Text

Pressure and Temperature Limits Reports, Revisions 8 and 9
	ML17277B091
Person / Time
Site:	Beaver Valley
Issue date:	10/04/2017
From:	Bologna R; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-17-277
	Download: ML17277B091 (62)
	v • d • e

FENOC' FirstEnergy Nuclear Operating Company Richard D. Bologna Site Vice President October 4, 2017 L-17-277 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Pressure and Temperature Limits Report Revision Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 724-682-5234 Fax: 724-643-8069 Pursuant to the requirements of Beaver Valley Power Station, Unit No. 1 (BVPS-1)

Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," FirstEnergy Nuclear Operating Company (FENOC) hereby submits the BVPS-1 PTLR, Revisions 8 and 9. Technical Specification 5.6.4.c requires that the PTLR be provided to the Nuclear Regulatory Commission (NRC) upon issuance for any revision or supplement thereto.

Revision 8 of the BVPS-1 PTLR was made effective on July 19, 2017, and was updated to provide pressure and temperature limit curves and low temperature over-pressure protection system setpoints that are valid through 50 effective full power years of BVPS-1 operation. This revision of the BVPS-1 PTLR incorporates the Capsule X fluence analysis results, sister plant surveillance capsule test results, and revised unirradiated nil-ductility reference temperature values for each of the four reactor vessel beltline plate materials. The revised unirradiated nil-ductility reference temperature values were previously reported in FENOC letter dated September 20, 2016 (Accession No. ML16265A047).

Revision 9 of the BVPS-1 PTLR was made effective on September 7, 2017, and was updated to correct typographical errors discovered in Revision 8 of the PTLR. A condition report was written when errors were discovered.

The BVPS-1 PTLR, revisions 8 and 9, are provided as enclosures A and B, respectively.

Beaver Valley Power Station, Unit No. 1 L-17-277 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at 330-315-6810.

Richard D. Bologna

Enclosure:

A. Beaver Valley Power Station Unit No. 1, Pressure and Temperature Limits Report, Revision 8

8. Beaver Valley Power Station Unit No. 1, Pressure and Temperature Limits Report, Revision 9 cc:

NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Enclosure A L-17-277 Beaver Valley Power Station, Unit No. 1 Pressure and Temperature Limits report, Revision 8 (29 pages follow)

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Pressure and Temperature Limits Report BVPS-1 Technical Specification to PTLR Cross -Reference Technical PTLR Specification Section Figure Table 3.4.3 5.2.1.1 5.2-1 N/A 5.2-2 3.4.6 N/A NIA 5.2-3 3.4.7 N/A N/A 5.2-3 3.4.10 NIA N/A 5.2-3 3.4.12 5.2.1.2 N/A 5.2-3 5.2.1.3 3.5.2 N/A N/A 5.2-3 BVPS-1 Licensing Requirement to PTLR Cross-Reference Licensing PTLR Requirement Section Figure Table LR 3.1.2 N/A N/A 5.2-3 LR 3.1.4 N/A N/A 5.2-3 LR 3.4.6 NIA N/A 5.2-3 Beaver Valley Unit 1 5.2-i PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 5.2 5.2.1 5.2.1.1 Pressure and Temperature Limits Report Reactor Coolant System (RCS} Pressure and Temperature Limits Report (PTLR}

The PTLR for Unit 1 has been prepared in accordance with the requirements of Technical Specification 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.

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

1.

LCO 3.4.3 Reactor Coolant System Pressure and Temperature (PIT)

Limits, 2.

LCO 3.4.6 RCS Loops - MODE 4, 3.

LCO 3.4.7 RCS Loops - MODE 5, Loops Filled, 4.

LCO 3.4.10 Pressurizer Safety Valves, 5.

LCO 3.4.12 Overpressure Protection System (OPPS),

6.

LCO 3.5.2 ECCS - Operating, 7.

LR 3.1.2 Boration Flow Paths - Operating, 8.

LR 3.1.4 Charging Pump - Operating, and 9.

LR 3.4.6 Pressurizer Safety Valve Lift Involving Liquid Water Discharge.

Operating Limits The PTLR limits for Beaver Valley Power Station (BVPS) Unit 1 have been prepared in accordance with the requirements of Technical Specification 5.6.4, using the methodology contained in Reference 1.

RCS Pressure and Temperature (PIT} Limits (LCO 3.4.3}

The RCS temperature rate-of-change limits are defined as:

a.

A maximum heatup of 100°F in any one hour period (Reference 2).

b.

A maximum cooldown of 100°F in any one hour period (Reference 2), and c.

A maximum temperature change of less than or equal to 5°F in any one hour period during inservice hydrostatic testing operations above system design pressure. This rate-of-change limit ensures that thermal gradient stress resulting from temperature change is not induced in the reactor vessel during inservice hydrostatic testing operations above system design pressure.

Beaver Valley Unit 1 5.2 - 1 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2 The RCS PIT limits for heatup, leak testing, and criticality are specified by Figure 5.2-1 and Table 5.2-1. The RCS PIT limits for cooldown are shown in Figure 5.2-2 and Table 5.2-2. These limits are defined in Reference 2.

Consistent with the methodology described in Reference 1, the RCS PIT limits for heatup and cooldown shown in Figures 5.2-1 and 5.2-2 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 10 CFR 50, Appendix G (Reference 5). The heatup and cooldown curves also include the effect of the reactor vessel flange.

The PIT 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 PIT curve for heatup and cooldown.

Pressure-temperature limit curves shown in Figure 5.2-3 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 Ill, 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.

- NOTE -

Pressure limits are considered to be met for pressures that are below 0 psig (i.e., up to and including full vacuum conditions) since the resulting PIT combination is located in the region to the right and below the operating limits provided in Figures 5.2-1, 5.2-2, and 5.2-3.

Figures 5.2-1 and 5.2-2 and Tables 5.2-1 and 5.2-2 are based upon analysis of all applicable surveillance capsules per Reference 2. Reference 2 provides an updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factor values, and an updated fluence evaluation. Therefore, the development of the PIT limit curves (Reference 2) utilized the revised information. Taking into account the updated surveillance data credibility evaluation, the Position 2.1 chemistry factor values, and the fluence analysis summarized in Reference 2, the limiting material for the current BVPS-1 PIT limits continues to be the lower shell plate B6903-1 at 50 EFPY.

Beaver Valley Unit 1 5.2-2 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.2 5.2.1.3 Using the fluence analysis provided in Section 2 of Reference 2, the neutron fluence value for lower shell plate 86903-1 at 50 EFPY is determined to be 5.89 x 1019 n/cm2 (E > 1.0 MeV). Using this updated fluence value along with the updated Position 2.1 chemistry factor value (Table 5.2-4) for this material, the limiting 1/4T and 3/4T adjusted reference temperature (ART) values are 244.0°F and 208.8°F, respectively, at 50 EFPY. Note that for conservatism, PIT limit curves were developed using 1/4T and 3/4T ART values of 244.5°F and 2O9.5°F, respectively (Reference 2).

Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12)

The power operated relief valves (PORVs) shall each have a nominal maximum lift setting and enable temperature in accordance with Table 5.2-3. The lift setting provided does not impose any reactor coolant pump restrictions.

The PORV setpoint is based on PIT 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 WCAP-14040-A, Revision 4 (Reference 1 ). The PORV lift setting (Reference 10) shown in Table 5.2-3 accounts for appropriate instrument error.

OPPS Enable Temperature (LCO 3.4.12)

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. Based on this method, the arming temperature (Reference 10) is 347°F with uncertainty for 50 EFPY.

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 RT NDT + 50°F), whichever is greater. The calculated enable temperature (Reference 10) is 345°F with uncertainty for 50 EFPY.

As the arming temperature is higher and, therefore, more conservative than the calculated enable temperature, the OPPS enable temperature, as shown in Table 5.2-3, is set to equal the arming temperature.

Beaver Valley Unit 1 5.2 - 3 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.4 5.2.2 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 5.2-3, 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 LCO 3.4.12. 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.

Reactor Vessel Boltup Temperature {LCO 3.4.3)

The minimum boltup temperature for the Reactor Vessel Flange shall beY 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.

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 5.2-1 and 5.2-2, and Tables 5.2-1 and 5.2-2 in this report. The time of specimen withdrawal may be modified to coincide with those refueling outages nearest 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, RT Nor, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RT Nor 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 8 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.

Beaver Valley Unit 1 5.2 - 4 PTLR Revision 8 LRM Revision 97

3 Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.3 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 PIT limits.

Table 5.2-4 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5.2-4a shows the Calculation of Chemistry Factors based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data.

Table 5.2-4b shows the St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Weld Data.

Table 5.2-5, taken from Reference 2, provides the reactor vessel beltline material property table.

Table 5.2-6, taken from Reference 2, shows the reactor vessel extended beltline material properties.

Table 5.2-7, taken from Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 50 EFPY.

Table 5.2-8, taken from Reference 2, shows the calculation of ARTs for 50 EFPY.

Table 5.2-9, taken from Reference 9, provides RT Prs values for the beltline materials at 50 EFPY.

Table 5.2-10, taken from Reference 9, provides RT Prs values for the extended beltline materials at 50 EFPY.

Table 5.2-11, provides Reactor Vessel Toughness Data (Unirradiated)

Beaver Valley Unit 1 5.2 - 5 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report 5.2.4 References Pressure and Temperature Limits Report 5.2 1.

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al., May 2004.

2.

WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," B.E. Mays, et al., June 2017.

3.

WCAP-17896-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," E.J. Long and E.T. Hayes, September 2014.

4.

WCAP-8457, "Duquesne Light Company, Beaver Valley Unit No. 1 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, October 1974.

5.

10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.

6.

10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No.

243, December 19, 1995. (PTS Rule) 7.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.

8.

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

9.

WCAP-15571, Supplement 1, Revision 2, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program,"

A. E. Freed, September 2011.

10.

L TR-SCS-16-58 Rev. 0, L TOPS Setpoint Evaluation for 50 EFPY for Beaver Valley Unit 1, June 2017.

11.

NUREG-0800, BTP 5-2 and 5-3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," March 2007.

Beaver Valley Unit 1 5.2 - 6 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate 86903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY:

1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 2250 2000 1750 1500 C,

"cii

1250 en en

CL 1000 ni u

750 500 250 0

0 joperlim Version:5. Run:1944 Operlin:1.xlsm Ver ion: 5.4!

I I

---. r****

. L I I -lr 50 Criticality Limit based on inservice hydrostatic test temperature (301 °F) for the service period up to 50 EFPY 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 5.2-1 (Page 1 of 1)

Reactor Coolant System Heatup Limitations Applicable for 50 EFPY (LCO 3.4.3)

Beaver Valley Unit 1 5.2 - 7 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate 86903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY:

1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 ---r-;::::======================--i---i----r---::--7

'in Q)...

