DCL-02-079, Units, 1 and 2, License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report

Units, 1 and 2, License Amendment Request 02-04, Revision of Technical Specification 5.6.6 - Reactor Coolant System Pressure and Temperature Limits Report

ML022200120
Person / Time
Site:	Diablo Canyon
Issue date:	07/31/2002
From:	Rueger G Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-02-079
Download: ML022200120 (126)

Text

Pacific Gas and Electric Company° Gregory M. Rueger US Mail Senior Vice President- Mail Code B32 Generation and Pacific Gas and Electric Company Chief Nuclear Officer P0 Box 770000 July 31, 2002 San Francisco, CA 94177-0001 Overnight Mail.

Mail Code B32 PG&E Letter DCL-02-079 Pacific Gas and Electric Company 77 Beale Street, 32nd Floor San Francisco, CA 94105-1814 U.S. Nuclear Regulatory Commission 415.9734684 Attn: Document Control Desk Fax 4159732313 Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 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 accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively. This License Amendment Request (LAR) proposes to revise Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to obtain approval of WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," for use at DCPP to allow changes to the PTLR without prior NRC approval. This LAR are also incorporates NRC approved Industry/TSTF Standard Technical Specification Change Traveler No. 419.

PG&E initially submitted the DCPP PTLR for NRC review in PG&E Letter DCL-99-146 dated November 24, 1999. Following submittal, the NRC staff requested additional information, due in part, to the diverse references PG&E used to document its PTLR methodology. After discussing the questions with the NRC staff, PG&E decided to perform a new analysis consolidating the supporting calculations into a more cohesive and comprehensive calculation and to resubmit the PTLR. In the meantime, PG&E stated it would continue to operate DCPP within the existing pressure/temperature (P/T) and low temperature overpressure protection (LTOP) limits approved in License Amendment Nos. 133 and 131, dated May 3, 1999, for DCPP Units 1 and 2, respectively. (Reference PG&E Letter DCL-00-070 dated April 26, 2000).

The changes proposed by this LAR are not required to address an immediate safety concem. New PTLR P/T and LTOP curves will be needed for the cycle 13 A member of the STARS (Strategic Teaming and Resource Sharing) Attiance Callaway e Comanche Peak

• Diabto Canyon

• Palo Verde

• South Texas Project e Wolf Creek K ,

Document Control Desk PG&E Letter DCL-02-079 )

July 31,2002 Page 2 reload cores for both Unit 1 and Unit 2. The cycle 13 core for Unit 1, the lead unit, will be installed during Refueling Outage No. 12, scheduled for February 2004. To provide sufficient lead time to prepare new curves and implement procedure changes, PG&E requests that the NRC staff complete its review and approval of this LAR by August 1, 2003. PG&E requests that the proposed TS change become effective immediately, to be implemented within 30 days from the date of issuance.

Sincerely, G g M. Rueger cc: Edgar Bailey, DHS Ellis W. Merschoff David L. Proulx Girija S. Shukla Diablo Distribution Enclosures JER

PG&E Letter DCL-02-079 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

) Docket No. 50-275 In the Matter of ) Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY SNo. DPR-80

)

Diablo Canyon Power Plant ) Docket No. 50-323 Units 1 and 2 ) Facility Operating License No. DPR-82 AFFIDAVIT Gregory M. Rueger, of lawful age, first being duly sworn upon oath says that he is Senior Vice President - Generation and Chief Nuclear Officer of Pacific Gas and Electric Company; that he has executed LAR 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.

Senior ice President - Generation and Chief Nuclear Officer Subscribed and sworn to before me this 30O'day of July, 2002 Notary blic L County of San Francisco State of California AMY EMIKO DONG 14 Commission # 1206749 zi Notary Pubric- COaifrnlf "SonFrancisco CountY 4 .___n22003 i

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Enclosure 1 PG&E Letter DCL-02-079 LICENSE AMENDMENT REQUEST FOR DIABLO CANYON POWER PLANT REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS REPORT

1.0 DESCRIPTION

This License Amendment Request (LAR) proposes to revise Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to obtain approval of WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

for use at Diablo Canyon Power Plant (DCPP) to allow changes to the PTLR without prior NRC approval. The PTLR is submitted in accordance with the guidance of Generic Letter 96-03 for NRC staff review to allow the plant specific application of the WCAP-14040-NP-A PTLR methodology to calculate new plant pressure/temperature (P/T) and low-temperature overpressure protection (LTOP) limits in the future, without prior NRC staff approval.

2.0 PROPOSED CHANGE

The proposed change would revise TS 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," by adding the phrase "and LTOP" to TS 5.6.6.b, and replacing the references under TS 5.6.6.b with WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RSC Heatup and Cooldown Limit Curves." In addition, the definition of PTLR in TS 1.1, "Definitions," would be revised to delete the reference to TS containing the limits specified in the PTLR. The proposed TS changes to reference the topical report WCAP-14040-NP-A by number and title, and to revise the PTLR definition, are consistent with NRC approved Industry/TSTF Standard Technical Specification Change Traveler No. 419 (TSTF-419).

Specifically, the proposed changes are:

TS 1.1 - PTLR definition - Delete the last sentence which states in part; "Plant operation within.. .(LTOP) System."

TS 5.6.6.b - Add "and LTOP" after "temperature" and replace the PTLR methodology references in subparagraphs 1, 2 and 3 with the following subparagraph:

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Enclosure 1 PG&E Letter DCL-02-079

1. WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RSC Heatup and Cooldown Limit Curves."

The proposed TS changes are noted on the marked up TS pages in Enclosure 2. The proposed TS pages are provided in Enclosure 3. The proposed PTLR is provided in Enclosure 4. PG&E Calculation STA-1 38, which contains the revised LTOP analysis to support the PTLR, is submitted for information in Enclosure 5. Supporting references are included in Enclosure 6 for information.

There are no changes required to the TS Bases.

3.0 BACKGROUND

NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,"

dated January 31, 1996, provides guidance for relocating P/T limit curves and LTOP system limits from TS to a PTLR or a similar document. GL 96-03 states that this alternative was based on a change included in the improved standard technical specifications (STS) to remove the P/T limit curves and LTOP system limits from the TS and relocate them to a PTLR or a similar document to reduce the number of amendment requests associated with changes to these curves and limits.

GL 96-03 states that since an LAR must be submitted whenever a change is made to the TS, the relocation of the P/T curves and LTOP system limits will result in a resource savings for the licensees and the NRC by eliminating unnecessary license amendment requests for changes to the P/T limit curves and LTOP system limits in TS.

In NRC letter to PG&E, "Conversion to Improved Technical Specifications for Diablo Canyon Power Plant, Units 1 and 2 - Amendment No. 135 to Facility Operating License Nos. DPR-80 and DPR-82," License Amendment (LA) 135/135) dated May 28, 1999, the NRC staff approved conversion of the DCPP TS to the Improved TS. As part of this conversion, the RCS P/T and LTOP limits were relocated from the TS to the DCPP PTLR. The safety evaluation (SE) for LA 135/135 stated that the limits addressed in the PTLR for TS 5.6.6 are the limits that the NRC staff previously approved in LA 133/131, dated May 13, 1999. These LAs approved P/T limit curves that are valid for 16 effective full power years (EFPY). The SE for LA 135/135 further stated that the NRC staff will review PG&E's plant-specific application of the PTLR methodology to allow PG&E to use the PTLR methodology in the future to calculate new P/T and LTOP limits without prior NRC approval.

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Enclosure 1 PG&E Letter DCL-02-079 In PG&E Letter to NRC, DCL-99-146, "Request for NRC Approval of Diablo Canyon Methodology for Establishing Pressure/ Temperature and Low Temperature Overpressure Protection Limits Using WCAP-14040-NP-A in Accordance with Generic Letter 96-03," dated November 24, 1999, PG&E initially submitted its plant specific PTLR methodology for approval. Following submittal, the NRC staff requested additional information due, in part, to the diverse references PG&E used to document its PTLR methodology. After discussing the questions with the NRC staff, PG&E decided (Reference PG&E letter to NRC, DCL-00-070, "Supplement to Reactor Coolant System Pressure and Temperature Limits Report," dated April 26, 2000) to perform a new analysis consolidating the supporting calculations into a more cohesive and comprehensive calculation, and to resubmit the PTLR methodology for approval. In the meantime, PG&E stated it would operate DCPP within the existing RCS P/T limit curves approved in LA 133/131.

TSTF-419 allows revising the TS definition of PTLR. The NRC safety evaluation that approved TSTF-419 (Reference 14) states that the definition of PTLR identifies the TS in which the P/T limits are addressed. However, TS 5.6.6.a requires that the individual TS that address RCS P/T limits be referenced. The proposed change to the definition eliminates the duplication between the definition of PTLR and TS 5.6.6.

TSTF-419 also allows topical reports identified in TS 5.6.6 to be identified by number and title. Reference 14 states this will allow licensees to use current topical reports to support limits in the PTLR without having to submit an LAR every time the topical report is revised. The PTLR would provide the specific information identifying the particular approved topical report used to determine the P/T limits or LTOP system limits.

4.0 TECHNICAL ANALYSIS

The analysis below explains how PG&E utilizes the guidance of GL 96-03 and the methodology of WCAP-14040-NP-A, Revision 2, dated January 1996, with minor variations to establish the DCPP PTLR methodology. The following discussion includes; (1) differences between the DCPP PTLR methodology and WCAP-14040-NP-A; (2) changes between the PTLR that PG&E submitted initially (Reference 7) and the PTLR submitted with this LAR; and (3) a summary of the PTLR methodology proposed for approval.

4.1 Differences from WCAP-14040-NP-A In a follow up to its initial PTLR methodology submittal (Reference 7), PG&E submitted a response (Reference 8) to an NRC staff request for additional information that identified several areas where there were differences between PG&E's determination of LTOP setpoints and WCAP-14040-NP-A.

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Enclosure 1 PG&E Letter DCL-02-079 These differences have been resolved and the DCPP PTLR methodology is now completely consistent with the WCAP-14040-NP-A methodology, but also includes the following programmatic enhancements which are not explicitly discussed in the WCAP.

1. The WCAP-14040-NP-A methodology evaluates the RCS pressure overshoot for starting a reactor coolant pump (RCP) with a maximum RCS/steam generator (SG) temperature difference of 501F and the RCS in a water solid condition. DCPP TSs 3.4.6 and 3.4.7 allow starting a RCP without any RCS/SG temperature restrictions when the pressurizer level indicates less than or equal to 50 percent. PG&E has added to the methodology an additional heat injection analysis with the pressurizer partially full to reiterate the bases for this TS limiting condition for operation (LCO) allowance. The heat injection evaluation demonstrated that the reduced pressurizer level provides enough additional expansion volume to ensure that the maximum RCS/SG temperature difference allowed within the operating procedures guidelines does not challenge the ASME Section III Appendix G P/T limits.

2. The stress correction factors, Mm and Mb, used in WCAP 14040-NP-A are taken from graphs within the WCAP and reference Welding Research Council Bulletin No. 175. The DCPP PTLR stress correction factors are graphically presented in Figures A-3300-3 and A-3300-5 of ASME Code Section Xl Appendix A. The equation for the flaw shape factor, Q, is from EPRI NP-1 181, "Computational Method to Perform the Flaw Evaluation Procedure as Specified in ASME Code, Section Xl, Appendix A," dated September 1979. The use of ASME Section Xl Appendix A is approved and discussed in LA 133/131 SE, section 3.2.2. As stated in the SE, PG&E opted to use the technical methods provided in nonmandatory Appendix A to Section Xl of the ASME Code as the methodology for generating the DCPP P/T limit curves. To test the validity of PG&E's proposed curves, the staff performed an independent assessment of the PG&E's submittal. PG&E's proposed P/T limit curves for normal operating and pressure testing conditions, effective to 16 effective full power years, were found to be slightly more conservative than the P/T limit curves generated by the staff in accordance with the methods of Appendix G to the Code.

3. The WCAP-14040-P-A methodology evaluates a variable cold overpressure mitigation system (COMS) power operated relief valve (PORV) setpoint that decreases with RCS temperature and the associated Appendix G P/T limit. The DCPP LTOP system requires only a single constant lift setpoint and administrative controls on mass injection capability and RCP operation to ensure acceptable margin as the RCS temperature decreases.

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Enclosure 1 PG&E Letter DCL-02-079

4. The structural integrity of the PORV and the associated discharge piping are assured throughout the LTOP range based on safety and relief valve testing performed in accordance with NUREG-0737. The testing demonstrated the ability of the PORVs to mitigate cold overpressurization events (Ref. NRC Safety Evaluation Report: TMI Action NUREG-0737 (ll.D.1), "Relief and Safety Valve Testing for Diablo Canyon Units 1 and 2," dated January 27, 1986). This is different from the approach specified in WCAP 14040-NP-A which was based on a generic study by Westinghouse for COMS (variable pressure setpoint system) using a type of PORV which would cause maximum back pressure in the piping during an overpressure transient. DCPP LTOP system is not a COMS system and the DCPP PORVs are not of this limiting type.

5. The neutron fluence calculated at the 1 t and 3/ t locations by PG&E is in accordance with the guidance given in Regulatory Guide (RG) 1.99, Rev. 2. That is, the fluence attenuation is from the wetted vessel surface, versus from the clad/base metal interface specified by WCAP-14040-NP-A.

6. The existing heatup/cooldown curves are based on the "best-estimate" methodology of WCAP-14040-NP-A. However, the ongoing in-vessel and ex-vessel reactor vessel (RV) radiation surveillance programs, as well as the generation of future heatup/cooldown curves (after 16 EFPY) will utilize projected fluences based on pure transport theory rather than the "best-estimate" methodology of WCAP-14040-NP-A. These pure transport theory projected fluences will be compared to plant specific measured dosimetry results for validation only. Measured results will not be used to modify future transport theory projections without prior NRC approval.

4.2 Changes from Previous PTLR Submittal There are three significant changes between the previously submitted PTLR (Reference 7) and the PTLR submitted with this LAR.

1. The previous DCPP LTOP methodology was based on a Westinghouse parametric study of DCPP mass injection cases with the LOFTRAN computer code, while the heat injection and RCS undershoot results were based on DCPP evaluations of Westinghouse generic results. The proposed PTLR is based on using the approved RETRAN computer code for the complete spectrum of mass injection, heat injection, and RCS undershoot cases consistent with the WCAP-14040-NP-A methodology.

PTLR Section 2.2, "Low Temperature Overpressure Protection (LTOP) 5

Enclosure 1 PG&E Letter DCL-02-079 Setpoints (LCO 3.4.12)," provides a detailed discussion of the DCPP LTOP methodology.

2. The LTOP administrative temperature restrictions in PTLR Table 2.2-2 have changed slightly due to minor variations in the latest RCS overpressure results obtained with the DCPP RETRAN model as compared to the original Westinghouse LOFTRAN results. In particular, the RETRAN model uses revised emergency core cooling system (ECCS) and charging injection profiles which are explicitly based on DCPP pump characteristics, and which has resulted in reduced temperature restrictions related to securing the safety injection (SI) and charging system flow paths. Also, the RETRAN model generates a more conservative estimate of the dynamic head effects on the RCS wide range pressure transmitters, which has resulted in slightly more restrictive (greater) temperature limits for RCP operation and establishing the RCS vent.

3. The bolt up temperature, based on ASME Appendix G and 10 CFR 50 Appendix G, Table 1, is required to be the initial nil-ductility temperature (RTNDT) of the flange plus any irradiation effects. The highest initial RTNDT 0

of the vessel and closure head flange materials is 53 F. The flange area is sufficiently distant from the fuel region, that the fluence has negligible affect on the RTNDT of the materials in this area. Currently, the bolt up temperature of 70°F is based on the value given in the original Combustion Engineering (CE) instruction manual for the RV. The proposed curves set the temperature at 60°F based on the Westinghouse WCAP-14040-NP-A position of Section 2.7 and correspondence from CE that upgraded the original instruction manual in conformance with ASME Code requirements (Reference 10). A copy of Reference 10 is included in Enclosure 6 for information.

In addition, to provide the option in the future, separate curves could be developed, if warranted, for Unit I and Unit 2. Currently, the composite curves are common to both units as the most limiting adjusted reference temperature (ART) between the units at the % t and % t locations is utilized for the stress intensity factor, Kia, calculation (Kia is defined under the response to Provision 5, below). Should the ART difference between the units justify separate P/T curves, those curves will be generated using the same methodologies described herein.

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Enclosure 1 PG&E Letter DCL-02-079 4.3 Methodology and PTLR To relocate the PIT curves and LTOP system limits from the TS, GL 96-03 requires the licensee to; (1) reference a methodology approved by the NRC for deriving the parameters used for constructing the curves and setpoints, (2) develop a PTLR or a similar document, and (3) make appropriate changes to the applicable sections of the TS.

The first two of the three GL 96-03 requirements for relocating the P/T curves and LTOP system limits are to use an NRC-approved methodology, and to develop a PTLR. GL 96-03, Attachment 1 contains seven provisions for PTLR methodology from the administrative controls section of the STS. The following information explains how PG&E meets each of these seven provisions.

Provision 1: "The methodology shall describe how the neutron fluence is calculated (reference new regulatoryguide when it is issued)."

The existing heatup/cooldown curves for DCPP Units 1 and 2 are based on the Westinghouse "best-estimate" methodology for calculation of neutron fluence, whereby pure transport fluences are modified with both in-vessel and ex-vessel measured dosimetry results. This methodology is as described in WCAP-14040-NP-A, and as implemented in WCAP-14284, Revision 0, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 1 - Cycles 1 through 6," dated January 1995, and WCAP-14350, Revision 0, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 2 - Cycles 1 through 6," dated November 1995.

The neutron fluence calculations for DCPP were carried out using forward and adjoint formulations in r, 0 geometry of the two-dimensional Discrete Ordinates Transport (DORT) code. The anisotropic scattering was treated with a P 3 expansion of the scattering cross section and the angular discretization was modeled with an S 8 order of angular quadrature. The core power distribution and neutron source distribution were estimated conservatively, accounting for spectral changes due to plutonium accumulation. The methodology used the BUGLE-93 cross section library which is based on the data set of the Evaluated Nuclear Data File/B-VI (ENDF/B-VI).

The fast neutron fluence is calculated for any depth into the vessel wall in conformance with RG 1.99, Revision 2, as follows:

f = fsurtce [exp (-0.24x)]

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Enclosure 1 PG&E Letter DCL-02-079 where fsufce is the calculated value of the neutron fluence at the inner wetted surface of the vessel at the location of the postulated defect and x (in inches) is the depth into the vessel wall measured from this surface.

WCAP 14040-NP-A methodology differs from this approach as it calculates fluence attenuation from the clad/base metal interface. Also, Westinghouse typically only provides a clad/base metal interface fluence (fclbm) in their surveillance capsule and cavity dosimetry reports. To resolve this difference, DCPP vessel wetted surface fluence, fs, is calculated per Reference 11 as follows:

fs = 1.029 fcdbm A copy of Reference 11 is included in Enclosure 6 for information.

As discussed above, the existing heatup/cooldown curves are based on the "best-estimate" methodology of WCAP-14040-NP-A. However, the ongoing in-vessel and ex-vessel RV radiation surveillance programs, as well as the generation of future heatup/cooldown curves (after 16 EFPY) will utilize projected fluences based on pure transport theory rather than the "best estimate" methodology of WCAP-14040-NP-A. These pure transport theory projected fluences will be compared to plant specific measured dosimetry results for validation only. Measured results will not be used to modify future transport theory projections without prior NRC approval. The latest pure transport theory methodology for DCPP, as described in WCAP-1 5423, Revision 0 (Unit 2), WCAP-1 5780, Revision 0 (Unit 1), and WCAP-1 5782, Revision 0 (Unit 2), follows the methods outlined in RG 1.190. All DCPP transport calculations are currently carried out using DORT Version 3.1 and the BUGLE-96 cross-section library. In these calculations, anisotropic scattering is treated with a P5 legendre expansion and the angular discretization is modeled with an S 16 order of angular quadrature.

Provision 2 "The Reactor Vessel MaterialSurveillance Programshall comply with Appendix H to 10 CFR Part 50. The reactorvessel materialirradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves."

The RV surveillance program is designed to monitor radiation effects on RV materials under actual operating conditions. The radiation effects are determined from changes in fracture toughness of the material obtained by pre- and post-irradiation testing of vessel material specimens removed from the surveillance capsules. 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," requires that the surveillance program satisfy ASTM Standard E-1 85 which specifies material selection, material testing, specimen sizes and specimen quantities.

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Enclosure 1 PG&E Letter DCL-02-079 The DCPP Units 1 and 2 RV surveillance programs are in compliance with 10 CFR Part 50, Appendix H. The material test requirement and the acceptance standard utilize the nil-ductility temperature, RTNDT, which is determined in accordance with ASTM E208. The empirical relationship between RTNDT and the fracture toughness of the RV steel is developed in accordance with ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, "Protection Against Non-Ductile Failure." The surveillance capsule removal schedule is presented in Final Safety Analysis Update (FSARU) Table 5.2-22 and meets the requirements of ASTM E185-70 (Unit 1) and E185-73 (Unit 2).

