L-14-141, Pressure and Temperature Limits Reports and Unit No. 2 Cycle 18 Core Operating Limits Report
| ML14133A107 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/12/2014 |
| From: | Emily Larson FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-14-141 | |
| Download: ML14133A107 (74) | |
Text
Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 FirstEnergy Nuclear Operating Company Eric A. Larson 724-682-5234 Site Vice President Fax: 724-643-8069 May 12, 2014 L-14-141 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 Unit Nos. 1 and 2 Pressure and Temperature Limits Reports and Unit No.2 Cycle 18 Core Operating Limits Report Pursuant to the requirements of BeaverValley Power Station, Unit Nos. 1 (BVPS-1) and 2 (BVPS-2) 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, Revision 7 and the BVPS-2 PTLR, Revision 7. Technical Specification 5.6.4.c requires, in part, that the PTLR be provided to the Nuclear Regulatory Commission upon issuance for any revision or supplement thereto. Revision 7 of the BVPS-1 and BVPS-2 PTLRs became effective on April 1.,8, 2014. The PTLRs were revised to add a note clarifying that the pressure limits continue to be met if the measured pressures are less than 0 psi g.
Pursuant to the requirements of BVPS-2 Technical Specification 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," FENOC hereby submits the BVPS-2 COLR for Cycle 18. Technical Specification 5.6.3.d requires, in part, that the COLR be provided to the Nuclear Regulatory Commission upon issuance for each reload cycle.
The Cycle 18 BVPS-2 COLR is effective May 2, 2014.
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.
Sincerely, zai-Eric A. Larson
Beaver Valley Power Station, Unit Nos. 1 and 2 L-14-141 Page 2
Enclosures:
A. Beaver Valley Power Station, Unit No. 1, Pressure and Temperature Limits Report, Revision 7 B. Beaver Valley Power Station, Unit No. 2, Pressure and Temperature Limits Report, Revision 7 C. Beaver Valley Power Station, Unit No. 2, Core Operating Limits Report, Cycle 18 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Enclosure A L-14-141 Beaver Valley Power Station, Unit No. 1 Pressure and Temperature Limits Report, Revision 7 (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 3.4.1 0 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 PTLR Revision 7 Beaver Valley Unit 1 5.2- i LRM Revision 84
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 (P!T)
Limits,
- 4. LCO 3.4.1 0 Pressurizer Safety Valves,
- 5. LCO 3.4.12 Overpressure Protection System (OPPS),
- 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 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."
5.2.1.1 RCS Pressure and Temperature (P!T) 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 any one hour period.
- b. A maximum cooldown of 1oooF in any one hour period, and PTLR Revision 7 Beaver Valley Unit 1 5.2- 1 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report
- 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.
The RCS P!T limits for heatup, leak testing, and criticality are specified by Figure 5.2-1 and Table 5.2-1. The RCS P!T 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 P!T 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 cooldown curves also include the effect of the reactor vessel flange.
The P!T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding P!T curve for heatup and cooldown.
Pressure-temperature limit curves shown in Figure 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 P!T 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 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 P!T 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 P!T limits continues to be the lower shell plate 86903-1 at 30 EFPY.
PTLR Revision 7 Beaver Valley Unit 1 5.2-2 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 10 19 n/cm 2 (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 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 P!T 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 P!T limit curves remain bounding for 30 EFPY, the 30 EFPY P!T limits remain valid as documented in Reference 12.
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 P!T limits which were established in accordance with 10 CFR 50, Appendix G Without allowancedor 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.
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. 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 347°F.
The calculated enable temperature is based on either a RCS temperature of less than 200°F or materials concerns (reactor vessel metal temperature less than 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 318°F.
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 7 Beaver Valley Unit 1 5.2-3 LRM Revision 84
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/enabledY 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 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 Surveillanc*e Program." The material test requirements and the acce'ptance standards utilize the reference nil-ductility temperature, RT NoT, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RT NOT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.
Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 1. This commitment is a condition of license Amendment 256 and will remain in effect until the NRC staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.
PTLR Revision 7 Beaver Valley Unit 1 5.2-4 LRM Revision 84
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 Prr 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 RTPTs values for the beltline materials at 50 *EFPY. -
- Table 5.2-10, taken from Reference 11, provides RTprs values for the extended beltline materials at 50 EFPY.
- Table 5.2-11, Reactor Vessel Toughness Data (Unirradiated)
PTLR Revision 7 Beaver Valley Unit 1 5.2-5 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 al., January 1996.
- 2. Deleted
- 3. WCAP-15571, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," C. Brown, et. al., November 2000.
- 4. WCAP-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 al., November 2000.
- 6. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 7. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 8. *Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
- 9. Deleted
- 10. FirstEnergy Nuclea*r 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 ofLTOPS SetpointAnalysis Report, July 26, 2007.
- 14. Westinghouse Calculation CN-SCS-07-27, Revision 0, LTOPS 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.
PTLR Revision 7 Beaver Valley Unit 1 5.2-6 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE B6903-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 245.7°F 3/4T, 207.6°F 2500 2250 I Leak Test Limit I
........ I I I 2000 7I I 1750 I ~
IHeatup Rate I IUnacceptable I 100°F/Hr
~
v I I t --
~ Operation 1500
~
~Critical Limit I i~ 1250 100°F/Hr
~ 1000 II I i I Acceptable Operation I
J v
750
_-/
500 Boltup Criticality Limit based on Temperature inservice hydrostatic test 250 60°F temperature (302°F) for the _
service period up to 30 EFPY i/
0 0 50 100 150 200 250 300 350 400 450 500 550 INDICATED TEMPERATURE (°F)
Figure 5.2-1 (Page 1 of 1)
Reactor Coolant System Heatup Limitations Applicable for the Fi~st 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 1 5.2-7 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE B6903-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 245.rF 3/4T, 207.6°F 2600 I
2260 2000 I
~,
1760 IUnacceptable Operation I
! 1600 i~ I v
1260 I Acceptable Operation I
~ 1000 A 760 ./. ~
~
7 Cooldown Rates 0°F/Hr (steady-state) 600
~.---
20°F/Hr 40°F/Hr
....... 60°F/Hr 260 Boltup r--- - 1 00°F/Hr
/
v Temperature 60°F 0
0 60 100 160 200 260 300 360 400 450 500 650 INDICATED TEMPERATURE (°F)
Figure 5.2-2 (Page 1 of 1)
Reactor Coolant System Cooldown Limitations Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 1 5.2-8 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 2500 2000 6 "1500 /
Ci5 e:..
w
/
c::
- J ~
__..... ~
CIJ CIJ w
. g: . "1000 ..-
I 500 "I I "
I I
0 I 50 60 70 80 90 "100 i iO "120 TEMPERATURE (°F)
Figure 5.2-3 (Page 1 of 1)
Isolated Loop Pressure- Temperature Limit Curve (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 1 5.2-9 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 1 of 1)
Heatup Curve Data Points for 30 EFPY (LCO 3.4.3) 100°F/hr 100°F/hr 100°F/hr Heatup Heatup Criticality T p T p T p (oF) (psig) (oF) (psig) (oF) (psig) 60 0 245 840 302 0 60 554 250 876 302 981 65 554 255 917 305 1010 70 554 260 961 310 1064 75 554 265 1010 315 1124 80 554 270 1064 320 1189 85 554 275 1124 325 1262 90 554 280 1189 330 1342 95 554 285 1262 335 1431 100 554 290 1342 340 1528.