U)

Q)

a.

"C Q)

Operlim Version:5.4 Run:19454 Operlim.xlsm Version: 5.4 2250 2000 1750 1500 1250 1000 750 Cooldown Rates

°F/Hr 500 Steady-State 20 40 60 100 I

250 0,-....-,-,--1.....-!L....,---,--.---l--,--,-i--.--r--,---,---1--,---,-,-,-+-,---,--,---,-..!--,--,-..,.....,.....h-,---r-,---l--,---,-,-,-+-,--,---,--r-+-,--,-

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 5.2-2 (Page 1 of 1)

Reactor Coolant System Cooldown Limitations Applicable for 50 EFPY (LCO 3.4.3)

Beaver Valley Unit 1 5.2 - 8 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual 2500 2000 6 1500 ci5 a:

en en w

Pressure and Temperature Limits Report 5.2

/

V

/

V

1000

i.---

500 0

50 60 70 80 90 100 TEMPERATURE (°F)

Figure 5.2-3 (Page 1 of 1) 110 120 Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)

Beaver Valley Unit 1 5.2-9 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 1 of 2)

Heatup Curve Data Points for 50 EFPY (LCO 3.4.3)

G0°F/hr Heatup T (°F)

P (psig) 60 0

60 602 65 602 70 602 75 602 80 602 85 602

90 602 95 602 100 602 105 602 110 603 115 604 120 606 125 609 130 612 135 616 140 621 145 627 150 633 155 640 160 648 165 657 170 667 175 678 180 691 185 704 190 719 195 736 200 755 205 775 210 798 215 823 220 851 225 882 Beaver Valley Unit 1 G0°F/hr 100° F /hr Heatu p Criticality T (°F)

P (psig)

T {°F)

P (psig) 301 0

60 0

301 1190 60 552 305 1241 65 552 310 1303 70 552 315 1358 75 552 320 1417 80 552 325 1483 85 552 330 1555 90 552 335 1636 95 552 340 1724 100 552 345 1821 105 552 350 1929 110 552 355 2048 115 552 360 2179 120 552 365 2324 125 552 370 2483 130 552 135 552 140 553 145 555 150 557 155 561 160 565 165 570 170 575 175 582 180 590 185 598 190 608 195 619 200 631 205 645 210 660 215 677 220 696 225 717 5.2 - 10 100°F/hr Criticality T (°F) 301 301 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 P (psig) 0 947 990 1042 1099 1162 1232 1310 1395 1488 1592 1706 1832 1971 2124 2292 2464 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 2 of 2)

Heatup Curve Data Points for 50 EFPY (LCO 3.4.3) 60°F/hr Heatup 60°F/hr 100°F/hr Heatup 100°F/hr Criticality Criticality T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig)

T (°F)

P (psig) 230 915 230 741 235 953 235 766 240 994 240 795 245 1040 245 827 250 1085 250 861 255 1132 255 900 260 1184 260 943 265 1241 265 990 270 1303 270 1042 275 1358 275 1099 280 1417 280 1162 285 1483 285 1232 290 1555 290 1310 295 1636 295 1395 300 1724 300 1488 305 1821 305 1592 310 1929 310 1706 315 2048 315 1832 320 2179 320 1971 325 2324 325 2124 330 2483 330 2292 335 2464 Leak Test Limit T (°F)

P (psig) 283 301 Beaver Valley Unit 1 5.2 - 11 2000 2485 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 1 of 2)

Cooldown Curve Data Points for 50 EFPY (LCO 3.4.3)

Steady State 20°F/hr 40°F/hr T

p T

p (OF)

(psig)

{OF)

{psig)

{OF)

(psig) 60 0

60 0

60 621 60 607 60 563 65 621 65 608 65 564 70 621 70 609 70 565 75 621 75 610 75 566 80 621 80 611 80 567 85 621 85 613 85 569 90 621 90 614 90 570 95 621 95 616 95 572 100 621 100 618 100 574 105 621 105 621 105 576 110 621 110 621 110 579 115 621 115 621 115 582 120 621 120 621 120 585 125 621 125 621 125 589 130 621 130 621 130 593 130 680 130 637 135 598 135 684 135 641 140 603 140 689 140 646 145 609 145 694 145 652 150 615 150 700 150 658 155 623 155 706 155 665 160 630 160 713 160 672 165 639 165 721 165 680 170 649 170 729 170 689 175 660 175 739 175 700 180 672 180 749 180 711 185 685 185 761 185 723 190 700 190 774 190 737 195 717 195 788 195 752 200 735 200 803 200 769 205 755 205 821 205 788 210 778 210 840 210 808 215 802 215 861 215 831 220 830 220 884 220 856 225 860 225 910 225 884 230 894 230 938 230 915 235 931 235 970 235 949 240 973 240 1004 240 987 245 1018 245 1043 245 1029 250 1069 250 1085 250 1075 255 1125 Beaver Valley Unit 1 5.2 - 12 60°F/hr T

p (OF)

(psig) 60 0

60 518 65 519 70 520 75 521 80 522 85 523 90 525 95 527 100 529 105 531 110 534 115 537 120 541 125 545 130 549 135 554 140 559 145 566 150 572 155 580 160 588 165 598 170 609 175 620 180 633 185 648 190 664 195 682 200 702 205 724 210 748 215 775 220 805 225 838 230 875 235 916 240 961 245 1011 250 1067 255 1125 100°F/hr T

p

{OF)

(psig) 60 0

60 426 65 426 70 427 75 428 80 429 85 431 90 432 95 434 100 436 105 439 110 442 115 445 120 449 125 453 130 458 135 464 140 470 145 477 150 485 155 494 160 504 165 515 170 527 175 541 180 556 185 573 190 593 195 614 200 637 205 664 210 693 215 725 220 761 225 801 230 846 235 895 240 949 245 1010 250 1067 255 1125 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 2 of 2)

Cooldown Curve Data Points for 50 EFPY (LCO 3.4.3)

Steady State 20°F/hr T

p T

p (OF)

(psig)

(OF)

(psig) 255 1132 255 1127 260 1184 260 1183 265 1241 265 1241 270 1305 270 1305 275 1375 275 1375 280 1452 280 1452 285 1537 285 1537 290 1632 290 1632 295 1736 295 1736 300 1851 300 1851 305 1979 305 1979 310 2120 310 2120 315 2275 315 2275 320 2448 320 2448 Beaver Valley Unit 1 40°F/hr 60°F/hr T

p T

p (OF)

(psig)

(OF)

(psig) 260 1183 260 1183 265 1241 265 1241 270 1305 270 1305 275 1375 275 1375 280 1452 280 1452 285 1537 285 1537 290 1632 290 1632 295 1736 295 1736 300 1851 300 1851 305 1979 305 1979 310 2120 310 2120 315 2275 315 2275 320 2448 320 2448 5.2 - 13 100°F/hr T

p (OF)

(psig) 260 1183 265 1241 270 1305 275 1375 280 1452 285 1537 290 1632 295 1736 300 1851 305 1979 310 2120 315 2275 320 2448 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-3 (Page 1 of 1)

Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12)

FUNCTION OPPS Enable Temperature PORV Setpoint Beaver Valley Unit 1 5.2 - 14 SETPOINT 347°F s 397 psig PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4 (Page 1 of 1)

Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f(a)

FF<b) lRT NDic)

FF *lRTNDT FF2 Lower Shell V

0.297 0.6677 127.9 85.40 0.446 Plate u

0.618 0.8652 118.3 102.35 0.749 B6903-1(d) w 0.952 0.9862 147.7 145.66 0.973 (Longitudinal) y 2.10 1.2018 141.7 170.30 1.444 X

4.99 1.4020 175.8 246.46 1.965 Lower Shell V

0.297 0.6677 138.0 92.14 0.446 Plate u

0.618 0.8652 132.1 114.29 0.749 B6903-1(d) w 0.952 0.9862 180.2 177.72 0.973 (Transverse) y 2.10 1.2018 166.9 200.58 1.444 X

4.99 1.4020 179.0 250.95 1.965 SUM:

1585.86 11.154 CF = L(FF

lRT NDT) + L(FF2) = (1585.86) + (11.154) = 142.2°F(e)

V 169.4 0.297 0.6677 (159.8) 113.10 0.446 Beaver Valley u

174.8 0.618 0.8652 (164.9) 151.23 0.749 Unit 1 197.5 Surveillance w

0.952 0.9862 (186.3) 194.76 0.973 Weld Metal<d)

(Heat # 305424) y 189.2 2.10 1.2018 (178.5) 227.40 1.444 X

252.1 4.99 1.4020 (237.8) 353.39 1.965 SUM:

1039.87 5.577 CF = L{FF

lRT NDT) + L{FF2) = {1039.87) + {5.577) = 186.5°F(e)

Notes:

(a) f = Calculated surveillance capsule neutron fluence (x 1019 n/cm2, E > 1.0 MeV). The surveillance capsule fluence results are contained in Table 4-1 of Reference 2.

(b)

FF = fluence factor = f <0 0-1

109 f)_

(c) lRT NoT values are the measured 30 ft-lb shift values. The Beaver Valley Unit 1 lRT NDT values for the surveillance weld data are adjusted by a ratio of 1.06. Pre-adjusted values are listed in parentheses, and were taken from Table 4-1 of Reference 2.

NOTE:

Per Regulatory Guide 1.99, Revision 2 (Reference 7), section 2.1 "Radiation Embrittlement of Reactor Vessel Materials," the vessel weld chemistry factor is divided by the surveillance weld chemistry factor to obtain a ratio factor to multiply the lRT Nor values by to obtain adjusted lRT NDT values. In Table 5-2 of Reference 2, the ratio is determined to be 1.06 or (191.7/181.6).

(d)

The plate and weld surveillance data is deemed non-credible per Appendix D of Reference 2.

(e)

Position 2.1 chemistry factor values are summarized in Table 5-4 of Reference 2.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 15 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4a (Page 1 of 2)

Calculation of Chemistry Factors<a)

(Based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data)

Material Capsule Capsule f(b)

FF<c) eRT NDid)

FF *eRTNDT FF2 97° 0.5174 0.8160 82.6 67.44 0.666 (72.34)

Weld Metal 81.1 Heat # 90135<e) 104° 0.7885 0.9333 (67.4) 75.68 0.871 (St. Lucie Unit 1) 83.8 284° 1.243 1.0606 (68.0) 88.85 1.125 Weld Metal 67.5 Heat # 90135<e) 97° 0.324 0.6902 (65.93) 46.61 0.476 (Millstone Unit 2) 57.0 104° 0.949 0.9853 (52.12) 56.18 0.971 61.4 83° 1.74 1.1523 (56.09) 70.74 1.328 SUM:

405.50 5.437 CF = L(FF

eRT NDT) + L(FF2) = (405.50) + (5.437) = 74.6°F<9)

W-225 0.488 0.800 197.30 157.83 0.640 Weld Metal (210)

Heat # 305414<t)

W-265 0.847 0.953 218.30 208.13 0.909 (Fort Calhoun (225)

Unit 1)

W-275 1.54 1.119 215.90 241.68 1.253 (219)

SUM:

607.64 2.802 CF= L(FF

eRT Nor)+ L(FF2) = (607.64) + (2.802) = 216.9°F<9>

Notes for Table 5.2-4a are on the following page.

