DCL-23-117, Reactor Coolant System Pressure and Temperature Limits Report

Reactor Coolant System Pressure and Temperature Limits Report

ML23298A107
Person / Time
Site:	Diablo Canyon
Issue date:	10/25/2023
From:	Darrell Adams Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
DCL-23-117
Download: ML23298A107 (1)

Text

Pacific Gas and Electric Company" PG&E Letter DCL-23-117 U.S. Nuclear Regulatory Commission A TIN: 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 Dallas L Adams Manager, Program Engineering Diablo Canyon Power Plant Mail code 104/41427 P.O. Box 56 Avila Beach, CA 93424 805.545.6182 Dallas.Adams@pge.com 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 current Revision 16a of the PTLR for Units 1 and 2.

Additionally, review of previous PTLR submittals indicated that Revision 11 of the PTLR was not previously submitted. Therefore, also enclosed in this letter is Revision 11 of the PTLR for Units 1 and 2, dated February 29, 2012. The untimely submittal of Revision 11 has been entered into the Diablo Canyon corrective action program.

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 Mr. Rob O'Sullivan, Design Engineering Manager, at (805) 545-6873.

Dallas L. Adams armb/4743/51194829 & 51205795 Enclosure cc:

Diablo Distribution cc/enc:

Mahdi 0. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRC Senior Project Manager John D. Monninger, NRC Region IV Administrator A member of the STARS Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek PG&E Letter DCL-23-117 PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)-1 REVISION 16A EFFECTIVE DATE: April 19, 2022

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 NUCLEAR POWER GENERATION REVISION 16A DIABLO CANYON POWER PLANT PAGE 1 OF 37 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE:

PTLR for Diablo Canyon 1

AND 2 INFO ONLY EFFECTIVE DATE CLASSIFICATION: QUALITY RELATED PTLR-1u3r16.DOC 04B 0416.1038

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-TABLE OF CONTENTS SECTION PAGE REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 2 OPERATING LIMITS...................................................................................................................................... 2 RCS Pressure and Temperature (P/T) 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 60°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) 9 2.1-2 Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) 12 List of Tables Table 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2) With Margins for Instrumentation Errors 10 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2) With Margins for Instrumentation Errors 13 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 5.0-4 Farley Unit 1 and Calvert Cliffs Unit 1 Surveillance Capsule Data 24

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1.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

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

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems The limits provided in this report remain valid until 35 EFPY on Unit 1 and Unit 2.

2.

OPERATING LIMITS 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3)

The RCS temperature rate-of-change limits are:

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

A maximum cooldown of 100F in any 1-hour period.

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

The RCS P/T 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 P/T limits as shown in Figure 2.1-1 and Figure 2.1-2, and the LTOP P/T 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 P/T 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 (KIR) as a function of vessel metal temperature, define the size of the assumed flaw, and apply specified safety factors.

The reference stress intensity (KIR) is the combined thermal and pressure stress intensity limit at a given temperature. The assumed crack has a radial depth of 1/4 of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

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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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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 RTNDT of 35°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 psig pressure is 125°F (RTNDT + 90°F). For Condition 1.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 (RTNDT + 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 RTNDT 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 psig. 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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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 P/T 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 200F or RTNDT 50F 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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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 (SI) 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 P/T 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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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 F 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 P/T 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 RCS/SG temperature restrictions for starting an RCP, since even the maximum credible RCS/SG temperature differential will not challenge the Appendix G P/T 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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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 P/T 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 P/T 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 RCS/SG 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 PORVs 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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FIGURE 2.1-1: Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr)

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50 100 150 200 250 300 3 50 400 450 RCS TEMPERATURE (F)

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TABLE 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2)

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

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig) 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

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TABLE 2.1-1 Diablo Canyon Heatup Data at 35 EFPY (Unit 1 and Unit 2)

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

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig) 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

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FIGURE 2.1-2: Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100°F/hr) Applicable to 35 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) j I

J 2000 -+--1--t--+--+-+--+--+--+-+-1--t---+-+--+-1-J

_i'+ *-,,f-E1GE-P.:i:A --+-+-+-+--+-+-+--+---1--t-+-+----1-+-+-'C:lP-~R-A:r: G l'll---,l--l-+-+----l-+-+--1--t--+-,1---t--1--t-+-+----l-+-+--t----11--i I

I I

I,

1 '.,00 --+--+--+--+--+--+--+--+---+---+---+---+----+---+--+--+--+--+--+---t---t---tl--ll--l---1--f-'--+---+--

--+--+--+--+--+--+--+---+--+---t---t-l--t

~

vi

~

w 0::

~

w 0::..

I II

~

o:: 1 mo -+-+--+--+--+--+--+--+--+--+--+-+--+-+-+-+-+--+--+--+---t---t----11--f-,,~..._

1 1--1--1--1--+--+-+'""-l--+--+--+--+---+--+--+--+-1--1 500 u

,'.'.:~~

~I

==t=t=t:::::t=:t=t=l=:t=t=t:::=t=t===t=t=t:::=l=:t=tj :;;;~~t-,,,z~-,,,,~=t===t=:t=t=l=:t=fO Eeifilirl __,_

f-

-+---1---+--+--+--l--i

'00 doy.m Rate c..,,1---':::::: --::::~

me rnl~ u..o._,.N,_,_-+---+--+-+--+--1--i 1--1---1---1--+(F ~r 0

~

2 7

+--+--+--+--<1'/ *111A-B0 tlclp-+---+--+--+--+--+---+---+---+--+--+--+--+---+--+--+--+---tl--ll--1--t--t--t--+--+---+--+--+--+---+---i

-+--+--+-+--HT-er~- O-f--+--+-+-+-+-+-+-+--+-+--+--+--+---+--+--+-+-+-+-+-+--t--t--t-+--t--t--t---1--1--t so 100 1SO 700 7SO 400 4SO RCS TEMPERATURE (F)

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TABLE 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2)

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

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig) 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

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TABLE 2.1-2 Diablo Canyon Cooldown Data at 35 EFPY (Unit 1 and Unit 2)

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

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig)

Temp.

(F)

Press.

(psig) 225 737.7 225 696.7 225.0 656.7 225.0 617.9 225.0 581.2 220 717.4 220 674.6 220.0 632.6 220.0 591.5 220.0 552.1 215 698.4 215 654.1 215.0 610.2 215.0 567.0 215.0 525.1 210 680.8 210 635.0 210.0 589.4 210.0 544.2 210.0 500.0 205 664.3 205 617.1 205.0 570.0 205.0 523.0 205.0 476.7 200 648.9 200 600.5 200.0 551.9 200.0 503.3 200.0 455.0 195 634.6 195 585.0 195.0 535.1 195.0 485.0 195.0 435.0 190 621.1 190 570.5 190.0 519.5 190.0 468.0 190.0 416.4 185 608.6 185 557.1 185.0 504.9 185.0 452.2 185.0 399.1 180 596.9 180 544.5 180.0 491.4 180.0 437.5 180.0 383.0 175 586.0 175 532.8 175.0 478.8 175.0 423.8 175.0 368.2 170 575.8 170 521.9 170.0 467.0 170.0 411.2 170.0 354.4 165 566.3 165 511.8 165.0 456.2 165.0 399.4 165.0 341.7 160 557.4 160 502.3 160.0 446.0 160.0 388.5 160.0 329.8 155 549.1 155 493.5 155.0 436.6 155.0 378.4 155.0 318.9 150 541.4 150 485.3 150.0 427.9 150.0 369.0 150.0 308.8 145 534.2 145 477.7 145.0 419.8 145.0 360.4 145.0 299.4 140 527.5 140 470.6 140.0 412.2 140.0 352.3 140.0 290.8 135 521.2 135 464.0 135.0 405.3 135.0 344.9 135.0 282.8 130 515.3 130 457.9 130.0 398.8 130.0 338.0 130.0 275.4 125 509.9 125 452.2 125.0 392.8 125.0 331.7 125.0 268.7 120 504.8 120 446.9 120.0 387.3 120.0 325.8 120.0 262.4 115 500.0 115 442.0 115.0 382.1 115.0 320.4 115.0 256.7 110 495.6 110 437.4 110.0 377.4 110.0 315.5 110.0 251.5 105 491.5 105 433.2 105.0 373.0 105.0 310.9 105.0 246.6 100 487.6 100 429.2 100.0 369.0 100.0 306.7 100.0 242.2 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

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Table 2.2-1 Low Temperature Over-Pressure (LTOP)

System Setpoints Function Setpoint PORV Arming Temperature(1) 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 (LTOP)

Temperature Restrictions Restriction Setpoint RSGs(1)

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

Calc. STA-249, Rev 4 Assumptions:

1) PORV Stroke Time of 2.9 seconds.

