ML13150A010
| ML13150A010 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/29/2013 |
| From: | Emily Larson FirstEnergy Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| Download: ML13150A010 (59) | |
Text
FENOC FirstEnergy Nuclear Operating Company Eric A. Larson Site Vice President May 29,2013 L-13-084 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 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 Nos. 1 (BVPS-1) and 2 (BVPS-2) Technical Specification (TS) 5.6.4, "Reactor Coolant System (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," FirstEnergy Nuclear Operating Company (FENOC) hereby submits the BVPS-1 PTLR, Revision 6 and the BVPS-2 PTLR, Revision 5. TS Section 5.6.4.c requires that the PTLR be provided to the Nuclear Regulatory Commission (NRC) upon issuance for any revision or supplement thereto.
In the spring of 2000, Surveillance Capsule Y was pulled and the analysis was documented in WCAP-15571, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program." For the 60-year license, WCAP-15571, Supplement 1, Revision 2, dated September 2011, documents the end-of-license extension analysis for neutron fluence values. The BVPS-1 PTLR was revised on May 24,2013 to include the WCAP-15571, Supplement 1, Revision 2, updated fluence values. The BVPS-1 PTLR is provided in Enclosure A.
In the spring of 2005, Surveillance Capsule X was pulled and the analysis was documented in WCAP-16527-NP, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program." For the 60-year license, WCAP-16527-NP, Supplement 1, Revision 1, dated September 2011, documents the end-of-license extension analysis for neutron fluence values. The BVPS-2 PTLR was revised on May 24, 2013 to include the WCAP-16527-NP, Supplement 1, Revision 1, updated fluence values. The BVPS-2 PTLR is provided in Enclosure B. Additionally, the BVPS-2 PTLR was revised to identify that the actual initial nil-ductility transition reference temperature values are located in the BVPS-2 Updated Final Safety Analysis Report.
Beaver Valley Power Station, Units No.1 and 2 L-13-084 Page 2 WCAP-15571, Supplement 1, Revision 2 and WCAP-16527 -N P, Supplement 1,
Revision 1, were previously provided to the NRC by correspondence dated May 28, 2013.
There are no regulatory commitments contained in this letter. If there are any questions, or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.
Sincerely,
£""(/) ;i~
Eric A. Larson
Enclosures:
A Beaver Valley Power Station, Unit No.1, Pressure and Temperature Limits Report, Revision 6 B
Beaver Valley Power Station, Unit No.2, Pressure and Temperature Limits Report, Revision 5 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site Representative (BRP/DEP)
Enclosure A L-13-084 Beaver Valley Power Station, Unit No.1 Pressure and Temperature Limits Report, Revision 6 (26 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 N/A 5.2-3 3.4.7 N/A N/A 5.2-3 3.4.10 N/A 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 N/A N/A 5.2-3 Beaver Valley Unit 1 5.2 - i PTLR Revision 6 LRM Revision 80
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.
- 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.
- 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 were developed using a methodology specified in the Technical Specifications. The methodology listed in Reference 1 was used with two exceptions:
a)
Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1," and b)
Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure."
RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)
The RCS temperature rate-of-change limits defined in Reference 2 are:
- a.
A maximum heatup of 100°F in anyone hour period.
- b.
A maximum cooldown of 100°F in anyone hour period, and Beaver Valley Unit 1 5.2 - 1 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2
- c.
A maximum temperature change of less than or equal to 5°F in anyone hour period during inservice hydrostatic testing operations above system design pressure.
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 12.
Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 5.2.1, 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. The heatup and cool down 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 III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.
Figures 5.2-1 and 5.2-2 and Tables 5.2-1 and 5.2-2 are based upon analysis of Capsule Y per Reference 12. Reference 11 provides an updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factor values, and an updated fluence evaluation. Therefore, the applicability of the PIT limit curves (Reference 12) was assessed based on 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 11, the limiting material for the current BVPS-1 PIT limits continues to be the lower shell plate B6903-1 at 30 EFPY.
Since the adjusted reference temperature (ART) calculation is based on surveillance data for this limiting material, updated ART values are needed in order to assess the applicability of the existing curves. Using the fluence analysis provided in Table 5-1 of Reference 11, the maximum neutron fluence value at 30 EFPY is 3.33 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 Beaver Valley Unit 1 5.2 - 2 PTLR Revision 6 LRM Revision 80
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 this material, the limiting 1/4T and 3/4T ART values would be 242.9°F and 203.6°F, respectively. These values are less conservative than the limiting ART values summarized in Tables 5.2-6 and 5.2-7 (see Reference 12). The Reference 12 values were used to develop the 30 EFPY PIT limit curves provided in Figures 5.2-1 and 5.2-2 along with the data points contained in Tables 5.2-1 and 5.2-2. Since the ART values used in the development of the Capsule Y PIT limit curves remain bounding for 30 EFPY, the 30 EFPY PIT limits remain valid as documented in Reference 12.
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 Reference 1, including the exceptions noted in Section 5.2.1. The PORV lift setting 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. At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature is 34?DF.
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 does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 31SoF.
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.
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.
Beaver Valley Unit 1 5.2 - 3 PTLR Revision 6 LRM Revision SO
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.4 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 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 NDT, which is determined in accordance with ASME,Section III, NB-2331. The empirical relationship between RT NDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.
Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 1. This commitment is a condition of license Amendment 256 and will remain in effect until the NRC staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.
Beaver Valley Unit 1 5.2 - 4 PTLR Revision 6 LRM Revision 80
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 and Fort Calhoun Surveillance Capsule Data.
Table 5.2-4b shows the St. Lucie and Fort Calhoun Surveillance Weld Data.
Table 5.2-5, taken from Reference 12, provides the reactor vessel beltline material property table.
Table 5.2-6, taken from Reference 12, provides a summary of the Adjusted Reference Temperature (ARTs) for 30 EFPY.
Table 5.2-7, taken from Reference 12, shows the calculation of ARTs for 30 EFPY.
Table 5.2-8, taken from Reference 11, shows the reactor vessel extended beltline material properties.
Table 5.2-9, taken from Reference 11, provides RT PTS values for the beltline materials at 50 EFPY.
Table 5.2-10, taken from Reference 11, provides RT PTS values for the extended beltline materials at 50 EFPY.
Table 5.2-11, Reactor Vessel Toughness Data (Unirradiated)
Beaver Valley Unit 1 5.2 - 5 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report 5.2.4 References Pressure and Temperature Limits Report 5.2
- 1.
WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et aI., January 1996.
- 2.
Deleted
- 3.
WCAP-15571, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," C. Brown, et. aI., November 2000.
- 4.
WCAP-8457, "Duquesne Light Company, Beaver Valley Unit No.1 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, October 1974.
- 5.
WCAP-15569, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 1," C. Brown, et aI., November 2000.
- 6.
10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 7.
10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 8.
Regulatory Guide 1.99, Revision 2, "Radiation EmbriUlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
- 9.
Deleted
- 10.
FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
- 11.
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.
- 12.
WCAP-16799-NP, Revision 1, "Beaver Valley Power Station Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," B. N. Burgos, June 2007.
- 13.
FENOC-07-120, Transmittal of LTOPS Setpoint Analysis Report, July 26, 2007.
- 14.
Westinghouse Calculation CN-SCS-07-27, Revision 0, L TOPS Setpoint Evaluation for Beaver Valley Unit 1 at 30 EFPY.
- 15.
NUREG-0800, MTEB 5-2 and 5-3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," June 1987.
