L-03-151, Original Issuance of the Pressure and Temperature Limits Report
| ML032751497 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/29/2003 |
| From: | Pearce L FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-03-151 | |
| Download: ML032751497 (48) | |
Text
FENOC FirstEniergy Nuclear Operating Company Beaver Valley Power Station Route 168 PO. Box 4 Shippingport, PA 15077-0004 L William Pearce Site Vice President 724-682-5234 Fax: 724-643-8069 September 29, 2003 L-03-151 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Original Issuance of the Pressure and Temperature Limits Report FirstEnergy Nuclear Operating Company (FENOC) hereby submits the initial issuance of the Pressure and Temperature Limits Report (PTLR) for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2. The PTLR for each BVPS unit is provided as an Enclosure to this letter.
The initial issuance, Revision 0, of the PTLR for each unit is being provided in accordance with Technical Specification 6.9.6, "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." Technical Specification 6.9.6.c states that the PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto. Creation of the PTLR was NRC approved by License Amendment Nos. 256 (Unit No. 1) and 138 (Unit No. 2) dated July 15, 2003. The BVPS Unit Nos. 1 and 2 PTLRs were issued as part of Revision Nos. 35 (Unit No. 1) and 33 (Unit No. 2) of the respective unit's Licensing Requirements Manual, coincident with the implementation of the amendments on September 9, 2003.
No new commitments are contained in this submittal.
If there are any questions concerning this matter, please contact Mr. Larry R. Freeland, Manager, Regulatory Affairs/Performance Improvement at 724-682-5284.
Enclosures:
- 1.
BVPS Unit 1 Revision 0 of the Pressure and Temperature Limits Report
- 2.
BVPS Unit 2 Revision 0 of the Pressure and Temperature Limits Report Aoc1
Beaver Valley Power Station, Unit No. 1 and No. 2 Original Issuance of the Pressure and Temperature Limits Report L-03-151 Page 2 c:
Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)
BVPS-I LICENSING REQUIREMENTS MANUAL SECTION 4.2 PRESSURE AND TEMPERATURE LIMITS REPORT BVPS-1 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.1.2.4 N/A N/A 4.2-3 3.4.1.3 N/A N/A 4.2-3 3.4.9.1 4.2.1.1 4.2-1 N/A 4.2-2 3.4.9.3 4.2.1.2 N/A 4.2-3 4.2.1.3 3.5.3 N/A N/A 4.2-3 3.5.4.1.2 N/A N/A 4.2-3 3.10.3 N/A 4.2-1 N/A
_ _ _ _ _ _ _4
.2 -2 PTLR Revision 0 4.2-i LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
This Pressure and Temperature Limits Report (PTLR) for Unit I has been prepared in accordance with the requirements of Technical Specification 6.9.6. Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications (TS) addressed in, or make reference to, this report are listed below:
TS 3.1.2.4 Reactivity Control Systems - Charging Pumps - Operating, TS 3.4.1.3 Reactor Coolant System - Shutdown, TS 3.4.9.1 Reactor Coolant System - Pressure/Temperature Limits, TS 3.4.9.3 Overpressure Protection Systems, TS 3.5.3 ECCS Subsystems - T8vs < 3501F, TS 3.5.4.1.2 Boron Injection Tank < 3500F, and TS 3.10.3 Special Test Exceptions - Pressure/Temperature Limitations -Reactor Criticality.
4.2.1 Operating Limits The PTLR limits for Beaver Valley Power Station (BVPS) Unit 1 were developed using a methodology specified in the Technical Specifications.
The methodology listed in Reference 1 was used with two exceptions:
a)
Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1", and b)
Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure".
4.2.1.1 RCS Pressure and Temperature (P/I) Limits (TS 3.4.9.1)
The RCS temperature rate-of-change limits defined in Reference 2 are:
- a. A maximum heatup of 1000F in any one hour period.
- b. A maximum cooldown of 1 000F in any one hour period, and
- c. A maximum temperature change of less than or equal to 5F in any one hour period during inservice hydrostatic testing operations above system design pressure.
The RCS P/T limits for heatup, leak testing, and criticality are specified by Figure 4.2-1 and Table 4.2-1. The RCS P/T limits for cooldown are shown in Figure 4.2-2 and Table 4.2-2.
These limits are defined in Reference 2.
Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 4.2.1, the RCS P/T limits for heatup and cooldown shown in Figures 4.2-1 and 4.2-2 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to 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.
PTLR Revision 0 4.2-1 LRM Revision 35
A BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 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 4F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.
Pressure-temperature limit curves shown in Figure 4.2-3 were developed for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This figure provides the ASME III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.
4.2.1.2 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)
The power operated relief valves (PORVs) shall each have maximum lift setting and enable temperature in accordance with Table 4.2-3. The lift setting provided does not impose any reactor coolant pump restrictions.
The PORV setpoint is based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 4.2.1. The PORV lift setting shown in Table 4.2-3 accounts for appropriate instrument error.
4.2.1.3 OPPS Enable Temperature (TS 3.4.9.3)
Two different temperatures are used to determine the OPPS enable temperature, they are the arming temperature and the calculated enable temperature. The arming temperature (when the OPPS rendered operable) is established per ASME Section XI, Appendix G. At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature is 343 0F.
The calculated enable temperature is based on either a RCS temperature of less than 200F or materials concerns (reactor vessel metal temperature less than RTNDT + 501F), whichever is greater. The calculated enable temperature does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 308 0F.
As the arming temperature is higher and, therefore, more conservative than the calculated enable temperature, the OPPS enable temperature, as shown in Table 4.2-3, is set to equal the arming temperature.
The calculation method governing the heatup and cooldown of the RCS requires the arming of the OPPS at and below the OPPS enable temperature specified in Table 4.2-3, and 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.
PTLR Revision 0 4.2-2 LRM Revision 35
i BVPS-1 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT From a plant operations viewpoint the terms "armed" and "enabled" are synonymous when it comes to activating the OPPS. As stated in the applicable operating procedure, the OPPS is activated (armed/enabled) manually before entering the applicability of TS 3.4.9.3. This is accomplished by placing two keylock switches (one in each train) into their "automatic" position.
Once OPPS is activated (armed/enabled) reactor coolant system pressure transmitters will signal a rise in system pressure above the OPPS setpoint. This will initiate an alarm in the control room and open the OPPS PORVs.
4.2.1.4 Reactor Vessel Boltup Temperature (TS 3.4.9.1)
The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60'F. Boltup is a condition in which the reactor vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.
4.2.2 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The capsule withdrawal schedule is provided in Table 4.5-3 of the UFSAR. Also, the results of these analyses shall be used to update Figures 4.2-1 and 4.2-2, and Tables 4.2-1 and 4.2-2.
