L-05-063, Pressure and Temperature Limits Reports, Revision 1

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Pressure and Temperature Limits Reports, Revision 1
ML050960554
Person / Time
Site: Beaver Valley
Issue date: 03/31/2005
From: Pearce L
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-05-063
Download: ML050960554 (50)
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Text

FENOC Beaver Valley Power Station PO.Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077-0004 L William Pearce 724-082-5234 Vice President Fax: 724-643-8069 March 31, 2005 L-05-063 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Pressure and Temperature Limits Reports, Revision 1 FirstEnergy Nuclear Operating Company (FENOC) hereby submits Revision 1 of the Pressure and Temperature Limits Report (PTLR) for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2. The revised PTLRs are provided as Enclosures to this letter.

The PTLR revisions are being provided in accordance with Technical Specification 6.9.6, "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," which requires that the PTLR be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto. These revisions were issued as part of Revision Nos. 44 (Unit No. 1) and 42 (Unit No. 2) of the respective unit's Licensing Requirements Manuals.

Each unit's revised PTLR contains editorial changes to reflect the Technical Specifications and Licensing Requirements that reference the PTLR. The revision to the Unit No. 2 PTLR also contains changes to reflect the Capsule W analysis and corresponding low temperature overpressure protection system analysis.

No new commitments are contained in this submittal. If you have questions or require additional information, please contact Mr. Henry L. Hegrat, Supervisor - Licensing, at 330-315-6944.

Sincerely, William Pearce 3/4u1

Beaver Valley Power Station, Unit Nos. 1 and 2 Pressure and Temperature Limits Reports, Revision 1 L-05-063 Page 2

Enclosures:

1 Beaver Valley Power Station Unit No. 1 Pressure and Temperature Limits Report, Revision 1.

2 Beaver Valley Power Station Unit No. 2 Pressure and Temperature Limits Report, Revision 1.

c: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Sr. Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

L-05-063 Enclosure 1 Beaver Valley Power Station Unit No. 1 Pressure and Temperature Limits Report, Revision I

p BVPS-1 LICENSING REQUIREMENTS MANUAL SECTION 4.2 PRESSURE AND TEMPERATURE LIMITS REPORT BVPS-1 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.4.1.3 N/A N/A 4.2-3 3.4.3 N/A N/A 4.2-3 I 3.4.9.1 4.2.1.1 4.2-1 N/A 4.2-2 3.4.9.3 4.2.1.2 N/A 4.2-3 4.2.1.3 3.5.2 N/A N/A 4.2-3 I 3.5.3 N/A N/A 4.2-3 3.5.4.1.2 N/A N/A 4.2-3 3.10.3 N/A 4.2-1 N/A 4.2-2 _-

BVPS-1 Licensing Requirement to PTLR Cross-Reference Licensing PTLR Requirement Section Figure Table 2.2 N/A N/A 4.2-3 12.4 N/A N/A 4.2-3 PTLR Revision I 4.2-i LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This Pressure and Temperature Limits Report (PTLR) for Unit 1 has been prepared in accordance with the requirements of Technical Specification 6.9.6. Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications (TS) and Licensing Requirements (LR) addressed in, or make reference to, this report are listed below:

TS 3.4.1.3 Reactor Coolant System- Shutdown, TS 3.4.3 Reactor Coolant System - Safety Valve, TS 3.4.9.1 Reactor Coolant System - Pressure/Temperature Limits, TS 3.4.9.3 Overpressure Protection Systems, TS 3.5.2 ECCS Subsystems - Tavg 2 350'F, TS 3.5.3 ECCS Subsystems - Tavg < 350'F, TS 3.5.4.1.2 Boron Injection Tank < 350 0 F, TS 3.10.3 Special Test Exceptions - Pressure/Temperature Limitations -Reactor Criticality, LR 2.2 Boration Flow Paths - Operating, and LR 2.4 Charging Pumps - Operating.

4.2.1 Operating Limits The PTLR limits for Beaver Valley Power Station (BVPS) Unit 1 were developed using a methodology specified in the Technical Specifications. The methodology listed in Reference 1 was used with two exceptions:

a) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1", and b) Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure".

4.2.1.1 RCS Pressure and Temperature (PHT) Limits (TS 3.4.9.1)

The RCS temperature rate-of-change limits defined in Reference 2 are:

a. A maximum heatup of 1000 F in any one hour period.
b. A maximum cooldown of 1000 F in any one hour period, and
c. A maximum temperature change of less than or equal to 50 F in any one hour period during inservice hydrostatic testing operations above system design pressure.

The RCS P/T limits for heatup, leak testing, and criticality are specified by Figure 4.2-1 and Table 4.2-1. The RCS P/T limits for cooldown are shown in Figure 4.2-2 and Table 4.2-2.

These limits are defined in Reference 2. Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 4.2.1, the RCS P/T limits for heatup and cooldown shown in Figures 4.2-1 and 4.2-2 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to PTLR Revision 1 4.2-1 LRM Revision 44

BVPS-l LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. The heatup and cooldown curves also include the effect of the reactor vessel flange.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

Pressure-temperature limit curves shown in Figure 4.2-3 were developed for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This figure provides the ASME III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.

4.2.1.2 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)

The power operated relief valves (PORVs) shall each have maximum lift setting and enable temperature in accordance with Table 4.2-3. The lift setting provided does not impose any reactor coolant pump restrictions.

The PORV setpoint is based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 4.2.1. The PORV lift setting shown in Table 4.2-3 accounts for appropriate instrument error.

4.2.1.3 OPPS Enable Temperature (TS 3.4.9.3)

Two different temperatures are used to determine the OPPS enable temperature, they are the arming temperature and the calculated enable temperature. The arming temperature (when the OPPS rendered operable) is established per ASME Section XI, Appendix G. At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature is 343°F.

The calculated enable temperature is based on either a RCS temperature of less than 200'F or materials concerns (reactor vessel metal temperature less than RTNDT + 50'F), whichever is greater. The calculated enable temperature does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 3080 F.

As the arming temperature is higher and, therefore, more conservative than the calculated enable temperature, the OPPS enable temperature, as shown in Table 4.2-3, is set to equal the arming temperature.

The calculation method governing the heatup and cooldown of the RCS requires the arming of the OPPS at and below the OPPS enable temperature specified in Table 4.2-3, and PTLR Revision I 4.2-2 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT disarming of the OPPS above this temperature. The OPPS is required to be enabled, i.e.,

OPERABLE, when any RCS cold leg temperature is less than or equal to this temperature.

