DCL-08-030, Transmittal of Pressure and Temperature Limits Report, Revision 9
| ML081050033 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/31/2008 |
| From: | Becker J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-08-030, OL-DPR-80, OL-DPR-82 | |
| Download: ML081050033 (33) | |
Text
'4.
ElectricCompany Pacific Gas and Diablo Canyon Power Plant James R. Becker Site Vice President and Mail Code 104/5/502 Station Director P. 0. Box 56 March 31, 2008 Avila Beach, CA 93424 805.545.3462 PG&E Letter DCL-08-030 Internal: 691.3462 Fax: 805.545.4234 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 Power Plant Units 1 and 2 Pressure and Temperature Limits Report, Revision 9
Dear Commissioners and Staff:
In accordance with Diablo Canyon Power Plant Technical Specification 5.6.6.c, enclosed is the Pressure and Temperature Limits Report (PTLR) for Units 1 and 2.
It was issued as PTLR-1, Revision 9, effective March 26, 2008.
As provided under 10 CFR 50.59, the PTLR changes were nmade without prior NRC approval, utilizing the methodology approved in License Amendment 170 and 171, dated May 13, 2004, for Units 1 and 2, respectively. The PTLR continues to meet the requirements of 10 CFR 50, Appendix G, "Fracture Toughness Requirements,"
and ASME Boiler and Pressure Vessel Code,Section XI, Appendix G. This change revises the Unit 2 setpoints and their references due to the replacement of steam generators. The changes are denoted with revision bars in the right margin.
There are no new or revised regulatory commitments in this report. If there are any questions regarding this submittal, contact Mr. Steve Hamilton at (805) 545-3449.
ddm/469/A0701121 Enclosure cc: Diablo Distribution cc/encl: Elmo E. Collins, NRC Region IV Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRR Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance A-4-x7 Callaway o Comanche Peak ° Diablo Canyon
- Palo Verde
- South Texas Project 9 Wolf Creek
Enclosure PG&E Letter DCL-08-030 PRESSURE AND TEMPERATURE LIMITS REPORT DIABLO CANYON POWER PLANT, UNITS 1 AND 2 PTLR-1, REVISION 9 (31 Pages)
EFFECTIVE DATE MARCH 26, 2008
- ISSUED FOR USE BY: DATE: EXPIRES:____
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 NUCLEAR POWER GENERATION REVISION 9 DIABLO CANYON POWER PLANT PAGE 1 OF 31 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE: PTLR for Diablo Canyon AND2 03/26/08 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).2 O PE R A TIN G L IM IT S ....................................................................................................................................... 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 A D D ITION A L CON SID ERA TION S ............................................................................................................ 15 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM .......................................................... 15 REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY ............................................................ 16 SUPPLEM EN TA L D A TA TA BLES .............................................................................................................. 20 PRESSURIZED THERMAL SHOCK (PTS) SCREENING .................................................................... 21 RE F E REN C E S ................................................................................................................................................ 21 List of Figures Figure PAGE 2.1-1 Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 7 60°F/hr) Applicable to 23 EFPY (Unit I and Unit 2) (Without Margins for Instrumentation Errors) 2.1-2 Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 10 0, 25, 50, 75 and 100°F/hr) Applicable to 23 EFPY (Unit I andUnit 2) (Without Margins for Instrumentation Errors)
List of Tables Table 2.1-1 Diablo Canyon Heatup Data at 23 EFPY (Unit I and Unit 2) With Margins for 8 Instrumentation Errors 2.1-2 Diablo Canyon Cooldown Data at 23 EFPY (Unit I and Unit 2) With Margins for 11 Instrumentation Errors 2.2-1 LTOP System Setpoints 13 2.2-2 LTOP Temperature Restrictions 13 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data 17 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data 18 IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 2 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2
This PTLR for Diablo Canyon has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TSaddressed 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 23 EFPY on Unit I and Unit 2.
The RCS temperature rate-of-change limits are:
- A maximum heatup of 60'F in any 1-hour period.
o A maximum cooldown.of 100'F in any 1-hour period.