Beaver Valley Unit 1 5.2 - 6 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LIMITING ART VALUES AT 30 EFPY:
2500 LOWER SHELL PLATE B6903-1 1/4T,245.7°F 3/4T,207.6°F 2250 I Leak Test Limi~ I I I 2000 1 I I I ~
I Heatup Ratel /
100°F/Hr 1750
(
1500
/
I Unacceptable I Operation
"" / t 1
Critical Limit I I
1250
~ i 1000 750 500 250 o
o 50 100°F/Hr
/ /
I Acceptable I Operation
/
/
~
/
Boltup Criticality Limit based on Temperature inservice hydrostatic test GO°F temperature (302°F) for the
/
service period up to 30 EFPY 100 150 200 250 300 350 INDICATED TEMPERATURE (OF)
Figure 5.2-1 (Page 1 of 1)
Reactor Coolant System Heatup 400 Limitations Applicable for the First 30 EFPY (LCO 3.4.3) 450 500 I-550 Beaver Valley Unit 1 5.2 - 7 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LOWER SHELL PLATE B6903-1 1/4T, 245.7°F LIMITING ART VALUES AT 30 EFPY:
2600 2260 2000 1760 760 600 260 o
o 3/4T,207.6°F I
I II I
Unacceptable I Operation I /
V I Acceptable I Operation
~
./ II 3 r Cooldown Rates O°F/Hr (steady-state)
I---
20°F/Hr
--~
40°F/Hr r-- 60°F/Hr i---r-100°F/Hr Boltup
'/
Temperature
/
60°F 50 100 150 200 250 300 350 400 INDICATED TEMPERATURE (OF)
Figure 5.2-2 (Page 1 of 1)
Reactor Coolant System Cooldown Limitations Applicable for the First 30 EFPY (LCO 3.4.3) 450 500 Beaver Valley Unit 1 5.2 - 8 PTLR Revision 6 LRM Revision 80 550
Licensing Requirements Manual 2500 2000 G' 1500 CiS
~
w c::
- J en en w
~
Pressure and Temperature Limits Report 5.2
/
V '/
~
g:
1000 L-----'L-----'
500 a
50 I I 60 I
70 80 90 100 110 120 TEMPERATURE (OF)
Figure 5.2-3 (Page 1 of 1)
Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)
Beaver Valley Unit 1 5.2 - 9 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 100°F/hr Heatup T
p (OF)
(psig) 60 0
60 554 65 554 70 554 75 554 80 554 85 554 90 554 95 554 100 554 105 554 110 554 115 554 120 554 125 554 130 554 135 554 140 555 145 557 150 560 155 563 160 567 165 573 170 579 175 585 180 593 185 602 190 613 195 624 200 637 205 651 210 667 215 685 220 705 225 727 230 751 235 778 240 807 Leak Test Limit Beaver Valley Unit 1 Table 5.2-1 (Page 1 of 1)
Heatup Curve Data Points for 30 EFPY (LCD 3.4.3) 100°F/hr Heatup T
P (OF)
(psig) 245 840 250 876 255 917 260 961 265 1010 270 1064 275 1124 280 1189 285 1262 290 1342 295 1431 300 1528 305 1636 310 1754 315 1885 320 2029 325 2151 330 2282 335 2426 336.8 2485 100°F/hr Criticality T
p (OF)
(psig) 302 0
302 981 305 1010 310 1064 315 1124 320 1189 325 1262 330 1342 335 1431 340 1528 345 1636 350 1754 355 1885 360 2029 365 2151 370 2282 375 2426 376.8 2485 284 302 2000 2485 5.2 - 10 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Steady State T
p JOFJ (psig) 60 0
60 621 65 621 70 621 75 621 80 621 85 621 90 621 95 621 100 621 105 621 110 621 115 621 120 621 125 621 130 621 135 621 140 621 145 621 150 621 155 621 160 621 165 621 170 621 175 621 180 621 180 747 185 758 190 771 195 784 200 800 205 816 210 835 215 856 220 878 225 903 230 931 235 962 240 996 245 1033 250 1075 Beaver Valley Unit 1 Table 5.2-2 (Page 1 of 2)
Cooldown Curve Data Points for 30 EFPY (LCO 3.4.3) 20°F/hr 40°F/hr 60°F/hr T
P T
P T
P (OF)
(psig)
(OF)
(psig)
(OF)
(psig) 60 0
60 0
60 0
60 606 60 563 60 518 65 607 65 563 65 519 70 608 70 564 70 519 75 609 75 565 75 520 80 611 80 567 80 522 85 612 85 568 85 523 90 614 90 570 90 525 95 616 95 571 95 526 100 618 100 574 100 528 105 620 105 576 105 531 110 621 110 578 110 533 115 621 115 581 115 536 120 621 120 585 120 540 125 621 125 588 125 544 130 621 130 592 130 548 135 621 135 597 135 553 140 621 140 602 140 558 145 621 145 607 145 564 150 621 150 614 150 571 155 621 155 621 155 578 160 621 160 621 160 586 165 621 165 621 165 595 170 621 170 621 170 606 175 621 175 621 175 617 180 621 180 621 180 621 180 708 180 669 180 630 185 720 185 682 185 644 190 733 190 696 190 660 195 748 195 713 195 677 200 765 200 730 200 697 205 783 205 750 205 718 210 803 210 772 210 742 215 825 215 796 215 768 220 850 220 823 220 797 225 877 225 853 225 830 230 908 230 886 230 866 235 941 235 922 235 906 240 978 240 962 240 950 245 1019 245 1007 245 999 250 1064 250 1056 250 1053 5.2 - 11 100°F/hr T
p (OF)
(psig) 60 0
60 425 65 426 70 427 75 428 80 429 85 430 90 432 95 433 100 435 105 438 110 441 115 444 120 448 125 452 130 457 135 462 140 468 145 475 150 483 155 491 160 501 165 512 170 524 175 537 180 552 185 569 190 588 195 608 200 631 205 657 210 685 215 717 220 752 225 791 230 835 235 883 240 936 245 995 250 1053 255 1111 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Steady State T
p (OF)
(psig) 255 1121 260 1171 265 1227 270 1289 275 1357 280 1433 285 1516 290 1608 295 1710 300 1823 305 1947 310 2085 315 2237 320 2405 322.1 2485 Beaver Valley Unit 1 Table 5.2-2 (Page 2 of 2)
Cooldown Curve Data Points for 30 EFPY (LCO 3.4.3) 20°F/hr 40°F/hr 60°F/hr T
P T
P T
P (OF)
(psig)
(OF)
(psig)
(OF)
(psig) 255 1114 255 1111 255 1111 260 1169 260 1169 260 1169 265 1227 265 1227 265 1227 270 1289 270 1289 270 1289 275 1357 275 1357 275 1357 280 1433 280 1433 280 1433 285 1516 285 1516 285 1516 290 1608 290 1608 290 1608 295 1710 295 1710 295 1710 300 1823 300 1823 300 1823 305 1947 305 1947 305 1947 310 2085 310 2085 310 2085 315 2237 315 2237 315 2237 320 2405 320 2405 320 2405 322.1 2485 322.1 2485 322.1 2485 5.2 - 12 100°F/hr T
P (OF)
(psig) 260 1169 265 1227 270 1289 275 1357 280 1433 285 1516 290 1608 295 1710 300 1823 305 1947 310 2085 315 2237 320 2405 322.1 2485 PTLR Revision 6 LRM Revision 80
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 - 13 SETPOINT 347°F
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 ra)
FF(b)
~RTNOT(C)
FF *~RTNOT FF2 Lower Shell V
0.299 0.669 128.49 86.01 0.448 Plate U
0.604 0.859 118.93 102.14 0.738 B6903-1(d)
W 0.930 0.980 148.52 145.50 0.960 (Longitudinal) y 2.05 1.196 142.18 169.98 1.429 V
0.299 0.669 137.81 92.25 0.448 Lower Shell U
0.604 0.859 131.84 113.23 0.738 Plate B6903-1(d)
W 0.930 0.980 179.99 176.33 0.960 (Transverse) y 2.05 1.196 166.93 199.58 1.429 SUM:
1085.02 7.150 CF = L:(FF * ~RT NOT) + L:(FF2) = (1085.02) + (7.150) = 151.8°F(e)
V 0.299 0.669 169.30 113.33 0.448 (159.72)
Beaver Valley U
0.604 0.859 176.30 151.41 0.738 (166.32)
Unit 1 198.99 Surveillance W
0.930 0.980 194.95 0.960 Weld Metal(d)
(187.73)
(Heat # 305424)
Y 2.05 1.196 190.47 227.72 1.429 (179.69)
SUM:
687.41 3.575 CF = L:(FF * ~RTNOT) + L:(FF2) = (687.41) + (3.575) = 192.3°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 8-1 of Reference 11.
(b)
FF = fluence factor = f (0.28 - 0.1 *Iog t).
(c)
~RT NOT values are the measured 30 ft-Ib shift values. The Beaver Valley Unit 1
~RT NOT 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 A-1 of Reference 11.
NOTE:
Per Regulatory Guide 1.99, Revision 2, 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 ~RT NOT values by to obtain adjusted ~RT NOT values.
In Table 6-1 of Reference 11, the ratio is determined to be 1.06 or (192.3/191.7).
(d)
The plate and weld surveillance data is deemed non-credible per Appendix A of Reference 11.
(e)
Position 2.1 chemistry factor values are summarized in Table 6-1 of Reference 11.
Beaver Valley Unit 1 5.2 - 14 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4a (Page 1 of 1)
Calculation of Chemistry Factors(a)
(Based on St. Lucie and Fort Calhoun Surveillance Capsule Data)
Material Capsule Capsule fb)
FF(C) ilRT NOT(d)
FF *ilRTNOT FF2 97 0
0.5174 0.816 82.65 67.44 0.666 (72.34)
Weld Metal 81.08 Heat # 90136(e) 104 0
0.7885 0.933 (67.4) 75.68 0.871 (St. Lucie Unit 1) 284 0
83.77 1.243 1.061 (68.0) 88.85 1.125 SUM:
- ilRT NOT) + L:(FF2) = (231.97) + (2.662) = 87.1°F(g)
W-225 0.488 0.800 197.30 157.83 0.640 (210)
Weld Metal 218.30 Heat # 305414(1)
W-265 0.847 0.953 (225) 208.13 0.909 (Fort Calhoun 215.90 Unit 1)
W-275 1.54 1.119 1219J 241.68 1.253 SUM:
- ilRTNOT) + L:(FF2) = (607.64) + (2.802) = 216.9°F(g)
Notes:
(a)
Use of St. Lucie and Fort Calhoun Surveillance Capsule Data approved by NRC letter dated February 20,2002, "BEAVER VALLEY POWER STATION, UNIT 1 -
ISSUANCE OF AMENDMENT RE: AMENDED PRESSURE-TEMPERATURE LIMITS (TAC NO. MB2301)."
