•** **- UNITED STATES * * NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

October 26, 2000

Mr. Gregory M. Rueger Senior Vice President, Generation and

Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P. O. Box 3 Avila Beach, CA 94177

SUBJECT: DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: UNIT 1 REACTOR CORE THERMAL POWER UPRATE (TAC NO. MA7813)

Dear Mr. Rueger:

The Commission has issued the enclosed Amendment No. 143 to Facility Operating License No. DPR-80 for the Diablo Canyon Nuclear Power Plant, Unit No. 1. The amendment consists of changes to Facility Operating License No. DPR-80 and to the Technical Specifications (TS) in response to your application dated December 31, 1999, as supplemented by letters dated January 18, July 7, September 22, and 29, and October 12, 2000.

The amendment revises Section 2.C.(1) of Facility Operating License No. DPR-80 to authorize operation at reactor core power levels not to exceed 3411 megawatts thermal (100 percent rated power). The amendment also (1) revises the definition of rated thermal power in Section 1.1 of the TS to reflect Unit 1 operation at the uprated reactor core power level, (2) changes the reactor core safety limits in TS Figure 2.1.1-1 to reflect the current fuel type, and (3) provides additional margin for Overtemperature AT (OTAT) and Overpressure AT (OPAT) setpoint calculations and changes the nominal full power Tavg in the OTAT and OPAT function in Notes 1 and 2 to TS Table 3.3.1-1.

Please note that your analysis and our evaluation are based on the 17X1 7 VANTAGE 5 fuel assemblies, and ZIRLO cladding. Therefore, use of any other fuel design or cladding material is subject to our review and approval. We understand that you have identified two errors in your large break loss-of-coolant-accident (LBLOCA) analysis. Although your evaluation of these errors shows that the peak clad temperature would be 2009 OF, and below the 10 CFR 50,46 acceptance limit of 2200 OF, by letter dated October 12, 2000, you committed to complete the LBLCOA reanalysis by July 26, 2003.

Mr. Gregory M. Rueger

A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely,

IRA/

Docket No. 50-275

Enclosures: 1. Amendment No. 143 to D[ 2. Safety Evaluation

cc w/encls: See next page

Accession No. ML0037

L. Raghavan, Sr. Project Manager, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation

DISTRIBUTION: PUBLIC GHill (2) PDIV-2 Reading WBeckner

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NAME JZwolinski BSh o SCollins

DATE L-\10 I I_______VDOCUMENT NAME: G:\PDIV-2\DiabloCanyon\Amda7813final.wpd

OFFICIAL RECORD COPY

-2- October 26, 2000

SJ

Diablo Canyon Power Plant, Unit 1

cc: NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Commission P.O. Box 369 Avila Beach, CA 93424

Dr. Richard Ferguson, Energy Chair Sierra Club California 1100 11th Street, Suite 311 Sacramento, CA 95814

Ms. Nancy Culver San Luis Obispo

Mothers for Peace P.O. Box 164 Pismo Beach, CA 93448

Chairman San Luis Obispo County Board of

Supervisors Room 370 County Government Center San Luis Obispo, CA 93408

Mr. Truman Burns Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, CA 94102

Mr. Steve Hsu Radiologic Health Branch State Department of Health Services P.O. Box 942732 Sacramento, CA 94327-7320

Diablo Canyon Independent Safety Committee

ATTN: Robert R. Wellington, Esq. Legal Counsel

857 Cass Street, Suite D Monterey, CA 93940

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064

Christopher J. Warner, Esq. Pacific Gas & Electric Company Post Office Box 7442 San Francisco, CA 94120

Mr. David H. Oatley, Vice President Diablo Canyon Operations and

Plant Manager Diablo Canyon Nuclear Power Plant P.O. Box 3 Avila Beach, CA 93424

Telegram-Tribune ATTN: Managing Editor 1321 Johnson Avenue P.O. Box 112 San Luis Obispo, CA 93406

Mr. Ed Bailey, Radiation Program Director Radiologic Health Branch State Department of Health Services P.O. Box 942732 (MS 178) Sacramento, CA 94327-7320

Mr. Robert A. Laurie, Commissioner California Energy Commission 1516 Ninth Street (MS 31) Sacramento, CA 95814

N E UNITED STATES *NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

PACIFIC GAS AND ELECTRIC COMPANY

DOCKET NO. 50-275

DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 143

License No. DPR-80

The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas and Electric Company (the licensee) dated December 31, 1999, as supplemented by letters dated January 18, July 7, September 22, and 29, and October 12, 2000, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to Facility Operating License No. DPR80 and the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

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(2) Technical SDecifications

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 143 , are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

.Seiuel J. C ?s, Director " O Office of Nuclear Reactor Regulation

Attachment: Changes to the Operating License and Technical Specifications

Date of Issuance: October 26, 2000

ATTACHMENT TO LICENSE AMENDMENT NO. 143

TO FACILITY OPERATING LICENSE NO. DPR-80

DOCKET NO. 50-275

Replace the following pages of Facility Operating License No. DPR-80 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT

Operating License

3

Technical Specifications

1.1-5 2.0-2 3.3-17 3.3-18

Operating License

3

Technical Specifications

1.1-5 2.0-2 3.3-17 3.3-18

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(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level

The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2) Technical Specifications

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3) Initial Test Program

The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:

a. Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential;

Amendment No. 143

Definitions 1.1

1.1 Definitions (continued)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

QUADRANT POWER TILT RATIO (QPTR)

RATED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM (RTS) RESPONSE TIME

SHUTDOWN MARGIN (SDM)

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the power operated relief velve (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, 'tow Temperature Overpressure Protection (LTOP) System."

QPTR shall be the ratio of the maximum upper excore 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.

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

The RTS RESPONSE TIME shall be that time interval from 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.

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. t-5, 143 Unit 2 - Amendment No. 135

Sis 2.0

UNITS I & 2

40% 60% 80% 100%

Percent of Rated Thermal Power (RTP)

When operating in the reduced RTP region of Technical Specification LCO 3.4.1 (Table 3.4.1-1 for Unit 1 and Table 3.4.1-2 for Unit 2) the restricted power level must be considered 100% for this Figure.

FIGURE 2.1.1-1 REACTOR CORE SAFETY LIMIT

DIABLO CANYON - UNITS 1 & 2 2.0-2 Unit 1-Amendment No. *95-;143 Unit 2 -Amendment No. 135

680

670

660

650

640

U

0,

630

620

610

600

590

580

570

56o L

0% 120%

RTS Instrumentation 3.3.1

Table 3.3.3-1 (page 6 of 7)

Reactor Trip System Instrumentation

Note 1: Overtemperature AT

The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 0.46% of AT span for/hot leg or cold leg temperature inputs, 0.14% AT span for priessurizer pressure input, 0.19% AT span for Al inputs.

AT (1l+4s)<ATO {K - I~ ) (1+r<s K -1 Is) T-T•+ K3(P-P')-f 1 (AI)

Where: AT is measured RCS AT, °F.

ATo is the loop specific indicated AT at RTP, OF.

s is the Laplace transform operator, sec-1 .

T is the measured RCS average temperature, °F.

T' is the nominal loop specific indicated Tvg at RTP, < 577.3 (Unit 1) & 577.6 (Unit 2)°F.

P is the measured pressurizer pressure, psig

P' is the nominal RCS operating pressure, = 2235 psig

K1 = 1.20 K2 = 0.01821°F K3 = 0.000831/psig t 1 30 sec r2 4 sec

T4 = 0sec T5=0 sec

f1(AI) = - 0.0275{ 19 + (q, - qb)} when q, - qb < - 19% RTP

0% of RTP when -19% RTP < q - qb•-< 7% RTP

0.0238{(qt - qb) - 7) when q, - qb > 7% RTP

Where q, and qb are percent RTP in the upper and lower halves of the core, respectively, and q, + qb is the total THERMAL POWER in percent RTP.

DIABLO CANYON - UNITS 1 & 2 3.3-17 Unit 1 -Amendment No. 13 143 Unit 2 - Amendment No. 135

RTS Instrumentation 3.3.1

Table 3.3.3-1 (page 7 of 7)

Reactor Trip System Instrumentation

Note 2: Overpower AT

The Overpower AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 0.46% of AT span for hot leg or cold leg temperature inputs.

AT (I+ 4 s)<A'r K4 -K, 1 T-KIT-T'I-f,(A)} (1+r 5s) l+13s

Where: AT is measured RCS AT, OF.

ATo is the loop specific indicated AT at RTP, °F.

s is the Laplace transform operator, sec1 .

T is the measured RCS average temperature, OF.

T" is the nominal loop specific indicated Tvg at RTP, (Unit 2)°F.

K4 = 1.072 K5 = 0.01 74/0F for increasing Tavg 0/°F for decreasing Tav,

13 -10 sec ' 4 = 0 sec

f2(AI) = 0% RTP for all Al.

Note 3: Steam Generator Water-Level Low Low Time Delay

< 577.3 (Unit 1) & 577.6

Ks = 0.00145/°F when T > "T"0/*F when T• T"

5 =0 sec

The Steam Generator Water Level-Low Low time delay function power allowable value shall not exceed the following trip setpoint power by more than 0.7% RTP.

TD = BI (P) 3 + B2(P) 2 + B3(P) + B4 Where: P = RCS Loop AT Equivalent to Power (%RTP), P 5 50% RTP

TD = Time delay for Steam Generator Water Level Low-Low Reactor Trip (in seconds).

