RCS Heatup/Cooldown Rate Limits 100.-F per hour Maximum RCS Heatup Rate 100 °F per hour Maximum RCS Cooldown Rate Referenced in: TS 3.4.3, 1
1 SR 3.4.3.1 Over Pressure Protection System (OPPS) PORV Setpoint 500 psig*
Referenced in: TS 3.4.12I
- This setpoint accounts for instrument channel uncertainty (Reference 5.8).
3
Pressure and Temperature limits Report Revision 3 (Effective until 35 EFPY)
RCS Minimum Temperature When Not Vented 86 OF*
I Referenced in: TS 3.4.3, 1
1 TS 5.6.6
- Analytical limit [680F] plus indicating instrument channel uncertainty [1 80F]
(Reference 5.11)
Minimum Boltup Temperature 60 OF**
Referenced in: I TS 5.6.6
- No instrument uncertainty included.
Pressurizer Temperature Limits 100 OF per hour Maximum Pressurizer Heatup Rate 200 OF per hour Maximum Pressurizer Cooldown Rate 320 OF Maximum Temperature Difference Between the Pressurizer and the Spray Fluid for which the Pressurizer Spray can be used.
Referenced in: I TRM Specification 13.4.4 II Steam Generator Temperature/Pressure Limit 200 psig Maximum secondary side Pressure if the temperature of the steam generator is below 70 OF.
Referenced in: TRM Specification 13.4.5 II 4
I P
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Pressure and Temperature Limits Report Revision 3 (Effective until 35 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.
- Adiusted 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 is the Unit 1 nozzle to intermediate shell forging circumferential weld material (Table 6.4 and 6.5). The limiting ARTs are as follows:
1/4T= 154 OF 3/4T = 136 °F
References:
j 5.3 1
1 5.6 5
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
End of Life Fluence Reference Temperature (RTptl 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 35 Effective Full Power Years (35 EFPY). Comparison of RTpts for all materials indicates that the limiting material is the Unit 1 nozzle to intermediate shell forging circumferential weld material. The limiting RTpts is as follows:
RTpts = 162 'F
References:
I5.4 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.
The neutron fluences Used in determining the 35 EFPY limiting ART for the reactor vessels are as follows:
Units are 1019 n/cm2, for energies > 1.0 MeV at 35 EFPY Clad/Base Metal Interface = 2.2 1/4T= 1.47 3/4T= 0.66
References:
I 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 Unit 1 nozzle to intermediate shell forging circumferential weld. The highest fluences were obtained at the unit 2 intermediate to lower shell forging circumferential weld. (Reference 5. 5).
6
Pressure and Temperature Umits Report Revision 3 (Effective until 35 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 are as follows.
1/4T = 79.5 °F 3/4T = 79.5 'F
References:
j 5.3 15.6 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.
I
References:
5.3 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 7
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY) capsule data gives lower values, either may be used. In the case of the Prairie Island limiting material, the Unit 1 nozzle to intermediate shell forging circumferential weld, surveillance data was not available and Position 1.1 of Regulatory Guide 1.99, Revision 2, was applied. 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.
8
Pressure and Temperature Umits Report Revision 3 (Effective until 35 EPPY) 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 Capsule S 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 I 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 1 PR-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.
9
Pressure and Temperature Umits Report Revision 3 (Effective until 35 EFPY) 6.0 Tables and Fiqures Table 6.1 Table 6.2 Table 6.3 Table 6.4 Table 6.5 Figure 6.1 Figure 6.2 35 EFPY Heatup Data Points 35 EFPY Cooldown Data Points Reactor Vessel Material Surveillance Ca~psule Removal Schedule.
Prairie Island Unit 1 1/4T and 3/4T ART Calculations at 35 EFPY Prairie Island Unit 2 1/4T and 3/4T ART Calculations at 35 EFPY Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY.
Prairie Island Reactor Coolant System Cooldown Limitations Applicable to 35 EFPY.
10
Pressure and Temperature Umits Report Revision 3 (Effective until 35 EFPY)
TABLE 6.1 35 EFPY Heatup Data Points (Without Instrumentation Uncertaintv Marains)
Heatup Curves 60 Critical. Umit 100 Heatup Critical. Limit Leak Test Limit Heatup MT P -
T T
P T
P T
P 60 60 65 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 0
584 584 584 584 584 586 591 597 604 613 622 633 645 658 672 687 704 722 741 761 784 808 833 861 891 923 957 994 1033 1076 1121 1170 1223 1279 1339 1404 1473 1548 1628 1713 1805 1903 2007 2119 2231 2347 2471 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 0
594 587 584 584 586 591 597 604 613 622 633 645 658 672 687 704 722 741 761 784 808 833 861 891 923 957 994 1033 1076 1121 1170 1223 1279 1339 1404 1473 1548 1628 1713 1805 1903 2007 2119 2231 2347 2471 60 60 65 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315
- 0.