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

The Unit 1 surveillance program is described in WCAP-8465, "Pacific Gas and Electric Co. Diablo Canyon Unit No. 1 Reactor Vessel Radiation Surveillance Program," and WCAP-1 3440, "Supplemental Reactor Vessel Radiation Surveillance Program for the [PG&E Co.] Diablo Canyon Unit No. 1." The Unit 2 surveillance program is described in WCAP-8783, "[PG&E Co.] Diablo Canyon Unit No. 2 Reactor Vessel Radiation Surveillance Program." The withdrawal schedules are in accordance with ASTM E185, and have been reviewed and accepted by the NRC (References 12 and 13).

The current surveillance capsule reports are as follows:

1. WCAP-14284, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 1 - Cycles 1 through 6,"

January, 1995.

a. WCAP-1 1567, "Pacific Gas and Electric Company Analysis of Capsule S From Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," December, 1987.

b. WCAP-13750, "Pacific Gas and Electric Company Analysis of Capsule Y From Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," July, 1993.

2. WCAP-14350, "Pacific Gas and Electric Company Reactor Cavity Neutron Measurement Program for Diablo Canyon Unit 2 - Cycles 1 through 6,"

November, 1995.

a. WCAP-1 1851, "Pacific Gas and Electric Company Analysis of Capsule U From Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," May, 1988.

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Enclosure 1 PG&E Letter DCL-02-079

b. WCAP-12811, "Pacific Gas and Electric Company Analysis of Capsule X From Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," December, 1990.

c. WCAP-14363, "Pacific Gas and Electric Company Analysis of Capsule Y From Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," August, 1995.

PG&E letter DCL-01-004, "Reactor Vessel Material Surveillance Program Capsule V Technical Report," dated January 12, 2001, transmitted Westinghouse technical report, WCAP-15423, Revision 0, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," dated September 2000.

PG&E stated in DCL-01-004 that: (1) the results of the Capsule V specimen testing show that the limiting vessel beltline plate and weld material are behaving in accordance with previous predictions; (2) the results do not indicate any changes needed to the LTOP setpoints or P/T curves currently approved to 16 EFPY; and (3) that the PTLR will be updated with the Capsule V data upon approval of the PTLR methodology.

Provision 3: "Low temperature overpressureprotection (L TOP) system limits developed using NRC-approved methodologies may be included in the PTLR."

The DCPP LTOP system limits are determined in Calculation STA-138, "RETRAN Evaluation of DCPP LTOP Parameters," which is included in for information, and summarized in the following discussion.

The DCPP LTOP system protects the RCS from overpressure transients that could occur at low operating temperatures such as during plant startups and shutdowns. The LTOP system consists of two mutually redundant and independent systems, which can each open a PORV as described in FSARU Section 5.2.2.4. The LTOP system will maintain the RCS pressure below the ASME Section III, Appendix G isothermal pressure-temperature limit curve in accordance with ASME Code Case N-514. The LTOP system is manually enabled by reactor operators via a switch on the main control board prior to decreasing the RCS temperature below the predetermined enable temperature. The LTOP system then automatically engages the PORV LTOP lift setpoint when the RCS temperature reaches the enable temperature. The LTOP lift setpoint and enable temperature are established using the COMS methodology in the Westinghouse WCAP-14040-NP-A.

This COMS methodology has been approved by the NRC and was evaluated in the applicable safety evaluation report dated October 16, 1995, to satisfy the requirements of Standard Review Plan (SRP) Section 5.2.2 and Branch 10

Enclosure 1 PG&E Letter DCL-02-079 Technical Position RSB 5-2. The LTOP enable temperature of 270°F was calculated using ASME Section Xl, Code Case N-514 and found to ensure LTOP would be operational at the proper water temperature, which, consistent with Code Case N-514, is the greater of: (1) that associated with the vessel metal temperature at least 50°F above the RTNDT at the limiting beltline location; or (2) 2000 F. The use of ASME Code Case N-514 is described in WCAP-14040-NP-A and was approved for DCPP in LA 133/131.

As discussed in Section 4.1.3, the DCPP LTOP system requires only one constant lift setpoint as compared to the variable COMS setpoints discussed in WCAP-14040-NP-A.

The design basis of the DCPP LTOP system includes both mass injection and heat injection events as documented in Calculation STA-1 38. These events are assumed to occur with the RCS in water solid conditions and letdown isolated, so that the RCS pressure rapidly increases to the PORV actuation setpoint. The RCS pressure continues increasing even after the lift setpoint is reached and until the PORV has sufficiently opened so that the relief capacity equals the RCS 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 delay time, opening time and flow characteristics. The DCPP LTOP setpoints incorporate an appropriate conservative instrumentation delay and PORV opening time to ensure the peak RCS overshoot does not exceed the applicable Appendix G limit as documented in Calculation STA-1 38.

The design basis mass injection event is defined as the 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 temperature range is established by the TS restriction to secure the SI pumps, one centrifugal charging pump and isolating all SI accumulators prior to entering the LTOP mode of operation. As the RCS temperature decreases, successive administrative controls are implemented to prevent the ECCS charging injection discharge valves from opening, and to limit the number of charging pumps that can be operated to ensure that the maximum mass injection capability remains bounded by the LTOP design basis. As the RCS temperature continues to decrease, the number of operating RCPs is administratively limited to decrease the dynamic pressure of the system as discussed in Calculation STA-1 38, and included in the PTLR as referenced in the Bases of TS 3.4.12. The LTOP design basis includes an administrative temperature limit for establishing an RCS vent based on determining the temperature at which the reduced Appendix G P/T limit no longer has sufficient margin to accommodate the mass injection RCS overshoot associated with the PORV response time. Conversely, during a RCS heatup, the administrative restrictions are removed as the RCS temperature increases. The LTOP administrative controls are implemented 11

Enclosure 1 PG&E Letter DCL-02-079 within the Operating Procedures (OP) L-1, "Plant Heatup From Cold Shutdown to Hot Standby, OP L-5, "Plant Cooldown From Minimum Load to Cold Shutdown, and OP L-6, " Refueling."

The design basis heat injection event is described in Calculation STA-138. It is defined as the starting of the first RCP during water solid conditions with a maximum allowable temperature difference of 50°F between the RCS and the SGs as allowed by TSs 3.4.6 and 3.4.7, while the residual heat removal system is isolated. This results in a sudden heat input from the SGs and causes a pressure increase transient due to the thermal expansion of the water solid RCS. Since the residual heat removal (RHR) system is isolated, the RHR relief valves are not available. 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.

In addition to the spectrum of heat injection cases specified in WCAP-14040-NP-A, the DCPP LTOP design basis also includes an analysis which establishes that there are no RCS/SG temperature restrictions for starting a RCP when the pressurizer level indicates less than or equal to 50 percent.

This additional heat injection case evaluated the additional expansion volume provided by the reduced pressurizer level with respect to the maximum RCS/SG temperature difference allowed within the operating procedures guidelines. The SGs temperature can only exceed the RCS during a plant shutdown when the steam generator heat removal lags behind that of the RCS. The steam generators must remain slightly cooler than the RCS since they are the primary source for RCS heat removal down to a temperature of 350 0 F, at which time the RHR system could be placed into service. The DCPP operating procedures and chemistry guidelines then maintain at least one RCP operating for complete mixing and a uniform cooldown of the RCS and steam generators until the temperature has decreased below 160 0 F. At this time, all RCPs could be secured and additional RCS cooling could be performed with only RHR flow. Depending on the RHR flow conditions, the main RCS vessel and loop volumes could be cooled down while the steam generator tube bundle secondary liquid volumes remain at the higher temperature.

A worst case operational scenario for LTOP consideration assumes that the steam generator liquid volumes remain at 160 0 F, while the RCS has been cooled down to the minimum temperature at which an RCS vent must be established. A assumed RCS vent temperature of 50°F is well below the lowest expected value, and creates a potential RCS/SG temperature differential of 11 0°F. Assuming the maximum temperature measurement 12

Enclosure 1 PG&E Letter DCL-02-079 uncertainties of 150 F, starting an RCP at this time would represent a LTOP heat injection event with a net RCS/SG temperature differential of 140 0 F.

Since heat injection events become more severe with increasing RCS temperature, a LTOP heat injection analysis was performed assuming the RCS temperature was at the maximum LTOP value of 2700 F, the SG temperature was 150OF greater, and the pressurizer was at 50 percent indicated narrow range level. The heat injection analysis showed that the additional expansion volume provided by the reduced pressurizer level ensures that even these bounding LTOP RCS/SG conditions will not challenge the Appendix G P/T limits.

The major function of the LTOP system is to protect the structural integrity of the reactor pressure vessel from brittle fracture at low temperatures. In order to achieve this purpose, P/T limits established in accordance with the requirements of Appendix G to 10 CFR 50 and ASME Code Case N-514 are considered as the upper limits for the RCS during postulated transient conditions. However, since the overpressure events most likely occur at isothermal conditions in the RCS, the steady state Appendix G limits are used for the design of the LTOP system.

The structural integrity of the PORV and the associated discharge piping are assured throughout the LTOP range based on safety and relief valve testing performed in accordance with NUREG-0737. The testing demonstrated the ability of the PORVs to mitigate cold overpressurization events (

Reference:

NRC Safety Evaluation Report: TMI Action NUREG-0737 (ll.D.1), "Relief and Safety Valve Testing for Diablo Canyon Units 1 and 2," dated January 27, 1986). This is different from the approach specified in WCAP 14040-NP-A, which was based on a generic study by Westinghouse for COMS (variable pressure setpoint system) using a type of PORV, which would cause maximum back pressure in the piping during an overpressure transient.

DCPP LTOP system is not a COMS system and the DCPP PORVs are not of this limiting type.

The DCPP LTOP setpoints are established in accordance with the ASME Code Case N-514 as provided in LAs 133/131 for Unit 1 and Unit 2, respectively. The LTOP methodology provides adequate protection for the RV integrity and maintains proper operating margins. In establishing the LTOP setpoints, the DCPP specific plant parameters and transient conditions listed in Section 3.2.1 of WCAP-14040-NP-A are considered. This includes the initial volume and fluid conditions for the RCS and SGs, the PORV opening and relief characteristics, the mass input and heat input to the RCS, and the pressure limits which protect the DCPP RV. The LTOP setpoint evaluation also includes the RCP startup dynamics, and the dynamic and static pressure difference between the limiting RV weld elevation and the LTOP pressure transmitters. These LTOP analysis assumptions, including 13

Enclosure 1 PG&E Letter DCL-02-079 those listed in the PTLR, may be revised based on changes to the applicable P/T limits, changes in plant equipment, or changes in operating strategy as long as they remain consistent with the approved LTOP methodology.

A RETRAN computer model of DCPP is used to evaluate the DCPP plant response during the various LTOP transient conditions. The RETRAN code is a thermal hydraulic computer code developed by the Electric Power Research Institute (EPRI) for analyzing plant transients. PG&E previously received NRC approval (Reference LAs 108/107 dated October 1, 1995) for use of the RETRAN code in analyzing the loss of load event which was submitted as part of LAR 95-06 (Reference PG&E Letter DCL-95-220, dated September 30, 1995) to revise the main steam safety valve setpoint tolerance. The DCPP LTOP analysis model used in Calculation STA-138 is developed directly from this NRC approved RETRAN model of DCPP. In addition, this RETRAN analysis of the LTOP event has been verified to be consistent with the applicable restrictions and conditions of the latest RETRAN SER which includes the original Mod 4 SER condition responses provided by DCPP in Attachment E.1 to LAR 95-06.

The RETRAN computer code model of DCPP is benchmarked to verify it generates comparable results to the specialized LOFTRAN model used in WCAP-14040-NP-A. The two LTOP PORVs have the same lift pressure setpoint, such that if one fails the other is individually capable of mitigating an overpressure event. The opening of both PORVs simultaneously is also considered for verifying that an excessive pressure undershoot condition would not challenge the RCP number one seal performance criteria.

Consistent with WCAP-14040 Section 3.2.2, when there is insufficient range between the upper (ASME overpressure) and lower (RCP seal undershoot) pressure limits, the DCPP LTOP methodology establishes the precedence for selecting the LTOP lift setpoint to provide protection against the upper limit.

The uncertainties in the pressure and temperature instrumentation used by the LTOP system are explicitly accounted for in the determination of the LTOP setpoints and operating margins. These instrumentation uncertainties are determined using the process described in ISA Standard S67.04.01-2000.

Provision 4: "The adjusted reference temperature (ART) for each reactor beltline material shall be calculated, accounting for irradiationembrittlement, in accordancewith Regulatory Guide 1.99, Revision 2."

As described in DCPP FSARU, Section 5.2.4.1, due to the fabrication dates of the DCPP RVs, the Charpy impact test orientation for the vessel plate material was in the longitudinal direction. However, full Charpy test curves in 14

Enclosure 1 PG&E Letter DCL-02-079 the transverse direction were subsequently obtained for the intermediate and lower shell course plates of both vessels. The transverse Cv data given in FSARU Tables 5.2-17A & 5.2-17B are either the results of the available data, or estimated by adding 20°F to the longitudinal Cv data, per the guidance provided in Branch Technical Position MTEB 5-2. The initial RTNDT is then taken to be the higher of either the measured TNDT or (Tcv - 600 F), per the guidance of ASME Code,Section III, NB-2331 and WCAP-14040-NP-A.

As identified in FSARU Tables 5.2-21A & 5.2-21B, the initial RTNDT was not determined by testing for all beltline weld material. Where not available, a value of -56 0 F is used per the guidance provided in 10 CFR 50.61. The use of generic values is allowed by RG 1.99 and the specific value has been reviewed and approved by the NRC in NRC Letter dated June 28, 1996, "Diablo Canyon 1: Assessment Of Diablo Canyon Surveillance Material For Issuance Of Revision 1 Of The Reactor Vessel Integrity Database."

The method for calculating ART is performed as described in WCAP-14040-NP-A, which conforms with RG 1.99, Revision 2. The ART is calculated by adding the initial nil-ductility transition reference temperature of the unirradiated material (IRTNDT), the shift in reference temperature caused by irradiation (ARTNDT), and a margin to account for uncertainties as follows:

ART = IRTNDT + ARTNDT + margin The determination of ARTNDT due to irradiation conforms to RG 1.99, Revision 2, and is calculated as follows:

ARTNDT = CF x f**(0.28 - 0.10 log f) where CF is the chemistry factor and f is the neutron fluence at a specific depth calculated as described in the Provision 1 discussion. The CF is taken from RG 1.99, Revision 2, based on the copper and nickel content of the vessel material. Alternatively, the CF is calculated using credible surveillance data.

The margin included in the ART calculation, in conformance with RG 1.99, Revision 2, is included to account for uncertainties in the values of IRTNDT, copper and nickel contents, fluence and the calculation procedures. The margin is calculated as follows:

Margin = 2 x [ai**2 + a*'2]**1/2 where o, is the standard deviation for IRTNDT and 0 A is the standard deviation for ARTNDT. If IRTNDT is a measured value, ot is estimated from the precision of the test method. For generic mean values, ai is the standard deviation 15

Enclosure 1 PG&E Letter DCL-02-079 from the set of data used to establish the mean. The ARTNDT standard deviation, 0A, is 28 0 F for welds and 170F for base metal in accordance with RG 1.99, Revision 2, except that a, need not exceed half the mean value of ARTNDT. a,& is reduced by half when credible surveillance data are used.

Provision 5: "The limiting ART shall be incorporatedinto the calculation of the pressureand temperature limit curves in accordance with NUREG-0800, SRP Section 5.3.2, Pressure-TemperatureLimits."

PG&E used linear elastic fracture mechanics from ASME Section III, Appendix G, and ASME Section Xl, Appendix G, for calculating the P/T limits.

The method is based on restricting the stress intensity factor of the postulated defect to be less than the reference stress intensity factor of the RV material, KIR. This factor is denoted as Kia in ASME Section Xl equations and in WCAP-14040-NP-A and will be used going forward. The Kia is determined by the metal temperature and RTNDT at the tip of the postulated flaw. The flaw is assumed to have a depth of one-fourth of the beltline thickness and a length of 1.5 times the beltline thickness. The Kia curve in the ASME Code is given as follows:

KIa = 26.78 + 1.223 x exp[O.0145(T - RTNDT + 160)]

where T is the metal temperature and RTNDT is the ART value of the limiting vessel material at the / t and / t locations of the vessel wall. The stress intensity factor caused by the postulated crack is limited to the reference stress intensity factor of the vessel material as follows:

C x KIM + KIT < Kia where KiM is the stress intensity factor caused by pressure (membrane) stress, KIT is the stress intensity factor caused by the thermal stress and C is a safety factor that is 2 for the heatup and cooldown, and 1.5 for the hydrostatic and leak test conditions when the reactor is not critical.

Equations used in determination of KIT are in accordance with Westinghouse Electric Corporation Proprietary computer program "OPERLIM" which was verified by Westinghouse in the requests for additional information contained in WCAP-14040-NP-A. The KIT solution is used in combination with Kia determined above to solve for the limiting KIM as follows:

KIm(max) = (KIa - KIT)/2 Using the methodology from ASME Section Xl, Appendix A Article A-3000, KIM calculated above is related to the specific vessel geometry and the postulated flaw size by the following equation:

16

Enclosure 1 PG&E Letter DCL-02-079 KIM = (I-aMm + O'bMbVl where am is the membrane stress and ob is the bending stress in psi, "a" is 0.5 times the axis of the elliptical flaw, Q is the flaw shape parameter, Mm and Mb are the correction factor for membrane stress and bending stress, respectively, from Section Xl Appendix A Figures A-3300-3 and 3300-5. The equation for the flaw shape factor, Q, is from EPRI NP-1 181 "Computational Method to Perform the Flaw Evaluation Procedure as Specified in ASME Code,Section XI, Appendix A", September, 1979, and for a/t = 0.25 and a/l = 0.167 (I is the major axis of the flaw) is as follows:

(ICrm I+I+ b "1)2 Q= 1.2404- 0.212 2 o-ys where oa is the minimum specified yield strength for ASME SA-533, the vessel beltline material.

This ASME Section Xl Appendix A and EPRI NP-1 181 based methodology for determining the flaw shape factor is different from that used in WCAP-14040-NP-A, Revision 2, and SRP Section 5.3.2. This difference is approved and discussed in LAs 133/131 SE, Section 3.2.2.

The general equation for hoop stress due to internal pressure in a thick wall pressure vessel:

P = a [(Ro**2 - Ri**2)/(Ri**2 + Ro**2)]

is applied in an iterative process to calculate the allowable pressure. The steady state, cooldown and heatup P/T curves are determined using this process. For steady state, KIT is zero and Kia is determined at the 1/4 t location. For cooldown, KIT and Kia are determined at the 1 t location. The P/T curve at 1/ t is compared with the steady-state curve. The allowable pressure for cooldown is determined by the lesser of the two values, and the resulting curve is the composite cooldown limit curve. For heatup, KIT and Kla are determined at the % t and / t locations. The P/T curves at % t, % t, and steady-state are compared. The lowest of the three for each heatup rate is used to generate the composite heatup limit curve. The composite cooldown limit curve and composite heatup limit curve provide the allowable operating range for operation. The composite curves are common to both units as the most limiting ARTNDT between the Units at the / t and % t locations is utilized for the Kia calculation. Should the ARTNDT difference between the units justify separate P/T curves, those curves will be generated using the same methodologies described herein. This composite curve construction is in accordance with The WCAP-14040-NP-A methodology.

17

Enclosure 1 PG&E Letter DCL-02-079 Provision 6: "The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporatedinto the pressure and temperature curves."

10 CFR Part 50, Appendix G, imposes a minimum temperature at the closure flange based on the reference temperature for the flange material. With the core not critical; (1) when pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120°F for normal operation and by 90°F for hydrostatic pressure tests and leak tests. This "flange notch" has been incorporated into the P/T curves; (2) When the pressure is less than 20 percent of the pre-service system hydrotest pressure, the temperature of the closure flange region's highly stressed by the bolt preload must be greater than the reference temperature for the flange material. This value therefore becomes the minimum bolt up temperature The bolt up temperature is based on ASME Appendix G, and RG 1.99 that require the bolt up temperature to be the initial RTNDT of the flange plus any irradiation effects. The flux exposed in the RV flange and RV head flange result in negligible RTNDT shift, and, thus minimum bolt up temperature does not change with time. The boltup temperature of 70°F in the previously provided PTLR is based on the value given in the original CE Instruction Manual for the Reactor Vessel. The highest flange RTNDT, calculated in accordance with Branch Technical Position MTEB 5-2, between DCPP Unit 1 and 2 is 53 0F (Unit 1 RV closure head flange). The proposed curves set the temperature at 60°F based on the Westinghouse WCAP 14040-NP-A position of Section 2.7 and correspondence from CE (Reference 10).

Between the minimum bolt up temperature and the minimum LTOP operating temperature (72 0 F), a 2.07 square inch opening is relied on for RCS venting when the RV head is bolted up.

When the core is critical the minimum temperature for the RV, in accordance with 10 CFR 50, Appendix G, is the larger of (1) the minimum permissible temperature for the in-service hydrostatic pressure test, or (2) the limiting flange RTNDT +160 0 F. These minimum temperature requirements are reflected in the DCPP P/T heatup and cooldown curves.