2) Apply 10 % per Code Case N-514.

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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 of Capsule 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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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.

4.17 WCAP-18566-NP Rev 0, Ex-Vessel Neutron Dosimetry Program for Diablo Canyon Unit 2 Cycle 21, July 2020.

4.18 WCAP-18655-NP Rev 0, Ex-Vessel Neutron Dosimetry Program for Diablo Canyon Unit 1 Cycle 22, August 2021.

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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 CFR 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 1/4t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel and the most limiting 3/4t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel.

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 irradiated and unirradiated 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 of the RTNDT at 30 ft-lb and upper shelf energy.

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Criterion 3:

Where there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28F for welds and 17F 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 RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 should be less than 1 (standard deviation) of 17F for base metal and 28F for weld material.

The Diablo Canyon Unit 1 Surveillance Capsule S 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.1-8), 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.2-10, 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 +/- 25F.

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 25F. 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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Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data Material Capsule CF(a)

FF Best Fit RTNDT(b)

Measured RTNDT(c)

Scatter in RTNDT Inter Shell Plate B4106-3 S

32.2 0.655 21.07

-0.38 21.5 Inter Shell Plate B4106-3 Y

1.014 32.59 48.26 15.7 Inter Shell Plate B4106-3 V

1.085 34.9 33.22 1.7 Surveillance Weld DCPP Heat 27204 S

199.6 0.655 130.79 112.19 18.6 Surveillance Weld DCPP Heat 27204 Y

1.014 202.31 232.19 29.9 Surveillance Weld DCPP Heat 27204 V

1.085 216.64 199.97 16.7 Source: WCAP-17315-NP Table A.1-4 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see WCAP-17315-NP Table A.1-3).

(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.1-1 (see WCAP-17315-NP), the Diablo Canyon Unit 1 surveillance capsules are adjusted by the temperature adjustment values summarized in Table A.1-2 (see WCAP-17315-NP).

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Table 5.0-2(d)

Diablo Canyon Unit 1 & Palisades Unit 1 Surveillance Capsule Data Material Capsule CF(a)

FF Best Fit RTNDT(b)

Measured RTNDT(c)

Scatter in RTNDT Surveillance Weld DCPP Heat 27204 S

210.4 0.655 137.88 116.24 21.6 Surveillance Weld DCPP Heat 27204 Y

1.014 213.27 237.44 24.2 Surveillance Weld DCPP Heat 27204 V

1.085 228.38 204.9 23.5 Surveillance Weld Palisades Heat 27204 SA-60-21 210.4 1.112 234.02 245.92 11.9 Surveillance Weld Palisades Heat 27204 SA-240-1 1.234 259.6 261.16

1.6 Source

WCAP-17315-NP Table A.1-8 (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 RTNDT = CF

• FF.

(c)

Measured RTNDT 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 RTNDT 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.1-6 (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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Table 5.0-3(d)

Diablo Canyon Unit 2 Surveillance Capsule Data Material Capsule CF(a)

FF Best Fit RTNDT(b)

Measured RTNDT(c)

Scatter in RTNDT Inter Shell Plate B5454-1 (Long)

U 99.0 0.695 68.80 65.4 3.4 Inter Shell Plate B5454-1 (Long)

X 0.972 96.25 100.1 3.9 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 99.0 0.695 68.80 73.3 4.5 Inter Shell Plate B5454-1 (Trans)

X 0.972 96.25 99.5 3.3 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

197.9 0.695 137.53 173 35.5 Surveillance Weld X

0.972 192.40 203.2 10.8 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 RTNDT = CF

• FF.

(c)

Measured RTNDT values are derived from the measure 30 ft-lb shift values from Table 4.2-1 (see WCAP-17315-NP). These measuredRTNDT 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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Table 5.0-4(d)

Farley Unit 1 and Calvert Cliffs Unit 1 Surveillance Capsule Data Material Capsule CF(a)

FF Best Fit RTNDT(b)

Measured RTNDT(c)

Scatter in RTNDT Surveillance Weld Heat

1. 33A277 (Farley)

Y 79.2 0.862 68.27 79.2 10.9 Surveillance Weld Heat

1. 33A277 (Farley)

U 1.151 91.09 84.4 6.7 Surveillance Weld Heat

1. 33A277 (Farley)

X 1.295 102.54 99.5 3.1 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

1. 33A277 (Calvert Cliffs) 263° 79.2 0.866 68.56 48.5 20.1 Surveillance Weld Heat

1. 33A277 (Calvert Cliffs) 79.2° 1.26 99.72 74.3 25.4 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 RTNDT 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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6.

SUPPLEMENTAL DATA TABLES Table 6.0-1 Comparison of Diablo Canyon Unit 1 Surveillance Material 30 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 Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 6.0-3A/

B/C Calculation of Chemistry Factors Using Surveillance Capsule Data 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, 1/4t and 3/4t 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, 1/4t and 3/4t Locations at 35 EFPY Table 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 35 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t 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

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7.

PRESSURIZED THERMAL SHOCK (PTS) SCREENING 10 CFR 50.61 requires that RT PTS be determined 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 of PTS 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 Cooldown 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"

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Table 6.0-1 Comparison of Diablo Canyon Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Materials Capsule Fluence (d)

(X 1019 n/cm2) 30 ft-lb Transition Temperature Shift Upper Shelf Energy Decrease Predicted (F) (a)

Measured (F) (b)

Predicted

(%) (a)

Measured

(%) (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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Table 6.0-2 Comparison of Diablo Canyon Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Materials Capsule Fluence (c)

(X 1019 n/cm2) 30 ft-lb Transition Temperature Shift Upper Shelf Energy Decrease Predicted (F) (a)

Measured (F) (b)

Predicted

(%) (a)

Measured

(%) (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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Table6.0-3A Calculation of Diablo Canyon Unit1 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV)

FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF2 IS Plate B4106-3 (Longitudinal)

S 0.283 0.655 6.00 (0(d))

3.93 0.429 Y

1.05 1.014 52.86 (48.66) 53.58 1.027 V

1.36 1.085 37.82 (34.32) 41.05 1.178 SUM:

98.56 2.635 CFIS Plate B4106-3 = (FF

• RTNDT) ÷ (FF2) = (98.56) ÷ (2.635) = 37.4F Weld Metal Heat # 27204 (Diablo Canyon Unit1 data)

S 0.283 0.655 119.13 (110.79) 78.06 0.429 Y

1.05 1.014 241.53 (232.59) 244.82 1.027 V

1.36 1.085 208.66 (201.07) 226.49 1.178 Weld Metal Heat # 27204 (Palisades data)

SA-60-1 1.50 1.112 250.10 (253.1) 278.18 1.237 SA-240-1 2.38 1.234 265.50 (267.8) 327.59 1.522 SUM:

1155.14 5.395 CFHeat # 27204 = (FF

• RTNDT) ÷ (FF2) = (1155.14) ÷ (5.395) = 214.1F (a) f = fluence.

