DCL-03-177, Response to NRC RAI Regarding License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report.

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Response to NRC RAI Regarding License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report.
ML040070364
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 12/23/2003
From: Oatley D
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-03-177
Download: ML040070364 (41)
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Text

Pacific Gasand ElectricCompany David H.Oatley Diablo Canyon Power Plant Vice President and P.O.Box 56 General Manager Avila Beach, CA 93424 December 23, 2003 Fax:805.545A234 PG&E Letter DCL-03-177 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 Response to NRC Request for Additional Information Reqarding "License Amendment Request 02-04. Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report"

Dear Commissioners and Staff:

In Pacific Gas & Electric Company's (PG&E) Letter DCL-02-079, "License Amendment Request (LAR) 02-04, Revision to Technical Specification (TS) 5.6.6 -

Reactor Coolant System Pressure and Temperature Limits Report (PTLR)," dated July 31, 2002, PG&E requested NRC approval of the methodology to be used to make changes to the Diablo Canyon Power Plant PTLR without prior NRC approval.

In PG&E Letter DCL-03-101 dated August 15, 2003, PG&E submitted its response to a request for additional information (RAI) regarding LAR 02-04.

On August 29, 2003, and September 3, 2003, the NRC staff requested additional information needed to complete their review of LAR 02-04. PG&E's response to this RAI is provided in Enclosures 1 and 2.

This additional information does not affect the results of the technical evaluation and no significant hazards consideration determination previously transmitted in PG&E Letter DCL-02-079.

If you have any questions or require additional information, please contact Stan Ketelsen at (805) 545-4720.

Sincerely David H. Oatley A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde . South Texas Project
  • NVolfCreek

Document Control Desk PG&E Letter DCL-03-177 December 23, 2003 Page 2 jer/3664 Enclosures cc: Edgar Bailey, DHS Bruce S. Mallett David L. Proulx Diablo Distribution cc/enc: Girija S. Shukla A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • ComanchePeak
  • DiabloCanyon
  • PaloVerde
  • SouthTexasProject
  • WolfCreek

PG&E Letter DCL-03-177 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Docket No. 50-275 In the Matter of ) Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY) No. DPR-80 Diablo Canyon Power Plant ) Docket No. 50-323 Units1 and2 ) Facility Operating License No. DPR-82 AFFIDAVIT David H. Oatley, of lawful age, first being duly sworn upon oath states that he is Vice President and General Manager - Diablo Canyon of Pacific Gas and Electric Company; that he has executed this response to the NRC request for additional information on License Amendment Request 02-04 on behalf of said company with full power and authority to do so; that he is familiar with the content thereof; and that the facts stated therein are true and correct to the best of his knowledge, information, and belief.

David H. Oatley Vice President and General Manager - Diablo Canyon Subscribed and sworn to before me this 23rd day of December, 2003.

otary Public County of San Luis Obispo SANDRA L. RECTOR Commission # 1339380 State of California § -. Notary Public - Californiaz San Luis Obispo County PMyComm. Egires Jan 12, 2006

Enclosure 1 PG&E Letter DCL-03-177 Response to Request for Additional Information Concerning License Amendment Request 02-04 "Revision of Technical Specification 6.6.6 - Reactor Coolant System Pressure and Temperature Limits Report" On August 29, 2003, and September 3, 2003, the NRC staff provided the following requests for additional information (RAI) regarding Pacific Gas & Electric Company's (PG&E) Letter DCL-02-079, "License Amendment Request 02-04, Revision to Technical Specification (TS) 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report (PTLR)," dated July 31, 2002.

NRC Request No. I Table 1 of 10 CFR Part 50, Appendix G, specifies six different minimum temperature requirements that must be met when generating the pressure-temperature (P/T) limits for U.S. operating pressurized water reactors (PWRs):

A. Those forpressure test conditions with the reactor coolant system (RCS) pressure less than or equal to 20% of the reactor's preservice hydrostatic test pressure.

B. Those for pressure test conditions with the RCS pressure greater than 20% of the reactor's preservice hydrostatic test pressure.

C. Those for normal operating conditions (including heatups and cooldowns of the reactor and transient operating conditions) with the RCS pressure less than or equal to 20% of the reactor's preservice hydrostatic test pressure, at times the reactor is not in the critical operating mode.

D. Those for normal operating conditions (including heatups and cooldowns of the reactor and transient operating conditions) with the RCS pressure greater than 20% of the reactor's preservice hydrostatic test pressure at times the reactor is not in the critical operating mode.

E. Those for normal operating conditions (including heatups and cooldowns of the reactor and transient operating conditions) with the RCS pressure less than or equal to 20% of the reactor's preservice hydrostatic test pressure at times the reactor is in the critical operating mode.

F. Those for normal operating conditions (including heatups and cooldowns of the reactor and transient operating conditions) with the RCS pressure greater than 20% of the reactor's preservice hydrostatic test pressure at times the reactor is not in the critical operating mode.

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Enclosure 1 PG&E Letter DCL-03-177 Section 2.1.2 of DCPP Report PTLR-1, Revision 2, provides a discussion of the minimum temperature requirements that have been incorporated into the PIT limits for the Diablo Canyon Power Plant (DCPP); however, the discussion only addresses the minimum temperature requirements for normal operating conditions or pressure test conditions when the operating pressure for the RCS is less than or equal to 20% of the preservice hydrostatic test pressure for the RCS and when the reactor is not in the critical operating Mode (i.e., those for items A. and C. above). To ensure that future 10 CFR 50.59 changes of the DCPP P/T limit figures will continue to comply with the minimum temperature requirements of Table 1 of 10 CFR Part 50, Appendix G, and that the PTLR will continue to conform to the provisions of PTLR Criterion 6 in Attachment 1 to Generic Letter (GL) 96-03, update Section 2.1.2 of the PTLR, the next time it is revised, to provide a discussion on how the PIT limit curves will continue to meet all of minimum temperature requirements mandated by Table I of 10 CFR Part 50, Appendix G.