105 554 295 1431 345 1636 110 554 300 1528 350 1754 115 554 305 1636 355 1885 120 554 310 1754 360 2029 125 554 315 1885 365 2151 130 554 320 2029 370 2282 135 .. 554 325 2151 375 2426 140 555 330 2282 376.8 2485 145 557 335 2426 150 560 336.8 2485 155 563 160 567 165 573 170 579 175 585 180 593 185 602 190 613 195 624 200 637 205 651 210 667 215 685 220 705 225 727 230 751 235 778 240 807 284 302 Leak Test Limit 2000 2485 PTLR Revision 7 Beaver Valley Unit 1 5.2- 10 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 1 of 2)
Cooldown Curve Data Points for 30 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 60 606 60 563 60 518 60 425 65 621 65 607 65 563 65 519 65 426 70 621 70 608 70 564 70 519 70 427 75 621 75 609 75 565 75 520 75 428 80 621 80 611 80 567 80 522 80 429 85 621 85 612 85 568 85 523 85 430 90 621 90 614 90 570 90 525 90 432 95 621 95 616 95 571 95 526 95 433 100 621 100 618 100 574 100 528 100 435 105 621 105 620 105 576 105 531 105 438 110 621 110 621 110 578 110 533 110 441 115 621 115 621 115 581 115 536 115 444 120 621 120 621 120 585 120 540 120 448 125 621 125 621 125 588 125 544 125 452 130 621 130 621 130 592 130 548 130 457 135 621 135 621 135 597 135 553 135 462 140 621 140 621 140 602 140 558 140 468 145 621 145 621 145 607 145 564 145 475 150 621 150 621 150 614 150 571 150 483 155 621 155 621 155 621 155 578 155 491 160 621 160 621 160 621 160 586 160 501 165 621 165 621 165 621 165 595 165 512 170 621 170 621 170 621 170 606 170 524 175 621 175 621 175 621 175 617 175 537 180 621 180 621 180 621 180 621 180 552 180 747 180 708 180 669 180 630 185 569 185 758 185 720 185 682 185 644 190 588 190 771 190 733 190 696 190 660 195 608 195 784 195 748 195 713 195 677 200 631 200 800 200 765 200 730 200 697 205 657 205 816 205 783 205 750 205 718 210 685 210 835 210 803 210 772 210 742 215 717 215 856 215 825 215 796 215 768 220 752 220 878 220 850 220 823 220 797 225 791 225 903 225 877 225 853 225 830 230 835 230 931 230 908 230 886 230 866 235 883 235 962 235 941 235 922 235 906 240 936 240 996 240 978 240 962 240 950 245 995 245 1033 245 1019 245 1007 245 999 250 1053 250 1075 250 1064 250 1056 250 1053 255 1111 PTLR Revision 7 Beaver Valley Unit 1 5.2- 11 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 2 of 2)
Cooldown Curve Data Points for 30 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 1121 255 1114 255 1111 255 1111 260 1169 260 1171 260 1169 260 1169 260 1169 265 1227 265 1227 265 1227 265 1227 265 1227 270 1289 270 1289 270 1289 270 1289 270 1289 275 1357 275 1357 275 1357 275 1357 275 1357 280 1433 280 1433 280 1433 280 1433 280 1433 285 1516 285 1516 285 1516 285 1516 285 1516 290 1608 290 1608 290 1608 290 1608 290 1608 295 1710 295 1710 295 1710 295 1710 295 1710 300 1823 300 1823 300 1823 300 1823 300 1823 305 1947 305 1947 305 1947 305 1947 305 1947 310 2085 310 2085 310 2085 310 2085 310 2085 315 2237 315 2237 315 2237 315 2237 315 2237 320 2405 320 2405 320 2405 320 2405 320 2405 322.1 2485 322.1 2485 322.1 2485 322.1 2485 322.1 2485 PTLR Revision 7 Beaver Valley Unit 1 5.2- 12 LRM Revision 84
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 34rF PORV Setpoint ::; 397 psig PTLR Revision 7 Be.aver Valley Unit 1 5.2- 13 LRM Revision 84
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 t<aJ FF(bJ ~RTNoT(c) FF *~RTNOT FF 2 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(dJ (Longitudinal) w 0.930 0.980 148.52 145.50 0.960 y 2.05 1.196 142.18 169.98 1.429 v 0.299 0.669 137.81 92.25 0.448 Lower Shell Plate u 0.604 0.859 131.84 113.23 0.738 B6903-1(dJ 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 = ~(FF * ~RT NOT) + ~(FF 2 ) = (1 085.02) + (7.150) = 151.8°F(e) 169.30 v 0.299 0.669 (159. 72) 113.33 0.448 176.30 Beaver Valley u 0.604 0.859 (166.32) 151.41 0.738 Unit 1 198.99 Surveillance w 0.930 0.980 (187.73) 194.95 0.960 Weld Metal(dJ (Heat# 305424) y 190.47 2.05 1.196 227.72 1.429 (179.69)
SUM: 687.41 3.575 CF = ~(FF * ~RT NOT) + ~(FF 2 ) = (687 .41) + (3.575) = 192.3°F(e)
Notes:
(a) f = 2 Calculated surveillance capsule neutron fluence (x 10 19 n/cm , E > 1.0 MeV).
The surveillance capsule fluence results are contained in Table 8-1 of Reference 11.
(b) FF = fluencefactor = f(o. 2s-o. 1 'lagfl.
(c) ~RT NOT values are the measured 30 ft-lb 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.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 14 LRM Revision 84
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 f<bl FF(cl ~RTNoT(d) FF *~RTNoT FF 2 82.65 9r 0.5174 0.816 67.44 0.666 (72.34)
Weld Metal 81.08 Heat# 90136(e) 104° 0.7885 0.933 75.68 0.871 (67.4)
(St. Lucie Unit 1) 83.77 284° 1.243 1.061 88.85 1.125 (68.0)
SUM: 231.97 2.662 2
- 11RTNOT) + L:(FF ) = (231.97) + (2.662) = 87.1°F(g) 197.30 W-225 0.488 0.800 157.83 0.640 (21 0)
Weld Metal 218.30 Heat# 305414(1) W-265 0.847 0.953 208.13 0.909 (225)
(Fort Calhoun 215.90 Unit 1) W-275 1.54 1.119 241.68 1.253 (219)
SUM: 607.64 2.802 CF = L:(FF * ~RTNoT) + L:(FF 2) = (607.64) + (2.802) = 216.9°F 19l 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)." . -
19 2 (b) f =calculated surveillance capsule fluence values (x 10 n/cm , E > 1.0 MeV). The surveillance capsule"fluence results are contained in Tables A-3 and A-5 of Reference 11.
(c) FF = fluencefactor = f( 0 .28 " 0 *1 'Jogfl.
(d) t1RT NoT values are the measured 30ft-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 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 t1RT NOT values for the weld data are adjusted by a ratio of 1.17. The Fort Calhoun ~RT 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.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 15 LRM Revision 84
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<aJ(bl Irradiated Capsule fcl Cu Ni 2 ~RTNDT(d)
Material Capsule Temperature (x1 0 19 n/cm ,
(wt. %) (wt. %) (oF)
(oF) E > 1.0 MeV)
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)
(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 = calculate:d surveillance capsule fluence values.
(d) ~RT Nor values are the measured 30 ft-lb shift values from Tables A-3 and A-5 of Reference 11.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 16 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-5 (Page 1 of 1)
Reactor Vessel Beltline Material Properties Chemistry Initial Cu Ni RT NDT(a)
Material Description Factor (wt. %) (wt. %)
(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&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 Nor values for the plates are based on measured data while the weld values are generic.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 17 LRM Revision 84
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<eJ 30 EFPY Material Description 1/4T ART<aJ 3/4T ART<aJ (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 S/C Data<bJ 245.7 207.6 Intermediate Shell Longitudinal Weld 19-714A/B 161.9 115.4
- UsinQ SIC Data<bl 159.6 113.8 Intermediate to Lower Shell Circ. Weld 11-714 163.4 131.7
- Using S/C Data (c) 93.0 71.4 Lower Shell Longitudinal Weld 20-714A/B 176.8 125.8
- Using S/C Data<dJ
- 187.5 ..
133.2 Notes:
(a) ART= I + LlRT Nor+ M.
(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full cr 6 .)
(c) Based on St. Lucie Unit 1 surveillance data. (Data credible. ART calculated with a reduced cr 6 .)
(d) Base~ on Fort Calhoun Unit 1 surveillance data. (Data not credi.ble. ART calculated with a full cr 6 .)
(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.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 18 LRM Revision 84
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(°F) 3/4T ART(°F)
Chemistry Factor, CF (°F) 149.2 149.2 Fluence (f), n/cm 2 (E>1.0 Mev)(a) 2.4194 X 10 19 9.404 X 10 18 Fluence Factor, FF 1.238 .9828
~RT NDT = CF X FF(°F)(c) 184.7 (C) 146.6 Initial RTNor, I(OF)<aJ 27 27
~J.arQin, M(°F) 34(C) 34 ART= I+(CF*FF)+M, °F(llJ per RG 1.99, Revision 2 245.7 207.6 Notes:
(a) Initial RT Nor 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 crll.)