Beaver Valley Unit 1 5.2 - 16 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4a (Page 2 of 2)

Calculation of Chemistry Factors<a>

(Based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data)

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 )." As a result of the unclear identification of the Millstone Unit 2 surveillance weld heat number, the Millstone Unit 2 data was not originally incorporated into Beaver Valley Unit 1 chemistry factor calculations. Since the Millstone Unit 2 surveillance weld contains specimens made of Heat# 90136, the use of this data is appropriate. See Appendix D of Reference 2 for more details.

(b) f = calculated surveillance capsule fluence values (x 1019 n/cm2, E > 1.0 MeV). The surveillance capsule fluence results for St. Lucie Unit 1 and Millstone Unit 2 are contained in Table 4-2 of Reference 2. The surveillance capsule fluence results for Fort Calhoun Unit 1 are contained in Table D-5 of Reference 3.

(c)

FF= fluence factor= f <0-28-0*1 *logf)_

(d)

LiRT NOT values are the measured 30 ft-lb. shift values. \\RT NOT values for the surveillance weld data are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the beltline weld chemistry. Pre-adjusted values are listed in parentheses, and were taken from Tables 4-2 of Reference 2 and Table A-5 of Reference 9. The temperature adjustments for each capsule were calculated from the data in Table 5.2-4b and the average plant irradiation temperature for BV-1. The St. Lucie Unit 1 \\RT NOT values for the weld data are adjusted by a ratio of 1.17. The Millstone Unit 2 and Fort Calhoun \\RT NOT values were not adjusted since the ratio was less than 1.00; therefore, a conservative value of 1.00 was used.

(e)

The St. Lucie Unit 1 and Millstone Unit 2 surveillance data is deemed credible per Appendix D of Reference 2; however, a full margin term should be utilized for conservatism when this data is applied as a result of the unclear identification of the Millstone Unit 2 weld specimen heat numbers. See Appendix D of Reference 2 for more details.

(f)

The Fort Calhoun Unit 1 surveillance data is deemed non-credible per Appendix D of Reference 3.

(g)

Position 2.1 chemistry factor values are summarized in Table 5-4 of Reference 2.

Beaver Valley Unit 1 5.2 - 17 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4b (Page 1 of 1)

St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Weld Data<a}(b)

Cu Ni Irradiated Capsule f(d)

RT NDie)

Material Capsule (wt.%)

(wt.%)

Temperature<c>

(x1019 n/cm2,

(OF)

E > 1.0 MeV)

Weld Metal 97° 0.23 0.07 541 0.5174 72.34 Heat# 90136 104° 0.23 0.07 544.6 0.7885 67.4 (St. Lucie Unit 1) 284° 0.23 0.07 546.3 1.243 68.0 Weld Metal 97° 0.30 0.06 544.3 0.324 65.93 Heat# 90136 104° 0.30 0.06 547.6 0.949 52.12 (Millstone Unit 2) 83° 0.30 0.06 548.0 1.74 56.09 Weld Metal W-225 0.35 0.60 530 0.488 210 Heat W-265 0.35 0.60 536 0.847 225

305414 (Fort Calhoun W-275 0.35 0.60 539.6 1.54 219 Unit 1)

(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 )." As a result of the unclear identification of the Millstone Unit 2 surveillance weld heat number, the Millstone Unit 2 data was not originally incorporated into Beaver Valley Unit 1 chemistry factor calculations. Since the Millstone Unit 2 surveillance weld contains specimens made of Heat# 90136, the use of this data is appropriate. See Appendix D of Reference 2 for more details.

(b)

Data contained in this table was obtained from Reference 2, unless otherwise noted.

(c)

Irradiated temperatures are the average inlet temperatures over the specific cycles corresponding to the operating time experienced by each of the respective capsules.

(d) f = calculated surveillance capsule fluence values.

(e)

[RT Nor values are the measured 30 ft-lb shift values from Table 4-2 of Reference 2 and Table D-5 of Reference 3.

Beaver Valley Unit 1 5.2 - 18 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-5 (Page 1 of 1)

Reactor Vessel Beltline Material Properties Cu Ni Position 1. 1 Initial Material Description (wt.%)

(wt.%)

Chemistry RT Noia)

Factor (OF)

(OF)

Intermediate Shell Plate 86607-1 0.14 0.62 100.5 26.8 Intermediate Shell Plate 86607-2 0.14 0.62 100.5 53.6 Lower Shell Plate 86903-1 0.21 0.54 147.2 13.1 Lower Shell Plate 87203-2 0.14 0.57 98.7 0.4 Intermediate to Lower Shell Weld 0.27 0.07 124.3

-56 Seam (Heat# 90136)11-714 Intermediate Longitudinal Shell Weld 0.28 0.63 191.7

-56 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 RT NDT values for the plates are based on measured data while the weld values are generic.

Beaver Valley Unit 1 5.2-19 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 1 of 2)

Reactor Vessel Extended 8eltline Material Properties<a>

Material Heat Number Wt%

Wt%

Material Description ID (Lot Number)

Cu Ni Upper Shell Forging 86604 123V339VA1 o.12

0.68 305414 0.34 0.61 (3951 & 3958)

Upper to Intermediate 10-714 AOFJ 0.03 0.93 Shell Girth Weld FOIJ 0.03 0.94 EODJ 0.02 1.04 HOCJ 0.02 0.93 B6608-1 95443-1 0.10 0.82 Inlet Nozzles B6608-2 95460-1 0.10 0.82 B6608-3 95712-1 0.08 0.79 EODJ 0.02 1.04 FOIJ 0.03 0.94 1-717B HOCJ 0.02 0.93 Inlet Nozzle Welds 1-717D O8IJ 0.02 0.97 1-717F EOEJ 0.01 1.03 ICJJ 0.03 0.99 JACJ 0.04 0.97 B6605-1 95415-1 0.13(d) 0.77 Outlet Nozzles 86605-2 95415-2 0.13(d) 0.77 B6605-3 95444-1 0.09 0.79 ICJJ 0.03 0.99 IOBJ 0.02 0.97 1-717A JACJ 0.04 0.97 Outlet Nozzle Welds 1-717C HOCJ 0.02 0.93 1-717E EODJ 0.02 1.04 FOIJ 0.03 0.94 Notes for Table 5.2-6 are on the following page.

Beaver Valley Unit 1 5.2 - 20 Initial RT NDic)

(OF) 40

-56 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 48.5

-15.2 11.4 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen)

-26.2 3.3 10.1 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen)

PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 2 of 2)

Reactor Vessel Extended Beltline Material Properties<a)

Notes:

(a) Data obtained from Table 3-2 of Reference 2.

(b) The Cu wt % was not available from the CMTR so in accordance with Regulatory Guide 1.99, Revision 2, a standard deviation analysis (average+ standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Shell Forgings (55 data points).

(c) The initial RT NDT value for the upper shell forging, inlet nozzle forgings, and outlet nozzle forgings are based on measured values. The generic initial RT NDT values for the weld materials were determined in accordance with NUREG-0800 [Reference 11] and 10 CFR 50.61 [Reference 6].

(d) The Cu wt% was not available from the CMTR, so in accordance with Regulatory Guide 1.99, Revision 2, a standard deviation analysis (average+ standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Nozzle Forgings (178 data points).

Beaver Valley Unit 1 5.2 - 21 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-7 (Page 1 of 1)

Calculation of Adjusted Reference Temperatures (ARTs) for 50 EFPY(d) 50 EFPY Material Description 1I4T ART(a) 314T ART(a)

(OF)

Intermediate Shell Plate 86607-1 195.2 Intermediate Shell Plate 86607-2 222.0 Lower Shell Plate 87203-2 166.4 Lower Shell Plate 86903-1 244.0(f)

- Using SIC Data(b) 237.3 Intermediate Shell Longitudinal Weld 19-714A/B 182.4

- Using SIC Data(b) 177.7 Intermediate to Lower Shell Circ. Weld 11-714 175.7

- Using SIC Data (c) 109.3 Lower Shell Longitudinal Weld 20-714AIB 199.9

- Using SIC Data(d) 205.6 Upper Shell Forging 86604 139.4 Upper Shell to Intermediate Shell Girth Weld 10-714 172.9 (Heat # 305414)

-Using SIC Data(d) 177.9 Upper Shell to Intermediate Shell Girth Weld 10-714 88.4 (Heat #'s AOFJ and FOIJ)

Upper Shell to Intermediate Shell Girth Weld 10-714 44.0 (Heat #'s EODJ and HOCJ)

Inlet and Outlet Nozzle Welds (All Heat #'s) 44.0 Notes:

(a) ART= I+ dRT Nor+ M.

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

(OF) 171.2 198.0 142.8 208.8(t) 203.3 133.5 130.2 146.0 91.4 146.2 150.3 119.2 122.5 125.9 44.0 44.0 44.0 (c) Based on St. Lucie Unit 1 and Millstone Unit 2 surveillance data. (Data credible.

ART calculated with a full crA per Appendix D of Reference 2.)

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

(e) Data obtained from Tables 7-2 and 7-3 of Reference 2. Nozzle ART values are excluded from this table, as these values are calculated using surface fluence values. See Reference 2 for nozzle ART values.

(f) For the purposes of PIT limit curve development, a 1I4T ART value of 244.5°F and a 314T ART value of 209.5°F were used for conservatism.

Beaver Valley Unit 1 5.2 - 22 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-8 (Page 1 of 1)

Calculation of Adjusted Reference Temperatures (ARTs) for 50 EFPY(c)

Parameter VALUES Operating Time 50 EFPY Material Plate 86903-1 Location Lower Shell Plate 1/4T ART(°F)

Chemistry Factor, CF (°F) 147.2 Fluence (f), n/cm2 (E>1.0 Mev) 3.672 X 1019 Fluence Factor, FF 1.3374

[RT NOT = CF X FF(°F) 196.9(b}

Initial RT NOT, I(°F)(a) 13.1 Margin, M(°F) 34(b)

ART = l+(CF*FF)+M, °F per RG 1.99, Revision 2 244.Q(d}

Notes:

(a)

Initial RT NOT values are measured values for plate material.

