ML20135B459

Reload Analysis Rept for San Onofre Nuclear Generating Station 2,Cycle 3

ML20135B459
Person / Time
Site:	San Onofre
Issue date:	09/05/1985
From:	SOUTHERN CALIFORNIA EDISON CO.
To:
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ML13303B407	List: ML20135B459
References
NUDOCS 8509110063
Download: ML20135B459 (172)

Text

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I RELOAO ANALYSIS REPORT FOR SONGS 2 CYCLE 3 i

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8509110063 850905 PDR ADOCK 05000361 P PDR

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TABLE OF CONTENTS I

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i Section g

1. Introduction and Sucmary 1-1 i

2. Operating History of the Reference Cycle 2-1

3. General Cescription 3-1 4-1

4. Fuel Syscem Oesign ,

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5. Nuclear Oesian

o. *hermal-Hydraulic Design 5-1 t

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7. Transient Analysis 7-1

3. ECOS Analysis 3-1 t

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9. Reactor Protection and Monitoring Systes 3-1 l

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10. Tecnnical Specifications 10-1 j

11. Startup Tes .ing 11-1 f, 4 i

12. References 12-1  ;

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.(,--) List of Tables s

3-1 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Core Loading 5-1 SONGS-2 Cycle 3 Nominal Physics Characteristics a

5-2 SONGS-2 Cycle 3 Limiting Values of Reactivity Worths and Allowances for Hot Full Power Steam Line Break, %ao End-of-Cycle (EOC) 4 5-3 SONGS-2 Cycle 3 Reactivity Worth of CEA Regulating Groups at Hot Cull Power, % 3, 6-1 SONGS-2 Cycle 3 Thermal Hydraulic Parameters at Full Power 7.0-1 SONGS Unit 2, Design Basis Events Considered in the Cycle 3 Safety Analysis i

7.0-2 DBEs Evaluated with Respect to offsite Oase Criterion 7.0-3 DBEs Evaluated with Respect to RCS Pressure Criterion 7.0 2 DBEs Evaluated with Respect to Fuel Performance

,s 7.0-5 OBEs Evaluated with Respect to Shutdown Margin Criterion 7.0-6 SONGS Unit 2, Cycle 3 Core Parameters Input to Safety Analyses 7.1.3-1 Key Parameters Assumed for the Increased Main Steam Flow Event 7.1.3-2 Sequence of Events for the Increased Main Steam Flod Event 31us a Single Failure 7.1.5b-1 Key Parameters assumed for the Steam Line Break' Event 7.1.5b-2 Sequence of Events' for the Hot Full Power, 7.41 ft2, inside Containment Steam Line Break with Loss of Offsite oower 7.2.6-1 Key Parameters Assumed for the Feedwater System Pipe Break Event 7.2.6-2 Sequence of Events for the Feedwater System Dipe Rreak Event 7.3.2-1 Xey Parameters Assumed for the Total loss of Forced Reactor Coolant Flow Event

i. 7.3.7-2 Sequence of Evants for Total loss of Forced Reactor Coolant Flow Event 7.3.3-1 Key Parameters Assumed for the' Single Peactor Coolant Duno Sheared Shaft Event

p. 7.a.1-1 '<ey ~ Darameters Assumed in the CEA withdrawal from Succritical

\ 1 Conditions Event G

7.4.1-2 Sequence , of Events for the -CEA Withdrawal from Subcritical Conditions Event II

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List of Tables  ;

(continued) 7.4.1-3 Key Parameters Assumed in - the CEA Withdrawal from Low Powers l- Event 7.4.1 4 Sequence of Events for the CEA Withdrawal from Low Powers Event 7.4.4-1 Key Parameters Assumed in the Inadvertent Boron Oilution Event 7.4.4-2 Results of the Inadver;ent Boron Dilution Event 7.7.1-1 Key Parameters Assumed for the loss of Load to One Steam l Generator Event j 7.7.1-2 Sequence of Events for the Loss of Load to One Steam Generator (

Event 8-1 SONGS Unit 2 Cycle 3 Core and System Parameters ,

R-2 SONGS Unit 2 Cycle 3 Limiting Break Size (1.0 DEG/PO)

S-3 SONGS Unit 2 Cycle 3 Variables Plotted as a Function of Time for the Lir :ing large Break 9-1 CPC System Software Algorithm Changes for Cycle 3 9-2 CPC System Addressable Constant Changes for Cycle 3 O

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List of Figures O 3-1 Enrichment Zoning Pattern for SONGS Batch E Fuel Assemblies 3-2 San Onofre Nuclear Generating Station - Unit 2 Cycle 3 Core Map I 3-3 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly Average Burnup and Initial Enrichment Distribution

! 3-4 CEA Bank Identification

3-5 In-Core Instrument Assemblies - Core Locations 5-1 SONGS-2 Cycle 3 POIL for Regulating Groups 5-2 SONGS-2 Cycle 3 Part length CEA Insertion Limit vs. Thermal Power 5-3 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly Relative Power Density, HFP at 800, Equilibrium Xenon, ARO 5-4 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly '

Relative Power Density, HFP at 8 GWD/T, Equilibrium Xenon, ARO 5-5 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly l Relative Power Density, HFP at EOC, Equilibrium Xenon, ARO 5-6 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly

. Relative Power Density, HFP at BOC, Equilibrium Xenon, With PLCEAs 5-7 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly Relative Power Density, HFP at BOC, Equilibrium Xenon, With Bank 6 i

i 5-R San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly Relative Power Density, HFP at ROC, Equilibrium Xenon, With Bank '

6 and PLCEAs

) 5-9 San Onofre Nuclear Generating Station Unit 2 Cycle 3 Assembly

Relative Power Density, HFP at EOC, Equilibrium Xenon, With PLCEAs 5-10 San Onofre Nuclear Generating Station . Unit 2 Cycle 3 Assembly Relative Power Density, HFP at EOC, Equilibrium Xenon, With Bank 6

, 5-11 San Onofre Nuclear Generating Station Unit 2 Cycle 3 ' Assembly Relative Power Density, HFP at EOC, Equilibrium Xenon, With Bank 6 and PLCEAs

. 7.1.3-1 Increased Main Steam Flow Plus Single Failure Core Power vs. Time i 7.1.3-2 Increased Main Steam Flow Plus Single Failure Core Heat Flux vs.

Time 4

7.1.3-3 Increased Main Steam - Flow Plus Single Failure RCS Pressure vs.

Time 7.1.3 4 Increased Main Steam Flow Plus Single Failure RCS Temperatures vs. Time-IV

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List of Figures (continued)

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V 7.1.3-5 Increased Main Steam Flow Plus Single Failure Steam Generator Pressure vs. Time 7.1.3-6 Increased Main Steam Flow Plus Single Failure DNBR vs. Time 7.1.5b-1 Steam Line Break Comparative Moderator Reactivities .

7.1.5b-2 Full Power Steam Line Break with loss of AC Power Core Power vs.

