ML20247E267
| ML20247E267 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/05/1998 |
| From: | Breene T, Eckholt G, Waterman R NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20247E264 | List: |
| References | |
| NUDOCS 9805180283 | |
| Download: ML20247E267 (19) | |
Text
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l Prairie Island Nuclear Generating Plant Units One and Two i
Pressure and Temperature Limits Report Revision 0 (Effective until 35 EFPY) l Prepared by: /An:_ SS 7 RpfWaterman D' ate Sr. Engineer Nuclear Engineering Reviewed by: bm b9pDE. 5/r/Sg Gene Bckholt Date Licensing Project Manager Reviewed by: Me [h- TM///
Tom Breene Date Superintendent of Nuclear Engineering Approved by: esA /n If.A. MS/98 lid' mundson O date' General Superintendent of Engineering l
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9805180283 980506 PDR ADOCK 05000282 P PDR ,
Pressure and Temperature umrts Report Ravison 0 (Effectrve untd 35 EFPY)
Table of Contents Section Page 1.0 Purpose 1 l
2.0 Applicability 1 l
3.0 Operating Limits 3 Over Pressure Protection System (OPPS) Enable Temperature 3 j Temperature for Disabling Both Safety injection Pumps 3 l RCS Pressure / Temperature (P/T) Limits 3 Instrumentation Uncertainty for P/T Curves 4 RCS Heatup/Cooldown Rate Limits 4 Over Pressure Protection System (OPPS) PORV Setpoint 4 RCS Minimum Temperature When Not Vented 4 Minimum Boltup Temperature 4 4.0 Discussion 5 Adjusted Reference Temperature (ART) 5 End of Life Fluence Reference Temperature (RTpts) 6 Neutron Fluences (f) 6 Chemistry Factor (CF) 7 Reactor Vessel Material Surveillance Program 7 Supplemental Data Tables 7 Surveillance Data Credibility 7 RCS Minimum Temperature When Not Vented 8 Minimum Boltup Temperature 8 5.0 References 9 6.0 Tables and Figures 10 Table 6.1 35 EFPY Heatup Curve Data Points 11 Table 6.2 Cooldown Curves Data Points Applicable to 35 EFPY 12 Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedt.le 13 Table 6.4 Prairie Island Unit 11/4T and 3/4T ART Calculations
- at 35 EFPY 14 l
Table 6.5 Prairie Island Unit 21/4T and 3/4T ART Calculations at 35 EFPY 15 Figure 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY 16 Figure 6.2 Prairie Island Reactor Coolant System Cooldown Limitations to 35 EFPY 17 i
Prassure and Tsmper ture Limits Rtport Revision O (Effectrvs untd 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, (2) heatup and cooldown rates, and (3) the Over Pressure Protection System (OPPS) setpoint. This report has l been prepared in accordance with the requirements of Technical Specification 6.7.A.7.
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: l l
3.1. A.1.c(4) Reactor Coolant System - Operational Components - Reactor Coolant System Average Temperature Below 350 F (and Reactor Coolant Level Above the Reactor Vessel Flange).
3.1. A.2.c(2) Reactor Coolant System - Operational Components -
Pressurizer Power Operated Relief Valves - Reactor Coolant System average temperature greater than or equal to the temperature specified in PTLR for disabling both safety injection pumps and below the Over Pressure Protection System enable temperature specified in the PTLR.
3.1. A.2.c(3) Reactor Coolant System - Operational Components -
Pressurizer Power Operated Relief Valves - Reactor Coolant System average temperature below the temperature specified in PTLR for disabling both safety injection pumps.
3.1. B.1. 3 Reactor Coolant System - Pressure / Temperature Limits - !
- 3.3.A.3 Engineered Safety Features - Safety Injection and Residual .'
Heat Removal Systems -
3.3.A.4 Engineered Safety Features - Safety injection and Residual Heat Removal Systems.
3.3.A.5 Engineered Safety Features - Safety injection and Residual Heat Removal Systems.
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I I* Pressurs and Tamperrtura umits Report
, Revision 0 (Effectwo until 35 EFPY)
Table TS 4.1-1c Miscellaneous Instrumentation Surveillance Requirements -
Table Notation - (38).
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l Prsseurs and Temper:tura Limits Report Revision 0 (Effectivs until 35 EFPY) 3.0 Operatina Limits ,
1 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 F*
Referenced in: TS 3.1.A.1.c(4),
TS 3.1.A.2.c(2),
TS 3.3.A.3, TS 3.3.A.5, and Table TS 4.1-1c
- Analytical limit [225 *F) plus indicating instrument channel uncertainty [18 F]
(Reference 5.11) plus additional margin for operational simplicity.