(b)

Based on Regulatory Guide 1.99, Revision 2 Position 1.1.

Plate 86903-1 Lower Shell Plate 3/4T ART(°F) 147.2 1.427 X 1019 1.0987 161.7(b) 13.1 34(b) 208.B(d>

(Surveillance data not credible. ART calculated with a full crti.)

(c)

Data obtained from Tables 7-2 and 7-3 of Reference 2.

(d)

For the purposes of PIT limit curve development, a 1/4T ART value of 244.5°F and a 3/4T ART value of 209.5°F were used for conservatism.

Beaver Valley Unit 1 5.2 - 23 PTLR Revision 8 LRM Revision 97

Licensing Requirements Manual Table 5.2-9 (Page 1 of 2)

Pressure and Temperature Limits Report 5.2 RT PTs Calculation for 8eltline Region Materials at Life Extension (50 EFPY)(a)

Material Heat Surface Fluence Chemistry Material Description Intermediate Shell Plate Intermediate Shell Plate Lower Shell Plate ID 86607-1 86607-2 86903-1 Number

+ Using non-credible surveillance data(9>

Lower Shell Plate Intermediate to Lower Shell Girth Weld 87203-2 11-714 90136

+ Using credible surveillance data(h)

Intermediate Shell 19-714 305424 Longitudinal Weld A&B

+ Using non-credible surveillance data(9>

Lower Shell Longitudinal 20-714 305414 Weld A&8

+ Using non-credible surveillance data(i)

Notes:
(a) Data obtained from Table 6-3 of Reference 9.
(b)
FF = fluence factor= f (0-28-0*10109 (t))_
Fluence
Factor, (x1019 n/cm2)
FF(b) 5.57 1.4231 5.57 1.4231 5.57 1.4231 5.57 1.4231 5.57 1.4231 5.55 1.4225 5.55 1.4225 1.08 1.0224 1.08 1.0224 1.09 1.0241 1.09 1.0241 (c)
Initial RT NOT values are measured values with the exception of the vessel welds.
(d) cRT PTS = CF* FF.
(e) M = 2 *(cru2 + crl)112*
(f)
RT PTs = Initial RT NOT+ cRT PTs + Margin.
Beaver Valley Unit 1 5.2 - 24 Factor (OF) 100.5 100.5 147.2 151.8 98.7 124.3 87.1 191.7 192.3 210.5 216.9 Initial cRT PTS(d)
Ou RT No/c)
(°F)
(OF)
(OF) 43 143.0 0
73 143.0 0
27 209.5 0
27 216.0 0
20 140.5 0
-56 176.8 17
-56 123.9 17
-56 196.0 17
-56 196.6 17
-56 215.6 17
-56 222.1 17 a
(OF) 17 17 17 17(g) 17 28 14(h) 28 28(g) 28 28(i)
Margin(e)
(OF) 34 34 34 34 34 65.5 44.0 65.5 65.5 65.5 65.5 RT PTS(f)
(OF) 220.0 250.0 270.5 277.0 194.5 186.3 111.9 205.5 206.1 225.1 231.6 PTLR Revision 8 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 2 of 2)
RT PTS Calculation for Beltline Region Materials at Life Extension (50 EFPY)(a)
Notes continued:
(g) The BVPS-1 surveillance weld metal is the same weld heat as the BVPS-1 intermediate shell longitudinal welds (heat 305424). The BVPS-1 surveillance weld data is non credible; therefore, the higher ob. term of 28°F was utilized for BVPS-1 weld heat 305424.
The BVPS-1 surveillance plate material is representative of the BVPS-1 lower shell plate 86903-1. The surveillance plate material is non-credible; therefore, the higher Ob. term of 17° F was utilized for BVPS-1 plate 86903-1. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
(h) The St. Lucie Unit 1 surveillance weld metal is the same weld heat as the BVPS-1 intermediate to lower shell girth weld (heat 90136). The St. Lucie Unit 1 surveillance weld data is credible; therefore, the reduced Ob. term of 14°F was utilized for BVPS-1 weld heat 90136. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
(i)
The Fort Calhoun surveillance weld metal is the same weld heat as the BVPS-1 lower shell longitudinal welds (heat 305414). The Fort Calhoun surveillance weld data is non credible; therefore, the higher ob. term of 28°F was utilized for BVPS-1 weld heat 305414.
The credibility evaluation conclusions are contained in Appendix A of Reference 9.
Beaver Valley Unit 1 5.2 - 25 PTLR Revision 8 LRM Revision 97
Licensing Requirements Manual Table 5.2-10 (Page 1 of 2)
Pressure and Temperature Limits Report 5.2 RT Prs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)(a)
Material Material Heat Number Surface Fluence Chemistry Fluence
Factor, Factor Description ID (Lot Number)
(x1019 n/cm2)
FF(b)
(OF)
Upper Shell 86604 123V339VA1 0.625 0.8685 84.2 Forging Upper to 305414 Intermediate 10-714 (3951 & 3958) 0.625 0.8685 209.11 Shell Girth Weld
--+ Using non-credible surveillance data<g>
0.625 0.8685 216.9 Upper to AOFJ 0.625 0.8685 41.0 FOIJ 0.625 0.8685 41.0 Intermediate 10-714 EODJ 0.625 0.8685 27.0 Shell Girth Weld HOCJ 0.625 0.8685 27.0 86608-1 95443-1 0.016 0.1513 67.0 In let Nozzles 86608-2 95460-1 0.016 0.1513 67.0 86608-3 95712-1 0.016 0.1513 51.0 EODJ 0.016 0.1513 27.0 FOIJ 0.016 0.1513 41.0 Inlet Nozzle 1-717 8 HOCJ 0.016 0.1513 27.0 1-717 D D8IJ 0.016 0.1513 27.0 Welds 1-717 F EOEJ 0.016 0.1513 20.0 ICJJ 0.016 0.1513 41.0 JACJ 0.016 0.1513 54.0 86605-1 95415-1 0.011 0.1191 95.25 Outlet Nozzles 86605-2 95415-2 0.011 0.1191 95.25 86605-3 95444-1 0.011 0.1191 58.0 ICJJ 0.011 0.1191 41.0 1-717 A IO8J 0.011 0.1191 27.0 Outlet Nozzle JACJ 0.011 0.1191 54.0 Welds 1-717 C HOCJ 0.011 0.1191 27.0 1-717 E EODJ 0.011 0.1191 27.0 FOIJ 0.011 0.1191 41.0 Notes:
(a)
Data obtained from Table 6-4 of Reference 9.
Beaver Valley Unit 1 5.2 - 26 Initial RT PTS(d)
RT NDic)
Ou (OF)
(OF)
(OF) 40 73.1 0
-56 181.6 17
-56 188.4 17 10 35.6 17 10 35.6 17 10 23.4 17 10 23.4 17 60 10.1 17 60 10.1 17 60 7.7 17 10 4.1 17 10 6.2 17 10 4.1 17 10 4.1 17 10 3.0 17 10 6.2 17 10 8.2 17 60 11.3 17 60 11.3 17 60 6.9 17 10 4.9 17 10 3.2 17 10 6.4 17 10 3.2 17 10 3.2 17 10 4.9 17 0,1 (OF) 17 28 28(9) 17.8 17.8 11.7 11.7 5.1 5.1 3.9 2.0 3.1 2.0 2.0 1.5 3.1 4.1 5.7 5.7 3.5 2.4 1.6 3.2 1.6 1.6 2.4 Margin(e)
RT prs(f)
(OF)
(OF) 34 147.1 65.5 191.1 65.5 197.9 49.2 94.8 49.2 94.8 41.3 74.8 41.3 74.8 35.5 105.6 35.5 105.6 34.9 102.6 34.2 48.3 34.6 50.8 34.2 48.3 34.2 48.3 34.1 47.2 34.6 50.8 35.0 53.1 35.8 107.2 35.8 107.2 34.7 101.6 34.3 49.2 34.2 47.4 34.6 51.0 34.2 47.4 34.2 47.4 34.3 49.2 PTLR Revision 8 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 0 (Page 2 of 2)
RT PTs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)'a)
Notes continued:
(b)
FF = fluence factor= f (0 0-10109 (t))_
(c)
Initial RT NOT value for the upper shell forging is a measured value. All other values are generic.
(d) QRT PTS = CF* FF.
(e)
M = 2 *(cru2 + crl)112.
(f)
RT PTs = Initial RT NOT+ QRT PTs + Margin.
(g)
The Fort Calhoun surveillance weld metal is the same weld heat as the BVPS-1 upper to intermediate shell girth weld (heat 305414). The Fort Calhoun surveillance weld data is non-credible; therefore, the higher Ot,. term of 28°F was utilized for BVPS-1 weld heat 305414. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
Beaver Valley Unit 1 5.2 - 27 PTLR Revision 8 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table5.2-11 (Page 1 of1)
Reactor Vessel Toughness Data (Unirradiated)
Cu Ni p
TNDT COMPONENT HEAT NO.
CODE NO.
MATERIAL TYPE
(%)
(%)
(%)
(OF)
Closure Head C6213-1 B 86610 A533B CL. 1
.15
.010
-40 Dome Closure Head A5518-2 86611 A533B CL. 1
.14
.015
-20 Seg.
Closure Head ZV3758 A508 CL. 2
.08
.007 60*
FlanQe Vessel Flam:ie ZV-3661 FV-2961 A508 CL. 2
.12
.010
-54.7**
Inlet Nozzle 9-5443-1 86608-1 A508 CL. 2
.10
.82
.008 35.8**
Inlet Nozzle 9-5460-1 86608-2 A508 CL. 2
.10
.82
.010
-18.3**
Inlet Nozzle 9-5712-1 86608-3 A508 CL. 2
.08
.79
.007
-2.5**
Outlet Nozzle 9-5415-1 B6605-1 A508 CL. 2
.13
.77
.008
-26.2**
Outlet Nozzle 9-5415-2 86605-2 A508 CL. 2
.13
.77
.007 3.0**
Outlet Nozzle 9-5444-1 86605-3 A508 CL. 2
.09
.79
.007 10.1**
Upper Shell 123V339VA1 A508 CL. 2
.12
.68
.010 40 Inter Shell C4381-2 86607-2 A5338 CL. 1
.14
.62
.015
-10 Inter Shell C4381-1 86607-1 A5338 CL. 1
.14
.62
.015
-10 Lower Shell C6317-1 86903-1 A5338 CL. 1
.21
.54
.010
-50 Lower Shell C6293-2 87203-2 A5338 CL. 1
.14
.57
.015
-20 Trans Ring 123V223 A508 CL. 2 30 Bottom Hd Seg C4423-3 86618 A5338 CL. 1
.13
.008
-30 Bottom Hd Dome C4482-1 86619 A5338 CL. 1
.13
.015
-50 Inter to Lower 90136
.27
.07 Shell Weld Inter Shell Long.
305424
.28
.63 Weld Lower Shell 305414
.34
.61 Long. Weld Weld HAZ
-40
Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2
Estimated Per BWRVIP-173-A, Alternate Approach 2 MWD - Major Working Direction NMWD - Normal to Major Working Direction RTNDT UPPER SHELF ENERGY (FT-LB)
(OF)
MWD NMWD 0*
121
-20*
131 60*
>100 10**
166 48.5**
82.5
-15.2**
94 11.4**
97
-26.2**
93 3.3**
112.5 10.1 **
103 40*
155 101 53.6 123 83 26.8 128.5 94 13.1 134 83 0.4 129.5 85 30*
143
-29*
124
-33*
125.5
-56
> 100
-56
> 100
-56
> 100
-40 136.5 Note:
For evaluation of lnservice 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.
Beaver Valley Unit 1 5.2 - 28 PTLR Revision 8 LRM Revision 97 L-17-277 Beaver Valley Power Station, Unit No. 1 Pressure and Temperature Limits report, Revision 9 (29 pages follow)
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Pressure and Temperature Limits Report BVPS-1 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.4.3 5.2.1.1 5.2-1 NIA 5.2-2 3.4.6 NIA NIA 5.2-3 3.4.7 NIA NIA 5.2-3 3.4.10 NIA NIA 5.2-3 3.4.12 5.2.1.2 NIA 5.2-3 5.2.1.3 3.5.2 NIA NIA 5.2-3 BVPS-1 Licensing Requirement to PTLR Cross-Reference Licensing PTLR Requirement Section Figure Table LR 3.1.2 NIA NIA 5.2-3 LR 3.1.4 NIA NIA 5.2-3 LR 3.4.6 NIA NIA 5.2-3 Beaver Valley Unit 1 5.2 - i PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 5.2 Pressure and Temperature Limits Report Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