From a plant operations viewpoint the terms "armed" and "enabled" are synonymous when it comes to activating the OPPS. As stated in the applicable operating procedure, the OPPS is activated (armed/enabled) manually before entering the applicability of TS 3.4.9.3. This is accomplished by placing two keylock switches (one in each train) into their "automatic" position. Once OPPS is activated (armed/enabled) reactor coolant system pressure transmitters will signal a rise in system pressure above the OPPS setpoint. This will initiate an alarm in the control room and open the OPPS PORVs.

4.2.1.4 Reactor Vessel Boltup Temperature (TS 3.4.9.1)

The minimum boltup temperature for the Reactor Vessel Flange shall be 2 601F. Boltup is a condition in which the reactor vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.

4.2.2 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The capsule withdrawal schedule is provided in Table 4.5-3 of the UFSAR. Also, the results of these analyses shall be used to update Figures 4.2-1 and 4.2-2, and Tables 4.2-1 and 4.2-2. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

The pressure vessel material surveillance program (References 3 and 4) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME,Section III, NB-2331. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.

Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 1.

This commitment is a condition of license Amendment 256 and will remain in effect until the NRC staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.

PTLR Revision I 4.2-3 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.3 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.

Table 4.2-4, taken from Reference 5, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 4.2-4a, taken from Reference 2, shows the Calculation of Chemistry Factors based on St. Lucie and Fort Calhoun Surveillance Capsule Data.

Table 4.2-4b, taken from Reference 3, shows the St. Lucie and Fort Calhoun Surveillance Weld Data.

Table 4.2-5, taken from Reference 2, provides the reactor vessel beltline material property table.

Table 4.2-6, taken from Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 22 EFPY.

Table 4.2-7, taken from Reference 2, shows the calculation of ARTs for 22 EFPY.

Table 4.2-8 shows the Reactor Vessel Toughness Data (Unirradiated).

Table 4.2-9, taken from Reference 5, provides RTpTS values for 28 EFPY.

Table 4.2-10, taken from Reference 5, provides RTpTs values for 45 EFPY.

PTLR Revision I 4.24 LRM Revision 44 l

BVPS-- I LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.4 References

1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al., January 1996.
2. WCAP-15570, Revision 2, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, April 2001.
3. WCAP-15571, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," C. Brown, et. al-, November 2000.
4. WCAP-8475, "Duquesne Light Company, Beaver Valley Unit No. I Reactor Vessel Radiation Surveillance Program," J. A. Davidson, October 1974.
5. WCAP-15569, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 1,"

C. Brown, et al., November 2000.

6. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
7. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
8. Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

9. Westinghouse Report, "Beaver Valley Unit 1 FirstEnergy Nuclear Operating Company -

Overpressure Protection System - Setpoints for Y-Capsule", Revision 1, April 2001.

10. FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.

PTLR Revision I 4.2-5 LRM Revision 44 l

i BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL PLATE LIMITING ART VALUES AT 22 EFPY: 1/4T, 233°F 3/4T, 196°F 2500 -

2250 --

Leak Test Umitj 2000 _ 1_

Unacceptable / Acceptable Operation Operation 1500 Heatup Rate Critical Limit

1. 1250 - T _t U_-

looI zLI 0 so 1Q0 150 200 250 300 350 400 450 500 550 0

INDICATED TEMPERATURE ( F)

Figure 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 22 EFPY (TS 3.4.9.1)

PTLR Revision I 4.2-6 LRM Revision 44 l

BVPS-I LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL PLATE LIMITING ART VALUES AT 22 EFPY: 1/4T, 2330 F 3/4T, 196-F 2500 2250 2000 1750 efi 1500 to 0

1250 w

a.

1000 z

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 0

INDICATED TEMPERATURE ( F)

Figure 4.2-2 Reactor Coolant System Cooldown Limitations Applicable for the First 22 EFPY (TS 3.4.9.1)

PTLR Revision I 4.2-7 LRM Revision 44 l

BVPS- 1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2000

& 1500 CO) w co co 1L 500 0

50 60 70 80 90 100 110 120 TEMPERATURE (OF)

Figure 4.2-3 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1)

PTLR Revision I 4.2-8 LRM Revision 44 l

b BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-1 Heatup Curve Data Points for 22 EFPY (TS 3.4.9.1)

I 00-F/HR HEATUP I 00OF/HR CRITICALITY LEAK TEST LIMIT Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(-F) (psig) (F) (psig) (F) (psig) ('F) (psig) (F) (psig) 60 0 200 677 289 0 289 716 271 2000 60 564 205 696 289 564 289 739 289 2485 65 564 210 716 289 565 289 764 70 564 215 739 289 565 289 792 75 564 220 764 289 566 289 822 80 564 225 792 289 566 289 856 85 564 230 822 289 568 289 894 90 564 235 856 289 569 289 936 95 564 240 894 289 571 290 982 100 564 245 936 289 572 295 1033 105 564 250 982 289 575 300 1089 110 564 255 1033 289 577 305 1151 115 564 260 1089 289 580 310 1219 120 564 265 1151 289 583 315 1294 125 564 270 1219 289 586 320 1378 130 565 275 1294 289 591 325 1470 135 566 280 1378 289 593 330 1571 140 568 285 1470 289 600 335 1682 145 571 290 1571 289 601 340 1806 150 575 295 1682 289 611 345 1941 155 580 300 1806 289 612 350 2091 160 586 305 1941 289 621 355 2222 165 593 310 2091 289 621 360 2361 170 601 315 2222 289 621 175 611 320 2361 289 621 180 621 289 621 180 621 289 633 180 621 289 646 185 633 289 661 190 646 289 677 195 661 289 696 PTLR Revision I 4.2-9 LRM Revision 44 l

BVPS- l LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 (Page 1 of 2)

Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1)

STEADY 20'F/HR. 40'F/HR. 60F/IHRM IOOF/HR.

STATE Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(F) (psig) ('F) (psig) ( F) (psig) (F) (psig) (IF) (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 609 60 566 60 521 60 430 65 621 65 611 65 567 65 523 65 431 70 621 70 612 70 568 70 524 70 432 75 621 75 614 75 570 75 525 75 433 80 621 80 615 80 572 80 527 80 435 85 621 85 617 85 574 85 529 85 437 90 621 90 619 90 576 90 531 90 439 95 621 95 621 95 578 95 534 95 442 100 621 100 621 100 581 100 536 100 445 105 621 105 621 105 584 105 540 105 448 110 621 110 621 110 587 110 543 110 452 115 621 115 621 115 591 115 547 115 457 120 621 120 621 120 596 120 552 120 462 125 621 125 621 125 600 125 557 125 468 130 621 130 621 130 606 130 562 130 474 135 621 135 621 135 612 135 569 135 481 140 621 140 621 140 618 140 576 140 490 145 621 145 621 145 621 145 584 145 499 150 621 150 621 150 621 150 592 150 509 155 621 155 621 155 621 155 602 155 520 160 621 160 621 160 621 160 613 160 533 165 621 165 621 165 621 165 621 165 547 170 621 170 621 170 621 170 621 170 563 175 621 175 621 175 621 175 621 175 581 180 621 180 621 180 621 180 621 180 600 180 621 180 621 180 621 180 621 185 622 180 778 180 742 180 706 180 670 190 647 185 792 185 757 185 723 185 689 195 674 190 808 190 775 190 742 190 709 200 704 195 826 195 794 195 762 195 732 205 737 PTLR Revision I 4.2-10 LRM Revision 44 I

BVPS- l LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 (Page 2 of 2)

Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1)

STEADY 20'F/HR. 40IF/HR. 60'FIHR. lOOTF/HR.