Time 7.1.5b-3 Full Power Steam Line Break with Loss of AC Power Core Heat Flux vs. Time 7.1.5b 4 Full Power Steam Line Break with Loss of AC Power RCS Pressure vs. Time 7.1.5b-5 Full Power Steam Line Rreak with Loss of AC Power RCS Temperatures vs Time 7.1.5b-6 Full Power Steam Line Break with Loss of AC Power Steam Generator Pressure vs. Time 7.1.5b-7 Full Power Steam Line Break with Loss of AC Power Reactivities s

vs. Time ,

7.2.6-1 Feedwater System Pipe Rreak Core Power vs. Time 7.2.6-2 Feedwater System Pipe Break Core Heat Flux vs. Time 7.2.6-3 Feedwater System Pipe Break RCS Pressure vs. Time 7.2.6-4 Feedwater System Pipe Break RCS Temperature vs. Time 7.2.6-5 Feedwater System Pipe Break Steam Generator Pressure vs. Time 7.3.2-1 Total Loss of Forced Reactor Coolant Flow Core Flow vs. Time 7.3.2-2 Total Loss of Forced Reactor Coolant Flow Core Power vs. Time 7.3.2-3 Total loss of Forced Reactor Coolant Flow Core Heat Flux vs. Time 7.3.2-4 Total Loss of Forced Reactor Coolant Flow RCS Pressure vs. Time 7.3.2-5 . Total Loss of Forced Reactor Coolant Flow RCS Temperatures vs.

Time 7.4.1-1 CEA Withdrawal from Subcritical Core Power vs. Time

(~'T 7.4.1-2 CEA Withdrawal from Subcritical Core Heat Flux vs. Time V

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t, List of Figures

('5 (continued) 7.4.1-3 CEA Withdrawal from Subcritical RCS Pressure vs. Time 7.4.1 4 CEA Withdrawal from Subcritical RCS Temperatures vs. Time 1

7.4.1-5 CEA Withdrawal from Subcritical Steam Generator Pressure vs. Time 7.4.1-6 CEA Withdrawal at Low Power Core Oower vs. Time 7.4.1-7 CEA Withdrawal at low Power Core Heat Flux vs. Time 7.4.1-8 CEA Withdrawal at Low Power RCS Pressure vs. Time 7.4.1-9 CEA Withdrawal at low Power RCS Temperatures vs. Time i

I 7.4.1-10 CEA Withdrawal at low Power Steam Generation Pressure vs. Time 7.7.1-1 Asymmetric Steam Generator Event Core Power vs. Time l

7.7.1-2 Asymmetric Steam Generator Event Core Heat Flux vs. Time 7.7.1-3 Asymmetric Steam Generator Event RCS Pressure vs. Time g-'s 7.7.1 4 Asymmetric Steam Generator Event RCS Temperatures vs. Time 4

\- 7.7'1-5 Asymmetric Steam Generator Event Steam Fenerator Pressure vs. Time i

R-1 Core Power 1.0 X Double Ended Guillotine Break in Pump Discharge Leg 8-2 Pressure in Center Hot Assembly Node 1.0 X Double Ended Guillotine Break in Pump Discharge leg a  ? Leak Flow 1.0 x Double Ended Guillotine Break in Pump Discharge leg 84 Flow in Hot Assembly-Path 16,. Below Hot Spot 1.0 x Double Ended -

Guillotine Break in Pump Discharge Leg 8-5 Flow in Hot Assembly-Path 17, Above Hot Spot 1.0 x Double Ended Guillotine Break in Pump Discharge Leg 8-6 Hot Assembly Ouality 1.0 x Double Ended Guillotine Break in Pump Discharge leg

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I 8-7 Containment Pressure 1.0 X Doubled Ended Guillotine Break in Pump l

Discharge Leg l

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List of Figures (continued) 8-8 Mass Added to Core During Reflood 1.0 X Double Ended Guillotine

' Break in Pump Discharge leg l- 8-9 Peak Clad Temperature 1.0 X Oouble Ended Guillotine Rreak in Pump Discharge Leg l 8-10 Hot Spot Gap Conductance 1.0 X Double Ended Guillotine Break in Pump Discharge Leg 8-11 Local Clad Oxidation 1.0 X Double Ended Guillotine Break in Pump Disepargeleg i

8-12 Clad Centerline Average Fuel and Coolant Temperature for Hottest Node 1.0 X Double Ended Guillotine Break in Pump Discharge leg 8-13 Hot Spot Heat Transfer Coefficient 1.0 X Double Ended Guillotine l

Break in Pump Discharge leg 8-14 Hot Rod Internal Gas Pressure 1.0 X Double Ended Guillotine Break in Pump Discharge leg O

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V) 1.0 Introduction and Sununary This report provides an evaluation of the design and performance of San Onofre Nuclear Generating Station Unit 2 during its third cycle of operation at 100% rated core power of 3390 MWt and NSSS power of 3410 MWt.

Operating conditions for Cycle 3 have been assumed to be consistent with those of the previous cycle and are sumarized as full power operation under base load conditions. The core will consist of irradiated Batch A, C, and 0 assemblies, along with fresh Batch E assemblies. The Cycle 2 termination burnup has been assumed to be between 9,800 and 10,200 MWO/T.

The second cycle of SONGS-2 will hereafter be referred to in this report as the " Reference Cycle."

The safety criteria (trip setpoints, margins of safety, dose limits, etc.1 apolicable for SONGS-2 were established in the Cycle 1 FSAR (Reference 1-1) and the Reference Cycle (Reference 1-2). A review of all postulated accidents and anticipated operational' occurrences has shown that the Cycle 3 core design meets these safety criteria.

The evaluations of the Cycle 3 reload core characteristics have been examined with respect to the Reference Cycle. Specific differences in core fuel loadings have been accounted for in the present analysis. The status of the postulated accidents and anticipated operational occurrences for Cycle 3 can be summarized as follows:

1. transient data are less severe than those of the Reference Cycle analysis, therefore, no reanalysis is necessary, and

2. transient data are not bounded by those of the Reference Cycle analysis, therefore, reanalysis is required, n

For those transients requiring reanalysis (Type ?l, analyses are presented in Sections 7 and 8 showing results that meet the established safety criteria, i

The Technical Specification changes needed. for Cycle 3 are described both in Section 10 and in separate license amendment applications.

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l I. i Modifications to the Core Protection Calculator (CPC) System and to the i

Core Operating Limit Supervisory System (COLSS) are being made to improve 1

performance and reflect the Cycle 3 core configuration. The data base i

changes are a result of the Extenced Cycles Program (ECP), are applicable I to Cycle 3 and should be applicabla to future cycles of SONGS-2. Algorithm changes are a result of the CPC Improvement Program (CIP) and are

! summarized in Section 9. A description of the ECP and CIP and their relationship to Cycle 3 are discussed in Reference 1-3.

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2.0 Ooerating History of the Reference Cycle SONGS-2 Unit 2 is currently in its .second fuel cycle which began with initial criticality on April 12, 1985. Low Power physics Testing was satisfactorily completed on April 10, 1985, and on May 2,1985 the unit I reached full power.

It is presently estimated that Cycle 2 will terminate on or about January 15, 1986. The Cycle 2 termination point can vary between 9800 P610/T and 10,200 PaWD/T to accornodate the plant schedule and still be

1. within the assumptions of the Cycle 3 analyses.

As of June 2a,1985 the unit has had no major o'utages. The Cycle 2 average burnup achieved to this date is 2260 MWD /T.

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f'N 3.0 General Description The Cycle 3 core will ;onsist of those assembly types and numbers listed in Table 3-1. Eighty Batch B assemblies and eight Batch C will be removed from the Cycle 2 core to make way for 88 fresh, Batch E assemblies. Fifty-six Batch C and all Batch 0 assemblies now in the core will be retained.

One Batch A assembly now in the core will be replaced with one Batch A assembly discharged after Cycle 1.