0 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.
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
'/ of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 3 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 affectedby 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 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 cooldown curves extract the values that are based on the highest magnitude combined stress at either the 1/4t or 3/4t location.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 4 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2.1.2 RCS Pressure Test Limits:
10 CFR 50, Appendix G establishes the pressure and temperature requirements for pre-ser'vice hydrostatic test (no fuel) and hydro test and leak tests performed 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 I headflange 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 hydrotests 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 do exceed 621 psig pressure is 143°F (RTNDT + 90'F). For Condition L.c, the limiting material is Unit I lower shell weld 3-442 C based an ART of 198.3°F. For this pre-service hydro test, with no fuel 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 for Condition 1.c. Thus, the pressure temperature limits foe 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 53 deg F (Unit I 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 (72 deg F), a 2.07 sq. in. opening is relied on for RCS venting. This satisfies Condition 2.a of the 0 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 173°F (RTNDT 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 I Conditions 2.c and 2.d require that the temperature be at least 40'F greater than the corresponding ASME Appendix Glimit. 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 temperature at with the core may be critical is 330'F.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 5 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 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 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 perform the thermal hydraulic analysis and verify that the LTOP setpoints and temperature restrictions .are acceptable as documented in the calculation STA-197 (Ref. 8.7) for Unit I w/Original Steam Generators (OSG's) and STA-249 (Ref. 8.11) with input from STA-197 for 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.
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.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 6 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 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.
Theadministrative 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.
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.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 7 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 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 injectioncase 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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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 8 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500 2250 2000 1750
(.)
w 1500 cc (if aLM C/) 1250 U) w 1000 U)
C~)
0* 750 11 i i ; i ii ii I I -ii i i 500
-linr it 250 0
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 0 F/hr)
Applicable to 23 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 9 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 2.1-1 Diablo Canyon Heatup Data at 23 EFPY (Unit I and Unit 2)
With Margins for Instrumentation Errors 25°F/hr 6 0 'F/hr 60°F/hr Crit. Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press.
(Ff) (psig) (OF) (psig) (-F) (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 95 424.3 100 470.9 100 424.0 105 474.1 105 424.2 110 478.0 110 424.5 115 482.8 115 425.4 120 488.2 120 426.6 125 494.4 125 428.5 130 501.1 130 431.0 135 508.5 135 434.3 140 516.5 140 438.2 145 524.8 145 442.8 150 533.0 150 448.2 155 541.7 155 453.4 160 550.8 160 460.6 165 559.0 165 468.7 170 567.8 170 475.9 175 577.1 175 485.3 180 587.1 180 496.0 185 597.8 185 506.8 190 609.2 190 518.4 195 621.5 195 531.0 200 634.6 200 544.5 205 648.6 205 559.1 210 663.6 210 574.7 215 679.6 215 591.6 220 696.8 220 609.7 225 715.2 225 629.1 230 734.9. 230 650.0 235 756.0 235 672.4 1GRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 10 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 2.1-1 Diablo Canyon Heatup Data at 23 EFPY (Unit 1 and Unit 2)
With Margins for Instrumentation Errors 250 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. Calc. N-291 I GRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 11 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500 2250 2000 1750 I I I I I P I I I I I I
(-
1500 C,) _UIN PERATI 1250 ZD co)
Cl) 1000 Uj 750 El I JA L L IACL LENKTA 1- LE ---
C/) - - - C66 d Ratel 0
500 1:7-1 7ý I I I 1 11 1 1 1 I.Up 250 -pl?, ; P M, 041 0
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 23 EFPY (Unit 1 and Unit 2) (Without Margins for Instrumentation Errors)
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 12 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 2.1-2 Diablo Canyon Cooldown Data at.23 EFPY (Unit 1 and Unit 2)
With Margins for Instrumentation Errors 0
Steady State 25 F/hr 50°F/hr 75*F/hr 100 0 F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.