(b) FF = fluence factor = f(0.28 - 0.10*log f).

(c)

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

1.02.

All Palisades values are taken from Table 4.1-2 of WCAP-17315-NP. The Palisades surveillance weld RTNDT 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 Weld/CFSurv. 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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Table6.0-3B Calculation of Diablo Canyon Unit2 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV)

FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF2 IS Plate B5454-1 (Longitudinal)

U 0.330 0.695 72.4 (65.4) 50.32 0.483 X

0.906 0.972 107.1 (100.1) 104.14 0.945 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 IS Plate B5454-1 (Transverse)

U 0.330 0.695 80.3 (73.3) 55.81 0.483 X

0.906 0.972 106.5 (99.5) 103.55 0.945 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 CFIS Plate B5454-1 = (FF

• RTNDT) ÷ (FF2) = (887.76) ÷ (8.400) = 105.7F Diablo Canyon Unit2 Weld Metal (Heat # 21935/12008)

U 0.330 0.695 180.0 (173.0) 125.10 0.483 X

0.906 0.972 210.2 (203.2) 204.38 0.945 Y

1.53 1.118 218.4 (211.4) 244.09 1.249 V

2.38 1.234 231.5 (224.5) 285.64 1.522 SUM:

859.22 4.200 CFHeat # 21935/12008 = (FF

• RTNDT) ÷ (FF2) = (859.22) ÷ (4.200) = 204.6F (a) f = fluence.

(b)

FF = fluence factor = f(0.28 - 0.10*log f).

(c)

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

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Table6.0-3C Calculation of Diablo Canyon Unit2 Weld Heat # 33A277 Chemistry Factors Using Surveillance Capsule Data from Farley Unit1 and Calvert Cliffs Unit1 Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV) FF(b)

RTNDT(c)

(°F)

FF*RTNDT

(°F)

FF2 Weld Metal Heat # 33A277 (Farley Unit1 data)

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 X

3.06 1.295 146.2 (87.4) 189.42 1.678 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 (Calvert Cliffs Unit1 data) 263° 0.62 0.866 73.1 (59) 63.35 0.750 97° 2.64 1.260 109.2 (93) 137.54 1.587 SUM:

1436.99 12.396 CFHeat # 33A277 = (FF

• RTNDT) ÷ (FF2) = (1436.99) ÷ (12.396) = 115.9F (a) f = fluence.

(b)

FF = fluence factor = f(0.28 - 0.10*log f).

(c)

RTNDT values are the measured 30 ft-lb shift values. All values are taken from Table 4.2-2 of WCAP-17315-NP. RTNDT 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) Ratio = CFVessel Weld/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°F/119.4°F = 1.06

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TABLE 6.0-4 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Material Description Cu (%)

Ni(%)

Initial RTNDT (F)

Upper Shell Plate (b)

B4105-1 B4105-2 B4105-3 0.12 0.12 0.14 0.56 0.57 0.56 28 9

14 Inter Shell Plate B4106-1 B4106-2 B4106-3 0.125 0.12 0.086 0.53 0.50 0.476

-10

-3 30 Lower Shell Plate B4107-1 B4107-2 B4107-3 0.13 0.12 0.12 0.56 0.56 0.52 15 20

-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) Per CE NPSD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E + 17.

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TABLE 6.0-5 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu (%)

Ni(%)

Initial RTNDT (F)

Upper Shell Plate (b)

B5453-1 B5453-3 B5011-1R 0.11 0.11 0.11 0.60 0.60 0.65 28 5

0 Inter Shell Plate B5454-1 B5454-2 B5454-3 0.14 0.14 0.15 0.65 0.59 0.62 52 67 33 Lower Shell Plate B5455-1 B5455-2 B5455-3 0.14 0.14 0.10 0.56 0.56 0.62

-15 0

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) Per CE NSPD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E + 17.

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TABLE 6.0-6 DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 1/4t, and 3/4t Locations at 35 EFPY Material (a)

Fluence f1/4t Fluence f3/4t Inter Shell Plate B4106-1 B4106-2 B4106-3 7.98 E + 18 7.98 E + 18 7.98 E + 18 2.83 E + 18 2.83 E + 18 2.83 E + 18 Lower Shell Plate B4107-1 B4107-2 B4107-3 7.98 E + 18 7.98 E + 18 7.98 E + 18 2.83 E + 18 2.83 E + 18 2.83 E + 18 Inter Shell Long Welds 2-442 A,B Weld 2-442 C 5.89 E + 18 3.07 E + 18 2.09 E + 18 1.09 E + 18 Inter Shell to Lower Shell Weld 9-442 7.98 E + 18 2.83 E + 18 Lower Shell Long Welds 3-442 A,B Weld 3-442 C 4.87 E + 18 7.98 E + 18 1.73 E + 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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TABLE 6.0-7 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 1/4t and 3/4t Locations at 35 EFPY Material (a)

Fluence f1/4t Fluence f3/4t Inter Shell Plate B5454-1 B5454-2 B5454-3 9.04 E + 18 9.04 E + 18 9.04 E + 18 3.21 E + 18 3.21 E + 18 3.21 E + 18 Lower Shell Plate B5455-1 B5455-2 B5455-3 9.04 E + 18 9.04 E + 18 9.04 E + 18 3.21 E + 18 3.21 E + 18 3.21 E + 18 Inter Shell Long Weld 2-201 A Weld 2-201 B Weld 2-201 C 5.01 E + 18 6.06 E + 18 5.16 E + 18 1.78 E + 18 2.15 E + 18 1.83 E + 18 Inter Shell to Lower Shell Weld 9-201 9.04 E + 18 3.21 E + 18 Lower Shell Long Weld 3-201 A Weld 3-201 B Weld 3-201 C 5.16 E + 18 5.01 E + 18 6.06 E + 18 1.83 E + 18 1.78 E + 18 2.15 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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TABLE 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 35 EFPY 35 EFPY ART(a)

Material RG 1.99 Rev 2 Method 1/4t (F) 3/4t (F)

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 RTNDT + RTNDT + Margin (°F)

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16A PAGE 36 OF 37 TITLE:

PTLR for Diablo Canyon UNITS 1 AND 2 PTLR-1u3r16.DOC 04B 0416.1038

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TABLE 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 35 EFPY 35 EFPY ART(a)

Material RG 1.99 Rev 2 Method 1/4t (F) 3/4t (F)

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 RTNDT + RTNDT + Margin (°F)

(b)

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

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 16A PAGE 37 OF 37 TITLE:

PTLR for Diablo Canyon UNITS 1 AND 2 PTLR-1u3r16.DOC 04B 0416.1038

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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 Location 1/4t(d) 3/4t(e)

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

Fluence 1019 n/cm2 (E > 1.0 MeV), f(a) 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 (F)(c) 34 34 ART = I + (CF x FF) + M (F) per Regulatory Guide 1.99, Rev 2 197.8(f) 169.5(f)

Calc N-288 Rev 4 (a) Fluence, f, is based upon f1/4t and f3/4t 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(0.28 - 0.10*log f).

(c) Margin is calculated as M = 2(I2+ 2)0.5. The standard deviation for the initial RTNDT margin term I, is 0F for plate since the initial RTNDT is a measured value. The standard deviation for RTNDT term,

is 17F for the plate, except that 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 1/4t.

(e) DCPP-2 intermediate shell plate B5454-2 is limiting for the heatup and cooldown Appendix G curves at 3/4t.

(f) The higher CF based on CE 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 1/4t and 169.5°F for 3/4t.

