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 "

Beaver Valley Power Station P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 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 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 PTLR Revision 8 Beaver Valley Unit 1 5.2- i LRM Revision 97

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2 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 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.

5.2.1.1 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.

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

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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.

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

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 10 19 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).

5.2.1.2 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.

5.2.1.3 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 RTNDT + 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.

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

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 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 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.

5.2.1.4 Reactor Vessel Boltup Temperature {LCO 3.4.3)

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

5.2.2 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 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.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 4 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 RTPrs values for the beltline materials at 50 EFPY.

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

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

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

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

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. LTR-SCS-16-58 Rev. 0, LTOPS 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.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 6 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 joperlim Version:5. Run:1944 Operli .xlsm Verion: 5.4!

n:1 I

--- . r**** I

. L -l I

I 2250 r

2000 1750 1500 C,

"cii 1250 en en CL

.$ 1000 ni u

750 Criticality Limit based on 500 inservice hydrostatic test temperature (301 °F) for the service period up to 50 EFPY 250 0

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

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 7 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 Operlim Version:5.4 Run:19454 Operlim.xlsm Version: 5.4 2250 2000 1750 1500

'in Q) 1250 U)

U)

Q) a.

"C Q) 1000 750 Cooldown Rates

°F/Hr Steady-State 500 20 40 60 100 250 I

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)

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

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 2500 2000 6 1500 /

ci5

/

a:

V V

en en w

1000 i.---

500 0 100 50 60 70 80 90 110 120 TEMPERATURE (°F)

Figure 5.2-3 (Page 1 of 1)

Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)

PTLR Revision 8 Beaver Valley Unit 1 5.2- 9 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 100 ° F/hr G0 ° F/hr Heatup 100 ° F/hr Heatup Criticality Criticality T (° F) P (psig) T (° F) P (psig) T { ° F) P (psig) T (° F) P (psig) 60 0 301 0 60 0 301 0 60 602 301 1190 60 552 301 947 65 602 305 1241 65 552 305 990 70 602 310 1303 70 552 310 1042 75 602 315 1358 75 552 315 1099 80 602 320 1417 80 552 320 1162 85 602 325 1483 85 552 325 1232

90 602 330 1555 90 552 330 1310 95 602 335 1636 95 552 335 1395 100 602 340 1724 100 552 340 1488 105 602 345 1821 105 552 345 1592 110 603 350 1929 110 552 350 1706 115 604 355 2048 115 552 355 1832 120 606 360 2179 120 552 360 1971 125 609 365 2324 125 552 365 2124 130 612 370 2483 130 552 370 2292 135 616 135 552 375 2464 140 621 140 553 145 627 145 555 150 633 150 557 155 640 155 561 160 648 160 565 165 657 165 570 170 667 170 575 175 678 175 582 180 691 180 590 185 704 185 598 190 719 190 608 195 736 195 619 200 755 200 631 205 775 205 645 210 798 210 660 215 823 215 677 220 851 220 696 225 882 225 717 PTLR Revision 8 Beaver Valley Unit 1 5.2 - 10 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 100 ° F/hr 60° F/hr Heatup 100 ° F/hr Heatup 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 2000 301 2485 PTLR Revision 8 Beaver Valley Unit 1 5.2 - 11 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 60° F/hr 100° F/hr T p T p T p T p T p (OF) (psig) {OF) {psig) {OF) (psig) (OF) (psig) {OF) (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 607 60 563 60 518 60 426 65 621 65 608 65 564 65 519 65 426 70 621 70 609 70 565 70 520 70 427 75 621 75 610 75 566 75 521 75 428 80 621 80 611 80 567 80 522 80 429 85 621 85 613 85 569 85 523 85 431 90 621 90 614 90 570 90 525 90 432 95 621 95 616 95 572 95 527 95 434 100 621 100 618 100 574 100 529 100 436 105 621 105 621 105 576 105 531 105 439 110 621 110 621 110 579 110 534 110 442 115 621 115 621 115 582 115 537 115 445 120 621 120 621 120 585 120 541 120 449 125 621 125 621 125 589 125 545 125 453 130 621 130 621 130 593 130 549 130 458 130 680 130 637 135 598 135 554 135 464 135 684 135 641 140 603 140 559 140 470 140 689 140 646 145 609 145 566 145 477 145 694 145 652 150 615 150 572 150 485 150 700 150 658 155 623 155 580 155 494 155 706 155 665 160 630 160 588 160 504 160 713 160 672 165 639 165 598 165 515 165 721 165 680 170 649 170 609 170 527 170 729 170 689 175 660 175 620 175 541 175 739 175 700 180 672 180 633 180 556 180 749 180 711 185 685 185 648 185 573 185 761 185 723 190 700 190 664 190 593 190 774 190 737 195 717 195 682 195 614 195 788 195 752 200 735 200 702 200 637 200 803 200 769 205 755 205 724 205 664 205 821 205 788 210 778 210 748 210 693 210 840 210 808 215 802 215 775 215 725 215 861 215 831 220 830 220 805 220 761 220 884 220 856 225 860 225 838 225 801 225 910 225 884 230 894 230 875 230 846 230 938 230 915 235 931 235 916 235 895 235 970 235 949 240 973 240 961 240 949 240 1004 240 987 245 1018 245 1011 245 1010 245 1043 245 1029 250 1069 250 1067 250 1067 250 1085 250 1075 255 1125 255 1125 255 1125 PTLR Revision 8 Beaver Valley Unit 1 5.2 - 12 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 40° F/hr 60° F/hr 100° F/hr T p T p T p T p T p (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 255 1132 255 1127 260 1183 260 1183 260 1183 260 1184 260 1183 265 1241 265 1241 265 1241 265 1241 265 1241 270 1305 270 1305 270 1305 270 1305 270 1305 275 1375 275 1375 275 1375 275 1375 275 1375 280 1452 280 1452 280 1452 280 1452 280 1452 285 1537 285 1537 285 1537 285 1537 285 1537 290 1632 290 1632 290 1632 290 1632 290 1632 295 1736 295 1736 295 1736 295 1736 295 1736 300 1851 300 1851 300 1851 300 1851 300 1851 305 1979 305 1979 305 1979 305 1979 305 1979 310 2120 310 2120 310 2120 310 2120 310 2120 315 2275 315 2275 315 2275 315 2275 315 2275 320 2448 320 2448 320 2448 320 2448 320 2448 PTLR Revision 8 Beaver Valley Unit 1 5.2 - 13 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 SETPOINT OPPS Enable Temperature 347 ° F PORV Setpoint s 397 psig PTLR Revision 8 Beaver Valley Unit 1 5.2 - 14 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) RTNDic) FF *RTNDT 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