(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.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 19 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-8 (Page 1 of 1)
Reactor Vessel Extended Beltline Material Properties(al Initial Material Description Material Heat Number Wt% Wt% RT NOT(c)
ID (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 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 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 1-717D DBIJ 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)
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:
(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).
PTLR Revision 7 Beaver Valley Unit 1 5.2- 20 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-9 (Page 1 of 2)
RTPTs Calculation for Beltline Region Materials at Life Extension (50 EFPY)Cal Surface Fluence Chemistry Initial Material Heat LlRTPTS(d) cru a,.,. MarginCel RTPTS(f)
Material Description Fluence Factor, Factor RTNoT(c)
ID Number (oF) (oF) (oF) (oF) (oF)
(x1 0 19 n/cm 2 ) FFCbl CF) (oF)
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
------------- ---------------- ------------ -------------- -------- --=j7<9f- ------------- -----------
---+ Using non-credible suNeillance dataC l 9 5.57 1.4231 151.8 27 216.0 0 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~;ediblesuNeillance data<11r---------- ------------------- ------------- ---------------- ------------ -------------- --------
4<W ------------- -----------
5.55 1.4225 87.1 -56 123.9 17 44.0 111.9 Intermediate Shell 19-714
____h<~_mgitudinal Weld A&B 305424 1.08 1.0224 191.7 -56 196.0 17 28 65.5 205.5
---+ Using non-credible suNeillance dataC 9l *
------------- ---------------- ----------- -------------- -------- -28<9)- ------------- -----------
1.08 1.0224 192.3 -56 196.6 17 65.5 206.1 Lower Shell Longitudinal 20-714 Weld 305414 1.09 1.0241 210.5 -56 215.6 17 28 65.5 225.1 A&B
---+ Using non-credible suNeillance datal!J
-28Cif- ------------- -----------
1.09 1.0241 216.9 -56 222.1 17 65.5 231.6 Notes:
(a) Data obtained from Table 6-3 of Reference 11.
(b) FF = fluence factor= fC 0 0* 101 og(fll.
(c) Initial RT NoT values are measured values with the exception of the vessel. welds.
- FF.
(e) M =2 *(au 2
+ a,.,.2 ) 112 *
(f) RT PTs = Initial RTNOT+ LlRT PTs + Margin.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 21 LRM Revision 84
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 ol!. 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 ol!. term of 1rF was utilized for BVPS-1 plate 86903-1. 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 ol!. 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 ol!. term of 28°F was utilized for BVPS-1 weld heat 305414.
The credibility evaluation conclu'sions are contained in Appendix A of Reference 11.
PTLR Revision 7 Beaver Valley Unit 1 5.2-22 LRM Revision 84
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-10 (Page 1 of 2)
RTPTS Calculation for Extended Beltline Region Materials at Life Extension (50 EFPY)<al Surface Fluence Chemistry Initial Material Material Heat Number ~RTPTS(d) cru a~ Margin<el RTPTS(f)
Fluence Factor, Factor RTNDT(c)
Description ID (Lot Number) (oF) (oF) (oF) (oF) (oF)
(x1 0 19 n/cm 2) FF(bl CF) COF)
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 Shell Girth Weld . ( & 3958 )
t-*.---+ Using non-credible surveiil~~~~-datar9r--- 0.625 0.8685 ---28\gf-- ------------- ----------------
216.9 -56 188.4 17 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 Shell Girth Weld EODJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 .
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 D81J 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 108J 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 1-717 c 51.0 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 11.
PTLR Revision 7 Beaver Valley Unit 1 5.2-23 LRM Revision 84
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 .28 - 0
- 101 og(fll.
(c) Initial RT NOT value for the upper shell forging is a measured value. All other values are generic.
(d) ~RTPTs = CF
- FF.
(e) M = 2 *(cru 2 + crll2) 112 .
(f) RT PTs = Initial RT NoT + ~RT 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 afl term of 28°F was utilized for BVPS-1 weld heat 305414. The credibility evaluation conclusions are contained in Appendix A of Reference 11.
PTLR Revision 7 Beaver Valley Unit 1 5.2- 24 LRM Revision 84
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 RTNDT 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 ---
Se~.
Closure Head ZV3758 --- A508 CL. 2 .08 --- .007 60* 60* >100 ---
Flan~e Vessel Flan~e ZV3661 --- A508 CL. 2 .12 --- .010 60* 60* 166 ---
Inlet Nozzle 9-5443 --- A508 CL. 2 .10 --- .008 60* 60* 82.5 ---
Inlet Nozzle 9-5460 --- A508 CL. 2 .10 --- .010 60* 60* 94 ---
Inlet Nozzle 9-5712 --- A508 CL. 2 .08 --- .007 60* 60* 97 ---
Outlet Nozzle 9-5415 --- A508 CL. 2 --- --- .008 60* 60* 97 ---
Outlet Nozzle 9-5415 --- A508 CL. 2 --- --- .007 60* 60* 112.5. ---
Outlet Nozzle 9-5444 --- A508 CL. 2 .09 --- .007 60* 60* 103 ---
Upper Shell 123V339 --- A508 CL. 2 --- -.:.- .010 40 40* 155 ---
Inter Shell C4381-2 B6607-2 A533B CL. 1 .14 .62 .015 -10 73 123 82.5 Inter Shell C4381-1 B6607-1 A533B CL. 1 .14 .62 .015 -10 43 128.5 90 Lower Shell C6317-1 B6903-1 A533B CL. 1 .20 .54 .010 -50 27 134 80 Lower Shell C6293-2 B7203-2 A533B CL. 1 .14 .57 .015 -20 20 129.5 83.5 Trans Rinl:J 123V223 --- A508 CL. 2 --- --- " --- 30 30* 143 ---
Bottom Hd Se~ 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 Long. Weld Weld HAZ --- --- --- -40 -40 --- 136.5
- Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Direction NMWD- Normal to Major Working Direction Note: For evaluation of 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 7 Beaver Valley Unit 1 5.2- 25 LRM Revision 84
Enclosure B L-14-141 Beaver Valley Power Station, Unit No. 2 Pressure and Temperature Limits Report, Revision 7 (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 3.4.1 0 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/k N/A 5.2-3 LR 3.4.6 N/A N/A 5.2-3 PTLR Revision 7 Beaver Valley Unit 2 5.2- i LRM Revision 79
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 (P!T)
Limits,
- 4. LCO 3.4.1 0 Pressurizer Safety Valves,
- 5. LCO 3.4.12 Overpressure Protection System (OPPS),
- 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 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 u~ed 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."
5.2.1.1 RCS Pressure and Temperature (P!T) Limits (LCO 3.4.3)
The RCS temperature rate-of-change limits defined in Reference 14 are:
- a. A maximum heatup of 60°F in any one hour period.
- b. A maximum cooldown of 100°F in any one hour period, and PTLR Revision 7 Beaver Valley Unit 2 5.2- 1 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report
- 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.
The RCS P!T limits for heatup, leak testing, and criticality are specified by Figure 5.2-1 and Table 5.2-1. The RCS P!T 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 14. Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 5.2.1, the RCS P/T 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 P!T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding P!T curve for heatup and cooldown.
The pressure-temperature limit curve shown in Figure 5.2-7 was 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 P!T combination is located in the region to the right and below the operating limits provided in Figures 5.2-1, 5.2-2, 5.2-3, 5.2-4, 5.2-5, 5.2-6, and 5.2-7.
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 P!T 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 BVPS-2 P!T limits continues to be the intermediate shell plate 89004-1 at 30 EFPY.
PTLR Revision 7 Beaver Valley Unit 2 5.2-2 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 19 2 value at 30 EFPY is 3.03 x 10 n/cm (E > 1.0 MeV). This value was calculated by interpolating the fluence at the oo azimuthal position for BVPS-2 from the end of Cycle 15 to the fluence value at the future projection out to 32 EFPY. The 19 fluence of 3.39 x 10 n/cm 2 (E > 1.0 MeV) used to develop the 30 EFPY PIT limit curves generated as a result of the Capsule X analysis (Reference 12), is more 19 2 conservative than the updated fluence of 3.03 x 10 n/cm (E > 1.0 MeV).
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 that varies with RCS temperature and which does not exceed the limits in Figure 5.2-8 (Reference 9). 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 137°F. 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 P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 5.2.1. The PORV lift setting shown in Figure 5.2-8 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. 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.