The reload batch will consist of 40 type E0 assemblies, 8 type El assemblies with a burnable poison shims per assembly, 28 type E2 assemblies with 8 burnable poison shims per assembly and 12 type E3 assemblies with 16 burnable poison shims per assembly. These sub-batch types are zone-enriched and their configurations are shown in Figure 3-1.

The loading pattern for Cycle 3, showing fuel type and location, is

- displayed in Figure 3-2.

4 O 4 Figure 3-3 displays the beginning of Cycle 3 assembly average burnup

! distribution along with the initial assembly average fuel enrichment. The burnup distribution is based on a Cycle 2 length of 10,000 MWD /T.

Control element assembly patterns and in-core inst rument locations will remain unchanged f rom Cycle 2 and are shown in Figure 3 4 ano Figure 3-5.

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3-1

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< , 1 San Onofre Nuclear Generating Station Unit 2

.t , Cycle 3 Core Loading 1

t initial Total Number Assemoly . Fuel Rods Initial Number Shim of Desig- Number of .per Enrichment Shims / Loading Fuel Shim nation Assemblies Assembly (w/o U-235) Assembly (gm B10 /I"I R0d5 R0d5 1

1.87 0 0 236 0

,I - A 1 236 1

2.91 0 0 8960 0 C 40 224 4A0 t 12 2.31 1

i 2.91 12 .01034 1696 c6

<C. A 212 '

12 2.31 06 I ,

! C+ A 20A 2.91 16 .01034 1664 12A 12 2.31 06

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! O SF 194 3.65 0 0 10304 52 2.7A 2912 4

i 0 0 3594 0 0* 16 224 2.7A 12 1.92 142 i

40 1AA 4.05 0 0 7360 0

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-5? 3.40 2080 a

-l (,'; ' : , e a .0192 laan 32 El 8 190 4.05 52 '3.40 416 E2 2A 216 3.40 -8 0242 6048 22a I

12 2.78 336 i

{ 2nA 3.40 16 0102 2a06 102

. E3 17 12 2.7A laa Total 217 50540 672 3-2 l

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E0 E0 E0 E0 E0 E0 El D D* D El EO E0 D C E2 C D C. D E0 E0 D C.

E2 D* E2 C E3 D E2 E0 E0 E2 D E3 C E2 D* C+ D D E0 E0 D D C+ D* E2 C E3 C C D D C D* E3 C. E0 E0 C. E3 D* C D D C+ E2 D C+ D E2 C D El El D C E2 D D EO E0 D D C C C D D C E2 C D C E2 C D E0 O E0 E0 D* E2 D* E3 C E2 C A C E C U D* E2 D*

E0 D D C C C D D C E2 C D D C E2 C E0 E2 D C+ D E2 C D El El D C E2 D C+ D D C D D C D* E3 C. E0 E0 C. E3 D* C D E3 C E2 D* C+ D D E0 E0 D D C+ D* E2 C E2 D* E2 C .E3 D E2 EO E0 E2 D E3 C C. D C E2 C D C. D EO E0 D

! E0 E0 El D D* D El E0 E0  ;

1 E0 E0 E0 EO l SA;i O!;0FRE ? SAN ONOFRE NUCLEAR GENERATING STATION - UNIT 2 UCLEAR l CYCLE 3 CORE M.AP 3-2 GENER.2T!NG 57;;;;NI .

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XXX INITIAL ENRIOtENT W/0 U-3 YYYY BOC ASSEGLY AVERAGE BURNUP b D/T)

EoC2=10000 Mwn/T

( l 3.91 3.91 l 0 0 l

3.91 3.91 3.90 3.46 2.74 0 0 0 11504 12368 3.91 3.46 2.88 3.46 2.88 3.37 0 6753 24081 9641 19876 0 3.91 3.37 3.46 3.37 2.88 3.37 2.74 0 0 7357 0 20355 0 12714 2.83 2.74 3.37 2.88 3.37 0 .

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.46 6753 3.46 7357 22233 12526 0 22188 0 3.91 2.88 3.37 2.74 2.88 3.46 3.46 2.88 0 24081 0 12526 18857 8461 10967 20411 3.90 3.46 2.88 3.37 3.46 2.88 3.46 3.37 i 3.91 0 9641 20355 0 8461 24697 7559 0 0 3.46 2.88 3.37 2.83 3.46 3.46 2.88 2.88 1 3.91 11504 19876 0 22188 10967 7559 20305 21247 0 2.74 3.37 2.74 3.37 2.88 3.37 2.88 1.87 _

G 4 12368 0 12714 0 20411 0 21247 11407 l

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V SN4 OJCFRE NUCLEAR GE4ERATING STATICtd UNIT 2 CYCLE 3 ASSESY l

AVER /4E SUR*1UP AND INITIAL ENRICWENT DISTRIBUTION 3_3

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6 - LEAD REGULATING B ANK 5 - SECOND REGULATING BANK 4 -THIRD REGULATING BANK 3 - FOURTH REGULATING BANK 1 2 I 2 - FIFTH REGULATING BANK SA 1 - LAST REGULATING B ANK S S 6 7 4 5

, B h 0WN B 3

2 SA- SHUTDOWN BANK A 8 9 10 11 12 13 3 4 SB 15 16 17 18 '19 20 1 S 3 A 3 24 25 26 27 28 J 5 PLR 6 33 34 35 36 S 3 A 3 42 43 44 4 1 52 53

• SHUTDOWN ROD IN POSITION S[

. 52 IS AVAILABLE FOR ONLY TWO DIAGONALLY OPPOSITE CORE 62 QUADRANTS. 2 SAN ONOFRE NUCLEAR GENERATING STATION O Unit 2 CEA 3A:iK IDE::TIFICATIO :

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IN-CO R E INSTRUM ENT O ASSEMBLY LOCATION IN-C0 R E INSTRUMEN T

@ ASSEMBLY WITH BACKGROUNO SAN ONOFRE CETECTOR LOCATION NUCLEAR GENERATING STATION f1 Unit 2 L) IN-CORE INSTRCIENT ASSE!BL'ES -

} CORE LOCATIONS

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~ System Design 4.1 Mechanical Design The mechanical design for the standard Batch E reload fuel is essentially identical to that of Batch 0 fuel used in SONGS-2 Cycle 2 and described in the reload analysis report for the Reference Cycle (Reference 4-1), with the following exception:

The designs of the CEA guide tubes and wear sleeves have been modified to permit installation of. the wear sleeves completely within the guide tubes. This permits a design in which the sleeve is expanded along its entire length, thereby eliminating the need for vent holes in the

' sleeve and facilitates, when necessary, fuel bundle reconstitution.

Reference 4-2 is C-E's submittal discussing the CEA guide tube wear sleeve modification and Reference 4-3 is the NRC's acceptance of the design change.

w C-E has performed analytical predictions of cladding creep-collapse time for all SONGS-2 fuel batches that will be irradiated in Cycle 3 and has concluded that the collapse resistance of all fuel pins is sufficient to preclude collapse during Cycle 3. These analyses utilized the CED AN computer code (Reference 4 4) and the procedures described in Reference 4 7 and included as input conservative values of internal pressure, cladding dimensions, cladding temperatures and neutron fluence.

4.2 Mitigation of Guide Tube Near 4

All fuel assemblies which will be placed in CEA locations in Cycle 3 will have stainless steel sleeves installed in the guide- tubes- to prevent guide tube wear. The design of the sleeves for the Batch E fuel is discussed in Section 4.1 above. For all other batches of fuel a detailed discussion of the design of the sleeves and their effect on reactor operation is contained in Reference 4-12.

