DCL-14-011, Reactor Coolant System Pressure and Temperature Limits Report

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Reactor Coolant System Pressure and Temperature Limits Report
ML14044A042
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 02/13/2014
From: Allen B
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-14-011
Download: ML14044A042 (36)
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Text

Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.4888 February 13, 2014 Internal: 691.4888 Fax: 805.545. 6445 PG&E Letter DCL-14-011 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Reactor Coolant System Pressure and Temperature Limits Report for Units 1 and 2

Dear Commissioners and Staff:

In accordance with Diablo Canyon Power Plant Technical Specification 5.6.6.c, Pacific Gas & Electric Company (PG&E) is submitting the enclosed Revision 13 of the Pressure and Temperature Limits Report (PTLR) for Units 1 and 2, dated January 14, 2014.

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. Hector Garcia at (805) 545-3942.

Sincerely, J813/4486/50554955 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Thomas R. Hipschman, NRC Senior Resident Inspector James S. Kim, NRR Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Enclosure PG&E Letter DCL-14-011 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

DIABLO CANYON POWER PLANT, UNITS 1 AND 2 EFFECTIVE DATE: January 14, 2014

      • ISSUED FOR USE B Y : - - - - - - - - DATE:----- EXPIRES:-----***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 NUCLEAR POWER GENERATION REVISION 13 DliffiLOCANYONPOWERPLANT PAGE 1 OF 34 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE: PTLR for Diablo Canyon 1 AND2 01/14/14 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).2 OPERATING LIMITS ....................................................................................................................................... 2 RCS Pressure and Temperature (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 SURVEILLANCEDATACREDIBILITY .................................................................. 18 SUPPLEMENTAL DATA TABLES .............................................................................................................. 23 PRESSURIZED THERMAL SHOCK (PTS) SCREENING *************************************************~************************24 REFERENCES ................................................................................................................................................ 24 List of Figures Figure PAGE 2.1-1 Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 8 60°F!hr) Applicable to 27 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors) 2.1-2 Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 11 0, 25, 50, 75 and 100°F/hr) Applicable to 27 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

List of Tables Table 2.1-1 Diablo Canyon Heatup Data at 27 EFPY (Unit 1 and Unit 2) With Margins for 9 Instrumentation Errors 2.1-2 Diablo Canyon Cooldown Data at 27 EFPY (Unit 1 and Unit 2) With Margins for 12 Instrumentation Errors 2.2-1 LTOP System Setpoints 14 2.2-2 LTOP Temperature Restrictions 14 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data 19 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data 20 PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 2 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2

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 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems The limits provided in this report remain valid until 27 EFPY on Unit 1 and Unit 2.
2. OPERATING LIMITS 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)

The RCS temperature rate-of-change limits are:

  • A maximum heatup of 60°F in any 1-hour period.
  • A maximum cooldown of 100°F in any 1-hour period.
  • A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

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

2.1.1 RCS PIT Limits:

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

The reference stress intensity (Km) is the combined thermal and pressure stress intensity limit at a given temperature. The assumed crack has a radial depth of

~ of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPOWERPLANT REVISION 13 PAGE 3 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 10 CFR 50 Appendix G and Reg. Guide 1.99 provide guidelines for determining the maximum permissible (allowable) stress intensity, based on nil-ductility of the reactor vessel metals during the operational life of the reactor. The transition temperature at which the metal becomes acceptably ductile is affected by neutron radiation embrittlement over the course of reactor operation. Appendix G and Reg. Guide 1.99 provide formulas which are used to calculate this Adjusted Reference Temperature based on fluence and vessel material chemistry. The shift in nil-ductility resulting from the fluence effect is added to the unirradiated nil-ductility transition temperature and, with Reg.

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

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

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

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

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

The membrane (pressure induced) stress can then be determined as a function of the allowable stress intensity reduced by thermal stress intensity and that difference divided by 2 as specified in ASME Section ITI 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 114t or 3/4t location.