(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 350 2009.6 350 2009.6 350 2009.6 350 2009.6 350 2009.6 345 1904.1 .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 1713.5 335 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 .7958 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 IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 13 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 23 EFPY (Unit I and Unit 2)
With Margins for Instrumentation Errors Steady State 25°F/hr 5 0 °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) 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 A150 362.0 150 300.9 145 529.2 145 472.3 145 414.0 145 354.0 145 292.4 140 522.9 140 465.7 140 407.0 140 346.6 140 284.5 135 517.0 135 459.6 135 400.6 135 339.8 135 277.2 130 511.5 130 453.8 130 394.5 130 333.4 130 270.4 125 506.4 125 448.5 125 388.9 125 327.5 125 264.2 120 501.5 120 443.5 120 383.7 120 322.1 120 258.4 115 497.0 115 438.9 115 378.9 115 317.0 115 253.1 110 492.8 110 434.5 110 374.4 110 312.4 110 248.2 105 488.9 105 430.5 105 370.3 105 308.1 105 .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 85 417.1 85 356.6 85 294.1 85 229.2 80 472.7 80 414.1 80 353.7 80 291.1 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 Calc. N-291 IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 14 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 2.2-1 Low Temperature Over-Pressure (LTOP)
System Setpoints Function Setpoint PORV Arming TemperatureM' > 280 OF PORV Pressure Setpoint(2) 435 psig (1) Calc. N-298, Rev. 0. Valid to 23 EFPY (2) STA-197, Rev. 0 Table 2.2-2 Low Temperature Over-Pressure (LTOP)
Temperature Restrictions Restriction Setpoint OSGs(1) RSGs(1,2)
SI Pumps Secured, CCP 1 or CCP 2 Secured, SI Accumulators Isolated, < 280 OF < 280 OF CCP 3 aligned for LTOP operation Safety Injection Flowpath Blocked, and SI Blocked _ 166 OF _ 169 OF 2 of 3 Charging Pumps Secured _ 152 OF _156 OF I of 4 RCPs Secured *145 OF _148 OF 2 of 4 RCPs Secured *128 OF 132 OF 3 of 4 RCPs Secured _114 OF _117 OF 4 of 4 RCPs Secured _104 OF 108 OF RCS Vent Path of 2.07 in2 Established *84 OF 90 OF (1) Calc. STA-197, Rev. 0 (2) Calc. STA-249, Rev. 0 Assumptions: 1) PORV Stroke Time of 2.9 seconds.
- 2) Apply 10 % per Code Case N-514.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 15 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2
- 3. ADDITIONAL CONSIDERATIONS Revisions tothe 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 5.2.4A4 of the Final Safety Analysis Report (FSAR). The withdrawal schedule is presented in FSAR Table 5.2-22.
Diablo Canyon Units I & 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-1 1567, 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-1 5423, Analysis of Capsule V from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, September 2000.
I GRSATO9.DOC 04B 3110 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 16 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 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 I - cycles I through 6, January, 1995.
4.12 WCAP-15780, Fast Neutron Fluence and Neutron Dosimetry Evaluations for the Diablo Canyon Unit I 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-1 5782, Fast Neutron Fluence and Neutron Dosimetry Evaluations for the Diablo Canyon Unit 2 Reactor Pressure Vessel, December, 2001.
- 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 1/4/4t location is found in Seam Weld 3-442 C in the Unit 1 reactor vessel while the most limiting 3/4tlocation 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).
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 17 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 ARTNDT 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 ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 should be less than 1 cy (standard deviation) of 17'F for base metal and 28°F for weld material.
The Diablo Canyon Unit I Surveillance Capsule S for the Intermediate Shell Plate B4106-3 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 I 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.
Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.
The capsule specimens are located in the reactor between the thermal shield (Unit 1) or neutron pads (Unit 2) and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield (Unit 1) or neutron pads (Unit 2). The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25'F. Hence this criteria is met.
Criterion 5: The surveillance datafor the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.