PTLR Revision 7 Beaver Valley Unit 2 5.2-3 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 15, and are based upon analysis of Capsule X. 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 30 EFPY. As a result, Tables 5.2-3 and 5.2-4 and Figure 5.2-8 remain valid for Capsule X up to 30 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.
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.
PTLR Revision 7 Beaver Valley Unit 2 5.2-4 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report 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 4 and 13) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RT NoT, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RT NOT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.
Reference 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 P!T 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 4-7 and 4-13 of Reference 14, provides a summary of the Adjusted Reference Temperature (ARTs) for 30 EFPY.
PTLR Revision 7 Beaver Valley Unit 2 5.2-5 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report Table 5.2-9, taken from Tables 4-7 and 4-8 of Reference 14, shows the calculation of ARTs for 30 EFPY.
Table 5.2-10, taken from Table 6-3 of Reference 13, provides RTprs values for the Beltline Region Materials at 54 EFPY.
Table 5.2-11, taken from Table 6-4 of Reference 13, provides RT Prs values for the Extended Beltline Region Materials at 54 EFPY.
Note that Tables 5.2-5, 5.2-8 and 5.2-9 reflect Capsule X analysis and fluence data.
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. (Deleted)
- 3. (Deleted)
- 4. WCAP-9615, Revision 1, "Duquesne Light Company, Beaver Valley Unit No.2 Reactor Vessel Radiation Surveillance Program," P. A. Peter, June 1995.
- 5. WCAP-15676, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2,~* J. H. Ledger, August 2001.
- 6. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 7. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 8. Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
- 9. FENOC Calculation No. 10080-SP-2RCS-006, Revision 4, Addendum 1, "BV-2 LTOPS Setpoint Evaluation Capsule W for 22 EFPY."
- 10. FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
PTLR Revision 7 Beaver Valley Unit 2 5.2-6 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.2 Pressure and Temperature Limits Report
- 11. (Deleted)
- 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."
PTLR Revision 7 Beaver Valley Unit 2 5.2-7 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVES APPLICABLE FOR HEATUP RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 I I Leak Test Limi!J-.
2250 2000 Unacceptable 1 Operation I IAcceptable Operation 5
(/)
e:.
1750 w 1500 u Heatup rate to 60°F I Hr.
I vI I
0::
(/)
(/) Criticality Limit for 60°F/Hrj w 1250 II 0::
a.
c w
1- 1000
<C
(.)
5 z
/'I 750 I Criticality-Lim it based on inservice 500 . + hydrostatic test tern peratue (199°~) for
--i Boltup Temperature1 I the service period up to 30 EFPY 250 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-1 (Page 1 of 1)
Reactor Coolant System Heatup Limitations* Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2-8 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 0°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 2250 I I
2000 IUnacceptablel II
~
Operation IAcceptable I Operation
...... 1750 I Q
en e:. Ir---
rv w 1500 0::
- l en 1---.... -j Cooldown Rate ooF/Hrj en w 1250 0::
a.
c w
1-
<( 1000 J
(.)
c I
~
750 500 I.._ H Boltup Temperature 250 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-2 (Page 1 of 1)
Reactor Coolant System Cooldown (steady state- 0°F/Hr.)
Limitations Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2-9 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 20°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 2250 I
2000 Unacceptable I Operation IAcceptable' Operation 1750 J
~
en e:.
w 1500 I ~
~~
0::
- J en en w 1250 0::
a.
c II N_Cooldown Rate 20°F/Hrl w
1- 1000 I
~
{)
25 I
~
750 500
~ Boltup
-+-- Temperature 250 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-3 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 20°FH:Ir.)
Limitations Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 10 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 40°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 2250 L 2000 Unacceptable I II Operation lAcceptablel Operation J
.-.. 1750
(!)
U5 e:.
w 1500 I~
~
- l C/)
C/)
w 1250
~
a..
c Iv ~ NCooldown Rate 40°F/Hrj w
1- 1000 I
<(
(,)
i5
~
I 750 500
.,....__ ~ Boltup Temperature l
250 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-4 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 40°F/Hr.)
Limitations Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 11 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 2250 I 2000 Unacceptable Operation IAcceptable j Operation
§' 1750 en e::. ~
w 1500 I ~ NCooldown Rate 60°F/Hrl 0::
- J en en w 1250 0::
a..
c w
1- 1000 J
v
<(
(.)
25 z . 750 I
500 250
-H Boltup Temperature
!I 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-5 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 60°F/Hr.)
Limitations Applicable for the Firs~ 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 12 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 30 EFPY: 1/4T, 143°F 3/4T, 132°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 30 EFPY.
2500 2250 I 2000 Unacceptable Operation IAcceptable Operation I
(5' 1750
(/)
~
~ 1500
/ ~
~~
- l
(/)
(/)
I 'j Cooldown Rate 100°F/HrJ
~ 1250 a..
0 w
1- 1000 J v
(3 5
z 750 IJ 500
/ ,,
250
-H Boltup Temperature
,I 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (°F)
Figure 5.2-6 (Page 1 of 1)
Reactor Coolant System Cooldown (up to 100°F/Hr.)
Limitations Applicable for the First 30 EFPY (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 13 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 2500
! I I I i
I I
2000
-0 /
v U) 1500 w
0..
a:
- > ~
~
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0..
1000 t.------
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500 0
0 50 60 70 80 90 100 110 120 TEMPERATURE (°F)
Figure 5.2-7 (Page 1 of 1)
Isolated Loop Pressure- Temperature Limit Curve (LCO 3.4.3)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 14 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 See Table 5.2-4 for RCP restrictions.
725 700 675 650 l Unacceptable L Operation I 625 1Acceptable I 600 1 Operation 1 575 550 J
525
~l -
500 v
I 475 450 50 75 100 125 150 175 200 225 250 275 300 325 350 0
TRro*AUCTIONEERED LOW-MEASURED RCS TEMPERATURE ( F)
Figure 5.2-8 (Page 1 of 1)
Maximum Allowable Nominal PORV Setpoint for the Overpressure Protection System (LCO 3.4.12)
PTLR Revision 7 Beaver Valley Unit 2 5.2- 15 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 (Page 1 of 1)
Heatup Curve Data Points for 30 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) (psig) eF) (psi g) 60 0 199 0 181 2000 60 621 199 621 199 2485 65 621 199 621 70 621 199 621 75 621 199 621 80 621 199 621 85 621 199 621 90 621 199 621 95 621 199 621 100 621 199 621 105 621 199 777 110 621 199 793 115 621 199 813 120 621 199 835 120 621 199 861 120 777 199 889 125 793 199 921 ..
130 813 199 957 135 835 200 996 140 861 205 1040 145 889 210 1089 150 921 215 1143 155 957 220 1203 160 996 225 1269 -
165 1040 230 1342 170 1089 235 1423 175 1143 240 1512 180 1203 245 1611 185 1269 250 1719 190 1342 255 1840 195 1423 260 1972 200 1512 265 2118 205 1611 270 2280 210 1719 275 2458 215 1840 220 1972 225 2118 230 2280 235 2458 PTLR Revision 7 Beaver Valley Unit 2 5.2- 16 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-2 (Page 1 of 1)
Cooldown Curve Data Points for 30 EFPY (LCO 3.4.3) oaF/HR 20°F/HR 40°F/HR 60°F/HR 100°F/HR Temp. Press. Press. Press. Press. Press.
