ML17334A105

Reactor Coolant System Pressure and Temperature Limits Report

ML17334A105
Person / Time
Site:	Diablo Canyon
Issue date:	11/29/2017
From:	Hamzehee H Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-17 -099
Download: ML17334A105 (39)

Text

Pacific Gas and Electric Company" Diablo Canyon Power Plant Hossein G. Hamzehee Manager Mail Code 104/5/535 Regulatory Services P.O. Box 56 Avila Beach, CA 93424 805.545.4720 Internal: 691 .4720 Fax: 805.545. 3459 November 29, 2017 PG&E Letter DCL-17 -099 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Reactor Coolant System Pressure and Temperature Limits Report for Units 1 and 2

Dear Commissioners and Staff:

In accordance with Diablo Canyon Power Plant Technical Specification 5.6.6.c, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

Pacific Gas and Electric Company (PG&E) is submitting the enclosed Revision 16 of the PTLR for Units 1 and 2, dated July 31, 2017.

PG&E makes no new or revised regulatory commitments in this submittal (as defined by NEI 99-04).

If there are any questions regarding the PTLR, please contact Ms. Candice Chou at (805) 545-6164.

in G. Hamzehee dqmg/6192/50943731 Enclosure cc: Diablo Distribution cc/enc: Kriss M. Kennedy, NRC Region IV Administrator Christopher W. Newport, NRC Senior Resident Inspector Balwant K. Singal, NRC Senior Project Manager A member of the STARS Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Enclosure PG&E Letter DCL-17-099 PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)-1 REVISION 16 EFFECTIVE DATE: July 31, 2017

• • • ISSUED FOR USE BY: _ _ _ _ _ _ _ _ _ DATE: _ _ _ _ _ EXPIRES: _ _ _ _ _ ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 NUCLEAR POWER GENERATION REVISION 16 DUffiLOCANYONPOWERPLANT PAGE 1 OF 37 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE: PTLR for Diablo Canyon 1 AND2 07/31/17 EFFECTIVE DATE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).2 OPERATING LIMITS ....................................................................................................................................... 2 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) ............................................................................ 2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12) .................................. 5 ADDITIONAL CONSIDERATIONS ............................................................................................................. 16 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM .............................................................. 16 REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY .................................................................. 18 SUPPLEMENTAL DATA TABLES .............................................................................................................. 24 PRESSURIZED THERMAL SHOCK (PTS) SCREENING .......................................................................... 25 REFERENCES ................................................................................................................................................ 25 List of Figures Figure PAGE 2.1-1 Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 9 60°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) 2.1-2 Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 12 0, 25, 50, 75 and 100°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

List ofTables Table 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2) With Margins for 10 Instrumentation Errors 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2) With Margins for 13 Instrumentation Errors 2.2-1 LTOP System Setpoints 15 2.2-2 LTOP Temperature Restrictions 15 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data 20 5.0-2 Diablo Canyon Unit 1 & Palisades Unit 1 Surveillance Capsule Data 21 5.0-3 Diablo Canyon Unit 2 Surveillance Capsule Data 22

_y, 5.0-4 Farley Unit 1 and Calvert Cliffs Unit 1 Surveillance Capsule Data 24 PTLR-1u3r16.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 2 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2

1. REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LllviiTS REPORT (PTLR)

This PTLR for Diablo Canyon has been prepared in accordance with the requirements of Technical Specification CI'S) 5.6.6. The TS addressed in this report are listed below:

• LC03.4.3 RCS Pressure and Temperature (PIT) Limits

• LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems The limits provided in this report remain valid until35 EFPY on Unit 1 and Unit 2.

2. OPERATING LIMITS 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)

The RCS temperature rate-of-change limits are:

• A maximum heatup of 60°F in any 1-hour period.

• A maximum cool down of 100°F in any 1-hour period.

• A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Tables 2.1-1 and 2.1-2.

As documented in the Reference 8.12 evaluation, the RCS pressure and temperature conditions implemented during the Vacuum Refill process per procedure OP A-2:IX (Ref. 8.11) remain bounded by the RCS PIT limits as shown in Figure 2.1-1 and Figure 2.1-2, and the LTOP PIT limits established in Section 2. The RCS Vacuum Refill restricts RCS pressure criteria to values above 0 psia to ensure RHR system operability.

2.1.1 RCS PIT Limits:

The parameter limits for the specifications listed in section 1 are presented in the following subsections. The limits were developed using a methodology that is in accordance with the NRC approved methodology provided in WCAP 14040-NP-A (Ref. 8.4). The analysis methods implemented per ASME B&PV Code Section III Appendix G utilize linear elastic fracture mechanics, determine the maximum permissible stress intensity correlated to the reference stress intensity (KrR) as a function of vessel metal temperature, define the size of the assumed flaw, and apply specified safety factors.

The reference stress intensity (KrR) is the combined thermal and pressure stress intensity limit at a given temperature. The assumed crack has a radial depth of Y<1 of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 3 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 10 CFR 50 Appendix G and Reg. Guide 1.99 provide guidelines for determining the maximum permissible (allowable) stress intensity, based on nil-ductility of the reactor vessel metals during the operational life of the reactor. The transition temperature at which the metal becomes acceptably ductile is affected by neutron radiation embrittlement over the course of reactor operation. Appendix G and Reg. Guide 1.99 provide formulas which are used to calculate this Adjusted Reference Temperature based on fluence and vessel material chemistry. The shift in nil-ductility resulting from the fluence effect is added to the unirradiated nil-ductility transition temperature and, with Reg.

Guide 1.99 defined margins included, the Adjusted Referenced Temperature (ART) is established for a specified neutron fluence.

The allowable stress intensity is determined from ASME Code formula and is based on the difference between any given vessel metal temperature and the ART.

The thermal stress intensities were provided by Westinghouse (Appendix A to PG&E Technical & Ecological Services - TES -Letter file no. 89000571 -

Chron. no. 126962- RLOC 04014-1712) over the 70 deg to 550 deg range for various heat up and cool down rates. The stress intensities are dependent on geometry and temperature change rate and are not affected by embrittlement.

Thus, the Westinghouse provided values remain valid throughout Plant life.

The membrane (pressure induced) stress can then be determined as a function of the allowable stress intensity reduced by thermal stress intensity and that difference divided by 2 as specified in ASME Section III Appendix G. Several safety factors and conservative assumptions are incorporated into the calculation process for determining the remaining allowable pressure stress.

The RCS pressure that imposes this Pressure Stress can then be determined at the various temperatures. Note that during heatup the Thermal Stress can be offset by the pressure stress on an internal crack and conversely during cooldown, the thermal stress can offset the pressure stress on an external crack.

The heat up and cooldown curves extract the values that are based on the highest magnitude combined stress at either the 1/4t or 3/4t location.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 4 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.1.2 RCS Pressure Test Limits:

10 CFR 50, Appendix G establishes the pressure and temperature requirements for pre-service hydrostatic test (no fuel) and hydro test and leak tests performed with fuel in the core.

To meet Condition 1.a of 10 CFR Appendix G, Table 1, the limiting temperature for the closure flange is the Unit 1 head flange that has an R TNDT of35°F. The 20% of pre-service system hydrostatic test pressure is 621 psig.

Thus, the minimum RCS temperature for the hydro tests and leak tests with fuel in the vessel and core not critical that do not exceed 621 psig pressure is 35°F. For Condition 1.b, the minimum RCS temperature for the hydro tests and leak tests with fuel in the vessel and core not critical that do exceed 621 psigpressure is 125°F (RTNnT+ 90°F). For Condition l.c, the limiting material is Unit 2 intermediate shell plate B5454-2 based on an ART of 197.8°F. For this pre-service hydro test, with no fuel in the vessel, the minimum RCS temperature for all pressures is 257.8°F (RTNnT+ 60°F). The limiting temperature for all these conditions is for Condition 1.c. Thus, the pressure temperature limits for leak testing are imposed starting with a minimum temperature of 260°F.