PTLR-lu3rl3.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 4 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.1.2 RCS Pressure Test Limits:

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

To meet Condition 1.a of 10 CFR Appendix G, Table 1, the limiting temperature for the closure flange is the Unit 1 head flange that has an 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 l.c, the limiting material is Unit 1lower shell weld 3-442 C based on an ART of207.8°F. For this pre-service hydro test, with no fuel in the vessel, the minimum RCS temperature for all pressures is 267.8°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 270°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 (96 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 of35°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 337.3°F. Thus, the minimum temperature at which the core may be critical is 340°F.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 5 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12)

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

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

2.2.1 LTOP Enable Setpoints:

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

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

2.2.2 RCS Pressure Overshoot:

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

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 6 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 The RCS peak system pressure occurs at the bottom of the reactor vessel requiring that the elevation head be accounted for between this peak location and the RCS wide range pressure transmitters that generate the PORV open signal. In addition, the RHR pump and RCP flow impacts the PORV setpoint by generating a dynamic pressure drop across the reactor vessel which increases the difference between the RCS wide range pressure transmitters and the bottom of the reactor vessel. The magnitude of the total pressure drop determines the limiting RCS pressure at the bottom of the vessel for a given RCS overshoot case. An appropriate range of mass injection and heat injection cases are evaluated to ensure they conservatively bound the dynamic pressure drop effects due to the RCS flow conditions.

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. Per Note 2 on Table 2.2-2, an administrative exception has been established for the RCS vent temperature restriction when performing the RCS vacuum refill per procedure OP A-2:IX. Calculation STA-298 documents that when the RCS level is maintained at an elevation of less than 123', there is more than adequate time for operators to take action and preclude any credible water solid challenge to the LTOP system.

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 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.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DVffiLOCANYONPO~RPLANT REVISION 13 PAGE 7 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2.4 LTOP Heat Injection Case:

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

2.2.5 RCS Pressure Undershoot:

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 8 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 2.2.6 Measurement Uncertainties:

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

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

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

PTLR-1 u3r 13 .DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DliffiLOCANYONPO~RPLANT REVISION 13 PAGE 9 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 0 50 100 150 200 250 300 350 400 450 RCS TEMPERATURE {F)

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

Applicable to 27 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 10 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-1 Diablo Canyon Heatup Data at 27 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.

eF) (psig) eF) (psig) (oF) (psig) (oF) (psig) 75 467.0 75 465.4 80 469.3 80 466.1 85 465.8 85 452.5 90 463.9 90 435.9 95 465.6 95 421.5 100 468.0 100 413.7 105 471.6 105 416.6 110 475.9 110 418.9 115 481.1 115 421.1 120 486.9 120 423.0 125 493.3 125 425.1 130 500.3 130 427.4 135 507.9 135 430.2 140 514.5 140 433.6 145 521.1 145 437.5 150 527.5 150 442.0 155 534.2 155 446.4 160 541.4 160 452.6 165 549.1 165 459.6 170 557.4 170 466.6 175 566.3 175 473.6 180 575.8 180 482.8 185 586.0 185 492.9 190 596.9 190 503.3 195 608.6 195 514.4 200 621.1 200 526.3 205 634.6 205 539.2 210 648.9 210 553.3 215 664.3 215 568.2 220 680.8 220 584.4 225 698.4 225 601.6 230 717.4 230 620.3 235 737.7 235 640.3 240 759.4 240 661.6 PTLR-1 u3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 11 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-1 Diablo Canyon Heatup Data at 27 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.