4.3 Thermal 1esign The thermal performance of composite. fuel pins that envelope the various pins of the various fuel batches present in Cycle 3 (fuel batches A, C, 9 4-1

and El have been evaluated using the FATES 3A version of the fuel evaluation model (References 4-5 and 4-8) as approved by the NRC (Reference 4-9). The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of Cycle 3.

Results of these burnup dependent fuel performance calculations were used in the Transient Analysis presented in Section 7 and in the ECCS Analysis presented in Section 8.

4.4 Chemical Design 4 The metallurgical design specifications of the fuel cladding and the fuel

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assembly structural members for the Batch E fuel are identical to those of the Batches A, B and C fuel as described in Reference 4-6 and the Batch D fuel as described in Reference 4-1.

4.5 Shoulder Gap Adecuacy Calculations using the methods described in Reference 4-10 indicate that adequate shoulder gap can be provided for all fuel assemblies that will ~ be irradiated in Cycle 3. The NRC review conducted on these methods (Reference 4-11) concluded that additional data were necessary before the methods were useable on fuel accumulating fluences exceeding 6.5x10 21 nyt. Therefore, an inspection program and an evaluation will be performed to ensure that adequate shoulder gap remains on fuel scheduled for its third cycle of service in Cycle 3 I

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5.1 Physics Characteristics 5.1.1 Fuel Management The Cycle 3 loading pattern is characterized by loading approximately half of the fresh fuel on the core periphery and shuffling to the interior the fuel assemblies previously located on the periphery in Cycle 2. Forty fresh fuel assemblies have a lower assembly average enrichment than those on the periphery and are mixed with the previously burned fuel in the central region of the core in a pattern which minimizes power peaking.

With this loading and a Cycle 2 endpoint at 10,000 WD/T, the Cycle 3 reactivity lifetime for full power operation is expected to be 14,500 MWO/T. Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 2 termination burnup of between 4,800 and 10,200 MWD /T and for a Cycle 3 length up to 16,000 MWO/T.

b U Characteristic physics parameters for Cycle 3 are compared to those of the Reference Cycle in Table 5-1. The values in this table are intended to represent nominal core parameters. Those values used in the safety analysis (see Sections 7 and 8) contain appropriate uncertainties, or incorporate values from the Extended Cycles Program (Reference 5-1) to bound future operating cycles, and in all cases are conservative with respect to the values reported in Table 5-1.

Table 5-2 presents a summary of CEA _ reactivity worths and allowances for the end of Cycle 3 full power steam line break transient with a comparison to the Reference Cycle data. The full power steam line break was chosen to illustrate differences in CEA reactivity worths for the two cycles.

1 The CEA core locations and group identifications remain the sane as in t9 Reference Cycle. The power dependent insertion limit (P0ll) for regulating groups and part length CEA groups remains the same as in the Reference

- Cycle and is shown in Figures 5-1 and 5-2 respectively. Table 5-3 shows the reactivity worths of various CEA groups calculated at full power v conditions for Cycle 1 and the Reference Cycle.

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l O 5.1.2 Power Distribution Figures 5-3 through 5-5 illustrate the calculated All Rods Out ( ARO) planar radial power distributions during Cycle 3. The one-pin planar radial power peaks presented in these figures represent the maximum that could be expected between about 20 and 80 percent of core height. Power peaks outside this axial region were examined and found not to be limiting at any time during the cycle. Time points at the beginning, middle, and end of cycle were chosen to display the variation in maximum planar radial peak as a function of burnup.

Radial power distributions for rodded configurations are given for ROC and EOC in Figures 5-6 through 5-11. The rodded configurations shown are those allowed by the POIL at full power: part length CEAs (PLCEAs), Bank 6, and Rank 6 plus the PLCEAs. As is the case for unrodded configurations, the largest planar radial peak for each of these rodded configurations occurs at beginning of Cycle 3.

O The radial power distributions described in this section are calculated data which do not include any uncertainties or allowances. The calculations performed to determine these radial power peaks explicitly account for augmented power peaking which is characteristic of fuel rods adjacent to the water holes.

Nominal axial peaking factors are expected to range from 1.24 at 30C3 to 1.09 at EOC3.

5.2 Safety Related Data 5.2.1 Augmentation Factors A recently completed analysis performed by C-E for EPRI, Referanca 5-2, demonstrated that the increased power peaking associated with the small interpellet gaos found in C-E's modern fuel rods (non-densifying fuel in pre-pressurized tubes) is insignificant compared to the uncertainties in d the safety analyses. The report concluded that augmentation factors can be

! eliminated from the ' reload analyses of any reactor loaded exclusively with this type of fuel. This discussion of the elimination of the augmentation 5-2

-- - . - . ~ . - _ .- . .-

I factors was used by BGAE in Reference 5-3 and accepted by the NRC in Reference 5 4 Since the manufacturing process of C-E's modern fuel is the same for both BG4E and SCE, and the fuel differs only in dimensions, it is C-E's conclusion that the peaking factor penalty due to fuel densification is insignificant compared to the uncertainties incorporated into COLSS and CPC and thus the augmentation factors have been eliminated for Cycle 3.

3 5.3 Physics Analysis Methods 5.3.1 Analytical Input to In-Core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in accordance with Reference 5-

5. As in the Reference Cycle, ROCS-0IT with the MC module will be used.

ROCS-DIT and the MC module have been aDDroved for this application in Reference 5-6.

5.3.2 Uncertainties in Measured Power Distributions

The planar radial power distribution measurement uncertainty of 5.3%, based on Reference 5-5, will be applied to the Cycle 3 C 1.SS and CPC on-line calculations which use planar radial power peaks. The axial and three 2 dimensional power distribution neasurement uncertainties are determined in conjunction with _other monitoring and protection system measurement ,

uncertainties, as was done for Cycle 2.

5.3.3 Nuclear Design Methodology l

As 'in the Reference Cycle, the Cycle 3 nuclear design was performed with two and three dimensional core models using the ROCS computer code and t

employing OIT calculated cross : sections. The ROCS-DIT and the MC modula

was described in Reference 5-6.

i O

l 5-3

TARLE 5-1 SONGS-2 CYCLE 3 l NOMINAL PHYSICS CHARACTERISTICS Reference

! Dissolved Roron Units Cycle Cycle 3 Dissolved Boron Concentration for-Criticality, CEAs Withdrawn, Hot Full Power PPM 945 1186 Equilibrium Xenon, ROC i

Baron Worth

,i Hot Full Power, BOC PPM /% ao 9A 114 t

! Hot Full Power, EOC PPM /* ao 84 44 Moderator Temoerature Coef ficients Hot Full Power, Equilibrium Xenon Beginning of Cycle 10*#ac/0F 0.4 -0.2 4 End of Cycle 10"#ac/0F -2.4 -2.6 Goooler Coefficient i

Hot Zero Power, ROC 10-53 ,f og _g,73 ,7,4g Hot Full Power, ROC 10-6ao/oF -1.25 -1.21 Hot Full Dower, EOC 10-53 ,foF -1.1A -1.41 Total Delayed Neutron Fraction,.3ef f l ROC ------- 0.0n66 0.0066 j EOC ------- 0.0046 0.0051 i j I NeutronGenerationTime,l' SOC 10~6 sec 23.7 22.3 s

EOC 10 6 sec 34.1 27.1 5-4

i h

TABLE 5-2 SONGS-2 CYCLE 3 LIMITING VALUES OF .