2.1 .3 Reactor Vessel Bolt-up and Criticality Temperature Limits:

Operating restrictions illustrated on the P-T curve also include reactor flange bolt up temperature. This is based on ASME Appendix G and 10 CFR 50 Appendix G that require the bolt-up temperature to be the initial RTNnT of the flange plus any irradiation effects. The flux exposed in the R.V. Flange and R.V. Head Flange result in negligible RTNDT shift, and, thus minimum Bolt Up Temperature does not change with time. The highest flange RTNDT between DCPP Unit 1 and 2 is 35 deg F (Unit 1 R.V. closure head). The curves conservatively set the temperature at 60 deg F based on WCAP 14040-NP-A minimum temperature. Between the minimum bolt up temperature and the minimum LTOP operating temperature (86 deg F), a 2.07 sq. in. opening is relied on for RCS venting. This satisfies Condition 2.a of the 10 CFR Appendix G, Table 1.

To comply with Condition 2.b of 10 CFR Appendix G, Table 1, the pressure temperature limits impose a minimum temperature of 155°F (RTNDT of 35°F + 120°F) at pressures not exceeding the 20% hydro test pressure or 621 psi g. These portions of the Figures 2.1-1 and 2.1-2 curves are graphically bounded by the heatup and cooldown curves and are not visible.

When the core is critical, the 10 CFR Appendix G, Table 1 Conditions 2.c and 2.d require that the temperature be at least 40°F greater than the corresponding ASME Appendix G limit. The minimum temperature for criticality is equal to the minimum temperature for the in-service system hydrostatic pressure of 2459 psig, which is 327.5°F. Thus, the minimum temperature at which the core may be critical is 330°F.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 5 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12)

The power-operated relief valves (PORVs) shall each have a lift setting and an arming temperature in accordance with Table 2.2-1.

Operation of plant equipment shall comply with the temperature restrictions of Table 2.2-2.

2.2.1 LTOP Enable Setpoints:

The LTOP lift setpoint and arming temperature are based on the methodology established in the Westinghouse WCAP- 14040- NP- A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996. The lift setpoint is 435 psig based on limiting the maximum RCS pressure overshoot to a value below the Appendix G PIT curve and limiting the minimum RCS undershoot to maintain a nominal operating pressure drop across the number one RCP seal.

The arming temperature setpoint is 200°F or RTNDT + 50°F whichever is greater in accordance with ASME Code Case N-514. The RETRAN-02 Mod3 computer code (Ref. 8.6) was used to perform the thermal hydraulic analysis and to ensure that the LTOP setpoints and temperature restrictions are acceptable as documented in the calculation STA-249 (Ref. 8.10) with input from STA-197 (Ref. 8.7) for Unit 1 and Unit 2 w/Replacement Steam Generators (RSG's).

2.2.2 RCS Pressure Overshoot:

The mass injection and heat injection events are assumed to occur with the RCS in water solid conditions and letdown isolated, so the RCS pressure rapidly increases to the PORV actuation setpoint. The RCS pressure continues increasing even after the PORV setpoint is reached until the PORV has sufficiently opened so that the relief capacity equals the RCS mass increase or volumetric expansion. The magnitude of the RCS pressure overshoot above the PORV setpoint is dependent on the mass injection and heat injection rates, and the associated PORV electronic delay time and valve opening time. The LTOP analysis assumes a conservative PORV lift setpoint, PORV opening time, and also includes appropriate instrumentation delays. Even considering the limiting single failure of one pressurizer PORV to open, there is still a qualified PORV available to adequately relieve the RCS system pressure.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 6 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 The RCS peak system pressure occurs at the bottom of the reactor vessel requiring that the elevation head be accounted for between this peak location and the RCS wide range pressure transmitters that generate the PORV open signal. In addition, the RHR pump and RCP flow impacts the PORV setpoint by generating a dynamic pressure drop across the reactor vessel which increases the difference between the RCS wide range pressure transmitters and the bottom of the reactor vessel. The magnitude of the total pressure drop determines the limiting RCS pressure at the bottom of the vessel for a given RCS overshoot case. An appropriate range of mass injection and heat injection cases are evaluated to ensure they conservatively bound the dynamic pressure drop effects due to the RCS flow conditions.

2.2.3 LTOP Mass Injection Case:

The LTOP mass injection analysis is based on an inadvertent initiation of the maximum injection flow capability for the applicable Mode of operation into a water solid RCS with letdown isolated. The initial mass injection capability within the LTOP range is established by Tech Spec. 3.4.12 restriction to secure the safety injection (Sl) pumps and one ECCS centrifugal charging pump (CCP), isolate all SI Accumulators, and align CCP 3 for LTOP operation prior to entering the LTOP mode of operation. The administrative temperature limit for blocking the SI signal is based on a mass injection case with one ECCS CCP and CCP 3 aligned for LTOP operation injecting through the SI injection flowpath. The administrative temperature limit for operating with a maximum of one charging pump is based on a mass injection case with one ECCS CCP (which bounds operation with CCP 3 aligned for LTOP operation) injecting through the normal and the alternate charging flowpaths. The administrative temperature limits for starting and stopping RCPs are based on limiting the dynamic pressure drop increase on the RCS overshoot for a mass injection case with one CCP injecting through the normal and alternate charging flowpaths.

The administrative temperature limit for establishing an RCS vent is based on determining the temperature at which the reduced Appendix G PIT limit no longer has additional margin to accommodate the mass injection RCS overshoot associated with the PORV response time. All mass injection cases account for a conservative RCP seal injection flow into the RCS and the dynamic effects of both RHR pumps running.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 7 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2.4 LTOP Heat Injection Case:

The heat injection cases are based on starting an RCP in one loop with a maximum allowable measured temperature difference of 50 op between the RCS and the Steam Generators (SGs). The heat injection cases are evaluated at various RCS temperature conditions which bound the potential volumetric expansion effects of water on the RCS overshoot within the LTOP range. The heat injection RCS overshoot cases were determined to remain below the Appendix G PIT curve and are conservatively bounded by the mass injection overshoot results throughout the LTOP temperature range. The heat injection cases establish that there are no LTOP administrative RCS temperature restrictions for starting an RCP when the measured SG temperature does not exceed the RCS by more than 50 °F. A bounding heat injection case was also evaluated to establish that if the pressurizer level indicates less than or equal to 50%, there are no RCSISG temperature restrictions for starting an RCP, since even the maximum credible RCSISG temperature differential will not challenge the Appendix G PIT limit in the LTOP range.

2.2.5 RCS Pressure Undershoot:

Once an LTOP PORV has opened to mitigate the pressure transient due to a mass injection or heat injection case, the RCS pressure continues decreasing even after the close setpoint has been reached and until the PORV has fully closed. The limiting RCS undershoot case is based on the maximum RCS pressure relief capacity associated with both LTOP PORVs opening and closing simultaneously during the least severe mass injection and heat injection overshoot case, respectively. The RCS undershoot evaluation is based on maintaining the RCS pressure above the minimum value which is considered acceptable for the number one RCP seal operating conditions. The PORV lift setpoint in Table 2.2-1 was evaluated to adequately limit the RCS undershoot to an acceptable value for the applicable mass injection and heat injection cases within the LTOP range.

Where there is insufficient range between the upper and lower pressure limits to select a PORV setpoint to provide protection against violation of both limits, setpoint selection to provide protection against the upper pressure limit violation shall take precedence.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 8 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2.6 Measurement Uncertainties:

The LTOP mass injection and heat injection overshoot analyses incorporate the appropriate measurement uncertainties associated with the RCS wide range pressure transmitters and the RCS wide range RTDs. Since these two measurement processes are independent of each other, they are statistically combined into one equivalent pressure error term with respect to the Appendix G PIT curve that is added onto the calculated peak pressure. This bounding peak pressure is then used to determine the corresponding temperature limit which ensures compliance with the applicable Appendix G PIT curve.