The PTLR for Unit 1 has been prepared in accordance with the requirements of Technical Specification 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications (TS) and Licensing Requirements (LR) addressed, or made reference to, in this report are listed below:
1.
LCO 3.4.3 Reactor Coolant System Pressure and Temperature (PIT)
Limits, 2.
LCO 3.4.6 RCS Loops - MODE 4, 3.
LCO 3.4.7 RCS Loops - MODE 5, Loops Filled, 4.
LCO 3.4.10 Pressurizer Safety Valves, 5.
LCO 3.4.12 Overpressure Protection System (OPPS),
6.
LCO 3.5.2 ECCS - Operating, 7.
LR 3.1.2 Boration Flow Paths - Operating, 8.
LR 3.1.4 Charging Pump - Operating, and 9.
LR 3.4.6 Pressurizer Safety Valve Lift Involving Liquid Water Discharge.
5.2.1 Operating Limits 5.2.1.1 The PTLR limits for Beaver Valley Power Station (BVPS) Unit 1 have been prepared in accordance with the requirements of Technical Specification 5.6.4, using the methodology contained in Reference 1.
RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)
The RCS temperature rate-of-change limits are defined as:
a.
A maximum heatup of 100°F in any one hour period (Reference 2).
b.
A maximum cooldown of 100°F in any one hour period (Reference 2), and c.
A maximum temperature change of less than or equal to 5°F in any one hour period during inservice hydrostatic testing operations above system design pressure. This rate-of-change limit ensures that thermal gradient stress resulting from temperature change is not induced in the reactor vessel during inservice hydrostatic testing operations above system design pressure.
Beaver Valley Unit 1 5.2 - 1 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2 The RCS PIT limits for heatup, leak testing, and criticality are specified by Figure 5.2-1 and Table 5.2-1. The RCS PIT limits for cooldown are shown in Figure 5.2-2 and Table 5.2-2. These limits are defined in Reference 2.
Consistent with the methodology described in Reference 1, the RCS PIT limits for heatup and cooldown shown in Figures 5.2-1 and 5.2-2 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 10 CFR 50, Appendix G (Reference 5). The heatup and cooldown curves also include the effect of the reactor vessel flange.
The PIT 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 PIT curve for heatup and cooldown.
Pressure-temperature limit curves shown in Figure 5.2-3 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 Ill, 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.
- NOTE Pressure limits are considered to be met for pressures that are below 0 psig (i.e., up to and including full vacuum conditions) since the resulting PIT combination is located in the region to the right and below the operating limits provided in Figures 5.2-1, 5.2-2, and 5.2-3.
Figures 5.2-1 and 5.2-2 and Tables 5.2-1 and 5.2-2 are based upon analysis of all applicable surveillance capsules per Reference 2. Reference 2 provides an updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factor values, and an updated fluence evaluation. Therefore, the development of the PIT limit curves (Reference 2) utilized the revised information. Taking into account the updated surveillance data credibility evaluation, the Position 2.1 chemistry factor values, and the fluence analysis summarized in Reference 2, the limiting material for the current BVPS-1 PIT limits continues to be the lower shell plate 86903-1 at 50 EFPY.
Beaver Valley Unit 1 5.2-2 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.2 5.2.1.3 Using the fluence analysis provided in Section 2 of Reference 2, the neutron fluence value for lower shell plate 86903-1 at 50 EFPY is determined to be 5.89 x 1019 n/cm2 (E > 1.0 MeV). Using this updated fluence value along with the updated Position 2.1 chemistry factor value (Table 5.2-4) for this material, the limiting 1/4T and 3/4T adjusted reference temperature (ART) values are 244.0°F and 208.8°F, respectively, at 50 EFPY. Note that for conservatism, PIT limit curves were developed using 1/4T and 3/4T ART values of 244.5°F and 209.5°F, respectively (Reference 2).
Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12)
The power operated relief valves (PORVs) shall each have a nominal maximum lift setting and enable temperature in accordance with Table 5.2-3. The lift setting provided does not impose any reactor coolant pump restrictions.
The PORV setpoint is based on PIT 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 WCAP-14040-A, Revision 4 (Reference 1 ). The PORV lift setting (Reference 10) shown in Table 5.2-3 accounts for appropriate instrument error.
OPPS Enable Temperature (LCO 3.4.12)
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. Based on this method, the arming temperature (Reference 10) is 34 7°F with uncertainty for 50 EFPY.
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 RT NDT + 50°F), whichever is greater. The calculated enable temperature (Reference 10) is 345°F with uncertainty for 50 EFPY.
As the arming temperature is higher and, therefore, more conservative than the calculated enable temperature, the OPPS enable temperature, as shown in Table 5.2-3, is set to equal the arming temperature.
Beaver Valley Unit 1 5.2-3 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.4 5.2.2 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 5.2-3, 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 LCO 3.4.12. 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.
Reactor Vessel Boltup Temperature (LCO 3.4.3)
The minimum boltup temperature for the Reactor Vessel Flange shall beX 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.
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 5.2-1 and 5.2-2, and Tables 5.2-1 and 5.2-2 in this report. The time of specimen withdrawal may be modified to coincide with those refueling outages nearest 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, RT NOT, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RT NOT 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 8 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.
Beaver Valley Unit 1 5.2 - 4 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report 5.2.3 Supplemental Data Tables Pressure and Temperature Limits Report 5.2 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 PIT limits.
Table 5.2-4 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
Table 5.2-4a shows the Calculation of Chemistry Factors based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data.
Table 5.2-4b shows the St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Weld Data.
Table 5.2-5, taken from Reference 2, provides the reactor vessel beltline material property table.
Table 5.2-6, taken from Reference 2, shows the reactor vessel extended beltline material properties.
Table 5.2-7, taken from Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 50 EFPY.
Table 5.2-8, taken from Reference 2, shows the calculation of ARTs for 50 EFPY.
Table 5.2-9, taken from Reference 9, provides RT PTs values for the beltline materials at 50 EFPY.
Table 5.2-10, taken from Reference 9, provides RT PTs values for the extended beltline materials at 50 EFPY.
Table 5.2-11, provides Reactor Vessel Toughness Data (Unirradiated)
Beaver Valley Unit 1 5.2 - 5 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report 5.2.4 References Pressure and Temperature Limits Report 5.2 1.
WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al., May 2004.
2.
WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," B.E. Mays, et al., June 2017.
3.
WCAP-17896-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," E.J. Long and E.T. Hayes, September 2014.
4.
WCAP-8457, "Duquesne Light Company, Beaver Valley Unit No. 1 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, October 1974.
5.
10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
6.
10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No.
243, December 19, 1995. {PTS Rule) 7.
Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
8.
FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
9.
WCAP-15571, Supplement 1, Revision 2, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program,"
A. E. Freed, September 2011.
10.
L TR-SCS-16-58 Rev. 0, L TOPS Setpoint Evaluation for 50 EFPY for Beaver Valley Unit 1, June 2017.
11.
NUREG-0800, BTP 5-2 and 5-3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," March 2007.
Beaver Valley Unit 1 5.2 - 6 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate 86903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY:
1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 2250 2000 1750 1500 en