STATE Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(-F) (psig) (F) (psig) (F) (psig) (F) (psig) (F) (psig) 200 846 200 815 200 785 200 757 210 774 205 868 205 839 205 811 205 785 215 815 210 892 210 865 210 839 210 815 220 861 215 918 215 894 215 871 215 850 225 911 220 947 220 925 220 905 220 888 230 967 225 980 225 961 225 944 225 930 235 1030 230 1016 230 1000 230 986 230 976 240 1099 235 1055 235 1043 235 1033 235 1028 245 1147 240 1099 240 1090 240 1086 240 1085 250 1201 245 1147 245 1143 245 1143 245 1147 255 1260 250 1201 250 1201 250 1201 250 1201 260 1325 255 1260 255 1260 255 1260 255 1260 265 1397 260 1325 260 1325 260 1325 260 1325 270 1477 265 1397 265 1397 265 1397 265 1397 275 1565 270 1477 270 1477 270 1477 270 1477 280 1662 275 1565 275 1565 275 1565 275 1565 285 1770 280 1662 280 1662 280 1662 280 1662 290 1888 285 1770 285 1770 285 1770 285 1770 295 2020 290 1888 290 1888 290 1888 290 1888 300 2165 295 2020 295 2020 295 2020 295 2020 305 2325 300 2165 300 2165 300 2165 300 2165 305 2325 305 2325 305 2325 305 2325 PTLR Revision I 4.2-11 LRM Revision 44 I

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)

FUNCTION SETPOINT OPPS Enable Temperature 343°F -

PORV Setpoint < 403 psim PTLR Revision I 4.2-12 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule apsule f( FF(b) ARTNDT () FF *ARTNDT FF2 V .323 .689 128.49 88.53 .475 Lower Shell Plate B6903-l(d) U .646 .878 118.93 104.42 .771 (Longitudinal) W .986 .996 148.52 147.93 .992 Y 2.15 1.21 142.18 172.04 1.464 V .323 .689 137.81 94.95 .475 Lower Shell Plate B6903-l(d) U .646 .878 131.84 115.76 .771 (Transverse) W .986 .996 179.99 179.27 .992 Y 2.15 1.21 166.93 201.99 1.464 SUM: 1104.89 7.404 CF = (FF*RTNDT) I(FF 2 ) = (1104.89) * (7.404) = 149.2 0 F V .323 .689 169.30 116.65 .475 Beaver Valley U .646 .878 176.30 154.79 .771 Surv. Weld Material 305424) W .986 .996 198.99 198.19 .992 Y 2.15 1.21 189.41 229.19 1.464 SUM: 698.82 3.702

___ CF = I(FF*RTNDT) + Y(FF2 ) = (698.82) + (3.702) = 188.80 F Notes:

(a) F= Calculated fluence from Beaver Valley Unit 1 capsule Y dosimetry analysis results, (x 10O9 n/cm2, E > 1.0 Mev).

(b) FF = fluence factor = f (0. 2 8 - 0.1 log f).

(c) The surveillance weld metal ARTNDT values have been adjusted by a ration factor of 1.06.

(d) Data not credible.

PTLR Revision I 4.2-13 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4a Calculation of Chemistry Factors(a)

(Based on St. Lucie and Fort Calhoun Surveillance Capsule Data)

Material Capsule Capsule f (D) FFc_ &RTNDT,() FF *ARTNDT FF_

St. Lucie 970 0.627 0.869 72.3 76.1 0.755 Surveillance 1040 0.909 0.973 67.4 79.7 0.947 Weld Metal 2840 1.41 1.10 68.0 90.9 1.21 Heat 90136 SUM: 246.7 2.91 CF =2(FF*RTNDT) + X(FF) = (246.7)-+ (2.91) = 84.80F Fort Calhoun W-225 0.553 0.834 238 183.0 0.696 Surveillance W-265 0.771 0.927 221 194.1 0.859 Weld Metal W-275 1.28 1.07 219 226.2 1.14 Heat 305414 SUM: 603.3 2.695 CF = 2(FF*RTNDT) + X(FF2) = (603.3) + (2.695) = 223.90 F Notes:

(a) Use of St. Lucie and Fort Calhoun Surveillance Capsule Data approved by NRC letter dated February 20,2002, "BEAVER VALLEY POWER STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: AMENDED PRESSURE-TEMPERATURE LIMITS (TAC NO.

MB2301)".

(b) f= Calculated fluence (x 1019 n/cm 2 , E> 1.0 Mev) from Reference 2.

(c) FF = fluence factor = f (0-28-0.1 0og0.

(d) ARTNDT values are the measured 30 ft-lb. shift values taken from Reference 2.

PTLR Revision I 4.2-14 LRM Revision 44 l

BVPS- 1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4b St. Lucie and Fort Calhoun Surveillance Weld Data(aO)

Material Capsule Cu Ni Irradiated Fluence ARTNDT Temperature 'F 1019 n/cm2 St. Lucie 970 0.2291 0.0699 546.7 0.627 72.3 Weld Metal 1040 0.2291 0.0699 546.7 0.909 67.4 Heat 90136 2840 0.2291 0.0699 546.7 1.41 68.0 Fort Calhoun W-225 0.35 0.60 527 0.553 238 Weld Metal W-265 0.35 0.60 534 0.771 221 Heat 305414 W-275 0.35 0.60 538 1.28 219 Notes:

(a) Use of St. Lucie and Fort Calhoun Surveillance Capsule Data approved by NRC letter dated February 20, 2002, "BEAVER VALLEY POWER STATION, UNIT I -ISSUANCE OF AMENDMENT RE: AMENDED PRESSURE-TEMPERATURE LIMITS (TAC NO.

MB230 1)".

(b) Data contained in this table was obtained from Reference 3.