I REACTIVITY WORTHS AND ALLOWANCES FOR HOT l FULL POWER STEAM LINE BREAK, % END-0F-CYCLE (EOC)

J

' Reference

}

Cycle Cycle 3 1

-10.A -11.A 1.. Worth of all CEAs Inserted

- +1.9

?. Stuck CEA Allowance +2.A5 i-i

• 1 ,

i j 3. Worth of all CEAs Less Highest i Worth CEA Stuck Out -7.95 4.5

a. Full Power Dependent Insertion j

i Limit CEA Bite +0.2 +0.2

}

S. Calculated Scram Worth 7.75 -9.3 l

i J

6. Pnysics Uncertainty + 65

. +.an ,

7 Other Allowances (worth losses l

due to voiding and moderator temperature axial redistribution) +0.2 +0.2  ;

1 l 8. Net Available Scran Worth -6.9 -R.3 '

1 i

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TABLE S-3.

SONGS-2 CYCLE 3

j. REACTIVITY WORTH OF CEA REGllLATING GROUPS

.AT HOT FULL POWER, ?ac

' Beginning of Cycle End of Cycle i

i Regulating Reference Reference j Cycle 3 CEAs Cycle Cycle 3 Cycle l

1 Group 6 0.4 0.3 04 0*b j

i ~

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d t

ASSEMBLY RELATIVE POWER DENSITY O.73 1.00 X.

0.69 0.94 1.16 1.12. 1.03 0.83 1.05 0.82 1.10 0.94 1.24 0.83 1.16 1.17 1.09 0.87 1.18 0.99 0.69 1.05 1.17 0.87 0.93 1.20 0.87 1.09 b) 0.94 0.82 1.09 0.93 0.90 1.19 1.14 0.90 1.16 1.10 0.87 1.20 1.19 0.90 1.16 1.18 0.73 .

1.12 0.94 1.17 0.37 1.13 1.15 0.78 0.70 0.99 E 1.02 1.24 0.99 1.08 0.89 1.17 0.69 0.55 -

k X = LCCATIOfJ OF MAXIMUM 1 PIN PEAK = 1.ll8 I

I

~ Figure SCt;nEM CALIFoMIA SA'1 CfCFRE t1UCLEAR GENE?ATI'lG STATICri Ut41T 2 CYCLE 3 ASSEGIY ED! son Co. REMTIVE PCWER CE!4SITY, HFP AT BOC, ECUILIBR!t.fi XE?Crl, 5-3 soscs-2 MO

' 5-9

l l

E ASSEMBLY RELATIVE POWER DENSITY

' O.71 0.88

'I 0.67 0.89 1.07 1.02 0.95 0.81 1.01 0.84 1.08 0.93 1.24 0.81 1.14 1.16 1.22 0.93 1.25 1.05 X

C 1.25 U 0.67 1.01 1.15 0.91 0.98 1.26 0.94 0.89 0.84 1.22 0.98 0.93 1.16 1.13 0.94 1.07 1.08 0.93 1.26 1.16 0.90 1.13 1.20 0.71 1.02 0.93 1.25 0.94 1.13 1.12 0.81 0.76 0.88 E 0.95 1.24 1.05 1.25 0.93 1.19 0.75 0.64 -E I

X = LCCATION OF MAXIMUM 1. PIN PEAK 1.42 G

O II

SOUTHERN CALIFCRNIA SM CtCFRE NUCLEAR GE4ERATING STATION tritT 2 CYCLE 3 ASSSSQ' EDISCN CO. RELATIVE POWER DENSITY, HFP AT 8 GWD/T, ECUILIBRILN XEtCN, 3-TNGs-2 m 5-10

0  :

t ASSEMBLY RELATIVE POWER DENSITY t

0.70 0.86 0.69 0.89 1.06 0.99 0.93 0.81 1.00 0.87 1.06 0.94 1.24 0.81 1.15 1.14 1.23 0.96 1.27 1.06 X

I I 0.69 1.00 1.14 0.92 0.99 1.26 0.96 1.31 0.89 0.87 1.28 0.99 0.91 1.10 1.08 0.94 1.06 1.06 0.96 1.26 1.10 0.88 1.00 1.19 0.70 0.99 0.94 1.27 0.96 1.08 1.08 0.83 0.79 O.36 E 0.93 1.25 1.06 1.31 0.94 1.19 0.79 0.70 -!

I

! X = LCCATION OF MAXIMUM 1 Pl!J PEAK = 1.li6 1

I J

v ~" -

SOUTHERN CALIFCRNIA SAN GCFFE NUCLEAR GENERATING STATICN UNIT 2 CYCLE 3 ASSEMBLY RELATIVE POWER CENSITY, HFP AT EOC, ECUILIBR!Lti XE?CN, 5-3 EDISCN CO.

SONGS-2 mo 3-11

O LOCATION OF PLCEAS N\ ASSEMBLY RELATIVE POWER DENSITY O.79 1.01 X

0.f0 0.95 1.17 1.14 1.05 0.84 1.06 0.83 1.11 0.95 1.26 0.84 1.16 1.18 1.09 0.36 1.17 1.00 N

g 0.70 1.06 1.18 0.89 0.91 1

!Le.0 0.86 1.09 0.94 0.83 1.09 0.91 0.89 1.17 1.14 0.91 1.17 1.11 0.86

\

. 09 1.16 0.92 1.13 1.20 N\\'3N 0.79 1.14 0.95 1.17 0.35 1.13 1.17 0.80 0.73 1.01

( 1.04 1.25 0.99 1.09 0.90 1.19 0.72 0.58 -(

I X = LCCATION OF MAXIMUM 1. PIN PEAK - 1.51 Ol Figwre SN1 CtCFRE NUCLEAR GENERATING STATICN t.ft!T 2 CYCLE 3 ASSEMBLY SCUTHE?ff CALIFCANIA '

REl.ATIVE POWER CENSITY, HFP AT BOC, ECUILIBRIUM XE?Oi, WITH 3-4 CISCN Co.

SONGS-2 ptCEAs 5-12

LOCATION OF BM4K 6 ASSEMBLY RELATIVE POWER DENSITY 1

0.81 1.03 l 9

0.74 0.99 1.21 1.16 1.06 0.89 1.13 0.87 1.14 0.94 1.23 X

0.89 1.24 1.26 1.15 0.88 1.10 0.87 0 0.74 1.13 1.25 0.92 0.97 1.18 0.76 0.99 0.37 1.15 0.97 0.92 1.17 1.05 0.77 1.20 1.14 0.88 1.18 1.17 0.90 1.12 1.12 0.81 1.16 0.94 1.10 0.76 1.05 1.12 0.77 0.69 E 1.06 1.22 0.86 0.77 1.12 0.69 0.56 -(

I X = LCCATION OF MAXIMUM 1 PIN PEAK 1.54

( )

I Fi g .r e SOU HERN CALIFORNIA Ski CtCFRE NUCLEAR GE4ERATING STATICN UNIT 2 CYC E 3 ASSEM 5-7 E o co. RE!.ATIVE POWER C04 SIT (, HFP AT BOC, ECUILIBRith XENCN, WITH BANK 6 5-13

l l

l N' t LOCATION CF PLCEAS ASSEMBLY RELATIVE POWER DENSITY 1 0.84 1.08 LOCATIONOFBM4K6 )(,

O 76 1.02 1.24 1.20 1.10 0.92 1.15 0.89 1.16 0.96 1.25 0.92 1.27 1.27 1.15 0.86 1.09 0.87 0 0.76 1.15 1.27 0.92 0.94 0.73 1.02 0.89 1.14 0.93 0.88 1.12 1.02 0.76 1.24 1.16 0,86 1.12 0.87 1.10 1.11 0.34 1.20 0.96 1.09 0.73 1.02 1.10 0.76 0.69

( 1.10 1.25 0.87 0.75 1.10 0.68 0.56 -(

I X = LCCATION OF MAXlMUM 1. PIN PEAK 1.59 q,

(_/

SN4 CtCFRE NUCLEAR GEhERATING STATICil UNIT 2 CvCLE 3 Asseet3 "i"

SCUTHERN C & !FORNIA RELATIVE PCWER CD4SITY, HFP AT BOC, ECUIL!BRILfi XENCN, WITH 5-3 ED!SCt1 CO.