RTNDT) + L(FF ) = (1585.86) + (11.154) = 142.2°F(e) 2 169.4 V 0.297 0.6677 113.10 0.446 (159.8)

Beaver Valley u 0.618 0.8652 174.8 (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) 189.2 y 2.10 1.2018 227.40 1.444 (178.5) 252.1 X 4.99 1.4020 353.39 1.965 (237.8)

SUM: 1039.87 5.577 CF = L{FF

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

Notes:

(a) f = Calculated surveillance capsule neutron fluence (x 10 19 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) RTNoT values are the measured 30 ft-lb shift values. The Beaver Valley Unit 1 RT 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 RTNor values by to obtain adjusted RTNDT 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) RTNDid) FF *RTNDT FF2 82.6 97° 0.5174 0.8160 67.44 0.666 (72.34)

Weld Metal 81.1 Heat # 90135<e) 104° 0.7885 0.9333 75.68 0.871 (67.4)

(St. Lucie Unit 1) 83.8 284° 1.243 1.0606 88.85 1.125 (68.0)

Weld Metal 67.5 97 ° 0.324 0.6902 (65.93) 46.61 0.476 Heat # 90135<e)

(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

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

W-225 0.488 0.800 197.30 157.83 0.640 (210)

Weld Metal 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

RTNor)+ L(FF2)= (607.64)+ (2.802)= 216.9°F<9>

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

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 16 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 10 19 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.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 17 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)

Irradiated Capsule f(d)

Cu Ni RT NDie)

Material Capsule Temperature<c> (x10 19 n/cm2 ,

(wt. %) (wt.%) (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 Nor values are the measured 30 ft-lb shift values from Table 4-2 of Reference 2 and Table D-5 of Reference 3.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 18 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 Position 1. 1 Initial Cu Ni Material Description Chemistry RT Noia)

(wt.%) (wt.%)

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 RTNDT values for the plates are based on measured data while the weld values are generic.

PTLR Revision 8 Beaver Valley Unit 1 5.2-19 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>

Heat Number Initial Material Description Material Wt% Wt%

RT NDic)

ID (Lot Number) Cu Ni (OF)

Upper Shell Forging 86604 123V339VA1 o.12 0.68 40 305414 0.34 0.61 -56 (Gen)

(3951 & 3958)

Upper to Intermediate AOFJ 0.03 0.93 10 (Gen)10-714 Shell Girth Weld FOIJ 0.03 0.94 10 (Gen)

EODJ 0.02 1.04 10 (Gen)

HOCJ 0.02 0.93 10 (Gen)

B6608-1 95443-1 0.10 0.82 48.5 Inlet Nozzles B6608-2 95460-1 0.10 0.82 -15.2 B6608-3 95712-1 0.08 0.79 11.4 EODJ 0.02 1.04 10 (Gen)

FOIJ 0.03 0.94 10 (Gen) 1-717B HOCJ 0.02 0.93 10 (Gen)

Inlet Nozzle Welds 1-717D O8IJ 0.02 0.97 10 (Gen) 1-717F EOEJ 0.01 1.03 10 (Gen)

ICJJ 0.03 0.99 10 (Gen)

JACJ 0.04 0.97 10 (Gen)

B6605-1 95415-1 0.13(d) 0.77 -26.2 Outlet Nozzles 86605-2 95415-2 0.13(d) 0.77 3.3 B6605-3 95444-1 0.09 0.79 10.1 ICJJ 0.03 0.99 10 (Gen)

IOBJ 0.02 0.97 10 (Gen) 1-717A JACJ 0.04 0.97 10 (Gen)

Outlet Nozzle Welds 1-717C HOCJ 0.02 0.93 10 (Gen) 1-717E EODJ 0.02 1.04 10 (Gen)

FOIJ 0.03 0.94 10 (Gen)

Notes for Table 5.2-6 are on the following page.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 20 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 RTNDT value for the upper shell forging, inlet nozzle forgings, and outlet nozzle forgings are based on measured values. The generic initial RTNDT 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).