The heat injection case overshoot analysis also incorporates the measurement uncertainty associated with establishing the SG secondary temperature prior to starting an RCP. The RCS and SG measurement uncertainties are then assumed to be in the worst case opposite direction to establish a conservatively bounding RCSISG temperature difference for the heat injection analysis.

The LTOP mass injection and heat injection undershoot analyses incorporate the appropriate measurement uncertainty for the RCS wide range pressure transmitters associated with both PORV s opening and closing simultaneously.

Since each PORV has a normal and independent setpoint uncertainty distribution, they are statistically combined into a value which represents the lowest simultaneous drift setpoint with a 95% probability.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 9 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 50 150 200 2.50 350 RCS TEMPERATURE {F)

FIGURE 2.1-1: Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr)

Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 10 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2)

With Margins for Instrumentation Errors 25°F/hr 60°F/hr 60°F/hr Crit. Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(oF) (psi g) (oF) (psi g) (oF) (psi g) (oF) (psi g) 75 471.9 75 470.3 80 464.3 80 462.4 85 459.1 85 443.7 90 458.4 90 424.3 95 459.7 95 409.9 100 461.8 100 407.7 105 465.1 105 410.3 110 469.0 110 412.3 115 473.7 115 414.3 120 478.9 120 416.0 125 484.8 125 417.9 130 491.2 130 419.9 135 498.2 135 422.4 -

140 505.7 140 425.5 145 513.8 145 428.9 150 522.5 150 433.1 155 531.9 155 437.0 160 541.9 160 442.7 165 552.7 165 449.1 170 564.2 170 455.5 175 576.5 175 462.0 180 589.8 180 470.4 185 604.0 185 479.7 190 618.7 190 489.1 195 633.8 195 499.3 200 648.9 200 510.3 205 664.3 205 522.1 210 680.8 210 535.0 215 698.4 215 548.7 220 717.4 220 563.5 225 737.7 225 579.2 230 759.4 230 596.4 235 782.7 235 614.8 240 807.6 240 634.3 PTLR-lu3r16.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 11 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2)

With Margins for Instrumentation Errors 25° F/hr 60°F/hr 60°F/hr Crit. Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press.

CF) (psi g) (oF) (psi g) (oF) (psig) (oF) (psi g) 245 834.4 245 655.5 250 863.1 250 678.2 255 893 .9 255 702.7 260 926.9 260 728.9 265 962.3 265 757.0 270 1000.3 270 786.7 275 1041.0 275 819.1 280 1084.7 280 853.9 285 1131.6 285 891.2 290 1181.7 290 931.3 285.0 1505.9 295 1234.6 295 974.2 290.0 1571.9 300 1287.7 300 1020.4 295.0 1642.8 305 1344.1 305 1069.6 300.0 1718.7 310 1404.6 310 1121.8 305.0 1800.1 315 1469.4 315 1178.6 355.0 1449.4 310.0 1887.2 320 1538.8 320 1239.5 360.0 1510.5 315.0 1980.4 325 1613.3 325 1304.8 365.0 1575.9 320.0 2080.2 330 1693.0 330 1374.9 370.0 1645.5 325.0 2187.0 335 1778.2 335 1449.9 375.0 1720.2 330.0 2301.1 340 1869.6 340 1530.2 380.0 1800.2 335.0 2422.9 345 1967.2 345 1616.3 385.0 1885.7 340.0 2552.9 350 2071.6 350 1708.0 390.0 1977.1 345.0 2691.5 355 2183 .3 355 1805.3 395.0 2074.9 350.0 2839.1 360 2302.5 360 1911.8 400.0 2179.2 355.0 2996.2 365 2429.6 365 2022.5 405.0 2290.7 360.0 3163.0 370 2565 .3 370 2142.6 410.0 2409.6 365.0 3339.9 375 2709.9 375 2270.8 415.0 2536.4 370.0 3527.2 380 2863.7 380 2405.1 420.0 2671.5 375.0 3725.2 385 3026.9 385 2537.5 425.0 2815 .2 380.0 3933 .9 390 3200.4 390 2672.5 430.0 2967.9 385.0 4153.4 395 3383.8 395 2816.1 435.0 3130.1 390.0 4383 .6 400 3578.2 400 2968.8 440.0 3302.0 395.0 4624.2 Ref. Calc. N-291 PTLR-1 u3r16.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 12 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 0 50 100 150 200 250 300 350 400 450 RCS TEMPERATURE (F}

FIGURE 2.1-2: Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates ofO, 25, 50, 75 and 100°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

PTLR-lu3rl6.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 13 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2)

With Margins for Instrumentation Errors Steady State 25°F/hr 50°F/hr 75°F/hr 100°F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(oF) (psig) (oF) (psi g) (oF) (psi g) (oF) (psi g) (oF) (psi g) 390 3398.7 390 3398.7 390.0 3398.7 390.0 3398.7 390.0 3398.7 385 3208 .9 385 3208.9 385.0 3208.9 385.0 3208.9 385.0 3208.9 380 3029.6 380 3029.6 380.0 3029.6 380.0 3029.6 380.0 3029.6 375 2860.8 375 2860.8 375.0 2860.8 375.0 2860.8 375.0 2860.8 370 2701.9 370 2701.9 370.0 2701.9 370.0 2701.9 370.0 2701.9 365 2552.5 365 2552.5 365.0 2552.5 365.0 2552.5 365.0 2552.5 360 2412.4 360 2412.4 360.0 2412.4 360.0 2412.4 360.0 2412.4 355 2281.0 355 2281.0 355.0 2281.0 355.0 2281.0 355.0 2281.0 350 2157.9 350 2157.9 350.0 2157.9 350.0 2157.9 350.0 2157.9 345 2042.7 345 2042.7 345.0 2042.7 345.0 2042.7 345.0 2042.7 340 1935.0 340 1935.0 340.0 1935.0 340.0 1935.0 340.0 1935.0 335 1834.3 335 1834.3 335.0 1834.3 335.0 1834.3 335.0 1834.3 330 1740.2 330 1740.2 330.0 1740.2 330.0 1740.2 330.0 1740.2 325 1652.4 325 1652.4 325.0 1652.4 325.0 1652.4 325.0 1652.4 320 1570.3 320 1570.3 320.0 1570.3 320.0 1570.3 320.0 1570.3 315 1493.8 315 1493.8 315.0 1493.8 315.0 1493.8 315.0 1493.8 310 1422.4 310 1422.4 310.0 1422.4 310.0 1422.4 310.0 1422.4 305 1355.8 305 1355.8 305.0 1355.8 305.0 1355.8 305.0 1355.8 300 1293.7 300 1293.7 300.0 1293.7 300.0 1293.7 300.0 1293.7 295 1235.9 295 1235.9 295.0 1235.9 295.0 1235.9 295.0 1235.9 290 1181.9 290 1180.7 290.0 1181.9 290.0 1181.9 290.0 1181.9 285 1131.6 285 1127.9 285.0 1129.8 285.0 1131.6 285.0 1131.6 280 1084.7 280 1076.8 280.0 1074.8 280.0 1078.0 280.0 1084.7 275 1041.0 275 1028.9 275.0 1022.0 275.0 1021.2 275.0 1027.0 270 1000.3 270 984.0 270.0 972.5 270.0 966.7 270.0 967.6 265 962.3 265 942.3 265.0 926.5 265.0 915.7 265.0 911.1 260 926.9 260 903.5 260.0 883 .7 260.0 868.3 260.0 858.5 255 893.9 255 867.3 255.0 843.8 255.0 824.2 255.0 809.6 250 863.1 250 833.6 250.0 806.7 250.0 783 .3 250.0 764.2 245 834.4 245 802.2 245.0 772.2 245.0 745.2 245.0 722.0 240 807.6 240 772.9 240.0 740.1 240.0 709.8 240.0 682.8 235 782.7 235 745.7 235.0 710.3 235.0 676.9 235.0 646.4 230 759.4 230 720.3 230.0 682.5 230.0 646.3 230.0 612.6 PTLR-1 u3rl6.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 14 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2)