PG&E Letter DCL-23-117 PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)-1 REVISION 11 EFFECTIVE DATE: February 29, 2012

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PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT NUMBER PTLR-1 REVISION 11 PAGE 1 OF 32 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE:

PTLR fol' Diablo Canyon 1

AND 2 INFO ONLY EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 2 OPERATING LIMITS...................................................................................................................................... 2 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3)............................................................................ 2 Low Temperature Overpressure Protection (L TOP) System Setpoints (LCO 3.4.12).................................. 5 ADDITIONAL CONSIDERATIONS............................................................................................................. 15 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM.............................................................. 15 REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY.................................................................. 16 SUPPLEMEN.TAL DATA TABLES.............................................................................................................. 21 PRESSURIZED THERMAL SHOCK (PTS) SCREENING.......................................................................... 22 REFERENCES................................................................................................................................................ 22 Figure 2.1-1 2.1-2 Table

2. 1-1 2.1-2 2.2-1 2.2-2 5.0-1 5.0-2 List of Figures Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr) Applicable to 23.8 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100°F/hr) Applicable to 23.8 EFPY (Unit I and Unit 2) (Without Margins for Instrumentation Errors)

List of Tables Diablo Canyon Heatup Data at 23.8 EFPY (Unit 1 and Unit 2) With Margins for Instrumentation Errors Diablo Canyon Cooldown Data at 23.8 EFPY (Un it 1 and Unit 2) With Margins for Instrumentation Errors L TOP System Setpoints LTOP Temperature Restrictions Diablo Canyon Unit 1 Surveillance Capsule Data Diablo Canyon Unit 2 Surveillance Capsule Data v13-PTLR-1 u3rl I.doc 048 1004.1056 PAGE 8

11 9

12 14 14 19 20

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PACIFIC GAS AND ELECTRIC COMP ANY DIABLO CANYON POWER PLANT NUMBER PTLR-1 REVISION 11 PAGE 2 OF 32 TITLE:

PTLR for Diablo Canyon UNITS 1AND2

1.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR for Diab lo Canyon has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems The limits provided in this repo1t remain valid until 23.8 EFPY on Unit 1 and Unit 2.

2.

OPERATING LIMITS 2.1 RCS Pressure and Temperature (P/T) 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 cooldown of 100°F in any 1-hour period.

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

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

2.1.1 RCS PIT Limits:

vl3-PTLR-lu3rl I.doc 04B 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, detennine the maximum pe1111issible stress intensity correlated to the reference stress intensity (KJR) as a function of vessel metal temperature, define the size of the assumed flaw, and apply specified safety factors.

The reference stress intensity (Krn) is the combined thennal and pressure stress intensity limit at a given temperah1re. The assumed crack has a radial depth of 1/4 of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

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PTLR for Diablo Canyon PAGE UNITS 3 OF 32 1AND2 vl3-PTLR-lu3rl I.doc 04B 10 CFR 50 Appendix G and Reg. Guide 1.99 provide guidelines for determining the maximum pennissible (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 tluence 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 fonnula 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 Teclrnical & 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 geometTy 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 assumption 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 during heatup.

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 11 PAGE 4 OF 32 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 perfonned with fuel in the core.

To meet Condition l.a of 10 CFR Appendix G, Table 1, the limiting temperature for the closure flange is the Unit 1 head flange that has an RTNDT of 53 °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 53°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 d6 exceed 621 psig pressure is 143°F (RTNDT+ 90°F). For Condition l.c, the limiting material is Unit 1 lower shell weld 3-442 C based on an ART of 198.3°F. For this pre-service hydro test, with no fi.1el in the vessel, the minimum RCS temperature for all pressures is 258.3°F (RTNDT+ 60°F). The limiting temperature for all these conditions is fm 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 Lim its:

v13-PTLR-lu3rll.doc 04B 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 RT NDT shift, and, thus minimum Bolt Up Temperahll'e does not change with time. The highest flange RTNDT between DCPP Unit 1 and 2 is 53 deg F (Unit 1 R.V. closure head). The curves conservatively set the temperatme at 60 deg F based on WCAP 14040-NP-A minimum temperature. Between the minimum bolt up temperature and the minimum LTOP operating temperahire (72 deg F), a 2.07 sq. in. opening is relied on for RCS venting. This satisfies Condition 2.a of the 10 CFRAppendix G, Table l.

To comply with Condition 2.b of 10 CFR Appendix G, Table 1, the pressure temperature limits impose a minimum temperahire of 173°F (RTi'-lm of 53°F + 120°F) at pressures not exceeding the 20% hydro test pressure or 621 psig. 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 temperahll'e be at least 40°F greater than the corresponding ASME Appendix G limit. The minimum temperature for criticality is a minimum temperature for the In-service system hydrostatic pressure temperature, which is 2459 psig. The corresponding temperature for a hydrostatic test at 2459 psig is 327.9°F. Thus, the minimum temperahire at with the core may be critical is 330°F.

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PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT NUMBER PTLR-1 REVISION 11 PAGE 5 OF 32 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 settings and an arming temperature in accordance with Table 2.2-1.

Plant equipment shall be operated in accordance with the restrictions of Table 2.2-2.

2.2.1 L TOP Enable Setpoints:

The LTOP lift setpoint and arming tempernture 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, Januaiy 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 which ever is greater in accordance with ASME Code Case N-514. The RETRAN-02 Mod3 computer code (Ref. 8.6) was used to petform the thennal hydraulic analysis and verify 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:

v13-PTLR-I u3rl I.doc 04B 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 1 im iting single failure of one pressurizer POR V to open, there is still a qualified PORV available to adequately relieve the RCS system pressure.

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 overs~oot 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.

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PACIFIC GAS AND ELECTRIC COMP ANY DIABLOCANYONPOWERPLANT NUMBER PTLR-1 REVISION 11 PAGE 6 OF 32 TITLE:

PTLR for Diablo Canyon UNITS 1AND2 The administrative temperature restrictions in Table 2.2-2 are established based on the most limiting RCS overshoot results obtained from the spectrum of mass injection and heat injection cases evaluated at the specified RCS 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 (SI) 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 L TOP 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.

2.2.4 L TOP Heat Injection Case:

v13-PTLR-lu3rll.doc 048 The heat injection cases are based on starting an RCP in one loop with a maximum allowable measured temperature difference of 50 °F 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 withi11 the LTOP range. The heat injection RCS overshoot cases were determined to remai11 below the Appendix G PIT curve and are conservatively bounded by the mass injection overshoot results throughout the L TOP temperature range. The heat injection cases establish that there are no LTOP admi11istrative RCS temperature restrictions for starting an RCP when the measured SG temperatme 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.

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PTLR for Diablo Canyon UNITS 1AND2 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.

2.2.6 Measurement Unce1tainties:

vl 3-PTLR-lu3rl I.doc 04B The LTOP mass injection and beat 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 unce1tainties 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 unce1tainty for the RCS wide range pressure transmitters associated with both PORVs opening and closing simultaneously.

Since each PORV has a normal and independent setpoint unce1tainty 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 COMP ANY DIABLO CANYON POWER PLANT TITLE:

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PTLR-1 11 8 OF 32 1AND2 FIGURE 2.1-1: Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/lu")

Applicable to 23.8 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) v 13-PTLR-I u3rl I.doc 048 1004.1056

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PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT TITLE:

PTLR for Diablo Canyon TABLE 2.1-l NUMBER PTLR-1 REVISION 11 PAGE 9 OF 32 UNITS 1AND2 Diablo Canyon Heatup Data at 23.8 EFPY (Unit 1 and Unit 2)

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

Press.

Temp.

Press.

Temp.

Press.

Temp.

Press.