Bi = -0.007128 sec/(RTP) 3

B2 = +0.8099 sec/(RTP) 2

B3 = -31.40 sec/(RTP)

B4 = +464.1 sec

DIABLO CANYON - UNITS 1 & 2 3.3-18 Unit 1 -Amendment No. 1-35;143 Unit 2 - Amendment No. 135

" NCER UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 143 TO FACILITY OPERATING LICENSE NO. DPR-80

PACIFIC GAS AND ELECTRIC COMPANY

DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-275

1.0 INTRODUCTION

By application dated December 31, 1999, (Reference 1) as supplemented by letters dated January 18, July 7, September 22, and 29, and October 12, 2000, Pacific Gas and Electric Company (PG&E or the licensee) requested changes to the operating license and Technical Specifications (Appendix A to Facility Operating License No. DPR-80) for the Diablo Canyon Nuclear Power Plant, Unit 1. The proposed changes revise Section 2.C.(1) of Facility Operating License No. DPR-80 to authorize operation at reactor core power levels not to exceed 3411 megawatts thermal (MWt) (100 percent rated power). This amendment would also (1) revise the definition in Section 1.1 of the TS of rated thermal power to reflect Unit 1 operation at the uprated reactor core power level, (2) change the reactor core safety limits in TS Figure 2.1.1-1 to reflect the current fuel type, and (3) provide additional margin for Overtemperature AT (OTAT) and Overpressure AT (OPAT) setpoint calculations, and change the nominal full power Tavg in the OTAT and OPAT functions in Notes 1 and 2 to TS Table 3.3.1-1.

2.0 BACKGROUND

During the design of Diablo Canyon Units 1 and 2, although the reactors, structures, and auxiliary equipment are almost identical for the two units, the difference in the reactor internal design resulted in a lower coolant flow rate for Unit 1. The reactor coolant system (RCS) minimum thermal design flow for Diablo Canyon Unit 1 is 359,200 gpm as compared to the Diablo Canyon Unit 2 value of 362,500 gpm. Consequently, the license application reactor ratings were 3338 MWt for Diablo Canyon Unit 1 and 3411 MWt for Diablo Canyon Unit 2. These power levels included inherent margins since the design criteria and expected ultimate reactor core power was 3488 MWt for Diablo Canyon Unit 1 and 3568 MWt for Diablo Canyon Unit 2. The amendment would increase the Diablo Canyon Unit 1 reactor thermal power by 2.2 percent to the higher thermal power level of 3411 MWt permitted for Diablo Canyon Unit 2 and would result in identical power ratings for both units. The revised reactor thermal power level is within the initial design rating of Diablo Canyon Unit 1 and does not require physical modifications to Diablo Canyon Unit 1.

The supplemental letters dated July 7, September 22, and 29, and October 12, 2000, provided additional clarifying information, did not expand the scope of the application as originally

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noticed, and did not change the staff's original proposed no significant hazards consideration determination published in the Federal Register on April 19, 2000 (65 FR 21037).

3.0 EVALUATION

3.1 Reactor Systems

In support of the Diablo Canyon Unit 1 power uprate request, the licensee provided the Diablo Cayon Unit I power uprating program licensing report WCAP-14819 (Reference 2) and an addendum to WCAP-14819, respectively (Enclosures B and C to Reference 1). WCAP-14819, which was completed in 1997, provided the safety analysis and evaluation results to support operation of Diablo Canyon Unit 1 with a nuclear steam supply system (NSSS) vendor power level of 3425 MWt, consisting of 3411 MWt of core power and 14 MWt of reactor coolant pump heat. The safety analyses include accident analyses and evaluations, fluid and auxiliary systems, primary components, fuel, balance of plant systems, and turbine generator system analyses. In its analyses and evaluations, the licensee used the NSSS design parameters, which feature a reactor vessel outlet temperature of 610.1 OF, a steam generator tube plugging (SGTP) level up to 15 percent, and a thermal design flow (TDF) of 87,700 gallons per minute (gpm) per loop (or a total reactor coolant system (RCS) flow of 350,800 gpm). The addendum to WCAP-14819 addresses changes in the Diablo Canyon Unit 1 licensing basis since the WCAP was written.

3.1.1 LOCA Analyses

3.1.1.1 Large Break LOCA

The large break LOCA (LBLOCA) analysis in support of the Diablo Canyon Unit 1 power uprate is documented in detail in WCAP-14775 (Reference 3), which was submitted by the licensee in its letter of July 7, 2000 (Reference 4). The licensee's analysis uses the Westinghouse best estimate (BE) LBLOCA evaluation model (EM) described in Topical Report WCAP-12945-P-A (Reference 5), which has been accepted by the NRC for use at Diablo Canyon Unit 1 (Reference 6). This BE LBLOCA methodology complies with the requirements of 10 CFR 50.46, by using a realistic calculation of LOCA behavior, accounting for uncertainties in the analysis. The BE LBLOCA methodology uses the WCOBRAITRAC computer code for calculation of the transient response to an LBLOCA event. The BE LBLOCA methodology follows the basic steps in the code scaling, applicability and uncertainty (CSAU) methodology developed by the NRC (Reference 7). It includes a detailed treatment of the uncertainties associated with the computer code models used to analyze the accident scenario, and the uncertainties associated with plant operation, which provides initial condition and power distributions input to the analysis.

It is noted that the LBLOCA analysis in WCAP-14775 was performed to support 24-month fuel cycles for both Diablo Canyon Units 1 and 2, and support Diablo Canyon Unit 1 power uprate. Application of the analysis to Diablo Canyon Unit 2, 24-month fuel cycle, however, is beyond the scope of this review, and this evaluation is restricted to the analysis in support of Diablo Canyon Unit 1 power uprate.

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The fuel analyzed is 17x17 VANTAGE 5 fuel with ZIRLO cladding. Since Diablo Canyon Units 1 and 2 have different flow characteristics due to different reactor internals designs in the areas of upper and lower support plates and thermal shield, the licensee developed two separate models for the Diablo Canyon Unit 1 and Diablo Canyon Unit 2 reactor vessels to determine the most limiting BE LBLOCA analysis. The licensee determined that the Diablo Canyon Unit 1 BE LBLOCA analysis was bounding for Unit 1 uprate power conditions. The BE LBLOCA analysis for Diablo Canyon Units 1 and 2 documented in WCAP-14775 calculated the 95th percentile peak cladding temperature (PCT) of 1976 °F, maximum cladding oxidation of 11 percent, and maximum total hydrogen generation of 0.89 percent. These results are within the respective acceptance criteria of 2200 OF, 17 percent, and 1 percent specified in 10 CFR 50.46.

Subsequently, in letters of July 24, 1998 (Reference 8) and April 5, 2000 (Reference 9), the licensee identified two errors in the analysis that affect PCT calculations: an intercell force gap numbering error and a vessel channel DX error in connection with the calculation of momentum in the reactor vessel. The licensee assessed the effects of the errors and determined that the PCT would be 2009 OF with the errors corrected. In response to staff questions (Reference 10), the licensee provided a description of the two errors and how their impacts on the LOCA analysis results were assessed. The intercell force gap numbering error occurs when sequential numbers are skipped in the input deck, such that the highest numbering gap exceeds the total number of gaps specified with the code variable "NK." Therefore, the gaps with numbers that exceed NK do not have the intercell interphase drag force cleared at the end of a time step, resulting in an incorrect force applied in the gap momentum equation. The licensee performed a plant-specific reanalysis of the reference transient with the input error corrected, and the result showed increases of 33 OF and 67 OF, respectively, for the Reflood 1 and Reflood 2 PCTs. The vessel channel DX error occurs for any section in the reactor vessel that is modeled using multiple levels. In these sections, an error in the computer code caused the cell height used to calculate the interfacial drag and wall friction for lateral flow to be erroneously based on the cell height of the top node in the section below. A plant-specific reanalysis of the reference transient with the code error corrected showed that the effect of the code error correction was a minor decrease in the Reflood 1 PCT, which was conservatively assessed as a 0 OF impact, and a large decrease in Reflood 2 PCT, which offset the 67 OF penalty calculated for the gap numbering error. Therefore, with the correction of both errors, the licensee determined that Reflood 1 with a 33 OF error correction is more limiting, and the PCT is 2009 OF. By letter dated October 12, 2000, the licensee has committed to perform a reanalysis of the LBLOCA by July 26, 2003. Based on the margin between the PCT of 2009 OF and the acceptance criterion of 2200 OF, the staff concludes that the PCT is sufficiently below the acceptance criterion to justify its proposed schedule for reanalysis. This satisfies the 10 CFR 50.46(a)(3)(ii) requirements to estimate the effects of these errors on the PCT and to propose a schedule for reanalysis.

3.1.1.2 Small Break LOCA

Section 3.1.2 of WCAP-14819 describes a small break LOCA (SBLOCA) analysis performed with the Diablo Canyon Unit 1 power uprate of 3411 MWt. The analysis is performed with NRCapproved SBLOCA EM, which uses the NOTRUMP code to calculate the transient mass and energy release of the fluid through the break, as well as the depressurization of the RCS. The approved NOTRUMP EM includes an improved condensation model for the pumped safety

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injection (SI) into both the broken and intact loops. The VANTAGE 5 fuel design with the annular fuel pellet blanket was explicitly modeled. Fuel cladding thermal analyses are performed with the LOCTA-IV code using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core st.dam flow and mixture heights as boundary conditions.

The analysis is performed with the following key assumptions: (1) the most limiting single active failure of a safety system power supply train, resulting in the loss of one complete train of ECCS components; (2) a loss of offsite power (LOOP) coincident with reactor trip; (3) a total 4.7 seconds delay time from the reactor trip signal to the full rod insertion for the rod cluster control assembly (RCCA); (4) the most reactive RCCA stuck in the full out position; (5) degraded highand intermediate-head SI pumps, delivering only the minimum ECCS flow; and (6) a total 27second SI delay to account for signal initiation, diesel generator startup and emergency power bus loading, as well as the time involved in aligning the valves and bringing the SI pumps up to full speed.

The break spectrum analysis is performed based on the Westinghouse SBLOCA ECCS EM generic study and previous Diablo Canyon SBLOCA analyses results, which conclude that the cold leg break is the limiting break location. The analysis considered the effect of the RCS operating temperature by performing a sensitivity study of an RCS temperature window bounded by high Tavg of 582.3 OF and low Tavg of 560 OF to bound a plant operating range and uncertainty. This RCS temperature window covers the nominal Tvg of 577.3 OF for the OTAT and OPAT reactor trip functions. A spectrum of cold leg break sizes is analyzed with high T,,g conditions to determine the limiting break size. A sensitivity run is then made with the low Tag conditions for the limiting break to verify the high Tavg conditions are limiting. The analysis found the limiting break to be a 3-inch diameter cold leg break. The resulting PCT is 1304 OF, well below the acceptance criterion of 2200 OF specified in 10 CFR 50.46.