SONGS-2 BANK 6 At0 PLCEAS 5-14

l

\ LOCATION OF PLCEAS

\\ - -

ASSEMBLY RELATIVE POWER DENSITY 1 0.72 0.88 0.$ 0.91 1.08 1.02 0.95 X

0.83 1.02 0.88 1.08 0.95 1.27 i

0.83 1.17 1.15 1.28 0.94 1.27 1.07 W

\

1.30 C' 0.70 1.02 1.15 0.92 0.96 x1.15NN\\\g 0.93 0.91 0.88 1.28 0.96 0.88 1.07 1.07 0.94 1.08 1.08 0.94 1.07 0.87 1.08 1.20 LNNt:

0.72 1.02 0.96 1.27 0.93 1.07 1.08 0.83 0.80

,0.88

( 0.95 1.27 1.07 1.30 0.94 1.20 0.79 0.71

-(

X = LCCATICN CF MAXIMUM 1. PIN PEAK = 1.l}6 D'

(G *N i "

SCUTHERN CALIFCRNIA SAN CfJCFRE Ntr' EAR GENERATING STATICN UNIT 2 CYCLE 3 ASSEGLY EDISCN CO. E ATIVE PCWER CENSITY, HFP AT ECC, ECU! LIER!t.f4 XEC4, WITH 3-9 Scries-2 P CEAs 5-15

l 9

t.DCATIONOFB#4K6 ASSEMBLY RELATIVE POWER DENSITY t 0.76 0.91 0.f6 0.97 1.13 1.05 0.97 0.90 1.10 0.94 1.11 0.95 1.23 X

0.90 1.27 1.24 1.36 0.97 1.17 0.90 0.80 0 0.76 1.10 1.24 0.99 1.03 1.24 0.97 0.94 1.36 1.03 0.93 1.07 0.97 0.76 1.11 0.97 1.24 1.08 0.85 1.03 1.12 1.13 0.76 0.95 1.17 0.80 0.97 1.02 0.80 0.76 1.05

( 0.97 1.23 0.90 0.76 1.12 0.76 0.68 -(

X = LCCATIO.Y OF MAXIMUM 1 PIN PEAK - 1,54 O'

\v N"i sotmizu cALIFoutA 344 CfiCFRE fiUCLEAR GC4 ERAT!?tG STAT!Cri UNIT 2 CYCLE 3 ASSE.WLY V.L\TIVE PCWER CelsITY, HFP AT EOC, ECUIL!ERlt.N XBiCti, WITH 5-10 gkstjo,

.4MIK6 5-16

t.CCATION CF PLCEAS N\ ASSEMBLY RELATIVE POWER DENSITY 1 0.78 0.95 toCATtce4 cF M4K 6 0.7h 0.99 1.16 1.08 1.00 0.92 1.13 0.95 1.13 0.96 1.25 X

0.92 1.29 1.26 1.36 0.95 1.16 0.90 0.78 1.13 1.26 0.99 1.00 0.77 C) 0.99 0.93 1.36 1.00 0.89 1.03 0.95 0.76 N

1.16 1.13 0.95 ' .12 1.03 0.83 1.02 1.11

\\\\

0.78 1.08 0.96 1.16 0.77 0.95 1.02 0.80 0.77 E 1.00 1.25 0.90 0.75 1.11 0.76 0.69 -(

I X = LCCATION OF MAXIMUM 1. PIN PEAK 1.57 k

i O' SN4 CtCFRE NUCLEAR GE!4E?ATING STAT!Off (f4tT 2 CYCLE 3 ASseety Fis r. ,

SCUTHERti CALIFCRNIA RELATIVE OER OCl! STY, HFP AT EOC, ECU! LIER!lti XBO4, WITH !41 EDISO4 CO.

Saecs-2 tenK 6 u40 ptCEAS 5-17 .

6.0 Thermal-Hydraulic Design 6.1 DNRR Analysis Steady state ONRR analyses of Cycle 3 at the rated power level of 3300 WT have been performed using the TORC computer code described in Reference 6-1, the CE-1 critical heat flux correlation described in Reference 6-2, the simplified TORC modeling methods described in Reference 6-3, and the CETOP code described in Reference 6 4 of pertinent thermal-hydraulic design Table 6-1 contains a list parameters. The Statistical Combination of Uncertainties (SCU) methodology presented in Reference 6-5 was applied with the calculational factors in Table 6-1 and other uncertainty factors at the 95/95 listed confidence / probability level to define a design limit of 1.31 on CE-1 minimum GNBR which was approved for use in the Reference Cycle. This finit has been verified for Cycle 3 Information on the HID-1 and Hin-2 grids is provided in References 6-6 and V 6-7 The use of both HID-1 and H10-2 grids has already been approved by NRC for the SONGS 2 and 3 cores (Reference 6-8). A penalty of 0.01 was inoosed by NRC on the CE-1 correlation DNBR limit for SONGS-? and 3 to address NRC concern about the effect of the HIO-1 and HID-2 spacer grids and a larger grid spacing. This penalty is included in the 1.31 DNRR limit, along with other penalties imposed by NRC in the review of previous SCO analyses (Reference 6-10).

6.2 Effects of Fuel Rod Bowing on ONRR Margin Effects of fuel rod bowing on DNB0 margin have bean incorporated in the safety and setpoint analyses in tha manner discussed in References 6 5 and 6-9. The penalty used for this analysis,1.75?, MONRR, is valid for hundle burnups up to 30,000 MWD / Mill. This penalty is included in the 1.31 ONBR limit.

O ,

6-1

i.

i i

i f

1 r 1 i i

l j

For - assemblies with burnup greater than 30 GWD/T sufficient available s margin exists to offset - red bow penalties due to the lower radial power j i

peaks in these higher burnup hatches. Hence the rod how penalty based upon Reference 6-Q for 30 GWD/T .is applicable for all assembly burnups expected .

for Cycle 3.  :

1 I

I k

i r

i e \

I i

i i

i l

I '

l  !

i i i i

i l

I I .

l l

1 6-2 l ~ ~ - ---- _ _._ _ __ _ ___, _ _._ _ e gmygp,,, ,,

TABLE 6-1 SONGS-2 Cycle 3 Thermal Hydraulic Parameters at Full Power 1

Reference General Characteristics Unit Cycle Cycle 3 i

Total Heat Output (Core only) 3390 3390

, MWg 10 Rtu/hr 11,570 11,570 Fraction of Heat Generated in -- 0.975 0.975 Fuel Rod Primary System Pressure psia 2250 2250 Nominal s 0 553.0 553.0 Inlet Temperature (Nominal) F Total Reactor Coolant Flow gpg 396,000 396,000 (Minimum Steady State). 10 lb/hr 148.0 148.0 1 Coolant Flow Through Core (Minimum) 106 lb/hr 143.9 143.6*