Provision 7: "Licensees who have removed two or more capsules should compare for each surveillance materialthe measured increasein reference temperature (RTNDT) to the predicted increase in RTNDT, where the predicted increasein RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2 a4) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increasein RTNDT + 2 cd ), the 18

Enclosure 1 PG&E Letter DCL-02-079 licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology."

Multiple surveillance capsules have been removed and evaluated for both units in compliance with 10 CFR Part 50, Appendix H. The PTLR presents the measured and predicted 30 ft-lb transition temperature shift (ARTNDT) for the plate and weld surveillance materials for all capsules, both credible and non-credible. Currently, the PTLR surveillance capsule data meets the required 2 0A between measured and predicted ARTNDT as required by this provision for credible surveillance data.

The credibility of each capsule's surveillance data has been evaluated in accordance with the five criteria of RG 1.99, Revision 2, Section B. Unit 1 does not have two or more sets of either credible surveillance plate or weld data at this time. Therefore, the requirement of this provision currently does not apply to Unit 1. For Unit 2, three sets of credible data are presented in the PTLR for both the surveillance plate and weld materials. The absolute difference between the measured and predicted ARTNDT was less than the plate 2 oA (340 F) for each of the plate samples and less than the weld 2 on (56°F) for each of the three surveillance weld samples. Therefore, measured surveillance data is currently consistent with the proposed methodology.

Technical Specifications GL 96-03 requires three separate actions to modify the plant TS:

Action (1) "Definitions"- the addition of the definition of a named formal report (PTLR or a similardocument) that would contain the explanations, figures, values, and parametersderived in accordance with an NRC approved methodology and consistent with all of the design assumptions and stress limits for cyclic operation; DCPP Unit I and Unit 2 TS contain the following definition in Section 1.1, "Definitions:"

19

Enclosure 1 PG&E Letter DCL-02-079 PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limits, including heatup and cooldown rates, and the power operated relief valve (PORV) lift settings and arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System."

This definition meets the requirements of GL 96-03 and is being revised in accordance with TSTF-419, to delete the last sentence which states in part; "Plant operation within.. .(LTOP) System."

• Action (2) LCOs - the addition of references to the PTLR noting that the P/T limits shall be maintained within the limits specified in the PTLR, DCPP Unit 1 and Unit 2 TS contain the following:

LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR.

LCO 3.4.12 An LTOP System shall be OPERABLE with no safety injection pumps and a maximum of one centrifugal charging pump capable of injecting into the RCS and the accumulators isolated and one of the following pressure relief capabilities:

a. Two Class I power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, or

b. The RCS depressurized and an RCS vent of > 2.07 square inches.

20

Enclosure 1 PG&E Letter DCL-02-079 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Protection (LTOP) arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.3, "RCS Pressure and Temperature (PIT)

Limits," and

2. Specification 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

These TS LCOs/requirements meet the requirements of GL 96-03.

Action (3) "AdministrativeControls"- the addition of a reporting requirement to submit the PTLR to the NRC, when it is issued, for each reactorvessel fluence period.

TS 5.6.6.c states "The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto."

This TS requirement meets the requirements of GL 96-03. However changes to the references to TS 5.6.6.b, as proposed under Section 2.0 of this enclosure, are needed to complete relocation of the P/T limit curves and LTOP system limits from TS to the PTLR in accordance with GL 96-03.

Conclusion The proposed methodology, and PTLR, and the existing TS as modified by the proposed changes, meet the provisions required by GL 96-03, for relocation of the PIT limit curves and LTOP system limits from TS to the PTLR. The proposed changes also meet the provisions of TSTF-419.

21

Enclosure 1 PG&E Letter DCL-02-079

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probabilityor consequences of an accidentpreviously evaluated?

Response: No The proposed Technical Specification (TS) changes provide the reference for the NRC approved methodology for the Diablo Canyon Power Plant (DCPP) Pressure And Temperature Limits Report (PTLR).

The TS and PTLR were developed using the guidance of NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31, 1996, which provides guidance on relocating reactor coolant system (RCS) pressure/temperature (P/T) limit curves and low-temperature overpressure (LTOP) system limits from TS to a PTLR. NRC approval of the DCPP specific application of the PTLR methodology will allow PG&E to use the approved PTLR methodology in the future to calculate new PIT and LTOP limits without prior NRC staff approval.

The proposed PTLR was developed using methodology previously approved by the NRC, primarily WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," dated January 1996. PG&E has evaluated this methodology and concludes it is applicable for use at DCPP. As a result, use of this methodology does not involve a significant increase in the probability or consequences of an accident previously evaluated.

22

Enclosure 1 PG&E Letter DCL-02-079

2. Do the proposed changes create the possibility of a new or different kind of accident from any accidentpreviously evaluated?

Response: No The proposed change completes relocation of the RCS P/T and LTOP limits from the TS to the PTLR. The DCPP PTLR submitted with this amendment has been developed primarily using the NRC-approved methodology of WCAP-14040-NP-A, Revision 2.

The proposed change makes no changes to plant equipment, and does not physically alter or change the function of any structures, systems or components that could initiate an accident. Through the PTLR, it provides operational controls to assure that current RCS P/T and LTOP limits are not violated. It provides for use of NRC-approved methodology for changing the RCS P/T and LTOP limits in the future without requiring prior NRC approval. As a result, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed change completes relocation of the RCS P/T and LTOP limits from the TS to the PTLR, and submits the DCPP PTLR methodology for NRC approval. The DCPP PTLR submitted with this amendment has been developed using the methodology of WCAP-14040-NP-A, Revision 2, which has previously been approved by the NRC.

The proposed change makes no changes to plant equipment, and does not physically alter or change the function of any structures, systems or components that could affect any margin of safety.

Through the PTLR, it provides operational controls to assure that current RCS P/T and LTOP limits are not violated. It provides for use of NRC approved methodology for changing the RCS P/T and LTOP limits in the future without requiring prior NRC approval. As a result, the proposed change has no affect on any margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

23

Enclosure 1 PG&E Letter DCL-02-079 Based on the above evaluations, PG&E concludes that the activities associated with the above described change present no significant hazards consideration under the standards set forth in 10 CFR 50.92 and accordingly, a finding by the NRC of no significant hazards consideration is justified.

5.2 Applicable Regulatory Requirements/Criteria In LA 135/135 for both DCPP Units 1 and 2, the NRC staff approved conversion of the DCPP TS to Improved TS. As part of this conversion, the RCS P/T and LTOP system limits were relocated from the TS to the DCPP PTLR. The SE for LA 135/135 stated that the limits addressed in the PTLR of TS 5.6.6 are the limits that the NRC staff previously approved in LA 133/131 dated May 3, 1999. These LAs approved P/T limit curves that are valid for 16 EFPY. The SE for LA 135/135 further stated that the NRC staff will review PG&E's plant-specific application of the PTLR methodology to allow PG&E to use the PTLR methodology in the future to calculate new P/T and LTOP limits without prior NRC approval.

The application of the PTLR methodology submitted for approval by this LAR was developed in accordance with the requirements of GL 96-03. The technical analysis presented in Section 4.0 above describes how the PTLR and associated TS meet the provisions of GL 96-03, including use of an NRC-approved methodology, specifically WCAP-14040-NP-A, Revision 2.

Submittal of this LAR completes a commitment made in PG&E letter DCL-00-070 in which PG&E stated it would resubmit the PTLR after performing a new analysis consolidating the supporting calculations into a more cohesive and comprehensive calculation. PG&E stated, In the meantime, it would operate DCPP within the existing RCS P/T limit curves approved in LA 133/131.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

PG&E has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the 24

Enclosure 1 PG&E Letter DCL-02-079 proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7.0 REFERENCES

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

2. NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31, 1996.

3. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements."

4. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

5. NRC letter to PG&E, "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit 1 and Unit 2," dated May 3, 1999. (License Amendments 133/131)

6. NRC letter to PG&E, "Conversion to Improved Technical Specifications for Diablo Canyon Power Plant, Units 1 and 2 - Amendment No. 135 to Facility Operating License Nos. DPR-80 and DPR-82," dated May 28, 1999.

7. PG&E Letter to NRC, DCL-99-146, "Request for NRC Approval of Diablo Canyon Methodology for Establishing Pressure/ Temperature and Low Temperature Overpressure Protection Limits Using WCAP-14040-NP-A in Accordance with Generic Letter 96-03," dated November 24, 1999.

8. PG&E Letter to NRC, DCL-00-039, "Supplement to PG&E's Request for NRC Approval of Diablo Canyon Methodology for Establishing Pressure/

Temperature and Low Temperature Overpressure Protection Limits Using WCAP-14040-NP-A in Accordance with Generic Letter 96-03,"

dated March 16, 2000.

9. PG&E letter to NRC, DCL-00-070, "Supplement to Reactor Coolant System Pressure and Temperature Limits Report," dated April 26, 2000.

10. CE Power Systems letter to Westinghouse Nuclear Energy Systems dated September 12, 1979, transmitted to PG&E by Westinghouse letter 25

Enclosure 1 PG&E Letter DCL-02-079 PGE-4083 dated July 21, 1980 as amended by Westinghouse letter PGE-6352 dated December 20, 1984.

11. Westinghouse letter PGE-88-765 dated December 14, 1988.

12. NRC Letter dated September 4, 1992, "Evaluation of Diablo Canyon Unit 1 Supplemental Reactor Vessel Radiation Surveillance Program (TAC No. M83285)."

13. NRC Letter dated February 10, 1998, "Pacific Gas & Electric Company's Revision to the Reactor Vessel Surveillance Capsule Withdrawal Schedule for Diablo Canyon Unit No. 2 (TAC No. M99917)."

14. NRC Letter from William D. Beckner to Anthony R. Pietrangelo, NEI, dated March 21, 2002.

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Enclosure 2 PG&E Letter DCL-02-079 Proposed Technical Specification Changes Remove Insert Page 1.1-5 Page 1.1-5 Page 5.0-28 Page 5.0-28

Definitions 1.1 1.1 Definitions (continued)

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, and the power operated relief valve (PORV) lift settings and arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. JPlan3.eperatiorffhin th Eoperatipg QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3411 MWt for each unit.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.

(continued)

DIABLO CANYON - UNITS 1 & 2 1.1-5 Unit 1 - Amendment No. -35,.ýA TAB 1.0 - R2 5 Unit

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.- J*C Letter frot f*RC to Gre ry M. Rueger d9 4 d May 28, 19,07 The anal "cal methods ed to determi hi tenme ure limits w developed in ccord, CFR 50, Ap ndix G nd H Regulatory e 1.99, Revi n 2 NUREG-* 0, Standa view Plan ctior Branch echnical Posi .in MTEB 5 AS B&PV Code ction III, A1endix G ME B&PV Co , Section , Appendix WCAP-1404 P-A, Secti 2.2  !

NUREG,,'800, Standpfd Review Plan Bra Technical sition RSB 5 I1tFR 50, Ap ndix G and H

,*egulatory 0ide 1.99, Re ion 2 Branch T hnical Positi MTEB 5-2 WCAP 4040-NP-A ection 2.2

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Not Used 5.6.8 PAM Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Not Used nued)

DIABLO CANYON - UNITS 1 & 2 5.0-28 Unit 1 - Amendment No.

TAB 5.0 - R2 29 Unit 2 -Amendment No..

Insert 1 for TS 5.6.6

1. WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

Enclosure 3 PG&E Letter DCL-02-079 Revised Technical Specification Pages

Definitions 1.1 1.1 Definitions (continued)

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, and the power operated relief valve (PORV) lift settings and arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

-t QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3411 MWt for each unit.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.

(continued)

DIABLO CANYON - UNITS 1 & 2 1.1-5 Unit I - Amendment No. 435,443 TAB 1.0 - RX 5 Unit 2 - Amendment No. 1-35

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

b. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

I

1. WCAP 14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Not Used 5.6.8 PAM Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Not Used (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-28 Unit 1 - Amendment No. --35 TAB 5.0 - RX 31 Unit 2 - Amendment No. 435

Enclosure 4 PG&E Letter DCL-02-079 Diablo Canyon PTLR

• • • 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 30 PRESSURE AND TEMPERATURE LIMITS REPORT UNITS TITLE: PTLR for Diablo Canyon I 2 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)2 OPERA TING LIM ITS ................................................................................................... 2 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) .................................................... 2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12) ..................... 4 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM .......................................... 14 REACTOR VESSEL SURVEILLANCE DATA CREDIBILITY .............................................. 15 SUPPLEMENTAL DATA TABLES ................................................................................ 19 PRESSURIZED THERMAL SHOCK (PTS) SCREENING ................................................... 20 REFER EN C ES .......................................................................................................... 20 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 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 100°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 8 Errors 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 1 Surveillance Capsule Data 17 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data 18 IGRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 2 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 I. REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

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

"* LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits

"* LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems

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

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

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

The RCS 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 'A of the reactor vessel wall thickness and an axial length of 1.5 times wall thickness and is elliptically shaped.

1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 3 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 10CFR50 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 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.

2.1.2 Reactor Vessel Bolt-up Temperature:

Operating Restrictions illustrated on the P-T curve also include Reactor Flange Boltup Temperature. This is based on ASME Appendix G and 10CFR50 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 53deg F (Unit 1 R.V. closure head). The curves conservatively set the temperature at 60 deg F based on WCAP 14040-NP-A minimum temperature. Between the minimum bolt up temperature and the minimum LTOP operating temperature (72 deg F), a 2.07 sq. in. opening is relied on for RCS venting.

IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-I DIABLO CANYON POWER PLANT REVISION 2 PAGE 4 OF 30 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-5 14.

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.

IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 5 OF 30 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 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 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.

IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 6 OF 30 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.

IGRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 7 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500.0 2250.0 2000.0 1750.0 LLJ 0x 1500.0 C/)

LLJJ1250.0 C/)

CC 1000.0 750.0 " 'I 26o~ -- ufcty-k 500.0 in-Boltui 250.0 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°F/hr)

Applicable to 16 EFPY (Without Margins for Instrumentation Errors) 1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 8 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 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.

(OF) (psig) (OF) (psig) (OF) (psi8) (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 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 9 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 2.1-1 Diablo Canyon Heatup Data at 16 EFPY Without 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) 240 951.68 240 890.92 245 988.04 245 928.66 245 1313.55 250 1027.10 250 969.17 250 1365.16 255 1069.05 255 1012.68 255 1420.55 260 1114.11 260 1059.36 260 1479.99 265 1162.49 265 1109.48 265 1543.76 270 1214.44 270 1163.26 270 1612.16 275 1266.63 275 1220.93 315 1220.93 275 1685.50 280 1321.05 280 1282.77 320 1282.77 280 1764.12 285 1379.42 285 1349.08 325 1349.08 285 1848.36 290 1442.01 290 1420.15 330 1420.15 290 1938.58 295 1509.11 295 1484.66 335 1484.66 295 2035.17 300 1581.04 300 1547.80 340 1547.80 300 2138.51 305 1658.10 305 1615.38 345 1615.38 305 2249.01 310 1740.65 310 1687.76 350 1687.76 310 2367.09 315 1829.05 315 1765.22 355 1765.22 315 2493.16 320 1923.67 320 1848.14 360 1848.14 320 2627.63 325 2024.86 325 1936.80 365 1936.80 325 2770.93 330 2133.08 330 2031.65 370 2031.65 330 2923.46 335 2248.68 335 2132.94 375 2132.94 335 3085.60 340 2372.13 340 2241.22 380 2241.22 345 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 2751.09 365 3121.20 365 2899.78 405 2899.78 370 3300.09 370 3057.72 410 3057.72 375 3489.35 375 3225.27 415 3225.27 Calc. N-NCM-97010 1GRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-I DIABLO CANYON POWER PLANT REVISION 2 PAGE 10 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 2500.0 2250.0 2000.0 1750.0 1500.0 w

1250.0 w.IL a

U 1000.0 750.0 500.0 250.0 0.0 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°F/hr) Applicable to 16 EFPY (Without Margins for Instrumentation Errors)

IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 11 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 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.

(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (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 7%.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 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 12 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 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°F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (psig) (OF) (psig) (OF) (psig) (*F) (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 L ---

Calc. N-NCM-97010 IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 13 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 2.2-1 Low Temperature Over-Pressure (LTOP)

System Setpoints Function Setpoint PORV Arming Temperature"'

P 270 OF PORV Pressure Setpointa2 435 psig (1) Calc. N-NCM-97011, Rev. 0 (2) STA-138, Rev. 0 Table 2.2-2 Low Temperature Over-Pressure (LTOP)

Temperature Restrictions Restriction Setpoint SI Pumps Secured, 1 CCP Secured, SI Accumulators Isolated _ 270 OF Safety Injection Flowpath Blocked, and SI Blocked *_ 153 OF 2 of 3 Charging Pumps Secured _<139 OF 1 of 4 RCPs Secured < 131 OF 2 of 4 RCPs Secured < 115 OF 3 of 4 RCPs Secured *101 OF 4 of 4 RCPs Secured *91 OF RCS Vent Path of 2.07 in 2 Established < 72 OF Calc. STA-138, 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 2 PAGE 14 OF 30 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 1 seal operation.

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

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

Diablo Canyon Units 1 & 2 each have their own independent material surveillance program allowing each to have its own unit specific heat up and cooldown curves and LTOP setpoints.

Both units. are currently operated using the same limitations resulting from the most conservative limitations in either unit.

The programs are described in the following:

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

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

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

The surveillance capsule reports are as follows:

4.4 WCAP- 11567, Analysis of Capsule S From Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program, December, 1987.

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

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

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

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

1GRSATY2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-I DIABLO CANYON POWER PLANT REVISION 2 PAGE 15 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Diablo Canyon Units 1 and 2 also have Reactor Cavity Neutron Measurement Programs described in:

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

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

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

4.12 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 two surveillance capsules removed and analyzed from the Diablo Canyon Unit 1 reactor vessel and three 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 'At location is found in Seam Weld 3-442 C in the Unit 1 reactor vessel while the most limiting U t location is found in the Intermediate Shell Plate B5454-2 in the Unit 2 reactor vessel. The Unit 1 Weld Surveillance Capsules are fabricated from a weld manufactured using the same weld wire heat number (Heat 27204).

1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 16 OF 30 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 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 1 a (standard deviation) of 17*F for base metal and 281F for weld material.

The Diablo Canyon Unit 1 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 1 and 2 of 10CFR50.61 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 may 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 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 17 OF 30 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 CFO' FF ARTrT(b) ARTNDT()4 RTNDT Inter Shell Plate S(d) 32.2 0.675 21.7 -2.0 23.7 B4106-3 Inter Shell Plate Y 1.006 32.4 46.9 -14.5 B4106-3 Surveillance Weld S(d) 211.2 0.675 142.6 110.0 32.6 Heat 27204 Surveillance Weld Y 1.006 212.5 234.3 -21.8 Heat 27204 WCAP 13771 (a) CF is calculated from surveillance data using Reg. Guide 1.99 Regulatory Position 2.1 (see Table 6.0-3).

Nb) Best fit ARTNDT = CF

• FF.

(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.

IGRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 18 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 Table 5.0-2 Diablo Canyon Unit 2 Surveillance Capsule Data Best Fit Measured Scatter in Material Capsule CF`a) FF ARTNDT W ARTNDT (0 ARTNT Inter Shell Plate U 0.716 73.4 65.9 7.5 B5454-1 (Long)

Inter Shell Plate X 102.5 0.960 98.4 101.0 -2.6 B5454-1 (Long)

Inter Shell Plate Y 1.08 110.7 113.0 -2.3 B5454-1 (Long)

Inter Shell Plate U 0.716 73.4 72.3 1.1 B5454-1 (Trans)

Inter Shell Plate X 102.5 0.960 98.4 98.9 -0.5 B5454-1 (Trans)

Inter Shell Plate Y 1.08 110.7 110.7 0.0 B5454-1 (Trans)

Surveillance Weld U 0.716 151.6 173.8 -22.2 Surveillance Weld X 211.7 0.960 203.2 204.2 -1.0 Surveillance Weld Y 1.08 228.6 212.5 16.1 WCAP- 14364 (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 2 PAGE 19 OF 30 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, 'At and 4 t 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, 'At and Ut Locations at 16 EFPY Table 6.0-8 Diablo Canyon Unit 1 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 'At and Mt Locations for 16 EFPY Table 6.0-9 Diablo Canyon Unit 2 Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the '4At and 'At 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 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 20 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2

7. PRESSURIZED THERMAL SHOCK (PTS) SCREENING 10 CFR 50.61 requires that RT prs be determined for each of the vessel beltline materials. The RT prs 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 .s for Units I and 2 is 259°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.