560 560 560 560 560 560 560 560 562 566 571 577 585 594 604 615 627 641 656 672 690 709 730 752 777 802 831 861 893 928 966 1006 1049 1096 1146 1199 1257 1318 1384 1455 1531 1612 1699 1792 1892 1998 2112 2233 2363 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 273 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 0
560 560 560 560 560 560 560 562 566 571 577 585 594 604 615 627 641 656 672 690 709 730 752 777 802 831 861 893 928 966 1006 1049 1096 1149 1199 1257 1318 1384 1455 1531 1612 1699 1792 1892 1998 2112 2233 2363 251 273 2000 2485 11
Pressure and Temperature Umits Report Revision 3 (Effective until 35 EFPY)
TABLE 6.2 35 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
T PP T
P T '
P
__x----- P 12 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
Pressure and Temperature Limits Report Revision 3 (Effective until 35 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 (degree)
Lead Factor(a)
EFPY°)
(n/cm2, E> 1.0 MeV)
V 77 2.94 1.34 5.630 x 10 Q1c)
P 247 1.72 4.60 1.318 x 10 ~c)
R 257 2.99 8.56 4.478 x 1 0uc)
S 57 1.77 18.12 4.017 x 10Vuc)
T 67 1.89 Standby N
237 1.77 Standby I
Recommended Surveillance Capsule Removal Schedule for Unit 2 Capsule Location Withdrawal Fluence(d)
Capsule
'(degree)
Lead Factor(d)
EFPY(b)
(n/cm 2, E> 1.0 MeV)
V 77 2.95 1.39 6.206 x 1 0Q(c)
T 67 1.75 4.00 1.199 x 10l 1c)
R 257 2.99 8.81 4.376 x 1091c)
P 247 1.84 17.24 4.165 x 10Vuc)
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.
13
Pressure and Temperature Limits Report Revision 3 (Effective until 35 EFPY)
Table 6.4 Prairie Island Unit I 1/4T and 314T ART Calculations at 35 EFPY Material CF I 35 II14TfI 1/4T Fd I I(e)
Mg (I
(RTN ART EFPY(a) 3/4Tf3/4T FF (F)
(OF)
T AOF R
(OF) 1/4T Calculations Nozzle (upper)Shell 51 2.20 1.47 1.11
-4 34 56.6 87 Forging B Nozzle to Inter. Shell 79.5 2.20 1.47-1.11 0*c) 66 88.2 154 Circ Weld (Heat 2269)
Intermediate Shell 44.0 3.95 2.64 1.26 14 34 55.4 103 Forging C Using S/C Data 54.7 3.95 2.64 1.26 14 3 4 (b) 68.9 117 Circumferential Weld 69.7 3.95 2.64 1.26
-13 56 87.8 131 Using S/C Data 80.8 3.95 2.64 1.26
-13 5 6(b) 101.8 145 Lower Shell Forging D 44.0 3.95 2.64 1.26
-4 34 55.4 85 3/4T Calculations Nozzle (upper) Shell 51 2.20 0.660 0.884
-4 34 45.1 75 Forging B Nozzle to Inter. Shell 79.5 2.20 0.660 0.884 0tcF 66 70.3 136 Circ Weld (Heat 2269)E Intermediate Shell 44.0 3.95 1.18 1.05 14 34 46.2 94 Forging C Using S/C Data 54.7 3.95 1.18 1.05 14 3 4 (b) 57.4 105 Circumferential Weld 69.7 3.95 1.18 1.05
-13 56 73.2 116 Using S/C Data 80.8 3.95 1.18 1.05
-13 5 6 (b) 84.8 128 Lower Shell Forging D 44.0 3.95 1.18 1.05
-4 34 46.2 76 NOTE:
(a)
Fluence values (f) are x 1019 n/cm2 (E > 1.0 MeV). In addition, the values used are the calculated values since they are higher than the best-estimate values (Ref. 5.3).