1250 U)
U) 1000 ni
(.)
750 500 250 0
0 Operiim Version:5.4 Run:19454 Operlim.xlsm Version: 5.4 50 Criticality Limit based on inservice hydrostatic test temperature (301 °F) for the service period up to 50 EFPY 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5.2-1 (Page 1 of 1)
Reactor Coolant System Heatup Limitations Applicable for 50 EFPY (LCO 3.4.3)
Beaver Valley Unit 1 5.2-7 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate 86903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY:
1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2250 2000 1750 1500 (1) 1250 ti) ti)
(1)
"C 1000 (1) 750 Cooldown Rates
°F/Hr 500 Steady-State 20 40 60 100 250 0
--+--,--r-r-,--+-+---,-,--,-+-,--,--,.-,-+-,--,--,-,--,i--,--,----,-,--+-r-,--,--,.---t-,--,--,--,--+-,r-r--r---,--+--,--,-,.......--+-,----,--,--,-+-,--,-,---,--1 0
Beaver Valley Unit 1 50 100 150
. 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5.2-2 (Page 1 of 1)
Reactor Coolant System Cooldown Limitations Applicable for 50 EFPY (LCO 3.4.3) 5.2 - 8 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual 2500 2000 8' 1500 ci5 w
w

Pressure and Temperature Limits Report 5.2
/
V V V
If 1000
----
500 0
50 60 70 80 90 100 TEMPERATURE (°F)
Figure 5.2-3 (Page 1 of 1) 110 120 Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)
Beaver Valley Unit 1 5.2 - 9 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 1 of 2)
Heatup Curve Data Points for 50 EFPY (LCO 3.4.3) 60°F/hr Heatup T (°F)
P (psig) 60 0
60 602 65 602 70 602 75 602 80 602 85 602 90 602 95 602 100 602 105 602 110 603 115 604 120 606 125 609 130 612 135 616 140 621 145 627 150 633 155 640 160 648 165 657 170 667 175 678 180 691 185 704 190 719 195 736 200 755 205 775 210 798 215 823 220 851 225 882 Beaver Valley Unit 1 60°F/hr 100°F/hr Heatup Criticality T (°F)
P (psig)
T (°F)
P (psig) 301 0
60 0
301 1190 60 552 305 1241 65 552 310 1303 70 552 315 1358 75 552 320 1417 80 552 325 1483 85 552 330 1555 90 552 335 1636 95 552 340 1724 100 552 345 1821 105 552 350 1929 110 552 355 2048 115 552 360 2179 120 552 365 2324 125 552 370 2483 130 552 135 552 140 553 145 555 150 557 155 561 160 565 165 570 170 575 175 582 180 590 185 598 190 608 195 619 200 631 205 645 210 660 215 677 220 696 225 717 5.2 - 10 100°F/hr Criticality T (°F) 301 301 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 P (psig) 0 947 990 1042 1099 1162 1232 1310 1395 1488 1592 1706 1832 1971 2124 2292 2464 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 2 of 2)
Heatup Curve Data Points for 50 EFPY (LCO 3.4.3) 60°F/hr Heatup 60°F/hr 100°F/hr Heatup 100°F/hr Criticalitv Criticality T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 230 915 230 741 235 953 235 766 240 994 240 795 245 1040 245 827 250 1085 250 861 255 1132 255 900 260 1184 260 943 265 1241 265 990 270 1303 270 1042 275 1358 275 1099 280 1417 280 1162 285 1483 285 1232 290 1555 290 1310 295 1636 295 1395 300 1724 300 1488 305 1821 305 1592 310 1929 310 1706 315 2048 315 1832 320 2179 320 1971 325 2324 325 2124 330 2483 330 2292 335 2464 Leak Test Limit T (°F)
P (psig) 283 301 Beaver Valley Unit 1 2000 2485 5.2 - 11 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 1 of 2)
Cooldown Curve Data Points for 50 EFPY (LCO 3.4.3)
Steady State 20°F/hr 40°F/hr T
p T
p T
p (OF)
(psig)
(OF)
(psig)
(OF)
(psig) 60 0
60 0
60 0
60 621 60 607 60 563 65 621 65 608 65 564 70 621 70 609 70 565 75 621 75 610 75 566 80 621 80 611 80 567 85 621 85 613 85 569 90 621 90 614 90 570 95 621 95 616 95 572 100 621 100 618 100 574 105 621 105 621 105 576 110 621 110 621 110 579 115 621 115 621 115 582 120 621 120 621 120 585 125 621 125 621 125 589 130 621 130 621 130 593 130 680 130 637 135 598 135 684 135 641 140 603 140 689 140 646 145 609 145 694 145 652 150 615 150 700 150 658 155 623 155 706 155 665 160 630 160 713 160 672 165 639 165 721 165 680 170 649 170 729 170 689 175 660 175 739 175 700 180 672 180 749 180 711 185 685 185 761 185 723 190 700 190 774 190 737 195 717 195 788 195 752 200 735 200 803 200 769 205 755 205 821 205 788 210 778 210 840 210 808 215 802 215 861 215 831 220 830 220 884 220 856 225 860 225 910 225 884 230 894 230 938 230 915 235 931 235 970 235 949 240 973 240 1004 240 987 245 1018 245 1043 245 1029 250 1069 250 1085 250 1075 255 1125 Beaver Valley Unit 1 5.2 - 12 60°F/hr T
p (OF)
(psig) 60 0
60 518 65 519 70 520 75 521 80 522 85 523 90 525 95 527 100 529 105 531 110 534 115 537 120 541 125 545 130 549 135 554 140 559 145 566 150 572 155 580 160 588 165 598 170 609 175 620 180 633 185 648 190 664 195 682 200 702 205 724 210 748 215 775 220 805 225 838 230 875 235 916 240 961 245 1011 250 1067 255 1125 100°F/hr T
p (OF)
(psig) 60 0
60 426 65 426 70 427 75 428 80 429 85 431 90 432 95 434 100 436 105 439 110 442 115 445 120 449 125 453 130 458 135 464 140 470 145 477 150 485 155 494 160 504 165 515 170 527 175 541 180 556 185 573 190 593 195 614 200 637 205 664 210 693 215 725 220 761 225 801 230 846 235 895 240 949 245 1010 250 1067 255 1125 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 2 of 2)
Cooldown Curve Data Points for 50 EFPY (LCO 3.4.3)
Steady State 20°F/hr T
p T
p (OF)
(psig)
(OF)
(psig) 255 1132 255 1127 260 1184 260 1183 265 1241 265 1241 270 1305 270 1305 275 1375 275 1375 280 1452 280 1452 285 1537 285 1537 290 1632 290 1632 295 1736 295 1736 300 1851 300 1851 305 1979 305 1979 310 2120 310 2120 315 2275 315 2275 320 2448 320 2448 Beaver Valley Unit 1 40°F/hr 60°F/hr T
p T
p (OF)
(psig)
(OF)
(psig) 260 1183 260 1183 265 1241 265 1241 270 1305 270 1305 275 1375 275 1375 280 1452 280 1452 285 1537 285 1537 290 1632 290 1632 295 1736 295 1736 300 1851 300 1851 305 1979 305 1979 310 2120 310 2120 315 2275 315 2275 320 2448 320 2448 5.2 - 13 100°F/hr T
p (OF)
(psig) 260 1183 265 1241 270 1305 275 1375 280 1452 285 1537 290 1632 295 1736 300 1851 305 1979 310 2120 315 2275 320 2448 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-3 (Page 1 of 1)
Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12)
FUNCTION OPPS Enable Temperature PORV Setpoint Beaver Valley Unit 1 5.2 - 14 SETPOINT 347°F s 397 psig PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4 (Page 1 of 1)
Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f(a)
FF<b)
ART NDic)
FF *ARTNDT FF2 Lower Shell V
0.297 0.6677 127.9 85.40 0.446 Plate u
0.618 0.8652 118.3 102.35 0.749 B6903-1(d) w 0.952 0.9862 147.7 145.66 0.973 (Longitudinal) y 2.10 1.2018 141.7 170.30 1.444 X
4.99 1.4020 175.8 246.46 1.965 Lower Shell V
0.297 0.6677 138.0 92.14 0.446 Plate u
0.618 0.8652 132.1 114.29 0.749 B6903-1<d) w 0.952 0.9862 180.2 177.72 0.973 (Transverse) y 2.10 1.2018 166.9 200.58 1.444 X
4.99 1.4020 179.0 250.95 1.965 SUM:
1585.86 11.154 CF= L(FF
ART Nor) + L(FF2) = (1585.86) + (11.154) = 142.2°F(e)
V 169.4 0.297 0.6677 (159.8) 113.10 0.446 Beaver Valley u
174.8 0.618 0.8652 (164.9) 151.23 0.749 Unit 1 197.5 SuNeillance w
0.952 0.9862 (186.3) 194.76 0.973 Weld Metal<d)
(Heat # 305424) y 189.2 2.10 1.2018 (178.5) 227.40 1.444 X
252.1 4.99 1.4020 (237.8) 353.39 1.965 SUM:
1039.87 5.577 CF = L(FF
ART NDT) + L(FF2) = (1039.87) + (5.577) = 186.5°F(e)
Notes:
(a) f = Calculated surveillance capsule neutron fluence (x 1019 n/cm2, E > 1.0 MeV). The suNeillance capsule fluence results are contained in Table 4-1 of Reference 2.
(b)
FF = fluence factor = f <0 0-1
109 f)_
(c)
ART Nor values are the measured 30 ft-lb shift values. The Beaver Valley Unit 1 ART Nor values for the suNeillance weld data are adjusted by a ratio of 1.06. Pre-adjusted values are listed in parentheses, and were taken from Table 4-1 of Reference 2.
NOTE:
Per Regulatory Guide 1.99, Revision 2 (Reference 7), section 2.1 "Radiation Embrittlement of Reactor Vessel Materials," the vessel weld chemistry factor is divided by the surveillance weld chemistry factor to obtain a ratio factor to multiply the ART Nor values by to obtain adjusted ART Nor values. In Table 5-2 of Reference 2, the ratio is determined to be 1.06 or (191.7/181.6).
(d)
The plate and weld suNeillance data is deemed non-credible per Appendix D of Reference 2.
(e)
Position 2.1 chemistry factor values are summarized in Table 5-4 of Reference 2.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 15 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4a (Page 1 of 2)
Calculation of Chemistry Factors(a)
(Based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data)
Material Capsule Capsule f(b)
FF(c) dRT NDid)
FF *dRTNDT FF2 97° 0.5174 0.8160 82.6 67.44 0.666 (72.34)
Weld Metal 81.1 Heat # 90136(e>
104° 0.7885 0.9333 (67.4) 75.68 0.871 (St. Lucie Unit 1) 83.8 284° 1.243 1.0606 (68.0) 88.85 1.125 Weld Metal 67.5 Heat # 90136(e) 97° 0.324 0.6902 (65.93) 46.61 0.476 (Millstone Unit 2) 57.0 104° 0.949 0.9853 (52.12) 56.18 0.971 61.4 83° 1.74 1.1523 (56.09) 70.74 1.328 SUM:
405.50 5.437 CF = L(FF
dRT NDT) + L(FF2) = (405.50) + (5.437) = 74.6°F<9>
W-225 0.488 0.800 197.30 157.83 0.640 Weld Metal (210)
Heat# 305414(f)