PG&E Response PG&E has revised PTLR-1, Revision 2, to include new Section 2.1.2, "RCS Pressure Test Limits," that explains how the PTLR meets 10 CFR Part 50, Appendix G, Table 1, Conditions 1.a, 1.b, and 1.c. Previous Section 2.1.2 has been renumbered and retitled Section 2.1.3, "Reactor Vessel Bolt-up and Criticality Temperature Limits." This section has been expanded to explain how the PTLR meets the remaining conditions of 10 CFR Part 50, Appendix G, Table 1, (i.e., Conditions 2.a, 2.b, 2c, and 2d). The revised PTLR-1, Revision 2, is included as Enclosure 2 of this submittal.

NRC Request No. 2 Westinghouse Topical Report Nos. WCAP-15958, Revision 0, and WCAP-15423, Revision 0, represent the NRC docketed surveillance capsule reports for each of the V surveillance capsules in the reactor vessels (RVs) for DCPP Units 1 and 2. The reports provided the neutron fluence dosimetry data, RV tensile test and RV Charpy impact data obtained from dosimetry measurements, tensile tests, and Charpy-V impact tests performed on test specimens removed from the DCPP Unit I and 2 "V" capsules. The reports also reanalyzed all previous fluence dosimetry data and RV surveillance capsule data reported and submitted onto the dockets for DCPP Units I and 2 (i.e., in the WCAP Reports for DCPP Unit 1 Capsules S and Y and in the WCAP Reports for DCPP Unit 2 Capsules U,X, and Y%. DCPP Report No. PTLR-1, Revision 2, does not include the impact of the latest dosimetry and charpy-impact data in WCAP-15958, Revision 0, and WCAP-15423, Revision 0, on the assessments performed in PTLR Tables 5.0-1 and 5.0-2, and PTLR Tables 6.0-1 through 6.0-10. To ensure conformance with the PTLR criteria 2 and 7 of to GL 96-03, make the following specific updates to the PTLR:

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Enclosure 1 PG&E Letter DCL-03-177 A. Update Section 4 of the PTLR report to include references to both WCAP-15958, Revision 0, and WCAP-15423, Revision 0.

B. Update the data in PTLR Table 5.0-1 and update the assessments in PTLR Tables 6.0-1 and 6.0-3 to incorporate the latest surveillance capsule data for DCPP Unit I (i.e., for capsules S, Y, and V), as analyzed in WCAP-15958, Revision 0, and update Table 6.0-6 to reflect the latest neutron fluences for the DCPP Unit I RV, as reported in WCAP-15958, Revision 0. Assess whether any of the data in Tables 6.0-8 and 6.0-10 need to be amended based on the impacts of the updates to the data in PTLR Tables 6.0-1, 6.0-3, and 6.0-6.

C. The data in PTLR Table 5.0-2 and the assessments in PTLR Tables 6.0-2 and 6.0-3 need to be updated to incorporate the latest surveillance capsule data for DCPP Unit 2 (i.e., capsules U, X, Y, and V), as analyzed in WCAP-15423, Revision 0, and update Table 6.0-7 to reflect the latest neutron fluences for the DCPP Unit 2 RV, as reported in WCAP-15423, Revision 0. Assess whether any of the data in Tables 6.0-9 and 6.0-10 need to be amended based on the impacts of the updates to the data in PTLR Tables 6.0-2, 6.0-3, and 6.0-7.

[By Fax] The staff is enclosing Reactor Vessel Integrity Database (RVID) Summary Sheets for all DCPP Unit I and 2 surveillance data that have been reported in compliance with the requirements of 10 CFR Part 50, Appendix H (i.e., surveillance capsules S, Y, and V for DCPP Unit I and surveillance capsules U, X, Y, and V for DCPP Unit 2). The RVID summary sheets are based on the updated dosimetry and Charpy impact data for the capsules, as reported in WCAP-15958, Revision 0, for DCPP Unit 1 and WCAP-15423, Revision 0, for DCPP Unit 2. These summary sheets may be used to assist you in making the changes to the tables identified in parts B. and C. of this RAL.

PG&E Response A. PG&E has revised PTLR-1, Revision 2, Section 4 to reference WCAP-1 5958 dated January 2003 (Revision 0), and WCAP-1 5423 dated September 2000 (Revision 0).

B. PG&E has updated the data in PTLR Table 5.0-1 and the assessments in PTLR Tables 6.0-1 and 6.0-3 to incorporate the latest surveillance capsule data for DCPP Unit 1 (i.e., for capsules S, Y, and V), as analyzed in WCAP-1 5958 dated January 2003 (Revision 0). PG&E has also updated Tables 6.0-6, 6.0-8 and 6.0-10 to reflect the latest neutron fluences for the DCPP Unit 1 RV, as reported in WCAP-1 5958.

C. PG&E has updated the data in PTLR Table 5.0-2 and the assessments in PTLR Tables 6.0-2 and 6.0-3 to incorporate the latest surveillance capsule data for 3

Enclosure 1 PG&E Letter DCL-03-177 DCPP Unit 2 (i.e., for capsules U, X, Y, and V), as analyzed in WCAP-1 5423 dated September 2000 (Revision 0). PG&E has also updated Tables 6.0-7, 6.0-9 and 6.0-10 to reflect the latest neutron fluences for the DCPP Unit 2 RV, as reported in WCAP-15423.

NRC Request No. 3 The RETRAN Code is used in the thermal hydraulic transient analysis for low temperature overpressureprotection (LTOP). Please list RETRAN in the LTOP section of the PTLR.

PG&E Response PG&E has revised PTLR-1, Revision 2, to include a discussion of the RETRAN Code in the LTOP section (Section 2.2.1) and has added it as Reference 8.6 to Section 8, "References."