(b) f = calculated surveillance capsule fluence values (x 1019 n/cm2, E > 1.0 MeV). The surveillance capsule fluence results are contained in Tables A-3 and A-5 of Reference 11.
(c)
FF = fluence factor = f (0.28-0.1 *Iogl).
(d) ilRT NOT values are the measured 30 ft-Ib. shift values. ilRT 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 A-3 and A-5 of Reference 11. 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 ilRT NOT values for the weld data are adjusted by a ratio of 1.17. The Fort Calhoun ilRT NOT values were not adjusted since the ratio was 0.99; therefore, a conservative value of 1.00 was used.
(e)
The St. Lucie Unit 1 surveillance data is deemed credible per Appendix A of Reference 11.
(f)
The Fort Calhoun Unit 1 surveillance data is deemed non-credible per Appendix A of Reference 11.
(g)
Position 2.1 chemistry factor values are summarized in Table 6-1 of Reference 11.
Beaver Valley Unit 1 5.2 - 15 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4b (Page 1 of 1)
St. Lucie and Fort Calhoun Surveillance Weld Data(a)(b)
Cu Ni Irradiated Capsule fe}
~RTNDT(d)
Material Capsule Temperature (X1019 n/cm2, (wt. %)
(wt. %)
(OF)
E> 1.0 MeV)
(OF)
Weld Metal 97° 0.23 0.07 541 0.5174 72.3 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 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)
Notes:
(a)
Use of St. Lucie and Fort Calhoun Surveillance Capsule Data approved by NRC letter dated February 20,2002, "BEAVER VALLEY POWER STATION, UNIT 1 -
ISSUANCE OF AMENDMENT RE: AMENDED PRESSURE-TEMPERATURE LIMITS (TAC NO. MB2301)."
(b)
Data contained in this table was obtained from Reference 3.
(c) f = calculated surveillance capsule fluence values.
(d)
~RT NDT values are the measured 30 ft-Ib shift values from Tables A-3 and A-5 of Reference 11.
Beaver Valley Unit 1 5.2 - 16 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5,2 Table 5,2-5 (Page 1 of 1)
Reactor Vessel 8eltline Material Properties Cu Ni Chemistry Initial Material Description (wt. %)
(wt. %)
Factor RTNDT(a)
(OF)
(OF)
Intermediate Shell Plate 86607-1 0,14 0,62 100,5 43 Intermediate Shell Plate 86607-2 0,14 0,62 100,5 73 Lower Shell Plate 86903-1 0,21 0,54 147,2 27 Lower Shell Plate 87203-2 0,14 0,57 98,7 20 Intermediate to Lower Shell Weld 0.27 0,07 124,3
-56 Seam (Heat 90136)11-714 Intermediate Longitudinal Shell 0,28 0,63 191,7
-56 Weld Seams (Heat 305424)19-714 A&8 Lower Longitudinal Weld Seams 0,34 0,61 210,5
-56 (Heat 305414)20-714 A&8 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, 8eaver Valley Unit 1 5,2 - 17 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 1 of 1)
Summary of Adjusted Reference Temperature (ARTs) for 30 EFPy(e) 30 EFPY Material Description 1/4T ART(a) 3/4T ART(a)
(OF)
(OF)
Intermediate Shell Plate B6607-1 201.4 175.8 Intermediate Shell Plate B6607-2 231.4 205.8 Lower Shell Plate B7203-2 176.2 151 Lower Shell Plate B6903-1 243.2 205.7
- Using SIC Data(b) 245.7 207.6 Intermediate Shell Longitudinal Weld 19-714A1B 161.9 115.4
- Using SIC Data(b) 159.6 113.8 Intermediate to Lower Shell Circ. Weld 11-714 163.4 131.7
- Using SIC Data (e) 93.0 71.4 Lower Shell Longitudinal Weld 20-714A/B 176.8 125.8
- Using SIC Data(d) 187.5 133.2 Notes:
(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full O'~.)
(c) Based on St. Lucie Unit 1 surveillance data. (Data credible. ART calculated with a reduced O'~.)
(d) Based on Fort Calhoun Unit 1 surveillance data. (Data not credible. ART calculated with a full O'~.)
(e) This table has not been updated to reflect updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factors, or updated fluence analysis (Reference 11); however, values listed here remain bounding. See Section 5.2.1.1 for additional information.
Beaver Valley Unit 1 5.2 - 18 PTLR Revision 6 LRM Revision 80
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 30 EFPy(d)
Parameter VALUES Operating Time 30 EFPY Material Plate B6903-1 Plate B6903-1 Location Lower Shell Lower Shell Plate Plate 1/4T ART(OF) 3/4T ART(OF)
Chemistry Factor, CF (OF) 149.2 149.2 Fluence (f), n/cm2 (E>1.0 Mev)(a) 2.4194 x 1019 9.404 X 1018 Fluence Factor, FF 1.238
.9828
~RT NOT = CF X FF(oF)(e) 184.7 (e) 146.6 Initial RT NOT, 1(0 F) (a) 27 27 Margin, M(OF) 34 (e) 34 ART = I+(CF*FF)+M, °F(b) per RG 1.99, Revision 2 245.7 207.6 Notes:
(a)
Initial RT NOT values are measured values for plate material.
(b)
This value was rounded per ASTM E29, using the "Rounding Method."
(c)
Based on Beaver Valley Unit 1 surveillance data. (Data not credible.
ART calculated with a full 0'[,.)
(d)
This table has not been updated to reflect updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factors, or updated fluence analysis (Reference 11); however, values listed here remain bounding. See Section 5.2.1.1 for additional information.
Beaver Valley Unit 1 5.2 - 19 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-8 (Page 1 of 1)
Reactor Vessel Extended 8eltline Material Properties(a)
Material Heat Number Wt%
Wt%
Initial Material Description RT NOT(C) 10 (Lot Number)
Cu Ni (OF)
Upper Shell Forging 86604 123V339VA1 0.12(b) 0.68 40 305414 (3951) 0.337 0.609
-56 (Gen) 305414 (3958) 0.337 0.609
-56 (Gen)
Upper to Intermediate 10-714 AOFJ 0.03 0.93 10 (Gen)
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 60 (Gen)
Inlet Nozzles 86608-2 95460-1 0.10 0.82 60 (Gen) 86608-3 95712-1 0.08 0.79 60 (Gen)
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 D81J 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 60 (Gen)
Outlet Nozzles 86605-2 95415-2 0.13(d) 0.77 60 (Gen) 86605-3 95444-1 0.09 0.79 60 (Gen)
ICJJ 0.03 0.99 10 (Gen) 1-717A I08J 0.02 0.97 10 (Gen)
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:
(a) Data obtained from Table 4-2 of Reference 11.
(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 is a measured value. The generic initial RT NOT values for the remaining materials were determined in accordance with NUREG-0800 [Reference 15] 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).
8eaver Valley Unit 1 5.2 - 20 PTLR Revision 6 LRM Revision 80
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 Material Description ID Number Intermediate Shell Plate 86607-1 Intermediate Shell Plate 86607-2 Lower Shell Plate 86903-1
~-USin-g-n-on~credlb-le-su-rVeiICance-data(gf-------
Lower Shell Plate 87203-2 Intermediate to Lower 11-714 90136 Shell Girth Weld
~ -Usln-g -credible-surveillance -data(hf----------
, _____ t;;;T!~_~r;~~~~~~ _____ J ___ ~~~~_~ __ t ____ ~~~_~~~ _____
I
-+ Using non-credible surveillance data(g) 1_ L_o:e~ ~ ~;~~n~itudl~al_ J __ ~~~~_4 J ____ 3~~
1_4 _____
I
-+ Using non-credible surveillance data(')
Notes:
(a) Data obtained from Table 6-3 of Reference 11.
(b) FF = fluence factor = f(o.28-o.10Jog(t)).
Surface Fluence Chemistry Fluence
- Factor, Factor (X1019 n/cm2)
FF(b)
(OF) 5.57 1.4231 100.5 5.57 1.4231 100.5 5.57 1.4231 147.2 5.57 1.4231 151.8 5.57 1.4231 98.7 5.55 1.4225 124.3 5.55 1.4225 87.1 1.08 1.0224 191.7 1.08 1.0224 192.3 1.09 1.0241 210.5 1.09 1.0241 216.9 (c)
Initial RT NOT values are measured values with the exception of the vessel welds.