The surveillance data for the correlation monitor material in the capsules fall within the scatter band for this (Correlation Monitor Material Heavy Section Steel Technology Plate 02) material.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 18 OF 31 J
TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data Best Fit Measured Scatter in Material Capsule CF(a) FF ARTNDT(b) ARTNDT(c) ARTNDT Inter Shell Plate S(d) 0.656 21.52 -1.78 -23.3 B4106-3 Inter Shell Plate Y 32.8 1.014 33.26 48.66 15.4 B4106-3 Inter Shell Plate V 1.087 35.65 34.32 -1.33 B4106-3 Surveillance Weld S(d) 0.656 131.00 110.79 -20.21 Heat 27204 Surveillance Weld Y 199.7 1.014 202.50 232.59 30.09 Heat 27204 I Surveillance Weld V 1.087 217.07 201.07 -16.0 Heat 27204 WCAP 15958 (a) CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see Table 6.0-3).
(b) Best fit ARTNDT = CF* FF.
(c) Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, r Revision 2.0.
(d) Diablo Canyon Surveillance Capsule S is currently notjudged Credible per Reg. Guide 1.99, Rev 2, Position 2.1.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 19 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND2 Table 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data Best Fit Measured Scatter in Material Capsule CF(a) FF ARTNDT(b) ARTNDT(C) ARTNDT Inter Shell Plate U 0.701 69.1 65.4 -3.7 B5454-1 (Long)
Inter Shell Plate X 98.6 0.976 96.2 100.1 3.9 B5454-1 (Long)
Inter Shell Plate Y 1.121 110.5 111.6 1.1 B5454-1 (Long)
Inter Shell Plate V 1.237 122.0 123.4 1.4 B5454-1 (Long)
Inter Shell Plate U 0.701 69.1 73.3 4.2 B5454-1 (Trans)
Inter Shell Plate X 98.6 0.976 96.2 99.5 3.3 B5454-1 (Trans)
Inter Shell Plate Y 1.121 110.5 111.6 1.1 B5454-1 (Trans)
Inter Shell Plate V 1.237 122.0 112.9 -9.1 B5454-1 (Trans)
Surveillance Weld U 0.701, 138.2 173.0 34.8 Surveillance Weld X 197.2 0.976 192.5 203.2 10.7 Surveillance Weld Y 1.121 221.1 211.4 -9.7 Surveillance Weld V 1.237 243.9 224.5 -19.4 WCAP-15423 (a) CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see Table 6.0-3).
(b) Best fit ARTNDT = CF
- FF.
(C) Calculated using measured Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 20 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2
- 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 6.0-3 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, '1/4tand 3/4tLocations at 23 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, 1/4/t and /t Locations at 23 EFPY Table 6.0-8 Diablo Canyon Unit I Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the '1/4tand %t Locations for 23 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4At and 3/4tLocations for 23 EFPY Table 6.0-10 Calculation of Adjusted Reference Temperature at 23 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 21 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2
- 7. PRESSURIZED THERMAL SHOCK (PTS) SCREENING 10 CFR 50.61 requires that RT p*s 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 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 Thermal Shock for Diablo Canyon Unit 1, July, 1993.
7.2 WCAP-14364, Evaluation of Pressurized Thermal Shock for the Diablo 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 (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 andRCS 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-I 850-CCM-A, Project 889-3, December, 1996 8.7 PG&E Calculation STA-197 Rev. 0, "LTOP Temperature Limits for 23 EFPY" 8.8 PG&E Calculation N-288, Rev. 0, "Adjusted RT-NDT Versus EFPY" 8.9 PG&E Calculation N-291, Rev. 1,,"Pressure-Temperature Limits for Heatup &'
Cooldown" 8.10 PG&E Calculation N-298, Rev. 0. "LTOP Enable Temperature for 23 EFPY" 8.11 PG&E Calculation STA-249 Rev. 0, "RSG - LTOP Analysis" IGRSAT09.DOC 04B 03118 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 22 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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/cm 2) Temperature Shift Decrease Predicted Measured Predicted Measured (OF) (a) (OF) (b) (%) (a) (%) (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 fluencevalues given here are the calculated fluence values, not'the best estimate.