NRC Request No. 4 The fluence calculation references in the pressurized thermal shock (PTS) section of the PTLR (i.e., the referenced reports in Section 7 of the PTLR) have not been updated as was done in the PIT limit curves section. Please update the fluence references in the PTS section of the PTLR.

PG&E Response PG&E has revised PTLR-1, Revision 2, to include the following fluence calculations as references in Section 7:

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

NRC Request No. 5 Calculation STA-138 states that a conservative calculation was performed for the heat injection transient. However, the licensee's submittal did not provide justification that the transient is bounding. Please demonstrate that the heat injection transient is bounding or report the results of the defining transient with a solid RCS and T = 500F.

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Enclosure 1 PG&E Letter DCL-03-177 PG&E Response As discussed in the following paragraphs, the maximum heat injection case was determined to be the case with the reactor coolant system (RCS) liquid at 285 0F, which bounds the LTOP enable temperature of 2700F. Calculation STA-138, "RETRAN Evaluation of LTOP Parameters," submitted as Enclosure 5 to LAR-02-04, documents the analysis results of nine heat injection cases. The results are shown in STA-138, Table 6-14, UDCPP LTOP Heat Input Peak Pressure Results," which represent a bounding range of conditions for DCPP. In addition to the conservative LTOP setpoints discussed in STA-128, Section 3, "Assumptions,"

the assumptions and initial conditions which establish these heat injection cases as bounding for DCPP are summarized below.

Maximum RCS/Steam Generator Temperature Difference DCPP Limiting Condition for Operation (LCO) 3.4.6 Note 2 states:

"No RCP [reactor coolant pump] shall be started with any RCS cold leg temperature < Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR unless the pressurizer water level is less than 50%, OR the secondary side water temperature of each steam generator (SG) is < 50'F above each of the RCS cold leg temperatures."

This LCO is verified by comparing the RCS wide range resistance temperature detector (RTD) cold leg temperature indication with a contact temperature probe measurement of the SGs. The LTOP heat injection cases assume that the secondary liquid in the SGs, the SG tube metal mass, and the RCS liquid within the SG tube primary side volume are all eighty degrees warmer than the steady state RCS liquid temperature. This assumption conservatively bounds the maximum RCS/SG temperature difference of 500 F allowed per the Technical Specifications and the maximum measurement uncertainties associated with the RCS wide range RTDs and SG contact temperature probe.

Water Solid RCS The heat injection cases assume that the RCS is in a water solid condition (the pressurizer is completely filled with liquid) when the RCP is started and the transient begins.

Maximum SG to RCS Heat Transfer The heat injection cases conservatively bound the maximum potential heat transfer by assuming 0 percent tube plugging in all four SGs, and artificially increasing the nominal heat transfer capability of the SGs. The nominal volumetric heat capacity of 5

Enclosure I PG&E Letter DCL-03-177 the Inconel SG tubes was increased by a factor of 1.5, while the thermal conductivity was increased by a factor of 10. The heat injection cases also assume the RCP achieves maximum RCS flow within 10 seconds of the pump start.

Bounding Spectrum of RCS/SG Temperatures The 9 heat injection cases documented in STA-138 Section 6 bound the variation in the volumetric expansion capability of the RCS liquid due to temperature for the complete LTOP range of allowable temperatures. The results listed in STA-138 Table 6-14 confirm that the most limiting heat injection case occurs at the maximum allowable temperature conditions, since the relative change in the specific volume of water increases with increasing temperature, and there is less RCS liquid mass available to absorb the secondary heat. Therefore, the maximum heat injection case was analyzed with the RCS liquid at 2850 F to bound the LTOP enable temperature of 2700 F and the associated RTD measurement uncertainty of 150F.

Additional heat injection cases were performed at subsequent intervals of RCS/SG temperature conditions to confirm the characteristic trend and dominant effect associated with the variation in the specific volume of water versus temperature.

The minimum RCS temperature evaluated for heat injection was 100F, since the results in STA-1 38 Figures 6-6, 6-7, and 6-8 demonstrate that the transient would be non-limiting for lower temperatures.

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Enclosure 2 PG&E Letter DCL-03-177 DIABLO CANYON POWER PLANT PRESSURE AND TEMPERATURE LIMITS REPORT NUMBER PTLR-1 REVISION 2

      • ISSUED FOR USE BY: DATE: EXPIRES:____

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 NUCLEAR POWER GENERATION REVISION 2 DIABLO CANYON POWER PLANT PAGE 1 OF 31 PRESSURE AND TEMPERATURE LIMITS REPORT TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 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 (PIT) Limits (LCO 3.4.3) .2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12) .5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM .15 REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY .16 SUPPLEMENTAL DATA TABLES .20 PRESSURIZED THERMAL SHOCK (PTS) SCREENING ......................................... 21 REFERENCES.21 List of Figures Figure PAGE 2.1-1 Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 7 600 F/hr) Applicable to 16 EFPY (Without Margins for Instrumentation Errors) 2.1-2 Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 10 0, 25, 50, 75 and 1000 F/hr) Applicable to 16 EFPY (Without Margins for Instrumentation Errors)

List of Tables Table 2.1-1 Diablo Canyon Heatup Data at 16 EFPY Without Margins for Instrumentation Errors 8 2.1-2 Diablo Canyon Cooldown Data at 16 EFPY Without Margins for Instrumentation 11 Errors 2.2-1 LTOP System Setpoints 13 2.2-2 LTOP Temperature Restrictions 13 5.0-1 Diablo Canyon Unit I Surveillance Capsule Data 17 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data 18 This procedure was rewritten; therefore, revision bars are not included.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 2 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2

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:

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

The RCS temperature rate-of-change limits are:

  • A maximum heatup of 60 0F in any 1-hour period.
  • A maximum cooldown of 1000 F in any 1-hour period.
  • A maximum temperature change of less than or equal to 100F in any I-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 Y/4 of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 3 OF 31 TITLE: PTLR for Diablo Canyon UNITS I 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 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 70deg to 550deg 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 II! 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 2 PAGE 4 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 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 hydrotest 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 head flange that has an RTNDT of 531F. 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 531F.