eF> (psig) (oF) (psig) (oF) (psig) (oF) (psig) 245 782.7 245 684.8 250 807.6 250 709.6 255 834.4 255 736.3 260 863.1 260 765.0 265 893.9 265 795.6 270 926.9 270 828.1 275 962.3 275 863.5 280 1000.3 280 901.4 285 1041.0 285 942.2 285 1386.7 290 1084.7 290 985.8 290 1444.2 295 1131.6 295 1032.6 295 1505.9 300 1181.7 300 1082.7 300 1571.9 305 1234.6 305 1134.2 305 1642.8 310 1287.8 310 1184.2 310 1718.7 315 1344.1 315 1237.9 355 1339.6 315 1800.1 320 1404.6 320 1292.9 360 1392.8 320 1887.2 325 1469.4 325 1342.3 365 1449.8 325 1980.4 330 1538.8 330 1395.3 370 1510.4 330 2080.2 335 1613.1 335 1452.2 375 1575.5 335 2187.0 340 1692.8 340 1512.6 380 1645.3 340 2301.1 345 1778.0 345 1577.6 385 1719.9 345 2422.9 350 1869.2 350 1647.2 390 1799.8 350 2552.9 355 1966.8 355 1721.7 395 1885.1 355 2691.5 360 2071.3 360 1801.4 400 1976.4 360 2839.1 365 2182.7 365 1886.7 405 2074.0 365 2996.2 370 2301.9 370 1977.9 410 2178.3 370 3163.0 375 2429.1 375 2075.4 415 2289.6 375 3339.9 380 2564.7 380 2179.6 420 2408.3 380 3527.2 385 2709.0 385 2290.8 425 2534.9 385 3725.2 390 2862.7 390 2409.4 430 2669.7 390 3933.9 395 3025.8 395 2536.0 435 2813.2 395 4153.4 400 3199.2 400 2670.7 440 2965.8 400 4383.6 Ref. Calc. N-291 PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 12 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 0 50 100 150 200 250 300 350 400 450 RCS TEMPERATURE {F}

FIGURE 2.1-2: Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100°F/hr) Applicable to 27 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 13 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 27 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) (oF) (psig) (oF) (psig) 390 3029.6 390 3029.6 390 3029.6 390 3029.6 390 3029.6 385 2860.8 385 2860.8 385 2860.8 385 2860.8 385 2860.8 380 2701.9 380 2701.9 380 2701.9 380 2701.9 380 2701.9 375 2552.5 375 2552.5 375 2552.5 375 2552.5 375 2552.5 370 2412.4 370 2412.4 370 2412.4 370 2412.4 370 2412.4 365 2281.0 365 2281.0 365 2281.0 365 2281.0 365 2281.0 360 2157.9 360 2157.9 360 2157.9 360 2157.9 360 2157.9 355 2042.7 355 2042.7 355 2042.7 355 2042.7 355 2042.7 350 1935.0 350 1935.0 350 1935.0 350 1935.0 350 1935.0 345 1834.3 345 1834.3 345 1834.3 345 1834.3 345 1834.3 340 1740.2 340 1740.2 340 1740.2 340 1740.2 340 1740.2 335 1652.4 335 1652.4 335 1652.4 335 1652.4 335 1652.4 330 1570.3 330 1570.3 330 1570.3 330 1570.3 330 1570.3 325 1493.8 325 1493.8 325 1493.8 325 1493.8 325 1493.8 320 1422.4 320 1422.4 320 1422.4 320 1422.4 320 1422.4 315 1355.8 315 1355.8 315 1355.8 315 1355.8 315 1355.8 310 1293.7 310 1293.7 310 1293.7 310 1293.7 310 1293.7 305 1235.9 305 1235.9 305 1235.9 305 1235.9 305 1235.9 300 1181.9 300 1181.9 300 1181.9 300 1181.9 300 1181.9 295 1131.6 295 1128.1 295 1129.7 295 1131.6 295 1131.6 290 1084.7 290 1075.3 290 1074.6 290 1077.8 290 1084.7 285 1041.0 285 1028.3 285 1021.7 285 1020.9 285 1026.9 280 1000.3 280 983.7 280 972.2 280 966.3 280 967.3 275 962.3 275 942.1 275 926.1 275 915.2 275 910.7 270 926.9 270 903.2 270 883.2 270 867.7 270 857.9 265 893.9 265 867.0 265 843.3 265 823.5 265 808.9 260 863.1 260 833.3 260 806.2 260 782.5 260 763.4 255 834.4 255 801.9 255 771.6 255 744.3 255 721.1 250 807.6 250 772.6 250 739.5 250 708.8 250 681.8 245 782.7 245 745.3 245 709.6 245 675.9 245 645.3 240 759.4 240 719.9 240 681.8 240 645.3 240 611.4 235 737.7 235 696.3 235 655.9 235 616.8 235 579.9 230 717.4 230 674.2 230 631.8 230 590.3 230 550.7 PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 14 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 27 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.