Temperature for Disablina both Safety iniection Pumos 218 F
- Referenced in: ' TS 3.1. A.2.c(2),
TS 3.1.A.2.c(3), and
, TS 3.3. A.4
- Analytical limit [200 F) plus indicating instrument channel uncertainty (18 F)
(Reference 5.11).
RCS Pressure / Temperature (P/T) Limits Figure 6.1* RCS P/T limits for heatup Figure 6.2* RCS P/T limits for cooldown "Neierenced in: [T5~5.1.B.1.a
- Figures are analytical limits and do not include instrumentation uncertainty.
Note: Tables 6.1 and 6.2 contain a tabulated version of the curves.
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Pressure and Tampartture umits Rrport Revision 0 (E*fsctus untd 35 EFPY) l instrumentation Uncertainty for P/T Curves 124 psig Pressure Uncertainty 18 F Temperature Uncertainty Note: These values must be applied to the P/T limit curves in operating procedures (Reference 5.10 and 5.11).
RCS Heatup/Cooldown Rate Limits 100 F per hour Maximum RCS Heatup Rate 100 F per hour Maximum RCS Cooldown Rate
[" leierenced in:]TS 3.1.B.1.a l Over Pressure Protection System (OPPS) PORV Setooint 500 psig*
- This setpoint accounts for instrument channel uncertainty (Reference 5.8).
RCS Minimum Temperature When Not Vented 86 F*
- Analyticallimit [68 F] plus indicating instrument channel uncertainty [18 F]
(Reference 5.11)
Minimum Boltup Temperature 60 F**
"No instrument uncertainty included.
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Prsteurs cnd Temper:turs Umrta R1 port Revision O (Effsetiva untd 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 6.7.A.7. 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. j 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.
I 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 l 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: l 1/4T = 154 F 3/4T = 136 F I
References:
~5'.'3 i 5.6 5
. Preature end Temperature Limits Rrport Revision 0 (Effectue untd 35 EFPY) 4 End of Life Fluence Reference Temperature (RTg}
The RT,,, 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 material in both Prairie Island Units 1 and 2. The projected end of life for the both units is 35 Effective Full Power Years (35 EFPY). Comparison of RT,,, for all materials indicates that the limiting material is the Unit 1 nozzle to intermediate shell forging circumferential weld material. The limiting RT,t, is as follows:
RT,,, = 162 F "Ikeferences: 5.4 5.7 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 Materia! 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 10" n/cm 2, 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 5.5 ,
Note: These values are not the highest fluences that were obtained in the reactor l vessels, but are the values determined for the most limiting material- the Unit 1 )
l nozzle to intermediate shell forging circumferential weld. The highest fluences l were obtained at the unit 2 intermediate to lower shell forging circumferential
! weld. (Reference 5. 5).
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. Pressure and Tempertturs Umsta Rtport Reveen 0 (Effectus untd 35 EFPY)
Chemistry Factor (CF)
I Chemistry Factors are parameters used in the development of the ARTS for the l beltline materials and account for the Copper and Nickel content in the reactor l vessel beltline materials. The chemistry factors determined for the limiting ARTS are as follows.
1/4T = 79.5 F l 3/4T = 79.5 F
References:
1 5.3
,5.6 ,
Reactor Vessel Material Surveillance Proaram 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:
5.3 5.6 ,
Surveillance Data Credibility 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, j Revision 2, Position 1.1, the surveillance data must be used. If the surveillance l capsule data gives lower values, either may be used. In the case of the limiting ;
material, the Unit 1 nozzle to intermediate shell forging circumferential weld, j l
surveillance data was not available and Position 1.1 of Regulatory Guide 1.99,
' Revision 2, was applied. For all beltline materials with surveillance data available credibility was determined (Reference 5.2 and 5.5).
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Prassura and Temperature Limds RIport Rsvtvon 0 (Effectnra until 35 EFPY) e 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 Boltuo 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.
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Pressure and Temper ture Umds Rtport Rowson 0 (Effactive untd 3s EFPY) 5.0 References 5.1 WCAP-14040-NP-A, Methodoloav Used to Develoo Cold Overpressure Mitiaation, Revision 2, January 1996.
5.2 WCAP-14779, Analysis of Caosule S from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Proaram, Revision 2 February 1998.