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 21 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) (OF)

Intermediate Shell Plate 86607-1 195.2 171.2 Intermediate Shell Plate 86607-2 222.0 198.0 Lower Shell Plate 87203-2 166.4 142.8 Lower Shell Plate 86903-1 244.0(f) 208.8(t)

- Using SIC Data(b) 237.3 203.3 Intermediate Shell Longitudinal Weld 19-714A/B 182.4 133.5

- Using SIC Data(b) 177.7 130.2 Intermediate to Lower Shell Circ. Weld 11-714 175.7 146.0

- Using SIC Data (c) 109.3 91.4 Lower Shell Longitudinal Weld 20-714AIB 199.9 146.2

- Using SIC Data(d) 205.6 150.3 Upper Shell Forging 86604 139.4 119.2 Upper Shell to Intermediate Shell Girth Weld 10-714 172.9 122.5 (Heat # 305414)

-Using SIC Data(d) 177.9 125.9 Upper Shell to Intermediate Shell Girth Weld 10-714 88.4 44.0 (Heat #'s AOFJ and FOIJ)

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

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

(a) ART= I+ RTNor+ M.

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

(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.

PTLR Revision 8 Beaver Valley Unit 1 5.2 - 22 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 Plate 86903-1 Location Lower Shell Lower Shell Plate Plate 1/4T ART(°F) 3/4T ART(°F)

Chemistry Factor, CF (°F) 147.2 147.2 Fluence (f), n/cm2 (E>1.0 Mev) 3.672 X 10 19 1.427 X 10 19 Fluence Factor, FF 1.3374 1.0987 RTNOT = CF X FF(° F) 196.9(b} 161.7(b)

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

ART = l+(CF*FF)+M, °F per RG 1.99, Revision 2 244.Q(d} 208.B(d>

Notes:

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

(b) Based on Regulatory Guide 1.99, Revision 2 Position 1.1.

(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.

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

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 1 of 2)

RT PTs Calculation for 8eltline Region Materials at Life Extension (50 EFPY) (a)

Surface Fluence Chemistry Initial Material Heat RT PTS(d) Ou a Margin (e) RT PTS(f)

Material Description Fluence Factor, Factor RT No/c)

ID Number (OF) (OF) (OF) (OF) (OF)

(x1019 n/cm 2) FF ( b) ( OF) ( °F)

Intermediate Shell Plate 86607-1 --- 5.57 1.4231 100.5 43 143.0 0 17 34 220.0 Intermediate Shell Plate 86607-2 --- 5.57 1.4231 100.5 73 143.0 0 17 34 250.0 Lower Shell Plate 86903-1 --- 5.57 1.4231 147.2 27 209.5 0 17 34 270.5

+ Using non-credible surveillance data (9> 5.57 1.4231 151.8 27 216.0 0 17(g) 34 277.0 Lower Shell Plate 87203-2 --- 5.57 1.4231 98.7 20 140.5 0 17 34 194.5 Intermediate to Lower 11-714 90136 5.55 1.4225 124.3 -56 176.8 17 28 65.5 186.3 Shell Girth Weld

+ Using credible surveillance data (h) 5.55 1.4225 87.1 -56 123.9 17 14( h) 44.0 111.9 Intermediate Shell 19-714 305424 1.08 1.0224 191.7 -56 196.0 17 28 65.5 205.5 Longitudinal Weld A&B

+ Using non -credible surveillance data (9 > 1.08 1.0224 192.3 -56 196.6 17 28(g) 65.5 206.1 Lower Shell Longitudinal 20-714 305414 1.09 1.0241 210.5 -56 215.6 17 28 65.5 225.1 Weld A&8

+ Using non-credible surveillance data(i) 1.09 1.0241 216.9 -56 222.1 17 28(i) 65.5 231.6 Notes:

(a) Data obtained from Table 6-3 of Reference 9.
(b) FF = fluence factor= f (0-28-0* 10109 (t))_
(c) Initial RT NOT values are measured values with the exception of the vessel welds .
(d) RT PTS = CF* FF.
(e) M= 2*(cru 2 + crl) 112*
(f) RT PTs= Initial RT NOT+ RT PTs + Margin.
PTLR Revision 8 Beaver Valley Unit 1 5.2 - 24 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 2 of 2)
RTPTS 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.
PTLR Revision 8 Beaver Valley Unit 1 5.2 - 25 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 T able 5.2-10 (Page 1 of 2)
RT Prs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)(a)
Surface Fluence Chemistry Initial Material Material Heat Number RT PTS (d) Ou 0,1 Margin(e) RT prs(f)
Fluence Factor, Factor RT NDic)
Description ID (Lot Number) (OF) (OF) (OF) (OF) (OF)
(x10 19 n/cm2) FF(b) (O F) O
( F)
Upper Shell 73.1 17 34 147.1 86604 123V339VA1 0.625 0.8685 84.2 40 0 Forging Upper to 305414 17 65.5 191.1 Intermediate 10-714 0.625 0.8685 209.11 -56 181.6 28 (3951 & 3958)
Shell Girth Weld
--+ Using non-credible surveillance data<g> 0.625 0.8685 216.9 -56 188.4 17 28(9) 65.5 197.9 AOFJ 0.625 0.8685 41.0 10 35.6 17 17.8 49.2 94.8 Upper to 41.0 10 35.6 17 17.8 49.2 94.8 FOIJ 0.625 0.8685 Intermediate 10-714 EODJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 Shell Girth Weld HOCJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 86608-1 95443-1 0.016 0.1513 67.0 60 10.1 17 5.1 35.5 105.6 Inlet Nozzles 86608-2 95460-1 0.016 0.1513 67.0 60 10.1 17 5.1 35.5 105.6 86608-3 95712-1 0.016 0.1513 51.0 60 7.7 17 3.9 34.9 102.6 EODJ 0.016 0.1513 27.0 10 4.1 17 2.0 34.2 48.3 FOIJ 0.016 0.1513 41.0 10 6.2 17 3.1 34.6 50.8 1-717 8 HOCJ 0.016 0.1513 27.0 10 4.1 17 2.0 34.2 48.3 Inlet Nozzle 4.1 17 2.0 34.2 48.3 1-717 D D8IJ 0.016 0.1513 27.0 10 Welds 1-717 F EOEJ 0.016 0.1513 20.0 10 3.0 17 1.5 34.1 47.2 ICJJ 0.016 0.1513 41.0 10 6.2 17 3.1 34.6 50.8 JACJ 0.016 0.1513 54.0 10 8.2 17 4.1 35.0 53.1 86605-1 95415-1 0.011 0.1191 95.25 60 11.3 17 5.7 35.8 107.2 Outlet Nozzles 86605-2 95415-2 0.011 0.1191 95.25 60 11.3 17 5.7 35.8 107.2 86605-3 95444-1 0.011 0.1191 58.0 60 6.9 17 3.5 34.7 101.6 ICJJ 0.011 0.1191 41.0 10 4.9 17 2.4 34.3 49.2 IO8J 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 1-717 A Outlet Nozzle JACJ 0.011 0.1191 54.0 10 6.4 17 3.2 34.6 51.0 1-717 C Welds HOCJ 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 1-717 E EODJ 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 FOIJ 0.011 0.1191 41.0 10 4.9 17 2.4 34.3 49.2 Notes:
(a) Data obtained from Table 6-4 of Reference 9.
PTLR Revision 8 Beaver Valley Unit 1 5.2 - 26 LRM Revision 97
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 0- 1 0109 (t))_
(c) Initial RTNOT value for the upper shell forging is a measured value. All other values are generic.
(d) RT PTS = CF* FF.
(e) M= 2*(cru 2 + crl) 112 .
(f) RTPTs= Initial RTNOT+ RTPTs + 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.
PTLR Revision 8 Beaver Valley Unit 1 5.2 - 27 LRM Revision 97
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table5.2-11 (Page 1 of1)
Reactor Vessel Toughness Data (Unirradiated)
UPPER SHELF ENERGY (FT-LB)
Cu Ni p TNDT RTNDT COMPONENT HEAT NO. CODE NO. MATERIAL TYPE
(%) (%) (%) (OF) (OF) MWD NMWD Closure Head C6213-1B 86610 A533B CL. 1 .15 --- .010 -40 0* 121 ---
Dome Closure Head A5518-2 86611 A533B CL. 1 .14 --- .015 -20 -20* 131 ---
Seg.
Closure Head ZV3758 --- A508 CL. 2 .08 --- .007 60* 60* >100 ---
FlanQe Vessel Flam:ie ZV-3661 FV-2961 A508 CL. 2 .12 --- .010 -54.7** 10** 166 ---
Inlet Nozzle 9-5443-1 86608-1 A508 CL. 2 .10 .82 .008 35.8** 48.5** 82.5 ---
Inlet Nozzle 9-5460-1 86608-2 A508 CL. 2 .10 .82 .010 -18.3** -15.2** 94 ---
Inlet Nozzle 9-5712-1 86608-3 A508 CL. 2 .08 .79 .007 -2.5** 11.4** 97 ---
Outlet Nozzle 9-5415-1 B6605-1 A508 CL. 2 .13 .77 .008 -26.2** -26.2** 93 ---
Outlet Nozzle 9-5415-2 86605-2 A508 CL. 2 .13 .77 .007 3.0** 3.3** 112.5 ---
Outlet Nozzle 9-5444-1 86605-3 A508 CL. 2 .09 .79 .007 10.1** 10.1** 103 ---
Upper Shell 123V339VA1 --- A508 CL. 2 .12 .68 .010 40 40* 155 101 Inter Shell C4381-2 86607-2 A5338 CL. 1 .14 .62 .015 -10 53.6 123 83 Inter Shell C4381-1 86607-1 A5338 CL. 1 .14 .62 .015 -10 26.8 128.5 94 Lower Shell C6317-1 86903-1 A5338 CL. 1 .21 .54 .010 -50 13.1 134 83 Lower Shell C6293-2 87203-2 A5338 CL. 1 .14 .57 .015 -20 0.4 129.5 85 Trans Ring 123V223 --- A508 CL. 2 --- --- --- 30 30* 143 ---
Bottom Hd Seg C4423-3 86618 A5338 CL. 1 .13 --- .008 -30 -29* 124 ---
Bottom Hd Dome C4482-1 86619 A5338 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 305414 --- --- .34 .61 --- --- -56 --- > 100 Long. Weld Weld HAZ --- --- --- -40 -40 --- 136.5