The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.
The pressure vessel material surveillance program (References 3 and 4) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNmT, which is determined in accordance with ASME,Section III, NB-233 1. The empirical relationship between RTmuT 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 0 4.2-3 LRM Revision 35
'a BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.3 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 4.2-4, taken from Reference 5, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
Table 4.24a, taken from Reference 2, shows the Calculation of Chemistry Factors based on St. Lucie and Fort Calhoun Surveillance Capsule Data.
Table 4.2-4b, taken from Reference 3, shows the St. Lucie and Fort Calhoun Surveillance Weld Data.
Table 4.2-5, taken from Reference 2, provides the reactor vessel beltline material property table.
Table 4.2-6, taken from Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 22 EFPY.
Table 4.2-7, taken from Reference 2, shows the calculation of ARTs for 22 EFPY.
Table 4.2-8 shows the Reactor Vessel Toughness Data (Unirradiated).
Table 4.2-9, taken from Reference 5, provides RTprs values for 28 EFPY.
Table 4.2-10, taken from Reference 5, provides RTprs values for 45 EFPY.
PTLR Revision 0 4.2-4 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.4 References
- 1.
WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al., January 1996.
- 2.
WCAP-15570, Revision 2, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, April 2001.
- 3.
WCAP-15571, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," C. Brown, et. al., November 2000.
- 4.
WCAP-8475, "Duquesne Light Company, Beaver Valley Unit No. I Reactor Vessel Radiation Surveillance Program," J. A. Davidson, October 1974.
- 5.
WCAP-15569, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 1,"
C. Brown, et al., November 2000.
- 6.
10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 7.
10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 8.
Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"
U.S. Nuclear Regulatory Commission, May 1988.
- 9.
Westinghouse Report, "Beaver Valley Unit 1 FirstEnergy Nuclear Operating Company -
Overpressure Protection System - Setpoints for Y-Capsule", Revision 1, April 2001.
- 10.
FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
PTLR Revision 0 4.2-5 LRM Revision 35
I BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LIMITING ART VALUES AT 22 EFPY:
INTERMEDIATE & LOWER SHELL PLATE 1/4T, 233-F 3/4T, 1960F a
vi w
or n
0 z
0 50 100 150 200 250 300 350 400 450 500 550 INDICATED TEMPERATURE IF)
Figure 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 22 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-6 LRM Revision 35
BVPS-I LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL:
INTERMEDIATE & LOWER SHELL PLATE LIMrITING ART VALUES AT 22 EFPY:
114T, 2330F 3/4T, 1960F 2500 2250 2000 Unacceptable
______Acptal 1750 Operation Operation 71500 u1250 a.
2 1000 2
750 50 005 Cool wn Rates Limitations ApplicableOFmr (ateadytate) 2Fr 250 r
0 50 100 150 200 250 300 350 400 450 500 550 INDICATED TEMPERATURE rF)
Figure 4.2-2 Reactor Coolant System Cooldown Limitations Applicable for the First 22 EFPY (TS 3.4.9. 1)
PTLR Revision 0 4.2-7 LRM Revision 35
4 BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2000 1500 co n
CO X
1 000 500 n 50 60 70 80 90 TEMPERATURE CF) 100 110 120 Figure 4.2-3 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1)
PTLR Revision 0 4.2-8 LRM Revision 35
BVPS-1 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-1 Heatup Curve Data Points for 22 EFPY (TS 3.4.9.1) 100F/HR HEATUP I 00F/HR CRITICALITY LEAK TEST LIMIT Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
( F)
(psig)
( F)
(psig)
(-F)
(psig)
( F)
(psig)
( F)
(psig) 60 0
200 677 289 0
289 716 271 2000 60 564 205 696 289 564 289 739 289 2485 65 564 210 716 289 565 289 764 70 564 215 739 289 565 289 792 75 564 220 764 289 566 289 822 80 564 225 792 289 566 289 856 85 564 230 822 289 568 289 894 90 564 235 856 289 569 289 936 95 564 240 894 289 571 290 982 100 564 245 936 289 572 295 1033 105 564 250 982 289 575 300 1089 110 564 255 1033 289 577 305 1151 115 564 260 1089 289 580 310 1219 120 564 265 1151 289 583 315 1294 125 564 270 1219 289 586 320 1378 130 565 275 1294 289 591 325 1470 135 566 280 1378 289 593 330 1571 140 568 285 1470 289 600 335 1682 145 571 290 1571 289 601 340 1806 150 575 295 1682 289 611 345 1941 155 580 300 1806 289 612 350 2091 160 586 305 1941 289 621 355 2222 165 593 310 2091 289 621 360 2361 170 601 315 2222 289 621 175 611 320 2361 289 621 180 621 289 621 180 621 289 633 180 621 289 646 185 633 289 661 190 646 289 677 195 661 289 696 PTLR Revision 0 4.2-9 LRM Revision 35 1
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 (Page 1 of 2)
Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1)
STEADY 20FIHR.
40'F/HR.
60'F/HR.
100F/HR.
STATE Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
(F)
(psig)
(F)
(psig)
(F)
(psig)
(F)
(psig)
(CF)
(psig) 60 0
60 0
60 0
60 0
60 0
60 621 60 609 60 566 60 521 60 430 65 621 65 611 65 567 65 523 65 431 70 621 70 612 70 568 70 524 70 432 75 621 75 614 75 570 75 525 75 433 80 621 80 615 80 572 80 527 80 435 85 621 85 617 85 574 85 529 85 437 90 621 90 619 90 576 90 531 90 439 95 621 95 621 95 578 95 534 95 442 100 621 100 621 100 581 100 536 100 445 105 621 105 621 105 584 105 540 105 448 110 621 110 621 110 587 110 543 110 452 115 621 115 621 115 591 115 547 115 457 120 621 120 621 120 596 120 552 120 462 125 621 125 621 125 600 125 557 125 468 130 621 130 621 130 606 130 562 130 474 135 621 135 621 135 612 135 569 135 481 140 621 140 621 140 618 140 576 140 490 145 621 145 621 145 621 145 584 145 499 150 621 150 621 150 621 150 592 150 509 155 621 155 621 155 621 155 602 155 520 160 621 160 621 160 621 160 613 160 533 165 621 165 621 165 621 165 621 165 547 170 621 170 621 170 621 170 621 170 563 175 621 175 621 175 621 175 621 175 581 180 621 180 621 180 621 180 621 180 600 180 621 180 621 180 621 180 621 185 622 180 778 180 742 180 706 180 670 190 647 185 792 185 757 185 723 185 689 195 674 190 808 190 775 190 742 190 709 200 704 195 826 195 794 195 762 195 732 205 737
-9
-762-19 PTLR Revision 0 4.2-10 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 (Page 2 of 2)
Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1)
STEADY 20F/HR.