3.1.2 Non-LOCA Safety Analyses

The majority of the currently applicable non-LOCA safety analyses for Diablo Canyon were performed using bounding assumptions for important plant parameters that envelop both Units 1 and 2. Therefore, the safety analyses of the majority of non-LOCA events, except for the three items identified below, need not be re-analyzed for the Diablo Canyon Unit 1 power uprate conditions. As indicated in Section 3.2.1 and Table 3.2-1 of WCAP-14819, these nonLOCA events are either (1) not affected by an uprate full power condition because they are either analyzed at no-load conditions or do not directly assume the specific core power, or (2) bounded by the currently applicable analyses because at-power safety analyses currently assume the lower design RCS flow rates associated with Diablo Canyon Unit 1 in combination with the higher licensed core and NSSS power and coolant average temperature of Diablo Canyon Unit 2. The only exceptions are the OTAT and OPAT reactor trip setpoint calculations, accidental depressurization of the RCS, and steamline break at full power. In order to support the Unit 1 uprate conditions, these items have been re-analyzed.

3.1.2.1 Reactor Core Safety Limit and OTAT and OPAT Reactor Trip Setpoint Calculations

TS 2.1.1 specifies the reactor core safety limits, depicted in TS Figure 2.1-1, which defines a region of permissible operation in terms of thermal power, and reactor coolant pressure and

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average temperature to prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. The core safety limits are that: (1) the loci of points of thermal power, RCS pressure, and average temperature for which the minimum departure from nucleate boiling ratio (DNBR) is not less than the safety analysis limit, (2) the fuel centerline temperature remains below melting, (3) the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or the exit quality is within the limits defined by the critical heat flux (CHF) correlation. The existing core safety limits, shown in TS Figure 2.1-1, were developed for the 17x1 7 standard fuel, which is more limiting than 17x17 VANTAGE 5 fuel currently used in both Diablo Canyon Units 1 and 2.

To ensure compliance with core safety limits, the OTAT and OPAT trip functions, respectively, provide protection of the core against exceeding the design DNBR limit and hot leg boiling, and fuel rod design limits under all possible overpower conditions. Westinghouse Topical Report WCAP-8746-A (Reference 11) describes the methodology used for the development of the OTAT and OPAT reactor trip setpoints. Though Diablo Canyon Units 1 and 2 use the same OTAT and OPAT trip setpoints (via the constants in the OTAT and OPAT functions), the calculations to confirm the acceptability of these setpoints are performed separately for the specific plant operating conditions of each unit.

The current OTAT and OPAT reactor trip setpoint constants were calculated based on the reactor core safety limits of TS Figure 2.1.1 -1 for the 17xl 7 standard fuel. The licensee has determined that there is insufficient departure from nucleate boiling (DNB) margin available to support the current setpoints assuming 17x17 standard fuel for Diablo Canyon Unit 1 uprate conditions. Therefore, the licensee has proposed to revise TS Figure 2.1.1-1, "Reactor Core Safety Limits."

The Diablo Canyon Unit 1 power uprate program includes reloads with 17X17 VANTAGE 5 fuel assembly design (Reference 16) and fully enriched annular axial blanket fuel pellets at the top and bottom of the fuel. The revised slightly higher core safety limits were developed based on exclusive use of 17x1 7 VANTAGE 5 fuel, and the uprated Unit 1 power conditions with higher Tavg of 577.3 OF. With the revised core thermal limits, the licensee performed setpoint calculations using the approved method of WCAP-8746-A. The results verify that the present OTAT and OPAT trip setpoint constants and associated axial power distribution f(AI) penalty function provide adequate protection for the revised core limits at the Diablo Canyon Unit 1 uprated power conditions.

Since these OTAT and OPAT reactor trip setpoints are developed for the 17xl 7 VANTAGE 5 fuel only, use of other fuel design is not permitted unless proper evaluation is performed to justify its use.

3.1.2.2 Accidental RCS Depressurization

An accidental depressurization of the RCS is assumed to be the result of a failed open pressurizer safety valve, which is designed to relieve approximately twice the steam flow rate of a relief valve.

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As part of the Diablo Canyon Unit 1 power uprate evaluation, a new analysis of accidental RCS depressurization is performed for the VANTAGE 5 fuel type instead of the limiting standard fuel, and uses conservative assumptions that bound both units. The transient is analyzed with the approved LOFTRAN code (Reference 12) and improved thermal design procedure (Reference 13). The analysis is made with the following conservative assumptions in the reactivity feedback calculation: (1) a positive moderator temperature coefficient for beginning of life (+7pcm/ *F) and a low (absolute value) Doppler-only power coefficient to provide a high amount of positive reactivity feedback and maximize any power increase, and (2) no consideration of the spatial effect of voids due to local and subcooled boiling, which would tend to flatten the core power distribution.

The results, shown in Reference 2, Figures 3.2.3-1 and 3.2.3-2 and Table 3.2-2, show that the DNBR remains above the safety analysis limit value throughout the transient, and is acceptable.

3.1.2.3 Steam Line Break at Full Power

A steam line break from the full power initial condition was not documented in the Final Safety Analysis Report (FSAR), but was performed to support an assumed increase in the OPAT trip response time for the resistance temperature detector (RTD) bypass elimination modification in conjunction with the Eagle 21 process protection system upgrade (Reference 14). Since this analysis did not bound the Unit 1 uprate conditions, a new steam line break analysis was performed using conservative assumptions that bound both units. Section 3.2.4 of WCAP14819 provides a summary of the new analysis for the steamline break core response at power event for the Diablo Canyon Unit 1 power uprate.

In response to a staff question, the licensee provided the analysis of the steamline break core response at power event for Diablo Canyon Unit 1 power uprate. The analysis uses the LOFTRAN code for the calculation of the overall plant behavior of the steamline break event, the THINC code for the core thermal hydraulic analysis, as well as the improved thermal design procedure. The same key core kinetics reload parameters (i.e., moderator temperature coefficient, Doppler coefficient, delay neutron fraction) and power distribution limits that established the bounding analysis limits for the hot zero power steam break case are used for the at power case. A spectrum of break sizes at the uprate conditions is analyzed to determine the most limiting conditions based on which reactor trip or safety injection signal is reached. The break spectrum analysis determined that the 0.53 ft2 case established the limiting DNBR since it was the largest steamline break at power case that did not generate a low steam pressure safety injection signal. For larger break sizes, a low steam pressure safety injection signal is generated within a few seconds and the power increase and minimum DNBR are less limiting. The minimum DNBRs for the 0.53 ft2 case are calculated to be 2.062 and 2.143 for the Cycle 10 thimble and typical cells, respectively, which meet the applicable DNBR limits. The licensee states that the acceptability of the calculated core design values with respect to these analysis limits will be verified for each core reload as part of the reload safety evaluation process of WCAP-9272-P-A. The licensee also commits to include a discussion in the FSAR Update summarizing how the steamline break at power core response has been analyzed to verify that the minimum DNBR that occurs prior to a reactor trip or safety injection remains within the appropriate limit.

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3.1.3 Steam Generator Tube Rupture

Section 3.4 of WCAP-14819 provides an evaluation of the impact of the Diablo Canyon Unit 1 power uprate on the steam generator tube rupture (SGTR) design basis to provide margin to steam generator overfill. The original SGTR overfill analysis for Diablo Canyon was documented in WCAP-11723 (Reference 15). A reanalysis, documented in PGE-92-685 and the Diablo Canyon FSAR Section 15.4.3, was performed with revised input assumptions related to the auxiliary feedwater flow rates and the steam generator (SG) power operated relief valve relief capacity. This reanalysis with the enhanced modeling of the turbine driven auxiliary feedwater (AFW) pump and AFW control valves as a function of SG level resulted in a net gain in overfill margin from the original 61 ft3 to 109 ft 3.

Both the original analysis and the reanalysis were performed to bound both Diablo Canyon units, by using the limiting parameters for either Unit 1 or Unit 2 such that the analysis is applicable for both units. The analyses were performed at 102 percent of 3423 MWt with a nominal TDF of 88,500 gpm per loop with up to 15 percent SG tube plugging and a RCS average temperature of 577.6 °F, which are slightly different from the Diablo Canyon Unit 1 power uprate conditions of 3425 MWt (i.e., RTP of 3411 MWt and RCS pump heat of 14 MWt), 87,700 gpm and 577.3 OF, respectively. The licensee has determined that the impact of the 2 MWt difference in thermal power (< .06 percent) and 800 gpm (< 1 percent) difference in loop flow rate are insignificant for the SGTR event, which is not a DNB related transient. The slightly higher RCS average temperature of .3 OF assumed in the analysis is conservative, as it results in an earlier reactor trip and, therefore, earlier auxiliary feedwater actuation for the SG overfill, as well as earlier steam releases to the environment for the offsite dose calculation. Therefore, these SGTR analyses are applicable to the Diablo Canyon Unit 1 power uprate conditions, and the conclusions of the FSAR remain valid for the Diablo Canyon Unit 1 power uprate.

3.1.4 lmpacts of Power Uprate on Primary Fluid Systems Functions

Section 4.1 provides an evaluation of the reactor coolant system and auxiliary systems to confirm their continued ability to meet the applicable design basis functional and performance requirements at the Diablo Canyon Unit 1 power uprate. For the RCS, the licensee performed a review of the "proof-of-design" calculations for the pressurizer relief tank low and high level alarm setpoints, pressurizer relief line pressure drop, pressurizer spray flow capability, and RCS loop pressure drops. The evaluation determined that the original calculations for these items remain applicable to the uprated conditions.