1GRSAT09.DOC. 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 23 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 Fluence (C) 30 ft-lb Transition Upper Shelf Energy Materials Capsule (X 1019 n/cm 2) Temperature Shift Decrease Predicted Measured Predicted Measured (OF) (a) (OF) (b) (%) (a) (%) (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 1,13.6 111.6 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 fluence values presented here are calculated fluence values, not the best estimate.
IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 24 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 6.0-3 Calculation of Chemistry Factors Using Surveillance Capsule Data Unit 1 - Material Capsule F~a) FF(b) Measured Unit__I_- Material Capsule _ F(ARTNDT(d) OF FFxARTNDT°F FF2 Intermediate 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 2
CF Plate = Z(FF* ARTNDT) + ý(FF ) = (86.65'F) + (2.64) = 32.8-F (c)
Weld Metal S (0) 0.284 0.656 110.79 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 2
CF weld = Y(FF* ARTNDT) - E(FF ) = (527.09) + (2.64) = 199.7-F (c)
Unit 2 - Material Capsule F(a) FF(b) Measured ARTNDT~d) "F FFxARTNDT°F FF2 Intermediate 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 Intermediate 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 = Y(FF* ARTNDT) + E(FF 2) (834.53°F) + (8.462) = 98.61F 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 1.121 211.39 236.97 1.257 V 2.41 1.237 224.47 277.67 1.530 SUM 834.26 4.231 CF Weld = (FF* ARTNDT) + E(FF 2 ) (834.26'F) + (4.23 1) = 197.2'F WCAP-15958 (Unit 1) WCAP-15423 (Unit 2)
(a) F = Calculated Fluence (1019 n/cm 2, E > 1.0 MeV).
(b) FF = Fluence Factor = F(° 28 - 0 1
- logF)
(c) Unit I Capsule S is 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.
IGRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 25 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-4 DCPP-1 Reactor Vessel Beltline Material, Chemistry, and Unirradiated Toughness Data Material Description Cu (%) Ni(%) Initial RTNDT (OF)
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 -4 4 2 (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.
1GRSAT09.DOC 04B 0311.1508
PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 26 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-5 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu (%) Ni(%) Initial RTNDT (OF)
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 - 2 0 1(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 expectedto exceed 1.OE + 17.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 ý PAGE 27 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 %t Locations at 23 EFPY Material Fluence f, Fluence f,/bm Fluence f/t Fluence f,/.t Upper Shell Plate~a)
B4105-1 2.01 E+ 17 1.96E+ 17 1.14 E+ 17 4.04E+ 16 B4105-2 2.01 E+ 17 1.96E+ 17 1.14 E + 17 4.04 E + 16 B4105-3 2.01 E+ 17 1.96 E + 17 1.14 E+ 17 4.04 E +'16 Inter Shell Plate B4106-1 .9.71 E + 18 9.44 E + 18 -5.49 E + 18 1.95 E + 18 B4106-2 9.71 E+ 18 9.44 E+18 5.49E+ 18 1.95 E+ i8 B4106-3 9.71 E+ 18 / 9.44 E+ 18 5.49 E+ 18 1.95 E+18 Lower Shell Plate B4107-1 9.71 E+ 18 9.44 E+18 5.49 E+ 18 1.95 E+ 18 B4107-2 9.71 E + 18 9.44 E + 18 5.49 E + 18 1.95 E + 18 B4107-3 9.71 E+ 18 9.44 E+ 18 5.49 E+ 18 1.95 E+ 18 Upper Shell Long(a)