(b)
The full Y& margin of 170F for the forging and 280F 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 *logf) (Ref. 5.3)
(e)
I is the unirradiated material reference temperature. (Ref. 5.3)
(g)
M is a margin term required for conservative results. (Ref. 5.3) 14
Pressure and Temperature Umits Report Revision 3 (Effective until 35 EFPY)
Table 6.5 Prairie Island Unit 2'114T and 3/4T ART Calculations at 35 EFPY Material
~
/4T f 1/4T F I
M I(O)ITND ART EFPY 3/4TP fa)/4T FF (OF) I (°F) I- (F)
(°F) 1/4T Calculations Upper Shell Forging B 44.0 2.379 1.59 1.13
-13 34 49.7
.71 Upper to Inter. Shell Weld 70.0 2.379 1.59 1.13
-13 56 79.1 122 W2 Using Unit 1 SIC Data(b) 80.8 2.379 1.59 1.13
-13 56 91.3 134 Intermediate Shell Forging 44.0 4.183 2.80 1.27 14 34 55.9 104 C
I Inter. to Lower Shell Weld 52.0 4.183 2.80 1.27
-31 56 66.0 91 W3 Using S/C Data 80.0 4.183 2.80 1.27
-31 28 101.6 99 Lower Shell Forging D 51.0 4.183 2.80 1.27
-4 34 64.8 95 Using S/C Data 60.0 4.183 2.80 1.27
-4 34
.76.2 106 3/4T Calculations Upper Shell Forging B 44.0 2.379 0.71 0.90
-13 34 39.7 61 Upper to Inter. Shell Weld 70.0 2.379 0.71 0.90
-13 56 63.0 106 W2 Using Unit 1 S/C Data(b) 80.8 2.379 0.71 0.90
-13 56 72.7 116 Intermediate Shell Forging 44.0 4.183 1.25 1.06 14 34 46.6 95 C
I Inter. to Lower Shell Weld 52.0 4.183 1.25 1.06
-31 55.1 55.1 79 W3 Using S/C Data 80.0 4.183 1.25 1.06
-31 28 84.8 82 Lower Shell Forging D 51.0 4.183 1.25 1.06
-4 34 54.1 84 Using SIC Data 60.0 4.183 1.25 1.06
-4 34 63.6 94 NOTE:
(a)
Fluence values (f) are x 1019 n/cm2 (E > 1.0MeV). In addition, the values used are the Best Estimate values since they are higher than the calculated values.
(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 ac, of 28°F was used in the margin term.
(c)
FF, Fluence Factor = f(0.28-0.1lIogf). (Ref. 5.3)
(d)
I is the unirradiated material reference temperature. (Ref. 5.3)
(e)
M is a margin term required for conservative results. (Ref. 5.3) 15
Pressure and Temperature Umits '
t Revision 3 (Effective until 35 L
,)
Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY (w/o Margins for Instrument Uncertainty) 2500 2000
- 0) a 1500 1000 50 100 150 200 250 300 400 Indicated Temperature (Deg. F)
I For each curve, acceptable operation Is to the right and below the curve.
16 Inservice Hydrostatic Pressure Test I
I __
r Heatup Rate
~---
~-.
I
-1~*-
Up to 60FIHr
_.-UNACCEPTABLE.
Critical*
.OPERATION-.
Heatup Rate
_ ~
Up to 100 F/Hr S...
Cri-tical*
S...
+/- OPERATION
-*-*', /1i iACCEPTABLE Heatup Rate OPERATION S....
Up to 60 F/Hr
/
Subcritical*
7 Criticality Limit Based on InservIce Hydrostatic Test Heatup Rate Temperature (238 F) For the
-Up to 100 F/Hr -
-i Service Period up to 35 EFPY, B )1tupSubcritical*
Temp.
I M~t 350 500 0
0 350
-I Pressure and Temperature Limits l' Revision 3 (Effective until 35 E, s Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 35 EFPY (w/o Margins for Instrument Uncertainty) 1800 1600 1400 1200 S1000
- 0.
S800 Is
.2
-V n
P h-Cooldown Rates*
Dea, F / Hr 0
20 40 60 100 9 Boitup Temp.
0 50 100
-J UNACCEPTABLE[
OPERATION lII lf 150 200
.4
'I/
1/
I
/ I
-i->-]A--f-+
i--I -H
_-H-H ACCEPTABLE' OPERATION
-t indicated Temperature (Deg. F)
Indicated Temperature (Deg. F)
For each curve, acceptable operation Is to the right and below the curve.
17 250 300 350
-I-400 200 0-
- i I.
I i I
--- t---i m
m 1,
I I
I I
I
- "i-IT-I I
I I..
S....
T-T 4#VV I
I i
i l
-- I--T--"
I I