(OF)

(psig)

(OF)

(psig)

(OF)

(psig)

(OF)

(psig) 75 469.7 75 469.7 80 471.0 80 468.9 85 468.0 85 453.2 90 466.3 90 435.8 95 467.9 95 424.3 100 470.9 100 424.0 105 474.l 105 424.2 110 478.0 110 424.5 115 482.8 115 425.4 120 488.2 120 426.6 125 494.4 125 428.5 130 501.1 130 431.0 135 508.5 135 434.3 140 516.5 140 438.2 145 524.8 145 442.8 150 533.0 150 448.2 155 541.7 155 453.4 160 550.8 160 460.6 165 559.0 165 468.7 170 567.8 170 475.9 175 577.1 175 485.3 180 587.1 180 496.0 185 597.8 185 506.8 190 609.2 190 518.4 195 621.5 195 531.0 200 634.6 200 544.5 205 648.6 205 559.1 210 663.6 210 574.7 215 679.6 215 591.6 220 696.8 220 609.7 225 715.2 225 629.1 230 734.9 230 650.0 235 756.0 235 672.4 vl3-PTLR-lu3rl I.doc 04B 1004.1056

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PACIFIC GAS AND ELECTRIC COMP ANY DIABLO CANYON POWER PLANT TITLE:

PTLR for Diablo Canyon TABLE 2.1-1 NUMBER PTLR-1 REVISION 11 PAGE 10 OF 32 UNITS 1AND2 Diablo Canyon Heatup Data at 23.8 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)

(psig)

(OF)

(psig)

(OF)

(psig)

(OF)

(psig) 240 778.6 240 696.5 245 802.9 245 722.3 250 829.0 250 750.0 255 856.9 255 779.7 260 886.8 260 811.6 260 1182.4 265 919.0 265 845.8 265 1224.8 270 953.5 270 882.5 270 1270.3 275 990.5 275 921.9 275 1319.1 280 1030.2 280 964.1 280 1371.4 285 1072.9 285 1009.5 285 1427.6 290 1118.6 290 1058.1 330.0 1058.1 290 1487.8 295 1167.6 295 1110.2 335.0 1110.2 295 1552.4 300 1220.2 300 1166.1 340.0 1166.1 300 1621.6 305 1272.7 305 1226.0 345.0 1226.0 305 1695.9 310 1327.8 310 1290.0 350.0 1290.0 310 1775.4 315 1386.9 315 1358.6 355.0 1358.6 315 1860.7 320 1450.3 320 1432.1 360.0 1432.1 320 1951.9 325 1518.2 325 1492.6 365.0 1492.6 325 2049.6 330 1590.9 330 1556.7 370.0 1556.7 330 2154.1 335 1668.9 335 1624.7 375.0 1624.7 335 2265.9 340 1752.3 340 1697.7 380.0 1697.7 340 2385.2 345 1841.7 345 1776.0 385.0 1776.0 345 2512.7 350 1937.3 350 1859.7 390.0 1859.7 350 2648.6 355 2039.5 355 1949.2 395.0 1949.2 355 2793.4 360 2148.8 360 2044.9 400.0 2044.9 360 2947.4 365 2265.6 365 2147.1 405.0 2147.1 370 2390.3 370 2256.3 410.0 2256.3 375 2523.3 375 2372.9 415.0 2372.9 380 2665.0 380 2497.1 420.0 2497.1 385 2815.9 385 2629.6 425.0 2629.6 390 2976.2 390 2770.6 430.0 2770.6 395 3146.5 395 2920.5 435.0 2920.5 400 3326.9 400 3079.7 440.0 3079.7 Ref. Cale. N-291 vl3-PTLR-lu3rll.doc 048 1004.1056

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PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT TITLE:

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I 450 FIGURE 2.1-2: Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100°F/lu') Applicable to 23.8 EFPY (Unit l and Unit 2) (Without Margins for Instrumentation Errors) vl3-PTLR-lu3rl I.doc 048 1004. 1056

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PTLR for Diablo Canyon TABLE 2.1-2 NUMBER PTLR-1 REVISION 11 PAGE 12 OF 32 UNITS 1AND2 Diablo Canyon Cooldown Data at 23.8 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)

(psig)

(OF)

(psig)

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(psig) 350 2009.6 350 2009.6 350 2009.6 350 2009.6 350 2009.6 345 1904. l 345 1904.1 345 1904.1 345 1904.1 345 1904.1 340 1805.6 340 1805.6 340 1805.6 340 1805.6 340 1805.6 335 1713.5 335 1713.5 335 1713.5 335 I 713.5 335 1713.5 330 1627.5 330 1627.5 330 1627.5 330 1627.5 330 1627.5 325 1547.3 325 1547.3 325 1547.3 325 1547.3 325 1547.3 320 1472.5 320 1472.5 320 1472.5 320 1472.5 320 1472.5 315 1402.7 315 1402.7 315 1402.7 315 1402.7 315 1402.7 310 1337.7 310 1337.7 310 1337.7 310 1337.7 310 1337.7 305 1277.0 305 1277.0 305 1277.0 305 1277.0 305 1277.0 300 1220.4 300 1220.4 300 1220.4 300 1220.4 300 1220.4 295 1167.8 295 1165.7 295 1167.8 295 1167.8 295 1167.8 290 1118.6 290 1113.8 290 1115.3 290 1118.6 290 1118.6 285 1072.9 285 1063.5 285 1059.7 285 1062.7 285 1072.9 280 1030.2 280 1016.6 280 1008.0 280 1005.4 280 1010.1 275 990.5 275 972.9 275 959.9 275 952.2 275 951.0 270 953.5 270 932.3 270 915.0 270 902.6 270 896.2 265 919.0 265 894.4 265 873.2 265 856.3 265 844.8 260 886.8 260 859.1 260 834.4 260 813.4 260 797.2 255 856.9 255 826.3 255 798.3 255 773.4 255 752.9 250 829.0 250 795.8 250 764.7 250 736.4 250 711.8 245 802.9 245 767.3 245 733.5 245 701.9 245 673.6 240 778.6 240 740.8 240 704.4 240 669.9 240 638.2 235 756.0 235 716.1 235 677.4 235 640.2 235 605.3 230 734.9 230 693.2 230 652.3 230 612.6 230 574.7 225 715.2 225 671.7 225 628.9 225 586.9 225 546.5 220 696.8 220 651.8 220 607.2 220 563.1 220 520.3 215 679.6 215 633.2 215 586.9 215 541.0 215 495.9 210 663.6 210 615.9 210 568.1 210 520.4 210 473.3 205 648.6 205 599.7 205 550.6 205 501.3 205 452.3 200 634.6 200 584.6 200 534.2 200 483.5 200 432.9 195 621.5 195 570.5 195 519.0 195 467.0 195 414.8 190 609.2 190 557.4 190 504.9 190 451.7 190 398.1 185 597.8 185 545.2 185 491.7 185 437.4 185 382.5 vl3-PTLR-lu3r11.doc 04B 1004.1056

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PTLR fol' Diablo Canyon TABLE 2.1-2 NUMBER PTLR-1 REVISION 11 PAGE 13 OF 32 UNITS 1AND2 Diablo Canyon Cooldown Data at 23.8 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)

(psig)

(OF)

(psig)

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(psig)

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(psig) 180 587. 1 180 533.7 180 479.4 180 424.1 180 368.1 175 577. 1 175 523. 1 175 468.0 175 411.8 175 354.7 170 567.8 170 513. 1 170 457.3 170 400.3 170 342.3 165 559.0 165 503.8 165 447.4 165 389.7 165 330.7 160 550.8 160 495.1 160 438.1 160 379.8 160 320.0 155 543.1 155 487.0 155 429.5 155 370.5 155 310.1 150 535.9 150 479.4 150 421.5 150 362.0 150 300.9 145 529.2 145 472.3 145 414.0 145 354.0 145 292.4 140 522.9 140 465.7 140 407.0 140 346.6 140 284.5 135 517.0 135 459.6 135 400.6 135 339.8 135 277.2 130 511.5 130 453.8 130 394.5 130 333.4 130 270.4 125 506.4 125 448.5 125 388.9 125 327.5 125 264.2 120 50 1.5 120 443.5 120 383.7 120 322.1 120 258.4 115 497.0 115 438.9 11 5 378.9 115 317.0 11 5 253.1 I 10 492.8 110 434.5 110 374.4 110 312.4 110 248.2 105 488.9 105 430.5 105 370.3 105 308.l 105 243.7 100 485.2 100 426.8 100 366.5 100 304.1 100 239.6 95 481.8 95 423.3 95 362.9 95 300.5 95 235.8 90 478.6 90 420.1 90 359.6 90 297.2 90 232.3 85 475.6 85 4 17.l 85 356.6 85 294.1 85 229.2 80 472.7 80 414.1 80 353.7 80 29 1.1 80 226. 1 75 469.9 75 411.4 75 350.9 75 288.5 75 223.3 70 467.2 70 408.7 70 348.3 70 285.7 70 220.7 Cale. N-291 v13-PTLR-lu3r l I.doc 048 1004.1056