- FF.
(e) M = 2 *(cru2 + cr/:,.2) 1/2.
(f)
RT PTS = Initial RT NOT +.6.RT PTS + Margin.
8eaver Valley Unit 1 5.2 - 21 Initial
(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 aLI.
Margin(e)
RTpTS(t)
(OF)
(OF)
(OF) 17 34 220.0 17 34 250.0 17 34 270.5 y(iif- ------------- -----------
34 277.0 17 34 194.5 28 65.5 186.3 4(f;r- ------------- -----------
44.0 111.9 28 65.5 205.5
-zS(iif- ------------- -----------
65.5 206.1 28 65.5 225.1
-2S-(ff- ------------- -----------
65.5
.231.6 PTLR Revision 6 LRM Revision 80
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 Of). 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 Of). term of 1rF was utilized for BVPS-1 plate B6903-1. The credibility evaluation conclusions are contained in Appendix A of Reference 11.
(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 Of). term of 14°F was utilized for BVPS-1 weld heat 90136. The credibility evaluation conclusions are contained in Appendix A of Reference 11.
(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 Of). term of 28°F was utilized for BVPS-1 weld heat 305414.
The credibility evaluation conclusions are contained in Appendix A of Reference 11.
Beaver Valley Unit 1 5.2 - 22 PTLR Revision 6 LRM Revision 80
Licensing Requirements Manual Table 5.2-10 (Page 1 of 2)
Pressure and Temperature Limits Report 5.2 RT PTS Calculation for Extended Beltline Region Materials at Life Extension (50 EFPy)(a)
Material Description Upper Shell Forging Material 10 B6604 Heat Number (Lot Number) 123V339VA1 Upper to Intermediate 110-714 j 305414 f_~_~~1~§1r:t_~_'t{~_19_____________
(3951 & 3958)
~ Usir:lfL non-credible surveiiiance-data(g'j----
Upper to Intermediate Shell Girth Weld Inlet Nozzles Inlet Nozzle Welds Outlet Nozzles Outlet Nozzle Welds Notes:
10-714 B6608-1 B6608-2 B6608-3 1-717 B 1-717 D 1-717 F B6605-1 B6605-2 B6605-3 1-717 A 1-717 C 1-717 E AOFJ FOIJ EODJ HOCJ 95443-1 95460-1 95712-1 EODJ FOIJ HOCJ DBIJ EOEJ ICJJ JACJ 95415-1 95415-2 95444-1 ICJJ IOBJ JACJ HOCJ EODJ FOIJ Surface Fluence (X1019 n/cm2) 0.625 0.625 0.625
-+
0.625 0.625 0.625 0.625 0.016 0.016 0.016 0.016 0.016 0.016 0.016 0.016 0.016 0.016 0.011 0.011 0.011 0.011 0.011 0.011 0.011 0.011 0.011 (a) Data obtained from Table 6-4 of Reference 11.
Beaver Valley Unit 1 Fluence
- Factor, FF(b) 0.8685 0.8685 0.8685 0.8685 0.8685 0.8685 0.8685 0.1513 0.1513 0.1513 0.1513 0.1513 0.1513 0.1513 0.1513 0.1513 0.1513 0.1191 0.1191 0.1191 0.1191 0.1191 0.1191 0.1191 0.1191 0.1191 5.2 - 23 Chemistry Factor CF) 84.2 209.11
---r-------
216.9 41.0 41.0 27.0 27.0 67.0 67.0 51.0 27.0 41.0 27.0 27.0 20.0 41.0 54.0 95.25 95.25 58.0 41.0 27.0 54.0 27.0 27.0 41.0 Initial RT NOT(e) I ilRT PTS(d)
(OF)
(OF) 40 73.1
-56 181.6
-56 188.4 10 35.6 10 35.6 10 23.4 10 23.4 60 10.1 60 10.1 60 7.7 10 4.1 10 6.2 10 4.1 10 4.1 10 3.0 10 6.2 10 8.2 60 11.3 60 11.3 60 6.9 10 4.9 10 3.2 10 6.4 10 3.2 10 3.2 10 4.9 au a",
(OF)
(OF) o I 17 17 I 28
---1 i --r --28(9)" --
17 I 17.8 17 I 17.8 17 I 11.7 17 I 11.7 17 I 5.1 17 I 5.1 17 I 3.9 17 I 2.0 17 I 3.1 17 I
2.0 17 I 2.0 17 I 1.5 17 I 3.1 17 I 4.1 17 I 5.7 17 I 5.7 17 I 3.5 17 I 2.4 17 I
1.6 17 I 3.2 17 I 1.6 17 I 1.6 17 I 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 RTpTs(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 6 LRM Revision 80
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 = fluencefactor = f(o.28-o.1olog(f)).
(c) Initial RT NDT value for the upper shell forging is a measured value. All other values are generic.
- FF.
(e) M = 2 *(O'u2 + 0'",2)112.
(f)
RT PTS = Initial RT NDT + ilRT 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 crt. term of 28°F was utilized for BVPS-1 weld heat 305414. The credibility evaluation conclusions are contained in Appendix A of Reference 11.
Beaver Valley Unit 1 5.2 - 24 PTLR Revision 6 LRM Revision 80
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 C6213-18 86610 A5338 CL. 1
.15
.010
-40 Dome Closure Head A551 B-2 86611 A5338 CL. 1
.14
.015
-20 Seg.
Closure Head ZV375B A50B CL. 2
.OB
.007 60' Flange Vessel Flange ZV3661 A50B CL. 2
.12
.010 60' Inlet Nozzle 9-5443 A50B CL. 2
.10
.OOB 60' Inlet Nozzle 9-5460 A50B CL. 2
.10
.010 60' Inlet Nozzle 9-5712 A50B CL. 2
.OB
.007 60' Outlet Nozzle 9-5415 A50B CL. 2
.OOB 60' Outlet Nozzle 9-5415 A50B CL. 2
.007 60' Outlet Nozzle 9-5444 A50B CL. 2
.09
.007 60' Upper Shell 123V339 A50B CL. 2
.010 40 Inter Shell C43B1-2 86607-2 A5338 CL. 1
.14
.62
.015
-10 Inter Shell C43B1-1 86607-1 A5338 CL. 1
.14
.62
.015
-10 Lower Shell C6317-1 86903-1 A5338 CL. 1
.20
.54
.010
-50 Lower Shell C6293-2 87203-2 A5338 CL. 1
.14
.57
.015
-20 Trans RinQ 123V223 A50B CL. 2 30 80ttom Hd Seg C4423-3 8661B A5338 CL. 1
.13
.OOB
-30 80ttom Hd Dome C44B2-1 86619 A5338 CL. 1
.13
.015
-50 Inter to Lower 90136
.27
.07 Shell Weld Inter Shell Long.
305424
.2B
.63 Weld Lower Shell 305414
.34
-40
- Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Direction NMWD - Normal to Major Working Direction RTNDT UPPER SHELF ENERGY (FT-LB)
(OF)
MWD NMWD O'
121
-20' 131 60'
>100 60' 166 60' B2.5 60' 94 60' 97 60' 97 60' 112.5 60' 103 40' 155 73 123 B2.5 43 12B.5 90 27 134 BO 20 129.5 B3.5 30' 143
-29' 124
-33' 125.5
-56
> 100
-56
> 100
-56
> 100
-40 136.5 Note:
For evaluation of Inservice Reactor Vessel Irradiation damage assessments, the best estimate chemistry values reported in the latest response to Generic Letter 92-01 or equivalent document are applicable.
Beaver Valley Unit 1 5.2 - 25 PTLR Revision 6 LRM Revision 80
Enclosure B L-13-084 Beaver Valley Power Station, Unit No.2 Pressure and Temperature Limits Report, Revision 5 (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:-2 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 5.2-3 5.2-4 5.2-5 5.2-6 3.4.6 N/A N/A 5.2-3 3.4.7 N/A N/A 5.2-3 3.4.10 N/A N/A 5.2-3 3.4.12 5.2.1.2 5.2-8 5.2-3 5.2.1.3
. 3.5.2 N/A N/A 5.2-3 BVPS-2 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 N/A N/A 5.2-3 Beaver Valley Unit 2 5.2 - i PTLR Revision 5 LRM Revision 76
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 2 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.
- 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.
- 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 Loop Seal or Water Discharge 5.2.1 Operating Limits 5.2.1.1 The PTLR limits for Beaver Valley Power Station (BVPS) Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in Reference 1 was used with two exceptions:
a)
Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1," and b)
Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure."
RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)
The RCS temperature rate-of-change limits defined in Reference 2 are:
- a.
A maximum heatup of 60°F in anyone hour period.
- b.
A maximum cooldown of 100°F in anyone hour period, and Beaver Valley Unit 2 5.2 - 1 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2
- c.