40'FARR.
60'F/HR.
lOOTF/HR.
STATE Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
Temp.
Press.
(F)
(psig)
(F)
(psig)
(F)
(psig)
(F)
(psig)
(F)
(psig) 200 846 200 815 200 785 200 757 210 774 205 868 205 839 205 811 205 785 215 815 210 892 210 865 210 839 210 815 220 861 215 918 215 894 215 871 215 850 225 911 220 947 220 925 220 905 220 888 230 967 225 980 225 961 225 944 225 930 235 1030 230 1016 230 1000 230 986 230 976 240 1099 235 1055 235 1043 235 1033 235 1028 245 1147 240 1099 240 1090 240 1086 240 1085 250 1201 245 1147 245 1143 245 1143 245 1147 255 1260 250 1201 250 1201 250 1201 250 1201 260 1325 255 1260 255 1260 255 1260 255 1260 265 1397 260 1325 260 1325 260 1325 260 1325 270 1477 265 1397 265 1397 265 1397 265 1397 275 1565 270 1477 270 1477 270 1477 270 1477 280 1662 275 1565 275 1565 275 1565 275 1565 285 1770 280 1662 280 1662 280 1662 280 1662 290 1888 285 1770 285 1770 285 1770 285 1770 295 2020 290 1888 290 1888 290 1888 290 1888 300 2165 295 2020 295 2020 295 2020 295 2020 305 2325 300 2165 300 2165 300 2165 300 2165 305 2325 305 2325 305 2325 305 2325 PTLR Revision 0 4.2-11 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)
FUNCTION SETPOINT OPPS Enable Temperature 3430F PORV Setpoint
< 403 psig PTLR Revision 0 4.2-12 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f a AFF R
ATNDT( )
FF *RTNDT V
.323
.689 128.49 88.53
.475 Lower Shell Plate B6903-1(d)
U
.646
.878 118.93 104.42
.771 (Longitudinal)
W
.986
.996 148.52 147.93
.992 Y
2.15 1.21 142.18 172.04 1.464 V
.323
.689 137.81 94.95
.475 Lower Shell Plate B6903-l(d)
U
.646
.878 131.84 115.76
.771 (Transverse)
W
.986
.996 179.99 179.27
.992 Y
2.15 1.21 166.93 201.99 1.464 SUM:
1104.89 7.404 CF =(FF*RTT)
+ E(FF2)
(1104.89) + (7.404) = 149.20F V
.323
.689 169.30 116.65
.475 Beaver Valley U
.646
.878 176.30 154.79
.771 Surv. Weld Material 305424 (d)
W
.986
.996 198.99 198.19
.992 Y
2.15 1.21 189.41 229.19 1.464 SUM:
698.82 3.702 CF =
W(FF*RTNDT) + (FF 2 ) = (698.82) + (3.702) = 188.80F Notes:
(a)
F= Calculated fluence from Beaver Valley Unit 1 capsule Y dosimetry analysis results, (x 1019 n/cm2, E > 1.0 Mev).
(b)
FF = fluence factor = f (0.2 8.1 I bgf)
(c)
The surveillance weld metal ARTNDT values have been adjusted by a ration factor of 1.06.
(d)
Data not credible.
PTLR Revision 0 4.2-13 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4a Calculation of Chemistry Factors(a)
(Based on St. Lucie and Fort Calhoun Surveillance Capsule Data)
Material Capsule Capsule f (b)
FF(c)
ARTNDT) )
FF *ARTNDT -
FF2 St. Lucie 970 0.627 0.869 72.3 76.1 0.755 Surveillance 1040 0.909 0.973 67.4 79.7 0.947 Weld Metal 2840 1.41 1.10 68.0 90.9 1.21 Heat 90136 SUM:
246.7 2.91 CF =(FFRTNDT) +
- (FF2) = (246.7) + (2.91) = 84.80F Fort Calhoun W-225 0.553 0.834 238 183.0 0.696 Surveillance W-265 0.771 0.927 221 194.1 0.859 Weld Metal W-275 1.28 1.07 219 226.2 1.14 Heat 305414 SUM:
603.3 2.695 CF = Q(FF*RTm) + 1;(FF2)
(603.3) + (2.695) = 223.91F 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.
MB2301y).
(b) f = Calculated fluence (x 1019 n/cm2, E > 1.0 Mev) from Reference 2.
(c)
FF = fluence factor f (0.28 -0.1 g.
(d)
ARTNDT values are the measured 30 ft-lb. shift values taken from Reference 2.
PTLR Revision 0 4.2-14 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-4b St. Lucie and Fort Calhoun Surveillance Weld Data(a)
Material Capsule Cu Ni Irradiated Fluence ARTmT Temperature OF 10' 9 n/cm 2 St. Lucie 970 0.2291 0.0699 546.7 0.627 72.3 Weld Metal 1040 0.2291 0.0699 546.7 0.909 67.4 Heat 90136 2840 0.2291 0.0699 546.7 1.41 68.0 Fort Calhoun W-225 0.35 0.60 527 0.553 238 Weld Metal W-265 0.35 0.60 534 0.771 221 Heat 305414 W-275 0.35 0.60 538 1.28 219 Notes:
(a)
Use of St. Lucie and Fort Calhoun Surveillance Capsule Data approved by NRC letter dated February 20,2002, "BEAVER VALLEY POWER STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: AMENDED PRESSURE-TEMPERATURE LIMITS (TAC NO.
MB2301)".
(b)
Data contained in this table was obtained from Reference 3.
PTLR Revision 0 4.2-15 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-5 Reactor Vessel Beltline Material Properties Material Description Cu(/o)
Ni(%)
Chemistry Initial Factor RTNDT(0F) (a)
Intermediate Shell Plate B6607-1 0.14 0.62 100.5 43 Intermediate Shell Plate B6607-2 0.14 0.62 100.5 73 Lower Shell Plate B6903-1 0.21 0.54 147.2 27 Lower Shell Plate B7203-2 0.14 0.57 98.7 20 Intermediate to Lower Shell Weld 0.27 0.07 124.3
-56 Seam (Heat 90136)11-714 Intermediate Longitudinal Shell 0.28 0.63 191.7
-56 Weld Seams (Heat 305424)19-714 A&B Lower Longitudinal Weld Seams 0.34 0.61 210.5
-56 (Heat 305414)20-714 A&B Surveillance Weld (Heat 305424) 0.26 0.61 181.6 Note:
(a)
The initial RTNmT values for the plates and are based on measured data while the weld values are generic.