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PTLR for Diablo Canyon Table 2.2-1 Low Temperature Over-Pressure (LTOP)

System Setpoints Function PORV Anning Temperature(ll PORV Pressure Setpoint<2l (1) Cale. N-298, Rev. 1. Valid to 23.8 EPPY (2) STA-249, Rev. 1 Table 2.2-2 Low Temperature Over-Pressure (LTOP)

Temperature Restrictions Restriction SI Pumps Secured, CCP 1 or CCP 2 Secured, SI Accumulators Isolated, CCP 3 aligned for LTOP operation Safety Injection Plowpath Blocked, and SI Blocked 2 of 3 Charging Pumps Secured 1 of 4 RCPs Secured 2 of 4 RCPs Secured 3 of 4 RCPs Secured 4 of 4 RCPs Secured RCS Vent Path of2.07 in2 Established (1) Cale. STA-249, Rev. 1 Assumptions:

1) PORV Stroke Time of 2.9 seconds.

2) Apply 10 % per Code Case N-514.

v I 3-PTLR-1 u3rl I.doc 04B 1004.1056 NUMBER PTLR-1 REVISION 11 PAGE 14 OF 32 UNITS 1AND2 Setpoint 2: 280 °P 435 psig Setpoint RSGs< 1l S 280 °P S 169 °P S 156 °P S 148 °P S 132 °P S 117 °P S 108 °F S 90 °F

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3.

ADDITIONAL CONSIDERATIONS NUMBER PTLR-1 REVISION 11 PAGE 15 OF 32 UNITS 1AND2 Revisions to the PTLR or its suppmting 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 irtiections case throughout the cmTent 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 Repmi (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 L TOP 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 Diab lo Canyon Unit J, 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 of Capsule S from Diablo Canyon Unit I Reactor Vessel Radiation Surveillance Program, December, 1987.

4.5 WCAP-13750, Analysis of Capsule Y from Diablo Canyon Unit I 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 Diab lo 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.

vl3-PTLR-tn3rt I.doc 048 1004.1056

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PTLR for Diablo Canyon NUMBER PTLR-1 REVISION 11 PAGE 16 OF 32 UNITS 1AND2 Diab lo Canyon Units 1 and 2 also have Reactor Cavity Neutron Measurement Programs described 111:

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 l Reactor Pressure Vessel, December, 2001.

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

4.14 WCAP-15782, Fast Neutron Fluence and Neutron Dosimet1y Evaluations for the Diablo Canyon Unit 2 Reactor Pressure Vessel, December, 2001.

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

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

5.

REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY Regulat01y 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 Diab lo Canyon Unit 2 reactor vessel. They must be shown to be credible in order to use these smveillance data sets. There are five requirements that must be met for the surveillance data to be judged credible in accordance with Regulatoty 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 CFR Patt 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."

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PTLR for Diablo Canyon NUMBER PTLR-1 REVISION 11 PAGE 17 OF 32 UNITS 1AND2 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 1/4t location is found in Seam Weld 3-442 C in the Unit I reactor vessel while the most limiting 3/4t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel. The Unit 1 Weld Surveillance Capsules are fabricated from a weld manufactured using the same weld wire heat number (Heat 27204).

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 propetiies (Cu content is identical and Ni content is 0.06% higher than the controlling material). The Diab lo Canyon Surveillance Program meets the intent of this criterion.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the i1rndiated and unirradiated 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 of the RT NDT at 30 ft-lb and upper shelf energy.

Criterion 3:

Where there are two or more sets of surveillance data from one reactor, the scatter of L'1RT NDT values about a best-fit line drawn as described in Regulato1y Position 2.1 normally should be less than 28°F for welds and I 7°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 E 185-82.

Tables 5.0-1 and 5.0-2 present the Surveillance Capsule Data for Diablo Canyon Units 1 and 2. The scatter of L'1RTNDT values about the functional form ofa best-fit line drawn as described in Regulatory Position 2.1 should be less than I CJ (standard deviation) of l 7°F for base metal and 28°F for weld material.

The Diab lo Canyon Unit 1 Surveillance Capsule S for the Intennediate Shell Plate 84106-3 and Surveillance Weld Heat 27204 both show scatter in excess of the Criterion 3 allowable values. The Diablo Canyon limiting CF values are based upon the CF Tables 1 and 2 of 10 CFR 50.61 and the chemistiy values provided by CE Repoti CE NPSD-1039, Rev 2. Should the credibility criteria be met upon future surveillance capsule withdrawal and evaluation, then Reg. Guide 1.99, Rev. 2, Position C.2 shall be utilized.

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.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 11 PAGE 18 OF 32 TITLE:

PTLR for Diablo Canyon UNITS 1AND2 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 con-elation 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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PTLR for Diablo Canyon Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data Best Fit Material Capsule CF<al FF LiRTNDT(b)

Inter Shell Plate S(d) 0.656 21.52 B4106-3 Inter Shell Plate y

32.8 1.014 33.26 B4106-3 fnter Shell Plate V

1.087 35.65 84106-3 Surveillance Weld S(d) 0.656 131.00 Heat 27204 Surveillance Weld y

199.7 1.014 202.50 Heat 27204 Surveillance Weld V

1.087 217.07 Heat 27204 WCAP 15958 NUMBER PTLR-1 REVISION 11 PAGE 19 OF 32 UNITS 1AND2 Measured Scatter in LiRTNoic)

LiRTNDT

-1.78

-23.3 48.66 15.4 34.32

-1.33 110.79

-20.21 232.59 30.09 201.07

-16.0 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulat01y Position 2.1 (see Table 6.0-3).

(b)

Best fit ti.RTNDT =CF* FF.

(cl Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

(dl Diablo Canyon Surveillance Capsule Sis currently not judged Credible per Reg. Guide 1.99, Rev 2, Position 2.1.

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PTLR for Diablo Canyon Table 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data Best Fit Material Capsule CF(*l FF ARTNoihl Inter Shell Plate u

0.701 69.)

B5454-l (Long)

Inter Shell Plate X

98.6 0.976 96.2 B5454-1 (Long)

Inter Shell Plate y

1.121 110.5 B5454-l (Long)

Inter Shell Plate V

1.237 122.0 B5454-1 (Long)

Inter Shell Plate u

0.701 69.1 B5454-1 (Trans)

Inter Shell Plate X

98.6 0.976 96.2 B5454-l (Trans)

Inter Shell Plate y

1.121 110.5 B5454-l (Trans)

Inter Shell Plate V

1.237 122.0 B5454-l (Trans)

Surveillance Weld u

0.701 138.2 Surveillance Weld X

197.2 0.976 192.5 Surveillance Weld y

1.121 221.1 Surveillance Weld V

1.237 243.9 WCAP-15423 NUMBER PTLR-1 REVISION 11 PAGE 20 OF 32 UNITS 1AND2 Measured Scatter in ARTNDT(c)

ARTNDT 65.4

-3.7 100.1 3.9 111.6 1.1 123.4 1.4 73.3 4.2 99.5 3.3 111.6 1.1 112.9

-9.1 173.0 34.8 203.2 10.7 211.4

-9.7 224.5

-19.4 (a)

CF is calculated from surveillance data using Reg. Guide 1.99 Regulat01y Position 2. I (see Table 6.0-3).