5.3 WCAP-14780, Prairie Island Unit 1 Heatuo 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, Anaivsis of Caosule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Proaram, Revision 2, February 1998.
5.6 WCAP-14637, Prairie Island Unit 2 Heatuo and Cooldown Limit Curves Normal Operation. Revision 2, February 1998.
5.7 WCAP-14638, Evaluation of Pressurized Thermal Shock for Prairie Island Unit 2, Revision 2, February 1998.
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" I 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.
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Pressure and Temperature Limits Report Revision 0 (Effective until 35 EFPY) 6.0 Fioures and Tables Table 6.1 35 EFPY Heatup Curve Data Points Table 6.2 Cooldown Curve Data Points Applicable to 35 EFPY Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule.
Table 6.4 Prairie Island Unit 11/4T and 3/4T ART Calculations at 35 EFPY Table 6.5 Prairie Island Unit 21/4T and 3/4T ART Calculations at 35 EFPY Fig >re 6.1 Prairie Island Reactor Coolant System Heatup Limitations Applicable to 35 EFPY.
Figure 6.2 Prairie Isisnd Pa9ctor Coolant System Cooldown Limitations to 36 EFFr.
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( Pressure and Temp:fttura Limits Riport Revision 0 (Effectiva until 35 EFF f)
TABLE 6.1 35 EFPY Heatup Curve Data Points (Without instrumentation Uncertainty Margins)
Heatuo Curves 60 Critical. Limit 100 Heatug 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 l 115 604 273 o13 115 562 273 566 f
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 17E 690 273 709 180 808 273 833 130 709 273 730 185 833 273 861 185 730 273 752 190 861 273 591 190 752 273 777 195 891 273 923 195 777 273 802 200 923 273 957 200 802 273 831 205 957 273 994 205 831 273 861 210 994 273 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 l 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 l
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. l Pressure and Ternperttura Lirruts Rrport Revision 0 (Effretive untd 35 EFPY)
TABLE 6.2 Cooldown Curves Data Points Applicable to 35 EFPY (Without Margins for Instrumentation Uncertainty) i 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 i 65 594 65 568 65 542 65 515 65 460 l 70 599 70 573 70 547 70 520 70 465 75 605 75 579 75 552 75 526 75 471 PO 611 80 585 80 558 80 532 80 478 65 617 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 i
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 l 140 726 140 704 150 715 150 695 150 653 720 714 674 145 741 145 155 734 155 155 150 757 150 736 160 754 160 735 160 697 155 774 155 754 165 776 165 757 165 722 1 160 793 160 773 170 799 170 782 170 748 l
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 l
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 l 250 1500 {
255 1572 1 260 1649 265 1732 270 1820 275 1915 j 280 2017 l 285 2126 290 2243 295 2367 l
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. Prtssure and Temp 3riture Limits Rrport Revision 0 (Effective until 35 EFPY)
Table 6.3 Reactor Vessel Material Surveillance Capsule Removal Schedule Recommended Surveillance Capsule Removal Schedule for Unit 1 Capsule l Location Withdrawal Fluence (')
l Capsule (degree) Lead Factor (') EFPY(*) (n/cm2, E> 1.0 MeV) l V 77 2.94 1.34 5.630 x 10**)
l P 247 1.72 4.60 1.318 x 10'*(
R 257 2.99 8.56 4.478 x 10**)
l S 57 1.77 18.12 4.017 x 10**'
T 67 1.89 Standby ---
N 237 1.77 Standby ---
Recommended Surveillance Capsule Removal Schedule for Unit 2 Capsule Location Withdrawal Fluence (*
Capsule (degree) Lead Factor (* EFPY(b) (n/cm2, E> 1.0 MeV)
V 77 2.95 1.39 6.206 x 10**)
T 67 1.75 4.00 1.199 x 10**)
R 257 2.99 8.81 4.376 x 10**) l P 247 1.84 17.24 4.165 x 10**'
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.