A maximum temperature change of less than or equal to 5°F in anyone hour period during inservice hydrostatic testing operations above system design pressure.
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 Figures 5.2-2 through 5.2-6 and Table 5.2-2. These limits are defined in Reference 2. Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 5.2.1, the RCS PIT limits for heatup and cooldown shown in Figures 5.2-1 through 5.2-6 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. 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 5.2-7 were developed for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region.
This figure provides the ASME III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.
Figures 5.2-1 thru 5.2-6 and Tables 5.2-1 and 5.2-2 are based upon analysis of Capsule W per Reference 2. The tables and curves generated as a result of the Capsule X analysis (Reference 12) and presented in Reference 14 are less conservative with respect to those for the Capsule Wanalysis. As a result, while Tables 5.2-5, 5.2-8, and 5.2-9 are updated with Capsule X fluence data and ART calculations, the pressure-temperature limits provided in Tables 5.2-1 and 5.2-2 and Figures 5.2-1 thru 5.2-6 continue to reflect Capsule W values through 22 EFPY and are bounding.
Additionally, Reference 13 provides an updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factor values, and an updated fluence evaluation. Therefore, the applicability of the PIT limit curves (Reference 14) was assessed based on 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 13, the limiting material for the current 8VPS-2 PIT limits continues to be the intermediate shell plate 89004-1 at 22 EFPY.
8eaver Valley Unit 2 5.2 - 2 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and TempeJature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.2 5.2.1.3 Since the adjusted reference temperature (ART) calculation is not based on surveillance data for this limiting material, only a fluence comparison is needed in order to assess the applicability of the existing curves. Using the fluence analysis provided in Table 5-1 of Reference 13, the maximum neutron fluence value at 22 EFPY is 2.29 x 1019 n/cm2 (E > 1.0 MeV). This value was calculated by interpolating the fluence at the 0° azimuthal position for BVPS-2 from the end of Cycle 15 to the fluence value at the future projection out to 32 EFPY. The fluence of 2.43 x 1019 n/cm2 (E > 1.0 MeV) used to develop the 22 EFPY PIT limit curves generated as a result of the Capsule X analysis (Reference 12), is more conservative than the updated fluence of 2.29 x 1019 n/cm2 (E > 1.0 MeV). The Capsule W results which were interpolated from Table 6-14 of WCAP-15675 have a 22 EFPY fluence of 2.61 x 1019 n/cm2 (E > 1.0 MeV). Since the fluence used in the development of the Capsule X PIT limit curves remains bounding for 22 EFPY, the 22 EFPY PIT limits remain valid as documented in Reference 14.
Furthermore, as discussed above, the PIT limits at 22 EFPY documented in Figures 5.2-1 thru 5.2-6 are based on the Capsule W analysis (Reference 2),
which is bounding of the Capsule X PIT limits (Reference 14).
Overpressure Protection System (OPPS) Setpoints (LCO 3.4.12)
The power operated relief valves (PORVs) shall each have a nominal maximum lift setting that varies with RCS temperature and which does not exceed the limits in Figure 5.2-8 (Reference 11). The OPPS enable temperature is in accordance with Table 5.2-3. The PORV lift setting provided is for the case with reactor coolant pump (RCP) restrictions. These restrictions are shown in Table 5.2-4, which is taken from Reference 9. Due to the setpoint limitations as a result of the reactor vessel flange requirements, there is no operational benefit achieved by restricting the number of RCPs running to less than two below an indicated RCS temperature of 13rF. Therefore, the PORV setpoints shown in Table 5.2-3 will protect the Appendix G limits for the combinations shown.
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 Reference 1, including the exceptions noted in Section 5.2.1. The PORV lift setting shown in Figure 5.2-8 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. At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature with uncertainty is 237°F.
Beaver Valley Unit 2 5.2 - 3 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 5.2.1.4 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 does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 240°F.
As the calculated enable temperature is higher and, therefore, more conservative than the arming temperature, the OPPS enable temperature, as shown in Table 5.2-3, is set to equal the calculated enable temperature.
The calculation method governing the heatup and cooldown of the RCS requires the arming of the OPPS at and below the OPPS enable temperature specified in Table 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.
The OPPS enable temperature, PORV setpoints, and RCP operating restrictions contained in Tables 5.2-3 and 5.2-4 and Figure 5.2-8 are as described in Reference 2, and are based upon analysis of Capsule W. The pressure-temperature limits provided in Reference 14 for Capsule X and setpoints evaluation per Reference 15 support the continued use of these existing OPPS/PORV setpoints and RCP operating restrictions for the period up to 22 EFPY. As a result, Tables 5.2-3 and 5.2-4 and Figure 5.2-8 continue to reflect Capsule W values and remain valid for Capsule X up to 22 EFPY.
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 switches (one in each train) into their "ARM" position. Once OPPS is activated (armed/enabled) reactor coolant system pressure transmitters will signal a rise in system pressure above the variable OPPS setpoint. This will initiate an alarm in the control room and open the OPPS PORVs.
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.
Beaver Valley Unit 2 5.2 - 4 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2 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 5.3-6 of the UFSAR. Also, the results of these analyses shall be used to update Figures 5.2-1 through 5.2-6, 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 NDT, which is determined in accordance with ASME,Section III, NB-2331. The empirical relationship between RT NDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.
Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 2. This commitment is a condition of License Amendment 138 and will remain in effect until the NRC staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.
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 5.2-5, taken from Table 2-4 of Reference 13, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
Table 5.2-6, taken from Table 2-1 of Reference 14, provides the reactor vessel beltline material property table.
Table 5.2-7, taken from Table 4-2 of Reference 13, provides the reactor vessel extended beltline material property table.
Table 5.2-8, taken from Tables 4-5 and 4-6 of Reference 14, provides a summary of the Adjusted Reference Temperature (ARTs) for 22 EFPY.
Beaver Valley Unit 2 5.2 - 5 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2 Table 5.2-9, taken from Tables 4-5 and 4-6 of Reference 14, shows the calculation of ARTs for 22 EFPY.
Table 5.2-10, taken from Table 6-3 of Reference 13, provides RTpTs values for the Beltline Region Materials at 54 EFPY.
Table 5.2-11, taken from Table 6-4 of Reference 13, provides RT PTS values for the Extended Beltline Region Materials at 54 EFPY.
Note that Tables 5.2-5, 5.2-8 and 5.2-9 have been updated to reflect Capsule X analysis and fluence data. This data has not, however, been incorporated into the pressure-temperature limits provided in Figures 5.2-1 thru 5.2-6 and Tables 5.2-1 and 5.2-2, which continue to reflect Capsule Wanalyses. See Section 5.2.1.1 for additional information.
5.2.4 References
- 1.
WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et aI., January 1996.'
- 2.
WCAP-15677, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," J. H. Ledger, August 2001.
- 3.
WCAP-15675, Revision 0, "Analysis of Capsule W from First Energy Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," J. H. Ledger, S. L. Anderson, J. Conermann, August 2001.
- 4.
WCAP-9615, Revision 1, "Duquesne Light Company, Beaver Valley Unit No.2 Reador Vessel Radiation Surveillance Program," P. A. Peter, June 1995.
- 5.
WCAP-15676, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2," J. H. Ledger, August 2001.
- 6.
10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 7.
10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 8.
Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
- 9.
FENOC Calculation No.1 0080-SP-2RCS-006, Revision 4, Addendum 0, "BV-2 L TOPS Setpoint Evaluation Capsule W for 22 EFPY."
Beaver Valley Unit 2 5.2 - 6 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual 5.2 Pressure and Temperature Limits Report Pressure and Temperature Limits Report 5.2
- 10.
FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
- 11.
Westinghouse Letter FENOC-04-31, dated April 14, 2004, "L TOPS Setpoint Evaluation for Beaver Valley Unit 2 Capsule W for 22 EFPY - Calculation Note."
- 12.
WCAP-16527, Revision 0, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," B. N. Burgos, J. Conermann, S. L. Anderson, March 2006.
- 13.
WCAP-16527, Supplement 1, Revision 1, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," A. E. Freed, September 2011.
- 14.
WCAP-16528, Revision 1, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," June 2008.
- 15.
Westinghouse Letter FENOC-07-92, dated June 8,2007, LTOPS Setpoint Evaluation for Beaver Valley Unit 2 Capsule X at 22 and 30 EFPY.
- 16.
Westinghouse Letter MCOE-L TR-13-19, Revision 0, dated March 6, 2013, "Acceptable Initial RT NDT Values for the Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle Materials."
Beaver Valley Unit 2 5.2 -7 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F LIMITING ART VALUES AT 22 EFPY:
3/4T, 129°F CURVES APPLICABLE FOR HEATUP RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2500.----r--~~---,---~----r_--~--_,----,_--~--~
6' (j) 0-2250 2000 1750
- 1500 0
- ::
- J (J)
(J)
~ 1250 0-c W
!;( 1000 u
is z
750 500 250 Leak Test Limit Unacceptable Operation Boltup Temperature I
I I 1...1 I
i I
. Iii ****Ii I.!