, Hydraulic Diameter (Nominal Channel) ft 0.039 0.039 6

Average Mass Velocity 10 lb/hr-ft 2.63 2.63 Pressure Orop Across Core (Minimum ' psi 20.0 19.9 steady state flow irreversible P over entire fuel assembly) -

Total Dressure Orop Across Vessel psi 43.6 43.6 ( ,

(Based on nominal dimensions and '

minimum steady state flow) ,

Core Average Heat Flux (Accounts RTU/hr-ft 2 1R2,400*** 178,900* -

for fraction of heat generated in fuel rod and axial densifica- ,

tion factor) 2

Total Heat Transfer Area (Ace'ounts ft 6?,000*** 63,000*

for axial densification factor)

Film Coefficient at Average BTU /hr-ft 20F 6200 6200 Conditions l

i Average Film Temperature Difference OF 29.4 28.8 Average Linear Heat Rate of Unden- kw/ft 5.34*** 5.23*

-sified Fuel Rod (Accounts for fraction of heat generated in t I <

fuel rod) '<

Average Core Enthalpy Rise RTU/lb 80.4 80.6 Il' 6-3

7. .

t

V t

TABLE 6-1 (continued) t j

Reference

\ Calculational Factors Unit Cycle Cycle 3 Maximum Clad Surface Temperature OF 656.7 656.7 1

Engineering Heat Flux Factor 1.03** 1.03**

Engineering Factor on Hot Channel 1.03** 1.03**

Heat Input ,

'- Rod Pitch, Bowing and Clad Diameter 1.05** 1.05**

i Factor Fuel Densification Factor ( Axial) 1.002 1.002 3

i

)

l' NOTES:

1 l-

• Based on 672 shins.

• • These factors have been combined statistically with other uncertainty factors at 95/95 confidence / probability level to define a new design limit

, on CE-1 minimum ONBR when iterating on power as discussed in Reference 6-5.

i

• • • Based on 1632 shtms.

' Design hyoass flowrate has increased from 2.8% to 3.0% of -total reactor i coolant flew.

i 1 g i

i I

i- ,

1 1

4 e

'l

" 6-4 I

.h

'l,.-....-,...~,,..._-,-..-

i I

7.0 Non-LOCA Safety Analysis O

U 7.0.1 Introduction This section presents the results of the San Onofre Nuclear Generating Station (SONGS) Unit 2, Cycle 3 Non-LOCA safety analyses at 3410 MWt.

The Design Bases Events (DREs) considered in the safety analyses are listed in Table 7.0-1. These events are categorized into three groups: Moderate Frequency, Infrequent and Limiting Fault events. For the purpose of this report, the Moderate Frequency and Infrequent Events will be termed Anticipated Operational Occurrences. The DBEs were evaluated with respect to four criteria: Offsite Dose, Reactor Coolant System Pressure, Fuel Performance (DNRR and Centerline Melt SAFDLs) and Loss of Shutdown Margin. Tables 7.0-2 through 7.0-5 present the list of events analyzed for each criterion. All events were re-evaluated to assure that they meet their respective criterion with the Cycle 3 fuel design. The 08Es chosen for _ analysis for each criterion are the limiting events with respect to that criterion.

The write-ups for those events presented are broken down into a discussion of the reason (s) for the reanalysis, a discussion of the cause(s) of the event, a description of the analyses performed, the results and conclusions. In the Reference Cycle (Reference 7-1), some events previously analyzed with and without a single failure in the Cycle 1 FSAR (Reference 7-2) had been combined into the same section for presentation. .This practice is repeated for Cycle 3.

7.0.2 Methods of Analysis The analytical methodology used is consistent with t% Reference Cycle analysis methods (Reference 7-1) unless otherwise stated in the e'ent presentation.

7.0.3 Mathematical Models The mathematical models and computer codes used' in the Cycle 3 Non-LOCA safety analysis are identical to those used in the Reference Cycle (Reference 7-1).

The exceptions to this are the application of the TORC code to the sheared shaft event and the HERMITE code to the Total Loss of Forced Reactor Coolant Flow and the Asynynetric Steam Generator Transient.

Plant response for Non-LOCA Events was simulated using the CESEC III computer code (Reference 7-10).

The STRIKIN II computer code (Section 15.0.4.1.2 of Reference 7-2 and Reference 7-3) was also used in the analysis of the CEA Ejection Event.

Simulation of the fluid conditions within the hot channel of the reactor core and calculation of DNBR was performed using the CETOP-0 computer code (Section 6.1 of this report and References 7-7 and 7-8). The number of fuel pins l predicted to experience DNR was calculated by the statistical convolution j method described in Reference 7 4 The TORC computer program is used to simulate 'the fluid conditions within the reactor core and- to calculate fuel pin DNBR for the sheared shaft event. The TORC program is described in References 7-14 and 7-15.

7-1

e s Determination of DNBR for the post trip return to power portion of the steam  ;

piping failure events is based on the correlation developed by R. V. tiacbeth (Reference 7-5) with corrections developed by 1.ee (Reference 7-6) to account for non-uniform axial heat flux. This methodology is consistent with that employed in the Reference Cycle analysis.

for The HERMITE code (Reference 7-11) was used to simulate the reactor core analyses which required more spatial detail than provided by a point kinetics model.

7.0.4 Input Parameters and Analysis Assumptions 4

Table 7.0-6 sumarizes the core parameters assumed in the Cycle 3 transient l

analysis and compares them to the values used in the Reference Cycle. Specific i

initial conditions for each event are tabulated in the section of the report summarizing that event. For some of the DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculated Cycle Those-3

' values fi.e., CEA worth at trip, moderator temperature coef ficient) .

values and ranges used for the core parameters resulted from the Extended Cycles Program (ECP) (Reference 7-16) for SONGS Units 2 and 1. The data base generated for the future, extended burnup cycles yielded parameters and range that not only bound the Cycle 3 generated data, but also should be applicable to future cycles as well.

7.0.5 Conclusion For all DBEs that have results bounded by the Reference Cycle, the margin of safety has not degraded from that of the Reference Cycle. Those events whose O results were. not bounded by the Reference Cycle are presented herein. All of these events have results within NRC acceptance criteria.

I J

f v

7-2 1

4 Table 7.0-1 SONGS Unit 2. Design Basis Events Considered in tne Cycle 3 Safety Analysis l

i 7.1 Increase in Heat Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature 7.1.2 Increase in Feedwater Flow

. 7.1.3 Increased Main Steam Flow 7.1.4 Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Oump Valve 7.1.5* Steam System Piping Failures 7.2 Decrease in Heat Removal by tne Secondary System I 7.2.1 Loss of External Load l 7.2.2 Turbine Trip 7.2.3 Loss of Condenser Vacuum 7.2.4 Loss of Normal AC Power 7.2.5 Loss of Normal Feedwater 7.2.6* Feedwater System Pipe Breaks 7.3 Decrease in Reactor Coolant Flowrate l 7.3.1 Partial loss of Forced Reactor Coolant Flow

, 7.3.2 Total Loss of Forced Reactor Coolant Flow 7.3.3* Single Reactor Coolant Pump Shaft Seizure / Sheared Shaft 7.4 Reactivity and Power Distribution Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or low Power Condition 7.4.2 Urcontrolled CEA Withdrawal at Power 7.4.3 CEA Misoperation Events 7.4.4 CVCS Malfunction (Inadvertent Boron Oilution) 7.4.5 Startup of an Inactive Reactor Coolant System Pumo 7.4.6* Control Element Assembly Ejection-7.5 Increase in Reactor Coolant System Inventory i

7.5.1 CVCS Malfunction 7.5.2 Inadvertent Operation of the ECCS Ouring power Operation O .

7-3

. . J-

t Table 7.0-1 (continued) 7.6 Decrease in Reactor Coolant System Inventory 7.6.1 ' Pressurizer Pressure Decrease Events 7.6.2* Small Primary Line Break Outside Containment

'l Steam Generator Tube Rupture 7.6.3*

7.7 Miscellaneous

] 7.7.1 Asymetric Steam Generator Events

• Categorized as Limiting Fault Events 4

e i.