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 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.
PTLR Revision 8 Beaver Valley Unit 1 5.2 - 28 LRM Revision 97
Enclosure 8 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 PTLR Revision 9 Beaver Valley Unit 1 5.2 - i LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2 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 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.
5.2.1.1 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 1 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2- 2 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 10 19 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).
5.2.1.2 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.
5.2.1.3 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 RTNDT + 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2- 3 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 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 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.
5.2.1.4 Reactor Vessel Boltup Temperature (LCO 3.4.3)
The minimum boltup temperature for the Reactor Vessel Flange shall be 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.
5.2.2 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 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, RTNOT, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RTNOT 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 4 LRM Revision 98
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 RTPTs values for the beltline materials at 50 EFPY.
Table 5.2-10, taken from Reference 9, provides RTPTs values for the extended beltline materials at 50 EFPY.
Table 5.2-11, provides Reactor Vessel Toughness Data (Unirradiated)
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 5 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.4 References
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, LTOPS 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 6 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 Operiim Version:5.4 Run:19454 Operlim.xlsm Version: 5.4 2250 2000 1750 1500 en 1250 U)
U)
.$ 1000 ni
(.)
750 Criticality Limit based on 500 inservice hydrostatic test temperature (301 °F) for the service period up to 50 EFPY 250 0
0 50 100 150 200 250 300 350 400 450 500 550 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)
PTLR Revision 9 Beaver Valley Unit 1 5.2- 7 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 (1) 1000 750 Cooldown Rates
°F/Hr Steady-State 500 20 40 60 100 250 0 --+--,--r-r-,--+-+---,-,--,-+-,--,--,.-,-+-,--,--,-,--,i--,--,----,-,--+-r-,--,--,.---t-,--,--,--,--+-,r-r--r---,--+--,--,-,.......--+-,----,--,--,-+-,--,-,---,--1 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)
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 8 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 2500 2000
/
V 8' 1500 ci5 w
V w
If 1000 ----
V 500 0 100 50 60 70 80 90 110 120 TEMPERATURE (°F)
Figure 5.2-3 (Page 1 of 1)
Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 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 100° F/hr 60 ° F/hr Heatup 100 ° F/hr Heatup Criticality Criticality T (° F) P (psig) T (° F) P (psig) T (° F) P (psig) T (° F) P (psig) 60 0 301 0 60 0 301 0 60 602 301 1190 60 552 301 947 65 602 305 1241 65 552 305 990 70 602 310 1303 70 552 310 1042 75 602 315 1358 75 552 315 1099 80 602 320 1417 80 552 320 1162 85 602 325 1483 85 552 325 1232 90 602 330 1555 90 552 330 1310 95 602 335 1636 95 552 335 1395 100 602 340 1724 100 552 340 1488 105 602 345 1821 105 552 345 1592 110 603 350 1929 110 552 350 1706 115 604 355 2048 115 552 355 1832 120 606 360 2179 120 552 360 1971 125 609 365 2324 125 552 365 2124 130 612 370 2483 130 552 370 2292 135 616 135 552 375 2464 140 621 140 553 145 627 145 555 150 633 150 557 155 640 155 561 160 648 160 565 165 657 165 570 170 667 170 575 175 678 175 582 180 691 180 590 185 704 185 598 190 719 190 608 195 736 195 619 200 755 200 631 205 775 205 645 210 798 210 660 215 823 215 677 220 851 220 696 225 882 225 717 PTLR Revision 9 Beaver Valley Unit 1 5.2 - 10 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 100 ° F/hr 60° F/hr Heatup 100° F/hr Heatup 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 2000 301 2485 PTLR Revision 9 Beaver Valley Unit 1 5.2 - 11 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 60° F/hr 100° F/hr T p T p T p T p T p (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 607 60 563 60 518 60 426 65 621 65 608 65 564 65 519 65 426 70 621 70 609 70 565 70 520 70 427 75 621 75 610 75 566 75 521 75 428 80 621 80 611 80 567 80 522 80 429 85 621 85 613 85 569 85 523 