L-13-360, Pressure and Temperature Limits Report Revision

Pressure and Temperature Limits Report Revision

ML13344A983
Person / Time
Site:	Beaver Valley
Issue date:	12/09/2013
From:	Emily Larson FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-13-360
Download: ML13344A983 (32)

Text

Beaver Valley Power Station P.O. Box 4 Sh~pmgpo~ PA 15077 FirstEnergy Nuclear Operating Company Eric A. Larson 724-682-5234 Site Vice President Fax: 724-643-8069 December 9, 2013 L-13-360 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Pressure and Temperature Limits Report Revision Pursuant to the requirements of Beaver Valley Power Station, Unit No. 2 (BVPS-2)

Technical Specification (TS) 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," FirstEnergy Nuclear Operating Company (FENOC) hereby submits the BVPS-2 PTLR, Revision 6. TS Section 5.6.4.c requires that the PTLR be provided to the Nuclear Regulatory Commission (NRC) upon issuance for any revision or supplement thereto.

The BVPS-2 pressure-temperature (P-T) limit curves for normal heatup and cooldown of the primary reactor coolant system were previously developed for 22 effective full power years (EFPY), 30 EFPY, 40 EFPY, and 54 EFPY. The enclosed BVPS-2 PTLR, Revision 6, revised the PTLR P-T limit curves from 22 EFPY to 30 EFPY.

There are no regulatory commitments contained in this letter. If there are any questions, or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at (330) 315-6810.

Eric A. Larson

Enclosure:

Beaver Valley Power Station, Unit No. 2, Pressure and Temperature Limits Report, Revision 6

Beaver Valley Power Station, Unit No. 2 L-13-360 Page2 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site Representative (BRP/DEP)

Enclosure L-13-360 Beaver Valley Power Station, Unit No. 2 Pressure and Temperature Limits Report, Revision 6 (29 Pages Follow)

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Pressure and Temperature Limits Report BVPS-2 Technical Specification to PTLR Cross-Reference Technical PTLR Specification Section Figure Table 3.4.3 5.2.1.1 5.2-1 N/A 5.2-2 5.2-3 5.2-4 5.2-5 5.2-6 3.4.6 N/A N/A 5.2-3 3.4.7 N/A N/A 5.2-3 3.4.10 N/A N/A 5.2-3 3.4.12 5.2.1.2 5.2-8 5.2-3 5.2.1.3 3.5.2 N/A N/A 5.2-3 BVPS-2 Licensing Requirement to PTLR Cross-Reference Licensing PTLR Requirement Section Figure T~ble 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 6 Beaver Valley Unit 2 5.2- i LRM Revision 78

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 (Pff)

Limits,

2. LCO 3.4.6 RCS Loops - MODE 4,

3. LCO 3.4. 7 RCS Loops - MODE 5, Loops Filled,

4. LCO 3.4.1 0 Pressurizer Safety Valves,

5. LCO 3.4.12 Overpressure Protection System (OPPS),

6. LCO 3.5.2 ECCS - Operating,

7. LR 3.1.2 Boration Flow Paths- Operating,

8. LR 3.1.4 Charging Pump- Operating, and

9. LR 3.4.6 Pressurizer Safety Valve Lift Involving 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 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 (Pff) 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 6 Beaver Valley Unit 2 5.2- 1 LRM Revision 78

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 PIT limits for heatup, leak testing, and criticality are specified by Figure 5.2-1 and Table 5.2-1. The RCS PIT limits for cooldown are shown in Figures 5.2-2 through 5.2-6 and Table 5.2-2. These limits are defined in Reference 14. Consistent with the methodology described in Reference 1, including the exceptions as noted in Section 5.2.1, the RCS PIT limits for heatup and cooldown shown in Figures 5.2-1 through 5.2-6 are provided without margins for instrument error. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. The heatup and cooldown curves also include the effect of the reactor vessel flange.

The PIT limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding PIT curve for heatup and cooldown.

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.

Reference 13 provides an updated surveillance capsule credibility evaluation, updated Position 2.1 chemistry factor values, and an updated fluence evaluation.