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

8.2 License Amendment No. 135 (U1)/135 (U2), dated May 28, 1999.

8.3 License Amendment No. 133 (U1)/131 (U2), dated May 3, 1999.

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

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

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 21 OF 30 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 30 ft-lb Transition Upper Shelf Energy (X 10" n/cm2 ) Tempe rature Shift Decrease Predicted

( OF) (Ja I Measured (OF) b Predicted Measured

(%) (,) (%) (b)W Plate B4106-3 S 0.305 35 -2 14 0 Y 1.02 52 47 19 3(10)

Surveillance Weld S 0.305 150 110 26 11 Metal Y 1.02 224 234 34 33 (39)

Heat Affected S 0.305 -- 77 -- 15 Zone Metal y 1.02 -- 84 -- 26 (26)

Correlation Monitor S 0.305 68 66 18 2 Plate HSST 02 y 1.02 102 112 23 2 (10)

WCAP-13750

( 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 by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

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

1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 22 OF 30 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 30 ft-lb Transition Upper Shelf Energy Materials Capsule (X 10'9 n/cm2) Temperature Shift Decrease Predicted Measured Predicted Measured (OF) (a) (oF) (%a) (

W(%)

Plate B5454-1 U 0.357 74.0 65.9 18 14.4 (Longitudinal) X 0.866 99.2 101.0 22 20.7 Y 1.320 111.3 113.0 24 18.4 Plate B5454-1 U 0.357 74.0 72.3 18 0.4 (Transverse) x 0.866 99.2 98.9 22 10.3 Y 1.320 111.3 110.7 24 6.6 Surveillance U 0.357 150.9 173.8 28 29.7 Weld Metal X 0.866 202.3 204.2 34 38.5 Y 1.320 226.9 212.5 38 36.1 Heat Affected U 0.357 -- 234.2 - 40.3 Zone Metal X 0.866 253.5 -- 31.4 Y 1.320 255.3 - 37.4 WCAP-14363 (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 by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 23 OF 30 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 FVa1 FF(b) Measured AR T mT(d) FFxARTND T FF2 Intermediate Shell S (C) 0.305 0.675 -2 0 0.456 Plate B4106-3 y 1.020 J 1.006 46.9 47.2 1.012 SUM 47.2 1.468 CF ple =(FF* ARTNDT) + X(FF 2) = (47.2 0 F) (1.468) = 32.2°F

• Weld Metal S (* 0.305 0.675 110 74.3 0.456 Y 1.020 1.006 234.3 235.7 1.012 SUM 310.0 1.468 2 = (310.0) + (1.468) = 211.2°F (c)

CF ,eld = F(FF* ARTNDT) + Y(FF )

Unit 2 - Material Capsule F FF Measured FFxARTND ARTNDT~d T Intermediate Shell U 0.357 0.716 65.9 47.2 0.513 Plate X 0.866 0.960 101.0 97.0 0.922 B5454-1 (Long) Y 1.320 1.080 113.0 121.7 1.160 Intermediate Shell U 0.357 0.716 72.3 51.8 0.513 Plate B5454-1 X 0.866 0.960 98.9 94.9 0.922 (Transverse) Y 1.320 1.080 110.7 119.2 1.160 SUM 531.8 5.190 CF Plt.,= Z(FF* ARTNDT) + X(FF 2) = (531.8°F) + (5.19) = 102.5°F U 0.357 0.716 173.8 124.4 0.513 Weld Metal X 0.866 0.960 204.2 196.0 0.922 Y 1.320 1.080 212.5 228.9 1.160 SUM 549.3 2.600 CF Weld = I(FF* ARTNDT) + Y(FF2 ) = (549.3°F) + (2.600) = 211.7°F I WCAP-13771 (Unit 1) WCAP-14364 (Unit 2)

"( F = Calculated Fluence (1019 nlcm 2 , E > 1.0 MeV).

(b) FF = Fluence Factor = 1° 28 -°0.1-lon Unit 1 Capsule S is not currently judged "credible" per RG 1.99, Rev 2. All other capsules are (C)

"credible" per RG 1.99, Position C.2. I (d) Calculated using Charpy data plotted by EPRI Hyperbolic Tangent Curve Fitting Routine, Revision 2.0.

1GRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 24 OF 30 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 RTNT (OF)

Upper Shell Plate ,

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(l) 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.

IGRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 25 OF 30 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 RTNT (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(`* -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

(') Per CE NSPD-1039, Rev. 2 (b) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.OE + 17.

IGRSATO2.doa 04B 0710-1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 26 OF 30 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 33/4t Locations at 16 EFPY Material J Fluence f, Fluence f,/. Fluence f,,, Fluence f.,

Upper Shell Plate(a) 1.59 E + 17 1.54 E + 17 9.00 E + 16 3.20 E + 16 B4105-1 1.59 E + 17 1.54 E + 17 9.00 E + 16 3.20 E + 16 B4105-2 1.59 E + 17 1.54 E + 17 9.00 E + 16 3.20 E + 16 B4105-3 Inter Shell Plate B4106-1 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 B4106-2 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 B4106-3 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 Lower Shell Plate B4107-1 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 B4107-2 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 B4107-3 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 Upper Shell Long(")

Welds 1-442 A,B,C 1.59 E + 17 1.54 E + 17 9.00 E + 16 3.20 E + 16 Upper Shell to Inter Shell Weld 8-442(a) 1.59 E + 17 1.54 E + 17 9.00 E + 16 3.20 E + 16 Inter Shell Long Welds 2-442 A,B 5.31 E + 18 5.16 E + 18 3.00 E + 18 1.07 E + 18 Weld 2-442 C 2.74 E + 18 2.66 E + 18 1.55 E + 18 5.50 E + 17 Inter Shell to Lower Shell Weld 9-442 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 Lower Shell Long Welds 3-442 A,B 4.24 E + 18 4.12 E + 18 2.40 E + 18 8.50 E + 17 Weld 3-442 C 7.68 E + 18 7.46 E + 18 4.34 E + 18 1.54 E + 18 Calc N-NCM-97009, Calc. 921130-0

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

1GRSATO2.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 27 OF 30 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, 1/4/t and 3At Locations at 16 EFPY Material Fluence f, Fluence fc,. Fluence f,, Fluence f,/.,

Upper Shell Plate")

B5453-1 1.57 E+ 17 1.53 E+ 17 8.90 E+ 16 3.10E + 16 B5453-3 1.57 E + 17 1.53 E + 17 8.90 E + 16 3.10 E + 16 B5011-IR 1.57 E + 17 1.53 E + 17 8.90 E + 16 3.10E + 16 Inter Shell Plate B5454-1 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 B5454-2 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 B5454-3 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 Lower Shell Plate B5455-1 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 B5455-2 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 B5455-3 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 Upper Shell Long"a)

Welds 1-201 A,B,C 1.57 E + 17 1.53 E + 17 8.90 E + 16 3.10 E + 16 Upper Shell to Inter Shell Weld 8-201"a) 1.57 E + 17 1.53 E + 17 8.90 E + 16 3.10 E + 16 Inter Shell Long Weld 2-201 A 3.81 E + 18 3.70 E + 18 2.15 E + 18 7.70 E + 17 Weldsc* 2-201 B, C 7.79 E + 18 7.57 E + 18 4.40 E + 18 1.56 E + 18 Inter Shell to Lower Shell Weld 9-201 7.58 E + 18 7.37 E + 18 4.29 E + 18 1.52 E + 18 Lower Shell Long Welds°') 3-201 A,C 7.79 E + 18 7.57 E + 18 4.40 E + 18 1.56 E + 18 Weld 3-201 B 3.81 E + 18 3.70 E + 18 2.15 E + 18 7.70 E + 17 Calc N-NCM-97009, Calc. 921130-0 (a) Upper shell materials are included for completeness since EOL exposure is expected to exceed 1.OE + 17.

(b) Fluence unreduced by neutron pads.

IGRSAT02.doa 04B 0710.1218 I

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-I DIABLO CANYON POWER PLANT REVISION 2 PAGE 28 OF 30 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 'At and 3at Locations for 16 EFPY 16 EFPY ART __)

Material RG 1.99 Rev. 2

'At (OF) /3/4t (OF)

_Method Upper Shell Plate'd)

B4105-1 Position 1.1 71.4 66.2 B4105-2 Position 1.1 52.4 47.2 B4105-3 Position 1.1 59.5 53.0 Inter Shell Plate B4106-1 Position 1.1 89.5 67.4 B4106-2 Position 1.1 93.2 72.2 B4106-3 Position 1.1 120.5 106.2 Lower Shell Plate B4107-1 Position 1.1 118.0 94.7 B4107-2 Position 1.1 117.1 95.8 B4107-3 Position 1.1 74.5 53.4 Upper Shell Long"')

Welds 1-442 A,B,C Position 1.1 24.0 0.7 Upper Shell to Interd)

Shell Weld 8-442 Position 1. 1 3.7 -11.3 Inter Shell Long Welds 2-442 A,B Position 1.1 161.5 107.2 Weld 2-442 C Position 1.1 125.2 79.4 Inter Shell to Lower Shell Weld 9-442 Position 1.1 141.8 97.1 Lower Shell Long Welds 3-442 A,B Position 1.1 148.7 96.9 Weld 3-442 C"') Position 1.1 183 .7Wb) 124.9 A

Calc N-NCM-97009 & Calc. N-282 (a) ART = Initial RTNDT + ARTNDT + Margin (*F) I Wb) This ART value is used to generate the heatup and cooldown curves.

(0 DCPP-1 Surveillance Capsule S was not judged "credible" per 10CFR50.61. 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 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 29 OF 30 TITLE: PTLR for Diablo Canyon UNITS 1 AND 2 TABLE 6.0-9 Diablo Canyon Unit 2 Adjusted Reference 1 Temperatures (ARTs) for the Reactor Vessel Beltline Materiak* at the /at and 3At I aw-tinn* fnr 1* 17.VPV 16 EFPY ART)' _

Material RG 1.99 Rev. 2 /4t (0 F) 3 1/4t (-F)

Method Upper Shell Plate()*

B5453-1 Position 1.1 43.1 35.0 B5453-3 Position 1.1 47.3 42.6 B5011-1R Position 1.1 42.4 37.7 Inter Shell Plate B5454-1 Position 2.1 147.4 120.8 B5454-2 Position 1.1 177.2 151. 4 (b)

B5454-3 Position 1.1 151.5 122.9 Lower Shell Plate B5455-1 Position 1.1 94.1 68.7 B5455-2 Position 1.1 109.1 83.7 B5455-3 Position 1.1 98.9 82.0 Upper Shell Long*')

Welds 1-201 A,B,C Position 2.1 -7.2 -30.1 Upper Shell to Inter(c)

Shell Weld 8-201 Position 1.1 -0.4 -13.0 Inter Shell Long Weld 2-201 A Position 2.1 102.2 55.6 Welds 2-201 B, C Position 2.1 141.4 86.3 Inter Shell to Lower Shell Weld 9-201 Position 1.1 12.8 -0.7 Lower Shell Long Welds 3-201 A,C Position 1.1 107.0 74.1 Weld 3-201 B Position 1.1 83.6 55.8 Calc N-NCM-97009 & Calc N-282

"(a) ART = Initial RTmDT + ARTNDT + Margin (1F)

(b) This ART value is used to generate the heatup and cooldown curves.

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

1GRSAT02.doa 04B 0710.1218

PACIFIC GAS AND ELECTRIC COMPANY NUMBER PTLR-1 DIABLO CANYON POWER PLANT REVISION 2 PAGE 30 OF 30 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 /4t~dt Chemistry Factor, CF (°F) 226.8"° 99.6 Fluence + 10"9 n/cm 2 (E > 1.0 MeV), f') 0.434 0.152 Fluence Factor, FF1b) 0.768 0.5058 ARTNDT = CF x FF, (OF) 174.2") 50.4 Initial RTNDT, I ('F) -56 67 Margin, M (°F)(') 65.5 34 ART = I + (CF x FF) + M (0 F) 183.7() 151.4 per Regulatory Guide 1.99, Rev. 2 Calc N-NCM-97009

() Fluence, f, is based upon f%, and f,, from Tables 6.0-6 and 6.0-7. The Diablo Canyon reactor vessel wall thickness is 8.625 inches at the beltline region.

(b) Fluence Factor (FF) per Regulatory Guide 1.99, Revision 2, is defined as FF = f(o.2 -o.,0-o0.

(c) Margin is calculated as M = 2(r,2 + CA2)o5. The standard deviation for the initial RTNDT margin term ar, is 0°F for plate since the initial RT1JDT is a measured value. The standard deviation for ARTNDT term aA, is 17°F for the plate, except that oA 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 1/t.

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

(0 DCPP-1 Surveillance Capsule S was not judged "credible" per 10CFRS0.61. The higher chemistry value of CE NPSD-1039, Rev 2 for this heat are used to generate the heatup and cooldown Appendix G curves.

IGRSATO2.doa 04B 0710.1218

Enclosure 5 PG&E Letter DCL-02-079 Calculation STA-138

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )l&2 SEE. NO. 0 SHEET NO. 1 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 TABLE of CONTENTS

1. PURPOSE 2

2. BACKGROUND 2

3. ASSUMPTIONS 2 3.1. General LTOP Methodology 2 3.2. Westinghouse Benchmark 3 3.3. Mass Input Evaluation 3 3.4. Heat Input Evaluation 4

4. DESIGN INPUT 4

5. METHODOLOGY 5 5.1. Develop RETRAN LTOP Model 5 5.2. RETRAN LTOP Model Benchmark 5 5.3. RETRAN Evaluation of DCPP LTOP Setpoints 5 5.4. LTOP Administrative Limits 6

6. ACCEPTANCE CRITERIA 6 6.1. RETRAN LTOP Model Benchmark 6 6.2. RETRAN Evaluation of DCPP LTOP Setpoints 6 6.3. DCPP LTOP Administrative Limits 7

7. CALCULATION 7 7.1. Develop RETRAN LTOP Model 7 7.2. RETRAN LTOP Model Benchmark 9 7.3. RETRAN Evaluation of DCPP LTOP Setpoints 12 7.4. LTOP Administrative Limits 16

8. RESULTS 18 8.1. RETRAN LTOP Model Benchmark 18 8.2. RETRAN Evaluation of DCPP LTOP Setpoints 19 8.3. LTOP Administrative Temperature Limits 20

9. CONCLUSIONS 20 9.1. RETRAN LTOP Model Benchmark 20 9.2. RETRAN Evaluation of DCPP LTOP Setpoints 20 9.3. LTOP Administrative Temperature Limits 20

10. IMPACT EVALUATION 21

11. REFERENCES 22 TABLES 24 FIGURES 43

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 2 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01

1. PURPOSE This calculation uses the RETRAN-02 Mod4 (RETRAN) computer code (Ref. 17) model of DCPP to evaluate the acceptability of the current LTOP setpoints for the applicable range of heat input and mass input cases consistent with the NRC approved LTOP methodology established in WCAP 14040 (Ref. I).

This calculation will establish the bases for the DCPP LTOP setpoint methodology in the Pressure Temperature Limits Report (PTLR) submittal to the NRC.

2. BACKGROUND In the process of answering NRC questions on the proposed LTOP evaluation methodology for the PTLR it became evident that DCPP needed to update several LTOP related evaluations to ensure consistency with the applicable Reference I Westinghouse methodology in WCAP 14040.

This calcnote forms the technical basis for the DCPP LTOP methodology within the PTLR license amendment. Therefore, it has been structured such that the main body contains the information and level of detail appropriate for direct compilation into a LAR, while the explicit details and documentation necessary for an independent technical verification have been provided in STA- 145 (Ref. 23).

The RETRAN code is a versatile thermal hydraulic computer code developed by EPRI for the purpose of analyzing various PWR and BWR transients. The NRC has reviewed the RETRAN code and issued an SER approving it for analyzing certain transients as delineated in NRC regulations. DCPP has already received NRC approval for use of the RETRAN code in analyzing the Loss of Load (LOL) event (calcnote N-098, Ref. 2) which was submitted as part of the License Amendment Request (LAR) 95-06 (Ref. 3) to revise the Main Steam Safety Valve (MSSV) setpoint tolerance. This analysis uses the same DCPP RETRAN model and code version which has already received NRC acceptance for use in analyzing the Loss of External Electrical Load and/or Turbine Trip (LOL/TT) event (calcnote N-098, Ref. 2). In addition, this RETRAN analysis of the LTOP event has been verified to be consistent with the applicable restrictions and conditions of the latest RETRAN SER which includes the original Mod 4 SER condition responses provided by DCPP in Attachment E.1 to LAR 95-06 (Ref. 3) This calculation will establish the bases for the DCPP LTOP setpoint methodology in the Pressure Temperature Limits Report (PTLR) submittal to the NRC.

3. ASSUMPTIONS To ensure conservative results consistent with the WCAP 14040 (Ref. 1) methodology, the following conservative assumptions are used in this calculation.

3.1. General L TOP Methodology I. The RCS including the pressurizer is initially at steady state water solid conditions.

2. There is no credit or modeling of any RCS metal expansion or heat transfer during the LTOP pressure transient.

3. Since the water inertia reduces the flow at the beginning of the mass injection LTOP event, it is conservative to neglect the inertia effect. The charging flow is determined by steady state flow calculation.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 3 OF 54 SHEET NO.

SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01

4. The pressurizer PORV flow is minimized by assuming the Pressurizer Relief Tank is at the maximum design discharge pressure of 50 psig per section 4.3.3.5.c of Ref. 22.

5. The mass injection and heat injection cases are all evaluated assuming that two RHR pumps are operating with a conservative maximum flow of 5000 gpm each. The RHR system is modeled such that there is no net effect on RCS mass injection or heat removal but the RHR flow does maximize the dynamic pressure drop across the RCS. This is conservatively bounding since the WCAP 14040 methodology assumes that the RHR system is isolated such that RHR heat removal and relief capability is not available.

6. The maximum RCS pressure overshoot is determined based on a limiting single failure for one of the two LTOP system PORVs fails to open.

7. Since both LTOP PORVs have the same nominal lift and reset setpoints, the minimum RCS pressure undershoot is determined based on both LTOP system PORVS opening and closing at the same time.

8. The PORV LTOP setpoint pressure uncertainty is assumed to be 32 psi to bound Design Input 5.

9. The LTOP RCS wide range temperature uncertainty is assumed equal to Design Input 6 (15 °F).

10. The SG secondary temperature uncertainty is assumed 15 'F to bound Design Input 9.

11. The pressurizer level uncertainty is assumed to be 15% to bound design input 12.

12. The maximum normal injection fluid temperature is assumed to be 100 'F consistent with the Westinghouse LTOP analyses for DCPP in Ref. 9.

13. The minimum acceptable RCS undershoot is verified by ensuring the RCP volume pressure does not decrease below 235 psig. This is based on the minimum 200 psid required across the number one RCP seal (Ref. 6) and a VCT backpressure of 35 psig which is the maximum of the normal operating range of 15-35 psig. (Ref. 6).

14. The RCP seal pressure is best represented as the average of the RCP volume and the RCP discharge or Cold Leg Volume pressures calculated by RETRAN. This is based on the RCP seal balancing chamber pressure increase during RCP flow conditions as established in Reference 21.

3.2. Westinghouse Benchmark

1. None.

3.3. Mass Input Evaluation I. The initial RCS pressurizer pressure is assumed to be between 300 psig and 400 psig as necessary to ensure subcooled liquid conditions at the RCS temperature and prevent cavitation code errors due to the conservatively fast start time of the RCP. The initial RCS pressure does not affect the transient since there is sufficient time to evaluate the transient response prior to reaching the PORV lift setpoint of 435 psig.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )l&2 I~ REV. NO.

~~SHEET NO.

0 4 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 3.4. Heat Input Evaluation

1. The RCS is assumed to have been cooled down to a steady state condition via RHR flow such that the RCS liquid in the SG tubes, the SG tube and shell, and the SG secondary liquid are all at a higher temperature than the remainder of the RCS.

2. The temperature difference between the RCS and the SG secondary side is assumed to be at the Tech Spec maximum plus uncertainty when an RCP is started in one loop.

3. The specific heat capacity of the SG tubes is increased by a factor of 1.5 to conservatively maximize the amount of metal heat transferred to the RCS.

4. The thermal conductivity of the SG tubes is increased by a factor of 10 to conservatively maximize the effective secondary side heat transfer efficiency to the RCS.

5. The SG secondary side temperature is manually measured using a WAHL Model 392 Digital Thermometer or an instrument of equivalent accuracy.

4. DESIGN INPUT

1. The water properties are from ASME Steam Tables (Ref. 16).

2. The PORV LTOP setpoint is 435 psig per the DCPP PTLR (Ref. 7)

3. The maximum PORV stroke time is 2.9 seconds per DCPP PTLR (Ref. 7)

4. The total PORV LTOP actuation time delay is 1.5 seconds, which consists of a 1.05 seconds electronic delay, and a 0.45 second process and pneumatic delay. The process and pneumatic delay are part of the PORV stroke time acceptance criteria in Design Input 3. (Ref. 19)

5. The PORV LTOP Pressure uncertainty is 31.7 psi (Ref. 13)

6. The LTOP RCS wide range temperature uncertainty is 15 'F (Ref. 14)

7. The charging injection flow versus RCS pressure bounds a Cv of 26 for both FCV-128 and HCV-142 as established in STA-143 Rev. 0 (Ref. 5)

8. SG Wide Range Pressure Uncertainty = 33.7 psi (Ref. 12)

9. WAHL Model 392 Digital Thermometer Accuracy at 250 'F is +/- 3.4 'F and at 400 'F is

+ 5 'F (Ref. 18).