85 431 90 621 90 614 90 570 90 525 90 432 95 621 95 616 95 572 95 527 95 434 100 621 100 618 100 574 100 529 100 436 105 621 105 621 105 576 105 531 105 439 110 621 110 621 110 579 110 534 110 442 115 621 115 621 115 582 115 537 115 445 120 621 120 621 120 585 120 541 120 449 125 621 125 621 125 589 125 545 125 453 130 621 130 621 130 593 130 549 130 458 130 680 130 637 135 598 135 554 135 464 135 684 135 641 140 603 140 559 140 470 140 689 140 646 145 609 145 566 145 477 145 694 145 652 150 615 150 572 150 485 150 700 150 658 155 623 155 580 155 494 155 706 155 665 160 630 160 588 160 504 160 713 160 672 165 639 165 598 165 515 165 721 165 680 170 649 170 609 170 527 170 729 170 689 175 660 175 620 175 541 175 739 175 700 180 672 180 633 180 556 180 749 180 711 185 685 185 648 185 573 185 761 185 723 190 700 190 664 190 593 190 774 190 737 195 717 195 682 195 614 195 788 195 752 200 735 200 702 200 637 200 803 200 769 205 755 205 724 205 664 205 821 205 788 210 778 210 748 210 693 210 840 210 808 215 802 215 775 215 725 215 861 215 831 220 830 220 805 220 761 220 884 220 856 225 860 225 838 225 801 225 910 225 884 230 894 230 875 230 846 230 938 230 915 235 931 235 916 235 895 235 970 235 949 240 973 240 961 240 949 240 1004 240 987 245 1018 245 1011 245 1010 245 1043 245 1029 250 1069 250 1067 250 1067 250 1085 250 1075 255 1125 255 1125 255 1125 PTLR Revision 9 Beaver Valley Unit 1 5.2 - 12 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 40° F/hr 60° F/hr 100° F/hr T p T p T p T p T p (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 255 1132 255 1127 260 1183 260 1183 260 1183 260 1184 260 1183 265 1241 265 1241 265 1241 265 1241 265 1241 270 1305 270 1305 270 1305 270 1305 270 1305 275 1375 275 1375 275 1375 275 1375 275 1375 280 1452 280 1452 280 1452 280 1452 280 1452 285 1537 285 1537 285 1537 285 1537 285 1537 290 1632 290 1632 290 1632 290 1632 290 1632 295 1736 295 1736 295 1736 295 1736 295 1736 300 1851 300 1851 300 1851 300 1851 300 1851 305 1979 305 1979 305 1979 305 1979 305 1979 310 2120 310 2120 310 2120 310 2120 310 2120 315 2275 315 2275 315 2275 315 2275 315 2275 320 2448 320 2448 320 2448 320 2448 320 2448 PTLR Revision 9 Beaver Valley Unit 1 5.2 - 13 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 SETPOINT OPPS Enable Temperature 347 ° F PORV Setpoint s 397 psig PTLR Revision 9 Beaver Valley Unit 1 5.2 - 14 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) ARTNDic) 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 0.9862 147.7 0.952 145.66 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 180.2 0.952 0.9862 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
ARTNor) + L(FF2) = (1585.86) + (11.154) = 142.2 ° F(e) 169.4 V 0.297 0.6677 113.10 0.446 (159.8)
Beaver Valley u 0.618 0.8652 174.8 (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) 189.2 y 2.10 1.2018 227.40 1.444 (178.5) 252.1 X 4.99 1.4020 353.39 1.965 (237.8)
SUM: 1039.87 5.577 CF= L(FF
ARTNDT) + L(FF2) = (1039.87) + (5.577)= 186.5° F(e)
Notes:
(a) f= Calculated surveillance capsule neutron fluence (x 10 19 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- 2 8 1
109 f)_
(c) ARTNor values are the measured 30 ft-lb shift values. The Beaver Valley Unit 1 ARTNor 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 ARTNor values by to obtain adjusted ARTNor 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) RTNDid) FF *RTNDT FF2 82.6 97° 0.5174 0.8160 67.44 0.666 (72.34)
Weld Metal 81.1 Heat# 90136(e> 104° 0.7885 0.9333 75.68 0.871 (67.4)
(St. Lucie Unit 1) 83.8 284° 1.243 1.0606 88.85 1.125 (68.0)
Weld Metal 67.5 97 ° 0.324 0.6902 (65.93) 46.61 0.476 Heat# 90136(e)
(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
RTNDT) + L(FF2) = (405.50)+ (5.437) = 74.6°F<9>
W-225 0.488 0.800 197.30 157.83 0.640 (210)
Weld Metal 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
RTNOT)+ L(FF ) = (607.64)+ (2.802)= 216.9 F(9) 2 Notes for Table 5.2-4a are on the following page.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 16 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- 2 8 1
109 f)_
(d) RTNor values are the measured 30 ft-lb. shift values. RTNor 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 RTNor values for the weld data are adjusted by a ratio of 1.17. The Millstone Unit 2 and Fort Calhoun RT 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 17 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)
Irradiated Capsule f(d)
Cu Ni RT NDie)
Material Capsule Temperature(c) (x10 19 n/cm 2 ,
(wt.%) (wt.%) (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.
PTLR Revision 9 Beaver Valley Unit 1 5.2-18 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 Position 1 . 1 Initial Cu Ni Material Description Chemistry RT NDT(a)