For Condition l .b, the minimum RCS temperature for the hydrotests and leak tests with fuel in the vessel and core not critical that do exceed 621 psig pressure is 1430 F (RTNDT+ 901F). For Condition l .c, the limiting material is Unit I lower shell weld 3442 C based an ART of 183.71F. For this pre-service hydrotest, with no fuel in the vessel, the minimum RCS temperature for all pressures is 243.7 0F (RTNDT+ 90'F). The limiting temperature for all these conditions is for Condition I.c. Thus, the pressure temperature limits for leak testing are imposed starting with a minimum temperature of 2450 F.

2.1.3 Reactor Vessel Bolt-up and Criticality Temperature Limits:

Operating restrictions illustrated on the P-T curve also include reactor flange boltup 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 RTNwT shift, and, thus minimum Bolt Up Temperature does not change with time. The highest flange RTNDT between DCPP Unit I and 2 is 53deg 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 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 1730 F (RTNDT of 531F + 900 F) at pressures not exceeding the 20% hydrotest pressure or 621 psig. This portion of the Figures 2.1 -1 and 2.1-2 curves are the notches about the 1730 F temperature.

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 G limit. The minimum temperature for criticality is a minimum temperature for the In-service system hydrostatic pressure temperature, which is 2459 psig. The corresponding temperature for a hydrostatic test at 2459 psig is 313.60 F. Thus, the minimum temperature at with the core may be critical is 314°F.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 5 OF 31 TITLE: PTLR for Diablo Canyon UNITS I 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 PIT curve and limiting the minimum RCS undershoot to maintain a nominal operating pressure drop across the number one RCP seal.

The arming temperature setpoint is 200'F or RTNDT + 50'F which ever is greater in accordance with ASME Code Case N-514. The RETRAN-02 Mod3 computer code (Ref. 8.6) was used to perform the thermal hydraulic analysis and verify that the LTOP setpoints and temperature restrictions are acceptable as documented in the calculation STA-138 (Ref. 8.7).

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 2 PAGE 6 OF 31 TITLE: PTLR for Diablo Canyon UNITS I 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 centrifugal charging pump (CCP), and isolate all SI Accumulators 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 CCP injecting through the SI injection flowpath and the positive displacement pump (PDP) injecting through the normal and the alternate charging flowpaths simultaneously. The administrative temperature limit for operating with a maximum of one charging pump is based on a mass injection case with one CCP injecting through the normal and the alternate charging flowpaths. The administrative temperature limits for starting and stopping RCPs are based on limiting the dynamic pressure drop increase on the RCS overshoot for a mass injection case with one CCP injecting through the normal and alternate charging flowpaths. The administrative temperature limit for establishing an RCS vent is based on determining the temperature at which the reduced Appendix G P/T limit no longer has additional margin to accommodate the mass injection RCS overshoot associated with the PORV response time. All mass injection cases account for a conservative RCP seal injection flow into the RCS and the dynamic effects of both RHR pumps running.

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 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 RCS/SG temperature restrictions for starting an RCP, since even the maximum credible RCS/SG temperature differential will not challenge the Appendix G PIT limit in the LTOP range.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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 injection case overshoot analysis also incorporates the measurement uncertainty associated with establishing the SG secondary temperature prior to starting an RCP. The RCS and SG measurement uncertainties are then assumed to be in the worst case opposite direction to establish a conservatively bounding RCS/SG temperature difference for the heat injection analysis.

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 8 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500.0 2250.0 2000.0 3-C/) 1750.0 EL w

Wl 1500.0 15)

W 1250.0 wi Cl) 1000.0 C) 750.0 500.0 I

0.0I .4-0 50 100 150 200 250 300 350 400 450 RCS TEMPERATURE ( 0F)

FIGURE 2.1 -1: Diablo Canyon Reactor Coolant System Heatup Limitations (Heatup Rates up to 60 0F/hr)

Applicable to 16 EFPY (Without Margins for Instrumentation Errors)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 9 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 TABLE 2.1-1 Diablo Canyon Heatup Data at 16 EFPY Without 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.

(°F) ~~(psig) (°F) (psig) (°F) (psig) (OF) (psig) 75 510.15 75 510.15 80 513.50 80 513.50 85 517.11 85 517.11 90 520.98 90 514.36 95 525.15 95 506.57 100 529.63 100 500.99 105 534.45 105 497.82 110 539.63 110 496.27 115 545.19 115 496.41 120 551.18 120 497.84 125 557.61 125 500.63 130 564.53 130 504.51 135 571.97 135 509.52 140 579.96 140 515.53 .

145 588.56 145 522.59 150 597.80 150 530.56 155 607.73 155 539.56 160 618.40 160 549.54 161.1 621.0 165 621.0 165 560.56 170 621.0 170 572.59 173 621.0 173 650.2 175 655.48 175 585.74 180 669.74 180 600.01 185 685.07 185 615.52 190 701.54 190 632.26 195 719.25 195 650.37 200 738.28 200 669.91 205 758.73 205 690.99 210 780.71 210 713.68 215 804.34 215 738.13 220 829.73 220 764.42 225 857.01 225 792.71 230 886.33 230 823.13 235 917.83 235 855.82 IGRSAT02.doa 04B 1223.0646

NUMBER PTLR-1 PACIFIC GAS AND ELECTRIC COMPANY REVISION 2 DIABLO CANYON POWER PLANT PAGE 10 OF 31 UNITS 1 AND 2 TITLE: PTLR for Diablo Canyon TABLE 2.1-1 Diablo Canyon Heatup Data at 16 EFPY Without Margins for Instrumentation Errors 60'F/hr 60'F/hr Crit. Limit Leak Test Limit 250 F/hr Temp. Press. Temp. Press. Temp. Press.

Temp. Press.