10. The ASME Appendix G steady state pressure limit (including a 10% limit relaxation per ASME Code Case 514 ) as established in PTLR Table 2.1-2 (Ref. 7). Here after in this report, the term Appendix G P/T curve limit is meant to include the 10% relaxation per ASME Case 514.

11. Not Used.

12. The pressurizer level indication uncertainty is + 6.1% (Ref.. 20)

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 SHEET NO. 5 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01

13. The 95% probability value for both PORVs drifting low simultaneously is 12.4 psi as established in STA-145. (Ref. 23)

5. METHODOLOGY 5.1. Develop RETRAN L TOP Model The DCPP LTOP analysis model is developed directly from the RETRAN-02 Mod3 (RETRAN) model of DCPP which analyzed the Loss of Load (LOL) event in calcnote N-098 (Ref. 2) and was submitted as part of the License Amendment Request (LAR) 95-06 (Ref. 3) to revise the Main Steam Safety Valve (MSSV) setpoint tolerance. This RETRAN model was demonstrated to accurately model the DCPP primary to secondary thermal hydraulic behavior in benchmark comparisons to DCPP start up tests, particularly the turbine trip from full power. This RETRAN model and the associated analysis results were accepted by the NRC for licensing applications and were incorporated into the DCPP Unit I and Unit 2 design basis per the License Amendments (LAs) 108 and 107, respectively.

The vast majority of the Ref. 2 RETRAN LOL Model has not been changed. The changes implemented to model the LTOP analyses are considered superficial in that they do not impact the basic thermal hydraulic performance of the RETRAN LOL model. These changes can be classified into three types. The first type involves simplifying changes to eliminate unnecessary modeling options and functions which are not needed for the LTOP analysis. These include eliminating the core neutronics and heat conductors models, all reactor protection and ECCS trip functions, the non equilibrium pressurizer model, and eliminating all secondary side volumes and components except for the steam generators. The second type of changes involve expanding the RETRAN DCPP model into four individual RCS loops and adding the components specifically required to evaluate the asymmetric LTOP mass input and heat input scenarios. The third type involves minor adjustments ID the RETRAN DCPP model for benchmarking to generic Westinghouse LOFTRAN code results and for establishing the conservatively bounding analysis assumptions specific to the DCPP LTOP methodology.

5.2. RETRAN LTOP Model Benchmark The DCPP RETRAN model is adjusted as necessary to provide a direct comparison of LTOP analysis results to comparable results obtained with the Westinghouse LOFTRAN model. The RETRAN model is used to generate RCS pressure overshoot and undershoot results for heat input and mass input scenarios. The comparisons cover a wide range of RCS conditions, heat addition rates, mass addition rates, and PORV actuation parameters. The benchmark results establish that the RETRAN model generates consistent thermal hydraulic results and is acceptable for use in evaluating the DCPP LTOP setpoints per the NRC approved methodology established in WCAP 14040 (Ref. 1).

5.3. RETRAN Evaluation of DCPPLTOP Setpoints The RETRAN model is used to evaluate the RCS response to an appropriate range of mass input and heat events using the applicable DCPP LTOP setpoints. The mass addition scenarios assume that

I Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 6 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 charging flow is injecting into a water solid RCS when the letdown flow is inadvertently isolated.

The RETRAN model is used to determine the RCS overshoot for a range of mass addition rates and for the various dynamic pressure drop effects associated with RCP and RHR pump operation.

The heat addition scenarios assume that an RCP is started in one loop, and the water solid RCS is 80 'F cooler than the steam generators (SGs) and RCS fluid within the tube bundle region. The RETRAN model is used to determine the RCS overpressure which occurs due to thermal expansion of the RCS fluid as it is heated by the SGs. Since the thermal expansion properties of water vary significantly with temperature, the heat addition scenario is evaluated over the full range of LTOP applicable temperatures. The LTOP PORV parameters evaluated for the impact on RCS overshoot include the lift setpoint, valve stroke time, flow capacity, instrument uncertainty, and electronic delays.

The RETRAN LTOP model is used to evaluate the RCS undershoot which could occur during a mass injection or heat injection event based on the DCPP PORV closing characteristics. This RCS undershoot section evaluates the minimum RCS pressure during a LTOP event with respect to maintaining an adequate operational pressure drop across the +number one RCP seal.

The RETRAN LTOP model is also used to establish that with the pressurizer level at 50% or less, an RCP may be started without temperature restrictions, since even the most limiting heat input transient will not challenge the Appendix G P/T curve limit.

5.4. L TOP Administrative Limits These LTOP analysis results establish that the DCPP LTOP setpoints in the PTLR ensure the maximum RCS pressure overshoot for the applicable range of mass input and heat events remains below the Appendix G steady state Pressure limits as allowed per the ASME Code Case N-514.

These results also establish the appropriate administrative controls to ensure that the actual DCPP operating conditions remain bounded by the range of RCS conditions, injection flow capability, and RCP operation assumed in the LTOP analyses.

6. ACCEPTANCE CRITERIA 6.1. RETRAN L TOP Model Benchmark There are no explicit or numerical acceptance criteria for the RETRAN benchmark evaluation.

However, a qualitative review shall conclude that the RETRAN DCPP model generates thermal hydraulic results consistent in both characteristic trend and magnitude compared to the available Westinghouse LOFTRAN data.

6.2. RETRAN Evaluation of DCPPL TOP Setpoints

1. The maximum RCS pressure (overshoot) shall not exceed the ASME Appendix G steady state pressure limit (including a 10% limit relaxation per ASME Code Case 514 ) as established in PTLR Table 2.1-2 (Ref. 7) for the applicable range of LTOP RCS temperatures, when accounting for instrument uncertainty.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )&2 SHEET NO. 7 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26 /01 CHK'D BY Dixon Yee DATE 11/26/01

2. In order to maintain appropriate operation of the number one RCP seal, the minimum RCS pressure at the RCP (undershoot) should not decrease below 235 psig as identified in assumption 13.

3. With the pressurizer level less than or equal to 50% narrow range, the worst case heat input due to the start of an RCP in one loop will not challenge the Appendix G P/T curve limit.

6.3. DCPPL TOP Administrative Limits

1. The Administrative Controls shall be implemented as necessary on RCS injection capability, RCS pressure relief capacity, and RCP operation to ensure that the actual plant configuration remains conservatively bounded with respect to the plant performance assumed in the LTOP analyses.

7. CALCULATION 7.1. Develop RETRAN L TOP Model 7.1.1. Base RETRAN LTOP Model for DCPP NRC Approved DCPP RETRAN LOL Model The DCPP LTOP analysis model is developed directly from the RETRAN model of DCPP, which analyzed the Loss of Load (LOL) event in calcnote N-098 (Ref. 2) and was submitted and approved per LAR 95-06 (Ref. 3) to revise the MSSV setpoint tolerance. The Ref. 2 model consisted of a single RCS loop and corresponding steam generator (SG), while the other three RCS loops were combined into one lumped RCS loop and combined SG. This methodology of using a lumped RCS loop is common in transient analysis when evaluating symmetric events. By scaling the appropriate physical and hydraulic parameters, the computer model is simplified to run faster, while preserving the accuracy of the overall thermal hydraulic plant response.

The conservative LTOP events as analyzed per WCAP-14040 (Ref. 1) involve subcooled water solid, and essentially isothermal shutdown conditions. The LTOP event is a much less complex thermal hydraulic event than the design basis LOL transient which occurs at dynamic full power conditions.

Therefore, the LTOP model can be made much simpler and does not require modeling numerous options and functions which are necessary for the LOL analysis model. Unnecessary model options which were eliminated for the LTOP model include the core neutronic and heat transfer models, the two region non-equilibrium pressurizer model, all reactor protection trips and ESF actuation functions, and all plant control system models. In addition, since the primary to secondary heat transfer is either ignored or conservatively controlled in the LTOP analysis, all of the secondary volumes except the SG themselves were eliminated from the LTOP model (i.e., steam lines, main steam isolation valves, condenser, steam dumps, etc.). This base LTOP model was originally documented in PG&E calculation STA-121 (Ref. 8).

I Pacilic Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 8 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 Reactor Coolant System The LTOP analysis is not a symmetric event, and requires evaluating any combination of RCP flow in each of the four RCS loops. Therefore, the lumped RCS loop of the Ref. 2 model was expanded into three individual loops, each with a corresponding steam generator and RCP. This expansion in the LTOP base model was documented in STA-121 (Ref. 8) and is considered cosmetic since all four RCS loops have the exact same dimensions and hydraulic properties as the existing single RCS loop in the LOL model. The LTOP model also uses the exact same nodalization, physical dimensions, and hydraulic parameters for the reactor vessel volumes as in the LOL model. Figure 6-1 shows the nodalization diagram of the DCPP LTOP four loop model. Tables 6-1 and 6-2 summarize the key RCS volume and junction parameters, respectively.

Reactor Coolant Pumps The input data tables used for modeling the RCPs include the same pump head and torque curve data for the Westinghouse Model 93A1 RCP that was used in the Ref. 2 LOL model. However, the LTOP model also included a minor change to allow initializing the model with no RCPs running and then start them as needed for the various cases analyzed. Each RCP is assigned a trip which allows it to be started as needed. An additional RETRAN pump curve data table is provided to establish the pump rpm as a function of time during the startup sequence. As discussed in the evaluation sections, an appropriately conservative pump startup time is used for the heat addition evaluations. The pump startup time does not impact the mass input evaluation results since this LTOP transient is initiated from steady state RCS flow conditions with the designated number of RCPs already running.

Pressurizer Model As discussed earlier, the limiting LTOP events are analyzed assuming a water solid and subcooled RCS conditions such that the two region non-equilibrium RETRAN pressurizer model is not needed.

In addition, the pressurizer heater and spray models including associated junctions and control systems are not needed, and were removed from the DCPP LOL RETRAN model. The LTOP model did require some additional PORV modeling options specific to the LTOP analysis. This included modifying the nominal trip setpoint, modifying the junction flow loss coefficient to generate the appropriate PORV liquid relief capacity. In addition, general data tables were added to model the PORV valve area as a function of the opening and closing times.

RHR, Charging, and RCP Seal Injection The LTOP model includes five additional RETRAN junctions to model the RHR letdown from the Loop 4 hot leg and RHR injection to each of the four individual loop cold legs, respectively. Normal RHR injection is actually only provided to the Loops I and 2 cold legs. However, this nodalization allows for future model flexibility in modeling ECCS cold leg injection flow if desired. Since there is no credit for any RHR heat removal in the LTOP analysis, the RHR system is modeled to inject liquid at the same enthalpy as is removed from the letdown volume. The only significant effect related to modeling the RHR flow is the impact on the total RCS flow through the core. This establishes the maximum associated dynamic pressure drop between the LTOP pressure sensor and the peak pressure location at the bottom of the RCS vessel. The total RCS pressure drop is not

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 SHEET NO. 9 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 sensitive to which cold legs receive RHR injection, but is only the total flow rate. The LTOP analysis assumes a conservatively bounding RHR flow rate of 5000 gpm per pump.

The LTOP model also includes four specific RETRAN junctions for modeling the combination of normal charging, alternate charging, and RCP seal injection flow to the four RCP loop cold legs.

These junctions are also used to model the charging injection flow though the ECCS injection path.

The flow rate into the RCS through the various injection junctions is established by using fill tables are which specify the injection flow rate as a function of RCS pressure. The injection flow rates that used are specifically identified and discussed for each appropriate analysis section.

Steam Generator The only other significant RETRAN model change for the LTOP analysis is the simplification of the multiple volume two phase steam generator model used for the LOL analysis, into a homogenous single volume steam generator with a bubble rise model. As stated earlier, the LTOP analysis does not credit any primary to secondary heat transfer in the mass input cases and the heat input cases only model reverse heat transfer from the SG to the RCS during shutdown conditions. The single volume SG model and the appropriately conservative RETRAN heat transfer characteristics are consistent with the Westinghouse LTOP methodology in WCAP 14040, as established in the benchmark studies of Section 6.2.

The RETRAN nodalization and input parameters used to model the SG heat conductors are the same as was used in the Ref. 2 LOL model. Table 6-3 lists the key RETRAN input data for the SG heat conductors, which are calculated from the geometry data for the Series 51 SGs at DCPP. The Series 51 SGs have 3388 tubes each with an inner diameter of 0.775 inches and an outer tube diameter of 0.875 inches. The total conductor left and right side heat transfer surface areas and the total conductor volume are also provided in Table 6-3. These LTOP heat conductors use the Inconel 600 thermal conductivity and specific heat properties as established in the Ref. 2 LOL deck.

7.2. RETRAN LTOP Model Benchmark The DCPP RETRAN LTOP model is benchmarked based on generating thermal hydraulic results, which are comparable to the West. LOFTRAN model used for the NRC approved methodology established in WCAP 14040 (Ref. 1).

7.2.1. Benchmark RETRAN Mass Input 7.2.1.1.RETRAN Mass Input Benchmark Model The Ref. 9 report (PGE-88-642 dated July 7, 1988) documents an extensive Westinghouse LOFTRAN parametric study of PORV setpoints, PORV stroke times, and mass injection rates, which establish the licensing basis for the current DCPP LTOP setpoints. Table 6-4 summarizes the key LTOP input parameters used by Westinghouse to evaluate the RCS overshoot and undershoot for various mass input events in the Ref. 9 report. The DCPP LTOP model developed in the previous section was modified slightly to more closely match the Westinghouse LOFTRAN model and the

L Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )l&2 REV. NO. 0 SHEET NO. 10 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 range of input assumptions such that the results of the two models can be directly compared. These input changes are summarized below.

RCS Volume As identified in Table 6-4, Westinghouse assumed an RCS volume of 12732 ft3 for both DCPP Units I and 2. However, the RETRAN LTOP model is based on the more conservative LOL model which used an RCS volume of 11700 ft3. Since the RCS pressure response is sensitive to the total RCS volume, the RETRAN benchmark model had the RCS volume increased by about 700 ft3 to more closely match the Westinghouse LOFTRAN model in Ref. 9. The key hydraulic parameters associated with the pressurizer and surge line have not been changed and are not impacted by this increase in volume. Since the mass injection event occurs at water solid, isothermal conditions, the RCS response is essentially a function of the total RCS volume, mass injection rate, and the PORV relief characteristics. It should be noted that a smaller RCS volume results in more limiting RCS pressure overshoot and undershoot results. Therefore, the original RETRAN RCS volume from the LOL analysis is conservatively bounding and is used for evaluating the DCPP LTOP setpoints in Section 6-4.

Pressurizer PORV The Ref. 9 Westinghouse LTOP analyses assume that the PORV valve position changes linearly as a function of stroke time for both the open and close cycles. Combining the linear valve stroke with the valve flow coefficient (Cv) as a function of valve position, results in the normalized PORV Cv value versus valve position data used in the Westinghouse LTOP analyses and shown in Table 6-5.

The RETRAN model uses these PORV flow and stroke characteristics for the benchmark comparisons. The general data tables number one and two were used in the RETRAN model to implement the PORV valve area versus time for the opening and closing sequence, respectively. The RETRAN trip numbers 8 and 9 were used to open the two PORVs, respectively, based on the specified pressure setpoint as measured in the hot leg volume 306. These trips were assigned a delay time of 1.1 seconds to match the West. analyses. It should be noted that since the Westinghouse LTOP analyses assume a limiting single failure of one PORV to open, one of the RETRAN PORV trips is assigned a very long delay time which prevents it from opening during the transient. The Westinghouse analyses assumed a constant PORV closure time of 2.0 seconds, as did the RETRAN benchmark model.

RCS Flow West. does not model the dynamic pressure drop effects due to RCS flow in the overshoot and undershoot cases in Ref. 9 (DCPP evaluated these effects separately in a subsequent calculation).

Therefore, the RCP and RHR pump trips were set to 1.0E6 seconds to ensure they did not run during the mass injection benchmark cases.

Initial RCS Conditions The RETRAN RCS and SG volumes were set to isothermal conditions of 200 psig and 100 'F to match the Westinghouse LOFTRAN model.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 IS Project: Diablo Canyon Unit ( )I ( )2 (X )1&2 REE. NO.

SHEET NO. 0 11 OF 54 1i4 SUB JECT RETRAN Evaluation of DCPP LTOP Parameters MAD)EBY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 7.2.2. Benchmark RETRAN Heat Input 7.2.2.1.RETRAN Heat Input Benchmark Model The Westinghouse heat input LTOP analyses, which establish the current DCPP license basis, are based on a generic LOFTRAN plant model that has been evaluated to be conservatively bounding for the DCPP parameters. These Westinghouse heat input results are documented in the Reference 10 report. Table 6-7 summarizes the key LOFTRAN model input parameters which Westinghouse used to generate the heat input results in Ref. 10. In order to perform an appropriate benchmark comparison, the DCPP RETRAN LTOP model is modified slightly to match the generic Westinghouse generic model as summarized below.

Initial RCS Conditions The Westinghouse heat input methodology and model are based on the starting of one RCP with the RCS liquid volume at a lower temperature than the SGs. The specific assumption is that the SG secondary liquid, SG tubes and shell, and RCS liquid inside the tubes are all at a higher temperature than the rest of the RCS. In order to model this conservative but physically unrealistic assumption, the RETRAN heat input model was modified to have a valve on the inlet and outlet of each SG tube bundle. The RETRAN model was then initialized with these valves closed such that the RCS liquid temperature within the SG tubes can be set to a value the same as the SG tubes and secondary liquid, and independently higher than the rest of the RCS. Once the RCP is started, the valves are very quickly opened (in less than 0.1 seconds) so that there is no delay in the RCS mass flow or the heat transfer to the RCS.

RCS Volume As listed in Table 6-7, the Westinghouse generic heat input model used an RCS volume of 13,000 Wt.

The DCPP RETRAN model has an RCS volume of 11700 WV. Since the RCS pressure increase response is sensitive to the total RCS volume, an additional 1300 Wt was added to the DCPP RETRAN model. This additional volume was included in the reactor head volume. Since very little mixing occurs in this volume, there is a minimal impact on the calculated heat conduction and volumetric expansion within the RCS loops and steam generator tube volumes. In addition, the Ref.

10 report indicates that the Westinghouse study used an artificially small pressurizer volume of only 100 ft3 to conservatively offset the LOFTRAN limitations related to modeling the pressurizer with subcooled liquid conditions. Although, RETRAN has no such limitation, the RETRAN heat input bench mark model also used a pressurizer volume of 100 ftto maintain consistency with the Westinghouse model. As in the Westinghouse model, the 1700 W (1800 - 100) removed from the pressurizer volume was relocated to an inactive RCS vessel volume to mininize any impact on the RCS / SG heat transfer rate. Thus, the RETRAN upper head volume was increased by a total of 3000 ft 3 to account for these two differences.

SG Heat Transfer Table 6-7 shows that the Westinghouse generic heat input model used a total steam generator 2

secondary side heat transfer surface of 58,000 ft . As shown in Table 6-3, the DCPP RETRAN model is based on a Series 51 SG with an effective area of about 51, 500 W. Therefore, the RETRAN SG heat conductor areas and volume were increased by the appropriate ratio to be

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 12 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 equivalent to the Westinghouse model. The LOFTRAN model assumed that the SG liquid inventory was at a conservatively maximum level and that all secondary liquid (not just the liquid near the tube bundles) was directly available for heat transfer to the RCS. The single volume SG in the RETRAN was similarly modeled. The original Westinghouse generic heat input analysis calculated a conservative RCS to SG heat transfer rate based only on the secondary free convection heat transfer coefficient in the SG. Westinghouse did not model the thermal conductance through the SG tubes, or the RCS forced convection on the inside of the tubes. The DCPP RETRAN model explicitly models all of the physical SG heat transfer phenomena. In order to more closely match the Westinghouse generic model for the benchmark study, the RETRAN SG heat transfer properties were adjusted to provide the appropriate overall heat transfer of energy from the SG into the RCS. This included increasing the volumetric heat capacity of the Inconel SG tubes by a factor of 1.5 and increasing the thermal conductivity by a factor often. These adjustments to the RETRAN heat transfer model effectively reproduce the Westinghouse results and conservatively bound the maximum potential secondary to RCS heat transfer. These conservative adjustments to the RETRAN heat transfer model are incorporated into the DCPP LTOP model.

Pressurizer PORV Model The RETRAN PORV model was revised to match the Westinghouse reference PORV model parameters as shown in Table 6-7. This reference pressurizer PORV model has a valve coefficient of Cv = 50 gpm/ psi' 2 which varies linearly with the valve stroke position and time. This reference PORV has 3 second opening time which consists of a 0.6 second delay and a 2.4 second valve stroke time. (Ref. 10) 7.3. RETRAN Evaluation of DCPPL TOP Setpoints The DCPP RETRAN model has been demonstrated to effectively model both the mass input and heat input analyses consistent with the WCAP 14040 methodology. This section summarizes the DCPP RETRAN evaluation of the DCPP LTOP setpoints established in the PTLR, which ensures that they provide adequate protection for the range of mass injection and heat injection scenarios consistent with the WCAP 14040 methodology. Table 6-9 lists the current DCPP LTOP setpoints and the other key LTOP related parameters, which are used in the evaluation. As discussed in Assumption 2.1.5, all of the mass injection and heat injection cases are evaluated assuming that two RHR pumps are operating at 5000 gpm each. This conservatively maximizes the RCS dynamic pressure drop across the RCS and the resultant effect on the peak pressure at the bottom of the RCS vessel. The PORV opening and closing sequence including delays is shown in Figure 6-3.