(bl Best fit ~RTNDr =CF* FF.

(cl Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

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PTLR for Diablo Canyon PAGE UNITS 21 OF 32 1AND2

6.

SUPPLEMENTAL DATA TABLES Table 6.0-1 Table 6.0-2 Table 6.0-3 Table 6.0-4 Table 6.0-5 Table 6.0-6 Table 6.0-7 Table 6.0-8 Table 6.0-9 Table 6.0-10 vl3-PTLR-lu3rl I.doc 04B Comparison of Diab lo Canyon Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Comparison of Diablo Canyon Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Calculation of Chernistty Factors Using Surveillance Capsule Data DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data DCPP-2 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the Vessel Surface, Clad to Base Metal Interface, 1/4t and 3/4t Locations at 23.8 EFPY DCPP-2 Summmy of the Projected Peak Pressure Vessel Neutron Fluence Values at the Vessel Surface, Clad to Base Metal Interface, 1/4t and 3/4t Locations at 23.8 EFPY Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 23.8 EFPY Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 23.8 EFPY Calculation of Adjusted Reference Temperature at 23.8 EFPY (Unit land Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials 1004.1056

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PTLR for Diablo Canyon

7.

PRESSURIZED THERMAL SHOCK (PTS) SCREENING PAGE UNITS 22 OF 32 1AND2 10 CFR 50.61 requires that RT rTs 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 of PTS require review and approval of the NRC. The maximum projected RT rrs for Units 1 and 2 is 250.9°F (Unit 1 Weld 3-442c), therefore, at a projected 32 EFPY at EOL, the PTS screening criteria is met. The PTS evaluations are described in the following reports:

7.1 WCAP-13771, Evaluation of Pressurized Thennal Shock for Diablo Canyon Unit I, July, 1993.

7.2 WCAP-14364, Evaluation of Pressurized Thennal Shock for the Diab lo Canyon Unit 2 Reactor Vessel, August, 1995.

7.3 PG&E Calculation N-287 (Unit 1) 7.4 PG&E Calculation N-272 (Unit 2)

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 (Ul)/135 (U2), dated May 28, 1999 8.3 License Amendment No. 133 (U 1 )/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 Cooldown Limit Cmves, Revision 2," January 1996 8.5 PG&E letter DCL-00-070, Supplement to Reactor Coolant System Pressure and Temperature Limits Repo1t 8.6 "RETRAN-02 A Program for Transient Thennal-Hydraulic Analysis of Complex Fluid Flow Systems", EPRI NP-1850-CCM-A, Project 889-3, December, 1996 8.7 PG&E Calculation N-288, Rev. 2, "Adjusted RT-NOT Versus EFPY" 8.8 PG&E Calculation N-291, Rev. 2, "Pressure-Temperature Limits for Heatup &

Cooldown" 8.9 PG&E Calculation N-298, Rev. 1. "LTOP Enable Temperature for 23.8 EFPY" 8.10 PG&E Calculation STA-249 Rev. 1, "RSG - LTOP Analysis" v13-PTLR-lu3rll.doc 048 1004.1056

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PTLR for Diablo Canyon Table 6.0-1 NUMBER PTLR-1 REVISION 11 PAGE UNITS 23 OF 32 1AND2 Comparison ofDiablo Canyon Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Materials Capsule Fluence (ct) 30 ft-lb Transition Upper Shelf Energy (X 1019 n/cm2)

Temperature Shift Decrease Predicted Measured Predicted Measured (oF) (a)

(OF) (h)

(%) (a)

(%) (c)

Plate 84106-3 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 C*l Based on Regulatmy 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 CV GRAPH, Version 4.1.

(cJ Values are based on the definition of upper shelf energy given in ASTM E 185-82.

(dJ The fluence values given here are the calculated tluence values, not the best estimate.

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PTLR for Diablo Canyon Table 6.0-2 N1JMBER PTLR-1 REVISION 11 PAGE UNITS 24 OF 32 1AND2 Comparison of Diab lo Canyon Unit 2 Surveillance Material 30 ft-lb Transition Temperatm*e 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 1019 n/cm2)

Temperature Shift Decrease P1*edicted Measured Predicted Measured (oF) (a)

(OF) (h)

(%) (n)

(%) (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 Cal 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 tluence values presented here are calculated fluence values, not the best estimate.

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PTLR for Diablo Canyon PAGE UNITS 25 OF 32 1AND2 Table 6.0-3 Calculation of Chemistry Factors Using Surveillance Capsule Data Unit 1 - Material Capsule F<*>

FF(b)

Measured FFxtiRTNDT°F FF2 6RTNDT(d) °F lntennediate Shell S (c) 0.284 0.656

-1.78 0

0.430 Plate B4106-3 y

1.050 1.014 48.66 49.34 1.028 V

1.37 1.087 34.32 37.31 1.182 SUM 86.65 2.64 CF Plate= 1(FF* 6RTNor)..;. L(FF2) = (86.65°F)..;. (2.64) = 32.8°F (c)

Weld Metal s<c) 0.284 0.656 l 10.79 72.68 0.430 y

1.050 1.014 232.59 235.85 1.028 V

1.37 1.087 201.07 218.56 1.182 SUM 527.09 2.64 CF weld= 1(FF* fiRTNDT)..;. L(FF2) = (527.09)..;. (2.64) = 199.7°F (c)

Unit 2 - Material Capsule F<n>

FF(b)

Measured FFx6.RTNvT°F FF2 6RTNDT(d) °F lnte1mediate Shell u

0.338 0.701 65.39 45.84 0.491 Plate X

0.919 0.976 100.06 97.67 0.953 B5454-1 (Long) y 1.55 1.121 111.58 125.08 1.257 V

2.41 1.237 123.43 152.68 1.530 Inte1mediate Shell u

0.338 0.701 73.30 51.38 0.491 Plate B5454-1 X

0.919 0.976 99.52 97.13 0.953 (Transverse) y 1.55 1.121 111.59 125.09 1.257 V

2.41 1.237 112.90 139.66 1.530 SUM 834.53 8.462 CF Plate= 1(FF* MT Nor)..;. L(FF2) = (834.53°F)..;. (8.462) = 98.6°F u

0.338 0.701 172.99 121.27 0.491 Weld Metal X

0.919 0.976 203.23 198.35 0.953 y

1.55 1.121 21 1.39 236.97 1.257 V

2.41 1.237 224.47 277.67 1.530 SUM 834.26 4.231 CF weld = 1(FF* tiRTNor)..;. L(FF2) = (834.26°F)..;. (4.231) = 197.2°F WCAP-15958 (Unit 1) WCAP-15423 (Unit 2)

(a)

F = Calculated Fluence (10 19 n/cm2, E > 1.0 MeV).

(b)

FF = Fluence Factor = F<0 0-1

• logF)

(c)

Unit 1 Capsule Sis not currently judged "credible" per RG 1.99, Rev 2. All other capsules are "credible" per RG 1.99, Position C.2.

(d)

Calculated using Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

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PTLR for Diablo Canyon TABLE 6.0-4 NUMBER PTLR-1 REVISION 11 PAGE 26 OF 32 UNITS 1AND2 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Mate1*ial Description Cu(%)

Ni(%)

Initial RTNDT (OF)

Upper Shell Plate (bl B4105-1 0.12 0.56 28 B4105-2 0.12 0.57 9

B4105-3 0.14 0.56 14 Inter Shell Plate B4 l 06-1 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 Cb)

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) 1.0 J8(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 o.203(a) 1.01 g(a)

-56 Cale N-NCM-97009 Cal Per CE NPSD-1039, Rev 2 (bl Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E+17.