PTLR Revision I 4.2-15 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-5 Reactor Vessel Beltline Material Properties Material Description Cu(%) Ni(%) Chemistry Initial Factor 0 RTNDT( F) (a)

Intermediate Shell Plate B6607-1 0.14 0.62 100.5 43 Intermediate Shell Plate B6607-2 0.14 0.62 100.5 73 Lower Shell Plate B6903-1 0.21 0.54 147.2 27 Lower Shell Plate B7203-2 0.14 0.57 98.7 20 Intermediate to Lower Shell Weld 0.27 0.07 124.3 -56 Seam (Heat 90136)11-714 Intermediate Longitudinal Shell 0.28 0.63 191.7 -56 Weld Seams (Heat 305424)19-714 A&B Lower Longitudinal Weld Seams 0.34 0.61 210.5 -56 (Heat 305414)20-714 A&B _

Surveillance Weld (Heat 305424) 0.26 0.61 181.6 Note:

(a) The initial RTNDT values for the plates and are based on measured data while the weld values are generic.

PTLR Revision I 4.2-16 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-6 Summary of Adjusted Reference Temperature (ARTs) for 22 EFPY MATERIAL DESCRIPTION 22 EFPY 1/4T ART(oF)(a) 3/4T ART(oF)(a)

Intermediate Shell Plate B6607-1 193 166 Intermediate Shell Plate B6607-2 223 196 Lower Shell Plate B7203-2 168 141 Lower Shell Plate B6903-1 230 191

- Using S/C Data(b) 233 193 Intermediate Shell Longitudinal Weld 19-714A/B 145 102

- Using S/C Data(b) 143 100 Intermediate to Lower Shell Circ. Weld 11-714 152 119

- Using S/C Data (c) 86 63 Lower Shell Longitudinal Weld 20-714A/B 159 111

- Using S/C Data(d) 168 117 Notes:

(a) ART = I + ARTNDT + M.

(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full car.)

(c) Based on credible St. Lucie Unit 1 surveillance data.

(d) Based on Fort Calhoun Unit 1 surveillance data. (Data not credible. ART calculated with a full OA.)

PTLR Revision I 4.2-17 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-7 Calculation of Adjusted Reference Temperatures (ARTs) for 22 EFPY PARAMETER VALUES Operating Time 22 EFPY Material Plate B6903-1 Plate B6607-2 Location Lower Shell Intermediate Plate Shell Plate 1/4T ART(QF) 3/4T ART(-F)

Chemistry Factor, CF (0F) 149.2 100.5 Fluence (f), n/cm2 (E>1.0 Mev)(a) 1.70 x 10'9 6.62 x I0Og Fluence Factor, FF 1.15 .884 ARTNDT = CF x FF(oF)(C) 171.6(C) 88.84 Initial RTNDT, I(oF)(a) 27 73 Margin, M(0 F) 34(C) 34 ART = I+(CF*FF)+M, oF(b) per RG 1.99, Revision 2 233 196 Notes:

(a) Initial RTNDT values are measured values for plate material.

(b) This value was rounded per ASTM E29, using the "Rounding Method."

(c) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full OA.)

PTLR Revision I 4.2-18 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-8 Reactor Vessel Toughness Data (IUnirradiated)

RatrV e Nir TNDT R UPPER SHELF ENERGY (FT-LB)

COMPONENT HEAT NO CODE NO. MATERIAL TYPE CF NF M N D

(%) (%/) (%f) (°F) (CF) MWD NMWD Closure Head C6213-1 B B6610 A533B CL.I .15 - .010 -40 0* 121 -

Dome IlII I Closure Head Seg. A5518-2 B6611 A533B CL. .14 - .015 -20 -20' 131 -

Closure Head ZV3758 _ A508 CL 2 .08 - .007 60 60* >100 -

Flange Vessel Flange ZV3661 _ A508 CL. 2 .12 - .010 60* 60* 166 -

Inlet Nozzle 9-5443 - A508 CL. 2 .10 - .008 60* 60* 82.5 _

Inlet Nozzle 9-5460 _ A508 CL. 2 .10 - .010 600 60* 94 l Inlet Nozzle 9-5712 _ A508 CL. 2 .08 _ .007 60* 60* 97 -

Outlet Nozzle 9-5415 _ A508 CL. 2 -- .008 60* 60* 97 Outlet Nozzle 9-5415 A508 CL 2 - _ .007 60* 60* 112.5 -

Outlet Nozzle 9-5444 A508 CL. 2 .09 .007 60* 60* 103 _

Upper Shell 123V339 A508 CL 2 -- .010 40 40* 155 -

Inter Shell C4381-2 B6607-2 A533B CL I .14 .62 .015 -10 73 123 82.5 Inter Shell C4381-i B6607-1 A533B CL 1 .14 .62 .015 -10 43 128.5 90 Lower Shell C6317-1 B6903-1 A533B CL. 1 .20 .54 .010 -50 27 134 80 Lower Shell C6293-2 B7203-2 A533B CL. 1 .14 .57 .015 -20 20 129.5 83.5 Trans Ring 123V223 -- A508 CL. 2 -- -- - 30 30* 143 _ _

Botton Hd Seg C4423-3 B6618 A533B CL. 1 .13 _ .008 -30 -29* 124 Bottom Hd Dome C4482-1 B6619 A533B CL. 1 .13 - .015 -50 -33* 125.5 Inter to Lower 90136 _ .27 .07 - - -56 - > 100 Shell Weld Inter Shell Long. 305424 _ _ .28 .63 - - -56 > 100 Weld Lower Shell Long. 305414 _ _ .34 .61 - - -56 > 100 Weld Weld HAZ _ _ - -40 -40 _ 136.5 Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Direction NMWD - Normal to Major Working Direction Note: For evaluation of Inservice Reactor Vessel Irradiation damage assessments, the best estimate chemistry values reported in the latest response to Generic Letter 92-01 or equivalent document are applicable.

PTLR Revision I 4.2-19 LRM Revision 44 l

BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-9 RTpms Calculation for Beltline Region Materials at EOL (28 EFPY)

Material Fluence FF CF A RTs(a) Margin RTNDT(In(b) RTpTs(

(I 0'9 n/cM2, (OF) (OF) (°F) O°F) (OF)

E>1.0 MeV)_

Intermediate Shell Plate B6607-1 3.54 1.329 100.5 133.6 34 43 211 Intermediate Shell Plate B6607-2 3.54 1.329 100.5 133.6 34 73 241 Lower Shell Plate B7203-2 3.54 1.329 98.7 131.2 34 20 185 Lower Shell Plate B6903-1 3.54 1.329 147.2 195.6 34 27 257

-* Using S/C Data(e) 3.54 1.329 149.2 198.3 34 27 259 Inter. Shell Long. Weld 19-714A/B 0.708 0.903 191.7 173.1 65.5 -56 183

-e Using S/C Data(6) 0.708 0.903 188.8 170.5 65.5 -56 180 Lower Shell Long. Weld 20-714A/B 0.708 0.903 210.5 190.1 65.5 -56 200

-~ Using S/C Data( 0 0.708 0.903 223.9 202.2 65.5 -56 212 Circumferential Weld 11-714 3.53 1.329 124.3 165.2 65.5 -56 175

-e Using S/C Data(d) 3.53 1.329 84.8 112.3 44 -56 101 Notes:

(a) ARTpTs = CF

(b) Initial RTNDT values of the plate material are measured values while the weld material values are generic.