7.3.1. RETRAN Evaluation of LTOP Mass Input Overshoot As identified in the DCPP PTLR, the Safety Injection (SI) pumps and one Centrifugal Charging Pump (CCP) are secured prior to entering the LTOP range. However, the SI signal and charging ECCS valves are still in service. Therefore, the most limiting mass injection case in the LTOP range assumes that one CCP and the Positive Displacement Charging Pump (PDP) are injecting through the ECCS flow path and through the RCP seal injection flow to all four RCPs. These injection assumptions are conservative since a Safety Injection signal would isolate the normal charging path except for the RCP seal injection. The total mass injection flow rates as a function of the RCS pressure for this ECCS injection case are listed in Table 6-10. These flow rates were calculated in

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 IS Project: Diablo Canyon Unit ( )1 ( )2 ( X )I&2 SHEET NO. 13 OF 54 SUB JECT RETRAN Evaluation of DCPP LTOP Parameters MAD 'E BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11 /26/01 Ref. 5 based on a conservatively bounding combination of maximum pump performance curves and minimum system line resistances. The ECCS flow and RCP seal injection flow is evenly distributed to each of the four RCS loops. The peak pressure results from this ECCS injection case establish the minimum administrative RCS temperature limit at which the charging system CCS flow path must be blocked.

Once the charging system CCS flow path is blocked, the most limiting mass injection case is one CCP and the PDP injecting simultaneously through both the normal and the alternate charging paths.

This is conservative since DCPP plant procedures specify that only one charging injection path is used at a time. Table 6-11 summarizes the mass injection flow rates versus RCS pressure for the CCP and the PDP. These flow rates were calculated in Ref. 5 based on a bounding combination of maximum pump performance curves and minimum system line resistarues. The RCP seal injection flow is evenly distributed to each of the four RCS loops. The alternate charging path into RCS Loop 3 has slightly less system resistance and slightly more charging flow than the normal injection path into RCS loop 4. The peak pressure results from this CCP/PDP charging injection case establish the minimum administrative RCS temperature limit at which only one CCP or one PDP at a time (but not both) can be available for RCS injection.

With only one charging pump allowed for injection, the most limiting mass injection case is one CCP injecting through the normal and the alternate charging path simultaneously. This is conservative since DCPP plant procedures specify that only one charging injection path is used at a time. Table 6 12 summarizes the mass injection flow rates versus RCS pressure for one CCP with the flow rates based on a bounding combination of a maximum CCP performance curve and minimum system line resistances. The RCP seal injection flow is evenly distributed to each of the four RCS loops. The CCP injection flow for the alternate charging path enters RCS Loop 3, while flow through the normal charging path enters RCS loop 4. The peak pressure results from this CCP charging injection case establish the minimum administrative RCS temperature limit at which all four RCPs may be operated.

The number of operating RCPs determine the dynamic pressure drop between the peak pressure which occurs at the bottom of the RCS vessel (RETRAN volume 3) and the RCS hot leg (RETRAN volume 306) where the LTOP pressure transmitter is located. Since the DCPP PORV actuation parameters (setpoint, delay time, and stroke time) are all constant throughout the LTOP range, the dynamic pressure drop translates into a direct increase in the RCS peak pressure overshoot. The mass injection flow capability versus RCS pressure remains the same as listed previously in Table 6

12. However, as the RCS temperature and the corresponding Appendix G P/T limit continue to decrease, the number of operating RCPs must be restricted to ensure the LTOP PORV parameters adequately protect the resulting peak pressure at the bottom of the RCS vessel. The next four mass injection cases evaluate the peak pressure results for one CCP charging injection with 3,2, and I RCPs operating, respectively. These cases then determine the minimum administrative RCS temperature limits for operating 3,2, and I RCPs , respectively.

Once all of the RCPs are secured, the next mass injection case evaluated is for the CCP charging injection through the normal and alternate paths with no RCS flow except that from the two RHR pumps operating. The mass injection flow capability versus RCS pressure remains the same as listed previously in Table 6-12. This case determines the minimum RCS peak pressure overshoot for the mass injection capability and LTOP actuation parameters established in the PTLR. The results for

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )l&2 REV. NO. 0 SHEET NO. 14 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 this case then determine the minimum RCS temperature at which the LTOP PORV parameters can still maintain the peak pressure overshoot below the Appendix G P/T limit curve and subsequently the minimum administrative RCS temperature limit at which an RCS vent must be established.

The last mass injection case is evaluated to establish that with the RCS vent open, there is no credible LTOP which could challenge the Appendix G pressure limit. This case evaluates the maximum possible ECCS mass injection as assumed in Case I combined at the lowest credible RCS temperature of 55 'F.

7.3.2. RETRAN Evaluation of LTOP Heat Input Overshoot The heat input cases evaluate the startup of an RCP in one loop assuming that there is a maximum allowable temperature difference between the RCS and the SG secondary side. These heat input cases are evaluated over the range of applicable RCS temperatures to ensure that the DCPP LTOP parameters established in the PTLR maintain the resulting peak pressure overshoot within the Appendix G P/T curve limits. Table 6-9 lists the current DCPP LTOP setpoints and the other key LTOP related parameters which are being used in both the mass injection and heat injection evaluations. As discussed in Section 6.2.2, the DCPP LTOP model uses conservative adjustments to the RETRAN heat transfer model to bound the maximum potential SG to RCS heat transfer capability. These adjustments include increasing the volumetric heat capacity of the Inconel SG tubes by a factor of 1.5 and increasing the thermal conductivity by a factor often.

As established in WCAP 14040, the variation in the volumetric expansion of water versus temperature causes the heat input peak pressure results to increase significantly at higher RCS temperatures. Therefore, the heat input cases are evaluated at the minimum and maximum RCS temperatures which bound the applicable LTOP range. The heat input cases are also evaluated at appropriate intervals of RCS temperatures which adequately define the variation in peak pressure results throughout the LTOP temperature range. In particular, the heat input cases are evaluated near each of the minimum administmtive RCS temperature limits established by the mass injection results of the previous section to identify whether the mass input or the heat input case is more limiting.

7.3.3. RETRAN Evaluation of LTOP RCS Undershoot The RCS undershoot evaluation is performed to determine if the LTOP PORV closes before the RCS pressure in the cold legs decreases below the minimum value needed to maintain operability of the number one RCP seal. Since the RCP seals are located between the impeller and the diffuser, the RCP seal pressure is best represented as the average of the RCP volume and the RCP discharge or Cold leg volume. The minimum RCS undershoot is determined based on the two LTOP PORVs opening and closing simultaneously assuming that both the lift and reset setpoints have drifted 13 psi low. This drift value represents a 95% probability that both PORVs would not have simultaneously drifted lower than a setpoint value of 422 psig as established in design input 3.13. The PORV performance characteristics (lift setpoint, stroke time and delays) are constant throughout the complete LTOP range at DCPP. Therefore, the most limiting RCS undershoot would occur for the LTOP event with the minimum RCS mass increase and/or thermal expansion, since by definition, these effects tend to offset the pressure relief capabilities of the LTOP PORVs. The WCAP 14040 methodology for RCS undershoot can be implemented by evaluating the limiting RCS undershoot

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REE. NO.

SHEET NO. 0 15 OF 54 SUB JECT RETRAN Evaluation of DCPP LTOP Parameters MADIEBY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 cases based on the least severe mass input, and the least severe heat input RCS overshoot event which causes the LTOP PORV to actuate, respectively.

The least severe mass input RCS overshoot in the DCPP LTOP range was established in Section 6.3.1 based on a single injecting through the normal and alternate flow paths with no RCS flow except that from the two RHR pumps operating. This represents the smallest mass input flow rate, which could be expected to result in an actuation of the LTOP circuitry. Since the RCS flow and the RCP developed head change significantly as additional RCPs operate, an additional RCS undershoot case is evaluated for all four RCPs operating. These two RCS undershoot evaluations bound any potential combination of RCP operating conditions.

The limiting RCS undershoot evaluation for the heat input case is performed for the start of one RCP, since by definition, there is no potential RCS/SG temperature mismatch if one or more RCPs are already operating. Table 6-14 shows that the least severe heat input case which would still cause the LTOP PORV to lift at the 95% minimum setpoint of 436.7 psig is the case with a RCS/SG temperature difference of 120/200 'F. In order to evaluate the sensitivity of the RCS undershoot due to the relative heat input, a case is evaluated with the RCS/SG temperature difference at 180/260 'F.

The results of these evaluations determine if the DCPP LTOP setpoints are adequate or whether additional administrative controls are necessary to demonstrate the RCP seal are not adversely impacted during an LTOP event.

7.3.4. RETRAN Evaluation of Heat Input at 50% Pressurizer Level The DCPP RETRAN LTOP model was used to verify that the Tech Spec LCO 3.4.6.2 restriction on RCP operation is consistent with and remains bounded by the DCPP LTOP analysis. The Tech Spec LCO 3.4.6.2 restriction specifies that an RCP can not be started with the RCS/SG temperature difference greater than 50 'F unless the pressurizer level is less than 50%. The reduced pressurizer level provides additional margin for RCS fluid expansion to ensure that a worst case SG heat transfer to the RCS could not increase the RCS pressure above the Appendix G P/T limit. This case was evaluated to bound any potential RCS/SG temperature difference within the LTOP range based on assuming the RCS was at 270 'F and the SG secondary liquid (and RCS tube volume) was at 420 'F.

This RCS/SG temperature difference conservatively bounds the 100 'F mismatch which Westinghouse established as the maximum physically credible difference in Ref. 10. This heat input case was then evaluated with the initial pressurizer liquid level set to a value of 67% of the total pressurizer volume of 1800 f-t which conservatively bounds the pressurizer level uncertainty of

+6.1% (Design Input 12, Reference 20).

This case models the pressurizer partially filled with liquid and the LTOP PORV will relieve an air and steam mixture upon opening. Therefore, the PORV junction area and loss coefficient were reset to their original values as established in the Ref. 2 LOL analysis.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 16 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 7.4. L TOP Administrative Limits 7.4.1. Statistical Treatment of Measurement Uncertainties This section summarizes the methodology for determining the appropriate App. G P/T curve limit value which bounds the mass input and heat input peak pressure results including measurement uncertainties.

7.4.1.1. Mass Input Measurement Uncertainties The uncertainty terms for the mass input analysis are the uncertainty associated with the pressurizer PORV LTOP pressure actuation setpoint (+/-32 psi) and the uncertainty associated with measuring the RCS wide range temperature (+ 15 'F). Since the measurement uncertainties for the RCS pressure and the RCS temperature are independent they may be statistically combined using the sum of the squares methodology. However, while the pressure error has a constant effect on the P/T curve limit, the 15 'F temperature error has a significantly greater affect on the Appendix G P/T limit curve as temperature increases. Therefore, the following process is used to determine the equivalent pressure error with respect to the Appendix G P/T limit curve which conservatively bounds the mass input peak pressure at the specified RCS temperature.

1. Determine P/T curve "analysis temperature limit" at the "analysis peak pressure"

2. Determine new P/T curve pressure limit at "analysis temperature limif' plus measurement uncertainty

3. Calculate the change in the P/T pressure limit ("P/T Temperature error") between temperature limits I and 2

4. Calculate equivalent "P/T pressure error" due to "P/T temperature error" and pressure measurement error

5. Add equivalent "P/T pressure error" to "analysis peak pressure" to obtain "peak pressure limit"

6. Determine final "P/T temperature limit" corresponding to "peak pressure limit" It should be noted that as discussed in Section 6.3.1, the mass input analyses were performed with the 32 psi pressure measurement uncertainty already applied to the PORV LTOP lift setpoint to ensure that the analysis bounded any dynamic peak pressure affects due to a delayed PORV actuation.

Therefore, the original 32 psi pressure error is subtracted from the equivalent "P/T pressure error",

and only the net "P/T temperature error" is added to the "analysis peak pressure" which already includes the pressure error. The corresponding P/T limits including the measurement uncertainties for the mass input analysis cases are summarized in Table 6-16.

7.4.1.2. Heat Input Measurement Uncertainties The heat input analysis must address the additional measurement uncertainty for the Steam Generator or secondary' water temperature, which is used to establish the initial RCS/SG temperature difference prior to starting an RCP. Feedwater temperature is not an accurate indicator of the overall SG liquid temperature. If the SG pressure is above atmospheric conditions, the indicated SG saturation pressure can be used to determine the corresponding liquid saturation temperature. However, as identified in Design Input 8 and Ref. 12. the SG pressure control room indication has a 34 psi

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 17 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26/01 uncertainty. As shown in Table 6.4.1.2-1 below, this pressure error translates into very large temperature errors as the SG pressure decreases to atmospheric conditions.

Table 6.4.1.2 SG Temperature Error vs. SG Pressure Indication Saturation Saturation Average Temperature Temperature Pressure Error Indication Error OF Press F/psia F/(34 psi) 212 14.70 220 17.19 3.21 109.2 240 24.97 2.57 87.4 260 35.43 1.91 65.0 280 49.20 1.45 49.4 In order to eliminate the potential for such a large error when determining the maximum RCS/SG temperature differential is < 50 OF, the SG temperature is determined by use of a digital thermometer used to measure the secondary side SG metal temperature. As identified in Design Input 9, Ref. 18, the digital thermometer accuracy at 250 OF is +/- 3.4 OF and at 400 OF is +/- 5 OF. Therefore, the heat input analysis assumes a secondary side temperature error of +/- 15 OF to conservatively bound the digital thermometer measurement process and the potential for minor variations between the SG liquid and metal temperatures. This SG temperature uncertainty is conservative since the WCAP 14040 methodology defines the LTOP events as occurring at essentially steady state thermal conditions and the SG liquid and metal temperatures would be very close.

The resulting RCS/SG temperature uncertainty is based on assuming a +/- 15 OF uncertainty for both the RCS and SG temperature measurements, respectively. Although these uncertainties are physically independent and may be statistically combined, the heat input analysis conservatively evaluates the worst case bounding values for the RCS/SG temperature difference. The Table 6.4.1.2 2 below shows the eight possible combinations of RCS and SG measurement errors and their impact on the actual RCS/SG temperature difference for a measured 50 OF RCS/SG limit. The RCS peak pressure results tend to become more severe as the RCS temperature increases, and as the RCS/SG temperature difference increases. The Table shows that due to the offsetting impact of the RCS and SG errors on these two effects, it is not possible to select one uncertainty combination that conservatively bounds all of the other cases. Therefore, the LTOP heat input analysis evaluates the maximum possible RCS/SG temperature difference of 80 OF for all RCS temperatures within the LTOP range, including at the maximum RCS temperature plus uncertainty of 285 OF. The peak RCS pressure for each heat input case is then evaluated with respect to the lower Appendix G P/T curve limit corresponding to the measured RCS temperature value without uncertainty. As an example, the RCS/SG temperature difference of 285/365 OF, is evaluated with respect to the Appendix P/T curve limit for an RCS temperature of 270 OF. This evaluation methodology ensures that the heat input peak pressure results conservatively bound any physical possible combination of RCS/SG temperature values and their associated measurement uncertainties.

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA - 138 Project: Diablo Canyon Unit ( )1 ( )2 ( X )l&2 REV. NO. 0 SHEET NO. 18 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerrv F RBallard AT: 11 / ')A /01 . .

........-. . ,~.. , ,k ,f l U :DT LiAUII tit: UATI I I /zblU Table 6.4.1.2 RCS./SG Temperature Error Combinations Case Measured RCS SG Actual RCS /SG Temperature Temperature ROCS/SG Temperature Error Error Temperat ure Difference niffinrcon OF OF 1 270/320 High High 285/335 2 270/320 High Low 285/305 3 270/320 Low High 255/335 4 270/320 Low Low 255/305 5 270/320 None High 270/335 6 270/320 None Low 270/305 7 270/320 High None 285/320 0 Z/U/IJZU LOW None 255/320 The heat input peak pressure results for each case are then evaluated for the impact of the RCS temperature and RCS pressure measurement uncertainties on the Appendix G P/T curve limit identical to that discussed previously for the mass input analysis results. Table 6-17 provides a summary of the heat input peak pressure results and compares them to the corresponding Appendix G

P/T limits when including the measurement uncertainties.

8. RESULTS 8.1. RETRAN L TOP Model Benchmark 8.1.1. RETRAN Mass Input Benchmark Table 6-6 summarizes the comparison of the RETRAN mass input results with those of the Westinghouse LOFTRAN model established in Ref. 9. The RETRAN LTOP model is evaluated over an applicable range of mass injection flow rates, PORV lift setpoints, and PORV opening times established by the Westinghouse LOFTRAN results. The RETRAN LTOP model generates RCS pressure overshoot and undershoot results which are comparable to the Westinghouse LOFTRAN model.

8.1.2. RETRAN Heat Input Benchmark Table 6-8 provides a parametric summary of the various heat input cases for which the DCPP RETRAN LTOP model was compared to the generic Westinghouse results. Figure 6-2 compares the RETRAN heat input results to the LOFTRAN results for three different PORV actuation setpoints.

The RETRAN model accurately models both the magnitude and characteristic thermal response of the heat input transient over an appropriate range of RCS conditions.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )1&2 REV. NO.

SHEET NO. 0 19 OF 54 SUEBJECT RETRAN Evaluation of DCPP LTOP Parameters MAIDE BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 8.2. RETRAN Evaluation of DCPPLTOP Setpoints 8.2.1. Mass Input RCS Overshoot A summary of the mass input cases evaluated and the peak RCS pressure results are provided in Table 6-13. Figure 6-4 compares the mass input peak pressure results while Figure 6-5 plots the peak pressure values versus the corresponding administrative temperature limit to demonstrate how they remain bounded by the Appendix G P/T curve limit.

8.2.2. Heat Input RCS Overshoot Table 6-14 lists the heat input cases and the peak RCS pressure results. Figure 6-6 and Figure 6-7 compare the RCS pressure and temperature responses versus time, respectively for the various heat input cases. The heat input peak pressure results are also plotted in Figure 6-8 to demonstrate how they remain bounded by the Appendix G P/T curve limit.

8.2.3. LTOP RCS Undershoot The RCS undershoot results are summarized in Table 6-15. This Table lists the PORV setpoints, the time of the minimum undershoot, and the relative RCS pressures including the minimum effective value at the RCP seal. Figures 6-8 and 6-9 plot the RCS pressure versus time for the mass input cases with one RCP and four RCPs operating, respectively. These results show that the RCS pressure in the RCP volume remains above 235 psig (250 psia). The limiting mass input RCS undershoot case does not challenge the operation of the number one RCP seals. The results indicate that as more than one RCPs operates, the relative RCP suction pressure increases and the RCS undershoot becomes less limiting.

The heat input undershoot results are also shown in Table 6-15 while the relative RCS volume pressures for the two cases evaluated are plotted in Figure 6-11 and 6-12, respectively. The results show that the RCS undershoot for the heat input case is more severe than for the mass input cases, since the RCS pressurization is significantly less, and there is no net RCS mass addition to offset the release through the PORVs. As expected, the undershoot becomes less severe as the heat input pressure transient increases. Figure 6-11 shows that following the initial RCS pressure increase, there are two brief periods where the RCS pressure experiences an additional decrease due to the mixing effects of the initially heated RCS fluid with that which is still relatively cool. After the first closure of the PORVS, this mixing effect causes the RCS pressure to continue to drop below the 250 psia value for a brief period of about one minute. After a couple of minutes, the RCS volume is thoroughly mixed such that the SG heat transfer becomes the only significant effect. As discussed in Reference 6, the RCP seals can operate for up to two minutes without any seal injection flow. Based on the conservative conditions established in this RCS undershoot evaluation, the small and brief pressure decrease below 250 psia is not considered to adversely impact the RCP seal performance.

8.2.4. LTOP Pressurizer Level LCO Figure 6-13 plots the RCS pressure and pressurizer liquid level versus time for the heat input case with the initial pressurizer level at 67%. Due to the additional expansion volume within the partially

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )I ( )2 ( X )l&2 REV. NO. 0 SSHEET NO. 20 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 filled pressurizer, the PORV lift pressure is not reached until a time of 128 seconds. The RCS pressure and pressurizer level continue to increase until about 480 seconds when the conservatively low PORV relief capacity is able to offset the thermal expansion effects. The RCS pressure increases to a peak value of 622 psia which remains bounded by the comparable water solid heat input case already evaluated. The heat input transient and the subsequent peak pressures become less limiting as the RCS and SG temperatures decrease. This confirms that as long as the pressurizer level indicates less than or equal to 50%, an RCP started with even the maximum credible RCS/SG temperature differential remains bounded by the heat input cases and does not challenge the Appendix G P/T limit.

8.3. L TOP Administrative TemperatureLimits Table 6-18 summarizes the DCPP LTOP administrative temperature limits based on the most limiting results obtained from the mass injection and heat injection evaluations including the appropriate measurement uncertainty. It should be noted that for the DCPP LTOP parameters established in the PTLR, the mass injection results establish the administrative temperature limits since they generate more limiting peak pressure results than the heat injection cases over the range of LTOP applicability. These administrative limits ensure that the mass injection capability and the dynamic pressure drop across the RCS remain bounded by the LTOP analysis assumptions.