W-265 0.847 0.953 218.30 208.13 0.909 (Fort Calhoun (225)
Unit 1)
W-275 1.54 1.119 215.90 241.68 1.253 (219)
SUM:
607.64 2.802 CF= L(FF
dRT NOT)+ L(FF2) = (607.64) + (2.802) = 216.9°F(9)
Notes for Table 5.2-4a are on the following page.
Beaver Valley Unit 1 5.2 - 16 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4a (Page 2 of 2)
Calculation of Chemistry Factors(a)
(Based on St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Capsule Data)
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 )." As a result of the unclear identification of the Millstone Unit 2 surveillance weld heat number, the Millstone Unit 2 data was not originally incorporated into Beaver Valley Unit 1 chemistry factor calculations. Since the Millstone Unit 2 surveillance weld contains specimens made of Heat# 90136, the use of this data is appropriate. See Appendix D of Reference 2 for more details.
(b) f = calculated surveillance capsule fluence values (x 1019 n/cm2, E > 1.0 MeV). The surveillance capsule fluence results for St. Lucie Unit 1 and Millstone Unit 2 are contained in Table 4-2 of Reference 2. The surveillance capsule fluence results for Fort Calhoun Unit 1 are contained in Table D-5 of Reference 3.
( c)
FF = fluence factor= f (0 0-1
109 f)_
( d) fRT Nor values are the measured 30 ft-lb. shift values. gRT Nor values for the surveillance weld data are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the beltline weld chemistry. Pre-adjusted values are listed in parentheses, and were taken from Tables 4-2 of Reference 2 and Table A-5 of Reference 9. The temperature adjustments for each capsule were calculated from the data in Table 5.2-4b and the average plant irradiation temperature for BV-1. The St. Lucie Unit 1 gRT Nor values for the weld data are adjusted by a ratio of 1.17. The Millstone Unit 2 and Fort Calhoun fRT NDT values were not adjusted since the ratio was less than 1.00; therefore, a conservative value of 1.00 was used.
(e)
The St. Lucie Unit 1 and Millstone Unit 2 surveillance data is deemed credible per Appendix D of Reference 2; however, a full margin term should be utilized for conservatism when this data is applied as a result of the unclear identification of the Millstone Unit 2 weld specimen heat numbers. See Appendix D of Reference 2 for more details.
(f)
The Fort Calhoun Unit 1 surveillance data is deemed non-credible per Appendix D of Reference 3.
(g)
Position 2.1 chemistry factor values are summarized in Table 5-4 of Reference 2.
Beaver Valley Unit 1 5.2 - 17 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4b (Page 1 of 1)
St. Lucie Unit 1, Millstone Unit 2, and Fort Calhoun Surveillance Weld Data(a)(b)
Cu Ni Irradiated Capsule f(d)
RT NDie)
Material Capsule (wt.%)
(wt.%)
Temperature(c)
(x1019 n/cm2, (OF)
(OF)
E>1.0 MeV)
Weld Metal 97° 0.23 0.07 541 0.5174 72.34 Heat# 90136 104° 0.23 0.07 544.6 0.7885 67.4 (St. Lucie Unit 1) 284° 0.23 0.07 546.3 1.243 68.0 Weld Metal 97° 0.30 0.06 544.3 0.324 65.93 Heat# 90136 104° 0.30 0.06 547.6 0.949 52.12 (Millstone Unit 2) 83° 0.30 0.06 548.0 1.74 56.09 Weld Metal W-225 0.35 0.60 530 0.488 210 Heat W-265 0.35 0.60 536 0.847 225
305414 (Fort Calhoun W-275 0.35 0.60 539.6 1.54 219 Unit 1)
(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 )." As a result of the unclear identification of the Millstone Unit 2 surveillance weld heat number, the Millstone Unit 2 data was not originally incorporated into Beaver Valley Unit 1 chemistry factor calculations. Since the Millstone Unit 2 surveillance weld contains specimens made of Heat# 90136, the use of this data is appropriate. See Appendix D of Reference 2 for more details.
(b)
Data contained in this table was obtained from Reference 2, unless otherwise noted.
(c)
Irradiated temperatures are the average inlet temperatures over the specific cycles corresponding to the operating time experienced by each of the respective capsules.
(d) f = calculated surveillance capsule fluence values.
(e)
_RT NDT values are the measured 30 ft-lb shift values from Table 4-2 of Reference 2 and Table D-5 of Reference 3.
Beaver Valley Unit 1 5.2-18 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-5 (Page 1 of 1)
Reactor Vessel Beltline Material Properties Cu Ni Position 1. 1 Initial Material Description (wt.%)
(wt.%)
Chemistry RT NDT(a)
Factor (OF)
(OF)
Intermediate Shell Plate 86607-1 0.14 0.62 100.5 26.8 Intermediate Shell Plate 86607-2 0.14 0.62 100.5 53.6 Lower Shell Plate 86903-1 0.21 0.54 147.2 13.1 Lower Shell Plate 87203-2 0.14 0.57 98.7 0.4 Intermediate to Lower Shell Weld 0.27 0.07 124.3
-56 Seam (Heat# 90136)11-714 Intermediate Longitudinal Shell Weld 0.28 0.63 191.7
-56 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 RT NDT values for the plates are based on measured data while the weld values are generic.
Beaver Valley Unit 1 5.2-19 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 1 of 2)
Reactor Vessel Extended Beltline Material Properties(a)
Material Heat Number Wt%
Wt%
Material Description ID (Lot Number)
Cu Ni Upper Shell Forging 86604 123V339VA1 0.12(b) 0.68 305414 0.34 0.61 (3951 & 3958)
Upper to Intermediate 10-714 AOFJ 0.03 0.93 Shell Girth Weld FOIJ 0.03 0.94 EODJ 0.02 1.04 HOCJ 0.02 0.93 86608-1 95443-1 0.10 0.82 Inlet Nozzles 86608-2 95460.:.1 0.10 0.82 86608-3 95712-1 0.08 0.79 EODJ 0.02 1.04 FOIJ 0.03 0.94 1-7178 HOCJ 0.02 0.93 Inlet Nozzle Welds 1-7170 O81J 0.02 0.97 1-717F EOEJ 0.01 1.03 ICJJ 0.03 0.99 JACJ 0.04 0.97 86605-1 95415-1 0.13(d) 0.77 Outlet Nozzles 86605-2 95415-2 0.13(d}
0.77 86605-3 95444-1 0.09 0.79 ICJJ 0.03 0.99 IOBJ 0.02 0.97 1-717A JACJ 0.04 0.97 Outlet Nozzle Welds 1-717C HOCJ 0.02 0.93 1-717E EODJ 0.02 1.04 FOIJ 0.03 0.94 Notes for Table 5.2-6 are on the following page.
Beaver Valley Unit 1 5.2 - 20 Initial RT NDic)
(OF) 40
-56 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 48.5
-15.2 11.4 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen)
-26.2 3.3 10.1 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen) 10 (Gen)
PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 2 of 2)
Reactor Vessel Extended Beltline Material Properties(a)
Notes:
(a) Data obtained from Table 3-2 of Reference 2.
(b) The Cu wt % was not available from the CMTR so in accordance with Regulatory Guide 1.99, Revision 2, a standard deviation analysis (average + standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Shell Forgings (55 data points).
(c) The initial RT NOT value for the upper shell forging, inlet nozzle forgings, and outlet nozzle forgings are based on measured values. The generic initial RT NOT values for the weld materials were determined in accordance with NUREG-0800 [Reference 11] and 1 0 CFR 50.61 [Reference 6].
(d) The Cu wt% was not available from the CMTR, so in accordance with Regulatory Guide 1.99, Revision 2, a standard deviation analysis (average + standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Nozzle Forgings
( 178 data points).
Beaver Valley Unit 1 5.2 - 21 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-7 (Page 1 of 1)
Summary of Adjusted Reference Temperatures (ARTs) for 50 EFPY(e) 50 EFPY Material Description 1/4T ART(a>
3/4T ART(a)
(OF)
Intermediate Shell Plate 86607-1 195.2 Intermediate Shell Plate 86607-2 222.0 Lower Shell Plate 87203-2 166.4 Lower Shell Plate 86903-1 244.0(f)
- Using S/C Data(b) 237.3 Intermediate Shell Longitudinal Weld 19-714A/B 182.4
- Using S/C Data(b) 177.7 Intermediate to Lower Shell Circ. Weld 11-714 175.7
- Using S/C Data (c) 109.3 Lower Shell Longitudinal Weld 20-714A/B 199.9
- Using S/C Data(d) 205.6 Upper Shell Forging 86604 139.4 Upper Shell to Intermediate Shell Girth Weld 10-714 172.9 (Heat # 305414)
-Using S/C Data(d) 177.9 Upper Shell to Intermediate Shell Girth Weld 10-714 88.4 (Heat #'s AOFJ and FOIJ)
Upper Shell to Intermediate Shell Girth Weld 10-714 44.0 (Heat #'s EODJ and HOCJ)
Inlet and Outlet Nozzle Welds (All Heat #'s) 44.0 Notes:
(a) ART= I+ fRTNoT + M.
(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full crt1.)
(OF) 171.2 198.0 142.8 208.8(t>
203.3 133.5 130.2 146.0 91.4 146.2 150.3 119.2 122.5 125.9 44.0 44.0 44.0 (c) Based on St. Lucie Unit 1 and Millstone Unit 2 surveillance data. (Data credible.
ART calculated with a full crt1 per Appendix D of Reference 2.)
(d) Based on Fort Calhoun Unit 1 surveillance data. (Data not credible. ART calculated with a full crt1.)
(e) Data obtained from Tables 7-2 and 7-3 of Reference 2. Nozzle ART values are excluded from this table, as these values are calculated using surface fluence values. See Reference 2 for nozzle ART values.
(f) For the purposes of PIT limit curve development, a 1/4T ART value of 244.5°F and a 3/4T ART value of 209.5°F were used for conservatism.
Beaver Valley Unit 1 5.2 - 22 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-8 (Page 1 of 1)
Calculation of Adjusted Reference Temperatures (ARTs) for 50 EFPY<c)
Parameter VALUES Operating Time 50 EFPY Material Plate 86903-1 Location Lower Shell Plate 1/4T ART(°F)