(oF) (psig) (psig) (psig) (psig) (psi g) 60 0 0 0 0 0 60 621 621 621 602 525 65 621 621 621 612 536 70 621 621 621 621 548 75 621 621 621 621 562 80 621 621 621 621 578 85 621 621 621 621 595 90 621 621 621 621 614 95 621 621 621 621 621 100 621 621 621 621 621 105 621 621 621 621 621 110 621 621 621 621 621 115 621 621 621 621 621 120 621 621 621 621 621 120 621 621 621 621 621 120 892 867 844 822 783 125 918 896 875 855 823 130 947 927 909 893 867 135 980 962 947 934 917 140 1016 1001 989 980 971 145 1055 1044 1036 1031 1031 150 1099 1092 1087 1087 1087 155 1147 1144 1144 '1144 1144 160 1201 1201 1201 1201 1201 165 1260 1260 1260 1260 1260 170 1325 1325 1325 1325 1325 175 1397 1397 1397 1397 1397 180 1477 1477 1477 1477 1477 185 1565 1565 1565 1565 1565 190 1662 1662 1662 1662 1662 195 1770 1770 1770 1770 1770 200 1888 1888 1888 1888 1888 205 2020 2020 2020 2020 2020 210 2165 2165 2165 2165 2165 215 2325 2325 2325 2325 2325 PTLR Revision 7 Beaver Valley Unit 2 5.2- 17 LRM Revision 79
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 PTLR Revision 7 Beaver Valley Unit 2 5.2- 18 LRM Revision 79
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
< 13rF 0-2
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 fal FF(bl D.RT NOT(c) FF*D.RTNoT FF 2 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(dl w 3.61 1.334 71.0 94.68 1.778 (Longitudinal) 139.65 2.031 X 5.63 1.425 98.0 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(dl w 3.61 1.334 63.4 84.55 1.778 (Transverse) 148.34 2.031 X 5.63 1.425 104.1 SUM: 631.87 12.284 CF =L(FF
- D.RTNoT) + 2 L(FF ) =(631.87) + (12.284) =51.4°F Beaver Valley u 0.615 0.864 4.1 3.54 0.746 Unit 2 Surveillance v 2.64 1.260 25.7 32.37 1.587 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: . 76.55 6.142 CF =L(FF
- D.RTNOT)+ L(FF ) =(76.55) + (6.142) =12.5°F 2
- Notes:
19 2 (a) f = calculated surveillance capsule neutron fluence (x 10 n/cm , E > 1.0 MeV). The surveillance capsule fluence results are contained in Table 8-1 of Reference 13.
(b) FF = fluencefactor = f( 0 28 0 1 1 9
. " * ' a fl.
(c) D.RT NOT values are the measured 30 ft~lb shift values. The BVPS-2 D.RTNOT 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.
PTLR Revision 7 Beaver Valley Unit 2 5.2-20 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-6 (Page 1 of 1)
Reactor Vessel Beltline Material Properties Cu Ni Initial RT NoT Material (F)(a)
(wt%) (wt%)
Closure Head Flange B9002-1 0.06(b) 0.74 -10 Vessel Flange B9001-1 0.06(b) 0.73 0 Intermediate Shell Plate B9004-1 0.065 0.55 60 Intermediate Shell Plate B9004-2 0.06 0.57 40 Lower Shell Plate B9005-1 0.08 0.58 28 Lower Shell Plate B9005-2 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 & B (Heat 83642) 0.046 0.086 -30 Lower Longitudinal Weld 101-142 A & B (Heat 83642) 0.046 0.086 -30 Plate Surveillance Material B9004-2 0.06 0.57 40 Surveillance Weld (Heat 83642) 0.065 0.065 -3Q(c)
(a) The initial RT NoT values for all of the beltline materials are based on measured data.
(b) According to the BVPS-2 reactor vessel CMTRs and MISC-PENG-ER-021, the material for the closure head flange (B9002-1) and vessel flange (B9001-1) forgings are ASTM A508 Class 2. The ASTM A508 material specification does not require analysis of copper content. The importance of copper content in the irradiation embrittlement of ferritic pressu-re vessel steel was not recognized or regulated by the NRC or nuclear steam supply system (NSSS) vendors when the BVPS-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 cop"per) were reported in MISC-PENGER-021. The copper values reported for both the closure head flange (B9002-1) and the vessel flange (B9001-1) was 0.06%.
(c) The initial RTNOT value is determined in accordance with the requirements of Subparagraph NB-2331 of Section Ill of the ASME B&PV Code, as specified by Paragraph II - D of 10 CFR Part 50, Appendix G. These fracture toughness requirements are also summarized in Branch Technical Position MTEB 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 ofWCAP-9615, the initial RTNoT value is concluded to be equal toT NoT (i.e., -30.0°F).
PTLR Revision 7 Beaver Valley Unit 2 5.2- 21 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-7 (Page 1 of 1)
Reactor Vessel Extended Beltline Material Properties (aJ Initial Material Material Wt% Wt%
Heat Number RTNW Description 10 Cu Ni (oF)' )
89003-1 A9406-1 0.13 0.60 50 Upper Shell 89003-2 84431-2 0.12 0.60 60 89003-3 A9406-2 0.13 0.60 50 51912 (3490) 0.156 0.059 -50 51912 (3536) 0.156 0.059 -70 101-122A Upper Shell EAIB 0.02 0.98 10(Gen) 101-1228 Longitudinal Welds lAG A 0.03 0.98 -30 101-122C BOHB 0.05 1.00 10(Gen)
BAOED 0.02 1.00 -50 4P5174 (1122) 0.09 1.00 -50 Upper Shell to -56 (Gen) 51922 (3489) 0.05 1.00 Intermediate Shell 103-121 AAGC 0.03 0.98 -70 Girth Weld KOIB 0.03 0.97 -60 89011-1 2V2436-01-002 0.11 0.85 60(C)
Inlet Nozzles 89011-2 2V2437-02-001 0.13 0.88 6o<cJ (Gen) 89011-3 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 BABBD 0.02 1.04 -70 105-121A FABGC 0.03 1.02 -80 Inlet Nozzle Welds 105-1218 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 89012-1 AV8080-2E9558 0.13 0.72 -10 Outlet Nozzles 89012-2 AV8120-2E9560 0.13 o:74 -10 89012-3 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 107-1218 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 NOT values were determined in accordance with NUREG-0800 and the 10 CFR 50.61.
(c) 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.
PTLR Revision 7 Beaver Valley Unit 2 5.2- 22 LRM Revision 79
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 30 EFPY(a)
Method Used To Material Description 30 EFPY ART Calculate the CF(b) 1/4T ART (°F) 3/4T ART (°F)
Intermediate Shell Plate 89004-1 Position 1.1 143 132 Position 1.1 119 109 Intermediate Shell Plate 89004-2 Position 2.1 119 106 Lower Shell Plate 89005-1 Position 1.1 123 110 Lower Shell Plate 89005-2 Position 1.1 120 109 Position 1.1 53 35 Vessel Beltline Welds(c)
Position 2.1 0 -6 (a) Table reflects Capsule X analysis per Reference 14.
(b) Regulatory Guide 1.99, Revision 2.
(c) All Beltline Welds are from Heat #83642, Linde 0091, Flux Lot #3536.
PTLR Revision 7 Beaver Valley Unit 2 5.2- 23 LRM Revision 79
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 30 EFPY(aJ PARAMETER VALUES Operating Time 30 EFPY Material-Intermediate Shell Plate B9004-1 B9004-1 Location 1/4T 3/4T Chemistry Factor, CF (°F) 40.5 40.5 Fluence, (f), (1 019 n/cm 2 )Cbl 2.113 0.8215 Fluence Factor, FF 1.203 0.9448
~RT NOT = CF X FF(°F) 48.74 38.27 lntitial RT NoT, ICF) 60 60 Margin, M(°F) 34 34 ART, per Regulatory Guide 1.99, Revision 2 143 132 (a) Table reflects Capsule X analysis per Reference 14.
19 2 (b) Fluence (f), is based upon fsurf (1 0 n/cm , E > 1.0 MeV)= 3.39 at 30 EFPY. The Beaver Valley Unit 2 reactor vessel wall thickness is 7.875 inches at the beltline region.
PTLR Revision 7 Beaver Valley Unit 2 5.2- 24 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-10 (Page 1 of 1)
RTPTs Calculation for 8eltline Region Materials at Life Extension (54 EFPY)cal
~~
Surface Neutron Fluence Chemistry Initial ~RTPTS(ct) MarginCel Material Description Material Heat Fluence Factor, Factor RTNDT(c) au a/\
ID Number FFCbl (oF) (oF) (oF) (oF) (oF)
Intermediate Shell Plate 89004-1 --- 5.18 1.4092 40.5 60 57.1 0 17 34 151.1 Intermediate Shell Plate 89004-2 --- 5.18 1.4092 37 40 52.1 0 17 34 126.1
---+ Using non-credible surveillance dataC9l 5.18 1.4092 51.4 40 72.4 0 17 34 146.4 Lower Shell Plate 89005-1 --- 5.21 1.4104 51 28 71.9 0 17 34 133.9 Lower Shell Plate 89005-2 --- 5.21 1.4104 44 33 62.1 0 17 34 129.1 Intermediate to Lower 101-171 83642 5.18 1.4092 34.4 -30 48.5 0 24.2 48.5 67.0 Shell Girth Weld
---+ Using credible surveillance dataC9l 5.18 1.4092 12.5 -30 17.6 0 8.8 17.6 5.2 Intermediate Shell Longitudinal Welds
.J. 101-12~J.