Therefore, the applicability of the PIT limit curves (Reference 14) was assessed based on the revised information. Taking into account the updated surveillance data credibility evaluation, the Position 2.1 chemistry factor values, and the fluence analysis summarized in Reference 13, the limiting material for the current BVPS-2 PIT limits continues to be the intermediate shell plate 89004-1 at 30 EFPY.

PTLR Revision 6 Beaver Valley Unit 2 5.2-2 LRM Revision 78

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 value at 30 EFPY is 3. 03 x 1019 n/cm 2 (E > 1. 0 MeV). This value was calculated by interpolating the fluence at the aa azimuthal position for BVPS-2 from the end of Cycle 15 to the fluence value at the future projection out to 32 EFPY. The fluence of 3.39 x 1019 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 conservative than the updated fluence of 3.03 x 1019 n/cm 2 (E > 1.0 MeV).

5.2.1.2 Overpressure Protection System COPPS) Setpoints CLCO 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 PIT limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1, including the exceptions noted in Section 5.2.1. The PORV lift setting shown in Figure 5.2-8 accounts for appropriate instrument error.

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 6 Beaver Valley Unit 2 5.2-3 LRM Revision 78

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 RTNOT + 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 6 Beaver Valley Unit 2 5.2-4 LRM Revision 78

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, RTNoT, which is determined in accordance with ASME, Section Ill, NB-2331. The empirical relationship between RTNoT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E 185-82.

Reference 10 is an NRC commitment made by FENOC to use only the calculated vessel fluence values when performing future capsule surveillance evaluations for BVPS Unit 2. This commitment is a condition of License Amendment 138 and will remain in effect until the NRC staff approves an alternate methodology to perform these evaluations. Best-estimate values generated using the FERRET Code may be provided for information only.

5.2.3 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the PIT limits.

Table 5.2-5, taken from Table 2-4 of Reference 13, shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5.2-6, taken from Table 2-1 of Reference 14, provides the reactor vessel beltline material property table.

Table 5.2-7, taken from Table 4-2 of Reference 13, provides the reactor vessel extended beltline material property table.

Table 5.2-8, taken from Tables 4-7 and 4-8 of Reference 14, provides a summary of the Adjusted Reference Temperature (ARTs) for 30 EFPY.

PTLR Revision 6 Beaver Valley Unit 2 5.2-5 LRM Revision 78

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 PTs 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 Develpp 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 6 Beaver Valley Unit 2 5.2-6 LRM Revision 78

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-LTR-13-19, Revision 0, dated March 6, 2013, "Acceptable Initial RTNor Values for the Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle Materials."

PTLR Revision 6 Beaver Valley Unit 2 5.2-7 LRM Revision 78

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE 89004-1 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.

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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 6 Beaver Valley Unit 2 5.2-8 LRM Revision 78

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.

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Reactor Coolant System Cooldown (steady state - 0°F/Hr.)

Limitations Applicable for the First 30 EFPY (LCO 3.4.3)

PTLR Revision 6 Beaver Valley Unit 2 5.2-9 LRM Revision 78

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.

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Limitations Applicable for the First 30 EFPY (LCO 3.4.3)

PTLR Revision 6 Beaver Valley Unit 2 5.2- 10 LRM Revision 78

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.

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Limitations Applicable for the First 30 EFPY (LCO 3.4.3)

PTLR Revision 6 Beaver Valley Unit 2 5.2- 11 LRM Revision 78

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.

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Limitations Applicable for the First 30 EFPY (LCO 3.4.3)

PTLR Revision 6 Beaver Valley Unit 2 5.2- 12 LRM Revision 78

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.

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Limitations Applicable for the First 30 EFPY (LCO 3.4.3)

PTLR Revision 6 Beaver Valley Unit 2 5.2- 13 LRM Revision 78

Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 2500 I

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PTLR Revision 6 Beaver Valley Unit 2 5.2- 14 LRM Revision 78

LicensingRequirementsManual PressureandTemperature LimitsReport 5.2 SeeTable5.2-4for RCPrestrictions.