9. CONCLUSIONS 9.1. RETRAN L TOP Model Benchmark The RETRAN LTOP model generates RCS pressure overshoot and undershoot results, which are comparable to the Westinghouse LOFTRAN model. Therefore, the DCPP RETRAN model is appropriate for evaluating the LTOP setpoints per the WCAP 14040 methodology.

9.2. RETRAN Evaluation of DCPPL TOP Setpoints The DCPP LTOP setpoints as established in the PTLR (Ref. 7) ensure that the maximum RCS overshoot results remain bounded by the Appendix G P/T curve limit, including the appropriate measurement uncertainty. The DCPP LTOP setpoints ensure that with a minimum RCS vent available that is equal to the pressurizer PORV area of 2.07 in2 no credible LTOP event can challenge the Appendix G pressure limit. The DCPP LTOP setpoints also ensure that the minimum RCS undershoot results do not adversely impact the operation of the number one RCP seal during an LTOP event. The Tech Spec LCO 3.4.6.2 restriction on RCP operation and pressurizer level remains bounded by the DCPP LTOP analysis.

9.3. L TOP Administrative Temperature Limits The LTOP administrative limits established in Table 6-18 ensure that the mass injection capability and the dynamic pressure drop across the RCS remain bounded by the LTOP analysis assumptions.

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )I ( )2 ( X )l&2 SHEET NO. 21 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard II DATE 11/26/01 CHKD BY Dixon Yee DATE 11/26/01

10. IMPACT EVALUATION These evaluation results will be used as the technical basis for a License Amendment Request to the DCPP PTLR, and will become the licensing basis for the DCPP LTOP setpoints, upon NRC approval.

• Pacilic Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canon Unit ( )I ( )2 ( X )1&2 SHEET NO. 22 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK-D BY Dixon Yee DATE 11 /26/01

11. REFERENCES I. WCAP 14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Curves, WOG Program MUHP-3024 Rev.

2, January 1996.

2. PG&E Calculation N-098, "Reanalyzing FSAR Loss of Load/Turbine Trip Transient", By H. Lee. Dated 10/4/93.

3. LAR 95-06, DCL 95-220

4. PG&E Calculation STA-063 Rev. 2."Temperature Restrictions for LTOP", By D).

Yee, Dated 8/25/99

5. PG&E Calculation STA-143 Rev. 0, "Proto-Flo CVCS Charging Model for LTOP Analysis:, By A. Lin, 5/10/01 .

6. Diablo Canyon System Training Guide, B-la - Rev. 7 "Chemical and Volume Control", and A6 - Rev. 6, Reactor Coolant Pumps".

7. DCPP Procedure PTLR-1 "Pressure and Temperature Limits Report (PTLR) for Diablo Canyon" Rev. 1 , 10/11/00.

8. PG&E Calculation STA-121 Rev. 0, "LTOP Evaluations - RCS Undershoot and Heat Injection", 7/5/00.

9. Westinghouse Letter PGE-88-642, from J. Hoebel to J. E. Tomkins "LTOP Setpoint Evaluation Final Report", Dated May 5, 1988.

10. "Pressure Mitigating Systems Transient Analysis Results", July 1977. (Westinghouse Report on RCS Water Overpressurization prepared for the W Owners Group on RCS Overpressurization)

11. Westinghouse Letter PGE-88-593, from J. Hoebel to B. Giffin, "Summary Report of LTOP Reanalysis". Dated May 5, 1988

12. PG&E Calculation PAM-0 04-514, R5, "Post Accident Steam Generator Pressure Indication Uncertainty", 9/24/98.

13. PG&E Calculation J-100 Rev. 1, "Basis for LTOP Nominal Setpoints", PAM-0 7 403, Rev. 5, 8/26/97.

14. PG&E Calculation PAM-0 7-413, Rev. 7 "Post Accident Monitoring Indication, RCS Cold Leg Water Temperature and Hot Leg Water Temperature",

15. Crane Manual, Technical paper No. 410,"Flow of fluids Through Valves, Fittings and Pipe", Twenty Fifth Printing- 1991.

16. ASME Steam Table Fifth Edition.

Pacific Gas and Electric Company CALC. NO. STA- 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 23 OF 54 LUBJECT UDE BY RETRAN Evaluation of DCPP LTOP Parameters Jerry E. Ballard DATE 11/26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Tables Table 6-1: RETRAN LTOP Model Volume Summary Volume Number Volume Description Volume Height Flow Flow Area Hydraulic Lowest Length Diameter Elevation (it3) (ft) (ft) (ft2) (ft) ELEV 1 Upper Downcomer 280.032 9 3.085 90.831 1.531 103.917 2 Lower Downcomer 408.593 15.292 0 26.72 0.629 88.625 3 Lower Plenum 1023.73 9.95 0 128.497 3.774 79.51 4 Lower Core Section 218 4.458 0 51.013 0.0362 89.4583 6 Middle Core Section 218 4.458 0 51.013 0.0362 93.9163 7 Upper Core Section 218 4.458 0 51.013 0.0362 98.3743 8 Core Bypass Volume 71.32 13.375 0 5.33 0.0375 89.4583 9 Upper Plenum 873.15 12.67 12.67 72.26 0.962 102.8323 10 Reactor Vessel Head 485.3 7.14 11.23 89.44 1.82 112.917 101,201,301,501 Reactor Coolant Pump 79 8.43 0 13.6 N/A 101.1875 102, 202, 302, 502 RCS Cold Leg 1 76.696 2.292 0 4.126 2.292 105.854 103, 203, 303, 503 RCS Cold Leg 2 26.888 2.612 0 4.778 2.344 105.854 105, 205, 305, 505 RCS Hot Leg 1 15.957 2.932 0 5.09 2.388 105.534 106, 206, 306, 506 RCS Hot Leg 2 60.324 2.42 0 4.6 2.42 105.79 107, 207, 307, 507 Steam Generator HL Inlet 19.459 3.648 0 4.914 2.5 105.79 108, 208, 308, 508 SG HL Plenum 158.771 5.234 0 30.335 9.022 108.689 111,211,311,511 SG CL Plenum 158.771 5.234 0 30.335 9.022 108.689 112, 212, 312, 512 SG CL Outlet 16.464 3.142 0 5.24 2.583 105.72 113, 213, 313, 513 RCS CL Cross Under 1 23.727 4.53 0 5.24 2.583 101.19 114, 214, 314, 514 RCS CL Cross Under 2 37.04 5.794 0 5.24 2.583 95.396 115, 215, 315, 515 RCS CL Cross Under 3 18.445 2.583 0 5.24 2.583 95.396 116, 216, 316, 516 RCP Suction 37.04 5.794 0 5.24 2.583 95.396 401 Pressurizer Surge Line 45.93 6.883 0 0.6829 0.933 106.534 402 Pressurizer 1800 52.8 0 38.485 7 113.16 141,241, 341, 541 RCS HL SG Tube Bundle 61.635 5.553 0 11.099 0.0646 119.476 142, 242, 342, 542 RCS HL SG Tube Bundle 143.872 12.963 0 11.099 0.0646 132.439 143, 243, 343, 543 RCS HL SG Tube Bundle 143.872 12.963 0 11.099 0.0646 145.402 144, 244, 344, 544 RCS SG U-Tube Bundle 77 4.987 0 11.099 0.0646 132.439 145, 245, 345, 545 RCS CL SG Tube Bundle 143.872 12.963 0 11.099 0.0646 119.476 146, 246, 346, 546 RCS CL SG Tube Bundle 143.872 12.963 0 11.099 0.0646 113.923 149, 249, 349, 549 RCS CL SG Tube Bundle 61.635 5.553 0 11.099 0.0646 113.923

Engineering - Calculation Sheet  %.,MLU,,. NIU. 0 1/M - I10 Project: Diablo Canyon Unit ( 1 ( 12 ( X )1&2 REV. NO. 0 I' 'A- SHEET NO. 24 (OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-2: RETRAN LTOP Model Junction Summary, I of 2 Junction No. Description From To Initial Junction Junction Forward Volume Volume Flow Area Reverse Hydraulic Elevation Loss Rate Loss Diameter Coefficieni Coefficient 101,201,301,501 RCP Discharge to RCS CL 1 (Ib/sec) (ft2) (ft2) 101 102 0 4.125 107 (ft) 102, 202, 302, 502 RCS CL 1 to RCS CL 2 0 102 103 - 0 4.125 107 0 2.292 104, 204, 304, 504 RCS CL Nozzle 0 103 1 0 6.752 107 - 0.036 2.292 T1.413 0.394 2.292 1 Upper Downcomer to Lower Downcomer 1 2 0 26.72 103.917 2 Lower Downcomer to Lower Plenum 0.686 3 2 3 0 26.72 88.625 0 0 Lower Plenum to Lower Core Inlet 1.03 5

3 5 0 51.013 89.4583 8.06 0 - o Lower Core to Middle Core 5 6 Middle Core to Upper Core 6 0 51.013 93.9163 9.33 6 7 0 51.013 98.3743 9.33 0 0 7 Upper Core to Upper Internals 0 0

-7 9 0 9 3 2.83235 51.01. 4041 0 0 8 Lower Plenum to Core Bypass I18et 348 0 15.0 89.4583 30.976 0 -0 9 Core Bypass to Upper Internals 8 19 0 151.0 102.8333 0 0 0 10 4U Head epuper Downocmer Bypass to Reactor 142 10 0 10.0167 112.917 0 0 0 0 64 11 Reactor Head to Upper Internals 10 1 0 0.9235 115.502 0 0 0.139 105, 106, 205,305, 505 RCS 206, 306,5506 RCS HLHL Nozzle 1 to RCS HL 2 1.07 2.9 9

105 105 106 00 6.752 4.6 107 107 1.06 F107, 2077,3077,507 RC HL2 to SG HL Inet 0

-106 107 4.6 1-07 U.45 2.42 108, 208, 308, 508 SG HL Inlet to SG HL Plenum .8 107 1-08 5.23 -108.6----9 1.0 0.071 2.42 140 S§G -HL Plenum to HL Tube B§undle 1 0.5 2. 58-3 108 141 _ 0 11.09 113.923 0.66L_ 1 0.0646 141, 241,341, 541 HL Tube Bundle 1 to HL Tube Bundle 2 141 14-2 -0 1-1.099 -119.47 142, 242,342, 542 HL Tubee Bundle 2 to HL Tube Bundle 3 142 14-3 -0 11.099 132.439 0 0.0646 143, 243,343, 543 HL Tubee Bundle 3 to U-Tube Bundle 0 14 144 0 11.09"----145.402 0 0.0646 14, 244,344, 544 1.729 U-Tube Bundle to CL Tube Bundle 1 144 1.31 145 0 1-1.099 745.4-02 -1.729 0.0646 145, 245,345, 545 U Tube Bundle 1 to CL Tube Bundle 2 1.31 0.0646 145 146 0 11.099 132.439 0 146, 246,346, 546 CL Tube-Bundle 3 to CL Tube Bundle 3 0 0.0646Z 146 147 0 11.099 119.476 0 0 -a-nAR4

"4"PacificGas and Electric Company W Engineering - Calculation Sheet ,JALC. NO. STA- 138 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 25 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-2: RETRAN LTOP Model Junction Summary . 2 of 2 T - T - -- - I F -

JUIILnLIUIi INU. Description From To Initial Junction Junction Forward Reverse Hydraulic Volume Volume Flow Area Elevation Loss Loss Diameter p:nfat 147, 247,347, 547 SG CL Tube Bundle 3 to CL Plenum 147 111 0 11.099 113.923 , ,i*iiolenrt 1.32 Uoetriclent 0.5 0.0646 112,212,312,512 SG CL Plenum to SG CL Outlet 111 112 0 5.24 108.689 0.153 0.818 2.583 113, 213, 313, 513 SG CL Outlet to RCS CL Cross Under 1 112 113 0 5.24 105.72 0 0.058 2.583 114,214,314,514 RCS CL Cross Under I to Cross Under 2 113 114 0 5.24 101.19 0.3 0.113 2.583 115,215,315,515 RCS CL Cross Under2 to Cross Under 3 114 115 0 5.24 96.688 0 0.114 2.583 116, 216, 316, 516 RCS CL Cross Under 3 to Cross Under 4 115 116 0 5.24 96.688 0.3 0.227 2.583 117, 217, 317, 517 RCS CL Cross Under 4to RCP Suction 116 101 0 5.24 101.1875 0 0 2.583 461 RCP Seal Injection to Loop 1 0 102 0 1 0 0 0 0 462 RCP Seal Injection to Loop 2 0 202 0 1 0 0 0 0 463 Alternate Charging & RCP Seal Injection to 0 302 0 1 0 0 0 0 Loop 3 464 Normal Charging & RCP Seal Injection to 0 502 0 1 0 0 0 0 Loop 4 401 RCS HL 1 to Pressurizer Surge line 206 401 0 0.6829 107 2.308 0.308 0.933 402 Pressurizer Surge Line to Pressurizer 401 402 0 0.6829 113.167 1.273 0.801 0.933 PORV MODEL 411 Pressurizer PORV # 1 to Containment 402 900 0 0.00824 165.96 0.0727 0 0.116 412 Pressurizer PORV# 2 to Containment 402 900 0 0.00824 165.96 0.0727 0 0.116 900 RHR Suction from RCS Loop 4 0 506 0 0 107 0 0 0 901 RHR Injection to RCS CL Loop 1 0 103 0 0 107 0 0 0 902 RHR Injection to RCS CL Loop 2 0 203 0 0 107 0 0 0 903 RHR Injection to RCS CL Loop 3 0 303 0 0 107 0 0 0 904 RHR Injection to RCS CL Loop 4 0 503 0 0 107 0 0 0

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA - 138 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 26 OF 54 UBJECT RETRAN Evaluation of DCPP LTOP Parameters ADE BY Jerry E. Ballard I DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 Table 6-3: RETRAN LTOP Model SG Heat Conductor Summarv

Pacific Gas and Electric Company CALC. NO. STA- 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 27 OF 54 I UBJECT RETRAN Evaluation of DCPP LTOP Parameters DE BY Jerry E. Ballard DATE 11/26 /01 CHKD BY Dixon Yee DATE 11 /26/01 Table 6-4: Westinghouse Benchmark LTOP Mass Input Parameters RCS Volume (ft) 12372 RCS Temperature (Deg. F) 100 PORV Opening Characteristics Per Table 6-5 PORV Opening Setpoint (psig) 350, 400, 450, 500 PORV Opening Times (sec) 2.0, 4.0, 6.0 PORV Closing Time (sec) 2.0 PORV Flow Rate Cv (gpm/(psi)`) 46 PORV Electronic Delay (sec) 1.1 Mass Injection Flow Rate (gpm) 200, 300,400

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA- 138 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 28 OF 54 LUBJECT RETRAN Evaluation of DCPP LTOP Parameters DEBY Jerrv S.... E. Ballard DATE 11/ 9r /nI ,H..

Jr ---

.-- v , '-,r*-TJ L.TiAJIl cc UDAIT I I iLDoUI Table 6-5: Westinghouse Benchmark PORV Cv vs. Stroke Normalized Normalized Time IFlow Cv 00 0 0.000 1 0 0.100 0.033 0.200 0.055 0.300 0.08 0.400 0.115 0.500 0.165 0.600 0.235 0.700 0.335 0.800 0.48 0.900 0.7 1.000 1

Pacific Gas and Electric Company CALC. NO. STA- 138 Engineedng - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 29 OF 54 IRI'g1

UBJECT RETRAN Evaluation of DCPP LTOP Parameters E BY Jerrv E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-6: Benchmark Comparison of RETRAN vs LOFTRAN Results Mass Input RCS Overshoot RCS Under Shoot RCS PORV PORV Mass RETRAN West Over RETRAN West Under temp Setpt Stroke Injection Over Shoot (psi) Under shoot (psi)

(F) (psig) time Flow Rate shoot Shoot (sec) (gpm) (psi) (psi) 100 450 4 200 31.8 32 75.9 71 100 450 4 300 52.3 54 64.2 63 100 450 4 400 76.5 78 54.6 59 100 350 4 300 59.7 56 55.1 55 100 400 4 300 55.9 55 59.9 59 100 450 4 300 52.3 54 64.2 63 100 500 4 300 50.9 53 69 67 100 450 2 300 37.3 36 64.4 63 100 450 4 300 52.3 54 64.2 63 100 450 6 300 68.4 72 64.7 64

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA - 138 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 30 OF 54 k DE BY UBJECT RETRANJerry Evaluation E. Ballard of DCPP LTOP DATE 11/ Parameters 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-7: Westinghouse Benchmark LTOP Heat Input Parameters RCS Volume (ft3 ) 13000 4 L ____P RCS Temperature (Deg. F) 100 100 SG Temperature (Dea. F)

PORV Openingq Characteristics Linear vs. time PORV Opening Setpoint (psig) 350,400, 450, 500 PORV Opening Times (sec) 3.0 PORV Closing Time (sec) 3.0 PORV Flow Rate Cv (gpm/(psi)l2) 150 PORV Electronic Delay (sec) 0.6 RCS/ SG Temperature Delta (F)) 50, 100 RCP Flow Rate (gpm) 95,000 d..

r1 OdSart I ime sec) I10U

Pacific Gas and Electric Company CAMC. NO. STA- 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 SHEET NO. 31 OF 54 sUBJECT RETRAN Evaluation of DCPP LTOP Parameters VIADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-8: Westinghouse Benchmark LTOP Heat Input Results RCS / SG PORV LOFTRAN RETRAN Temperatures (F) setpoint Peak RCS Peak RCS (pisg) Pressure Pressure

__(psia) (psia) 140/240 400 592 595 140/240 500 682 675 140/240 600 790 787

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA- 138 Project: Diablo Canyon Unit ( )1 ( )2 (X)1&2 REV. NO. 0 SHEET NO. 32 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters DE BY Jerry E. Ballard m DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-9: DCPP LTOP Parameters RCS Volume (ft3 ) 11716 Pressurizer Volume (ft3) 1800 RCS Flow Rate per Loop (gpm) 95,000 RCP Start Time (sec) 10 RHR Flow per Pump (gpm) 5000 PORV Opening Characteristics Figure 6-10i PORV Opening Setpoint (psig) 435 PORV Reseat Setpoint (psig) 415 PORV Stroke Time (sec) 2.9 PORV Actuation Delay Time (sec) 1.05 PORV Flow Rate Cv gpm/(psi) 1 2 ) 46 PORV Ventm Ature Unc y2.07 RCS Temperature Uncertainty (OF) 15 SG Temperature Uncertainty (OF) 15 RCS Pressure Uncertainty (psi) 32 Maximum Measured RCS/SG 50 Temperature Difference (2F)

Maximum LTOP RCS Temperature (OF) 285 Minimum LTOP RCS Temperature (OF) 55

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 SHEET NO. 33 OF 54 RETRAN Evaluation of DCPP LTOP Parameters MUBJECT DE BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 Table 6-10 "Table 6-10-P"-

DCPP CCP Iniection

.In .......

1ECCSt IE .. .I with with.....

PDP RCS ECCS/

Pressure PDP Flow (psia) (gpm) 14.7 595.2 114.7 582.8 214.7 571.2 314.7 559.4 414.7 547.6 514.7 535.7 614.7 523.9 714.7 512 814.7 500.1 914.7 486.1 1014.7 472.2 I

Pacific Gas and Electric Company CALC. NO. STA- 1 Engineering - Calculation Sheet 38 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEETNO. 34 )F 54 OUBJECT MADE BY RETRAN Evaluation of DCPP LTOP Parameters Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 4

Table 6-11 : DCPP CCP Charging Injection with PDP RCS RCP Seal Total Total Normal Total CCP/PDP Pressure Injection to Alternate Charging Injection Flow (psia) Each Loop Charging Flow to to RCS (gpm) Flow to Loop-4 (gpm)

Loop-3 (gpm)

(gpm) 14.7 21.75 223.75 205.75 473 114.7 21.25 219.25 201.25 463 214.7 20.75 214.75 197.75 454 314.7 20.50 211.50 193.50 446 414.7 20.00 207.00 190.00 437 514.7 19.75 202.75 185.75 428 614.7 19.25 198.25 182.25 419 714.7 18.75 193.75 177.75 409 814.7 18.50 189.50 174.50 401 914.7 18.00 185.00 170.00 391 1014.7 17.50 180.50 165.50 381

Pacific Gas and Electric Company CALC NO. STA- 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 35 OF 54 LUBJECT RETRAN Evaluation of DCPP LTOP Parameters E BY Jerrv S. E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Jerrv E. Ballard Table 6-12: DCPP CCP Charaina Injection RCS RCP Seal Total Total Normal Total CCP Pressure Injection to Alternate Charging Injection Flow (psia) Each Loop Charging Flow to to RCS (gpm) Flow to Loop-4 (gpm)

Loop-3 (gpm)

(gPm) 14.7 19.00 198.00 182.00 418 114.7 18.50 193.50 177.50 408 214.7 18.25 189.25 174.25 400 314.7 17.75 184.75 169.75 390 414.7 17.50 180.50 165.50 381 514.7 17.00 176.00 162.00 372 614.7 16.50 171.50 157.50 362 714.7 16.00 167.00 153.00 352 814.7 15.75 162.75 148.75 343

.914.7 15.25 157.25 144.25 332 1014.7 14.75 152.75 139.75 322 I

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA - 138 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 36 OF 54 kUBJECT RETRAN Evaluation of DCPP LTOP Parameters ADE BY Jerry E. Ballard DATE 11/ 2R/01 rmi ,-ov J

-.... . ..- ... *-,,,'..,. , , T.*"'JI

-ivCs V I *; Resul-- I I ILOIU-I U/*II:

Table 6-13: DCPP ITOP Mass Input Peak Pressure Results Mass Input Case Num. Time of P ressurizer Vessel Hot Leg RCS RCPs Peak P ressure Pressure Pressure Pressure (sec) Vcolume 402 Volume 3 Volume 306 Overshoot (F>SlA) (PSlA) (PSIA)

(PSID) r~r%-~.+ ~ - Irr - Iar CUI/ 1"U13 (U-US)

I 4

f 37.9 37.9 587.2 648.4 579.1 198.4

-- -- - I CCP and PDP 4 38.9 564.2 625.1 565.5 175.1 4 38.9I CCP -4 RCPs 4 39.5 553.3 614.0 545.4 164 CCP - 3 RCPs 3 39.5 544.7 593.8 546.3 143.8 CCP - 2 RCPs 2 39.5 543.7 579.1 546.6 129.1 CCP- 1 RCPs 1 39.5 543.1 570.7 546.6 120.7 CCP - 0 RCPs 0 39.2 531.9 557.0 547.3 107 CCP/ PDP (ECCS) w/vent 0 350 172.6 196.5 186.7 N/A

Pacific Gas and Electric Company CALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 37 OF 54 SHEET NO.