(c) RTm = RTNDTQJ) + ARTPTs + Margin ( 0F).

(d) Based on credible St. Lucie Unit I surveillance data.

(e) Based on non-credible Beaver Valley Unit 1 surveillance data with a full cYA.

(f) Based on non-credible Fort Calhoun Unit I surveillance data with a full GA.

PTLR Revision 1 4.2-20 LRM Revision 44 l

BVPS-1 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-10 RTprs Calculation for Beltline Region Materials at Life extension (45 EFPY)

Material Fluence FF CF A RTPrs(c) Margin RTNDT((a) RTprs(b)

(1019 n/cm2 , (OF) (OF) (OF) (OF) (OF)

E>1.0 MeV)

Intermediate Shell Plate B6607-1 5.85 1.43 100.5 143.7 34 43 221 Intermediate Shell Plate B6607-2 5.85 1.43 100.5 143.7 34 73 251 Lower Shell Plate B7203-2 5.85 1.43 98.7 141.1 34 20 195 Lower Shell Plate B6903-1 5.85 1.43 147.2 210.5 34 27 272

-4 Using S/C Data(e) 5.85 1.43 149.2 213.4 34 27 274 Inter. Shell Long. Weld 19-714A/B 1.13 1.03 191.7 197.5 65.5 -56 207

-* Using S/C Data(e) 1.13 1.03 188.8 194.5 65.5 -56 204 Lower Shell Long. Weld 20-714A/B 1.13 1.03 210.5 216.8 65.5 -56 226

-e Using S/C Data(0 1.13 1.03 223.9 230.6 65.5 -56 240 Circumferential Weld 11-714 5.82 1.43 124.3 177.7 65.5 -56 187

-e Using S/C Data(d) 5.82 1.43 84.8 121.3 44 -56 109 Notes:

(a) Initial RTNDT values of the plate material are measured values while the weld material values are generic.

(b) RTpTs = RTNDT) + ARTpTs + Margin ( 0F).

(c) ARTpms= CF

(d) Based on credible St. Lucie Unit I surveillance data.

(e) Based on non-credible Beaver Valley Unit I surveillance data with a full aA.

(f) Based on non-credible Fort Calhoun Unit 1 surveillance data with a full GA.

PTLR Revision I 4.2-21 LRM Revision 44 l

L-05-063 Enclosure 2 Beaver Valley Power Station Unit No. 2 Pressure and Temperature Limits Report, Revision 1

BVPS-2 LICENSING REOUIREMENTS MANUAL SECTION 4.2 PRESSURE AND TEMPERATURE LIMITS REPORT BVPS-2 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.4.1.2 N/A N/A 4.2-3 I 3.4.1.3 N/A N/A 4.2-3 3.4.3 N/A N/A 4.2-3 I 3.4.9.1 4.2.1.1 4.2-1 N/A 4.2-2 4.2-3 4.2-4 4.2-5 4 .2 -6 _ _ _ _ _ _ _ _ _

3.4.9.3 4.2.1.2 4.2-8 4.2-3

_ __ _ _ __ _ _ _ _ 4 .2. 1 .3_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

3.5.2 N/A N/A 4.2-3 3.5.3 N/AN/A 4.2-3 BVPS-2 Licensing Requirement to PTLR Cross-Reference Licensing PTLR Requirement Section Figure Table 2.2 N/A N/A 4.2-3 2.4 N/A N/A 4.2-3 PTLR Revision I 4.2-i LRM Revision 42 l

BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This Pressure and Temperature Limits Report (PTLR) for Unit 2 has been prepared in accordance with the requirements of Technical Specification 6.9.6. Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications (TS) and Licensing Requirements (LR) addressed in, or make l reference to, this report are listed below:

TS 3.4.1.2 Reactor Coolant System - Hot Standby, TS 3.4.1.3 Reactor Coolant System - Shutdown, TS 3.4.3 Reactor Coolant System - Safety Valve, TS 3.4.9.1 Reactor Coolant System - Pressure/Temperature Limits, TS 3.4.9.3 Overpressure Protection Systems, TS 3.5.2 ECCS Subsystems - Tavg > 350'F, TS 3.5.3 ECCS Subsystems - Tavg < 350'F, LR 2.2 Boration Flow Paths - Operating, and LR 2.4 Charging Pumps - Operating.

4.2.1 Operating Limits The PTLR limits for Beaver Valley Power Station (BVPS) Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in Reference I was used with two exceptions:

a) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1", and b) Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure".

4.2.1.1 RCS Pressure and Temperature (PIT) Limits (TS 3.4.9.1)

The RCS temperature rate-of-change limits defined in Reference 2 are:

a. A maximum heatup of 60'F in any one hour period.
b. A maximum cooldown of 1000 F in any one hour period, and
c. A maximum temperature change of less than or equal to 50 F in any one hour period during inservice hydrostatic testing operations above system design pressure.

The RCS P/T limits for heatup, leak testing, and criticality are specified by Figure 4.2-1 and Table 4.2-1. The RCS P/T limits for cooldown are shown in Figures 4.2-2 through 4.2-6 and Table 4.2-2. These limits are defined in Reference 2. Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 4.2.1, the RCS P/T limits for heatup and cooldown shown in Figures 4.2-1 through and 4.2-6 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in PTLR Revision I 4.2-1 LRM Revision 42

BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 10 CFR 50, Appendix G. The heatup and cooldown curves also include the effect of the reactor vessel flange.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

Pressure-temperature limit curves shown in Figure 4.2-7 were developed for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This figure provides the ASME 111, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.

4.2.1.2 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)

The power operated relief valves (PORVs) shall each have nominal maximum lift setting is in accordance with Figure 4.2-8 (Reference 11). The OPPS enable temperature is in accordance with Table 4.2-3. The PORV lift setting provided is for the case with reactor coolant pump (RCP) restrictions. These restrictions are shown in Table 4.2-4, which is taken from Reference 9. Due to the setpoint limitations as a result of the reactor vessel flange requirements, there is no operational benefit achieved by restricting the number of RCPs running to less than two below an indicated RCS temperature of 1370 F. Therefore, the PORV setpoints shown in Table 4.2-4 will protect the Appendix G limits for the combinations shown.