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PTLR Revision6 BeaverValleyUnit 2 5 . 2- 1 5 LRM Revision78

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) (psi g) (oF) (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 6 Beaver Valley Unit 2 5.2- 16 LRM Revision 78

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) 0°F/HR 20°F/HR 40°F/HR 60°F/HR 100°F/HR Temp. Press. Press. Press. Press. Press.

(oF) (psi g) (psig) (psig) (psi g) (psig) 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 6 Beaver Valley Unit 2 5.2- 17 LRM Revision 78

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 6 Beaver Valley Unit 2 5.2- 18 LRM Revision 78

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

Reactor Coolant Pump Restrictions TRcs Running RCPs

< 137°F 0-2

137°F 3 PTLR Revision 6 Beaver Valley Unit 2 5.2- 19 LRM Revision 78

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 f<al FF(bl .6.RTNoT(c) FF*L1RTNOT 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)

X 5.63 1.425 98.0 139.65 2.031 Intermediate u 0.615 0.864 17.7 15.29 0.746 Shell Plate v 2.64 1.260 46.1 58.07 1.587 B9004-2

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

• 1o9 f).

(c) .6.RTNOT values are the measured 30 ft-lb shift values. The BVPS-2 .6.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 6 Beaver Valley Unit 2 5.2-20 LRM Revision 78 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 RTNoT 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) Notes: (a) The initial RTNoT 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 pressure 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 copper) 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 RT NoT 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 6 Beaver Valley Unit 2 5.2- 21 LRM Revision 78 Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-7 (Page 1 of 1) Reactor Vessel Extended Beltline Material Properties (a) Initial Material Material Wt% Wt% Description ID Heat Number Cu Ni RTN~T (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 51922 (3489) 0.05 1.00 -56 (Gen) Intermediate Shell 103-121 AAGC 0.03 0.98 -70 Girth Weld KOIB 0.03 0.97 -60 89011-1 2V2436-0 1-002 0.11 0.85 60\C) Inlet Nozzles 89011-2 2V2437-02-001 0.13 0.88 60101 (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 -60 105-121 c 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 0.74 -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 RTNDT values were determined in accordance with NUREG-0800 and the 10 CFR 50.61. (c) As described in Reference 16, the reactor vessel initial RTNDT values for the inlet nozzles are conservatively assigned values. The actual initial RTNDT values for the reactor vessel inlet nozzles are located in BVPS-2 UFSAR Table 5.3-1. PTLR Revision 6 Beaver Valley Unit 2 5.2-22 LRM Revision 78 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(bl 1/4T ART CF) 3/4T ART CF) 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 Notes: (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 6 Beaver Valley Unit 2 5.2-23 LRM Revision 78 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(a) PARAMETER VALUES Operating Time 30 EFPY Material - Intermediate Shell Plate B9004-1 B9004-1 Location 1/4T 3/4T Chemistry Factor, CF CF) 40.5 40.5 Fluence, (f), (1 019 n/cm 2)(b) 2.113 0.8215 Fluence Factor, FF 1.203 0.9448 ~RT NOT= CF X FF(°F) 48.74 38.27 lntitial RTNoT, ICF) 60 60 Margin, M(°F) 34 34 ART, per Regulatory Guide 1.99, Revision 2 143 132 Notes: (a) Table reflects Capsule X analysis per Reference 14. (b) = Fluence (f), is based upon fsurf (1 019 n/cm 2 , 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 6 Beaver Valley Unit 2 5.2-24 LRM Revision 78 Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-1 0 (Page 1 of 1) RTPTs Calculation for Beltline Region Materials at Life Extension (54 EFPY)ca> Surface Neutron Fluence Chemistry Initial ~RTPTS(d) Margin(e) RTPTS(f) Material Description Material Heat RTNOT(c) au aLl Fluence Factor, Factor ID Number FFCbl (oF) (oF) (oF) (oF) (oF) (x1 0 19 n/cm 2 ) CF) CF) 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 ~--~------------------------------------------------------~ ---------- ------------------------ ------------- --------------- ------------ ------------- ------- ------- ------------- ---146-.4-- I ~ Using non-credible surveillance data<9> 1.4092 40 72.4 17 34 1 5.18 51.4 0 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 data<9> 5.18 1.4092 12.5 -30 17.6 0 8.8 17.6 5.2 Intermediate Shell 101-124 83642 1.76 1.1554 34.4 -30 39.7 0 19.9 39.7 49.5 Longitudinal Welds A&B ~ Using credible surveillance data<9> 1.76 1.1554 12.5 -30 14.4 0 7.2 14.4 -1.1 Lower Shell 101-142 83642 1.77 1.1569 34.4 -30 39.8 0 19.9 39.8 49.6 Longitudinal Welds A&B ~ Using credible surveillance data<9> 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 = fluence factor = f co.2a- 0*1 109 (f)). (c) Initial RTNOT values are measured values. (d) ~RT PTS = CF