A&8 83642 1.76 1.1554 34.4 -30 39.7 0 19.9 39.7 49.5 9
Using credible surveillance dataC l 1.76 1.1554 12.5 -30 14.4 14.4 -1.1
---+ 0 7.2 Lower Shell 101-142 83642 1.77 1.1569 34.4 -30 39.8 0 19.9 39.8 49.6 Longitudinal Welds A&8
---+ Using credible surveillance dataC9l 1.77 1.1569 12.5 -30 14.5 0 7.2 14.5 -1.1 Notes:
(a) Data obtained from Table 6-3 of Reference 13.
(b) FF =fluencefactor= fC 0 -28
- 0-1109 Cfll_
(c) Initial RT NDT values are measured values.
(d) ~RTPTS = CF
- FF.
(e) M = 2 *(cru2 + cri\2)112.
(f) RTPTS =Initial RTNoT + ~RTPTS +Margin.
(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 a~;. 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 a~;. term of 1rF was utilized for 8VPS-2 plate 89004-2. The credibility evaluation conclusions are contained in Appendix A of Reference 13.
PTLR Revision 7 8eaverValley Unit 2 5.2-25 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 1 of 3)
RTPrs Calculation for Extended B~ltline Region Materials at Life Extension (54 EFPY)<al Surface Neutron Fluence Chemistry Initial Heat Number llRTPTS(e) au ab. Margin<0 RTPTS(g)
Material Description MateriaiiD Fluence Factor, Factor RTNDT(c)
(Lot Number) FF(bl (oF) (oF) (oF) (oF) (oF)
(x1 019 (oF) (oF) n/cm 2 )
89003-1 A9406-1 0.515 0.8147 91.0 50 74.1 0 17 34 158.1 Upper Shell Plates 89003-2 84431-2 0.515 0.8147 83.0 60 67.6 0 17 34 161.6 89003-3 A9406-2 0.515 0.8147 91.0 50 74.1 0 17 34 158.1 51912 (3490)' 0.515 0.8147 73.71 -50 60.1 0 28 56 66.1 51912 (3536) 0.515 0.8147 73.71 -70 60.1 0 28 56 46.1 101-122A 10(d)
Upper Shell EAIB 0.515 0.8147 27.0 22.0 17 11.0 40.5 72.5 101-1228 Longitudinal Welds IAGA 0.515 0.8147 41.0 -30 33.4 0 16.7 33.4 36.8 101-122C 10(d)
BOHB 0.515 0.8147 68.0 55.4 17 27.7 65.0 130.4 BAOED 0.515 0.8147 27.0 -50 22.0 0 11.0 22.0 -6.0 4P5174 0.515 0.8147 122.0 -50 99.4 0 28 56.0 105.4 Upper to Intermediate 51922 0.515 0.8147 68.0 -56(d) 55.4 17 27.7 65.0 64.4 103-121 Shell Girth Weld AAGC 0.515 0.8147 41.0 -70 33.4 0 16.7 33.4 -3.2 KOIB 0.515 0.8147 41.0 -60 33.4 0 16.7 33.4 6.8 89011-1 2V2436-0 1-002 0.0298 0.2188 77.0 60(h) 16.8 0 8.4 16.8 93.7 Inlet Nozzles 89011-2 2V2437-02-001 0.0298 0.2188 96.0 60(d)(h) 21.0 17 10.5 40.0 121.0 89011-3 2V2445-02-003 0.0298 0.2188 96.0 70(h) 21.0 0 10.5 21.0 112.0 PTLR Revision 7 Beaver Valley Unit 2 5.2-26 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 2 of 3)
RT PTS Calculation for Extended Beltline Region Materials at Life Extension (54 EFPY)(a)
Surface Material Neutron Fluence Chemistry Initial Heat Number LlRTPTS(e) au all. Margin<fl RTPTS(g)
Description MaterialiD Fluence Factor, Factor RTNDT(c)
(Lot Number) FF(bl (oF) (oF) (oF) (oF) (oF)
(x10 19 (oF) (oF) n/cm 2 )
4P5174 0.0298 0.2188 122.0 -50 26.7 0 13.3 26.7 3.4 LOHB 0.0298 0.2188 41.0 -60 9.0 0 4.5 9.0 -42.1 HABJC 0.0298 0.2188 27.0 -70 5.9 0 3.0 5.9 -58.2 BABBD 0.0298 0.2188 27.0 -70 5.9 0 3.0 5.9 -58.2 105-121A Inlet Nozzle FABGC 0.0298 0.2188 41.0 -80 9.0 0 4.5 9.0 -62.1 105-121 B Welds EOBC 0.0298 0.2188 27.0 -60 5.9 0 3.0 5.9 -48.2 105-121C FAAFC 0.0298 0.2188 95.0 -60 20.8 0 10.4 20.8 -18.4 CCJC 0.0298 0.2188 27.0 -60 5.9 0 3.0 5.9 -48.2 FAGB 0.0298 0.2188 27.0 -30 5.9 0 3.0 5.9 -18.2 BAOED 0.0298 0.2188 27.0 -50 5.9 0 3.0 5.9 -38.2 89012-1 AV8080-2E9558 0.0151 0.1440 94.0 -10 13.5 0 6.8 13.5 17.1 Outlet Nozzles 89012-2 AV8120-2E9560 0.0151 0.1440 94.5 -10 13.6 0 6.8 13.6 17.2 I I
89012-3 AV8097-2E9559 0.0151 0.1440 93.5 -10 13.5 0 6.7 13.5 16.9 BABBD 0.0151 0.1440 27.0 -70 3.9 0 1.9 3.9 -62.2 FAAFC 0.0151 0.1440 95.0 -60 13.7 0 6.8 13.7 -32.6 107-121A Outlet Nozzle HAAEC 0.0151 0.1440 41.0 -80 5.9 0 3.0 5.9 -68.2 107-1218 Welds HABJC 0.0151 0.1440 27.0 -70 3.9 0 1.9 3.9 -62.2 107-121C HAGB 0.0151 0.1440 27.0 -40 3.9 0 1.9 3.9 -32.2 GACJC 0.0151 0.1440 41.0 -80 5.9 0 3.0 5.9 -68.2 JAHB 0.0151 0.1440 41.0 -40 5.9 0 3.0 5.9 -28.2 PTLR Revision 7 Beaver Valley Unit 2 5.2-27 LRM Revision 79
Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 3 of 3)
RT PTS Calculation for Extended Beltline Region Materials at Life Extension (54 EFPY)(a)
(a) Data obtained from Table 6-4 of Reference 13.
(b) FF =fluencefactor= f(o. 2 a-o. 1 log(fll.
(c) Initial RT NOT values are measured values, unless otherwise noted.
(d) Initial RTNoT values are generic.
- FF.
(g) RTPTs = Initial RT NoT + ~RT 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.
PTLR Revision 7 Beaver Valley Unit 2 5.2- 28 LRM Revision 79
Enclosure C L-14-141 Beaver Valley Power Station, Unit No. 2 Core Operating Limits Report, Cycle 18 (14 Pages Follow)
Licensing Requirements Manual Core Operating Limits Report 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Core Operating Limits Report This Core Operating Limits Report provides the cycle specific parameter limits developed in accordance with the NRC approved methodologies specified in Technical Specification Administrative Control 5.6.3.
5.1.1 SL 2.1.1 Reactor Core Safety Limits See Figure 5.1-1.
5.1.2 SHUTDOWN MARGIN (SDM)
In MODES 1, 2, 3, and 4, SHUTDOWN MARGIN shall be:::: 1.77% ~klk.( l 1
a.
- b. Prior to manually blocking the Low Pressurizer Pressure Safety Injection Signal, the Reactor Coolant System shall be borated to ::::the MODE 5 boron concentration and shall remain :::: this boron concentration at all times when this signal is blocked.
- c. In MODE 5, SHUTDOWN MARGIN shall be:::: 1.0% ~klk.
5.1.3 LCO 3.1.3 Moderator Temperature Coefficient (MTC)
- a. Upper Limit- MTC shall be maintained within the acceptable operation limit specified in Technical Specification Figure 3.1.3-1.
Lower Limit- MTC shall be maintained less negative than - 4.29 x 1o*
4 b.
~k/k/°F at RATED THERMAL POWER.