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PTLR for Diablo Canyon TABLE 6.0-5 NUMBER J>TLR-1 REVISION 11 PAGE 27 OF 32 UNITS 1AND2 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu(%)

Ni(%)

Initial RTNoT (OF)

Upper Shell Plate (bl 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. IO 0.62 15 Upper Shell Long<bJ Welds 1-201 A,B,C 0.22 0.87

-50 Upper Shell to Inter Shell Weld 8-20 I (bl 0.183(a) 0.704(a)

-56 Inter Shell Long Welds 2-20 l 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 Cale N-NCM-97009 (al Per CE NSPD-1039, Rev. 2 (bl Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E+ 17.

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PTLR for Diablo Canyon TABLE 6.0-6 NUMBER PTLR-1 REVISION 11 PAGE 28 OF 32 UNITS 1AND2 DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 1/4t, and 3/4t Locations at 23.8 EFPY Material (a)

Flucnce f1/4t Flucnce f3/4t Inter Shell Plate B4106-1 5.48 E + 18 1.95 E+ 18 B4106-2 5.48E + 18 1.95 E + 18 B4106-3 5.48 E+ 18 1.95 E + 18 Lower Shell Plate B4107-l 5.48E+18 1.95 E + 18 B4107-2 5.48E+18 1.95 E+ 18 B4107~3 5.48E+18 1.95 E + 18 Inter Shell Long Welds 2-442 A,B 4.01 E + 18 1.43 E + 18 Weld 2-442 C 2.06 E + 18 7.31 E + 17 Inter Shell to Lower Shell Weld 9-442 5.48 E + 18 1.95 E+ 18 Lower Shell Long Welds 3-442 A,B 3.25 E + 18 1.15E+l8 Weld 3-442 C 5.48E + 18 1.95 E + 18 Cale N-288 Rev. 2, (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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PTLR for Diablo Canyon TABLE 6.0-7 NUMBER PTLR-1 REVISION 11 PAGE 29 OF 32 UNITS 1AND2 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the 1/4t and %t Locations at 23.8 EFPY Material (a)

Fluence f*/4t Fluence f 3/.,t Inter Shell Plate B5454-l 6.02 E + 18 2.14E+l8 B5454-2 6.02 E+ 18 2.14E+l8 B5454-3 6.02 E + 18 2.14E+ 18 Lower Shell Plate B5455-1 6.02 E + 18 2.14E+18 B5455-2 6.02 E+ 18 2.14E + 18 B5455-3 6.02 E+ 18

2. 14E+l8 Inter Shell Long Weld 2-201 A 3.34 E+ 18

l. 19E+l8 Weld 2-201 B 4.11 E+ 18 1.46 E + 18 Weld 2-201 C 3.50 E+ 18 l.24E+l8 Inter Shell to Lower Shell Weld 9-201 6.02E +l8 2.14£+18 Lower Shell Long Weld 3-201 A 3.50E+l8 1.24 E + 18 Weld 3-201 B 3.34 E + 18 l.19E + 18 Weld 3-201 C 4.IIE + l8 1.46 E + 18 Cale N-288 Rev. 2.

(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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PTLR for Diablo Canyon TABLE 6.0-8 NUMBER PTLR-1 REVISION 11 PAGE UNITS 30 OF 32 1AND2 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4t and 3/4t Locations for 23.8 EFPY 23.8 EFPY ART<al Material RG 1.99 Rev. 2 1/4t (°F) 3/4t (°F)

Method Inter Shell Plate B4106-1 Position 1.1 95.0 72.0 B4106-2 Position 1. l 98.4 76.6 B4106-3 Position 1.1 124.0 I 07. l Lower Shell Plate B4107-1 Position l. 1 123.7 99.5 B4107-2 Position l. 1 122.4 100.3 84107-3 Position 1.1 79.7 57.8 Inter Shell Long Welds 2-442 A,B Position 1.1 178.9 121.0 Weld 2-442 C Position 1.1 140.2 90.5 Inter Shell to Lower Shell Weld 9-442 Position 1. l 152.7 106.4 Lower Shell Long Welds 3-442 A,B Position 1.1 166.3 110.7 Weld 3-442 c <cJ Position 1.1 198.2(b) 137.2 Cale N-288 Rev. 2 (a)

ART= Initial RTNDT + LIBTNm + Margin (°F)

(bl This limiting ART value is bounded by that used to generate heatup and cooldown curves (198.3°F).

(clDCPP-1 Surveillance Capsule data were not judged "credible" per 10 CFR 50.61.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 11 PAGE 31 OF 32 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 1/4t and 3/4t Locations for 23.8 EFPY 23.8 EFPY ART<al Material RG 1.99 Rev. 2 1/4t (°F) 3/4t (°F)

Method Inter Shell Plate B5454-1 Position 2.1 159.7 130.9 B5454-2 Position 1.1 186.4 159.J(b)

B5454-3 Position 1.1 161.8 131.7 Lower Shell Plate B5455-l Position 1.1 103.2 76.5 B5455-2 Position 1.1 118.2 91.5 B5455-3 Position 1.1 104.9 87.2 Inter Shell Long Weld 2-201 A Position 1.1 153.5 101.5 Weld 2-201 B Position 1.1 165.1 110.9 Weld 2-201 C Position 1.1 156.1 103.6 Inter Shell to Lower Shell Weld 9-201 Position 1.1 18.0 3.3 Lower Shell Long Weld 3-201 A Position 1.1 99.3 67.9 Weld 3-201 B Position 1. l 97.7 66.6 Weld 3-201 C Position 1.1 I 04.7 72.3 Cale N-288 Rev. 2 (a)

ART = Initial RTNDT + LiRTNDT + Margin (°F)

(bl This limiting ART value is bounded by that used to generate heatup and cooldown curves (159.8°F).

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PTLR for Diablo Canyon TABLE 6.0-10 NUMBER PTLR-1 REVISION 11 PAGE UNITS 32 OF 32 1AND2 Calculation of Adjusted Reference Tempernture at 23.8 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials Parameter ART Value Location 1/4t(d) 3/.if(e)

Chemistry Factor, CF (°F) 226.8Ct1 99.6 Fluence + 1019 n/cm2 (E > 1.0 MeV), fM 0.548 0.214 Fluence Factor, FF(b) 0.8318 0.5853

~RTNDT = CF X FF, (°F) 188.6(t) 58.3 Initial RTNDT, I (°F)

-56 67 Margin, M (°F)<°l 65.5 34 ART= I+ (CF X FF)+ M (°F) 198.2(t) 159 _3(t) per Regulatory Guide 1.99, Rev. 2 Cale N-288 Rev. 2 (a)

Fluence, f, is based upon f1/41 and fot from Tables 6.0-6 and 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 0-23 -o.I01ogt)_

(c)

Margin is calculated as M = 2(ai2+ a;/)°'5. The standard deviation for the initial RTNDT margin term a1, is 0°F for plate since the initial RT NOT is a measured value. The standard deviation for ~RT NOT term CTL\\,

is l 7°F for the plate, except that CTL\\ need not exceed the 0.5 times the mean value of ~RT NDT, (d)

DCPP-1 lower shell longitudinal weld 3-442 C is limiting for the heatup and cooldown Appendix G curves at 1/4t.

(e)

DCPP-2 intennediate shell plate B5454-2 is limiting for the heatup and cooldown Appendix G curves at 3/4t.

(t)

DCPP-1 Surveillance Capsule data were not judged credible" per 10 CFR 50.61. The higher chemistry value of CE NPSD-1039, Rev 2 for this heat are used to generate the heatup and cooldown Appendix G curves. The ART's used to generate the heatup and cooldown curves are 198.3°F for l/4t and 159.8°F for 3/4t.

vl3-PTLR-lu3r11.doc 048 1004.1056