RETRAN Evaluation of DCPP LTOP Parameters L UBJECT DE BY Jerry E. Ballard DATE 11/26/01 CHK'DBY Dixon Yee DATE 11/26/01 Table 6-14: DCPP LTOP Heat Input Peak Pressure Results RCS / SG RCS Time Pressurizer Vessel Hot Leg Volume Temperatures Pressure (sec) Volume 402 Volume 3 306 (psia)

(OF) (psia) (psia) (psia) 100/180 300 10.6 276.1 301.9 278.2 120/200 300 75 470.8 496.4 472.7 135/215 300 12.1 478.9 504.5 480.9 150/230 300 12.4 548.4 574.2 550.7 180/260 300 12.4 621.9 648.1 624.8 200/280 350 11.9 665.4 692.6 669.4 230/310 350 11.9 669.5 697.6 686.2 270/350 400 11.2 669.7 698.3 687 285/365 400 12.1 671.1 698.1 675.8

I-acmric (as ana liectric Company Engineering - Calculation Sheet UALC. NO. STA- 138 Project: Diablo Canyon Unit ( )l ( )2 ( X 1R)9 REV. NO. 0

' '"'" SHEET NO. 38 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHKD BY Dixon Yee DATE 11 /26/01 Table 6-15: DCPP LTOP RCS Undershoot Results Description PORV 1 PORV PORV PORV Time of m IPRES PRES PRES PRES PRES RCP Seal Lift 1 2 Lift 2 Reset Under 402 506 502 501 516 Pressure Reset Shoot (PSIA) (PSIA) (PSIA) (PSIA) (PSIA) (psia)

(sec)

Mass Input 1 RCP 436.7 416.7 436.7 416.7 40.3 194.3 Mass Input 4 RCP 436.7 289. 7 ý279.0 282.2 252.2 267.2 416.7 436.7 416.7 41.1 2731.4 283.5 340.6 291.4 222.8 316.0 Heat Input 436.7 416.7 -436.7 282 2ý31.44 232.96 238.6 209.64 152.95 RC/SG 120/200 416.7 4

224.1 Heat Input 436.7 416.7 436.7 16. 247.2- 248.3 253.9 225.8 169.9 RC/SG 180/260 239.9

raciv uas ano tiectric uompany LALC. NO. STA- 138 W Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 39 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11 /26/01 Table 6-16: DCPP LTOP Mass Input P/T Limit with Uncertainty MASS INPUT CASE RCS RCS Analysis LTOP Revised Revised Pressure Total Analysis Temp.

Press. Temp. Peak Temp. Limit Temp. Press. Error for Press. Over shoot Limit with Uncert. Uncert. Pressure @ Analysis Limit with Limit at Temp. Error Plus Total Total (psi). (OF) Overshoot Overshoot Temp. Temp. Uncert Sum of Press. Error Press.

(psia) (OF) Uncert. Uncert. (psi) Squares (psia) Error Added Added (psi) (OF)

(OF) (psia)

CCP/PDP (ECCS) 32 15 648.4 145.5 160.5 681.5 33.1 46.0 662.4 152.2 CCP and PDP 32 15 625.1 132.5 147.5 652.5 27.4 42.1 635.2 138.5 4 RCPs CCP - 4 RCPs 32 15 614 125.4 140.4 638.7 24.7 40.5 622.5 130.9 CCP - 3 RCPs 32 15 593.8 110.2 125.2 613.6 19.8 37.6 599.4 114.8 CCP - 2 RCPs 32 15 579.1 96.5 111.5 595.4 16.3 35.9 583.0 100.4 CCP- 1 RCPs 32 15 570.7 87.2 102.2 584.9 14.2 35.0 573.7 90.7 CCP - 0 RCPs 32 15 557 68.9 83.9 568.0 11.0 33.8 558.8 71.6

Engineering Calculation Sheet REV. NO. 0 mE Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 -w SHEET NO. 40 ()F 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26 /01 CHK'D BY Dixon Yee DATE 11/26/01 Table 6-17 : DCPP LTOP Heat Input PIT Limit with Uncertainty Heat Input Case RCS RCS Analysis LTOP Revised Revised Pressure 1total Analysis Temp.

Press. Temp. Peak Temp. Limit Temp. Press. Error for F'ress. Over shoot Limit with Uncert. Uncert. Pressure @ Analysis Limit with Limit at Temp. E:rror Plus Total Total (psi). (°F) Overshoot Overshoot Temp. Temp. Uncert (psia) Sum of Press. Error Press.

(OF) Uncert. Uncert. (psi) iquares (psia) Error Added Added (Ipsi) (OF)

(F (psia) 100/180 32 15 301.9 60.0 75.0 561.2 3.4 32.2 302.1 60.0 120/200 32 15 496.4 60.0 75.0 561.2 3.4 32.2 496.6 60.0 135/215 32 1___5 504.3 60.0 75.0 561.2 3.4 32.2 504.5 60.0 150/230 32 15 574.2 91.2 106.2 589.3 15.1 35.4 577.6 94.9 180/260 32 15 648.1 145.3 160.3 681.1 33.0 46.0 662.1 152.0 200/280 32 15 692.6 164.9 179.9 736.4 43.8 54.2 714.8 172.9 230/310 32 15 697.6 166.7 181.7 742.6 45.0 55.2 720.8 174.9 270/350 32 15 698.3 167.0 182.0 743.5 45.2 55.3 721.6 175.2 285/365 32 15 698.1 166.9 181.9 7 45.17 55.3 721.4 175.1

Pacific Gas and Electric Company CALC. NO. STA- 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 41 OF 54 RETRAN Evaluation of DCPP LTOP Parameters UBJECT Z

ETRAN Evaluation of QCPP LTOP Parameters E BY Jerry E. Ballard DATE 11/26/01 CHICD BY Dixon Yee DATE 11/26/01 Table 6-18: DCPP LTOP Administrative Temperature Limits LTOP Administrative Action RETRAN PTLR Code Case N-514 Temp. Limit LTOP Enable - Disable one CCP 270 Block ECCS Flow Path 153 Disable Second Charging Pump 139 Stop 1 of 4 RCPs 131 Stop 2 of 4 RCPs 115 Stop 3 of 4 RCPs 101 Stop 4 of 4 RCPs 91 Establish RCS Vent 72

Pacific Gas and Electric Company Engineering - Calculation Sheet uALC. NO. STA- 138 w Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 42 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/ 26 /01 Figures Figure 6-1 : DCPP RETRAN LTOP Evaluation Model LEGEND 453] Junction (103 ')Volumne

ý, . Heat Conductor

"--raLkaeou %ats ano ciectric trompany L,ALC. NO. STA - 138 Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 43 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/26 /01 Figure 6-2: RETRAN vs LOFTRAN LTOP Heat Input Results RETRAN vs LOFTRAN Heat Input Response RCSISG = 1401240F, RCS Volume = 13000ft3 900 800 700 EU 600 500 400 300 0 4 8 12 16 20 24 Time (sec)

vdLIIIL - ano tIzectric LUompany Engineering - Calculation Sheet LALC. NO. STA- 138 w Project: Diablo Canyon Unit ( '1 ( V' ( y \1 o REV. NO n SHEET NO. 44 C

)F 54 SUBJECT RETRAN Evaluation of DCPP LTOP MADE BY Parameters Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/ 26

/01 Figure 6-3: DCPP PORV Normalized Valve Cv vs Position DCPP Pressurizer PORV Stroke Sequence for LTOP Analysis

-- PORV Position Lift pressure @ 0.0 sec PORV Cv Reset pressure @ 5.0 sec (I 1 IL 0

0

'a 0

o..

0 Z

0 00 5.0 10.0 Time (sec)

vaciTic uas and Ilectric company .,ALC. NO. STA- 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 SHEET NO. 45 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 Figure 6- 4: DCPP RETRAN LTOP Mass Input Pressure Results DCPP LTOP Mass Input RCS Peak Pressure Comparison 650 600 550 41

" 500 4

a.450 0f 400 350 300 0 10 20 30 40 50 60 Time (Seconds)

Pacific Gas and Electric Company Engineering - Calculation Sheet CALC. NO. STA - 138 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 46 OF 54 UBJECT RETRAN Evaluation of DCPP LTOP Parameters ADE BY Jerry E. Ballard DATE 11/ 26 /01 CCHK'D BY Dixon Yee DATE 11/ 26 /01 Figure 6-5 DCPP RETRAN LTOP Mass Input P/T Summary Mass Injection Peak Pressure vs P/T Limit (with Uncertainties) 750 700 m

"650 Co60.

o 600 550 500 60 90 120 150 180 RCS Temperature (Deg. F)

r UI IOUn WaS 4111Iu &r. II II ., O I I IdI ly tALC. NO. STA- 138 w Engineering - Calculation Sheet Project: Diablo Canyon Unit( )1 ( )2 (X)1&2 REV. NO. 0 SHEET NO. 47 OF 54 SUBJECT RETRAN Evaluation of r)CPP LTOP Parameters MADE BY Jerrv E. Ballard DATE 1 1/26/01 CHK'D By Dixon Yee DATE 11/26/01 Figure 6- 6: DCPP RETRAN LTOP Heat Input Pressure Results DCPP LTOP Heat Input Comparison RCS Pressure vs Time 700 285/365F 650 " 230/31OF

• 200/280F x 150/230F 600

• 120/200 550 RI* .. x~xx_[

500 I

450 Ii 0

U 400 x

x x

350 x '0< ýXww*** X "x

"_J_qMM  % x )K A-Nin 300 250 200 0 10 20 30 40 50 60 Time (sec)

w Engineering - Calculation Sheet Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 48 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26/01 CHK'DBY DixonYee DATE 11/26/01 Figure 6- 7 : DCPP RETRAN LTOP Heat Input Temperature Results DCPP LTOP Heat Input Comparison RCS Temperature vs Time 350 300 F_ _-, -

v v i- 250

............ ,I E

r- W _= =1 .

0 4) 0 E

150 - '

a

• II I 0 0 *0 -

0 0 0 "-#--285/365F 100 -"-230/310 F 200/280

>. 150/230F 50

• 120/200F 0 5 10 15 20 25 30 Time (sec)

Pacific Gas and Electric Company CALC. NO. STA- 138 Engineering - Calculation Sheet 0 REV. NO.

Project: Diablo Canyon Unit ( )1 ( )2 (X)1&2 SHEET NO. 49 OF 54 bUBJECT RETRAN Evaluation of DCPP LTOP Parameters DE BY Jerry E. Ballard DATE 11/26/01 CHKID BY Dixon Yee DATE 11/26/01 Figure 6- 8: DCPP RETRAN LTOP Heat Input P/T Summary LTOP H-at InlM Peak Presiur vs P/T L*it 750 700 650 U

S S

S a.

U I

450 4O0 60 90 120 150 180 210 240 270 RCS TeTprahm (Dog%F)

W rdw.iIL vdn dimaUriectric %,ompany UALC. NO. STA - 138 Engineering - Calculation Sheet REV. NO. 0 Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 SHEET NO. 50 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 Figure 6- 9: DCPP RETRAN LTOP I RCP Mass Input RCS Undershoot RCS Undershoot - Mass Input with i RCP 95% Minimum Setpoints (436.7,416.7 psig) 500 450 400 0

0 (0

350 a)

I 0.

C.)

300 250 200 0 10 20 30 40 50 60 Time (sec)

"-'--alm,,0, *db dIUU rEiGetric t-ompany Engineering - Calculation Sheet LALC. NO. STA- 138 W Project: Diablo Canyon Unit ( )1 ( )2 ( X )1&2 REV. NO. 0 SHEET NO. 51 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY Dixon Yee DATE 11/ 26 /01 Figure 6- 10: DCPP RETRAN LTOP 4 RCP Mass Input RCS Undershoot RCS Undershoot - Mass Input with 4 RCPs 95% Minimum Drift Setpoint (436.7,416.7) 60(0 55' 0 5040 S45C 0 40(

E V3) 9 35(

300 250 200 0 10 20 30 40 50 60 Time (9ec)

Q -dl*d fl cll U *triIt; Lompany Engineering - Calculation Sheet uALC. NO. STA - 138 w Project: Diablo Canyon Unit ( )1 ( )2 ( X REV. NO. 0

)1&2 SHEET NO. 52 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/ 26 /01 CHK'DBY Dixon Yee DATE 11/ 26 /01 Figure 6- 11 : DCPP RETRAN LTOP Heat InDut RCS Undershoot A-120 F RCS Undershoot - I RCP- Heat Input w/RCS @ 120F 95% Minimum Setpoint (436.7,416.7 psia) 500 450 400 350 CL 300 L) a.

250 0.,

ci, 200 150 100 50 0

300 350 400 Time (sec)

Engineering - Calculation Sheet L,ALC. NO. STA - 138 w Project: Diablo Canyon Unit ( )1 ( )2 (X )1&2 REV. NO. 0 SHEET NO. 53 OF 54 SUBJECT RETRAN Evaluation of DCPP LTOP Parameters MADE BY Jerry E. Ballard DATE 11/26/01 CHK'D BY Dixon Yee DATE 11/26/01 Figure 6- 12: DCPP RETRAN LTOP Heat Input RCS Undershoot @. 180 F RCS Undershoot - I RCP Heat Input w/RCS @ 180F 95% Minimum Setpoint (436.7,416.7 psla) 600 550 500 450 U

400

a. 350 U,

300 250 200 150 0 10 20 30 40 50 60 70 80 90 Time (sec)

Engineering - Calculation Sheet LALC. NO. STA- 138 w Project: Diablo Canyon Unit ( )1 ( )2 REV. NO.

(X )1&2 SHEET NO. 0 54 OF 54 SUBJECT MADE BY RETRAN Evaluation of DCPP LTOP Parameters Jerry E. Ballard DATE 11/ 26 /01 CHK'D BY_ Dixon Yee DATE 11/26/01 Finure 6- 13 DCPP RETRAN LTOP Heat Input Pzr ( 50%

RETRAN Heat Input Response RCS/SG

= 270/420F, Pressurizer Level @ 67%

650 100 600 90 550 (U

80 a.

500 U",

70 (0

450 "U,

-J N

0, 60 0.

400 50 350 0

40 Time (seconds)

Enclosure 6 PG&E Letter DCL-02-079 Supporting References for DCPP PTLR This enclosure contains:

1. CE Power Systems letter to Westinghouse Nuclear Energy Systems dated September 12, 1979, transmitted to PG&E by Westinghouse letter PGE-4083 dated July 21, 1980 as amended by Westinghouse letter PGE-6352 dated December 20, 1984.

2. Westinghouse letter PGE-88-765 dated December 14, 1988.

i 0 C60 "S PSE-6352 oNow~w ea'S bsi Westingf louse Water Reactor Electric Corporationl Divisions arnM Msmugir ftmy~Iwa 15230 I December 20, 1984

3. V. Rocca Ref: PGE-4083 Chief Mechanical Engineer Pacific Gas A Electric Company c/o Bechtel Power Corporation Diablo Canyon Project 45 Fremont Street, 10th Floor, Room 028 San Francisco, CA 94602 Attention:

"17 J. J. McCracken r~k.PACIFIC GAS AND ELECTRIC COWPAY NUCLEAR PLANT, DIABLO CANYON UNITS; 1 & 2

-- - Reactor Vessel Minimum Bolt-Up Temperatures

Dear Mr. Rocca:

cc In the referenced letter, Westinghouse forwarded for inclusioan in the reactor Svessel manual a Combustion Engineering letter stating that the minimum bolt-up

.. , temperature is RTNDT per the ASME Code requirements. This RTNDT per the ASI4E Code is that of the affected areas, the upper shell and vessel flange. The all referenced letter itself stated that even if RTNDT was below 600 F a minimum bolt-up temperature of 60OF should be used.

CDSince and the for the RTNDT's Tech Specs the affected areas given in the Diablo Canyon Unit 1 Draft Diablo Canyon Unit 2 Tech Specs transmitted to PGandE by Westinghouse are below 600F, a minimu bolt-up temperature of 600 F Is applicable to both units. This bolt-up temperature is applicable for the life of the vessels.

Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION Manager Electric Project CWVernon/rcc/2479d V. Rocca IL

'3. E. Murphy (WSan Francisco office)

3. B. Hach 1L IL C.-Ioernry IL

DIABLO CANYON VAlit .I a 02 S) PGE-408 LTEK1 File.-: AL Westinghouse 0- Copy For: m rui ConmWrciai Water Reactor [3 Mn. File oewlatons Division Electric Corporation Divisions 0i Fle 1.13 B M355 PitsbigxhPmnnsy*Jan 15230 Date .July 21. 1980 Mr. D. V. Kelly Chief Mechanical Engineer S.O. No. PEG/PGE-IO.

PACIFIC GAS AND ELECTRIC COMPANY 0 2W Engr. Ltr. EP/SA-259t 0 N. "77 Beale Street all San Francisco, California 94106 3 2-6 1 PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR PLANT, DIABLO CANYON SITE I REVISION TO REACTOR VESSEL INSTRUCTION MANUAL I

.IWV-

> 'ie following documents are transndtted herewith for your use. The status of each docum cji5 one of the following as noted below:

Preliminary (PRE) Certified For Construction With Comments (CC a Approved For Layout. (AFL) As Manufactured (ASH)

Certified For Construction (CFC) Approved (APP) 0 to 0 SPIN No. Document No. Sht. Rev. Status Unit Document Title RCPCRV C.E. Letter - - APP I Revision to Reactor Vessel Sept. 12, 1979 Instruction Manual RCPCRV C.E. Letter - - APP 2 RevisIon to Reactor Vessel Sept. 12, 1979 Instruction Manual Consent: The attached letter relays the minimum bolt-up temperature for the reactor ves!

studs. Please insert this letter into your Reactor Vessel Instuction ManualsI modify your bolt-up procedures appropriately. However, Westinghouse adds an a, tional requirement that the bolt-up temperature shall not be below 60°F.

WESTINGHOUSE NUCLEAR ENERGY SYSTEMS J. F. Duran r-r-I AAG/YB/=

D. V. Kelly 6L, 17A biCr.

. 1cGangloff, Gas and Project Electric ManagerProject R. W. Beckwith IL (W San Francisco)

- *1

C-E Power Systems . Tel, 615/265-.631 "Combustion Engineering. Inc.

911 W. Main Street Chattanooga. Tennessee 37402

-3POWER SYSTEMS September 12, 1979 Westinghouse Electric Corporation Nucitar Energy Systems Nort.*orn Pike Road tbonrdeville, Pennsylvania 15146 Attention: Wr. M. A. Ditillo

Subject:

Reactor Vessel Bolt-up and Hydrotest Temperature Requirements Refarence: Contract PGE/67757/23066 Gen~temen:

The instruction manual for the reference contract is hereby upgraded in conformance 2222(c), Summer 1976 Addenda. This with AS%- Code Section III, Appendix G, Paragraph requirement.

applies only to bolt-up and hydrotest pressure Bolt-up temperature shall n ot be lower than RTNDT at a pressure not to exceed 600 psig.

Reactor Vessel pressurizati on above 600 psig will be at a temperature not lower than RTNDT +60 0 F.

Yours very truly, INC.

EiERINGs ENAG 0SUSTION CContract Administrator

. JLP/Jpu cc - Mr. J. A. Mbal. Jr.

IM. A. S. Harpor hr. J. S. W4eok S *