- c. 300 ppm Surveillance Limit: (- 35 pcm/°F)
- d. 60 ppm Surveillance Limit: (- 41 pcm/°F) 5.1.4 LCO 3.1.5 Shutdown Bank Insertion Limits The Shutdown Banks shall be withdrawn to at least 225 steps_t2l 5.1.5 LCO 3.1.6 Control Bank Insertion Limits 2
- a. Control Banks A and B shall be withdrawn to at least 225 steps.( l
- b. Control Banks C and D shall be limited in physical insertion as shown in Figure 5.1-2.(2)
- c. Sequence Limits - The sequence of withdrawal shall be A, B, C and D bank, in that order.
- d. Overlap Limits( 2l - Overlap shall be such that step 129 on banks A, B, and C corresponds to step 1 on the following bank. When C bank is fully withdrawn, these limits are verified by confirming D bank is withdrawn at least to a position equal to the all-rods-out position minus 128 steps.
- (1) The MODE 1 and MODE 2 with keff:::: 1.0 SDM requirements are included to address SDM requirements (e.g., MODE 1 Required Actions to verify SDM) that are not within the applicability of LCO 3.1.1, SHUTDOWN MARGIN (SDM).
(2) As indicated by the group demand counter CO LR Cycle 18 Beaver Valley Unit 2 5.1 - 1 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 5.1 Core Operating Limits Report 5.1.6 LCO 3.2.1 Heat Flux Hot Channel Factor (Fo(Z))
The Heat Flux Hot Channel Factor- F0 (Z) limit is defined by:
F0 (Z) :o; [C:Q}K(Z) for P > 0.5 F (Z):o;[CFQ]*K(Z) for P :o; 0.5 Q 0.5 THERMAL POWER Where: CFQ = 2.40 p = RATED THERMAL POWER K(Z) = the function obtained from Figure 5.1-3.
F5 (Z) = F~(Z)
- 1.0815 F~ (Z) = F5(Z)
- W(Z)
W(Z) values are provided in Table 5.1-1. The W(Z) values are generated assuming that they will be used for a full power surveillance. When a part power surveillance is performed, the W(Z) values should be multiplied by the factor 1/P, when P > 0.5. When P is :o; 0.5, the W(Z) values should be multiplied by the factor 1/(0.5), or 2.0. This is consistent with the adjustment in the F0 (Z) limit at part power _conditions.
The Fo(Z) penalty function, applied when the analytic Fo(Z) function increases from one monthly measurement to the next, is provided in Table 5.1-2.
5.1. 7 LCO 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor ( F~H)
Where: CFL'IH = 1.62 PFL'IH = 0.3 THERMAL POWER p = RATED THERMAL POWER 5.1.8 LCO 3.2.3 Axial Flux Difference CAFD)
The AFD acceptable operation limits are provided in Figure 5.1-4.
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 2 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 5.1 Core Operating Limits Report 5.1.9 LCO 3.3.1 Reactor Trip System Instrumentation - Overtemperature and Overpower f.. T Parameter Values from Table Notations 3 and 4
- a. Overtemperature f.. T Setpoint Parameter Values:
Parameter Overtemperature f.. T reactor trip setpoint K1 ~ 1.239 Overtemperature t...T reactor trip setpoint Tavg coefficient Overtemperature f.. T reactor trip setpoint pressure K3 :2: 0.001/psia coefficient Tavg at RATED THERMAL POWER Nominal pressurizer pressure P' :2: 2250 psia Measured reactor vessel f.. T lead/lag time constants 't 1 =0 sec*
(* The response time is toggled off to meet the analysis 't2= 0 sec*
value of zero.)
Measured reactor vessel f.. T lag time constant 't3 ~ 6 sees Measured reactor vessel average temperature lead/lag 't4 :2: 30 sees time constants 'ts ~ 4 sees Measured reactor vessel average temperature lag time 'ts ~ 2 sees constant f (t...l) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) For qt- qb between -37% and +15%, f 1 (t...l) = 0, where q1 and Qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q1 + qb is total THERMAL POWER in percent of RATED THERMAL POWER.
(1) T' represents the cycle-specific Full Power Tavg value used in core design.
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 3 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 5.1 Core Operating Limits Report (ii) For each percent that the magnitude of (q 1 - qb) exceeds -37%, the !1 T trip setpoint shall be automatically reduced by 2.52% of its value at RATED THERMAL POWER.
(iii) For each percent that the magnitude of (q 1 - qb) exceeds +15%, the !1 T trip setpoint shall be automatically reduced by 1.47% of its value at RATED THERMAL POWER.
- b. Overpower !1T Setpoint Parameter Values:
Parameter Overpower !1 T reactor trip setpoint K4:.:; 1.094 Overpower !1 T reactor trip setpoint Tavg K5 2 0.02/oF for increasing rate/lag coefficient average temperature
=
K5 0/°F for decreasing average temperature Overpower !1T reactor trip setpoint Tavg K6 2 0.0021/oF forT> T" heatup coefficient =
K6 0/oF for T :.:; T" Tavg at RATED THERMAL POWER Measured reactor vessel !1 T lead/lag t 1 = 0 sec*
time constants t2 = 0 sec*
(*The response time is toggled off to meet the analysis value of zero.)
Measured reactor vessel !1 T lag time
,constant Measured reactor vessel average t5:.:; 2 sees temperature l~g time constant Measured reactor vessel average t7 2 10 sees temperature rate/lag time constant (1) T" represents the cycle-specific Full Power Tavg value used in core design.
COLR Cycle 18 Beaver Valley Unit 2 5.1 -4 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 5.1 Core Operating Limits Report 5.1.1 0 LCO 3.4.1, RCS Pressure. Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Parameter Indicated Value Reactor Coolant System Tavg Pressurizer Pressure Pressure~ 2214 psia( 2l Reactor Coolant System Total Flow Rate Flow~ 267,300 gpm( 3l
- (1) The Reactor Coolant System (RCS) indicated Tavg value is determined by adding the appropriate allowances for rod control operation and verification via control board indication (3.6°F) to the cycle specific full power Tavg used in the core design.
(2) The pressurizer pressure value includes allowances for pressurizer pressure control operation and verification via control board indication.
(3) The RCS total flow rate includes allowances for normalization of the cold leg elbow taps with a beginning of cycle precision RCS flow calorimetric measurement and verification on a periodic basis via control board indication.
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 5 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report .
5.1 5.1 Core Operating Limits Report 5.1.11 LCO 3.9.1 Boron Concentration (MODE 6)
The boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained ;::: 2400 ppm. This value includes a 50 ppm conservative allowance for uncertainties.
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 6 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 5.1 Core Operating Limits Report 5.1.12 References
- 1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (Westinghouse Proprietary).
- 2. WCAP-87 45-P-A, "Design Bases for the Thermal Overtemperature L1 T and Thermal Overpower L1T Trip Functions," September 1986.
- 3. WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (Westinghouse Proprietary).
- 4. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control-Fa Surveillance Technical Specification," February 1994.
- 5. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.
- 6. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"
April 1995 (Westinghouse Proprietary).
- 7. WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids," April 1999.
- 8. Caldon, Inc. Engineering Report-SOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the
~EFMf"M System,~' Revision 0, March 1997. . _
- 9. Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM-f"M System," Revision 0, May 2000.
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 7 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 670 660 ~
~
"'I 2435 PSIA I I UNACCEPTABLE OPERATION I
650
.......___ ~\
~ ~
~2250 ~SIA I' 640 ~
~
~
G:;' 630 ~ ~
~ !'--..
~
~
&P Cd t-< 620
~~ ~~ ~I 2000 PSIA I
~ \
~~
/~ '~
I 1920 PSIA I 610
~~
600 I ACCEPTABLE OPERATION I
~
~
~
590
~
580 0 0.2 0.4 0.6 0.8 1 1.2 1.4 FRACTION OF RATED THERMAL POWER Figure5.1-1 (Page 1 of1)
REACTOR CORE SAFETY LIMIT THREE LOOP OPERATION (Technical Specification Safety Limit 2.1.1)
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 8 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 225 Vr (54.53, 1
I 225)
I 1
200 /
~" 1 (1oo, 187) 1
/ J BANKCJ c /
- s: 175 v
"0
.c
<tl
+-'
- 5: 150 .I
/ /
U) 0.
Q)
+-'
(/)
...__, /
/ /
125 z
0 I-vI (0,114) I I BANKO I /
(/)
0 100 /
a..