Section 4.1.2 describes the evaluation of the residual heat removal system (RHRS) capability to confirm that the RHRS can cool the RCS within the allowable limits based on the higher decay heat load associated with the uprated power level. The allowable cooldown limits are Westinghouse RHRS design criteria related to the sizing of RHRS heat exchanger that: (1) 2 RHR trains can cool the RCS to 140 OF within 20 hours after shutdown, and (2) one RHR train can cool the RCS to 200 OF within 36 hours after shutdown. The normal two-train and single-train cooldown performance of the RHRS was analyzed with the major system parameter inputs, e.g., the auxiliary heat loads on the component cooling water system (CCWS), applicable to the power uprated conditions. The results of this evaluation indicate that the RCS cools down to 140 °F at 17.4 hours after shutdown using two cooling trains. With only one

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cooling train, the RCS cooldown to 200 °F is achieved at 29.2 hours based on the design CCWS flow rate to the RHRS heat exchanger of 5000 gpm, and at 36 hours based on the minimum CCWS flow rate of 4076 gpm. Therefore, the RHRS remains capable of cooling the RCS to the required final temperature in the allowable length of time at the uprated conditions.

3.1.5 Overpressure Protection

Section 7.15 of WCAP-14819 provides an evaluation of the adequacy of primary/secondary overpressure protection at Unit 1 uprate conditions. Two FSAR design basis events, i.e., loss of load/turbine trip and locked rotor events, are evaluated. These overpressure events rely on the pressurizer safety valves (PSVs) and the main steam safety valves (MSSVs) for mitigation of the primary and secondary system overpressure challenges. As discussed in Section 3.2 of this safety evaluation, the safety analyses of the majority of non-LOCA events, including loss of load/turbine trip and locked rotor, for the Diablo Canyon Unit 1 power uprate conditions are bounded by the existing applicable analyses in the FSAR. This is because the existing analyses were performed with the Diablo Canyon Unit 2 power level, which bounds the Diablo Canyon Unit 1 power uprate conditions. Therefore, the Diablo Canyon Unit 1 power uprate has no impact on the PSVs and MSSVs for overpressure protection.

3.1.6 Fuel Design

The Diablo Canyon Unit 1 power uprate program includes reloads with 17X1 7 VANTAGE 5 fuel assembly design (Reference 16) and fully enriched annular axial blanket fuel pellets at the top and bottom of the fuel. The annular pellet blanket incorporates a cylindrical void in the center of the pellet to increase the available volume inside the fuel rod to gain margin for the internal pressurization that occurs as a result of fuel bumup. As described in Section J of WCAP12610-P-A (Reference 17), the annular pellet blanket is an optional feature, of the VANTAGE+ fuel design, which evolved from, and is comparable with, the VANTAGE 5 design with a new zircaloy cladding, known as ZIRLO. The licensee, in response to a staff question, stated that the fully enriched fuel pellet blanket design does not represent a new fuel design feature but has been in use at Diablo Canyon Unit 1 with the VANTAGE 5 fuel since Cycle 9 (April 1997), and was evaluated in accordance with 10 CFR 50.59 for the Cycle 9 core reload design using the reload safety analysis methodology of WCAP-9272-P-A (Reference 18).

The safety analyses did not explicitly model annular pellet blankets because they do not directly impact any fuel performance design feature assumed in these analyses. The potential impact of the annular pellet blanket fuel on the safety analyses results was evaluated with respect to its effects on the axial and radial power distributions within the core, and the evaluation through the power shape sensitivities performed as part of the normal reload safety analysis process. The evaluation concluded that the annular pellet blankets did not impact any non-LOCA, thermal hydraulic, or large break LOCA analysis results. These conclusions were based on the fact that the analysis results were not affected by the explicit internal design of the fuel pellet, but used limiting heat flux power distribution assumptions that were bounding for the calculated core design values. The evaluation did assign a 10 OF peak cladding temperature (PCT) penalty to the existing small break LOCA results to conservatively bound any potential impact of enriched annular pellet blankets. As discussed in Section 3.1.2 of this report, however, the small break LOCA analysis for the power uprate evaluation was performed with the annular

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pellet blankets explicitly modeled. Therefore, the staff concludes that the use of annular pellet blankets associated with the power uprate has been properly accounted for, and is acceptable.

Section 6.0, "Fuel Design," of WCAP-14819 describes the results of evaluations performed of the fuel for Diablo Canyon Unit 1 under the uprating program in the areas of fuel rod and fuel assembly structural integrity for the uprating conditions with ZIRLO cladding and annular pellet blankets. The evaluation included the impact of the NSSS performance parameters of power uprate on the ability of fuel to satisfy fuel rod design criteria and fuel assembly structural integrity. Using the fuel performance analysis code, PAD 3.4 (Reference 19), and the power uprate conditions and ZIRLO cladding, the results indicate that the rod internal pressure criterion and the clad stress limit are met. The licensee also performed clad oxidation analyses against clad corrosion design basis to determine whether excessive fuel clad oxidation can occur that results in fuel system failure. It concludes that sufficient margin exists to the cladding corrosion-related criteria at the longer cycle lengths with ZIRLO clad. However, the use of Zirc-4 clad fuel will require cycle-specific analysis to confirm its compliance to the new cladding corrosion model currently under development.

The licensee also performed an evaluation of the effect of power uprate on fuel assembly structural integrity. It concludes that the uprate has a negligible effect on reactor vessel and intemals LOCA forces, and does not increase the operating and postulated transient loads such that they will adversely affect the fuel assembly functional requirements. The staff concludes that the fuel design has been properly evaluated for operation under the Diablo Canyon Unit 1 power uprate conditions.

3.2 Electrical and Instrumentation and Controls

3.2.1 Electrical

The Diablo Canyon Unit 1 turbine is a 1137 MWe Westinghouse tandem compound turbine generator. The generator is a hydrogen-cooled unit with a water-cooled stator winding and is rated at 1300 MVA, 0.9 power factor. Its capacity is determined by its cooling system, the excitation design, and physical limitations of the windings. The maximum MW electric rating depends on the characteristics, described by mega volt-ampere reactives (MVARs) generated (overexcited or boosting) or absorbed (underexcited or bucking) by the generator. The licensee studied the generator and excitation system to determine the impact of the increase in thermal rating from 3350 MWt to 3425 MWt, which corresponds to a maximum 1163 MW electrical. At 1160 MWe, the generator curve envelops down to 0.9 power factor leading and 0.95 power factor lagging. The licensee states that the actual operation of the unit is very close to a unity power factor due to local grid characteristics. The increased power uprate of 2.2 percent is estimated to produce an additional 23 MWe, i.e., the electrical power will increase from 1137 MWe to 1160 MWe. This change will result in identical power ratings for both units. The revised reactor core thermal power level is within the initial design rating of Diablo Canyon Unit 1 and does not require physical modifications to the unit.

The balance-of-plant electrical loads (such as condensate booster pumps - to supply a higher flow rate) will increase about 1 percent over the current loads, which is within the design rating of the auxiliary power system. The licensee reviewed the impact of the uprate on electric grid

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stability and determined that it would have a negligible impact on the grid stability. The electrical systems are sized for the full loading of plant equipment. Since no change to the electrical components is required for the Diablo Canyon Unit 1 uprating, the capacity of the electrical system to respond to the uprate is assured.

The startup transformer receives power from the offsite power grid and supplies the electrical distribution system for startup and shutdown of the plant. The startup transformer is sized for full loading of the plant equipment. Since no change to the electrical components is required for the Diablo Canyon Unit 1 uprating, the adequacy of the startup transformer continues to be assured.

Main transformers deliver 96 percent of the generated power from the plant to the 500 kV grid (the additional 4 percent of the generator output makes up the house loads.) In addition, under certain plant configuration, the main transformers are used to backfeed power from the 500 kV switchyard (offsite power) to the plant electrical buses. The ratings of the main transformers remain unchanged at 1320 MVA. They can supply higher loads up to 1478.4 MVA, though this results in a temperature rise and accelerated aging. Since the transformers that see 96 percent of the generator output are rated at 1320 MVA, and the Diablo Canyon Unit 1 generator is rated at 1300 MVA, the transformer will never be the limiting equipment in a power uprate. The proposed Diablo Canyon Unit 1 power uprate will not challenge the transformer capability.

There are three emergency diesel generators (EDGs) at Diablo Canyon Unit 1 that provide vital power for a safe shutdown in the event of a loss of the preferred power source. Any transient that results in the need for EDG actuation will also result in a reactor trip. With the unit tripped, the electrical loads are identical to the non-uprated condition; thus, the emergency loads to the EDGs are not affected by the power uprate.

3.2.2 Environmental Qualification

The licensee evaluated the safety-related electrical equipment to ensure qualification for the normal and accident conditions expected in the areas in which the devices are located. Environmental qualification (EQ) is based on expected temperature and pressure resulting from accident conditions. For example, components within the containment that are relied upon for a LOCA or main steam line break (MSLB) mitigation, must be qualified to perform their function in the hot, moist, and potentially radioactive post-accident atmosphere (up to 47 psig, 100 percent humidity, and about 250 OF for LOCA and 347 OF for MSLB).

Currently, the EQ curves of pressure, temperature and radiation levels are identical between Diablo Canyon Units 1 and 2 and are based on a Unit 2 model. The Diablo Canyon Unit 2 model is based on the higher power level that the Diablo Canyon Unit 1 uprate program is seeking to justify. Since the Diablo Canyon Unit 1 electrical equipment was qualified to the Diablo Canyon Unit 2 pressure, temperature and radiation levels, the EQ curves continue to be applicable to Diablo Canyon Unit 1 in the uprated condition. Therefore, the power uprate of D:,ablo Canyon Unit 1 will not require any modification to the EQ plans for electrical equipment.

The proposed power uprate for Diablo Canyon Unit 1 is similar to the Diablo Canyon Unit 2 uprated power level at which it currently operates. Moreover, this request for power uprate is

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similar to that granted in 1986 for Salem Unit 1, where the power level was increased from 3338 MWt to 3411 MWt to be consistent with Salem Unit 2, which was operating at 3411 MW.

Based on the above evaluation, the staff concludes that the licensee has provided reasonable assurance that the safety functions as well as the environmental functions of the electrical power system will be maintained and would have a negligible impact on grid stability. This is consistent with General Design Criterion (GDC) 17 and the proposed change is, therefore, acceptable.