I i
~Criticality Limit for 60 of/Hr. I I
l I
I IIJ Criticality Limit based on inservice hydrostatic test temperature (196°F) for the service period up to 22 EFPY.
O~~~~~~~~~~~~~~~~~~~~~~~~
o 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (oF)
Figure 5.2-1 (Page 1 of 1)
Reactor Coolant System Heatup Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 8 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F LIMITING ART VALUES AT 22 EFPY:
3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO O°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2500~-----~----~--~--~----'---------'----r--~----.
I 2250 2000
& 1750 en a.. -
~ 1500
- l en en
~ 1250 D.. c
~ 1000
~
i5 z
750 500 250 o
Unacceptable Operation 50 Acceptable Operation Cooldown Rate OaF/Hr.
Boltup Temperature 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (oF)
Figure 5.2-2 (Page 1 of 1)
Reactor Coolant System Cooldown (up to O°F/Hr.)
Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 9 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F LIMITING ART VALUES AT 22 EFPY:
3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 20°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2250 2000
& 1750 en a.. -
~ 1500 en en
~ 1250 a.. o 5 1000 C
z 750 500 250 o
50 Acceptable Operation Cooldown Rate 20°F/Hr.
Boltup Temperature 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)
Figure 5.2-3 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 20°F/Hr.)
Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 10 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual MATERIAL PROPERTY BASIS LIMITING MATERIAL:
Pressure and Temperature Limits Report 5.2 LIMITING ART VALUES AT 22 EFPY:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F 3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 40°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2250 Unacceptable.
2000 G' 1750 (J)
- a. -
~ 1500
- J (J)
(J)
~ 1250
- a.
o
~ 1000 (3
c z 750 500 250 o
Operation Acceptable Operation I I I
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Boltup Temperature I
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50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)
Figure 5.2-4 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 40°F/Hr.)
Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 11 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F LIMITING ART VALUES AT 22 EFPY:
3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2500 2250 2000 G' 1750 Ci5 D. -
w 1500 0:::
en en w 1250 0:::
D.
C w 1000 e:(
() c z
750 500 250 o
Unacceptable Operation I
Acceptable Operation Cool down Rate GO°F/Hr.
Boltup Temperature 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)
Figure 5.2-5 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 60°F/Hr.)
Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 12 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 1/4T, 140°F LIMITING ART VALUES AT 22 EFPY:
3/4T, 129°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 22 EFPY.
2500~--~----~--~----~--~-----~--~----~--~--~
2250 2000 (3' 1750 C/)
D--
w 1500 0::.
C/)
C/)
~ 1250 D-C
~ 1000
~
C z
750 500 250 o
50 I
Acceptable Operation I
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Boltup Temperature I
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I 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)
Figure 5.2-6 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 100°F/Hr.)
Limitations Applicable for the First 22 EFPY (LCO 3.4.3)
NOTE: Values based upon analysis of Capsule W for 22 EFPY and are bounding for Capsule X at 22 EFPY (see Section 5.2.1.1). As a result, ART values shown do not coincide with Tables 5.2-8 and 5.2-9.
Beaver Valley Unit 2 5.2 - 13 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual 2500 2000 8'
Ci5 1500 c.. -
w a::
(f)
(f) w a:: 1000 c..
500 o
I
~
~
~
50 60 70
~
I Pressure and Temperature Limits Report 5.2 I
I I
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/
/
~
~
80 90 100 110 120 TEMPERATURE (OF)
Figure 5.2-7 (Page 1 of 1)
Isolated Loop Pressure - Temperature Limit Curve (LCO 3.4.3)
Beaver Valley Unit 2 5.2 - 14 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2
§'
(J)
- c. -
I-
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o C.
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- E.
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See Table 5.2-4 for RCP restrictions.
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Figure 5.2-8 (Page 1 of 1)
Maximum Allowable Nominal PORV Setpoint for the Overpressure Protection System (LCO 3.4.12)
NOTE: Data shown reflects analysis for Capsule W thru 22 EFPY and is the same or bounding for Capsule X thru 22 EFPY (See Section 5.2.1.3).
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Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 1 of 1)
Heatup Curve Data Points for 22 EFPY (LCO 3.4.3) 60°F/HR HEATUP 60°F/HR CRITICALITY LEAK TEST LIMIT Temp.
Press.
Temp.
Press.
Temp.
Press.
(OF)
(psig)
(OF)
(psi g)
(OF)
(psig) 60 0
196 0
178 2000 60 621 196 621 196 2485 65 621 196 621 70 621 196 621 75 621 196 621 80 621 196 621 85 621 196 621 90 621 196 621 95 621 196 621 100 621 196 621 105 621 196 621 110 621 196 621 115 621 196 621 120 621 196 779 120 621 196 799 120 779 196 821 125 799 196 846 130 821 196 874 135 846 196 905 140 874 196 940 145 905 196 978 150 940 200 1021 155 978 205 1068 160 1021 210 1120 165 1068 215 1178 170 1120 220 1242 175 1178 225 1312 180 1242 230 1390 185 1312 235 1476 190 1390 240 1571 195 1476 245 1675 200 1571 250 1791 205 1675 255 1919 210 1791 260 2060 215 1919 265 2215 220 2060 270 2387 225 2215 230 2387 NOTE: Data shown reflects analysis for Capsule W thru 22 EFPY and is the same or bounding for Capsule X thru 22 EFPY (See Section 5.2.1.3).
Beaver Valley Unit 2 5.2 - 16 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 1 of 1)
Cooldown Curve Data Points for 22 EFPY (LCO 3.4.3)
O°F/HR 20°F/HR 40°F/HR 60°F/HR 100°F/HR Temp.
Press.
Press.
Press.
Press.
Press.
(OF)
(psig)
(psig)
(psig)
(psig)
(psig) 60 0
0 0
0 0
60 621 621 621 608 532 65 621 621 621 618 544 70 621 621 621 621 557 75 621 621 621 621 572 80 621 621 621 621 588 85 621 621 621 621 606 90 621 621 621 621 621 95 621 621 621 621 621 100 621 621 621 621 621 105 621 621 621 621 621 110 621 621 621 621 621 115 621 621 621 621 621 120 621 621 621 621 621 120 621 621 621 621 621 120 907 884 862 842 807 125 935 914 895 877 849 130 966 948 932 917 897 135 1001 985 972 961 949 140 1039 1026 1017 1010 1007 145 1081 1072 1066 1064 1071 150 1127 1122 1121 1123 1127 155 1179 1178 1179 1179 1179 160 1235 1235 1235 1235 1235 165 1298 1298 1298 1298 1298 170 1367 1367 1367 1367 1367 175 1444 1444 1444 1444 1444 180 1528 1528 1528 1528 1528 185 1622 1622 1622 1622 1622 190 1725 1725 1725 1725 1725 195 1839 1839 1839 1839 1839 200 1966 1966 1966 1966 1966 205 2105 2105 2105 2105 2105 I 210 2259 2259 2259 2259 2259 215 2430 2430 2430 2430 2430 NOTE: Data shown reflects analysis for Capsule W thru 22 EFPY and is the same or bounding for Capsule X thru 22 EFPY (See Section 5.2.1.3).
Beaver Valley Unit 2 5.2 - 17 PTLR Revision 5 LRM Revision 76
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 240°F PORV Setpoint Figure 5.2-8 NOTE: Data shown reflects analysis for Capsule W thru 22 EFPY and is the same or bounding for Capsule X thru 22 EFPY (See Section 5.2.1.3).
Beaver Valley Unit 2 5.2 - 18 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-4 (Page 1 of 1)
Reactor Coolant Pump Restrictions TRcs Running RCPs
< 137°F 0-2
~ 137°F 3
NOTE: Data shown reflects analysis for Capsule W thru 22 EFPY and is the same or bounding for Capsule X thru 22 EFPY (See Section 5.2.1.3).
Beaver Valley Unit 2 5.2 - 19 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-5 (Page 1 of 1)
Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule fa)
FF(b) ilRTNoT(C)
FF*ilRTNOT FF2 Intermediate U
0.615 0.864 24.0 20.73 0.746 Shell Plate V
2.64 1.260 56.0 70.54 1.587 B9004-2(d)
W 3.61 1.334 71.0 94.68 1.778 (Longitudinal)
X 5.63 1.425 98.0 139.65 2.031 Intermediate U
0.615 0.864 17.7 15.29 0.746 Shell Plate V
2.64 1.260 46.1 58.07 1.587 B9004-2(d)
W 3.61 1.334 63.4 84.55 1.778 (Transverse)
X 5.63 1.425 104.1 148.34 2.031 SUM:
- ilRTNoT) -;- L(FF2) = (631.87) -;- (12.284) = 51.4°F Beaver Valley U
0.615 0.864 4.1 3.54 0.746 Unit2 V
2.64 1.260 25.7 32.37 1.587 Surveillance
. Weld Metal(e)
W 3.61 1.334 6.0 8.00 1.778 (Heat #83642)
X 5.63 1.425 22.9 32.63 2.031 SUM:
- ilRTNoT) -;- L(FF2) = (76.55) -;- (6.142) = 12.5°F 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 8-1 of Reference 13.