eF) (psig) eF) (psig) eF) (psig) (oF) (psig) eF) (psig) 225 698.4 225 653.7 225 609.4 225 565.8 225 523.5 220 680.8 220 634.5 220 588.5 220 542.9 220 498.4 215 664.3 215 616.7 215 569.1 215 521.7 215 475.0 210 648.9 210 600.0 210 551.0 210 501.9 210 453.3 205 634.6 205 584.5 205 534.2 205 483.6 205 433.2 200 621.1 200 570.1 200 518.5 200 466.5 200 414.5 195 608.6 195 556.6 195 504.0 195 450.7 195 397.2 190 596.9 190 544.0 190 490.4 190 436.0 190 381.1 185 586.0 185 532.3 185 477.8 185 422.4 185 366.3 180 575.8 180 521.4 180 466.1 180 409.7 180 352.4 175 566.3 175 511.3 175 455.2 175 397.9 175 339.6 170 557.4 170 501.8 170 445.0 170 387.0 170 327.8 165 549.1 165 493.0 165 435.6 165 376.8 165 316.8 160 541.4 160 484.8 160 426.8 160 367.4 160 306.6 155 534.2 155 477.2 155 418.7 155 358.8 155 297.3 150 527.5 150 470.1 150 411.2 150 350.7 150 288.6 145 521.2 145 463.5 145 404.2 145 343.3 145 280.6 140 515.3 140 457.3 140 397.7 140 336.4 140 273.2 135 509.9 135 451.6 135 391.8 135 330.0 135 266.4 130 504.8 130 446.3 130 386.2 130 324.2 130 260.2 125 500.0 125 441.4 125 381.0 125 318.8 125 254.5 120 495.6 120 436.8 120 376.3 120 313.8 120 249.2 115 491.5 115 432.6 115 371.9 115 309.2 115 244.4 110 487.6 110 428.7 110 367.8 110 305.0 110 240.0 105 484.1 105 425.0 105 364.2 105 301.2 105 236.0 100 480.7 100 421.7 100 360.7 100 297.6 100 232.3 95 477.6 95 418.6 95 357.5 95 294.4 95 229.0 90 474.7 90 415.7 90 354.6 90 291.5 90 226.0 85 472.0 85 413.0 85 352.0 85 288.8 85 223.2 80 469.5 80 410.5 80 349.5 80 286.4 80 220.8 75 467.1 75 408.2 75 347.2 75 284.3 75 218.5 70 464.7 70 405.9 70 345.0 70 281.8 70 216.3 Calc. N-291 PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 15 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 2.2-1 Low Temperature Over-Pressure (LTOP)

System Setpoints Function Setpoint PORV Arming Temperature(!) 2:::283 °F PORV Pressure Setpoint<2> 435 psig (1) Calc. N-298, Rev 3. Valid to 27 EFPY (2) STA-249, Rev 3 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, ~ 283 °F CCP 3 aligned for LTOP operation Safety Injection Flowpath Blocked, and SI Blocked ~ 174 °F 2 of 3 Charging Pumps Secured ~ 161 °F 1 of 4 RCPs Secured ~ 153 °F 2 of 4 RCPs Secured ~ 137 °F 3 of 4 RCPs Secured ~ 123 °F 4 of 4 RCPs Secured ~ 114 °F RCS Vent Path of 2.07 in2 Established ~ 96 op(2)

(1) Calc. STA-249, Rev 3 Assumptions: 1) PORV Stroke Time of2.9 seconds.

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

(2) Calculation STA-298 establishes an exception for vacuum refill that an RCS vent is not required as long as the RCS temperature is greater than 91 °F and when the RCS level < 123'.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 16 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2

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

3.1 The PORV piping qualification under LTOP conditions is bounded by testing performed in accordance with NUREG 073 7.

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 CapsuleS 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.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 17 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 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.