With Margins for Instrumentation Errors Steady State 25°F/hr 50°F/hr 75°F/hr 100°F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

CF) (psig) (oF) (psi g) (oF) (psi g) (oF) (psi g) (oF) (psi g) 225 737.7 225 696.7 225.0 656.7 225.0 617.9 225.0 581.2 220 717.4 220 674.6 220.0 632.6 220.0 591.5 220.0 552.1 215 698.4 215 654.1 215.0 610.2 215 .0 567.0 215 .0 525.1 210 680.8 210 635.0 210.0 589.4 210.0 544.2 210.0 500.0 205 664.3 205 617.1 205.0 570.0 205.0 523.0 205.0 476.7 200 648.9 200 600.5 200.0 551.9 200.0 503.3 200.0 455.0 195 634.6 195 585.0 195.0 535.1 195.0 485.0 195.0 435.0 190 621.1 190 570.5 190.0 519.5 190.0 468.0 190.0 416.4 185 608.6 185 557.1 185.0 504.9 185.0 452.2 185.0 399.1 180 596.9 180 544.5 180.0 49 1.4 180.0 437.5 180.0 383 .0 175 586.0 175 532.8 175.0 478.8 175.0 423 .8 175.0 368.2 170 575.8 170 521.9 170.0 467.0 170.0 411.2 170.0 354.4 165 566.3 165 511.8 165.0 456.2 165.0 399.4 165.0 341.7 160 557.4 160 502.3 160.0 446.0 160.0 388.5 160.0 329.8 155 549.1 155 493.5 155.0 436.6 155.0 378.4 155.0 318 .9 150 541.4 150 485.3 150.0 427.9 150.0 369.0 150.0 308.8 145 534.2 145 477.7 145.0 419.8 145.0 360.4 145.0 299.4 140 527.5 140 470.6 140.0 412.2 140.0 352.3 140.0 290.8 135 521.2 135 464.0 135.0 405.3 135.0 344.9 135.0 282.8 130 515.3 130 457.9 130.0 398.8 130.0 338.0 130.0 275.4 125 509.9 125 452.2 125.0 392.8 125.0 331.7 125.0 268.7 120 504.8 120 446.9 120.0 387.3 120.0 325.8 120.0 262.4 115 500.0 115 442.0 115.0 382.1 115.0 320.4 115.0 256.7 110 495.6 110 437.4 110.0 377.4 110.0 315.5 110.0 251.5 105 491.5 105 433.2 105.0 373.0 105.0 310.9 105.0 246.6 100 487.6 100 429.2 100.0 369.0 100.0 306.7 100.0 242.2 95 484.1

• 95 425.6 95.0 365.2 95.0 302.8 95.0 238.3 90 480.7 90 422.2 90.0 361.8 90.0 299.3 90.0 234.6 85 477.6 85 419.1 85 .0 358.7 85.0 296.2 85 .0 231.3 80 474.7 80 416.2 80.0 355.7 80.0 293 .2 80.0 228.3 75 472.0 75 413 .5 75 .0 353 .0 75 .0 290.7 75 .0 225.5 70 469.3 70 410.8 70.0 350.4 70.0 287.8 70.0 222.9 Calc. N-291 PTLR-1 u3r16 .DOC 048 0703.1 109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUillLOCANYONPO~RPLANT REVISION 16 PAGE 15 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 2.2-1 Low Temperature Over-Pressure (LTOP)

System Setpoints Function Setpoint PORV Arming Temperature(I) 2>: 273 °F PORV Pressure Setpoint(2) 435 psig (1)

Calc. N-298, Rev 4. Valid to 35 EFPY (2)

STA-249, Rev 4 Table 2.2-2 Low Temperature Over-Pressure (L TOP)

Temperature Restrictions Restriction Setpoint RSGsCt)

SI Pumps Secured, CCP 1 or CCP 2 Secured, SI Accumulators Isolated, CCP 3 aligned S 273 °F for LTOP operation Safety Injection Flowpath Blocked, and SI Blocked S 164 op 2 of3 Charging Pumps Secured S 151 °F 1 of 4 RCPs Secured S 143 op 2 of 4 RCPs Secured S 127 °F 3 of 4 RCPs Secured S 113 °F 4 of 4 RCPs Secured S 104 °F RCS Vent Path of 2.07 in2 Established S 86 °F (1)

Calc. STA-249, Rev 4 Assumptions: 1) PORV Stroke Time of2.9 seconds.

2) Apply 10% per Code Case N-514.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 16 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2

3. ADDITIONAL CONSIDERATIONS Revisions to the PTLR or its supporting analyses should include the following considerations to ensure that the assumptions are still valid:

3.1 The PORV piping qualification under LTOP conditions is bounded by testing performed in accordance with NUREG 0737.

3.2 At the LTOP setpoints, there is no credible way to challenge RCP number 1 seal operation.

3.3 LTOP heat injection case is bounded by the mass injections case throughout the current range of operation.

4. REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material surveillance program is in compliance with Appendix H to 10 CFR 50, entitled "Reactor Vessel Material Surveillance Program Requirements" and Section 5.2.4.4 of the Final Safety Analysis Report (FSAR). The withdrawal schedule is presented in FSAR Table 5.2-22.

Diablo Canyon Units 1 & 2 each have their own independent material surveillance program allowing each to have its own unit specific heat up and cooldown curves and LTOP setpoints. Both units are currently operated using the same limitations resulting from the most conservative limitations in either unit.

The programs are described in the following:

4.1 WCAP-8465, PG&E Diablo Canyon Unit 1 Reactor Vessel Surveillance Program, January, 1975.

4.2 WCAP-13440, Supplemental Reactor Vessel Radiation Surveillance Program for PG&E Diablo Canyon Unit 1, December, 1992.

4.3 WCAP-8783, PG&E Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, December, 1976.

The surveillance capsule reports are as follows:

4.4 WCAP-11567, Analysis ofCapsule S from Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, December, 1987.

4.5 WCAP-13750, Analysis of Capsule Y from Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, July, 1993.

4.6 WCAP-15958, Analysis of Capsule V from Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, January 2003.

4.7 WCAP-11851, Analysis of Capsule U from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, May, 1988.

4.8 WCAP-12811, Analysis of Capsule X from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, December, 1990.

4.9 WCAP-14363, Analysis of Capsule Y from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, August, 1995.

4.10 WCAP-15423, Analysis of Capsule V from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, September 2000.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 17 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Diablo Canyon Units 1 and 2 also have Reactor Cavity Neutron Measurement Programs described in:

4.11 WCAP-14284, Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 1 -cycles 1 through 6, January, 1995.

4.12 WCAP-15780, Fast Neutron Fluence and Neutron Dosimetry Evaluations for the Diablo Canyon Unit 1 Reactor Pressure Vessel, December, 2001.

4.13 WCAP-14350, Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 2 -cycles 1 through 6, November, 1995.

4.14 WCAP-15782, Fast Neutron Fluence and Neutron Dosimetry Evaluations for the Diablo Canyon Unit 2 Reactor Pressure Vessel, December, 2001.

4.15 WCAP-17472-NP Rev 1, Ex-Vessel Neutron Dosimetry Program for Diablo Canyon Unit 1 Cycle 16, October 2011.

4.16 WCAP-17528-NP Rev 0, Ex-Vessel Neutron Dosimetry Program for Diablo Canyon Unit 2 Cycle 16, February 2012.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 18 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2

5. REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there have been three surveillance capsules removed and analyzed from the Diablo Canyon Unit 1 reactor vessel and four from the Diablo Canyon Unit 2 reactor vessel. They must be shown to be credible in order to use these surveillance data sets. There are five requirements that must be met for the surveillance data to be judged credible in accordance with Regulatory Guide 1.99, Revision 2.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Diablo Canyon reactor vessel surveillance data.

Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CPR Part 50, "Fracture Toughness Requirements," as follows:

"The reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

The Diablo Canyon pressure and temperature limits are derived using the most limiting locations of both units to create a single set of limiting parameters. The most limiting %t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel and the most limiting %t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel.

I The Unit 2 Base Metal Surveillance Capsules are made using material from Intermediate Shell Plate

~

B5454-1. This material is the same type of material as the controlling material (B5454-2) and has nearly identical properties (Cu content is identical and Ni content is 0.06% higher than the controlling material). The Diablo Canyon Surveillance Program meets the intent of this criterion.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the inadiated and uninadiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

The Charpy energy versus temperature curves (irradiated and unirradiated) for the surveillance materials show reasonable scatter and allow determination ofthe RTNoT at 30 ft-lb and upper shelf energy.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 19 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Criterion 3: Where there are two or more sets of surveillance data from one reactor, the scatter of

!lRTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

Tables 5.0-1, 5.0-2, 5.0-3, and 5.0-4 present the Surveillance Capsule Data for Diablo Canyon

. Units 1 and 2, and sister plants. The scatter of !lRTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 should be less than 1 a (standard deviation) of 17°F for base metal and 28°F for weld material.

The Diablo Canyon Unit 1 Surveillance CapsuleS data sets for the Intermediate Shell Plate B4106-3 and Surveillance Weld Heat 27204 both show scatter in excess of the Criterion 3 allowable values.

However, when combined with the surveillance data for Palisades Weld Heat 27204 (WCAP 17315 NP Rev 0, Table A.l-8), the combined data is deemed credible per Criterion 3.

Per WCAP 17315 NP Rev 0, Table A.2-2, data for U2 Intermediate Shell Longitudinal Weld Metal Heat 21935/12008 indicates that three of the four surveillance data points fall within the 28°F scatter band for surveillance weld materials; therefore, the weld material (Heat 21935/12008) is deemed credible per Criterion 3.

Per WCAP 17315 NP Rev 0, Section A.2, data for U2 Intermediate Shell Longitudinal Weld Metal Heat 33A277 is not contained in the Diablo Canyon Unit 2 surveillance program. However, it is contained in the Farley Unit 1 and Calvert Cliffs Unit 1 surveillance programs and most closely represents the situation for Diablo Canyon Unit 2 weld Heat 33A277. WCAP 17315 NP Rev 0, Table A.2-1 0, indicates that all eight surveillance data points using Farley Unit 1 and Calvert Cliffs Unit 1 Data fall within the 28°F scatter band for surveillance weld materials; therefore, the weld material (Heat 33A277) is deemed credible per Criterion 3.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within+/- 25°F.

The capsule specimens are located in the reactor between the thermal shield (Unit 1) or neutron pads (Unit 2) and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield (Unit 1) or neutron pads (Unit 2). The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence this criteria is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

The surveillance data for the correlation monitor material in the capsules fall within the scatter band for this (Correlation Monitor Material Heavy Section Steel Technology Plate 02) material.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPO~RPLANT REVISION 16 PAGE 20 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data CF(a) Best Fit Measured Scatter in Material Capsule FF (b) (c)

~RTNDT ~RTNDT ~RTNDT Inter Shell Plate B4106-3 s 0.655 21.07 -0.38 21.5 Inter Shell Plate B4106-3 y 32.2 1.014 32.59 48.26 15.7 Inter Shell Plate B41 06-3 v 1.085 34.9 33.22 1.7 Surveillance Weld DCPP Heat 27204 s 0.655 130.79 112.19 18.6 Surveillance Weld DCPP y 199.6 1.014 202.31 232.19 29.9 Heat 27204 Surveillance Weld DCPP Heat 27204 v 1.085 216.64 199.97 16.7 Source: WCAP-17315-NP Table A.l-4 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see WCAP-17315-NP Table A.l-3).

(b)

Best fit ~RTNDT = CF

• FF.

(c)

Measured ~RTNDT values are derived from the measured 30 ft-lb shift values from Table 4.1-1 (see WCAP-17315-NP). These measured ~RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Reg. Guide 1.99, Rev. 2, Position 2.1 since this calculation is based on the actual surveillance weld metal shift values. Therefore, as shown in Table A.l-1 (see WCAP-17315-NP), the Diablo Canyon Unit 1 surveillance capsules are adjusted by the temperature adjustment values summarized in Table A.l-2 (see WCAP-17315-NP).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPO~RPLANT REVISION 16 PAGE 21 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-2(d)

Diablo Canyon Unit 1 & Palisades Unit 1 Surveillance Capsule Data CF(a) Best Fit Measured Scatter in Material Capsule FF (b) (c)

L\RTNDT L\RTNDT ARTNDT Surveillance Weld DCPP Heat 27204 s 0.655 137.88 116.24 21.6 Surveillance Weld DCPP y 210.4 1.014 213.27 237.44 24.2 Heat 27204 Surveillance Weld DCPP Heat 27204 v 1.085 228.38 204.9 23.5 Surveillance Weld Palisades SA-60-21 1.112 234.02 245.92 11.9 Heat 27204 210.4 Surveillance Weld Palisades SA-240-1 1.234 259.6 261.16 1.6 Heat 27204 Source: WCAP-17315-NP Table A.l-8 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see WCAP-17315-NP Table A.1-7).

(b)

Best fit ARTNDT = CF

• FF.

(c)

Measured ARTNDT values are derived from the measured 30 ft-lb shift values from Tables 4.1-1 and 4.1-2 (see WCAP-17315-NP) for Diablo Canyon Unit 1 and Palisades, respectfully. These ARTNDT values for the surveillance weld data are adjusted first by the difference in operating temperature then using the ratio procedure to account for difference in the surveillance weld chemistry and the beltline weld chemistry. The temperature adjustments are shown in Table A.l-6 (see WCAP-17315-NP). The ratios applied are 1.01 for Diablo Canyon Unit 1 and 0.99 for Palisades.

(d)

As established in WCAP-17315-NP Appendix A.1, specifically NRC Case 1 and Case 4 guidelines, the combined surveillance data from Diablo Canyon Unit 1 and Palisades may be applied to the Diablo Canyon Unit 1 reactor vessel weld Heat #27204.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUDLOCANYONPO~RPLANT REVISION 16 PAGE 22 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-3(d)

Diablo Canyon Unit 2 Surveillance Capsule Data CFCa) Best Fit Measured Scatter in Material Capsule FF (b) (c)

L1RTNDT L1RTNDT L1RTNDT Inter Shell Plate B5454-1 (Long) u 0.695 68.80 65.4 3.4 Inter Shell Plate B5454-1 (Long) X 0.972 96.25 100.1 3.9 99.0 Inter Shell Plate B5454-1 (Long) y 1.118 110.63 111.6 1.0 Inter Shell Plate B5454-1 (Long) v 1.234 122.14 123.4 1.3 Inter Shell Plate B5454-1 (Trans) u 0.695 68.80 73.3 4.5 Inter Shell Plate B5454-1 (Trans) X 0.972 96.25 99.5 3.3 99.0 Inter Shell Plate B5454-1 (Trans) y 1.118 110.63 111.6 1.0 Inter Shell Plate B5454-1 (Trans) v 1.234 122.14 112.9 9.2 Surveillance Weld u 0.695 137.53 173 35.5 Surveillance Weld X 0.972 192.40 203.2 10.8 197.9 Surveillance Weld y 1.118 221.16 211.4 9.8 Surveillance Weld v 1.234 244.15 224.5 19.7 Source: WCAP-17315-NP Table A.2-2 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see WCAP-17315-NP Table A.2-1).

(b)

Best fit L1RTNDT = CF

• FF.

(c)

Measured L1RTNDT values are derived from the measure 30 ft-lb shift values from Table 4.2-1 (see WCAP-17315-NP). These measured ~RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Reg. Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, all of the Diablo Canyon Unit 2 surveillance capsules were irradiated at the same temperature; therefore, no temperature adjustments are required.