LRM Revision 35 PTLR Revision 0 4.2-16
BVPS-I LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-6 Summary of Adjusted Reference Temperature (ARTs) for 22 EFPY MATERIAL DESCRIPTION 22 EFPY l/4T ART(OF)(a) 34T ART(OF)(a)
Intermediate Shell Plate B6607-1 193 166 Intermediate Shell Plate B6607-2 223 196 Lower Shell Plate B7203-2 168 141 Lower Shell Plate B6903-1 230 191
- Using S/C Data(b) 233 193 Intermediate Shell Longitudinal Weld 19-714A/B 145 102
- Using S/C Data(b) 143 100 Intermediate to Lower Shell Circ. Weld 11-714 152 119
- Using S/C Data (C) 86 63 Lower Shell Longitudinal Weld 20-714A/B 159 111
- Using S/C Data(d) 168 117 Notes:
(a) ART = I + ARTmT + M.
(b) Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a full cA.)
(c) Based on credible St. Lucie Unit 1 surveillance data.
(d) Based on Fort Calhoun Unit 1 surveillance data. (Data not credible. ART calculated with a full aA.)
PTLR Revision 0 4.2-17 LRM Revision 35
BVPS-1 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-7 Calculation of Adjusted Reference Temperatures (ARTs) for 22 EFPY PARAMETER VALUES Operating Time 22 EFPY Material Plate B6903-1 Plate B6607-2 Location Lower Shell Intermediate Plate Shell Plate I/4T ART(0F) 3/4T ART(0F)
Chemistry Factor, CF (F) 149.2 100.5 Fluence (f), n/cm2 (E>1.0 Mev)(a) 1.70 x IO' 6.62 x 10's Fluence Factor, FF 1.15
.884 ARTNDT = CF x FF(OF)(C) 171.6(c) 88.84 Initial RTNUT, I(CF)(a) 27 73 Margin, M(0F) 34 (C) 34 ART = I+(CF*FF)+M, oF(b) per RG 1.99, Revision 2 233 196 Notes:
(a)
(b)
(c)
Initial RTNDT values are measured values for plate material.
This value was rounded per ASTM E29, using the "Rounding Method."
Based on Beaver Valley Unit 1 surveillance data. (Data not credible. ART calculated with a fill ca.)
PTLR Revision 0 4.2-18 LRM Revision 35
BVPS-1 LICENSING REQUIREMTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-8 Reactor Vessel Tou ess Data (Unirradiated)
Cu Ni P
TMT RTmT UPERSHLFENRG_(T-B COMPONENT HEAT NO.
CODE NO.
MATERIAL TYPE Ni P
Tr m (F)
UPPE SHL N M(-
(%/)
(%/)
(%/) (°F)
(0F)
MWD NMWD Closure Head C6213-IB B6610 A533B CL. I
.15
.010
.40 0*
121 Dome Clure Head Seg.
A5518-2 B6611 A533B CL 1
.14
.015
-20
-20*
131 ClosureHead ZV3758 A508 CL 2
.08
.007 60*
60*
>100 Flange__
Vessel Flange ZV3661 A508 CL 2
.12
.010 60*
60*
166 Inlet Nozzle 9-5443 A508 CL 2
.10
.008 60*
60*
82.5 Inlet Nozzle 9-5460 A508 CL 2
.10
.010 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 InterShell C4381-2 B6607-2 AS33BCL 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
.14
.57
.015
-20 20 129.5 83.5 Trans Ring 123V223 A508 CL.2 30 30*
143 Bottom HdSeg C4423-3 B6618 A533B CL
.13
.008
-30
-29*
124 Bottom Hd Dome C4482-1 B6619 A533B CL I
.13
.015
-50
-33*
125.5 IntertoLower 90136
.27
.07
-56
> 100 Shell Weld InterShellLong.
305424
.28
.63
-56
> 100 W eld Lower Shel Long.
305414
.34
.61
-56
__-40_-40_
-40 136.5
- Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Direction NMWD - Normal to Major Working Direction Note:
For evaluation of Inservice Reactor Vessel Irradiation damage assessments, the best estimate chemistry values reported in the latest response to Generic Letter 92-01 or equivalent document are applicable.
PTLR Revision 0 4.2-19 LRM Revision 35
BVPS-1 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-9 RTprs Calculation for Beltline Region Materials at EOL (28 EFPY)
Material Fluence FF CF a RTrr,,a, Margin RTNDT(b)
RTM (1 019 n/cm2, (OF)
(°F)
(°F)
(OF)
(OF)
E>1.0 MeV)
Intermediate Shell Plate B6607-1 3.54 1.329 100.5 133.6 34 43 211 Intermediate Shell Plate B6607-2 3.54 1.329 100.5 133.6 34 73 241 Lower Shell Plate B7203-2 3.54 1.329 98.7 131.2 34 20 185 Lower Shell Plate B6903-1 3.54 1.329 147.2 195.6 34 27 257 Using S/C Data(")
3.54 1.329 149.2 198.3 34 27 259 Inter. Shell Long. Weld 19-714A/B 0.708 0.903 191.7 173.1 65.5
-56 183
-+ Using S/C Data(e) 0.708 0.903 188.8 170.5 65.5
-56 180 Lower Shell Long. Weld 20-714A/B 0.708 0.903 210.5 190.1 65.5
-56 200 Using S/C Data(f 0.708 0.903 223.9 202.2 65.5
-56 212 Circumferential Weld 11-714 3.53 1.329 124.3 165.2 65.5
-56 175 Using S/C Data(d) 3.53 1.329 84.8 112.3 44
-56 101 Notes:
(a)
ARTFm = CF
- FF.
(b)
Initial RTNDT values of the plate material are measured values while the weld material values are generic.
(c)
RTm = RTNDT(J) + ARTpis + Margin (OF).
(d)
Based on credible St. Lucie Unit 1 surveillance data.
(e)
Based on non-credible Beaver Valley Unit 1 surveillance data with a full 0a&.
(f)
Based on non-credible Fort Calhoun Unit 1 surveillance data with a full c.