~
z
<(
/
/
co 75 I 0
0 /
~
/
50 25
/
v .
v
~
VI (8, o) I 0
0 10 20 30 40 . 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER Figure 5.1-2 (Page 1 of 1)
CONTROL ROD INSERTION LIMITS AS A FUNCTION OF RATED POWER LEVEL COLR Cycle 18 Beaver Valley Unit 2 5.1 - 9 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 1.2 1 o.o, 1.00 1 16.0, 1.00 1 1.0 12.0, 0.925 1 0.8
§ 0.6 0.4 0.2 0.0 0 2 4 6 8 10 12 Core Height (feet)
Figure 5.1-3 (Page 1 of 1)
F0 T NORMALIZED OPERATING ENVELOPE, K(Z)
COLR Cycle 18 Beaver Valley Unit 2 5.1 - 10 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 120 I I I I I I ~
110 I I l I 8, i j
10 0)
I I I
(+8, 10 0)
I I I(
I I L 100 5 I I vI 1\ I 90 I IA I UNACCEPTABLE I I ~ /IACC EP TABLE I !\
UNACCEPTABLE i OPERATION OPERATION I
- I f OPERATION . \I I I I I I
... 80 I
0
~ I I I I I~ ~
I i
a.
C'a 70 I I I I \ i E
... I I I =
I I \ I I I I I Ji I \. I
(!)
..c: /!
1- 60 ~
I Ill I I I I "C
(!)
C'a a::
0 50 I
I I i if I I I
I I
I
\
\ I I
~
I I ,!(-25, 5o) I I I I J( +2 2, 50)
! I I !
i 40 I
I I I I
! I I I i
30 I I I I I I I I I I I I I I I I I I I 20 I I I I i I I I, I I I I I . I I 10 I I I 0 I I I I I I I I I I I *
-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Axial Flux Difference (Delta I)%
Figure 5.1-4 (Page 1 of 1)
AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF PERCENT OF RATED THERMAL POWER FOR RAOC COLR Cycle 18 Beaver Valley Unit 2 5.1 - 11 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 Table 5.1-1 (Page 1 of 2)
F0 Surveillance W(Z) Function versus Burn up Exclusion Axial Elevation 150 3000 8000 12000 16000 Zone Point (feet) (MWD/MTU) (MWD/MTU) (MWD/MTU) (MWD/MTU) (MWD/MTU)
- 1 12.1 1.0000 1.0000 1.0000 1.0000 1.0000
- 2 11.9 1.0000 1.0000 1.0000 1.0000 1.0000
- 3 11.7 1.0000 1.0000 1.0000 1.0000 1.0000
- 4 11.5 1.0000 1.0000 1.0000 1.0000 1.0000
- 5 11.3 1.0000 1.0000 1.0000 1.0000 1.0000
- 6 11.1 1.0000 1.0000 1.0000 1.0000 1.0000
- 7 10.9 1.0000 1.0000 1.0000 1.0000 1.0000 8 10.7 1.1506 1.1928 1.2322 1.2296 1.2203 9 10.5 1.1446 1.1860 1.2247 1.2235 1.2105 10 10.3 1.1390 1.1783 1.2162 1.2162 1.2053 11 10.1 1.1347 1.1700 1.2084 1.2084 1.1995 12 9.9 1.1345 1.1618 1.2027 1.2000 1.1931 13 9.7 1.1378 1.1570 1.1971 1.1909 1.1852 14 9.5 1.1418 1.1561 1.1907 1.1815 1.1798 15 9.3 1.1440 1.1542 1.1857 U719 1.1813 16 9.1 1.1472 1.1518 1.1850 1.1607 1.1864 17 8.9 1.1593 1.1556 1.1905 1.1597 1.1883 18 8.7 1.1758 1.1664 1.2013 1.1647 1.1929 19 8.5 1.1884 1.1745 1.2082 1.1750 1.2045 20 8.3 1.1987 1.1807 1.2130 1.1825 1.2133 21 8.1 1.2066 1.1849 1.2157 1.1880 1.2196 22 7.9 1.2121 1.1872 1.2162 1.1917 1.2239 23 7.6 .1.2155 1.1878 1.2148 1.1936 .1.2260 24 7.4 1.2168 1.1868 1.2115 1.1937 1.2260 25 7.2 1.2161 1.1843 1.2063 1.1922 1.2239
. 26 7.0 1.2133 1.1813 1.2002 1.1889 1.2197 27 6.8 1.2086 1.1771 1.1931 1.1849 1.2142 28 6.6 1.2021 1.1712 1.1845 1.1803 1.2080 29 6.4 1.1940 1.1639 1.1745 1.1740 1.2000 30 6.2 1.1842 1.1553 1.1630 1.1662 1.1901 31 6.0 1.1732 1.1456 1.1504 1.1571 1.1790 32 5.8 1.1605 1.1345 1.1365 1.1468 1.1657 Note: Top and Bottom 10% Excluded COLR Cycle 18 Beaver Valley Unit 2 5.1 - 12 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 TABLE 5.1-1 (Page 2 of 2)
F0 Surveillance W(Z) Function versus Burnup Exclusion Axial Elevation 150 3000 8000 12000 16000 Zone Point (feet) (MWD/MTU) (MWD/MTU) (MWD/MTU) (MWD/MTU) (MWD/MTU) 33 5.6 1.1477 1.1229 1.1233 1.1354 1.1523 34 5.4 1.1386 1.1141 1.1138 1.1243 1.1447 35 5.2 1.1308 1.1104 1.1036 1.1163 1.1414 36 5.0 1.1258 1.1092 1.0963 1.1112 1.1382 37 4.8 1.1233 1.1089 1.0924 1.1078 1.1336 38 4.6 1.1206 1.1098 1.0885 1.1047 1.1283 39 4.4 1.1172 1.1106 1.0849 1.1011 1.1222 40 4.2 1.1167 1.1112 1.0810 1.0970 1.1155 41 4.0 1.1194 1.1114 1.0773 1.0927 1.1084 42 3.8 1.1234 1.1119 1.0749 1.0881 1.1009 43 3.6 1.1266 1.1141 1.0746 1.0834 1.0933 44 3.4 1.1294 1.1168 1.0743 1.0799 1.0860 45 3.2 1.1329 1.1196 1.0747 1.0775 1.0773 46 3.0 1.1377 1.1266 1.0776 1.0790 1.0747 47 2.8 1.1466 1.1426 1.0893 1.0912 1.0839 48 2.6 1.1642 1.1673 1.1068 1.1050 1.0990 49 2.4 1.1855 1.1927 1.1251 1.1192 1.1143 50 2.2 1.2075 1.2180 1.1434 1.1337 1.1296 51 2.0 1.2290 1.2433 1.1619 1.1481 1.1447 52 1.8 1.2500 1.2680 1.1800 1.1622 1.1595 53 1.6 1.2701 1.2917 1.1976 1.1760 1.1739 54 1.4 1.2887 1.3135 1.2139 1.1888 1.1876
- 55 1.2 1.0000 1.0000 1.0000 1.0000 1.0000
- 56 1.0 1.0000 1.0000 1.0000 ' 1.0000 1.0000
- 57 0.8 1.0000 1.0000 1.0000 1.0000 1.0000
- 58 0.6 1.0000 1.0000' 1.0000 1.0000 1.0000
- 59 0.4 1.0000 1.0000 1.0000 1.0000 1.0000
- 60 0.2 1.0000 1.0000 1.0000 1.0000 1.0000
- '61 0.0 1.0000 1.0000 1.0000. 1.0000 1.0000 Note: Top and Bottom 10% Excluded COLR Cycle 18 Beaver Valley Unit 2 5.1 - 13 LRM Revision 80
Licensing Requirements Manual Core Operating Limits Report 5.1 Table 5.1-2 (Page 1 of 1)
F0 (Z) Penalty Factor versus Burnup Cycle Burnup (MWD/MTU) F0 (Z) Penalty Factor
>0 1.0200 Note: The Penalty Factor, to be applied to F0 (Z) in accordance with Technical Specification Surveillance Requirement (SR) 3.2.1.2, is the maximum factor by which F0 (Z) is expected to increase over a 39 Effective Full Power Day (EFPD) interval (surveillance interval of 31 EFPD plus the maximum allowable extension not to exceed 25% of the surveillance interval per Technical Specification SR 3.0.2) starting from the burnup at which the F0 (Z) was determined.