(°F) (psig) (OF)(psig) (° (psig)

(°F)~~(psig) 240 951.68 240 890.92 245 928.66 245 1313.55 245 988.04 250 969.17 250 1365.16 250 1027.10 255 1012.68 255 1420.55 255 1069.05 260 1059.36 260 1479.99 260 1114.11 265 1109.48 265 1543.76 265 1162.49 270 1163.26 270 1612.16 270 1214.44 275 1220.93 315 1220.93 275 1685.50 275 1266.63 280 1282.77 320 1282.77 280 1764.12 280 1321.05 285 1349.08 325 1349.08 285 1848.36 285 1379.42 290 1420.15 330 1420.15 290 1938.58 290 1442.01 295 1484.66 335 1484.66 295 2035.17 295 1509.11 300 1547.80 340 1547.80 300 2138.51 300 1581.04 305 1615.38 345 1615.38 305 2249.01 305 1658.10 310 1687.76 350 1687.76 310 2367.09 310 1740.65 315 1765.22 355 1765.22 315 2493.16 315 1829.05 320 1848.14 360 1848.14 320 2627.63 320 1923.67 325 1936.80 365 1936.80 325 2770.93 325 2024.86 330 2031.65 370 2031.65 330 2923.46 330 2133.08 335 2132.94 375 2132.94 335 3085.60 335 2248.68 2372.13 340 2241.22 380 2241.22 _

340 345 2503.81 345 2356.72 385 2356.72 350 2644.18 350 2479.95 390 2479.95 355 2793.59 355 2611.26 395 2611.26 360 2952.49 360 2751.09 400 2751.09 3121.20 365 2899.78 405 2899.78 365 370 3300.09 370 3057.72 410 3057.72 375 3489.35 375 3225.27 415 3225.27 Caic. N-NCM-97010 IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 11 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500.0 a

U)

U)

Lii a-C,,

C.

0 50 100 150 200 250 300 350 400 450 RCS TEMPERATURE (DEG F)

FIGURE 2.1-2: Diablo Canyon Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, 75 and 100 0F/hr) Applicable to 16 EFPY (Without Margins for Instrumentation Errors)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 12 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 16 EFPY Without Margins for Instrumentation Errors Steady State 25'F/hr 50'F/hr 75 0 F/hr 100 0F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF;) (Psig) (°IF) (Psig) (OF;) (psig) (OF) (psig) (OFi) (psig) 350 2787.30 350 2787.30 350 2787.30 350 2787.30 350 2787.30 345 2633.00 345 2633.00 345 2633.00 345 2633.00 345 2633.00 340 2488.11 340 2488.11 340 2488.11 340 2488.11 340 2488.11 335 2352.19 335 2352.19 335 2352.19 335 2352.19 335 2352.19 330 2224.83 330 2224.83 330 2224.83 330 2224.83 330 2224.83 325 2105.58 325 2105.58 325 2105.58 325 2105.58 325 2105.58 320 1994.03 320 1994.03 320 1994.03 320 1994.03 320 1994.03 315 1889.74 315 1889.74 315 1889.74 315 1889.74 315 1889.74 310 1792.30 310 1792.30 310 1792.30 310 1792.30 310 1792.30 305 1701.30 305 1701.30 305 1701.30 305 1701.30 305 1701.30 300 1616.37 300 1616.37 300 1616.37 300 1616.37 300 1616.37 295 1537.13 295 1537.13 295 1537.13 295 1537.13 295 1537.13 290 1463.22 290 1463.22 290 1463.22 290 1463.22 290 1463.22 285 1394.31 285 1394.31 285 1394.31 285 1394.31 285 1394.31 280 1330.07 280 1330.07 280 1330.07 280 1330.07 280 1330.07 275 1270.22 275 1270.22 275 1270.22 275 1270.22 275 1270.22 270 1214.44 270 1214.44 270 1214.44 270 1214.44 270 1214.44 265 1162.49 265 1162.20 265 1162.49 265 1162.49 265 1162.49 260 1114.11 260 1109.14 260 1109.76 260 1114.11 260 1114.11 255 1069.05 255 1058.79 255 1054.90 255 1057.27 255 1067.57 250 1027.10 250 1012.89 250 1003.87 250 1000.67 250 1004.66 245 988.04 245 970.00 245 956.45 245 948.10 245 946.25 240 951.68 240 930.26 240 912.34 240 899.24 240 891.96 235 917.83 235 892.57 235 871.38 235 853.90 235 841.61 230 886.33 230 858.23 230 833.29 230 811.77 230 794.83 225 857.01 225 826.13 225 797.94 225 772.69 225 751.48 220 829.73 220 796.36 220 765.07 220 736.39 220 711.22 215 804.34 215 768.60 215 734.58 215 702.74 215 673.93 210 780.71 210 742.65 210 706.23 210 671.49 210 639.32 205 758.73 205 718.65 205 679.95 205 642.55 205 607.29 200 738.28 200 696.51 200 655.52 200 615.67 200 577.57 195 719.25 195 675.93 195 632.88 195 590.79 195 550.08 190 701.54 190 656.26 190 611.84 190 567.69 190 524.59 185 685.07 185 638.52 185 592.35 185 546.33 185 501.04 IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2

'PAGE 13 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 TABLE 2.1-2 Diablo Canyon Cooldown Data at 16'EFPY Without 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.