PTLR Revision 0 4.2-20 LRM Revision 35
BVPS-I LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-10 RTs Calculation for Beltline Region Materials at Life extension (45 EFPY)
Material Fluence FF CF A RTpTs(c)
Margin RTmTw()
RTprs' (1 0'9 n/cM2, (OF)
(OF)
(OF)
(OF)
(OF)
E>1.0 MeV)
Intermediate Shell Plate B6607-1 5.85 1.43 100.5 143.7 34 43 221 Intermediate Shell Plate B6607-2 5.85 1.43 100.5 143.7 34 73 251 Lower Shell Plate B7203-2 5.85 1.43 98.7 141.1 34 20 195 Lower Shell Plate B6903-1 5.85 1.43 147.2 210.5 34 27 272
-4 Using S/C Data(e) 5.85 1.43 149.2 213.4 34 27 274 Inter. Shell Long. Weld 19-714A/B 1.13 1.03 191.7 197.5 65.5
-56 207
-4 Using S/C Data(e) 1.13 1.03 188.8 194.5 65.5
-56 204 Lower Shell Long. Weld 20-714A/B 1.13 1.03 210.5 216.8 65.5
-56 226
-e Using S/C Data(f 1.13 1.03 223.9 230.6 65.5
-56 240 Circumferential Weld 11-714 5.82 1.43 124.3 177.7 65.5
-56 187
-+ Using S/C Data(d) 5.82 1.43 84.8 121.3 44
-56 109 Notes:
(a)
(b)
(c)
(d)
(e)
(f)
Initial RTNDT values of the plate material are measured values while the weld material values are generic.
RTPTS = RTNDT(U) + ARTpTs + Margin ( 0F).
ARTpTs= CF
- FF.
Based on credible St. Lucie Unit 1 surveillance data.
Based on non-credible Beaver Valley Unit 1 surveillance data with a full aA.
Based on non-credible Fort Calhoun Unit 1 surveillance data with a full a,.
PTLR Revision 0 4.2-21 LRM Revision 35
BVPS-2 LICENSING REQUIREMENTS MANUAL SECTION 4.2 PRESSURE AND TEMPERATURE LIMITS REPORT BVPS-2 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.1.2.2 N/A N/A 4.2-3 3.1.2.4 N/A N/A 4.2-3 3.4.1.3 N/A N/A 4.2-3 3.4.9.1 4.2.1.1 4.2-1 N/A 4.2-2 4.2-3 4.2-6 3.4.9.3 4.2.1.2 4.2-8 4.2-3 4.2.1.3 3.5.2 N/A N/A 4.2-3 3.5.3 N/A N/A 4.2-3 PTLR Revision 0 4.2-i LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
This Pressure and Temperature Limits Report (PTLR) for Unit 2 has been prepared in accordance with the requirements of Technical Specification 6.9.6. Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications (TS) addressed in, or make reference to, this report are listed below:
TS 3.1.2.2 Reactivity Control Systems - Flow Paths - Operating, TS 3.1.2.4 Reactivity Control Systems - Charging Pumps - Operating, TS 3.4.1.3 Reactor Coolant System - Shutdown, TS 3.4.9.1 Reactor Coolant System - Pressure/Temperature Limits, TS 3.4.9.3 Overpressure Protection Systems, TS 3.5.2 ECCS Subsystems - Tayg 2 3501F, and TS 3.5.3 ECCS Subsystems - Tavg < 3500F.
4.2.1 Operating Limits The PTLR limits for Beaver Valley Power Station (BVPS) Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in Reference 1 was used with two exceptions:
a)
Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1", and b)
Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure".
4.2.1.1 RCS Pressure and Temperature (PIt) Limits (TS 3.4.9.1)
The RCS temperature rate-of-change limits defined in Reference 2 are:
- a. A maximum heatup of 600F in any one hour period.
- b. A maximum cooldown of 1000F in any one hour period, and
- c. A maximum temperature change of less than or equal to 5F in any one hour period during inservice hydrostatic testing operations above system design pressure.
The RCS P/T limits for heatup, leak testing, and criticality are specified by Figure 4.2-1 and Table 4.2-1. The RCS P/T limits for cooldown are shown in Figure 4.2-2 through 4.2-6 and Table 4.2-2.
These limits are defined in Reference 2.
Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 4.2.1, the RCS P/T limits for heatup and cooldown shown in Figures 4.2-1 through and 4.2-6 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. The heatup and cooldown curves also include the effect of the reactor vessel flange.
PTLR Revision 0 4.2-1 LRM Revision 33
BVPS-2 LICENSING REUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT The heatup and cooldown curves are shown for 14 effective full power years (EFPY), although the capsule data provided in Tables 4.2-5 through 4.2-10 state 15 EFPY. The EFPY for the heatup and cooldown curves were generated for 15 EFPY. However, the heatup and cooldown curves were revised by license amendment 243 (1.4 % power uprate). The change from 15 to 14 EFPY was done to impose a conservative administrative limit due to the increased neutron fluence associated with the 1.4% increase in reactor power. The capsule data tables however, continue to reflect 15 EFPYs to retain actual capsule analysis results.
The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.
Pressure-temperature limit curves shown in Figure 4.2-7 were developed for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This figure provides the ASME III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known.
The temperature range provided bounds the expected operating range for an isolated loop and Code Case N-640.
4.2.1.2 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)
The power operated relief valves (PORVs) shall each have nominal maximum lift setting is in accordance with Figure 4.2-8. The OPPS enable temperature is in accordance with Table 4.2-3.
The PORV lift setting provided is for the case with reactor coolant pump (RCP) restrictions.
These restrictions are shown in Table 4.2-4, which is taken from Reference 10. 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 1900F. Therefore, the PORV setpoints shown in Table 4.2-4 will protect the Appendix G limits for the combinations shown.
The PORV setpoint is based on PIT limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 4.2.1. The PORV lift setting shown in Figure 4.2-8 accounts for appropriate instrument error.
4.2.1.3 OPPS Enable Temperature (TS 3.4.9.3)
Two different temperatures are used to determine the OPPS enable temperature, they are the arming temperature and the calculated enable temperature. The arming temperature (when the OPPS rendered operable) is established per ASME Section XI, Appendix G.
At this temperature, a steam bubble would be present in the pressurizer, thus reducing the potential of a water hammer discharge that could challenge the piping limits. Based on this method, the arming temperature with uncertainty is 236.50F.
PTLR Revision 0 4.2-2 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL 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 RTNDT + 501F), whichever is greater. The calculated enable temperature does not address the piping limit attributed to a water hammer discharge. The calculated enable temperature is 3500F.
As the calculated enable temperature is higher and, therefore, more conservative than the arming temperature, the OPPS enable temperature, as shown in Table 4.2-3, is set to equal the calculated enable temperature.
The calculation method governing the heatup and cooldown of the RCS requires the arming of the OPPS at and below the OPPS enable temperature specified in Table 4.2-3, and disarming of the OPPS above this temperature. The OPPS is required to be enabled, i.e., OPERABLE, when any RCS cold leg temperature is less than or equal to this temperature.
From a plant operations viewpoint the terms "armed" and "enabled" are synonymous when it comes to activating the OPPS. As stated in the applicable operating procedure, the OPPS is activated (armed/enabled) manually before entering the applicability of TS 3.4.9.3.
This is accomplished by placing two keylock switches (one in each train) into their "ARM" position.