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_ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ ______.._______________j
. l Press ire and Ternpsrtture Lmts Raport
, Revision 0 (Effsetive unti! 3s EFPY)
Table 6.4 Prairie Island Unit 11/4T and 3/4T ART Calculations at 35 EFPY )
Material CF l f @35 1/4T f 1/4T FF I M l ARTnor ART lEFPY* 3/4T f 3/4T FF ( F) (cF)l ( F) (*F) 1/4T Calculations Nozzle (upper)Shell 51 2.20 1.47 1.1
- I -4 34 56.6 87 l Forging B '
Nozzle to Inter. Shell 79.5 2.20 1.47 1.1i 0(') 66 88.2 154 >
Circ Weld (Heat 2269) _
l Intermediate Shell 44.0 3.95 2.64 1.26 14 34 55.4 63
~
0 /U ~D ala~ ~ ~ ~ ~ ~5d.T~~i55~~~E64~~
~ ~ ~ 1726~~~ ~T 1 ~3d#~ ~55.5 ~ ~~ii7 ~
Circumferential Weld ~
69.7 3.95 2.64 1.26~~ -13~~
56 87.8 131
~Using S/U~ Data ~ ~~~~~~ ~ 56.5~~~E55"~~ ~2I64~ "~ 726 1 13 ~550'~ ~1IO'8~~~id5-'
Lower Shell Forging D 44.0 3.95 2.64 1.26 -4 34 55.4 85 i I
l 3/4T Calculations Nozzle (upper) Shell 51 2.20 0.660 0.884 -4 34 45.1 75 Forging B l Nozzle to Inter. Shell 79.5 2.20 0.660 0.884 0) 66 70.3 136 Circ Weld (Heat 2269)
Intermediate Shell 44.0 3.95 1.18 1.05 14 34 46.2 04
_Fof. gin _g_C _ _ _ _ _ _ _ _ _ .____________. , _ _ _ _ , _ . _ _ _ _ _ _____ ________ _ _ _ _ _ ,
Using S/C Data 54.7 3.95 1.18 ' 1.05 14 ___8 34' 57.4 105 Circumferential Weld ~ ~
69.7 3.95 1.18 1.05~~~~~~ -13 56 73.2 116
~Using S/5'Dala~~~~ ~~~ 56.5~~~i55~~~ ~1I18-~ ~ ~ ~1 765 -13 ~55#~ ~58.5-~IEE ~'
i Lower Shell Forging D 44.0 3.95 1.18 1.05 -4 34 46.2 76 l
l NOTE:
(a) Fluence values are x 10" n/cm2(E > 1.0 MeV). In addition, the values used are the calculated l
values since they are higher than the best-estimate values (Ref. 5 3).
(b) The full c3 margin of 17eF 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) 14
Prsssure and TsmpIrrture Limits Riport Revision 0 (Effictivs untd 33 EFPY)
Table 6.5 Prairie is!and Unit 21/4T and 3/4T ART Calculations at 35 EFPY Material CF f @35 1/4T f 1/4T FF 1 M ARTuor ART EFPY 3/4T f"' 3/4T FF ( F) ( F) ( 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 !
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~Using~Uniii57C bat?)~~"5E5~~ 2 375~~ ~ 3 ~55~ ~~ ~ lI13~~ -15~"~35~~ ~91 3 ~~~35I Interrnediate Shell 44.0 4.183 _.80 1.27 -4 34 55.9 86 Forging C Inter. to Lower Shell Weld 52.0 4.183 2.80 1.27 -31 56 66.0 91 l W
__3______________________ _ _ _ _ , , ,__________________________ _ _ _ _ _ _ , ._____ l Using S/C Data 80.0 4.183 2.80 1.27 -31 28 101.6 99
_L_ower phell.,F_orging D_ _ __51. 0 __ 4_183__ _ 2. 8 0, __ _1 J7_ __ -6____ _ _3,4_ _ _6_4 8, _ 2 _ __ _9_3_ _
Using S/C Data 60.0 4.183 2.80 1.27 -6 34 76.2 104 3/4T Calculations Upper Shell Forging 8 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 ~ ~ ~ ~ ~ ~ -
~Osin~g Unit 15IC 6ata*~~'~5E5~~ ~2 375 '~ ~ D~71~~ 0 90 ~~-15-" ~ 55 ~ ~ ~ ~f2 7 ~lI6 ~
Inte. mediate Shell 44.0 4.183 1.25 1.06 -4 34 46.6 77 Forging C 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
. L_o_wer phell Fo_rging D_ _ _ __51.Q __ ,4j 83__ _ ,1 ]5_ __ _ ,1_06_ __ _ 6_ _ ___,34__ _ _54;1_ ,__82__
Using G/C Data 60.0 4.163 1.25 1.06 -6 34 63.6 92 NOTE:
(a) Fluence values are x 105 n/cm 2(E > 1.0 MeV). In addition, the values used are the Best-l 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 a sof 28cF was used in the margin term.
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