(b)
FF = fluence factor = f (0.28" 0.1 *'og f).
(c) ilRT NOT values are the measured 30 ft-Ib shift values. The BVPS-2 ilRT NOT values for the surveillance weld data were not adjusted since the ratio was 0.91; therefore, a conservative value of 1.00 was used.
(d)
The surveillance plate data is deemed non-credible, per Appendix A of Reference 13.
(e)
The surveillance weld data is deemed credible, per Appendix A of Reference 13.
Beaver Valley Unit 2 5.2 - 20 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 1 of 1)
Reactor Vessel 8eltline Material Properties Material Cu Ni Initial RT NOT (wt%)
(wt%)
(F)(a)
Closure Head Flange 89002-1 0.06(b) 0.74
-10 Vessel Flange 89001-1 0.06(b) 0.73 0
Intermediate Shell Plate 89004-1 0.065 0.55 60 Intermediate Shell Plate 89004-2 0.06 0.57 40 Lower Shell Plate 89005-1 0.08 0.58 28 Lower Shell Plate 89005-2 0.07 0.57 33 Intermediate to Lower Shell Weld 101-171 (Heat 83642) 0.046 0.086
-30 Intermediate Longitudinal Weld 101-124 A & 8 (Heat 83642) 0.046 0.086
-30 Lower Longitudinal Weld 101-142 A & 8 (Heat 83642) 0.046 0.086
-30 Plate Surveillance Material 89004-2 0.06 0.57 40 Surveillance Weld (Heat 83642) 0.065 0.065
-30(C)
Notes:
(a)
The initial RT NOT values for all of the beltline materials are based on measured data.
(b)
According to the 8VPS-2 reactor vessel CMTRs and MISC-PENG-ER-021, the material for the closure head flange (89002-1) and vessel flange (89001-1) forgings are ASTM A508 Class2. The ASTM A508 material specification does not require analysis of copper content. The importance of copper content in the irradiation embrittlement of ferritic pressure vessel steel was not recognized or regulated by the NRC or nuclear steam supply system (NSSS) vendors when the 8VPS-2 reactor vessel was constructed. Even though the material specification did not require analysis of copper content for ASTM A508 Class 2 material, check analyses on chemistry measurements (including copper) were reported in MISC-PENGER-021. The copper values reported for both the closure head flange (89002-1) and the vessel flange (89001-1) was 0.06%.
(c)
The initial RT NOT value is determined in accordance with the requirements of Subparagraph N8-2331 of Section III of the ASME 8&PV Code, as specified by Paragraph II - 0 of 10 CFR Part 50, Appendix G. These fracture toughness requirements are also summarized in 8ranch Technical Position MTE8 Section 11.5-2 ("Fracture Toughness") of the NRC Regulatory Standard Review Plan. Following these requirements, along with the Charpy data reported in Table 3-3 of WCAP-9615 and the T NOT value of
-30°F defined on page 3-14 of WCAP-9615, the initial RT NOT value is concluded to be equal to TNOT (i.e., -30.0°F).
8eaver Valley Unit 2 5.2 - 21 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-7 (Page 1 of 1)
Reactor Vessel Extended Beltline Material Properties (a)
Material Material Wt%
Wt%
Initial Heat Number RTNW Description ID Cu Ni (OF)\\ )
B9003-1 A9406-1 0.13 0.60 50 Upper Shell B9003-2 B4431-2 0.12 0.60 60 B9003-3 A9406-2 0.13 0.60 50 51912(3490) 0.156 0.059
-50 101-122A 51912 (3536) 0.156 0.059
-70 Upper Shell 101-122B EAIB 0.02 0.98 10 (Gen)
Longitudinal Welds 101-122C IAGA 0.03 0.98
-30 BOHB 0.05 1.00 10 (Gen)
BAOED 0.02 1.00
-50 Upper Shell to 4P5174 (1122) 0.09 1.00
-50 Intermediate Shell 103-121 51922 (3489) 0.05 1.00
-56 (Gen)
Girth Weld AAGC 0.03 0.98
-70 KOIB 0.03 0.97
-60 B9011-1 2V2436-01-002 0.11 0.85 60\\C}
Inlet Nozzles B9011-2 2V2437 001 0.13 0.88 60\\C) (Gen)
B9011-3.
2V2445-02-003 0.13 0.84 70\\C}
4P5174 (1122) 0.09 1.00
-50 LOHB 0.03 1.03
-60 HABJC 0.02 1.02
-70 105-121A BABBD 0.02 1.04
-70 FABGC 0.03 1.02
-80 Inlet Nozzle Welds 105-121B EOBC 0.02 0.96
-60 105-121C FAAFC 0.07 1.04
-60 CCJC 0.02 0.99
-60 FAGB 0.02 1.06
-30 BAOED 0.02 1.00
-50 B9012-1 AV8080-2E9558 0.13 0.72
-10 Outlet Nozzles B9012-2 AV8120-2E9560 0.13 0.74
-10 B9012-3 AV8097 -2E9559 0.13 0.70
-10 BABBD 0.02 1.04
-70 FAAFC 0.07 1.04
-60 107-121A HAAEC 0.03 1.03
-80 Outlet Nozzle Welds 107-121B HABJC 0.02 1.02
-70 107-121C HAGB 0.02 1.04
-40 GACJC 0.03 1.00
-80 JAHB 0.03 0.97
-40 (a)
Materials information taken from Reference 13 (b)
Based on Reference 13, the generic Initial RT NDT values were determined in accordance with NUREG-0800 and the 10 CFR 50.61.
(c)
As described in Reference 16, the reactor vessel initial RTNDT values for the inlet nozzles are conservatively assigned values. The actual initial RT NDT values for the reactor vessel inlet nozzles are located in BVPS-2 UFSAR Table 5.3-1.
Beaver Valley Unit 2 5.2 - 22 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-8 (Page 1 of 1)
Summary of Adjusted Reference Temperature (ARTs) for 22 EFPy(a)
Material Description Method Used To 22 EFPY ART Calculate the CF(b) 1/4T ART CF) 3/4T ART CF)
Intermediate Shell Plate 89004-1 Position 1.1 139 128 Intermediate Shell Plate 89004-2 Position 1.1 115 103 Position 2.1 114 101 Lower Shell Plate 89005-1 Position 1.1 119 105 Lower Shell Plate 89005-2 Position 1.1 116 104 Vessel 8eltline Welds(C)
Position 1.1 47 29 Position 2.1
-2
-9 Notes:
(a)
Table updated to reflect Capsule X analysis per Reference 14; 1/4T and 3/4T ART values for 89004-1 will differ from as described on Figures 5.2-1 thru 5.2-6. See Section 5.2.1.1 for additional information.
(b)
Regulatory Guide 1.99, Revision 2.
(c)
All 8eltline Welds are from Heat #83642, Linde 0091, Flux Lot #3536.
8eaver Valley Unit 2 5.2 - 23 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 1 of 1)
Calculation of Adjusted Reference Temperatures (ARTs) for 22 EFPy(a)
PARAMETER VALUES Operating Time 22 EFPY Material - Intermediate Shell Plate 89004-1 89004-1 Location 1/4T 3/4T Chemistry Factor, CF CF) 40.5 40.5 Fluence, (f), (1019 n/cm2)(b) 1.515 0.589 Fluence Factor, FF 1.115 0.852
~RT NDT = CF x FFCF) 45.16 34.50 Intitial RT NDT, WF) 60 60 Margin, M(OF) 34 34 ART, per Regulatory Guide 1.99, Revision 2 139 128 Notes:
(a)
Table updated to reflect Capsule X analysis per Reference 14; 1/4T and 3/4T ART values for 89004-1 will differ from as described on Figures 5.2-1 thru 5.2-6. See Section 5.2.1.1 for additional information.
(b)
Fluence (f), is based upon fsurf (1019 n/cm2, E> 1.0 MeV) = 2.43 at 22 EFPY. The 8eaver Valley Unit 2 reactor vessel wall thickness is 7.875 inches at the beltline region.