3.2.3 Instrumentation and Controls

The following instrumentation TS changes related to power uprate have been proposed by the licensee: (1) overtemperature AT (Table 3.3.3-1, page 3.3-17), (2) change nominal full power Tavg from 576.6 °F to 577.3 OF, (3) overpower AT (Table 3.3.3-1, page 3.3-18), and (4) change nominal full power Ta,9 from 576.6 0F to 577.3 0F.

The staff reviewed the allowable value and nominal trip setpoint changes for the identified instrumentation.

The OTAT and OPAT functions have setpoints that are automatically varied based on both reactor core power level and reactor coolant system flow rate. The uprated OTAT and OPAT setpoint calculations for Diablo Canyon Unit 1 use the same methodology that was used for the pre-uprate setpoint calculations. This methodology is documented in Westinghouse Licensing Topical Report WCAP-8745-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," dated September 1986. The staff has previously reviewed and accepted this generic topical report.

The proposed OTAT and OPAT allowable value and nominal trip setpoint changes resulting from the power uprate are intended to maintain the existing margins between operating conditions and reactor trip setpoints and do not significantly increase the likelihood of a false trip nor failure to trip upon demand. Therefore, the existing licensing basis is not affected by the allowable value and nominal trip setpoint changes to accommodate the power uprate.

Based on the above evaluation, the staff concludes that the licensee's proposed instrumentation allowable value and nominal trip setpoint changes incorporated in the TS for power uprate are consistent with the Diablo Canyon Unit 1 setpoint methodology, licensing basis, and Westinghouse Standard Technical Specifications, and are, therefore, acceptable.

3.3 Radiological Consequences

The design-basis accident radiological consequences analyses described in the Diablo Canyon Unit 1 FSAR that are potentially affected by the change in reactor core thermal power are the LBLOCA, small break LOCA, SGTR, and MSLB. The original design criteria and expected ultimate reactor core thermal power level for Diablo Canyon were 3488 MWt for Unit 1 and 3568 MWt for Unit 2. The FSAR currently describes bounding design basis dose analyses for both units assuming the higher level of 3568 MWt. Since the current analyses assume a power

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level that bounds the requested uprate power, the licensee did not perform new dose calculations.

The licensee performed an evaluation of the radiological source terms to assess the impact of the increase in reactor thermal power. The current FSAR source term is based on a reactor core power level of 3568 MWt, which is approximately 105 percent of the proposed 3411 MWt rating. Therefore, the proposed power uprate has no impact on the assumed radiological source terms for the design-basis dose analyses. Core coolant activities used in the current FSAR design basis dose analyses also were based on a core thermal power level of 3568 MWt, which bounds the 102 percent rated power assumption for the proposed power level of 3411 MWt. The licensee further states that there is no impact on currently calculated radiological consequences of the SGTR since there are no changes in the calculated thermal and hydraulic analysis results of primary to secondary flow, flashed break flow, and steam released to the atmosphere. Similarly, the licensee states there is no impact on the currently calculated radiological consequences of the MSLB since the current design basis MSLB mass and energy releases and steam mass releases for dose considerations remain valid and bounding. Therefore, the staff concludes that design basis accident dose consequences of the proposed increase in thermal power are bounded by the current FSAR design basis dose analyses.

Based on the above discussion, the staff finds that the current Diablo Canyon design basis accident dose analyses, as described in the FSAR, remain bounding for the increase in reactor thermal power to 3411 megawatts thermal for Diablo Canyon Unit 1. The current FSAR dose analyses have previously been shown to meet the dose acceptance criteria given in NUREG0800 (Standard Review Plan), 10 CFR Part 100, and 10 CFR Part 50, Appendix A, General Design Criterion 19. Therefore, the staff finds the proposed power uprate for Diablo Canyon Unit 1 to be acceptable with regard to potential radiological consequences of design-basis accidents.

3.4 Plant Systems

The staff reviewed the licensee's submittals in the area of: (a) accident analyses and evaluations, (b) fluid and auxiliary systems, (c) balance of plant (BOP) systems, and (d) turbine generator systems and components review. The staff's evaluation of these areas is discussed below.

3.4.1 Accident Analyses and Evaluations

3.4.1.1 Containment Integrity Analyses

The design pressure for the Diablo Canyon containments is 47 psig. As part of the containment improvement program for both units, in 1993 the licensee re-performed the containment integrity analyses' including analyses for LOCA mass and energy releases; MSLB mass and

WCAP-13907, "Analysis of Containment Response Following Loss-of-Coolant

Accident for Diablo Canyon Units 1 and 2," December 1993.

WCAP-13908, "Analysis of Containment Response Following Main Steam Line Break Accident for Diablo Canyon Units 1 and 2," December 1993.

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energy releases; containment short-term subcompartment pressure responses; long term containment pressure and temperature responses; and environmental qualification. The analyses were performed based on the Diablo Canyon as-built specific information; new and more realistic containment analysis methodology; and the Diablo Canyon Unit 2 reactor licensed operating power of 3411 MWt. Results of the analyses indicated that the calculated containment peak pressures for the bounding LOCA2 and the MSLB 3 are 41.53.psig and 42.25 psig, respectively. These calculated containment peak pressures are well below the design pressure of 47 psig for the Diablo Canyon Units 1 and 2 containments.

Also, as part of the proposed power uprating program for Diablo Canyon Unit 1, the licensee performed a review to identify/evaluate any changes since 1993 that could affect the above calculated containment peak pressures. The licensee concluded that results of the containment integrity analyses performed in 1993 remain valid and bounding for Diablo Canyon Unit 1 operations at the proposed reactor power level.

Based on its review of the licensee's rationale and evaluation, the staff finds that the results of the licensee's 1993 containment integrity analyses would remain valid and bounding for Diablo Canyon Unit 1 operations at the proposed reactor power level of 3411 MWt.

3.4.1.2 Post-LOCA Hydrogen Generation

Following a LOCA, two independent electric hydrogen recombiners, each with 100 scfm, are available inside the containment to maintain the containment hydrogen concentration at or below the guidance of 4 percent by volume described in Regulatory Guide (RG)1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss-Of-Coolant Accident," for lower flammability limit. In addition, a containment hydrogen purge system serves as a backup system and provides the capability for controlled purging of the containment atmosphere in accordance with the guidance described in RG 1.7.

As part of the proposed power uprating program for Diablo Canyon Unit 1, the licensee reanalyzed the production and accumulation of hydrogen within containment based on plant operations at the current Unit 2 reactor licensed operating power of 3411 MWt, using the Diablo Canyon as-built specific information, and in accordance with the guidance described in RG 1.7. Results of the revised hydrogen production and accumulation analyses demonstrate that one

2 Double-ended hot leg (DEHL) break with minimum engineered safeguard

systems - 1 emergency core cooling train, 1 containment spray cooling train, and 2 containment fan coolers.

0.944 ft2 split break with the failure of a main steamline isolation valve and minimum engineered safeguard systems - 1 emergency core cooling train, 1 containment spray cooling train, and 2 containment fan coolers.

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hydrogen recombiner started when (approximately 4 days following a LOCA) the bulk containment hydrogen concentration reached 3.5 percent by volume. Therefore, one hydrogen recombiner will be adequate to ensure that the bulk containment hydrogen concentration is maintained at or below the RG 1.7 guidance of 4 percent by volume for lower flammability limit.

Based on its review of the licensee's rationale and evaluation, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the post-LOCA hydrogen control system.

3.4.2 Fluid and Auxiliary Systems

3.4.2.1 Containment Heat Removal Systems

The containment heat removal systems consist of the containment spray system (CSS), the containment fan cooler (CFC) system, and passive containment heat sinks such as containment walls, structures, and equipment located inside containment. The function of containment heat removal systems is to remove heat from the containment atmosphere to limit, reduce, and maintain at acceptably low containment pressure and temperature levels following a LOCA or MSLB. In addition, the CFC system is designed to maintain an acceptable environment inside the containment for equipment operation during normal plant operation.

3.4.2.1.1 Containment Spray System

The CSS, which consists of two independent trains, is designed in conjunction with containment fan cooler system and passive containment heat sinks to remove heat from the containment atmosphere to limit, reduce, and maintain at acceptably low containment pressure and temperature levels following a LOCA or MSLB. The CSS performs no function during normal plant operations.

3.4.2.1.2 Containment Fan Cooler System

The CFC system consists of five CFCs. During normal plant operations 4, four of the five CFCs are required to provide sufficient containment ventilation to maintain the average containment temperature below the TS limit of 120 OF. Following a LOCA or MSLB, two CFCs in conjunction with one CSS cooling train and the passive containment heat sinks are capable of maintaining the containment pressure below the design value.

The licensee stated that the CSS and the CFCs system are not affected by Diablo Canyon Unit 1 operations at the proposed reactor power level since these systems were initially designed based on the higher (Diablo Canyon Unit 2) reactor design power rating. Also, the major components for these systems were designed and procured identically for Diablo Canyon Units 1 and 2.

Impact of Diablo Canyon Unit 1 operations at the proposed reactor power level of 3411 MWt on the CFC system required for containment ventilation during plant normal operations is addressed in Section 4.6.

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As demonstrated in the containment integrity analyses performed in 1993, one CSS cooling train in conjunction with two CFC and the passive containment heat sinks is capable of maintaining the containment peak pressures of the bounding LOCA and the MSLB at 41.53 p.ig and 42.25 psig, respectively. These calculated peak containment pressures are well below the design pressure of 47 psig for the Diablo Canyon containments.

Based on the above discussion, the staff finds that Unit 1 operations at the proposed reactor power level of 3411 MWt will have no impact on the CSS and the CFC system.

3.4.2.2 Component Cooling Water System

The component cooling water (CCW) system was designed, based on the (higher) reactor design power rating of Diablo Canyon Unit 2, to supply cooling water to safety-related and nonsafety-related plant components during plant normal operations and following a LOCA or a MSLB. The licensee stated that the hardware in the CCW system is the same in Diablo Canyon Unit 1 as in Diablo Canyon Unit 2.