Diablo Canyon Units 1 and 2 also have Reactor Cavity Neutron Measurement Programs described in:

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

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

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

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

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUillLOCANYONPO~RPLANT REVISION 13 PAGE 18 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2

5. REACTOR VESSEL SURVEILLANCE DATA CREDffiiLITY 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 defmed 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."

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 19 OF 34 TITLE: PTLR for Diablo Canyon 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 ~t location is found in Seam Weld 3-442 C in the Unit 1 reactor vessel while the most limiting %t 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 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.

Criterion 3: Where there are two or more sets of surveillance data from one reactor, the scatter of ilRTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

Tables 5.0-1 and 5.0-2 present the Surveillance Capsule Data for Diablo Canyon Units 1 and 2. The scatter of MTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 should be less than 1 cr (standard deviation) of 17°F for base metal and 28°F for weld material.

The Diablo Canyon Unit 1 Surveillance Capsule S data sets for the Intermediate Shell Plate B41 06-3 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 chemistry values provided by CE Report 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.

Per Calculation N-288 Rev 3, data for U2 Intermediate Shell Longitudinal Weld Metal Heat 21935/12008 also shows scatter in excess of Criterion 3 allowable values.

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.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 20 OF 34 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 correlation monitor material in the capsules fall within the scatter band for this (Correlation Monitor Material Heavy Section Steel Technology Plate 02) material.

PTLR-1 u3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 21 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data Best Fit Measured Scatter in Material Capsule CF<a> FF (b) (c)

ARTNDT ARTNDT ARTNDT Inter Shell Plate s 0.655 24.51 6.00 -18.51 B4106-3 Inter Shell Plate y 37.4 1.014 37.92 52.86 14.94 B4106-3 Inter Shell Plate v 1.085 40.60 37.82 -2.78 B4106-3 Surveillance Weld s 0.655 136.62 119.13 -17.49 Heat27204 Surveillance Weld y 208.5 1.014 211.33 241.53 30.20 Heat27204 Surveillance Weld v 1.085 226.30 208.66 -17.64 Heat27204 Calculation N-288 Rev 3, Table 1 (a)

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

(b)

Best fit L\RTNDT = CF

(c)

Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0, and adjusted for the temperature difference between RV inlet temperature during capsule irradiation and 538°F.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 22 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data Best Fit Measured Scatter in Material Capsule CF<a> FF ARTNDT(b) (c)

ARTNDT ARTNDT Inter Shell Plate u 0.695 73.45 80.30 6.85 B5454-1 (Trans)

Inter Shell Plate X 105.7 0.972 102.76 106.50 3.74 B5454-1 (Trans)

Inter Shell Plate y 1.118 118.12 118.60 0.48 B5454-1 (Trans)

Inter Shell Plate v 1.234 130.40 119.90 -10.50 B5454-1 (Trans)

Inter Shell Plate u 0.695 73.45 72.40 -1.05 B5454-1 (Long)

Inter Shell Plate X 105.7 0.972 102.76 107.10 4.34 B5454-1 (Long)

Inter Shell Plate y 1.118 118.12 118.60 0.48 B5454-1 (Long)

Inter Shell Plate v 1.234 130.40 130.40 0.00 B5454-1 (Long)

Surveillance Weld u 0.695 142.19 180.00 37.81 Surveillance Weld X 204.6 0.972 198.92 210.20 11.28 Surveillance Weld y 1.118 228.64 218.40 -10.24 Surveillance Weld v 1.234 252.42 231.50 -20.92 Calculation N-288 Rev 3, Table 2 (a)

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

(b)

Best fit ~TNDT = CF

(c)

Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0, and adjusted for the temperature difference between RV inlet temperature during capsule irradiation and 538°F.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 23 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2

6. SUPPLEMENTAL DATA TABLES Table 6.0-1 Comparison ofDiablo Canyon Unit 1 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 6.0-2 Comparison of Diablo Canyon Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 6.0-3 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, ~t and %t Locations at 27EFPY Table 6.0-7 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the Vessel Surface, Clad to Base Metal Interface, ~t and %t Locations at 27 EFPY Table 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and %t Locations for 27 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and %t Locations for 27 EFPY Table 6.0-10 Calculation of Adjusted Reference Temperature at 27 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 24 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2