(d)

As established in WCAP-17315-NP Appendix A.2, Diablo Canyon Unit 2 surveillance and weld metal (Heat#21935/12008) were evaluated using Diablo Canyon Unit 2 Data and following NRC Case 1 guidelines.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPOWERPLANT REVISION 16 PAGE 23 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-4(d)

Farley Unit 1 and Calvert Cliffs Unit 1 Surveillance Capsule Data CF(a) Best Fit Measured Scatter in Material Capsule FF (b) (c)

~RTNDT ~TNDT ~RTNDT Surveillance Weld Heat y 0.862 68.27 79.2 10.9

1. 33A277 (Farley)

Surveillance Weld Heat

1. 33A277 (Farley) u 1.151 91.09 84.4 6.7 Surveillance Weld Heat X 1.295 102.54 99.5 3.1

1. 33A277 (Farley) 79.2 Surveillance Weld Heat

1. 33A277 (Farley) w 1.392 110.20 113.3 3.1 Surveillance Weld Heat

1. 33A277 (Farley) v 1.466 116.05 135.7 19.6 Surveillance Weld Heat

1. 33A277 (Farley) z 1.492 118.09 130.7 12.6 Surveillance Weld Heat 263° 0.866 68.56 48.5 20.1

1. 33A277 (Calvert Cliffs) 79.2 Surveillance Weld Heat 79.2° 1.26 99.72 74.3 25.4

1. 33A277 (Calvert Cliffs)

Source: WCAP-17315-NP Table A.2-10 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see WCAP-17315-NP Table A.2-9) .

(b)

Best fit ~RTNDT = CF

• FF.

(c)

Measured ~RT NDT values are derived from the measured 30 ft-lb shift values from Table 4.2-2 (see WCAP-17315-NP). These ~RTNDT values for the surveillance weld data are adjusted by the difference in operating temperature then using the ratio procedure to account for difference in the surveillance weld chemistry and the beltlines weld chemistry. The temperature adjustments are shown in Table A.2-8 (see WCAP-17315-NP). The ratios applied are 1.17 for Farley Unit 1 and Calvert Cliffs Unit 1, respectfully.

(d)

As established in WCAP-17315-NP Appendix A.2, Farley Unit 1 and Calvert Cliffs Unit 1 were evaluated using their respective surveillance data and following NRC Case 5 guidelines.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 24 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2

6. SUPPLEMENTAL DATA TABLES Table 6.0-1 Comparison of Diablo Canyon Unit 1 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 6.0-2 Comparison of Diablo Canyon Unit 2 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 6.0-3A/ Calculation of Chemistry Factors Using Surveillance Capsule Data B/C Table 6.0-4 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Table 6.0-5 DCPP-2 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Table 6.0-6 DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the Vessel Surface, Clad to Base Metal Interface, ~t arid %t Locations at 35 EFPY Table 6.0-7 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the Vessel Surface, Clad to Base Metal Interface, ~t and %t Locations at 35 EFPY Table 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and %t Locations for 35 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and %t Locations for 35 EFPY Table 6.0-10 Calculation of Adjusted Reference Temperature at 35 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials PTLR-1 u3r16.DOC 04B 0703.1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 25 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2

7. PRESSURJZED THERMAL SHOCK (PTS) SCREENING 10 CFR 50.61 requires that RT PTS be detennined for each of the vessel beltline materials. The RT PTS is required to meet the PTS screening criterion of 270°F for plates, forgings, and axial weld material, and 300°F for circumferential weld material. If the screening criterion is not met, specific actions taken to either meet the screening criterion or prevent potential reactor vessel failure as a result ofPTS require review and approval of the NRC. The maximum projected RT PTS for Units 1 and 2 is 243°F (Unit 1 Weld 3-442C), at 54 EFPY (EOL). Therefore at 35 EFPY the PTS screening criteria is met. The PTS evaluations are described in the following report:

7.1 WCAP-17315-NP, Rev. 0, "Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations", July 2011.

8. REFERENCES 8.1 Technical Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)"

8.2 License Amendment No. 135 (U1)/135 (U2), dated May 28, 1999 8.3 License Amendment No. 133 (U1)/131 (U2), dated May 3, 1999 8.4 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cool down Limit Curves, Revision 2,"

January 1996 8.5 PG&E letter DCL-00-070, Supplement to Reactor Coolant System Pressure and Temperature Limits Report 8.6 "RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", EPRI NP-1850-CCM-A, Project 889-3, December, 1996 8.7 PG&E Calculation STA-197, superseded by STA-249 8.8 PG&E Calculation N-291, Rev 5, "Pressure-Temperature Limits for Heatup &

Cooldown" 8.9 PG&E Calculation N-298, Rev 4, "LTOP Enable Temperature for 35 EFPY" 8.10 PG&E Calculation STA-249 Rev 4, "RSG- LTOP Analysis" 8.11 Operating Procedure OP A-2:IX, "Reactor Vessel- Vacuum Refill of the RCS" 8.12 Westinghouse Letter PGE 12, "Applicability of the Pressure-Temperature Limit Curves During Vacuum Refill of the RCS in Mode 5", February 21, 2014 8.13 PG&E Calculation N-288, Rev 4, "Reactor Vessel Adjusted RT-NDT Versus EFPY" PTLR-1u3r16.DOC 04B 0703 .1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 26 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-1 Comparison of Diablo Canyon Unit 1 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Materials Capsule Fluence (d) 30 ft-lb Transition Upper Shelf Energy (X 10 19 nlcm2) Temperature Shift Decrease Predicted Measured Predicted Measured (oF) (a) (oF) (b) (o/o) (a) (o/o) (c)

Plate B4106-3 s 0.284 36.2 -1.78 14 0 y 1.05 56.0 48.66 19 6.8 v 1.37 60.0 34.32 20 0 Surveillance Weld s 0.284 145.8 110.79 25.5 11 Metal y 1.05 225.4 232.59 34.5 34.1 v 1.37 241.6 201.07 36.5 27.5 Heat Affected s 0.284 -- 72.31 -- 8.1 Zone Metal y 1.05 -- 79.77 -- 19.9 v 1.37 -- 110.90 -- 14.7 Correlation Monitor s 0.284 73.01 65.62 -- 2.4 Plate HSST 02 y 1.05 112.9 115.79 -- 8.9 v 1.37 121.0 116.61 -- 4.9 WCAP-15958 (a)

Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b)

Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1.

(c)

Values are based on the definition of upper shelf energy given in ASTM E185-82.

(d)

The WCAP-15958 calculated fluence values given here are slightly higher than the more recent WCAP-17315-NP Rev 0 values.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 27 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-2 Comparison of Diablo Canyon Unit 2 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Fluence (c) 30 ft-lb Transition Upper Shelf Energy Materials Capsule (X 10 19 n/cm2) Temperature Shift Decrease Predicted Measured Predicted Measured COF) (a) COF) (b) (%) (a) (o/o) (b)

Plate B5454-1 u 0.338 71.0 65.4 18 11 (Longitudinal) X 0.919 98.9 100.1 22 20 y 1.55 113.6 111.6 25 18 v 2.41 125.3 123.4 28 24 Plate B5454-1 u 0.338 71.0 73.3 18 0 (Transverse) X 0.919 98.9 99.5 22 12 y 1.55 113.6 111.6 25 7 v 2.41 125.3 112.9 28 6 Surveillance u 0.338 148.1 173.0 28 31 Weld Metal X 0.919 206.1 203.2 35 38 y 1.55 236.8 211.4 40 40 v 2.41 261.3 224.5 44 40 Heat Affected u 0.338 -- 234.4 -- 41 Zone Metal X 0.919 -- 253.5 -- 31 y 1.55 -- 257.7 -- 40 v 2.41 -- 291.5 -- 52 WCAP-15423 (a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1.