In 1995, CCW heatup analyses5 for various scenarios of system/component failure (e.g., failure of auxiliary saltwater (ASW) pumps, residual heat removal pumps, etc.) at Diablo Canyon were performed to determine the limiting CCW temperature transients. The analyses were performed, based on the Diablo Canyon as-built specific information, new and more realistic containment analysis methodology, and the Diablo Canyon Unit 2 reactor licensed operating power of 3411 MWt. Results of these CCW heatup analyses demonstrate that the CCW system can maintain its intended cooling function during plant normal operations and following a LOCA or a MSLB. The licensee stated that the results of these CCW heatup analyses remain valid and bounding for Diablo Canyon Unit 1 operations at the proposed reactor power level.

Based on its review of the licensee's rationale and evaluation, the staff finds that the licensee's 1995 CCW heatup analyses would remain valid and bounding for Diablo Canyon Unit 1 operations at the proposed reactor power level. Therefore, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the CCW system.

3.4.2.3 Auxiliary Salt Water System

The ASW system was designed, based on the design power rating of the Diablo Canyon Unit 2 reactor, to provide cooling water from the Pacific Ocean to the CCW heat exchangers. The licensee stated that, as in the case of the CCW system, the hardware in the ASW system is the same in Diablo Canyon Unit 1 as in Diablo Canyon Unit 2.

WCAP-1 4282, "Evaluation of Peak CCW Temperature Scenarios for Diablo Canyon Units 1 and 2," March 1995

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The licensee performed a study6 in 1992 to evaluate the ASW system cooling capability as a function of ocean temperature based on the Diablo Canyon as-built specific information and the Unit 2 reactor licensed operating power level of 3411 MWt. Results of the study demonstrate that ample margin exists in the ASW system to maintain its intended cooling function during plant normal operations and following a LOCA or a MSLB. In addition, the ability of the Diablo Canyon Unit 1 ASW system to adequately support Diablo Canyon Unit 1 operations at the proposed reactor power level is demonstrated in the 1993 containment integrity analyses and the 1995 CCW heatup analyses.

Based on its review of the licensee's rationale and evaluation, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the ASW system.

3.4.2.4 Spent Fuel Pool Cooling System

The spent fuel pool cooling (SFPC) system was designed to remove heat generated by spent fuel stored in the spent fuel pool. As a result of Diablo Canyon Unit 1 operations at the proposed reactor power level, the decay heat load in the spent fuel pool for any specific fuel discharge scenario will increase slightly. The licensee stated that Diablo Canyon Unit 1 operations at the proposed reactor power level will not impact the heat removal capability of the SFPC system, and that the ability of the SFPC system to adequately support Diablo Canyon Unit 1 operations after the uprating is demonstrated by the present ability of the identical SFPC system of Diablo Canyon Unit 2 to support Diablo Canyon Unit 2 operations at the same reactor power level.

Based on its review of the licensee's rationale, and since the SFPC system was designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the SFPC system.

3.4.2.5 Auxiliary Feedwater/Condensate System

The AFW system serves as a backup system for supplying feedwater to the steam generators in the event of loss of main feedwater supply. The AFW system consists of one full capacity turbine-driven and two half capacity motor-driven auxiliary feedwater pumps. The AFW pumps are normally aligned to take suction from the condensate storage tank and, if needed, can be aligned to the fire water storage tank.

The AFW condensate storage tank has a capacity of 425,000 gallons. The licensee performed evaluations of the effects of plant operations at the proposed reactor power on the AFW/condensate system. The AFW condensate inventory required for Diablo Canyon Unit 1 operations at the proposed power level was determined to be 216,000 gallons, which is bounded by the current TS requirement of 222,600 gallons.

6 WCAP-1 2526, "Auxiliary Salt Water and Component Cooling Water Flow and

Temperature Study for Diablo Canyon," June 1992.

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Based on its review of the licensee's evaluation, and since the AFW system was designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, the staff concludes that Diablo Canyon Unit I operations at the proposed reactor power level will have an insignificant or no impact on the AFW/condensate system.

3.4.2.6 Main Steam System

Diablo Canyon Unit 1 operations at the proposed reactor power will result in an increase of the steam mass flow by approximately 2.5 percent. As indicated in the above sections, Diablo Canyon systems/components were designed based on a reactor power rating of 3568 MWt which envelops the proposed reactor operating power of 3411 MWt for Unit 1. In any event, the licensee performed evaluations of the effects resulting from Diablo Canyon Unit 1 operations at the proposed reactor power level on the main steam system including the main steam isolation valves (MSIVs), steam generator power operated atmospheric relief valves (ARVs) main steam bypass valves (MSBVs) and MSSVs. The licensee concluded that the designed performance of the main steam system and its associated components is not affected by Diablo Canyon Unit 1 operations at the proposed reactor power level.

Based on its review of the licensee's rationale, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the main steam system and its associated components.

3.4.2.7 Condensate and Feedwater System

The licensee performed evaluations of the effects of Diablo Canyon Unit 1 operations at the proposed reactor power level on the condensate and feedwater system. The licensee stated that all the system operating conditions are bounded by the existing design conditions.

Since the condensate and feedwater system does not perform any safety related function and its failure will not affect the performance of any safety-related system or component, the staff did not review the impact of Diablo Canyon Unit 1 operations at the proposed reactor power level on the design and performance of this system.

3.4.2.8 Radioactive Waste Processing System

The solid, liquid and gaseous radwaste activity is influenced by the reactor coolant activity, which is a function of the reactor power. However, the current reactor coolant activities used in the radiological offsite doses analysis were based on Diablo Canyon Unit 2 reactor original design power rating of 3568 MWt which envelops the proposed reactor operating power of 3411 MWt for Diablo Canyon Unit 1. Therefore, the designed performance of the radwaste systems is not affected by the plant operations at the proposed reactor power level.

Since the radwaste systems were designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, which bounds the proposed reactor operating power level of 3411 MWt, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor operating power level will have an insignificant or no impact on the radwaste systems.

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3.4.2.9 Containment Ventilation System

As stated in the above Section 4.1.4, the CFC system is also designed to operate with four out of five CFC during normal plant operations to maintain the average containment temperature below the plant TS limit of 120 "F, an acceptable environment inside the containment for equipment operation during normal plant operation.

Diablo Canyon Unit 1 operations at the proposed reactor power level will not cause a significant increase in containment heat load during plant normal operations. The licensee stated that containment fan coolers have more than enough capacity to perform their function, as demonstrated by their ability to maintain adequate cooling of the Diablo Canyon Unit 2 containment with only three or four of the five fans normally running. Therefore, the designed performance of the containment ventilation system is not affected by Diablo Canyon Unit 1 operations at the proposed reactor power level.

Based on its review, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level do not change the design and operation aspects of the containment ventilation system, and will have an insignificant or no impact on the system.

3.4.3 Balance-of-Plant Systems

3.4.3.1 Main Steam System

See the discussion in Section 3.4.2.6.

3.4.3.2 Circulating Water System

The circulating water system is designed to remove the heat rejected to the condenser by turbine exhaust and other exhausts over the full range of operating loads during normal plant operations. Since it does not perform any safety function and its failure will not affect the performance of any safety-related system or component, the staff did not review the impact of Diablo Canyon Unit 1 operations at the proposed reactor power level on the designs and performances of this system.

3.4.3.3 Condensate System

See discussion in Section 3.4.2.7.

3.4.3.4 Feedwater System

See discussion in Section 3.4.2.7.

3.4.3.5 Auxiliary Feedwater System

See discussion in Section 3.4.2.3.

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3.4.3.6 Emergency Diesel Generator

Diablo Canyon Unit 1 has three emergency diesel generators, which provide vital power for safe shutdown in the event of a loss of the preferred power source. The licensee stated that emergency diesel generators and their associated systems are not affected by plant operations at the proposed reactor power level.

Based on its review and the fact that emergency diesel generators and their associated systems were designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have no impact on the emergency diesel generators and their associated systems.

3.4.3.7 Instrument Air/Nitrogen System

The licensee stated that the instrument air/nitrogen system is not affected by plant operations at the proposed reactor power level.

Since the instrument air/nitrogen system was designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, which bounds the proposed reactor operating power level of '3411 MWt, the staff finds that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the instrument air/nitrogen system.

3.4.3.8 Auxiliary Building Ventilation System

The safety function of the auxiliary building ventilation system is to ensure the temperature in the rooms containing safety related equipment is sufficiently low for proper equipment operation under all circumstances including accident conditions. The licensee performed evaluations and stated that the Diablo Canyon Unit 1 uprating does not result in an increase in the heat load in the auxiliary building beyond that of Diablo Canyon Unit 2.

Since the auxiliary building ventilation system was designed based on a higher (Diablo Canyon Unit 2) reactor design power rating, which bounds the proposed reactor operating power level of 3411 MWt, the staff concludes that Diablo Canyon Unit 1 operations at the proposed reactor power level will have an insignificant or no impact on the auxiliary building ventilation system.

3.4.4 Turbine Generator Systems and Components Review

3.4.4.1 Heat Balance

The licensee performed a new turbine heat balance analysis to demonstrate the performances of the turbine generator systems and components resulting from operations of the Diablo Canyon Unit 1 reactor at the proposed power level. Since the turbine heat balance analysis is not a safety related function, the staff did not review the impact of Diablo Canyon Unit 1 operations at the proposed reactor power level on the designs and performances of the turbine generator systems and components.

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3.5 Structural and Pressure Boundary Integrity

3.5.1 General

The staff reviewed the effects of the power uprate on the structural and pressure boundary integrity of the NSSS and the BOP systems. Affected components in these systems included piping, in-line equipment and pipe supports, the reactor pressure vessel (RPV), CSS, reactor vessel internals (RVI), SGs, control rod drive mechanisms (CRDM), reactor coolant pumps (RCP), and the pressurizer. The staff's evaluation is discussed below.

3.5.2 Reactor Vessel

The licensee reported that the power increase will result in changing the design parameters given in Reference 1, Enclosure B, Table 2.1-1. Table 2.1-1 provides a comparison of the current design parameters and the corresponding revised parameters for use in the power uprate analysis at Diablo Canyon Unit 1.