The PORV setpoint is based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 4.2.1. The PORV lift setting shown in Figure 4.2-8 accounts for appropriate instrument error.

4.2.1.3 OPPS Enable Temperature (TS 3.4.9.3)

Two different temperatures are used to determine the OPPS enable temperature, they are the arming temperature and the calculated enable temperature. The arming temperature (when the OPPS rendered operable) is established per ASME Section XI, Appendix G. At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature with uncertainty is 2370 F.

The calculated enable temperature is based on either a RCS temperature of less than 2000 F or materials concerns (reactor vessel metal temperature less than RTNDT + 50'F), whichever is greater. The calculated enable temperature does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 2400 F.

As the calculated enable temperature is higher and, therefore, more conservative than the arming temperature, the OPPS enable temperature, as shown in Table 4.2-3, is set to equal the calculated enable temperature.

PTLR Revision 1 4.2-2 LRM Revision 42

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT The calculation method governing the heatup and cooldown of the RCS requires the arming of the OPPS at and below the OPPS enable temperature specified in Table 4.2-3, and disarming of the OPPS above this temperature. The OPPS is required to be enabled, i.e., OPERABLE, when any RCS cold leg temperature is less than or equal to this temperature.

From a plant operations viewpoint the terms "armed" and "enabled" are synonymous when it comes to activating the OPPS. As stated in the applicable operating procedure, the OPPS is activated (armed/enabled) manually before entering the applicability of TS 3.4.9.3. This is accomplished by placing two keylock switches (one in each train) into their "ARM" position.

Once OPPS is activated (armed/enabled) reactor coolant system pressure transmitters will signal a rise in system pressure above the variable OPPS setpoint. This will initiate an alarm in the control room and open the OPPS PORVs.

4.2.1.4 Reactor Vessel Boltup Temperature (TS 3.4.9.1)

The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60'F. Boltup is a condition in which the reactor vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.

4.2.2 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The capsule withdrawal schedule is provided in Table 5.3-6 of the UFSAR. Also, the results of these analyses shall be used to update Figures 4.2-1 through 4.2-6, and Tables 4.2-1 and 4.2-2. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

The pressure vessel material surveillance program (References 3 and 4) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME,Section III, NB-233 1. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.

Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 2. This commitment is a condition of License Amendment 138 and will remain in effect until the NRC l staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.

PTLR Revision I 4.2-3 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.3 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.

Table 4.2-5, taken from Table 4-9 of Reference 2, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 4.2-6, taken from Tables 4-8 and 4-10 of Reference 2, provides the reactor vessel beitline material property table.

Table 4.2-7, taken from Tables 4-15 and 4-16 of Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 22 EFPY.

Table 4.2-8, taken from Tables 4-15 and 4-16 of Reference 2, shows the calculation of ARTs for 22 EFPY.

Table 4.2-9 shows the Reactor Vessel Toughness Data (Unirradiated).

Table 4.2-10, taken from Table 6 of Reference 5, provides RTp-s values for 32 EFPY.

PTLR Revision I 4.2-4 LRM Revision 42

BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.4 References

1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al.,

January 1996.

2. WCAP-15677, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation,"

J. H. Ledger, August 2001.

3. WCAP-15675, Revision 0, "Analysis of Capsule W from First Energy Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," J. H. Ledger, S. L. Anderson, J. Conermann, August 2001.
4. WCAP-9615, Revision 1, "Duquesne Light Company, Beaver Valley Unit No. 2 Reactor Vessel Radiation Surveillance Program," P. A. Peter, June 1995.
5. WCAP-15676, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2," J. H. Ledger, August 2001.
6. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
7. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
8. Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

9. FENOC Calculation No. 10080-SP-2RCS-006, Revision 4, Addendum 0, "BV-2 LTOPS Setpoint Evaluation Capsule W for 22 EFPY."
10. FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
11. Westinghouse Letter FENOC-04-31, dated April 14, 2004, "LTOPS Setpoint Evaluation for Beaver Valley Unit 2 Capsule W for 22 EFPY - Calculation Note."

PTLR Revision I 4.2-5 LRM Revision 42 l

i, BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 22 EFPY: 1/4T, 140°F 3/4T, 129°F CURVES APPLICABLE FOR HEATUP RATES UP TO 60 °FIHR FOR THE SERVICE PERIOD UP TO 22 EFPY. I 2500 Leak Test Limit 2250 _ -t- -t __7 _

Unacceptabl e Operation tAcceptable Operation 1750 l

- eatup Rate to 60 VFIr.

",J15007t_==

O / Criticality Limit for 60 "F1Hr.

(I 120___ __// _ _ =__ ___

111250

<1000 _ __

_ i Criffcality Urnit based on Inservice 50S_ rBoto

< o p hydrostatictesttemperature(196F)forl ltho service period up to 22 EFPY.

I Temperature II 250 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE CF)

Figure 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-6 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 22 EFPY: 1/4T, 1400 F 3/4T, 1290 F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 0F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY. I 2500 2250 Unaccepable Operation Acceptable 2000 Operation

- 1750 0~

ax LU 1500 n

1*_____ _____ L...i.iYiI7I _____

\

I _____ _____

Cooldown Rate O 0 FIHr.

UJ 1250 c .-

0 w 1000

/

z 750 FI I_,

500 Boltup 1 _

Temperature 250 ________ _______ I ____ I_ ____i_____ _____

0 Yf¶ r 51T t 15! TT11 1T I TT

  • YYIr ,,-r-r + rrT-0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)

Figure 4.2-2 Reactor Coolant System Cooldown (up to 0 0F/Hr.)

Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-7 LRM Revision 42

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 22 EFPY: 1/4T, 140°F 3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 20°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.

2500 2250 -

Uncceptable Operation I

I I-I t

_ _ I - --

-t 2000 -4

- 1750 II--

./. Acceptable Operation 0

W 1500 i _

0 7 Cooldown Rate 20 'FIHr.

1250 + I w

I.- 1000 -4 -t t 4 4-500 z 750 I

500 l TBoltup - I 4 I

Temperature 250 0

0

,,L. 50

.j, 100 150

..i .i .

200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)

Figure 4.2-3 Reactor Coolant System Cooldown (up to 200F/Hr.)

Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-8 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 22 EFPY: 1/4T, 140°F 3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 40°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY. I 2500-2250 Una c ble Acceptable 2000 Op oOperation C1750 __ _

gg1500 _

D1250 -Cooldown -_ Rate W4FIHr.

LU 10100g 750 500 _ ___ Boltup _ c Temperature 250 _ _ _ = _

0 i II, r Y . 1,! l . T ,

0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (0F)

Figure 4.2-4 Reactor Coolant System Cooldown (up to 40 °F/Hr.)

Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-9 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 22 EFPY: 1/4T, 1400 F 3/4T, 1290 F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 60 0F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY. I 2500 -

2250 Unacceptabe Opration ly Op o Acceptable 2000 Operation 1750 Uj1500 -_

Cooldown Rate 60FIHr.

1250 7 t-_=

MI 750 Z 250 ------

250 mperature__ __T 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (0F)

Figure 4.2-5 Reactor Coolant System Cooldown (up to 600F/Hr.)

Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-10 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-l LIMITING ART VALUES AT 22 EFPY: l/4T, 140°F 3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY. I 2500 T 2250 Unceptabe Oeratlon . I i fl Acceptable 2000 Operation 1750 - __

1500 FC ooldown Rate 10D°FlHr.

LU 1250 tll 1

1 1000 Z 750 - -t 500 _ _ t

_Temperature 250 0 -r.,,

, lTIF

,,, lyl- lT-r-r- , ,,, ,

0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)

Figure 4.2-6 Reactor Coolant System Cooldown (up to l00°F/Hr.)

Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) I PTLR Revision I 4.2-11 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2000 (D 1500 a-w km ul) w 500 0

50 60 70 80 90 100 110 120 TEMPERATURE (F)

Figure 4.2-7 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1)

PTLR Revision 1 4.2-12 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT I See Table 4.2-4 for RCP restrictions.

750 .~. .

I II i i I i _

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l-U, _,__Operation 1>__

a.

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I , , I .

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____>._ _ Operation * ,

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_ _I I

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I 0 100 200 300 400 TRM-AUCTIONEERED LOW-MEASURED RCS TEMPERATURE (°F)

Figure 4.2-8 Maximum Allowable Nominal PORV Setpoint for the Overpressure Protection System (TS 3.4.9.3)

PTLR Revision 1 4.2-13 LRM Revision 42 I

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-1 Heatup Curve Data Points for 22 EFPY (TS 3.4.9.1) I 600 F/HR HEATUP 60 0 F/HR CRITICALITY LEAK TEST LIMIT Temp. Press. Temp. Press. Temp. Press.

(OF) (psig) (OF) (psig) (OF) (psig) 60 0 196 0 178 2000 60 621 196 621 196 2485 65 621 196 621 70 621 196 621 75 621 196 621 80 621 196 621 85 621 196 621 90 621 196 621 95 621 196 621 100 621 196 621 105 621 196 621 110 621 196 621 115 621 196 621 120 621 196 779 120 621 196 799 120 779 196 821 125 799 196 846 130 821 196 874 135 846 196 905 140 874 196 940 145 905 196 978 150 940 200 1021 155 978 205 1068 160 1021 210 1120 165 1068 215 1178 170 1120 220 1242 175 1178 225 1312 180 1242 230 1390 185 1312 235 1476 .

190 1390 240 1571 195 1476 245 1675 200 1571 250 1791 205 1675 255 1919 210 1791 260 2060 215 1919 265 2215 220 2060 270 2387 225 2215 230 2387 PTLR Revision I 4.2-14 LRM Revision 42 I

k BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1) I l0F/HR 20 0 F/HR 40 0 F/HR, 600 F/HR 100F/HR Temp. Press. Press. Press. Press. Press.

(0 F) (psig) (psig) (psig) (psig) (psig) 60 0 0 0 0 0 60 621 621 621 608 532 65 621 621 621 618 544 70 621 621 621 621 557 75 621 621 621 621 572 80 621 621 621 621 588 85 621 621 621 621 606 90 621 621 621 621 621 95 621 621 621 621 621 100 621 621 621 621 621 105 621 621 621 621 621 110 621 621 621 621 621 115 621 621 621 621 621 120 621 621 621 621 621 120 621 621 621 621 621 120 907 884 862 842 807 125 935 914 895 877 849 130 966 948 932 917 897 135 1001 985 972 961 949 140 1039 1026 1017 1010 1007 145 1081 1072 1066 1064 1071 150 1127 1122 1121 1123 1127 155 1179 1178 1179 1179 1179 160 1235 1235 1235 1235 1235 165 1298 1298 1298 1298 1298 170 1367 1367 1367 1367 1367 175 1444 1444 1444 1444 1444 180 1528 1528 1528 1528 1528 185 1622 1622 1622 1622 1622 190 1725 1725 1725 1725 1725 195 1839 1839 1839 1839 1839 200 1966 1966 1966 1966 1966 205 2105 2105 2105 2105 2105 210 2259 2259 2259 2259 2259 215 2430 2430 2430 2430 2430 PTLR Revision I 4.2-15 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3) l FUNCTION SETPOINT OPPS Enable Temperature 2400 F I PORV Seloint Figure 4.2-8 PTLR Revision I 4.2-16 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.24 Reactor Coolant Pump Restrictions TRCS Running RCPs

< 1370 F 0-2

Ž1370 F 3 PTLR Revision I 4.2-17 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-5 Calculation of Chemistry Factors Using Surveillance Capsule Data(a) I Material Capsule Capsulef(b) FF(c) ARTNDTrdY l FF *ARTNDT FF U 0.608 0.86 24.26 20.86 0.74 I Intermediate Shell Plate B9004-2 (Longitudinal) V 2.63 1.26 55.93 70.47 1.59 W 3.625 1.335 71.04 94.83 1.78 U 0.608 0.86 17.56 15.10 0.74 Intermediate Shell Plate B9004-2 (Transverse) V 2.63 1.26 46.27 58.30 1.59 W 3.625 1.335 63.39 84.63 1.78 SUM: 344.19 8.22 CF =£(FF*RTNDT) + XIFF) = 41.9 U 0.608 0.86 3.64 3.13 0.74 Weld Metal V 2.64 1.26 25.47 32.09 1.59 W 3.625 1.335 6.21 8.29 1.78 SUM: 43.51 4.11 CF = X(FF*RTNDT) + Z(FF2) = 10.6 Notes:

(a) Regulatory Guide 1.99, Revision 2, Position 2.1.

(b) f= fluence (1019 n/cr 2 ); Fluence values were taken from Capsule W analysis (Reference 3).

(c) FF = fluence factor = f(0. 28 -0.1

  • log f)

(d) ARTNDT values obtained from CVGRAPH Version 4.1. I PTLR Revision I 4.2-18 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-6 Reactor Vessel Beitline Material Properties I Material Method Used Average Average Chemistry Initial RTNDT(b)

To Calculate Fco f CF(a) Cu wt % Ni wt % Factor ((F)

Closure Head Flange N/A 0.74 N/A N/A -10 Vessel Flange N/A 0.73 N/A N/A 0 Intermediate Shell Position 1.1 0.065 0.55 40.5 60 Plate B9004-1 Intermediate Shell Position 1.1 0.06 0.57 37.0 40 Plate B9004-2 Position 2.1 N/A N/A 41.9 40 I Lower Shell Plate Position 1.1 0.08 0.58 51.0 28 B9005-1 Lower Shell Plate Position 1.1 0.07 0.57 44.0 33 B9005-2 Weld Metal Position 1.1 0.046 0.086 34.4 -30 (Longitudinal & Position 2.1 N/A N/A 10.6 -30 Circumferential Seams)

Notes:

(a) Regulatory Guide 1.99, Revision 2, Position.