(c) The WCAP-15958 calculated fluence values given here are slightly higher than the more recent WCAP-17315-NP Rev 0 values.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 28 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-3A Calculation of Diablo Canyon Unit 1 Chemistry Factors Using Surveillance Capsule Data (c)

Capsule :fa) ARTNDT FF*ARTNDT Material Capsule (x10 19 n/cm2 , E > 1.0 MeV) FF(b) (oF) (oF) FF 2 s 0.283 0.655 6.00 (O(d)) 3.93 0.429 IS Plate B41 06-3 y 1.05 1.014 52.86 (48.66) 53.58 1.027 (Longitudinal) v 1.36 1.085 37.82 (34.32) 41.05 1.178 SUM: 98.56 2.635 CFisPlateB4I06-3 = L:(FF

• L1RTNDT) + L:(FF2) = (98.56) + (2.635) = 37.4°F Weld Metal s 0.283 0.655 119.13 (110.79) 78.06 0.429 Heat#27204 y 1.05 1.014 241.53 (232.59) 244.82 1.027 (Diablo Canyon Unit 1 data) v 1.36 1.085 208.66 (201.07) 226.49 1.178 Weld Metal SA-60-1 1.50 1.112 250.10 (253.1) 278.18 1.237 Heat#27204 (Palisades data) SA-240-1 2.38 1.234 265.50 (267.8) 327.59 1.522 SUM: 1155.14 5.395 CFHeat#27204= L:(FF

• L1RTNDT) + L:(FF2) = (1155.14) + (5.395) = 214.1"F (a) f= fluence.

(b) FF = fluence factor= f 0*28 -0.lO*logf).

(c) L1RTNDT values are the measured 30 ft-lb shift values. All Diablo Canyon Unit 1 values are taken from Table 4.1-1 ofWCAP-17315-NP. The Diablo Canyon Unit 1 L1RTNDT values have been adjusted according to the temperature adjustments summarized in Table 4.1-1 ofWCAP-17315-NP. Then, the Diablo Canyon Unit 1 L1RTNDT values for the surveillance weld data are adjusted by a ratio of 1.02 (pre-adjusted values are listed in parentheses). Ratio= CFvessel we1iCFsurv. Weld= 226.8°F/222.3°F =

1.02.

All Palisades values are taken from Table 4.1-2 ofWCAP-17315-NP. The Palisades surveillance weld L1RTNDT values have been adjusted according to the temperature adjustments summarized in Table 4.1-2 (pre-adjusted values are listed in parentheses). No ratio is applied since the ratio was calculated to be 1.00. Ratio= CFvessel we1iCFsurv. Weld= 226.8°F/227.8°F = 1.00.

(d)

Measured ~RTNDT value was determined to be negative, but physically a reduction should not occur.

Therefore, a conservative value of zero will be used.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 29 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-3B Calculation of Diablo Canyon Unit 2 Chemistry Factors Using Surveillance Capsule Data Capsule fa) ARTNDT (c) FF*ARTNDT Material 19 2 Capsule (x10 n/cm , E > 1.0 MeV) FF(b) (oF) (oF) FF 2 u 0.330 0.695 72.4 (65.4) 50.32 0.483 IS Plate B5454-1 X 0.906 0.972 107.1 (100.1) 104.14 0.945 (Longitudinal) y 1.53 1.118 118 .6 (111. 6) 132.55 1.249 v 2.38 1.234 130.4 (123.4) 160.89 1.522 u 0.330 0.695 80.3 (73.3) 55.81 0.483 IS Plate B5454-1 X 0.906 0.972 106.5 (99.5) 103.55 0.945 (Transverse) y 1.53 1.118 118 .6 (111. 6) 132.55 1.249 v 2.38 1.234 119.9 (112.9) 147.94 1.522 SUM: 887.76 8.400 CFrsPiateB5454-I = I:(FF

• L1RTNDT)--;- I:(FF2) = (887.76) --;- (8.400) = 105.7aF u 0.330 0.695 180.0 (173.0) 125.10 0.483 Diablo Canyon Unit 2 X 0.906 0.972 210.2 (203.2) 204.38 0.945 Weld Metal y 1.53 1.118 218.4 (211.4) 244.09 1.249 (Heat# 21935/12008) v 2.38 1.234 231.5 (224.5) 285.64 1.522 SUM: 859.22 4.200 CFHeat#21935/12008= I:(FF

• LlRTNDT)--;- I:(FF2) = (859.22)--;- (4.200) = 204.6oF (a) f= fluence.

(b)

FF = fluence factor= f 0*2S-O.lO*Iogf).

(c) i1RTNDT values are the measured 30 ft-lb shift values. All values are taken from Table 4.2-1 ofWCAP-17315-NP. The Diablo Canyon Unit 2 L1RTNDT value~ have been adjusted according to the temperature adjustments summarized in Table 4.2-1 ofWCAP-17315-NP (pre-adjusted values are listed in parentheses). No ratio is applied to the i1RTNDT values for the surveillance weld data since the beltline weld and surveillance weld chemistry factors are identical.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 30 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-3C Calculation of Diablo Canyon Unit 2 Weld Heat# 33A277 Chemistry Factors Using Surveillance Capsule Data from Farley Unit 1 and Calvert Cliffs Unit 1 Capsule t'a) ARTNDT(c) FF*ARTNDT Material Capsule (x10 19 n/cm2 , E > 1.0 MeV) FFCb) (oF) (oF) FF 2

y 0.612 0.862 118.1 (66.9) 101.86 0.744 u 1.73 1.151 125.3 (75.1) 144.20 1.324 Weld Metal Heat# 33A277 X 3.06 1.295 146.2 (87.4) 189.42 1.678 (Farley Unit 1 data) w 4.75 1.392 165.3 (98.3) 230.15 1.938 v 7.14 1.466 196.4 (117.5) 287.90 2.149 z 8.47 1.492 189.4 (113.5) 282.59 2.225 Weld Metal Heat# 33A277 263° 0.62 0.866 73.1 (59) 63.35 0.750 (Calvert Cliffs Unit 1 data) 97° 2.64 1.260 109.2 (93) 137.54 1.587 SUM: 1436.99 12.396 CFHeat#33A277= :L(FF

• LlliTNDT) -;-- :L(FF2) = (1436.99)-;-- (12.396) = 115.9oF (a) f= fluence.

(b)

FF = fluence factor = f 0*28 - o.lO*log f).

(c)

L1RTNDT values are the measured 30 ft-lb shift values. All values are taken from Table 4.2-2 ofWCAP-17315-NP. L1RTNDT values for the surveillance weld data are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the beltline weld chemistry (pre-adjusted values are listed in parentheses). The temperature .

adjustments and ratios applied are as follows:

Farley Unit 1:

Temperature adjustment per Table 4.2-2 (on a capsule-by-capsule basis) ~atio = CFvessel we1/CFsurv. Weld

= 126.3°F/78.1 °F = 1.62 Calvert Cliffs Unit 1:

Temperature adjustment per Table 4.2-2 (+ 10.00°F for each capsule) Ratio= CFvessel Weld/CFsurv. Weld=

126.3°FI119.4°F = 1.06 PTLR-1 u3r16.DOC 04B 0703 .1109

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 31 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-4 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Material Description Cu ( 0/o) Ni( 0/o) Initial RT NDT eF)

Upper Shell Plate (b)

B4105-1 0.12 0.56 28 B4105-2 0.12 0.57 9 B4105-3 0.14 0.56 14 Inter Shell Plate B4106-1 0.125 0.53 -10 B4106-2 0.12 0.50 -3 B4106-3 0.086 0.476 30 Lower Shell Plate B4107-1 0.13 0.56 15 B4107-2 0.12 0.56 20 B4107-3 0.12 0.52 -22 Upper Shell Long (b)