The licensee evaluated the reactor vessel for the effects of the revised design conditions in Table 2.1-1 for the most limiting vessel locations with regard to ranges of stress intensity and fatigue usage factors (CUFs) in each of the regions, as identified in the reactor vessel stress reports. The evaluations considered the worst load sets of operating parameters that were identified for the uprated power condition. The regions of the reactor vessel affected by the power uprate include outlet and inlet nozzles, the RPV (main closure head flange, studs, and vessel flange), CRDM housing, bottom head to shell juncture, core support pads and the instrumentation tubes. The licensee evaluated the maximum ranges of stresses and cumulative fatigue usage factors for the critical components at the core power uprated conditions. The evaluation was performed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section III, 1965 Edition, with addenda through the Winter 1966 to assure compliance with the code of record.

The calculated maximum stresses and the maximum CUFs for the reactor vessel critical locations are provided in Table 4-1 of Reference 20. The table shows that the maximum stresses are within the allowable limits, and the CUFs remain below the allowable ASME Code iiriit of 1.0. The licensee concluded that the current design of the reactor vessel continues to be in compliance with licensing basis codes and standards for the power uprated conditions. Based on its review, the staff agrees with this conclusion.

3.5.3 Reactor Core Support Structures and Vessel Internals

By letter dated September 29, 2000, the licensee provided the additional information requested by the staff with regard to the evaluation of the reactor vessel core support and internal structures. The limiting reactor internal components evaluated include the lower core plate, core barrel, baffle plate, baffle/barrel region bolts, and the upper core plate. The licensee indicated that because the Diablo Canyon Unit 1 reactor internal components were designed and fabricated prior to the existence of ASME Code, Section III, Subsection NG, the evaluation is in compliance with the design criteria as documented in the Diablo Canyon FSAR. However,

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the lower core structural integrity was evaluated in Reference 20 in accordance with the 1989 Edition of the ASME Code Section II1. This is acceptable to the staff, in accordance with 10 CFR 50.55'a.

The licensee evaluated these critical reactor internal components considering the revised design conditions provided in Table 2.1-1 of Reference 1. The reactor coolant pressure remains unchanged at 2250 psia. The existing NSSS design transients are bounding for the power uprate. The licensee indicated that, for the baffle-barrel region and the upper core plate, the current structural and thermal analyses of record for Diablo Canyon Unit 1 remain bounding for the power uprate condition. Reference 20 identifies the maximum calculated stress intensity and CUF for the lower core plates. The calculated stress is less than the Code allowable limit based on the 1989 Edition ASME Code Section III. The CUFs are less than the limit of 1.0. The remaining reactor internal components are less limiting. In addition, the licensee's evaluation indicated that the potential for flow induced vibration does not increase for the power uprate. As a result of its evaluation, the licensee concluded that the reactor internal components at Diablo Canyon Unit 1 will be structurally adequate for the proposed power uprate conditions. Based on its review, the staff agrees with the licensee's assessment.

3.5.4 CRDMs

The pressure boundary portion of the CRDMs are those exposed to the vessel/core inlet fluid. The licensee evaluated the adequacy of the CRDMs by reviewing the Diablo Canyon Unit 1 current CRDM design specifications and stress report to compare the design-basis input parameters against the revised design conditions in Table 2.1-1 of Reference 1 for the power uprate. The reactor coolant pressure remains unchanged at 2250 psia. The existing design basis analysis for the CRDM code pressure boundary uses 650 'F, which is higher than the vessel outlet CRDM temperature of 610.1 °F for the power uprate. The licensee concluded that the existing design of the CRDM components will remain adequate for the power uprate. Because the existing pressure and temperature design parameters bound the power uprate conditions, the staff finds the licensee's conclusion regarding the acceptability of the Diablo Canyon Unit 1 CRDMs for the proposed power uprate to be acceptable.

3.5.5 Steam Generators

The licensee reviewed the existing structural and fatigue analyses of the SGs at Diablo Canyon Unit 1 and compared the power uprate conditions with the design parameters of the Model 51 SGs stress reports. The comparison of key parameters is shown in Table 1-1 of Reference 20. The licensee indicated that the existing NSSS design transients remain bounding for the Diablo Canyon Unit 1 uprated conditions. For evaluation of the critical SG components, the key input parameters were used to calculate the stress and fatigue usage for the power uprate condition. The critical SG components that have been evaluated for the power uprate condition are the tubesheet, the channel head and the divider plate, tubes, secondary side nozzles, and the secondary manway. The evaluation was performed in accordance with the requirements of the ASME Code, Section III, 1965 Edition through the Winter 1965 Addenda, which is the Code of record for SGs at Diablo Canyon Unit 1.

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The calculated maximum stresses and cumulative fatigue usage factors for the critical SG components are provided in Tables 1-2 and 1-3 of Reference 20, respectively. The results indicate that the maximum calculated stresses are all below the Code allowable limits. The res,!ts provided in Table 1-3 also show that the calculated CUFs are within the allowable limit of unity for the 40 years service life with the exception that the replacement interval from the SG secondary manway bolts is reduced from 34 years for the current operating conditions to 31 years for the power uprate condition.

On the basis of its review, the staff concludes that the licensee has demonstrated the maximum stresses and CUFs for the critical SG components to be within the Code allowable limits and therefore acceptable for the power uprate. With regard to the SG secondary side manway bolts, in its September 29, 2000, response (Reference 20) to the staff's request for additional information, the licensee indicated that the reduced service life replacement interval of 31 years for the manway bolts for the power uprate has been incorporated into the Diablo Canyon plant maintenance procedure MP-M-7.25, "Removal and Installation of Steam Generator Manway Covers." This is acceptable to the staff.

3.5.6 Reactor Coolant Pumps

The licensee reviewed the existing design basis analyses of the Model 93A RCPs for the Diablo Canyon Unit 1 to determine the impact of the revised design conditions in Table 2.1-1 of Reference 1. The RCP pressure remain unchanged at 2250 psia. The licensee also indicated that the NSSS design transients remain unchanged for the power uprate. Table 2.1-1 shows that for both the current and the power uprate conditions, the RCP inlet temperature for Diablo Canyon Unit 1 remains at 544.2 OF, which is bounded by the temperature of 545 °F used in the current stress report for the Model 93A RCPs at Diablo Canyon and the temperature of 550 OF used in the generic vendor stress report. Thus, the proposed power uprate conditions are bounded by the existing design basis conditions for the Model 93A RPC at Diablo Canyon Unit 1.

On the basis of its review, the staff finds that the current RCPs, when operating at the proposed conditions with 2.2 percent power increase from the current rated power, will remain in compliance with the requirements of the codes and standards under which Diablo Canyon Units 1 and 2 were originally licensed and are, therefore, acceptable.

3.5.7 Pressurizer

The licensee evaluated the structural adequacy of the pressurizer and components for limiting locations at the pressurizer spray nozzle, the surge nozzle, and upper shell for operation at the uprated conditions. The Code used in the evaluation is the ASME Code, Section III, 1965 Edition, through Winter 1965 addenda, which is the same as the current Code of record for the Diablo Canyon Unit 1 pressurizer. The evaluation was performed by comparing the key parameters in the current Diablo Canyon Unit 1 pressurizer stress report against the revised design conditions in Table 2.1-1 for the proposed power uprate. The table shows that the uprate results in a small increase in the reactor coolant system Thot from 608.8 to 610.1 OF, which decreases the temperature difference between the pressurizer (653 OF) and any in-surge of hot leg fluid that flows through the surge nozzle. This will slightly reduce the stresses in the

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surge line and the other lower head components. The cold leg temperature for the power uprate condition is also slightly higher (0.1 OF) than that for the current power level, which decreases the temperature difference between the pressurizer and the cold leg fluid that flows through the spray nozzle into the pressurizer. Thus, the existing evaluations of the spray nozzle and upper shell components remain bounding for the power uprate. The staff agrees with the licensee's assessment.

Table 3-1 of Reference 20 provides the calculated maximum stresses and CUFs for the critical pressurizer components for plant operation following the power uprate. The table shows that the maximum stresses are within the Code allowable stress limits. The cumulative fatigue usage factors are less than the ASME allowable limit of 1.0. The staff concludes that the existing pressurizer and components will remain adequate for plant operation with the proposed 2.2 percent power increase while the RCS pressure remains unchanged.

3.5.8 NSSS Piping and Pipe Supports

The proposed power uprate of Diablo Canyon Unit 1 involves an increase in the temperature difference across the RCS. The licensee evaluated the NSSS piping and supports by reviewing the design basis analysis against the uprated power condition with regard to the design system parameters, transients and the loss-of-coolant-accident (LOCA) dynamic loads. The evaluation was performed for the reactor coolant loop (RCL) piping, primary equipment nozzles, primary equipment supports, and the pressurizer surge line piping. The methods, criteria and requirements used in the existing design basis analysis for Diablo Canyon Unit 1 were used for the power uprate evaluation. The evaluation used the Code of record, the ANSI B31.1 Power Piping Code, 1967, and various other ANSI and ASME Codes as specified in the Diablo Canyon FSAR.

At Diablo Canyon Unit 1, the RCL piping was designed on the basis of postulated large loop pipe breaks. In 1992, Diablo Canyon Unit 1 implemented the leak-before-break (LBB) methodology in WCAP-13039, "Technical Justification for Eliminating Large Loop Rupture as a Structural Design Basis for Diablo Cayon Units 1 and 2 Nuclear Power Plants." As such, the conservatism exists in the original design basis analysis that used the forcing functions due to postulated large loop pipe breaks. Therefore, the effects of the 2.2 percent power uprate on the NSSS piping will become insignificant in comparison to the inherent conservatism of the original design analysis without implementing the LBB technology. The steam pressure decreases for the power uprate. There are no changes with regard to the forcing functions for the postulated pipe break at the secondary side nozzles for the power uprate condition.

The RCS pressure remains unchanged for the proposed core power uprate. The actual hot leg temperature for the power uprate is projected to be slightly greater (1.3 OF) than the hot leg temperature at the current rated power level. The cold leg temperature for the power uprate condition is also slightly higher (0.1 OF) than that for the current power level. The change in crossover leg temperature is determined from the SG outlet temperature, which does not change. The licensee indicated that there is sufficient margin in the existing analysis to envelop the thermal forces and stresses associated with the temperature changes defined in Table 2.1-1 of Reference 1.