Once OPPS is activated (armed/enabled) reactor coolant system pressure transmitters will signal a rise in system pressure above the variable OPPS setpoint. This will initiate an alarm in the control room and open the OPPS PORVs.
4.2.1.4 Reactor Vessel Boltup Temperature (TS 3.4.9.1)
The minimum boltup temperature for the Reactor Vessel Flange shall be 2 600F. Boltup is a condition in which the reactor vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.
4.2.2 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The capsule withdrawal schedule is provided in Table 5.3-6 of the UFSAR. Also, the results of these analyses shall be used to update Figures 4.2-1 through 4.2-6, and Tables 4.2-1 and 4.2-2. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.
The pressure vessel material surveillance program (References 3 and 4) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME,Section III, NB-233 1.
The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.
PTLR Revision 0 4.2-3 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Reference 12 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.
4.2.3 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 4.2-5, taken from Reference 2, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
Table 4.2-6, taken from Reference 2, provides the reactor vessel beltline material property table.
Table 4.2-7, taken from Reference 2, provides a summary of the Adjusted Reference Temperature (ARTs) for 15 EFPY.
Table 4.2-8, taken from Reference 2, shows the calculation of ARTs for 15 EFPY.
Table 4.2-9 shows the Reactor Vessel Toughness Data (Unirradiated).
Table 4.2-10, taken from Reference 6, provides RTpns values for 32 EFPY.
PTLR Revision 4.2-4 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 4.2.4 References
- 1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et al.,
January 1996.
- 2. WCAP-15139, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves During Normal Operation at 15 EFPY Using Code Case N-626," T. J. Laubham, January 1999.
- 3. WCAP-14484, Revision 0, "Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," P. A. Grendys, S. L. Anderson, J. F. Williams, February 1996.
- 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-14485, "Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation,"
P. A. Grendys, March 1986.
- 6. WCAP-14784, Revision 2, "Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2,"
T. J. Laubham, February 1996.
- 7. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.
- 8. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991. (PTS Rule)
- 9. Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials,"
U.S. Nuclear Regulatory Commission, May 1988.
- 10. Westinghouse Report NPD-OPES(99)-055, "Low Temperature Overpressure Protection System Setpoint Review for Beaver Valley Unit 2 15 EFPY Heatup and Cooldown Curves," March 1999.
- 11. Westinghouse Letter FENOC-01-261, COMS Arming Temperature", E. A. Dzenis, September 10, 2001.
- 12. FirstEnergy Nuclear Operating Company letter L-01-157, "Supplement to License Amendment Requests Nos. 295 and 167," dated December 21, 2001.
PTLR Revision 0 4.2-5 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INITIAL RTNDT:
INTERMEDIATE SHELL PLATE B9004-1 60°F RTNmi AFTER 14 EFPY:
1/4T, 140°F 3/4T, 128°F CURVES APPLICABLE FOR HEATUP RATES UP TO 60°F/HR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500 2250 2000 1750 A.
IL W 150 wo
- 0.
1250 0
C2o iooo 750 500 250 0
0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (F)
Figure 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 14 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-6 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INITIAL RTNDT:
INTERMEDIATE SHELL PLATE B9004-1 60°F RTmT AFTER 14 EFPY:
1/4T, 140°F 3/4T, 128°F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 0°F/HR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500-2250 2000 Unaccptab-Operation 1750__
- 5 1500
/
l
~~~~~~~~~~~~~A cptable poratlon 1250 LU 750 Cooldown Rate OF/hr.
750t 100 500 Boltup Temperatur 250-0 0
S0 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (F)
Figure 4.2-2 Reactor Coolant System Cooldown (up to 0°F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-7 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 INITIAL RTNry:
600F RTNur AFTER 14 EFPY:
1/4T, 1400F 3/4T, 1280F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 200F/HR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500 2250 2000 1750 a
co EL 1500
,,1 1250 uJ
!R 1 000 z
750 500 250 A-0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (F)
Figure 4.2-3 Reactor Coolant System Cooldown (up to 20'F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-8 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INTERMEDIATE SHELL PLATE B9004-1 INITIAL RTNDT:
600F RTNDT AFTER 14 EFPY:
1/4T, 1400F 3/4T, 1280F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 400F/HR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500 2250 2000 1750 to
- c. 1500 X 1250 Cl U,
o 1000 z
750 500 250 0
0 50 100 150 200 250 300 350 400 450 INDICATED TEMPERATURE (-F)
Figure 4.2-4 Reactor Coolant System Cooldown (up to 400F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 500 PTLR Revision 0 4.2-9 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INITIAL RTNDT:
INTERMEDIATE SHELL PLATE B9004-1 600F RTNM AFTER 14 EFPY:
1/4T, 1400F 3/4T, 1280F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 600F/HR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500 2250 2000 1750 6
S, 1500 W
W 1250 a.
Cla) 1000 aCr 750 500 250 a
0 50 100 150 200 250 300 350 400 450 INDICATED TEMPERATURE (F) 500 Figure 4.2-5 Reactor Coolant System Cooldown (up to 60VF/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-10 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
INITIAL RTNmr:
INTERMEDIATE SHELL PLATE B9004-1 600F RTNDT AFTER 14 EFPY:
1/4T, 1400F 3/4T, 1280F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 1000FIHR FOR THE SERVICE PERIOD UP TO 14 EFPY.
2500 2250 2000 1750 0:
OUJ 1250 IL 125 al NJ I-u 1000 750 500 250 0
0 50 100 150 200 250 300 350 400 450 S00 INDICATED TEMPERATURE (F)
Figure 4.2-6 Reactor Coolant System Cooldown (up to 1000F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1)
PTLR Revision 0 4.2-11 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2000 CD 1500 oo 00500 0o 50 60 70 80 90 100 110 TEMPERATURE (F) 120 Figure 4.2-7 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1)
PTLR Revision 0 4.2-12 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT 650 a
0)60 0
I-o 0
@?45 00 See Table 4.2-4 for RCP restrictions.I I
I Ira aT a
i a
a r -
I I
I I
I
~
~
~~I I- - -
a-i a
I I
I I
Operaato T
F I
i I
-- I L-
- - -Ac ep bl a a a a I
I I~~~~~~~~~~~~~~~~~~~~~~I a
-i-r-~
~~
~~-
J-*-
T 7
4 -
- - -L
- - -J
- - - - - a J - - -aL I
0 100 200 300 400 TRTC-AUCTIONEERED LOW-MEASURED RCS TEMPERATURE (F)
Figure 4.2-8 Maximum Allowable Nominal PORV Setpoint for the Overpressure Protection System (TS 3.4.9.3)
PTLR Revision 0 421 R
eiin3 4.2-13 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-1 Heatup Curve Data Points for 14 EFPY (TS 3.4.9.1) 600F/HR HEATUP 60OF/HR CRITICALITY LEAK TEST LIMIT Temp.