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 of270°F for plates, forgings, and axial weld material, and 300°F for circumferential weld material. If the screening criterion is not met, specific actions taken to either meet the screening criterion or prevent potential reactor vessel failure as a result ofPTS require review and approval of the NRC. The maximum projected RT PTS for Units 1 and 2 is 249°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 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", EPRINP-1850-CCM-A, Project 889-3, December, 1996 8.7 PG&E Calculation N-288, Rev 3, "Adjusted RT-NDT Versus EFPY" 8.8 PG&E Calculation N-291, Rev 4, "Pressure-Temperature Limits for Heatup &

Cooldown" 8.9 PG&E Calculation N-298, Rev 3, "LTOP Enable Temperature for 27 EFPY" 8.10 PG&E Calculation STA-249 Rev 3, "RSG- LTOP Analysis" PTLR-lu3rl3.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPOWERPLANT REVISION 13 PAGE 25 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 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) 30 ft-lb Transition Upper Shelf Energy (X 1019 n!cm2) Temperature Shift Decrease Predicted Measured Predicted Measured eF) <a> eF) (b) (o/o) (a) (0/o) (c)

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

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

(b)

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

(c)

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

(d)

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 26 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-2 Comparison of Diablo Canyon Unit 2 Surveillance Material30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Fluence (c) 30 ft-lb Transition Upper Shelf Energy Materials Capsule (X 1019 nlcm2) Temperature Shift Decrease Predicted Measured Predicted Measured eF) <a> eF) (b) (%)(a) (o/o) (b)

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

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 27 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 Table 6.0-3 Calculation of Chemistry Factors Using Surveillance Capsule Data F(a) FF(b) Measured Unit 1 - Material Capsule ARTNDT(c)~ FFxARTNnT~ FF2 Intermediate Shell s 0.283 0.655 6.00 3.93 0.429 Plate B4106-3 y 1.050 1.014 52.86 53.58 1.027 v 1.360 1.085 37.82 41.05 1.178 SUM 98.56 2.635 CF Plate= :L(FF* dRTNDT) + :L(FF2) = (98.56°F) + (2.635) = 37.4°F Weld Metal s 0.283 0.655 119.13 78.07 0.429 y 1.050 1.014 241.53 244.83 1.027 v 1.36 1.085 208.66 226.49 1.178 SUM 549.38 2.635 2

CF weld= :L(FF* i1RTNDT) + :L(FF ) = (549.38) + (2.635) = 208.5°F F(a) Measured Unit 2 - Material Capsule FF

ARTNDT(c)oF FFxt1RTNDT~ FF2 Intermediate Shell u 0.330 0.695 72.40 50.32 0.483 Plate X 0.906 0.972 107.10 104.14 0.945 B5454-1 (Long) y 1.530 1.118 118.60 132.55 1.249 v 2.380 1.234 130.40 160.89 1.522 Intermediate Shell u 0.330 0.695 80.30 55.81 0.483 Plate B5454-1 X 0.906 0.972 106.50 103.55 0.945 (Transverse) y 1.530 1.118 118.60 132.55 1.249 v 2.380 1.234 119.90 147.94 1.522 SUM 887.76 8.400 CF Plate= L(FF* dRTNDT) + :L(FF2) = (887.76°F) + (8.400) = 105.7°F u 0.330 0.695 180.00 125.10 0.483 Weld Metal X 0.906 0.972 210.20 204.38 0.945 y 1.530 1.118 218.40 244.09 1.249 v 2.380 1.234 231.50 285.64 1.522 SUM 859.22 4.200 2

CF Weld= :L(FF* dRTNDT) + :L(FF ) = (859.22°F) + (4.200) = 204.6°F Calculation N 288 Rev 3, Table 1 (Unit 1) and Table 2 (Unit 2)

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

(b) FF = Fluence Factor= F<0*28 - O.I

  • logF)

(c) Calculated using Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0, and adjusted for the temperature difference between RV inlet temperature during capsule irradiation and 538°F.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 28 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE6.0-4 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Material Description Cu(%) Ni( 0/o) Initial RTNDT eF)