l(°F) ~(psig) (OF) (psig) (OF;) (psig) (OF) (psig) (OF ) (psig) 180 669.74 180 622.04 180 574.25 180 526.51- 180 479.21 175 655.48 175 606.73 175 557.48 175 508.18 175 459.06 173 650.18 173 621.00 170 621.00 170 592.34 170 541.91 170 491.18 170 440.38 165 621.00 165 578.88 165 527.50 165 475.48 165 423.16 160 618.40 160 566.64 160 514.13 160 460.92 160. 407.21 155 607.73 155 555.11 155 501.76 155 447.49 155 392.53 150 597.80 150 544.57 150 490.27 150 435.04 150 378.93 145 588.56 145 534.59 145 479.67 145 423.56 145 366.43 140 579.96 140 525.35 140 469.82 140 412.93 140 354.87 135 571.97 135 516.95 135 460.73 135 403.15 135 344.25 130 564.53 130 509.15 130 452.30 130 394.09 130 334.44 125 557.61 125 501.75 125 444.53 125 385.77 125 325.46 120 551.18 120 494.92 120 437.32 120 378.06 120 317.16 115 545.19 115 488.72 115 430.69 115 371.00 115 309.57 110 539.63 110 482.81 110 424.54 110 364.46 110 302.57 105 534.45 105 477.50 105 418.88 105 358.49 105 296.20 100 529.63 100 472.58 100 413.64 100 352.96 100 290.32 95 525.15 95 467.76 95 408.83 95 347.92 95 284.99 90 520.98 90 463.56 90 404.38 90 343.27 90 280.09 85 517.11 85 459.68 85 400.30 85 339.04 85 275.65 80 513.50 80 455.91 80 396.53 80 335.14 80 271.59 75 510.15 75 452.53 75 393.09 75 331.61 75 267.92 70 507.03 70 449.30 70 389.90 70 328.36 70 264.57 Calc. N-NCM-97010 IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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') 270 OF PORV Pressure Setpoint(2) 435 psig (1) Calc. N-NCM-9701 1, Rev. 0 (2) STA-138, Rev. 0 Table 2.2-2 Low Temperature Over-Pressure (LTOP)

Temperature Restrictions Restriction Setpoint SI Pumps Secured, I CCP Secured, SI Accumulators Isolated

  • 270 OF Safety Injection Flowpath Blocked, and SI Blocked . 153 OF 2 of 3 Charging Pumps Secured S 139 °F I of 4 RCPs Secured S 131 OF 2 of 4 RCPs Secured S 115 OF 3 of 4 RCPs Secured S 101 OF 4 of 4 RCPs Secured < 91 °F RCS Vent Path of 2.07 in2 Established < 72 OF Calc. STA-138, Rev. 0 Assumptions: I) 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 2 PAGE 15 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2

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

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

3.2 At the LTOP setpoints, there is no credible way to challenge RCP number I 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 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 I 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 I Reactor Vessel Radiation Surveillance Program, December, 1987.

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

4.6 WCAP-1 595 8, Analysis of Capsule V from Diablo Canyon Unit I Reactor Vessel Radiation Surveillance Program, January 2003.

4.7 WCAP-1 1851, Analysis of Capsule U from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, May, 1988.

4.8 WCAP-1281 1,Analysis of Capsule X from Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program, December, 1990.

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON PONVER PLANT REVISION 2 PAGE 16 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 Diablo Canyon Units I 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-15782, 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 1/4t location is found in Seam Weld 3-442 C in the Unit I reactor vessel while the most limiting 3/4t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel. The Unit I Weld Surveillance Capsules are fabricated from a *veld 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 2 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 ARTNDr values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 281F for welds and 170F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed tvice 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 El 85-82.

Tables 5.0-1 and 5.0-2 present the Surveillance Capsule Data for Diablo Canyon Units I 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 I a (standard deviation) of 17'F for base metal and 28 0 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 +/- 250F.

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 251F. Hence this criteria is met.

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 18 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 5.0-1 Diablo Canyon Unit 1 Surveillance Capsule Data I I I 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 I I 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 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

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

(d) Diablo Canyon Surveillance Capsule S is currently not judged 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 2 PAGE 19 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data Matirial Material Capsule Capsule

~

CF(~&'

F FF j ~Best Fit

&RTNDT(b)

Measured ARTNDT( )

Scatter in I&RTNDT Inter Shell Plate U 0.701 69.1 65.4 -3.7 B5454-1 (Long) l Inter Shell Plate X 98.6 0.976 96.2 100.1 3.9 B5454-1 (Long) l Inter Shell Plate Y 1.121 110.5 111.6 1.1 B5454-1 (Long) l 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 (Transj _

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) II 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

(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 2 PAGE 20 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2

6. SUPPLEMENTAL DATA TABLES Table 6.0-1 Comparison of Diablo Canyon Unit I 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-I 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/4 and 3/4t Locations at 16 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, '/4 and %t Locations at 16 EFPY Table 6.0-8 Diablo Canyon Unit 1Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the '4t and 3/4t Locations for 16 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the /4t and /4t Locations for 16 EFPY Table 6.0-10 Calculation of Adjusted Reference Temperature at 16 EFPY for the Limiting Diablo Canyon Reactor Vessel Materials IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMIBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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 Pyrs 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 Prs for Units I and 2 is 2597F (Unit I Weld 3442c), 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 I) 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 (U1yl35 (U2), dated May 28, 1999.

8.3 License Amendment No. 133 (Ulyl3I (U2), dated May 3, 1999.

8.4 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Revision 2," January 1996.

8.5 PG&E letter DCL-00-070, Supplement to Reactor Coolant System Pressure and Temperature Limits Report.

8.6 "RETRAN-02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", EPRI NP-I 850-CCM-A, Project 889-3, December, 1996.

8.7 PG&E Calculation STA-138 Rev. 0, "RETRAN Evaluation of DCPP LTOP Parameters",

November 16, 2001.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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 109 n/cm2 ) Temperature Shift Decrease Predicted Measured Predicted Measured

([k (a)

) (o__) (b)_ (%)(2)

_ J 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 fluence values given here are the calculated fluence values, not the best estimate.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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/cm2 ) Temperature Shift Decrease Predicted Measured Predicted l Measured j j _____________________ (OF) (a) (ohF) (b) (% ) (a) (% ) (b)

Plate BS454-1 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.

(') The fluence values presented here are calculated fluence values, not the best estimate.

I GRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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 I - Material l Capsule l F' FFb) MauRT e d)F l FFxRTNDTF FF Intermediate Shell s( ) 0.284 0.656 -1.78 0 0.430 Plate B4106-3 Y 1.050 1.014 48.66 49.34 1.028 V 1.37 1.087 34.32 37.31 1.182 SUM 86.65 2.64 CF Plate = 7 (FF* ARTNDT) .(FF 2 ) = (86.65 0 F) -- (2.64) = 32.8 0 F (')

Weld Metal S () 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 CF weId = 7(FF* ARTNDT) - Z(FF ) = (527.09) --(2.64) = 199.7 0F (c) 2 Unit 2 - Material Capsule F-(s) FF<b) Measured ) FF FF2 I &~~~RTNTJTd)o0 F FFxART NDT0 F~

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 2 0

= (834.53 F) . (8.462) = 98.6°F CF Plate = Z(FF* ARTNDT) *.E(FF )

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 = Z(FF* ARTNDT) . Z(FF 2) = (834.26°F) . (4.23 1) = 197.2 0F 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(O.2S -0.1 ' logF)

(') 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.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 25 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.04 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 ABC 0.19 0.97 -20 Upper Shell to Inter Shell Weld 8-442(b) 0.25 0.73 -56 Inter Shell Long Welds 2-442 A,B,C 0.203(a) 1.01 8(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.OE + 17.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 26 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1AND 2 TABLE 6.0-5 DCPP-2 Reactor Vessel Beltline Material, and Chemistry, and Unirradiated Toughness Data Material Description Cu (%) Ni(%) Initial RTNDT l_________ (OF)

Upper Shell Plate (b)

B5453-1 0.11 0.60 28 B5453-3 0.11. 0.60 5 B5011-IR 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 Longeb)

Welds 1-201 A,B,C 0.22 0.87 -50 Upper Shell to Inter Shell Weld 8-201(') 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

( Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E+ 17.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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, %At,and 3/4t Locations at 16 EFPY Material Fluence f J Fluence fcbm j Fluence fJ,, Fluence f'>.,

Upper Shell Plate(a)

B4105-1 1.43 E + 17 1.39 E + 17 8.06 E + 16 2.86 E + 16 B4105-2 1.43 E + 17 1.39 E + 17 8.06 E + 16 2.86 E + 16 B4105-3 1.43 E + 17 1.39 E + 17 8.06 E + 16 2.86 E + 16 Inter Shell Plate B4106-1 6.88E+18 6.69E+18 3.89E+ 18 1.38E+ 18 B4106-2 6.88E+ 18 6.69E+ 18 3.89E+ 18 1.38E+ 18 B4106-3 6.88E+ 18 6.69E+ 18 3.89E+ 18 1.38E+ 18 Lower Shell Plate B4107-1 6.88 E + 18 6.69 E + 18 3.89 E + 18 1.38 E + 18 B4107-2 6.88 E + 18 6.69 E + 18 3.89 E + 18 1.38 E + 18 B4107-3 6.88 E + 18 6.69 E + 18 3.89 E + 18 1.38 E + 18 Upper Shell Long(a)

Welds 1-442 A,B,C 1.43 E + 17 1.39 E + 17 8.06 E + 16 2.86 E + 16 Upper Shell to Inter Shell Weld 8-442(a) 1.43 E+ 17 1.39 E+ 17 8.06 E + 16 2.86 E + 16 Inter Shell Long Welds 2-442 A,B 5.02 E + 18 4.88 E + 18 2.84 E + 18 1.01 E+ 18 Weld 2-442 C 2.51 E + 18 2.44 E + 18 1.42 E + 18 5.03 E + 17 Inter Shell to Lower Shell Weld9-442 6.88E+ 18 6.69E+ 18 3.89E+ 18 1.38E+ 18 Lower Shell Long Welds 3-442 A,B 4.05 E + 18 3.94 E + 18 2.29 E + 18 8.14 E + 17 Weld 3-442 C 6.88 E + 18 6.69 E + 18 3.89 E + 18 1.38 E + 18 Calc N-287, WCAP-15958 (a) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E + 17.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON PONWER PLANT REVISION 2 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, %Atand 3Y4t Locations at 16 EFPY Material l Fluence f, l Fluence fjbm l Fluence f.. Fluence fy, Upper Shell Plate(a)

B5453-1 1.40 E + 17 1.36 E + 17 7.94 E + 16 2.82 E + 16 B5453-3 1.40E+ 17 1.36E+ 17 7.94E+ 16 2.82E+ 16 B5011-IR 1.40 E + 17 1.36 E + 17 7.94 E + 16 2.82 E + 16 Inter Shell Plate B5454-1 6.78 E + 18 6.59 E + 18 3.83 E + 18 1.36 E + 18 B5454-2 6.78 E + 18 6.59 E + 18 3.83 E + 18 1.36 E + 18 B5454-3 6.78 E + 18 6.59E+ 18 3.83 E + 18 1.36 E + 18 Lower Shell Plate B5455-1 6.78 E+ 18 6.59 E+ 18 3.83 E+ 18 1.36 E+ 18 B5455-2 6.78 E + 18 6.59 E + 18 3.83 E + 18 1.36 E + 18 B5455-3 6.78 E + 18 6.59 E + 18 3.83 E + 18 1.36 E + 18 Upper Shell Long(a)

Welds 1-201 A,B,C 1.40 E + 17 1.36 E + 17 7.94 E + 16 2.82 E + 16 Upper Shell to Inter Shell Weld 8_201(a) 1.40E+ 17 1.36E+ 17 7.94E+ 16 2.82E+ 16 Inter Shell Long Weld 2-201 A 3.86 E+ 18 3.75 E+ 18 2.18 E+ 18 7.73 E+ 17 Weld 2-201 B 3.59 E + 18 3.49 E + 18 2.03 E + 18 7.23 E + 17 Weld 2-201 C 3.02 E + 18 2.94 E + 18 1.71 E + 18 6.06 E + 17 Inter Shell to Lower Shell Weld 9-201 6.78 E + 18 6.59 E + 18 3.83 E + 18 1.36 E + 18 Lower Shell Long Weld 3-201 A 3.02 E + 18 2.94 E + 18 1.71 E + 18 6.06 E + 17 Weld 3-201 B 3.86 E + 18 3.75 E + 18 2.18 E + 18 7.73 E + 17 Weld 3-201 C 3.59 E + 18 3.49 E + 18 2.03 E + 18 7.23 E + 17 Calc N-272, WCAP-15423 Rev. 0.