(wt.%) (wt.%)
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.
PTLR Revision 9 Beaver Valley Unit 1 5.2-19 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)
Initial Material Description Material Heat Number Wt% Wt%
RT NDic)
ID (Lot Number) Cu Ni (OF)
Upper Shell Forging 86604 123V339VA1 0.12 (b) 0.68 40 305414 0.34 0.61 -56 (Gen)
(3951 & 3958)
Upper to Intermediate AOFJ 0.03 0.93 10 (Gen)10-714 Shell Girth Weld FOIJ 0.03 0.94 10 (Gen)
EODJ 0.02 1.04 10 (Gen)
HOCJ 0.02 0.93 10 (Gen) 86608-1 95443-1 0.10 0.82 48.5 Inlet Nozzles 86608-2 95460.:.1 0.10 0.82 -15.2 86608-3 95712-1 0.08 0.79 11.4 EODJ 0.02 1.04 10 (Gen)
FOIJ 0.03 0.94 10 (Gen) 1-7178 HOCJ 0.02 0.93 10 (Gen)
Inlet Nozzle Welds 1-7170 O81J 0.02 0.97 10 (Gen) 1-717F EOEJ 0.01 1.03 10 (Gen)
ICJJ 0.03 0.99 10 (Gen)
JACJ 0.04 0.97 10 (Gen) 86605-1 95415-1 0.13(d) 0.77 -26.2 Outlet Nozzles 86605-2 95415-2 0.13(d} 0.77 3.3 86605-3 95444-1 0.09 0.79 10.1 ICJJ 0.03 0.99 10 (Gen)
IOBJ 0.02 0.97 10 (Gen) 1-717A JACJ 0.04 0.97 10 (Gen)
Outlet Nozzle Welds 1-717C HOCJ 0.02 0.93 10 (Gen) 1-717E EODJ 0.02 1.04 10 (Gen)
FOIJ 0.03 0.94 10 (Gen)
Notes for Table 5.2-6 are on the following page.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 20 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 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).
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 21 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) (OF)
Intermediate Shell Plate 86607-1 195.2 171.2 Intermediate Shell Plate 86607-2 222.0 198.0 Lower Shell Plate 87203-2 166.4 142.8 Lower Shell Plate 86903-1 244.0(f) 208.8(t>
- Using S/C Data(b) 237.3 203.3 Intermediate Shell Longitudinal Weld 19-714A/B 182.4 133.5
- Using S/C Data(b) 177.7 130.2 Intermediate to Lower Shell Circ. Weld 11-714 175.7 146.0
- Using S/C Data (c) 109.3 91.4 Lower Shell Longitudinal Weld 20-714A/B 199.9 146.2
- Using S/C Data(d) 205.6 150.3 Upper Shell Forging 86604 139.4 119.2 Upper Shell to Intermediate Shell Girth Weld 10-714 172.9 122.5 (Heat # 305414)
-Using S/C Data(d) 177.9 125.9 Upper Shell to Intermediate Shell Girth Weld 10-714 88.4 44.0 (Heat #'s AOFJ and FOIJ)
Upper Shell to Intermediate Shell Girth Weld 10-714 44.0 44.0 (Heat #'s EODJ and HOCJ)
Inlet and Outlet Nozzle Welds (All Heat #'s) 44.0 44.0 Notes:
(a) ART= I+ RTNoT+ M.
(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full crt1.)
(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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 22 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 Plate 86903-1 Location Lower Shell Lower Shell Plate Plate 1/4T ART(°F) 3/4T ART(°F)
Chemistry Factor, CF (°F) 147.2 147.2 Fluence (f), n/cm2 (E>1.0 Mev) 3.672 X 10 19 1.427 X 10 19 Fluence Factor, FF 1.3374 1.0987 LiRTNDT = CF X FF(°F) 196.9(b) 161.7(b)
Initial RTNDT, 1(°F)(a) 13.1 13.1 Margin, M(°F) 34(b) 34(b)
ART = l+(CF*FF)+M, °F per RG 1.99, Revision 2 244.0(d) 208.8(d)
Notes:
(a) Initial RTNor values are measured values for plate material.
(b) Based on Regulatory Guide 1.99, Revision 2 Position 1.1.
(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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 23 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 T able 5.2-9 (Page 1 of 2)
RT PTS Calculation for Beltline Region Materials at Life Extension (50 EFPY) <a>
Surface Fluence Chemistry Initial Material Heat RT PTS(d) Ou Oi1. Margin<e> RT PTS(f)
Material Description Fluence Factor, Factor RT ND/c)
ID Number (OF) (OF) (OF) (OF) (OF)
(x10 19 n/cm2) FF ( ° F) ( ° F)
Intermediate Shell Plate 86607-1 --- 5.57 1.4231 100.5 43 143.0 0 17 34 220.0 Intermediate Shell Plate 86607-2 - -- 5.57 1.4231 100.5 73 143.0 0 17 34 250.0 Lower Shell Plate 86903-1 --- 5.57 1.4231 147.2 27 209.5 0 17 34 270.5 Using non-credible surveillance data<9> 5.57 1.4231 151.8 27 216.0 0 17<9> 34 277.0 Lower Shell Plate 87203-2 --- 5.57 1.4231 98.7 20 140.5 0 17 34 194.5 Intermediate to Lower 11-714 90136 5.55 1.4225 124.3 -56 176.8 17 28 65.5 186.3 Shell Girth Weld Using credible surveillance data<h> 5.55 1.4225 87.1 -56 123.9 17 14(h) 44.0 111.9 Intermediate Shell 19-714 305424 1.08 1.0224 191.7 -56 196.0 17 28 65.5 205.5 Longitudinal Weld A&8 Using non-credible surveillance data<9> 1.08 1.0224 192.3 -56 196.6 17 28<9 > 65.5 206.1 Lower Shell Longitudinal 20-714 305414 1.09 1.0241 210.5 -56 215.6 17 28 65.5 225.1 Weld A&8 Using non-credible surveillance data(i> 1.09 1.0241 216.9 -56 222.1 17 28(i) 65.5 231.6 Notes:
(a) Data obtained from Table 6-3 of Reference 9.
(b) FF = fluence factor= f <0- 28-0-10109 <t>>_