Upper Shell Plate (b)

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

Welds 1-442 A,B,C 0.19 0.97 -20 Upper Shell to Inter Shell Weld 8-44ib) 0.25 0.73 -56 Inter Shell Long Welds 2-442 A,B,C 0.203(a) 1.018(a) -56 Inter Shell to Lower Shell Weld 9-442 0.183(a) 0.704(a) -56 Lower Shell Long Welds 3-442 A,B,C 0.203(a) 1.018(a) -56 Calc N-NCM-97009 (a) PerCE NPSD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed l.OE + 17.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 29 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-5 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu (%,) Ni( 0/o) Initial RTNnT eF)

Upper Shell Plate (b)

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

Welds 1-201 A,B,C 0.22 0.87 -50 Upper Shell to Inter Shell Weld 8-201 (b) 0.183(a) 0.704(a) -56 Inter Shell Long Welds 2-201 A,B,C 0.22 0.87 -50 Inter Shell to Lower Shell Weld 9-201 0.046(a) o.o8ia) -56 Lower Shell Long Welds 3-201 A,B,C 0.258(a) 0.165(a) -56 Calc N-NCM-97009 (a) PerCE NSPD-1039, Rev 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed l.OE+ 17.

PTLR-lu3rl3.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 30 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-6 DCPP-1 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the %t, and

%t Locations at 27 EFPY Material (a) Fluence f~t Fluence f%t Inter Shell Plate B4106-1 6.19E+18 2.20 E+ 18 B4106-2 6.19E+18 2.20 E + 18 B4106-3 6.19E+18 2.20 E + 18 Lower Shell Plate B4107-1 6.19E+18 2.20 E+ 18 B4107-2 6.19E+18 2.20 E+ 18 B4107-3 6.19E+18 2.20 E+ 18 Inter Shell Long Welds 2-442 A,B 4.55 E + 18 1.62E+18 Weld 2-442 C 2.35 E+ 18 8.34 E+ 17 Inter Shell to Lower Shell Weld 9-442 6.19E+18 2.20 E+ 18 Lower Shell Long Welds 3-442 A,B 3.71E+18 1.32 E + 18 Weld 3-442 C 6.19 E+ 18 2.20 E + 18 Calc N-288 Rev 3 (a) Only beltline materials are included. W CAP-1 7315-NP demonstrates that extended beltline materials are not limiting through at least 54 EFPY.

PTLR-lu3rl3.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUffiLOCANYONPO~RPLANT REVISION 13 PAGE 31 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-7 DCPP-2 Summary of the Projected Peak Pressure Vessel Neutron Fluence Values at the ~t and %t Locations at 27 EFPY Material (a) Fluence f~t Fluence f~t Inter Shell Plate B5454-1 6.88E+18 2.44E+18 B5454-2 6.88E+18 2.44 E+ 18 B5454-3 6.88E+18 2.44E+18 Lower Shell Plate B5455-1 6.88E+18 2.44 E + 18 B5455-2 6.88 E + 18 2.44E+18 B5455-3 6.88E+18 2.44 E + 18 Inter Shell Long Weld2-201 A 3.82 E+ 18 1.36E+18 Weld2-201 B 4.67 E+ 18 1.66 E + 18 Weld2-201 C 3.98 E+ 18 1.41 E + 18 Inter Shell to Lower Shell Weld 9-201 6.88E+18 2.44E+ 18 Lower Shell Long Weld 3-201 A 3.98 E + 18 1.41 E+ 18 Weld3-201 B 3.82 E + 18 1.36E+ 18 Weld 3-201 C 4.67 E+ 18 1.66E+18 Calc N-288 Rev 3 (a) Only beltline materials are included. WCAP-17315-NP demonstrates that extended beltline materials are not limiting through at least 54 EFPY.