(a) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.OE + 17.

IGRSATN2.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 29 OF 31 TITLE: PTLR for Diablo Canyon UNITS I AND 2 TABLE 6.0-8 Diablo Canyon Unit I Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the /4t and 3/4t Locations for 16 EFPY 16 EFPY ART(a)

Material l M ethod 2/t (-F) 3/t (oF)

Upper Shell Plate(d)

B4105-1 Position 1.1 70.6 65.8 B4105-2 Position 1.1 51.7 46.8 B4105-3 Position 1.1 58.5 52.6 Inter Shell Plate B4106-1 Position 1.1 87.0 65.4 B4106-2 Position 1.1 90.8 70.3 B4106-3 Position 1.1 118.9 100.0 Lower Shell Plate B4107-1 Position 1.1 115.3 92.5 B4107-2 Position 1.1 114.7 93.9 B4107-3 Position 1.1 72.1 51.5 Upper Shell Long(d)

Welds 1442 A,B,C Position 1.1 20.8 -1.1 Upper Shell to Inter d)

Shell Weld 8442 Position 1.1 1.4 -12.3 Inter Shell Long Welds 2442 A,B Position 1.1 158.4 104.5 Weld 2442 C Position 1.1 120.7 76.1 Inter Shell to Lower Shell Weld 9-442 Position 1.1 136.7 93.0 Lower Shell Long Welds 3-442 A,B Position 1.1 146.2 95.0 Weld 3-442 C(c) Position 1.1 177.1(b) 119.5 CalcN-287 & WCAP-15958 (a) ART = Initial RTNDT + ARTNDT + Margin (fF)

(b) This limiting ART value is bounded by that used to generate heatup and cooldown curves (183.70 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.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 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 V4t and 3/4t Locations for 16 EFPY 11 ~~~~~~~16 EFPY ART (a) _______

Material RG 1.99 Rev. 2 'At (°-F) 3Y4t (OF) ll Method Upper Shell Plate(c)

B5453-1 Position 1.1 41.9 34.4 B5453-3 Position 1.1 46.6 42.3 B5011-IR Position 1.1 41.7 37.4 Inter Shell Plate B5454-1 Position 2.1 141.4 116.5 B5454-2 Position 1.1 174.2 149.0(b)

B5454-3 Position 1.1 148.5 120.2 Lower Shell Plate B5455-1 Position 1.1 91.1 66.3 B5455-2 Position 1.1 106.1 81.3 B5455-3 Position 1.1 96.9 77.8 Upper Shell Long(')

Welds 1-201 A,B,C Position 2.1 -11.8 -32.3 Upper Shell to Inter(c)

Shell Weld 8-201 Position 1.1 -2.5 -13.8 Inter Shell Long Weld 2-201 A Position 2.1 122.3 78.4 Weld 2-201 B Position 2.1 119.1 76.0 Weld 2-201 C Position 2.1 110.9 69.9 Inter Shell to Lower Shell Weld 9-201 Position 1.1 8.3 -3.2 Lower Shell Long Weld 3-201 A Position 1.1 76.8 38.9 Weld 3-201 B Position 1.1 84.0 48.9 Weld 3-201 C Position 1.1 81.9 46.0 Calc N-272, Calc N-282 & WCAP-1 5423 (a) ART = Initial RTNDT + ARTNDT + Margin (fF)

(b) This limiting ART value is bounded by that used to generate heatup and cooldown curves (151.4 0F).

(C) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.0E + 17.

IGRSAT02.doa 04B 1223.0646

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWVER PLANT REVISION 2 PAGE 31 OF 31 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-10 Calculation of Adjusted Reference Temperature at 16 EFPY for the Limiting Diablo Canyon Reactor Vessel Materials Parameter ART Value Location " /td l Y4 t(e)

Chemistry Factor, CF (F) 226.8() 99.6 Fluence÷ 1019 n/cm 2 (E> 1.OMeV),f(a) 0.389 0.136 Fluence Factor, FF(b) 0.7387 0.4815 ARTNDT = CF x FF, (F) 167.5(0) 48.0 Initial RTNDT, I ( 0F) -56 67 Margin, M (OF)(C) 65.5 34 ART = I+ (CF x FF) + M (0 F) 177.1(0M 149.0(0 per Regulatory Guide 1.99, Rev. 2 Calc N-NCM-97009 (a) Fluence, f, is based upon fz,, and fyt from Tables 6.0-6 and 6.0-7. The Diablo Canyon reactor vessel wall thickness is 8.625 inches at the beltline region.

( Fluence Factor (FF) per Regulatory Guide 1.99, Revision 2, is defined as FF = f(0.28 -O.I Ogf)

(c) Margin is calculated as M = 2(012+ 0A2) 0 5 . The standard deviation for the initial RTNDT margin term 0o, is 0F for plate since the initial RTNDT is a measured value. The standard deviation for ARTNDT term ca, is 171F for the plate, except that a, need not exceed the 0.5 times the mean value of ARTNDT.

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

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

%t.

(0 DCPP-1 Surveillance Capsule S was not judged "credible" per 10 CFR 50.61. The higher chemistry value of CE NPSD-1 039, Rev 2 for this heat are used to generate the heatup and cooldown Appendix G curves. The calculated ART's are bounded by the temperatures used to generate the heatup and cooldown curves (183.71F for 1/4t and 151.4 0F for 3/4t).

IGRSAT02.doa 04B 1223.0646

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