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The licensee also indicated that the design transients used in the evaluation of the RCS piping systems and equipment nozzles are unchanged for Diablo Canyon Unit 1 power uprate. The licensee concluded that the existing stresses and loads remain bounding for the power uprate for the NSSS components such as: the reactor cooling loop piping, the primary equipment nozzles, the primary equipment supports, pipe supports and the auxiliary equipment (i.e., the heat exchangers, pumps, valves and tanks).

On the basis of its review of the licensee's submittal, the staff concurs with the licensee's conclusion that the existing NSSS piping and supports, the primary equipment nozzles, the primary equipment supports, and the auxiliary lines connecting to the primary loop piping will remain in compliance with the requirements of the design basis criteria, as defined in the Diablo Canyon FSAR, and are acceptable for the power uprate.

3.5.9 BOP Systems and Safety-Related Components

The licensee evaluated the adequacy of the BOP systems and safety-related components based on comparing the existing design basis parameters with the core power uprate conditions. The BOP piping systems evaluated for the power uprate are the main steam, feedwater, SG blowdown, auxiliary feedwater, extraction steam, heater drains, condensate, chemical and volume control, secondary sampling, spent fuel pool cooling, residual heat removal, component cooling, station service water, and containment spray systems. The safety-related components include valves, pumps, power-operated relief valves and safety relief valves.

The licensee evaluated the affected systems on the basis of the uprated input parameters in Table 2.1-1 (for RCS temperatures, and steam temperature and steam flow rate) and the heat balance at 3411 MWt reactor thermal power. As a result, the licensee concluded that the Diablo Canyon Unit 1 existing non-NSSS systems, structures, and components continue to remain in compliance with all Diablo Canyon Unit 1 design basis requirements at the uprated power level of 3411 MWt.

The licensee also reviewed the programs, components, structures, and non-NSSS system issues as they are affected by the power uprate. In Reference 20, the licensee indicated that there are no changes to the PG&E motor-operated valve (MOV) program as a result of the proposed Diablo Canyon Unit 1 power uprate. The safety-related valves were not found to be impacted by the 2.2 percent power uprate and are, therefore, acceptable. This determination was confirmed by verifying that changes in system operating temperature, pressure and flow rate were bounded by the requirements of the associated equipment specification. As such, the increased thrust required to operate the MOVs due to expected differential pressure conditions is within the capabilities of the existing valve actuators. Additionally, in its response dated September 29, 2000, (Reference 20) to the staff's request for additional information, the licensee assessed the impact of the 2.2 percent uprate on the GL 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," MOVs and found them to be acceptable.

The licensee reviewed the impact of the proposed uprated power on the evaluations performed in response to GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related PowerOperated Gate Valves," associated with the pressure locking and thermal binding of the gate

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valves. The licensee found that there are no existing gate valve designs susceptible to the thermal binding and pressure locking at Diablo Canyon Unit 1 and, thus, the existing analysis remains valid for the 2.2 percent power uprate. The licensee reviewed the evaluation of their GL 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," program regarding the over-pressurization of isolated pipe segments. The licensee concluded that the existing evaluation for GL 96-06 is valid and bounding for the proposed Diablo Canyon Unit 1 power uprate. On the basis of the above review, the staff concurred with the licensee's conclusions that the power uprate will have no adverse effects on the safety-related valves and that the conclusions from the Pacific Gas & Electric GL 95-07, and GL 96-06, as well as GL 89-10 programs, remain valid.

As a result of the above evaluation, the staff concluded that the BOP piping, pipe supports and equipment nozzles, and valves remain acceptable and continue to satisfy the design-basis requirements for the power uprate.

3.5.10 Summary

Based on its review of the licensee's evaluations of the NSSS and BOP piping, components, and supports, the reactor vessel and internal components, the CRDMs, SGs, RCPs and the pressurizer, the staff finds the licensee's evaluations bounds the licensing code of record and the original design basis and, therefore, concludes the foregoing components are acceptable for Diablo Canyon Unit 1 uprate operations at the proposed core power level of 3411 MWt.

4.0 FSAR UPDATE

Enclosure G to the December 31, 1999, letter, contains mark-ups of the Diablo Canyon Final Safety Analysis Report Update. These mark-ups generally reflect the conditions associated with the Diablo Canyon Unit 1 power uprate so that both units have the same rated thermal power. In Chapter 4, "Reactor," changes were made to reflect that a full core of VANTAGE 5 fuel assemblies is used. It states that some of the Chapter 15 accident analyses, including large- and small-break LOCAs, assume an all VANTAGE 5 core and, therefore, it is not expected that LOPAR fuel will be used without further analysis. In Chapter 15, changes are made in the DNBR safety limit values and the OTAT and OPAT trip functions and the core thermal limits in Section 15.1.3 and Figure 15.1-1 to reflect the use of full core VANTAGE 5 fuel assemblies. Sections 15.2.13, "Accidental Pressurization of the Reactor Coolant System," and 15.3.1, "Small Break Loss of Coolant Accident," are replaced with the new analyses of the respective events. These changes are acceptable.

The current FSAR Update did not document the steamline break at full power and, therefore no revision of the FSAR Update is made for this event. In response to a staff question, the licensee committed to include a discussion in the FSAR Update summarizing how the steamline break at power core response has been analyzed to verify that the minimum DNBR that occurs prior to a reactor trip or safety injection remains within the appropriate limit. The staff concludes that this is acceptable.

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5.0 SUMMARY

The staff has reviewed the Diablo Canyon Unit 1 TS amendment related to the proposed power uprate to 3411 MWt. Based on thr Chapter 15 design basis transients and accidents safety analysis, the staff found the power uprate to be acceptable subject to the following condition:

Since (1) the reactor core safety limit, as well as the OTAT and OPAT trip setpoints for the uprate power conditions are determined based on full core of VANTAGE 5 fuel assemblies, and (2) the fuel design evaluation is performed only for the ZIRLO cladding, use of other fuel design or cladding material must be evaluated.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on October 25, 2000 (65 FR 63900).

Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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.

9.0 REFERENCES

1. Letter from D. H. Oatley (PG&E) to US Nuclear Regulatory Commission, "Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit 1, License Amendment Request 99-03, Unit 1 Reactor Core Thermal Power Uprate," PG&E Letter DCL-99-170, December 31, 1999.

2. WCAP-14819, "Pacific Gas and Electric Company Diablo Canyon Power Plant, Unit 1, 3425 MWt Uprating Program Licensing Report," Enclosure B, PG&E Letter DCL-99-170, December 31, 1999.

3. WCAP-14775, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for Diablo Canyon Power Plant Units 1 & 2 to Support 24-Month Fuel Cycles and Unit 1 Uprating," Westinghouse Electric Corporation, January 1997.

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4. Letter from D. H. Oatley (PG&E) to US Nuclear Regulatory Commission, "Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit 1, Response to NRC Request for Additional Information Regarding License Amendment Request (LAR) 99-03, Unit 1 Reactor Core Thermal Power Uprate," PG&E Letter DCL-00-098, July 7, 2000.

5. WCAP-1 2945-P-A, "Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis," Westinghouse Electric Corporation, June 1996.

6. Letter from S. Bloom (USNRC) to G. M. Rueger (PG&E), "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No. 1 (TAC No. M98827) and Unit No. 2 (TAC No. M98828)," February 13, 1998.

7. NUREG/CR-5249, "Quantifying Reactor Safety Margin: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-ofCoolant-Accident," 1989.

8. Letter from L. F. Womack (PG&E) to US Nuclear Regulatory Commission, "Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, Diablo Canyon Units 1 and

2, 10 CFR 50.46 Annual Report of Emergency Core Cooling System Evaluation Model Changes," PG&E Letter DCL-98-101, July 24, 1998.

9. Letter from G. M. Rueger (PG&E) to US Nuclear Regulatory Commission, "Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, Diablo Canyon Units 1 and 2, 10 CFR 50.46 3-day Report of Emergency Core Cooling System Evaluation Model Changes," PG&E Letter DCL-98-1 01, April 5, 2000.

10. Letter from D. H. Oatley (PG&E) to US Nuclear Regulatory Commission, "Docket No. 50-275, OL-DPR-80, Diablo Canyon Unit 1, Response to NRC Request for Additional Information Regarding License Amendment Request (LAR) 99-03, Unit 1 Reactor Core Thermal Power Uprate," PG&E Letter DCL-00-1 23, September 22, 2000.

11. WCAP-8746-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.

12. WCAP-7907-A, "LOFTRAN code Description," April 1984.

13. WCAP-8567-A, "Improved Thermal Design Procedure," February 1989.

14. WCAP-1 3615-R2, "Summary Report, Eagle 21 Process Protection System Upgrade for Diablo Canyon Power Plant Units 1 and 2," June 1993.

15. WCAP-1 1723, "LOFTTR2 Analysis for a Steam Generator Tube Rupture for the Diablo Canyon Power Plant Units 1 and 2," February 1988.

16. WCAP-10444-P-A, "Westinghouse Reference Core Report, VANTAGE 5 Fuel Assembly," September 1985.

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17. WCAP-1261 0-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," Westinghouse Electric Cor,"poration, April 1995.

18. WCAP-9272-P-A, "Reload Safety Evaluation Methodology,". Westinghouse Electric Corporation, July 1985.

19. WCAP-1 1873-A, WCAP-1 0851 -P-A, "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," August 1988.

20. Letter from D. H. Oatley (PG&E) to US Nuclear Regulatory Commission, "Diablo Canyon Power Plant Unit 1-Response to NRC Requests for Additional Information Regarding License Amendment Request 99-03, Unit 1, Reactor Core Power Uprate," September 29, 2000.

Principal Contributors: Y. Hsii, DSSA, NRR B. Marcus, DE, NRR M. Hart, DSSA, NRR N. Trehan, DE, NRR C. Wu, DE, NRR D. Shum, DSSA, NRR

Date: Octbber 26, 2000