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPO~RPLANT REVISION 13 PAGE 32 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the ~t and %t Locations for 27 EFPY 27 EFPY ART<a>

Material RG 1.99Rev2

~teF) %teF)

Method Inter Shell Plate B4106-1 Position 1.1 97.9 74.5 B4106-2 Position 1.1 101.1 79.0 B4106-3 Position 1.1 125.9 109.9 Lower Shell Plate B4107-1 Position 1.1 126.7 102.2 B4107-2 Position 1.1 125.2 102.7 B4107-3 Position 1.1 82.5 60.2 Inter Shell Long Welds 2-442 A,B Position 1.1 186.6 127.4 Weld2-442 C Position 1.1 147.5 96.0 Inter Shell to Lower Shell Weld 9-442 Position 1.1 158.6 111.5 Lower Shell Long Welds 3-442 A,B Position 1.1 174.2 117.2 weld 3-442 c<c) Position 1.1 205.9(b) 143.9 Calc N-288 Rev 3 (a)

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

(b)

This limiting ART value is bounded by that used to generate heatup and cooldown curves (207.8°F, based on 28 EFPY).

(c)

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

PTLR-lu3r13.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 13 PAGE 33 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the %t and %t Locations for 27 EFPY 27 EFPY ART<a>

Material RG 1.99Rev2 %teF) %teF)

Method Inter Shell Plate B5454-1 Position 2.1 163.6 134.4 B5454-2 Position 1.1 190.2 162.6(b)

B5454-3 Position 1.1 165.9 135.3 Lower Shell Plate B5455-1 Position 1.1 106.9 79.7 B5455-2 Position 1.1 121.9 94.7 B5455-3 Position 1.1 107.4 89.3 Inter Shell Long Weld2-201 A Position 1.1 160.9 107.5 Weld2-201 B Position 1.1 172.4 117.0 Weld2-201 C Position 1.1 163.2 109.4 Inter Shell to Lower Shell Weld 9-201 Position 1.1 20.1 5.0 Lower Shell Long Weld 3-201 A Position 1.1 103.5 71.3 Weld3-201 B Position 1.1 102.2 70.2 Weld 3-201 C Position 1.1 109.0 75.9 Calc N-288 Rev 3 (a)

ART= Initial RTNDT + Lill.TNDT +Margin COF)

(b)

This limiting ART value is bounded by that used to generate heatup and cooldown curves (163.4°F, based on 28 EFPY).

PTLR-lu3rl3.DOC 04B 1224.0842

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DUBLOCANYONPOWERPLANT REVISION 13 PAGE 34 OF 34 TITLE: PTLR for Diablo Canyon UNITS 1AND2 TABLE 6.0-10 Calculation of Adjusted Reference Temperature at 27 EFPY (Unit 1 and Unit 2) for the Limiting

  • Diablo Canyon Reactor Vessel Materials Parameter ART Value

~t(d) %t(e)

Location Chemistry Factor, CF (°F) 226.8(+/-) 99.6 Fluence + 10 19 n/cm2 (E > 1.0 MeV), fa) 0.619 0.244 Fluence Factor, FF(b) 0.8658 0.6183

~RTNDT = CF X FF' COF) 196.4(+/-) 61.6 Initial RTNDT, I COF) -56 67 Margin, M (°Fic) 65.5 34 ART= I+ (CF xFF) + M (°F) 205.9(+/-) 162.6(+/-)

per Regulatory Guide 1.99, Rev 2 Calc N-288 Rev 3 (a) Fluence, f, is based upon f~t and f%t 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*28 -O.lOiogf).

(c) Margin is calculated as M = 2(ol+ cr!:l2) 05 . The standard deviation for the initial RTNDT margin term cr~,

is 0°F for plate since the initial RTNDT is a measured value. The standard deviation for ~RTNDT term cr/:i, is 17°F for the plate, except that cr!:i need not exceed the 0.5 times the mean value of ~RTNDT*

(d) DCPP-1lower shell longitudinal weld 3-442 Cis limiting for the heatup and cooldown Appendix G curves at ~t.

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

%t.

(f) The higher CF based onCE NPSD-1039, Rev 2 for these limiting materials is used to generate the heatup and cooldown Appendix G curves. The ART's used to generate the heatup and cooldown curves are bounding based on 28 EFPYvalues of207.8°F for 114t and 163.4°F for 3/4t.

PTLR-1 u3r13.DOC 04B 1224.0842

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