Press.
Temp.
Press.
Temp.
Press.
(O)
(psg (F)
(pig)sig) 60.00 621.00 196.00 0.00 178.00 2000.00 65.00 621.00 196.00 668.58 196.00 2485.00 85.00 621.00 196.00 693.10 90.00 621.00 196.00 687.74 95.00 621.00 196.00 686.96 100.00 621.00 196.00 689.93 105.00 621.00 196.00 696.41 110.00 621.00 196.00 705.98 115.00 621.00 196.00 718.57 120.00 621.00 196.00 734.02 120.00 794.02 196.00 752.40 X
125.00 812.40 196.00 773.69 130.00 833.69 196.00 798.06 135.00 858.06 196.00 825.62 140.00 885.62 196.00 856.64 145.00 916.64 196.00 891.32 150.00 951.32 196.00 930.00 155.00 990.00 200.00 973.00 160.00 1033.00 205.00 1020.74 165.00 1080.74 210.00 1073.65 170.00 1133.65 215.00 1132.24 175.00 1192.24 220.00 1197.06 180.00 1257.06 225.00 1268.73 185.00 1328.73 230.00 1347.95 190.00 1407.95 235.00 1435.48 195.00 1495.48 240.00 1532.16 200.00 1592.16 245.00 1638.92 205.00 1698.92 250.00 1756.81 210.00 1816.81 255.00 1886.95 215.00 1946.95 260.00 2030.62 X
220.00 2090.62 265.00 2189.19 225.00 2249.19 270.00 2364.21 230.00 2424.21 PTLR Revision 0 4.2-14 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-2 Cooldown Curve Data Points for 14 EFPY (TS 3.4.9.1)
O-F/HR. 20'---R. 400F/HR. 600F/HR. 100O'FiHk Temp.
Press.
Press.
Press.
Press.
Press.
(°F)~(psig)
(psig)
(psig)
(psig)
(psig) 60.00 621.00 621.00 621.00 607.90 531.77 65.00 621.00 621.00 621.00 618.15 543.64 70.00 621.00 621.00 621.00 621.00 556.89 75.00 621.00 621.00 621.00 621.00 571.67 80.00 621.00 621.00 621.00 621.00 588.13 85.00 621.00 621.00 621.00 621.00 606.46 90.00 621.00 621.00 621.00 621.00 621.00 95.00 621.00 621.00 621.00 621.00 621.00 100.00 621.00 622.00 621.00 621.00 621.00 105.00 621.00 621.00 621.00 621.00 621.00 110.00 621.00 621.00 621.00 621.00 621.00 115.00 621.00 621.00 621.00 621.00 621.00 120.00 621.00 621.00 621.00 621.00 621.00 120.00 907.20 883.94 862.00 841.62 806.75 125.00 935.36 914.39 894.99 877.44 849.30 130.00 966.47 948.05 931.50 917.13 896.50 135.00 1000.87 985.31 971.92 961.10 948.86 140.00 1038.88 1026.49 1016.66 1009.79 1006.92 145.00 1080.88 1072.06 1066.18 1063.72 1071.29 150.00 1127.31 1122.43 1120.97 1123.42 1123.42 155.00 1178.62 1178.62 1178.62 1178.62 1178.62 160.00 1235.32 1235.32 1235.32 1235.32 1235.32 165.00 1297.99 1297.99 1297.99 1297.99 1297.99 170.00 1367.25 1367.25 1367.25 1367.25 1367.25 175.00 1443.79 1443.79 1443.79 1443.79 1443.79 180.00 1528.38 1528.38 1528.38 1528.38 1528.38 185.00 1621.87 1621.87 1621.87 1621.87 1621.87 190.00 1725.19 1725.19 1725.19 1725.19 1725.19 195.00 1839.38 1839.38 1839.38 1839.38 1839.38 200.00 1965.58 1965.58 1965.58 1965.58 1965.58 205.00 2105.05 2105.05 2105.05 2105.05 2105.05 210.00 2259.18 2259.18 2259.18 2259.18 2259.18 215.00 2429.53 2429.53 2429.53 2429.53 2429.53 PTLR Revision 0 4.2-15 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3)
FUNCTION SETPOINT OPPS Enable Temperature I
3500 F PORV Se oint Fige 4.2-8 PTLR Revision 0 4.2-16 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.24 Reactor Coolant Pump Restrictions TRCS Running R I
< 190°F 0-2 2 190°F 3
PTLR Revision 0 4.2-17 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-5 Calculation of Chemistry Factors Using Surveillance Capsule Data(aXe)
Material Capsule Capsule f (b)
FF(c)
0.601 0.857 24.26 20.8 0.735 Intermediate Shell Plate B9004-2 (Longitudinal)
V 2.64 1.26 55.93 70.5 1.59 U
0.601 0.857 17.56 15.1 0.735 Intermediate Shell Plate B9004-2 (Transverse)
V 2.64 1.26 46.27 58.3 1.59 SUM:
164.7 4.66 CF = (FF*RTNDT) + ;(FF2 ) = 35.3 U
0.601 0.857 3.64 3.1 0.735 Weld Metal V
2.64 1.26 25.47 32.1 1.59 SUM:
35.2 2.32 CF = I(FF*RTNDT) + £(F2) = 15.2 Notes:
(a)
Regulatory Guide 1.99, Revision 2, Position 2.1.
(b) f= fluence (1019 n/cm2); Fluence values were taken from Capsule V analysis (Reference 4).
(c)
FF = fluence factor = f(0.28-o.l *log t)
(d)
ARTmT values obtained from CVGRAPH Version 4.0.
(e)
See Section 4.2.1.1 for a discussion of EFPY.
PTLR Revision 0 4.2-18 LRM Revision 33
I BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-6 Reactor Vessel Beltline Material Properties ()
Material Method Used Average Average Chemistry Initial RTNDT(b)
To aculate Cu wt %
Ni wt%
Factor (0F)(
0F)
Closure Head Flange N/A N/A 0.74 N/A
-10 Vessel Flange N/A N/A 0.73 N/A 0
Intermediate Shell Position 1.1 0.065 0.55 44.0 60 Plate B9004-1 Intermediate Shell Position 1.1 0.06 0.57 37.0 40 Plate B9004-2 Position 2.1 N/A N/A 35.3 40 Lower Shell Plate Position 1.1 0.08 0.58 51.0 28 B9005-1 Lower Shell Plate Position 1.1 0.07 0.57 44.0 33 B9005-2 Weld Metal Position 1.1 0.05 0.07 34.1
-30 (Longitudinal &
Position 2.1 N/A N/A 15.2
-30 Circumferential Seams)
Notes:
(a) Regulatory Guide 1.99, Revision 2, Position.