(b) Initial RTNDT values of the base metal and weld metal materials are measured values.

I PTLR Revision I 4.2-19 LRM Revision 42 l

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-7 Summary of Adjusted Reference Temperature (ARTs) for 22 EFPY I MATERIAL DESCRIPTION Method Used To 22 EFPY ART I Calculate the CF(a) 1/4T ART(0F) 3/4T ART(CF)

Intermediate Shell Plate B9004-1 Position 1.1 140 129 Intermediate Shell Plate B9004-2 Position 1.1 116 106 Position 2.1 104 94 Lower Shell Plate B9005-1 Position 1.1 120 106 Lower Shell Plate B9005-2 Position 1.1 117 105 Vessel Beltline Welds (b) Position 1.1 48 30 Position 2.1 -6 -12 Notes:

(a) Regulatory Guide 1.99, Revision 2.

(b) All Beltline Welds are from Heat #83642, Linde 0091, Flux Lot #3536. I PTLR. Revision I 4.2-20 LRM Revision 42

BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-8 Calculation of Adjusted Reference Temperatures (ARTs) for 22 EFPY I PARAMETER VALUES Operating Time 22 EFPY I Material - Intermediate Shell Plate B9004-1 B9004-1 Location I/4T ART 3/4T ART Chemistry Factor, CF (0F) 40.5 40.5 Fluence, (f), (1019 n/cm2)(a) 1.63 0.632 Fluence Factor, FF 1.13 0.87 ARTNDT = CF x FF(0 F) 45.8 35.2 Initial RTNDT, I(fF) 60 60 Margin, M(0 F) 34 34 ART, per Regulatory Guide 1.99, Revision 2 140 129 Notes:

(a) Fluence (f), is based upon ff (10'9 n/cm 2 , E > 1.0 MeV) = 1.81 at 22 EFPY.

The Beaver Valley Unit 2 reactor vessel wall thickness is 7.875 inches at the beltline region.

PTsLR Revision I 4.2-21 LRM Revision 42 l

BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-9 Reactor Vessel Toughness Data (Unirradiated)

MATERIAL TN= 50 FT/LB RTn USE COMPONENT CODE NO. SPEC. NO. Cu % Ni % P% OF 35 MIL OF FT-LBS.

Closure Head Dome B9008-1 A533B, CL. 1 .13 .54 .013 -20 50 -10 137 Closure Head Flange B9002-1 A508, CL.2 --- .74 .012 -10 <40 -10 136 Vessel Flange B9001 -1 A508, CL.2 .73 .010 0 <10 0 132.5 Inlet Nozzle B9011 -I A508, CL.2 .88 .006 0 <10 0 104 Inlet Nozzle B9011-2 A508, CL.2 . .88 .010 10 <10 10 115 Inlet Nozzle B9011-3 A508, CL. 2 --- .84 .009 20 <40 20 122 Outlet Nozzle B9012-1 A508, CL.2 _ .71 .007 -10 <0 -10 137 Outlet Nozzle B9012-2 A508, CL. 2 --- .74 .006 -10 <0 -10 121 Outlet Nozzle B9012-3 A508, CL. 2 -- .68 .008 -10 <0 -10 112 Nozzle Shell B9003-1 A533B, CL. 1 .13 .61 .008 -10 110 50 91 Nozzle Shell B9003-2 A533B, CL. 1 .12 .58 .009 0 120 60 79.5 Nozzle Shell B9003-3 A533B, CL. I .13 .61 .008 -10 110 50 97.5 Inter. Shell B9004-1 A533B, CL. 1 .07 .53 .010 0 120 60 83 Inter. Shell B9004-2 A533B, CL. I .07 .59 .007 -10 100 40 75.5 Lower Shell B9005-1 A533B, CL. I .08 .59 .009 -50 88 28 82 Lower Shell B9005-2 A533B, CL. I .07 .58 .009 -40 93 33 77,5 Bottom Head Torus B9010-1 A533B, CL. I .15 .49 .007 -30 56 -4 97 Bottom Head Dome B9009-1 A533B, CL. 1 .14 .53 .007 -30 35 -25 116 Weld (Inter. & Lower Shell Long. Seams & Girth Seam)* .08 .07 .008 -30 <30 -30 144.5 HAZ (Plate B9004-2) . --_ -80 40 -20 76 Same heat of wire and lot of flux used in all seams including surveillance weldment.

(1) For evaluation of Inservice Reactor Vessel Irradiation damage assessments, the best estimate chemistry values reported in the latest response to Generic Letter 92-01 or equivalent document are applicable.

(2) See Section 4.2.1.1 for a discussion of EFPY.

PTLR Revision 1 4.2-22 LRM Revision 42

BVPS-2 a

LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-10 RTvrs Calculation for Beltline Region Materials at EOL (32 EFPY)

I Material Method f (a) FF(b) CF A RTp~rs Margin RTNDT(U) RTpTs Fluence (OF) (OF) (°F) (OF) (OF)

Intermediate Shell Plate B9004-1 RG 1.99, R2, Pl.1 3.847 1.348 40.5 54.6 34 60 149 Intermediate Shell Plate B9004-2 RG 1.99, R2, Pl.l 3.847 1.348 37.0 49.9 34 40 124 RG 1.99, R2, P2.1 3.847 1.348 41.9 56.5 17 40 114 Lower Shell Plate B9005-1 RG 1.99, R2, Pl.1 3.847 1.348 51.0 68.7 34 28 131 Lower Shell Plate B9005-2 RG 1.99, R2, Pl.1 3.847 1.348 44.0 59.3 34 33 126 Vessel Beltline Welds RG 1.99, R2, P1.1 3.847 1.348 34.4 46.4 46.4 -30 63

__RG 1.99, R2, P2.1 3.847 1.348 10.6 14.3 14.3 -30 -1 Notes:

(a) f= peak clad/base metal interface fluence (10'9 n/cm2 , E>l.0 MeV) at 32 EFPY (450 fluence for longitudinal welds)

(b) FF = f(0.28 - 0.10 log f)

(c) RTNDTQJ) values are measured values.

(d) All BeItline Welds are from Heat #83642, Linde 0091, Flux Lot #3536. I PTLR Revision I 4.2-23 LRM Revision 42 l

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