Welds 1-442 A,BC 2.01 E + 17 1.96 E + 17 1.14 E + 17 4.04 E + 16 Upper Shell to Inter Shell Weld 8-442(a) 2..0 E +17 1.96 E + 17 1.14 E +17 4.04 E + 16 Inter Shell 'Long Welds 2-442 A,B 7.32E+ 18 7.12 E + 18 4.14E+ 18 1.47E+ 18 Weld 2-442 C 3.66 E + 18 3.56 E + 18 2.07 E + 18 7.36 E + 17 Inter Shell to Lower Shell Weld9-442 9.71 E+ 18 9.44E+ 18 5.49E+ 18 1.95E+ 18 Lower Shell Long Welds 3-442 A,B 6.01 E + 18 5.84 E + 18 3.40 E + 18 1.21E + 18 Weld 3-442 C 9.7.1 E + 18 9.44 E + 18 5.49 E + 18 1.95 E.+ 18 CalcN-288, WCAP-15958 (a) Upper shell materialsare included for completeness since EOL exposure is expected to exceed I.01E + 17.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 28 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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, 1/4tand %tLocations at 23 EFPY Material Fluence f, Fluence fc/bm Fluence fy. Fluence f&t Upper Shell Plate(a)
B5453-1 2.25 E + 17 2.19 E + 17 1.27 E + 17 4.52 E +- 16 B5453-3 2.25 E+ 17 2.19 E+ 17 1.27 E+ 17 4.52 E + 16 B5011-IR 2.25 E + 17 2.19 E+ 17 1.27 E+ 17 4.52 E + 16 Inter Shell Plate B5454-1 1.09 E + 19 1.06 E+ 19 6.15 E+ 18 2.19E+18 B5454-2 1.09 E + 19 1.06 E+ 19 6.15 E+ 18 2.19 E+ 18 B5454-3 1.09 E + 19 1.06 E + 19 6.15 E + 18 2.19 E + 18 Lower Shell Plate B5455-1 1.09 E +,19 1.06 E+ 19 6.15 E+ 18 2.19 E+ 18 B5455-2 1.09 E + 19 1.06 E+ 19 6.15 E+ 18 2.19 E+ 18 B5455-3 1.09E+ 19 1.06E+ 19 6.15E+ 18 2.19E+ 18 Upper Shell Long(a)
Welds 1-201 A,B,C 2.25 E+ 17 2.19 E+ 17 1.27 E+ 17 4.52 E+ 16 Upper Shell to Inter Shell Weld 8-201(a) 2.25 E + 17 2.19 E + 17 1.27 E + 17 4.52 E + 16 Inter Shell Long Weld 2-201 A 5.98 E+ 18 5.81 E+,18 3.38 E+ 18 1.20 E+ 18 Weld 2-201 B '5.62 E+ 18 5.47 E+ 18 3.18 E+ 18 1.13 E+ 18 Weld 2-201 C 4.72 E + 18 4.59 E+ 18 2.67 E+ 18 9.48 E+ 17 Inter Shell to Lower Shell Weld 9-201 1.09 E + 19 1.06 E + 19 6.15 E + 18 2.19 E + 18 Lower Shell Long Weld 3-201 A 4.72E + 18 4.59 E + 18 2.67 E + 18 9.48 E + 17 Weld 3-201 B 5.98 E + 18 5.81 E+ 18 3.38 E+ 18 1.20 E + 18 Weld 3-201 C 5.62 E+ 18 5.47 E+ 18 3.18 E+ 18 '1.13 E+ 18 Calc N-288, WCAP-15423 Rev. 0.
(a) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.OE + 17.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 29 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4tand %tLocations for 23 EFPY 23 EFPY ART(a)
Material RG 1.99 Rev. 2 t(-F) 3t (-F)
Method Upper Shell Plate(d)
B4105- 1 Position 1.1 73.2 67.0 B4105-2 Position 1.1 54.2 48.0 B4105-3 Position 1.1 61.7 54.1 Inter Shell Plate B4106-1 Position 1.1 95.0 72.1 B4106-2 Position 1.1 98.4 76.6 B4106-3 Position 1.1 124.0 107.2 Lower Shell Plate B4107-1 Position 1.1 123.7 99.6 B4107-2 Position 1.1 122.4 100.3 B4107-3 Position 1.1 79.7 57.9 Upper Shell Long(d)
Welds 1-442 A,B,C Position 1.1 31.6 4.6 Upper Shell to Inter(d)
Shell Weld 8-442 Position 1.1 9.0 -8.9 Inter Shell Long Welds 2-442 A,B Position 1.1 180.8 122.5 Weld 2-442 C Position 1.1 140.6 90.8 Inter Shell to Lower Shell Weld 9-442 Position 1.1 152.8 106.5 Lower Shell Long Welds 3-442 A,B Position 1.1 168.9 112.9.