• FF.

(e) M = 2 *(cru2 + crl) 112 * (f) RTPTS = Initial RTNOT + ~RTPTs + Margin. (g) The BVPS-2 surveillance weld metal is the same weld heat as the BVPS-2 beltline welds (heat 83642). The BVPS-2 surveillance weld data is credible; therefore, the reduced all term of 14°F was utilized for BVPS-2 weld heat 83642. The BVPS-2 surveillance plate material is representative of the BVPS-2 intermediate shell plate 89004-2. The surveillance plate material is non-credible; therefore, the higher all term of 1rF was utilized for BVPS-2 plate 89004-2. The credibility evaluation conclusions are contained in Appendix A of Reference 13. PTLR Revision 6 Beaver Valley Unit 2 5.2-25 LRM Revision 78 Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 1 of 3) RTPTs Calculation for Extended Beltline Region Materials at Life Extension (54 EFPY)(a) Surface Neutron Fluence Chemistry Initial ,1.RTPTS(e) Margin(fl RTPTS(g) Material Description Heat Number Fluence Factor, Factor RTNDT(c) cru 0'6, MateriaiiD (Lot Number) FF(bl (oF) (oF) (oF) (oF) (oF) (x1 019 CF) (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 BOHB 0.515 0.8147 68.0 10(d) 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 6 Beaver Valley Unit 2 5.2-26 LRM Revision 78 Licensing Requirements Manual Pressure and Temperature Limits Report 5.2 Table 5.2-11 (Page 2 of 3) RTPTs Calculation for Extended Beltline Region Materials at Life Extension (54 EFPY)Cal Surface Material Neutron Fluence Chemistry Initial ~RTPTS(e) MarginCfl RTPTS(g) Heat Number Fluence Factor, Factor RTNDT(c) cru cr"' Description MateriaiiD (Lot Number) FFCbl (oF) (oF) (oF) (oF) (oF) (x1 019 (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 -62.1 105-1218 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 FAGS 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 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 -68.2 107-121 B Welds HABJC 0.0151 0.1440 27.0 -70 3.9 0 1.9 3.9 -62.2 107-121C HAGS 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 6 Beaver Valley Unit 2 5.2-27 LRM Revision 78 Licensing Requirements Manual Pressure andTemperature LimitsReport 5.2 Table 5.2-11(Page3 of 3) RTprsCalculationfor ExtendedBeltlineRegionMaterialsat Life Extension(54 EFPY;t"r Notes: (a) Data obtainedfrom Table6-4 of Reference13. - 0'1rog (0). (b) FF = fluencefactor= 1(0'24 (c) InitialRTruor valuesare measuredvalues,unlessothenrvise noted. (d) valuesare generic. InitialRTruor (e) ARTprs = CF

• FF.

(0 M=Z*(ou'+ oo')t. (g) RTprs= InitialRTnot+ ARTprs+ Margin. (h) As describedin Reference16, the reactorvesselinitialRTltorvaluesfor the inletnozzles valuesfor the reactorvessel are conservativelyassignedvalues. The actual initialRTr.ror inletnozzlesare locatedin BVPS-2UFSARTable5.3-1. PTLR Revision6 BeaverValleyUnit 2 5.2- 28 LRM Revision78