(b) Initial RTNDT values of the base metal and weld metal materials are measured values.
(c) See Section 4.2.1.1 for a discussion of EFPY.
PTLR Revision 0 4.2-19 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-7 Summary of Adjusted Reference Temperature (ARTs) for 15 EFPY )
MATERIAL DESCRIPTION Method Used To 15 EFPY ART Calculate the CF(a) 1/4T ART(F) 3/4T ART(0F)
Intermediate Shell Plate B9004-1 Position 1.1 140 128 Intermediate Shell Plate B9004-2 Position 1.1 112 97 Position 2.1 94 84 Lower Shell Plate B9005-1 Position 1.1 115 101 Lower Shell Plate B9005-2 Position 1.1 112 101 Longitudinal Welds Position 1.1 19 3
(located at 450 azimuthal angle)
Position 2.1
-8
-15 Circumferential Weld Position 1.1 41 23 Position 2.1 1
-7 Notes:
(a)
Regulatory Guide 1.99, Revision 2.
(b)
See Section 4.2.1.1 for a discussion of EFPY.
PTLR Revision 0 4.2-20 LRM Revision 33
BVPS-2 LICENSING REOUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-8 Calculation of Adjusted Reference Temperatures (ARTs) for 15 EFPY (b)
PARAMETER VALUES Operating Time 15 EFPY Material - Intermediate Shell Plate B9004-1 B9004-1 Location 1/4T ART 3/4T ART Chemistry Factor, CF (F) 44.0 44.0 Fluence, (f), (10'9 n/cm) 1.13 0.439 Fluence Factor, FF 1.03 0.771 ARTmT = CF x FF(0F) 45.5 33.9 Initial RTNDT, IF) 60 60 Margin, M(0F) 34 33.9 ART, per Regulatory Guide 1.99, Revision 2 140 128 Notes:
(a)
Fluence (f), is based upon f urf (10'9 n/cm2, E > 1.0 MeV) = 1.81 at 15 EFPY.
(b)
See Section 4.2.1.1 for a discussion of EFPY.
The Beaver Valley Unit 2 reactor vessel wall thickness is 7.875 inches at the beltline region.
PTLR Revision 0 4.2-21 LRM Revision 33
I BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-9 Reactor Vessel Toughness Data (Unirradiated)
'4 MATERIAL TMT 50 FT/ILB RTmn USE COMPONENT CODE NO.
SPEC. NO.
Cu %
Ni %
P %
OF35 MIL OF FT-LBS.
TEMP 0F Closure Head Dome B9008-1 A533B, CL. 1
.13
.54
.013
-20 50
-10 137 Closure Head Flange B9002-1 A508, CL. 2
.74
.012
-10
<40
-10 136 Vessel Flange B9001-1 A508, CL. 2
.73
.010 0
<10 0
132.5 Inlet Nozzle B901 1-1 A508, CL.-2
.88
.006 0
<10 0
104 Inlet Nozzle B901 1-2 A508, CL.2
.88
.010 10
<10 10 115 Inlet Nozzle B9011-3 A58, CL.2-
.84
.009 20
<40 20 122 Outlet Nozzle B9012-1 A508, CL.2
.71
.007
-10
<0
-10 137 Outlet Nozzle B9012-2 A508, CL.2
.74
.006
-10
<0
-10 121 Outlet Nozzle B9012-3 A508, CL.2
.68
.008
-10
<0
-10 112 Nozzle Shell B9003-1 A533B, CL 1
.13
.61
.008
-10 110 50 91 Nozzle Shell B9003-2 A533B, CL. 1
.12
.58
.009 0
120 60 79.5 Nozzle Shell B9003-3 A533B, CL. 1
.13
.61
.008
-10 110 50 97.5 Inter. Shell B9004-1 A533B, CL. 1
.07
.53
.010 0
120 60 83 Inter. Shell B9004-2 A533B, CL. 1
.07
.59
.007
-10 100 40 75.5 Lower Shell B9005-1 A533B, CL. 1
.08
.59
.009
-50 88 28 82 Lower Shell B9005-2 A533B, CL. 1
.07
.58
.009
-40 93 33 77.5 Bottom Head Torus B9010-1 A533B, CL. 1
.15
.49
.007
-30 56
-4 97 Bottom Head Dome B9009-1 A533B, CL. 1
.14
.53
.007
-30 35
-25 116 Weld (Inter. & Lower Shell Long. Seams & Girth Seam)*
.08
.07
.008
-30
<30
-30 144.5 HAZ (Plate B9004-2)
I--
-80 40
-20 76 Same heat of wire and lot of flux used in all seams including surveillance weldment (1) For evaluation of Inservice Reactor Vessel Irradiation damage assessments, the best estimate chemistry values reported in the latest response to Generic Letter 92-01 or equivalent document are applicable.
(2) See Section 4.2.1.1 for a discussion of EFPY.
PTLR Revision 0 4.2-22 LRM Revision 33
BVPS-2 LICENSING REQUIREMENTS MANUAL PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2-10 RTpTs Calculation for Beltline Region Materials at EOL (32 EFPY) (d)
Material Method f (a)
FF(b)
CF A RTPmr Margin RTNDTJ ()
RTprs Fluence (OF)
(OF)
(OF)
(OF)
(OF)
Intermediate Shell Plate B9004-1 RG 1.99, R2, P1.1 3.85 1.348 44.0 54.6 34 60 149 Intermediate Shell Plate B9004-2 RG 1.99, R2, P1.1 3.85 1.348 37.0 49.9 34 40 124 RG 1.99, R2, P2.1 3.85 1.348 35.3 47.6 17 40 105 Lower Shell Plate B9005-1 RG 1.99, R2, P1.1 3.85 1.348 51.0 68.8 34 28 131 Lower Shell Plate B9005-2 RG 1.99, R2, P1.1 3.85 1.348 44.0 59.3 34 33 126 Circumferential Weld RG 1.99, R2, P1.1 3.85 1.348 34.1 45.9 45.9
-30 62 RG 1.99, R2, P2.1 3.85 1.348 15.2 20.5 20.5
-30 11 Longitudinal Weld RG 1.99, R2, P1.1 1.21 1.053 34.1 35.9 35.9
-30 42 RG 1.99, R2, P2.1 1.21 1.053 15.2 16.0 16.0
-30 2
Notes:
(a) f= peak clad/base metal interface fluence (10'9 n/cm2, E>1.0 MeV) at 32 EFPY (450 fluence for longitudinal welds)
(b)
FF = f (0.28 -0.10lg OS (c)
RTNDTU) values are measured values.
(d)
See Section 4.2.1.1 for a discussion of EFPY.