Chemistry Factor, CF (°F) 147.2 Fluence (f), n/cm2 (E>1.0 Mev) 3.672 X 1019 Fluence Factor, FF 1.3374 LiRT NDT = CF X FF(°F) 196.9(b)
Initial RT NDT, 1(°F)(a) 13.1 Margin, M(°F) 34(b)
ART = l+(CF*FF)+M, °F per RG 1.99, Revision 2 244.0(d)
Notes:
(a)
Initial RT Nor values are measured values for plate material.
(b)
Based on Regulatory Guide 1.99, Revision 2 Position 1.1.
Plate 86903-1 Lower Shell Plate 3/4T ART(°F) 147.2 1.427 X 1019 1.0987 161.7(b) 13.1 34(b) 208.8(d)
(Surveillance data not credible. ART calculated with a full crL1.)
(c)
Data obtained from Tables 7-2 and 7-3 of Reference 2.
(d)
For the purposes of PIT limit curve development, a 1/4T ART value of 244.5°F and a 3/4T ART value of 209.5°F were used for conservatism.
Beaver Valley Unit 1 5.2 - 23 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Table 5.2-9 (Page 1 of 2)
Pressure and Temperature Limits Report 5.2 RT PTS Calculation for Beltline Region Materials at Life Extension (50 EFPY)<a>
Material Heat Surface Fluence Chemistry Material Description Intermediate Shell Plate Intermediate Shell Plate Lower Shell Plate ID 86607-1 86607-2 86903-1 Number c Using non-credible surveillance data<9>
Lower Shell Plate Intermediate to Lower Shell Girth Weld 87203-2 11-714 90136 c Using credible surveillance data<h>
Intermediate Shell 19-714 305424 Longitudinal Weld A&8 c Using non-credible surveillance data<9>
Lower Shell Longitudinal 20-714 305414 Weld A&8 c Using non-credible surveillance data(i>
Notes:
(a) Data obtained from Table 6-3 of Reference 9.
(b)
FF = fluence factor= f <0-28-0-10109 <t>>_
Fluence
Factor, (x1019 n/cm2)
FF
5.57 1.4231 5.57 1.4231 5.57 1.4231 5.57 1.4231 5.57 1.4231 5.55 1.4225 5.55 1.4225 1.08 1.0224 1.08 1.0224 1.09 1.0241 1.09 1.0241 (c)
Initial RT NDT values are measured values with the exception of the vessel welds.
(d) bRT PTS = CF* FF.
(e) M = 2 *(cru2 + crl)112.
(f)
RT PTs = Initial RT NDT + bRT PTs + Margin.
Beaver Valley Unit 1 5.2 - 24 Factor
(°F) 100.5 100.5 147.2 151.8 98.7 124.3 87.1 191.7 192.3 210.5 216.9 Initial bRT PTS(d)
RT ND/c)
Ou
(°F)
(OF)
(OF) 43 143.0 0
73 143.0 0
27 209.5 0
27 216.0 0
20 140.5 0
-56 176.8 17
-56 123.9 17
-56 196.0 17
-56 196.6 17
-56 215.6 17
-56 222.1 17 Oi1.
(OF) 17 17 17 17<9>
17 28 14(h) 28 28<9>
28 28(i)
Margin<e>
RT PTS(f)
(OF)
(OF) 34 220.0 34 250.0 34 270.5 34 277.0 34 194.5 65.5 186.3 44.0 111.9 65.5 205.5 65.5 206.1 65.5 225.1 65.5 231.6 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 2 of 2)
RT Prs Calculation for Beltline Region Materials at Life Extension (50 EFPY)<a>
Notes continued:
(g) The BVPS-1 surveillance weld metal is the same weld heat as the BVPS-1 intermediate shell longitudinal welds (heat 305424). The BVPS-1 surveillance weld data is non credible; therefore, the higher Ot:,. term of 28°F was utilized for BVPS-1 weld heat 305424.
The BVPS-1 surveillance plate material is representative of the BVPS-1 lower shell plate B6903-1. The surveillance plate material is non-credible; therefore, the higher Ot:,. term of 17°F was utilized for BVPS-1 plate 86903-1. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
(h) The St. Lucie Unit 1 surveillance weld metal is the same weld heat as the BVPS-1 intermediate to lower shell girth weld (heat 90136). The St. Lucie Unit 1 surveillance weld data is credible; therefore, the reduced Ot:,. term of 14°F was utilized for BVPS-1 weld heat 90136. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
(i)
The Fort Calhoun surveillance weld metal is the same weld heat as the BVPS-1 lower shell longitudinal welds (heat 305414). The Fort Calhoun surveillance weld data is non credible; therefore, the higher Ot:,. term of 28°F was utilized for BVPS-1 weld heat 305414.
The credibility evaluation conclusions are contained in Appendix A of Reference 9.
Beaver Valley Unit 1 5.2 - 25 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Table 5.2-10 (Page 1 of 2)
Pressure and Temperature Limits Report 5.2 RT Prs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)<a)
Material Material Heat Number Surface Fluence Chemistry Fluence
Factor, Factor Description ID (Lot Number)
(x1019 n/cm2)
FF<b)
(°F)
Upper Shell 86604 123V339VA1 0.625 0.8685 84.2 Forging Upper to 305414 Intermediate 10-714 (3951 & 3958) 0.625 0.8685 209.11 Shell Girth Weld C Using non-credible surveillance data<g) 0.625 0.8685 216.9 Upper to AOFJ 0.625 0.8685 41.0 FOIJ 0.625 0.8685 41.0 Intermediate 10-714 EODJ 0.625 0.8685 27.0 Shell Girth Weld HOCJ 0.625 0.8685 27.0 86608-1 95443-1 0.016 0.1513 67.0 Inlet Nozzles 86608-2 95460-1 0.016 0.1513 67.0 86608-3 95712-1 0.016 0.1513 51.0 EODJ 0.016 0.1513 27.0 FOIJ 0.016 0.1513 41.0 Inlet Nozzle 1-717 8 HOCJ 0.016 0.1513 27.0 1-717 D D8IJ 0.016 0.1513 27.0 Welds 1-717 F EOEJ 0.016 0.1513 20.0 ICJJ 0.016 0.1513 41.0 JACJ 0.016 0.1513 54.0 86605-1 95415-1 0.011 0.1191 95.25 Outlet Nozzles 86605-2 95415-2 0.011 0.1191 95.25 86605-3 95444-1 0.011 0.1191 58.0 ICJJ 0.011 0.1191 41.0 1-717 A IO8J 0.011 0.1191 27.0 Outlet Nozzle JACJ 0.011 0.1191 54.0 Welds 1-717 C HOCJ 0.011 0.1191 27.0 1-717 E EODJ 0.011 0.1191 27.0 FOIJ 0.011 0.1191 41.0 Notes:
(a) Data obtained from Table 6-4 of Reference 9.
Beaver Valley Unit 1 5.2 - 26 Initial URTPTS(d)
Ou RT NDT(c)
(°F)
(OF)
(OF) 40 73.1 0
-56 181.6 17
-56 188.4 17 10 35.6 17 10 35.6 17 10 23.4 17 10 23.4 17 60 10.1 17 60 10.1 17 60 7.7 17 10 4.1 17 10 6.2 17 10 4.1 17 10 4.1 17 10 3.0 17 10 6.2 17 10 8.2 17 60 11.3 17 60 11.3 17 60 6.9 17 10 4.9 17 10 3.2 17 10 6.4 17 10 3.2 17 10 3.2 17 10 4.9 17 0
(OF) 17 28 28(9) 17.8 17.8 11.7 11.7 5.1 5.1 3.9 2.0 3.1 2.0 2.0 1.5 3.1 4.1 5.7 5.7 3.5 2.4 1.6 3.2 1.6 1.6 2.4 Margin<e)
(OF) 34 65.5 65.5 49.2 49.2 41.3 41.3 35.5 35.5 34.9 34.2 34.6 34.2 34.2 34.1 34.6 35.0 35.8 35.8 34.7 34.3 34.2 34.6 34.2 34.2 34.3 RT PTS(f)
(OF) 147.1 191.1 197.9 94.8 94.8 74.8 74.8 105.6 105.6 102.6 48.3 50.8 48.3 48.3 47.2 50.8 53.1 107.2 107.2 101.6 49.2 47.4 51.0 47.4 47.4 49.2 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-10 (Page 2 of 2)
RT PTs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)(a)
Notes continued:
(b) FF =fluence factor= f<0-2s-o.10 1og(f))_
(c)
Initial RT NDT value for the upper shell forging is a measured value. All other values are generic.
(d) PRT PTS = CF* FF.
(e) M = 2 *(cru2 + crl)112.
(f)
RT PTs = Initial RT NDT + PRT Prs + Margin.
(g)
The Fort Calhoun surveillance weld metal is the same weld heat as the BVPS-1 upper to intermediate shell girth weld (heat 305414). The Fort Calhoun surveillance weld data is non-credible; therefore, the higher 011 term of 28°F was utilized for BVPS-1 weld heat 305414. The credibility evaluation conclusions are contained in Appendix A of Reference 9.
Beaver Valley Unit 1 5.2 - 27 PTLR Revision 9 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 1 of 1)
Reactor Vessel Toughness Data (Unirradiated)
Cu Ni p
TNDT COMPONENT HEAT NO.
CODE NO.
MATERIAL TYPE
(%)
(%)
(%)
(OF)
Closure Head C6213-1 B B6610 A533B CL. 1
.15
.010
-40 Dome Closure Head A5518-2 B6611 A533B CL. 1
.14
.015
-20 Seq.
Closure Head ZV3758 A508 CL. 2
.08
.007 60*
Flange Vessel Flange ZV-3661 FV-2961 A508 CL. 2
.12
.010
-54.7**
Inlet Nozzle 9-5443-1 B6608-1 A508 CL. 2
.10
.82
.008 35.8**
Inlet Nozzle 9-5460-1 B6608-2 A508 CL. 2
.10
.82
.010
-18.3**
Inlet Nozzle 9-5712-1 B6608-3 A508 CL. 2
.08
.79
.007
-2.5**
Outlet Nozzle 9-5415-1 B6605-1 A508 CL. 2
.13
.77
.008
-26.2**
Outlet Nozzle 9-5415-2 B6605-2 A508 CL. 2
.13
.77
.007 3.0**
Outlet Nozzle 9-5444-1 B6605-3 A508 CL. 2
.09
.79
.007 10.1 **
Upper Shell 123V339VA1 ---
A508 CL. 2
.12
.68
.010 40 Inter Shell C4381-2 B6607-2 A533B CL. 1
.14
.62
.015
-10 Inter Shell C4381-1 B6607-1 A533B CL. 1
.14
.62
.015
-10 Lower Shell C6317-1 B6903-1 A533B CL. 1
.21
.54
.010
-50 Lower Shell C6293-2 B7203-2 A533B CL. 1
.14
.57
.015
-20 Trans Ring 123V223 A508 CL. 2 30 Bottom Hd Sea C4423-3 B6618 A533B CL. 1
.13
.008
-30 Bottom Hd Dome C4482-1 B6619 A533B CL. 1
.13
.015
-50 Inter to Lower 90136
.27
.07 Shell Weld Inter Shell Long.
305424
.28
.63 Weld Lower Shell 305414
.34
.61 Lono. Weld Weld HAZ
-40
Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2
Estimated Per BWRVIP-113-A, Alternate Approach 2 MWD - Major Working Direction NMWD - Normal to Major Working Direction RT Nor (OF) 0*
-20*
60*
1 0**
48.5**
-15.2**
11.4**
-26.2**
3.3**
10.1 **
40*
53.6 26.8 13.1 0.4 30*
-29*
-33*
-56
-56
-56
-40 UPPER SHELF ENERGY (FT-LB)
MWD NMWD 121 131
>100 166 82.5 94 97 93 112.5 103 155 101 123 83 128.5 94 134 83 129.5 85 143 124 125.5
> 100
> 100
> 100 136.5 Note:
For evaluation of lnservice 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.
Beaver Valley Unit 1 5.2 - 28 PTLR Revision 9 LRM Revision 98

L-17-277, Pressure and Temperature Limits Reports, Revisions 8 and 9

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L-17-277, Pressure and Temperature Limits Reports, Revisions 8 and 9

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