Weld 3-442 C(c) Position 1.1 198.3(b) 137.3 Calc N-288 & WCAP-15958 (a) ART = Initial RTNDT + ARTNDT + Margin (OF)
(b) This limiting ART value is bounded by that used to generate heatup and cooldown curves (198.3°F).
(c) DCPP-1 Surveillance Capsule S was not judged "credible" per 10 CFR 50.61. The higher chemistry values of CE NPSD-1039, Rev 2 for this heat are used to generate the heatup and cooldown Appendix G curves.
(d) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.OE + 17.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 30 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4tand 3/4tLocations for 23 EFPY 23 EFPY ART(a)
Material RG 1.99 Rev. 2 1/4t(-F) 3/4t(-F)
Method Upper Shell Plate(c)
B5453-1 Position 1.1 47.1 37.2 B5453-3 Position 1.1 49.9 43.9 B5011-1R Position 1.1 45.0 39.0 Inter Shell Plate B5454-1 Position 2.1 154.2 127.3 B5454-2 Position 1)1 187.1 159.8(b)
B545453 Position 1.1 162.5 132.3 Lower Shell Plate B5455-1 Position 1.1 103.8 77.0 B5455-2 Position 1.1 118.8 92.0 B5455-3 Position 1.1 105.3 87.5 Upper Shell Long(c)
Welds 1-201 A,B,C Position 1.1 4.5 -23.7 Upper Shell to Inter(')
Shell Weld 8-201 Position 1.1 5.8 -10.0 Inter Shell Long Weld 2-201 A Position 1.1 154.2 102.0 Weld 2-201 B Position 1.1 150.8 99.3 Weld 2-201 C Position 1.1 141.3 91.8 Inter Shell to Lower Shell Weld 9-201 Position 1.1 8.4 -2.6 Lower Shell Long Weld 3-201 A Position 1.1 92.3 59.0 Weld 3-201 B Position 1.1 100.2 68.3 Weld 3-201 C Position 1.1 98.1 66.6 Calc N-288 & WCAP-15423 (a) ART = Initial RTNDT +-ARTNDT + Margin (fF)
(b) This limiting ART value is bounded by that used to generate heatup and cooldown curves (159.8 0 F).
(c) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E + 17.
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 9 PAGE 31 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-10 Calculation of Adjusted Reference Temperature at 23 EFPY (Unit 1 and Unit 2) for the Limiting Diablo Canyon Reactor Vessel Materials Parameter ART Value Location 34t(e)
/t(d)
Chemistry Factor, CF (7F) 226.8(0 99.6 Fluence + 1019 n/cm 2 (E > 1.0 MeV), f(a) 0.549 0.219 Fluence Factor, FF(b) 0.8323 0.5908 ARTNDT = CF x FF, (0 F) 188.8(0 58.8 Initial RTNDT, I (-F) -56 67 Margin, M (OF)(C) 65.5 34 ART = I + (CF x FF) + M (7F) 198.3(0 159.8(0 per Regulatory Guide 1.99, Rev. 2 Calc N-288 (a) Fluence, f, is based upon f/4, and f/,tfrom 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(O. 28 -0.1OogO.
(c) Margin is calculated as M = 2((712+ CA2 ) 0 5 . The standard deviation for the initial RTNDT margin term a1, is 0°F for plate since the initial RTNDT is a measured value. The standard deviation for ARTNDT term CYA, is 17°F for the plate, except that GA need not exceed the 0.5 times the mean value of ARTNDT.
(d) DCPP'- lower shell longitudinal weld 3-442 C is limiting for the heatup and cooldown Appendix G curves at 1/t.
( DCPP-2 intermediate shell plate B5454-2 is limiting for the heatup and cooldown Appendix G curves at 3 t.
(0 DCPP-1 Surveillance Capsule S was 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 calculated ART's are used to generate the heatup and cooldown curves (198.3'F for 1/4t and 159.8°F for 3/4t).
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