8eaver Valley Unit 2 5.2 - 24 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Table 5.2-10 (Page 1 of 1)
Pressure and Temperature Limits Report 5.2 RT PTS Calculation for 8eltline Region Materials at Life Extension (54 EFPy)(a)
Material Heat Surface Neutron Material Description Fluence 10 Number (xi 019 n/cm2)
Intermediate Shell 89004-1 5.18 Plate Intermediate Shell 89004-2 5.18 Plate
-)- Using non-credible surveillance data(g) 5.18 Lower Shell Plate 89005-1 5.21 Lower Shell Plate 89005-2 5.21 Intermediate to Lower 101-171 83642 5.18 Shell Girth Weld
-)- Using credible surveillance data(g) 5.18 Intermediate Shell 101-124 83642 1.76 Longitudinal Welds A&8
-)- Using credible surveillance data(g) 1.76 Lower Shell 101-142 83642 1.77 Longitudinal Welds A&8
-)- Using credible surveillance data(g) 1.77 Notes:
(a)
Data obtained from Table 6-3 of Reference 13.
(b)
FF = fluence factor = f (0.28 - 0.1 log (I)).
(c)
Initial RT NDT values are measured values.
(d)
- FF.
(e)
M = 2 *(crl + cr~2)1/2.
(f)
RTpTS = Initial RTNDT + L1RTpTS + Margin.
Fluence Chemistry
- Factor, Factor FF(b)
CF) 1.4092 40.5 1.4092 37 1.4092 51.4 1.4104 51 1.4104 44 1.4092 34.4 1.4092 12.5 1.1554 34.4 1.1554 12.5 1.1569 34.4 1.1569 12.5 Initial L1RTpTS(d) au Margin(e)
RTpTS(f)
RTNDT(C) a~
CF)
(OF)
(OF)
(OF)
(OF)
(OF) 60 57.1 0
17 34 151.1 40 52.1 0
17 34 126.1 40 72.4 0
17 34 146.4 28 71.9 0
17 34 133.9 33 62.1 0
17 34 129.1
-30 48.5 0
24.2 48.5 67.0
-30 17.6 0
8.8 17.6 5.2
-30 39.7 0
19.9 39.7 49.5
-30 14.4 0
7.2 14.4
-1.1
-30 39.8 0
19.9 39.8 49.6
-30 14.5 0
7.2 14.5
-1.1 (g)
The 8VPS-2 surveillance weld metal is the same weld heat as the 8VPS-2 beltline welds (heat 83642). The 8VPS-2 surveillance weld data is credible; therefore, the reduced al1 term of 14°F was utilized for 8VPS-2 weld heat 83642. The 8VPS-2 surveillance plate material is representative of the 8VPS-2 intermediate shell plate 89004-2. The surveillance plate material is non-credible; therefore, the higher al1 term of 1 rF was utilized for 8VPS-2 plate 89004-2. The credibility evaluation conclusions are contained in Appendix A of Reference 13.
8eaver Valley Unit 2 5.2 - 25 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Table 5.2-11 (Page 1 of 3)
Pressure and Temperature Limits Report 5.2 RT PTS Calculation for Extended 8eltline Region Materials at Life Extension (54 EFPy)(a)
Surface Heat Number Neutron Fluence Chemistry Material Description MateriallD Fluence
- Factor, Factor (Lot Number)
(xi 019 FF(b)
(OF) n/cm2) 89003-1 A9406-1 0.515 0.8147 91.0 Upper Shell Plates 89003-2 84431-2 0.515 0.8147 83.0 89003-3 A9406-2 0.515 0.8147 91.0 51912 (3490) 0.515 0.8147 73.71 101-122A 51912 (3536) 0.515 0.8147 73.71 Upper Shell 101-1228 EAI8 0.515 0.8147 27.0 Longitudinal Welds IAGA 0.515 0.8147 41.0 101-122C 80H8 0.515 0.8147 68.0 8AOED 0.515 0.8147 27.0 4P5174 0.515 0.8147 122.0 Upper to Intermediate 103-121 51922 0.515 0.8147 68.0 Shell Girth Weld AAGC 0.515 0.8147 41.0 KOl8 0.515 0.8147 41.0 89011-1 2V2436-01-002 0.0298 0.2188 77.0 Inlet Nozzles 89011-2 2V2437 001 0.0298 0.2188 96.0 89011-3 2V2445-02-003 0.0298 0.2188 96.0 8eaver Valley Unit 2 5.2 - 26 Initial i1RTpTs(e) au RTNDT(C)
(OF)
(OF)
(OF) 50 74.1 0
60 67.6 0
50 74.1 0
-50 60.1 0
-70 60.1 0
10(d) 22.0 17
-30 33.4 0
10(d) 55.4 17
-50 22.0 0
-50 99.4 0
-56(d) 55.4 17
-70 33.4 0
-60 33.4 0
60(h) 16.8 0
60(d)(h) 21.0 17 70(h) 21.0 0
aLI.
(OF) 17 17 17 28 28 11.0 16.7 27.7 11.0 28 27.7 16.7 16.7 8.4 10.5 10.5 Margin(t)
RTpTs(g)
(OF)
(OF) 34 158.1 34 161.6 34 158.1 56 66.1 56 46.1 40.5 72.5 33.4 36.8 65.0 130.4 22.0
-6.0 56.0 105.4 65.0 64.4 33.4
-3.2 33.4 6.8 16.8 93.7 40.0 121.0 21.0 112.0 PTLR Revision 5 LRM Revision 76
Licensing Requirements Manual Table 5.2-11 (Page 2 of 3)
Pressure and Temperature Limits Report 5.2 RT PTS Calculation for Extended Beltline Region Materials at Life Extension (54 EFPy)(a)
I I
I Surface I I
I I
Material Heat Number Neutron Fluence Chemistry Description Material 10 (Lot Number)
Fluence
- Factor, Factor (xi 019 FF(b)
(OF) n/cm2) 4P5174 0.0298 0.2188 122.0 LOHB 0.0298 0.2188 41.0 HABJC 0.0298 0.2188 27.0 BABBD 0.0298 0.2188 27.0 105-121A FABGC 0.0298 0.2188 41.0 Inlet Nozzle 105-121 B Welds EOBC 0.0298 0.2188 27.0 105-121 C 0.2188 95.0 FAAFC 0.0298 CCJC 0.0298 0.2188 27.0 FAGB 0.0298 0.2188 27.0 BAOED 0.0298 0.2188 27.0 B9012-1 AV8080-2E9558 0.0151 0.1440 94.0 Outlet Nozzles B9012-2 AV8120-2E9560 0.0151 0.1440 94.5 B9012-3 AV8097 -2E9559 0.0151 0.1440 93.5 BABBD 0.0151 0.1440 27.0 107-121A FAAFC 0.0151 0.1440 95.0 Outlet Nozzle 107-121 B HAAEC 0.0151 0.1440 41.0 Welds HABJC 0.0151 0.1440 27.0 107-121C HAGB 0.0151 0.1440 27.0 GACJC 0.0151 0.1440 41.0 JAHB 0.0151 0.1440 41.0 Beaver Valley Unit 2 5.2 -27 I
I I
Initial
.6.RTpTs(e) au RTNDT(C)
(OF)
(OF)
(OF)
-50 26.7 0
-60 9.0 0
-70 5.9 0
-70 5.9 0
-80 9.0 0
-60 5.9 0
-60 20.8 0
-60 5.9 0
-30 5.9 0
-50 5.9 0
-10 13.5 0
-10 13.6 0
-10 13.5 0
-70 3.9 0
-60 13.7 0
-80 5.9 0
-70 3.9 0
-40 3.9 0
-80 5.9 0
-40 5.9 0
I a~
(OF) 13.3 4.5 3.0 3.0 4.5 3.0 10.4 3.0 3.0 3.0 6.8 6.8 6.7 1.9 6.8 3.0 1.9 1.9 3.0 3.0 I
Margin(f)
RTpTs(g)
(OF)
(OF) 26.7 3.4 9.0
-42.1 5.9
-58.2 5.9
-58.2 9.0
-62.1 5.9
-48.2 20.8
-18.4 5.9
-48.2 5.9
-18.2 5.9
-38.2 13.5 17.1 13.6 17.2 13.5 16.9 3.9
-62.2 13.7
-32.6 5.9
-68.2 3.9
-62.2 3.9
-32.2 5.9
-68.2 5.9
-28.2 PTLR Revision 5 LRM Revision 76 II
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 3 of 3)
RTpTS Calculation for Extended Beltline Region Materials at Life Extension (54 EFPy)(a)
Notes:
(a)
Data obtained from Table 6-4 of Reference 13.
(b)
FF = fluence factor = f(o.28-o.1Iog(f).
(c)
Initial RT NOT values are measured values, unless otherwise noted.
(d)
Initial RT NOT values are generic.
(e) ilRTpTS = CF
- FF.
(g)
RT PTS = Initial RT NOT + ilRT PTS + Margin.
(h)
As described in Reference 16, the reactor vessel initial RT NOT values for the inlet nozzles are conservatively assigned values. The actual initial RT NOT values for the reactor vessel inlet nozzles are located in BVPS-2 UFSAR Table 5.3-1.
Beaver Valley Unit 2 5.2 - 28 PTLR Revision 5 LRM Revision 76