(c) Initial RT NDT values are measured values with the exception of the vessel welds.
(d) RT PTS = CF* FF.
(e) M= 2*(cru 2 + crl) 112 .
(f) RT PTs= Initial RT NDT + RT PTs + Margin.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 24 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 2 of 2)
RTPrs 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 25 LRM Revision 98
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-10 (Page 1 of 2)
RTPrs Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)<a)
Surface Fluence Chemistry Initial Material Material Heat Number RTPTS(d) Ou 0 Margin<e) RTPTS(f)
Fluence Factor, Factor RT NDT(c)
Description ID (Lot Number) (OF) (OF) (OF) (OF) (OF)
(x10 19 n/cm2) FF<b) ( ° F) (° F)
Upper Shell 86604 123V339VA1 0.625 0.8685 84.2 40 73.1 0 17 34 147.1 Forging Upper to 305414 Intermediate 10-714 0.625 0.8685 209.11 -56 181.6 17 28 65.5 191.1 (3951 & 3958)
Shell Girth Weld Using non-credible surveillance data<g) 0.625 0.8685 216.9 -56 188.4 17 28(9) 65.5 197.9 AOFJ 0.625 0.8685 41.0 10 35.6 17 17.8 49.2 94.8 Upper to FOIJ 0.625 0.8685 41.0 10 35.6 17 17.8 49.2 94.8 Intermediate 10-714 EODJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 Shell Girth Weld HOCJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 86608-1 95443-1 0.016 0.1513 67.0 60 10.1 17 5.1 35.5 105.6 Inlet Nozzles 86608-2 95460-1 0.016 0.1513 67.0 60 10.1 17 5.1 35.5 105.6 86608-3 95712-1 0.016 0.1513 51.0 60 7.7 17 3.9 34.9 102.6 EODJ 0.016 0.1513 27.0 10 4.1 17 2.0 34.2 48.3 FOIJ 0.016 0.1513 41.0 10 6.2 17 3.1 34.6 50.8 1-717 8 HOCJ 0.016 0.1513 27.0 10 4.1 17 2.0 34.2 48.3 Inlet Nozzle 1-717 D D8IJ 0.016 0.1513 27.0 10 4.1 17 2.0 34.2 48.3 Welds 1-717 F EOEJ 0.016 0.1513 20.0 10 3.0 17 1.5 34.1 47.2 ICJJ 0.016 0.1513 41.0 10 6.2 17 3.1 34.6 50.8 JACJ 0.016 0.1513 54.0 10 8.2 17 4.1 35.0 53.1 86605-1 95415-1 0.011 0.1191 95.25 60 11.3 17 5.7 35.8 107.2 Outlet Nozzles 86605-2 95415-2 0.011 0.1191 95.25 60 11.3 17 5.7 35.8 107.2 86605-3 95444-1 0.011 0.1191 58.0 60 6.9 17 3.5 34.7 101.6 ICJJ 0.011 0.1191 41.0 10 4.9 17 2.4 34.3 49.2 IO8J 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 1-717 A Outlet Nozzle JACJ 0.011 0.1191 54.0 10 6.4 17 3.2 34.6 51.0 1-717 C Welds HOCJ 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 1-717 E EODJ 0.011 0.1191 27.0 10 3.2 17 1.6 34.2 47.4 FOIJ 0.011 0.1191 41.0 10 4.9 17 2.4 34.3 49.2 Notes:
(a) Data obtained from Table 6-4 of Reference 9.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 26 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- 2 s-o. 10 1og(f))_
(c) Initial RTNDT value for the upper shell forging is a measured value. All other values are generic.
(d) RT PTS = CF* FF.
(e) M=2*(cru 2 + crl) 112 .
(f) RT PTs = Initial RT NDT + RT 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 27 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)
UPPER SHELF ENERGY (FT-LB)
Cu Ni p TNDT RTNor COMPONENT HEAT NO. CODE NO. MATERIAL TYPE
(%) (%) (%) (OF) (OF) MWD NMWD Closure Head C6213-1B B6610 A533B CL. 1 .15 --- .010 -40 0* 121 ---
Dome Closure Head A5518-2 B6611 A533B CL. 1 .14 --- .015 -20 -20* 131 ---
Seq.
Closure Head ZV3758 --- A508 CL. 2 .08 --- .007 60* 60* >100 ---
Flange Vessel Flange ZV-3661 FV-2961 A508 CL. 2 .12 --- .010 -54.7** 10** 166 ---
Inlet Nozzle 9-5443-1 B6608-1 A508 CL. 2 .10 .82 .008 35.8** 48.5** 82.5 ---
Inlet Nozzle 9-5460-1 B6608-2 A508 CL. 2 .10 .82 .010 -18.3** -15.2** 94 ---
Inlet Nozzle 9-5712-1 B6608-3 A508 CL. 2 .08 .79 .007 -2.5** 11.4** 97 ---
Outlet Nozzle 9-5415-1 B6605-1 A508 CL. 2 .13 .77 .008 -26.2** -26.2** 93 ---
Outlet Nozzle 9-5415-2 B6605-2 A508 CL. 2 .13 .77 .007 3.0** 3.3** 112.5 ---
Outlet Nozzle 9-5444-1 B6605-3 A508 CL. 2 .09 .79 .007 10.1 ** 10.1 ** 103 ---
Upper Shell 123V339VA1 --- A508 CL. 2 .12 .68 .010 40 40* 155 101 Inter Shell C4381-2 B6607-2 A533B CL. 1 .14 .62 .015 -10 53.6 123 83 Inter Shell C4381-1 B6607-1 A533B CL. 1 .14 .62 .015 -10 26.8 128.5 94 Lower Shell C6317-1 B6903-1 A533B CL. 1 .21 .54 .010 -50 13.1 134 83 Lower Shell C6293-2 B7203-2 A533B CL. 1 .14 .57 .015 -20 0.4 129.5 85 Trans Ring 123V223 --- A508 CL. 2 --- --- --- 30 30* 143 ---
Bottom Hd Sea 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 305414 --- --- .34 .61 --- --- -56 --- > 100 Lono. Weld Weld HAZ --- --- --- -40 -40 --- 136.5
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 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.
PTLR Revision 9 Beaver Valley Unit 1 5.2 - 28 LRM Revision 98

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

Text

SUBJECT:

Enclosure:

Navigation menu

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

Text

SUBJECT:

Enclosure:

Navigation menu

Search