Welds 1-442 A,B,C 0.19 0.97 -20 Upper Shell to Inter Shell Weld 8-442(b) 0.25 0.73 -56 Inter Shell Long Welds 2-442 A,B,C 0.203(a) 1.018(a) -56 Inter Shell to Lower Shell Weld 9-442 0.183(a) 0.704(a) -56 Lower Shell Long Welds 3-442 A,B,C 0.203(a) 1.018(a) -56 Calc N-NCM-97009 (a) PerCE NPSD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed l.OE+ 17.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUillLOCANYONPO~RPLANT REVISION 16 PAGE 32 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-5 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu ( 0/o) Ni( 0/o) Initial RTNoT COF)

Upper Shell Plate (b)

B5453-1 0.11 0.60 28 B5453-3 0.11 0.60 5 B5011-1R 0.11 0.65 0 Inter Shell Plate B5454-1 0.14 0.65 52 B5454-2 0.14 0.59 67 B5454-3 0.15 0.62 33 Lower Shell Plate B5455-1 0.14 0.56 -15 B5455-2 0.14 0.56 0 B5455-3 0.10 0.62 15 Upper Shell Long(b)

Welds 1-201 A,B,C 0.22 0.87 -50 Upper Shell to Inter Shell Weld 8-201 (b) 0.183(a) 0.704(a) -56 Inter Shell Long Welds 2-201 A,B,C 0.22 0.87 -50 Inter Shell to Lower Shell Weld 9-201 0.046(a) 0.082(a) -56 Lower Shell Long Welds 3-201 A,B,C 0.258(a) 0.165(a) -56 Calc N-NCM-97009 (a) PerCE NSPD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed l.OE+17.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 33 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-6 DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 'i:lt, and %t Locations at 35 EFPY Material (a) Fluence f~t Fluence f~t Inter Shell Plate B4106-1 7.98 E + 18 2.83 E + 18 B4106-2 7.98E+18 2.83 E + 18 B4106-3 7.98 E + 18 2.83 E + 18 Lower Shell Plate B4107-1 7.98 E + 18 2.83 E + 18 B4107-2 7.98E+18 2.83 E + 18 B4107-3 7.98E+18 2.83 E + 18 Inter Shell Long Welds 2-442 A,B 5.89E+18 2.09 E + 18 Weld 2-442 C 3.07 E + 18 1.09E+18.

Inter Shell to Lower Shell Weld 9-442 7.98E+18 2.83 E + 18 Lower Shell Long Welds 3-442 A,B 4.87 E + 18 1.73 E + 18 Weld 3-442 C 7.98E+18 2.83 E + 18 Calc N-288 Rev 4 (a) Only beltline materials are included. WCAP-17315-NP demonstrates that extended beltline materials are not limiting through at least 54 EFPY.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 34 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-7 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 'i:1t and %t Locations at 35 EFPY Material (a) Fluence f~t Fluence ft.;t Inter Shell Plate B5454-1 9.04 E + 18 3.21 E + 18 B5454-2 9.04 E + 18 3.21 E + 18 B5454-3 9.04 E + 18 3.21E+18 Lower Shell Plate B5455-1 9.04 E + 18 3.21 E + 18 B5455-2 9.04 E + 18 3.21 E + 18 B5455-3 9.04 E + 18 3.21 E + 18 Inter Shell Long Weld 2-201 A 5.01 E + 18 1.78 E+ 18 Weld 2-201 B 6.06 E + 18 2.15E+18 Weld 2-201 C 5.16E+18 1.83E+18 Inter Shell to Lower Shell Weld 9-201 9.04 E + 18 3.21E+18 Lower Shell Long Weld 3-201 A 5.16E+18 1.83 E+ 18 Weld 3-201 B 5.01 E + 18 1.78 E+ 18 Weld 3-201 C 6.06E+18 2.15E+18 Calc N-288 Rev 4 (a) Only beltline materials are included. WCAP-17315-NP demonstrates that extended beltline materials are not limiting through at least 54 EFPY.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 16 PAGE 35 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 'l.:lt and %t Locations for 35 EFPY 35 EFPY ARTCa)

Material RG 1.99 Rev 2 l;.:lt (°F) %t COF)

Method Inter Shell Plate B4106-1 Position 1.1 103.9 79.9 B4106-2 Position 1.1 106.9 84.1 B4106-3 Position 1.1 129.8 114.3 Lower Shell Plate \

B4107-1 Position 1.1 133.1 107.9 B4107-2 Position 1.1 131.0 107.9 B4107-3 Position 1.1 88.2 65.4 Inter Shell Long Welds 2-442 A,B Position 2.1 170.4 112.3 Weld 2-442 C Position 2.1 132.8 81.1 Inter Shell to Lower Shell Weld 9-442 Position 1.1 170.8 122.4 Lower Shell Long Welds 3-442 A,B Position 2.1 159.2 102.7 Weld 3-442 C Position 2.1 188.6 128.4 Calc N-288 Rev 4 (a)

ART= Initial RTNnT + ~RTNDT +Margin (°F)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 16 PAGE 36 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and ~t Locations for 35 EFPY 35 EFPY ARTCa)

Material RG 1.99 Rev 2 ~tCOF) ~tCOF)

Method Inter Shell Plate B5454-1 Position 2.1 171.7 141.7 B5454-2 Position 1.1 197.8(b) 169.5(b)

B5454-3 Position 1.1 174.4 143.0 Lower Shell Plate B5455-1 Position 1.1 114.4 86.5 B5455-2 Position 1.1 129.4 101.5 B5455-3 Position 1.1 112.4 93.8 Inter Shell Long Weld 2-201 A Position 2.1 143.1 88.9 Weld 2-201 B Position 2.1 153.9 98.1 Weld 2-201 C Position 2.1 144.8 90.3 Inter Shell to Lower Shell Weld 9-201 Position 1.1 24.4 8.7 Lower Shell Long Weld 3-201 A Position 2.1 82.5 51.7 Weld 3-201 B Position 2.1 81.6 50.8 Weld 3-201 C Position 2.1 87.7 56.1 Calc N-288 Rev 4 (a)

ART = Initial R TNDT + ~RTNDT + Margin (°F)

(b)

These limiting ART values are used to generate heatup and cooldown curves (based on 35 EFPY).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16 PAGE 37 OF 37 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-10 Calculation of Adjusted Reference Temperature at 35 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials Parameter ART Value

'i.:tt(d) %t(e)

Location Chemistry Factor, CF (°F) 99.6(f) 99.6(f)

Fluence + 10 19 n/c~ (E > 1.0 MeV), fa) 0.904 0.321 Fluence Factor, FF(b) 0.9717 0.6878

~RTNDT = CF X FF, (°F) 96.8 68.5 Initial RTNDT, I (°F) 67 67 Margin, M (°Fic) 34 34 ART = I + (CF x FF) + M (°F) 197.8(f) 169.5(f) per Regulatory Guide 1.99, Rev 2 Calc N-288 Rev 4 (a) Fluence, f, is based upon f~t and f%t from Table 6.0-7. The Diablo Canyon reactor vessel wall thickness is 8.625 inches at the beltline region.

(b) Fluence Factor (FF) per Regulatory Guide 1.99, Revision 2, is defined as FF = f 0*2s- o.IO*logf).

(c) Margin is calculated as M = 2(o/+ a 112) 0*5

• The standard deviation for the initial RTNDT margin term ar, is 0°F for plate since the initial RTNDT is a measured value. The standard deviation for ~RTNDT term a 11 ,

is 17°F for the plate, except that a 11 need not exceed the 0.5 times the mean value of ~RTNDT*

(d) DCPP-2 intermediate shell plate B5454-2 is limiting for the heatup and cooldown Appendix G curves at

~t.

(e) DCPP-2 intermediate shell plate B5454-2 is limiting for the heatup and cooldown Appendix G curves at

%t.

(f) The higher CF based onCE NPSD-1039, Rev 2 for these limiting materials is used to generate the heatup and cooldown Appendix G curves. The ARTs used to generate the heatup and cooldown curves are bounding based on 35 EFPY values of 197.8°F for ~t and 169.5°F for %t.

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