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l 7-4

. ~ ..

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i Table 7.0-2 DBEs Evaluated with Respect to Offsite Dose Criterion

(}

Section Event Results A) Anticipated Operational Occurrences 7.1.4 1) Inadvertent Opening of a Steam Bounded by Generator Atmospheric Dump Valve Reference Cycle or Safety Valve 7.2.4 2) Loss of Normal AC Power Bounded by Paf&,ence Cycle B) Limiting Fault Events

1) Steam System Piping Failures; Presented 7.1.5a a) Pre-Trip Power Excursions 7.1.5b b) Post Trip Analysis 7.2.6 2) Feedwater System Pipe Breaks Bounded by Reference Cycle 7.3.3 3) Single Reactor Coolant Pump Presented Shaft Seizure l) 7.6.2 4) Small Primary Line Break Outside Bounded by Reference Cycle Containment 7.6.3 5) Steam Generator Tube Rupture Bounded by i Reference Cycle i

l v

l l

7-5

Table 7.0-3 OBEs Evaluated with Respect to RCS Pressure Criterion Section Event Results A) Anticipated Operational Occurrences 4

7.2.1 1) Loss of External Load Bounded by Reference Cycle 7.2.2 2) Turbine Trip Bounded by Reference Cycle 7.2.3 1) Loss of Condenser Vacuum Bounded by Reference Cycle 7.2.4 a) Loss of Normal AC Power Bounded by Reference Cycle 7.2.5 5) Loss of Normal Feedwater Bounded by Reference Cycle 7.4.1 6) Uncontrolled CEA Withdrawal from Presented A Subcritical or low Power Condition 7.4.2 7) Uncontrolled CEA Withdrawal at Power Bounded by O Referenca Cycle Bounded by 7.5.1 R) CVCS Malfunction Reference Cycle 7 . 5 . 7. 9) Inadvertent OpeFation of Bounded by ECCS During Power Operation Reference Cycle R) Limiting Fault Events t

7.2.6 1) Feedwater System Pipe Breaks Presented i

4 l

O l

7-6 l

~ . _..,_ , , . _ ._. _ _

Table 7.0 4

(

DREs Evaluated with Respect to Fuel Performance Event Results l Section A) Anticipated Operational Occurrences Decrease in Feedwater Temperature Rounded by 7.1.1 1) Reference Cycle Rounded by 7.1.2 2) Increase in Feedwater Flow Reference Cycle Increased Main Steam Flow Presented 7.1.3 3)

Partial loss of Forced Reactor Bounded by 7.3.1 4) Reference Cycle Coolant Flow

5) Total Loss of Forced Reactor Presented

• 7.3.2 Coolant Flow 1'

Uncontrolled CEA Withdrawal from a Presented l 7.4.1 6)

Subcritical or Low Power Condition Uncontrolled CEA Withdrawal Bounded by

) O 7.4.2 7) at Power Reference Cycle Rounded by

! 7.4.3 9) CEA Misoperation Events Reference Cycle Pressurizer Pressure Decrease Rounded by 7.6.1 Q)

Reference Cycle Events l-7.7.1 10) Asymmetric Steam Generator Events Presented

• B) Limiting Fault Events

+ 11 Steam System Piping Failures; 7.1.Sa a) Pre-Trio Power Excursions Presented 7.1.5b b) Post Trio Analysis Presented 7.3.1 2) Single Reactor Coolant Pump Presented Shaft Seizure / Sheared Shaft Control Element Assembly Ejection Bounded by 7.4.6. 3)

Reference Cycle

• The results of this event remain bounded by the Reference Cycle. The event is i

f e~g presented due to a change in analytical methodology.

( j 7-7 i

- _ _ _ _ . - _ _ _ ~ . . _ . - , . ~ . . . . , , _ . -. - - . , , . , , , ,

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l t

Table 7.0-5 OBEs Evaluated with Resoect to Shutdown Marain Criterion Ever.t Results Section A) Anticipated Operational Occurrences Inadvertent Opening of a Steam Bounded by 7.1.4 1) Reference Cycle

! Generator Safety Valve or Atmospheric Dump Valve CVCS Malfunction (Inadvertent Presented 7.4.4 2)

Boron Oilution)

B) Limiting Fault Events Steam System Piping Failure, Presented 7.1.5b 1)

Post Trip Analysis O

i 5

O 1

7-8

- . . -. . .-. . _ . ~- ._ .. - .

Table 7.0-6 I

('"Ns_-) SONGS Unit 2, Cycle 3 Core Parameters Inout to Safety Analyses Reference Cycle Safety Parameters Units Values Cycle 3 Values Total RCS Power HWt 3478 3478 (Core Thermal Power

+ Pump Heat) l Core Inlet Steady State C F 542 to 560 542 to 560 Temperature (70% power and (70% power and above) abovel 530 to 560 530 to 560 (below 70% power) (below 70% power)

Steady State psia 200n - 2300 2000 - 23n0 RCS Dressure Rated Reactor gpm 346,000 to 396,000 to Coolant Flow 410,000 110,000 4'

Axial Shape Index LCO ASI .3 to +.3 .3 to +.3 Band Assumed for Units All kowers f4aximum CEA Insertion  % Insertion 28 28 at Full Power of Lead Rank

% Insertion 25 25 of Part-Length Maximun Initial Linear KW/ft 13.4 13.9 Heat Rate for Transient i Steady State Linear KW/ft 21.0 21.0 i

Heat Rate for Fuel CTM i Assumed in the Safety '

, Analysis _

i- CEA Orop Time from sec 3.0 3.0 Removal of Power to

Holding Coils to 90%

! Insertion Minimum ONBR CE-1 1.31 1.31 Macbeth 1.30 1.30 f% -

V i

7-9 r

1

) Table i.0-6 (continued) 4 Reference Cycle Values Safety Parameters Units (Cycle 2) g_c'e3 Values-

?

i Moderator Temperature 10 #ac/0F -2.5 to +.5 -3.J to +.5 (below 70% (below 70% power)

Coefficient power) -3.3 to 0.0

-2.5 to 0.0 (70% power and (70% power and above) above)

Shutdown fiargin (Value  % ao -5.15 -5.15 Assumed in Limiting EOC Zero Power SLB) 1.

4 t

e >

i 7-10

i 1

l 7.1 Increase in Heat Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature V The results are bounded by the Reference C/ cle.

7.1.2 Increase in Feedwater Flow The results are hounded by the Reference Cycle.

7.1.3 Increased Main Steam Flow The Increased Main Steam Flow Event is analyzed to ensure that the Departure from Nucleate Boiling Ratio (DNBR) and Fuel Centerline Melt (CTM) Spectfied Acceptable Fuel Design Limits (SAFDLs) are not violated. This event was reanalyzed due to a more adverse pin census and an increased Doppler multiplier, and the availability of the Variable Overpower Trip (V0PT).

7.1.3.1 dentification of Caus_e_s_

An Increased Main Steam Flow Event is defined as any rapid increase in steam generator steam flow other than a steam line rupture (discussed in Section 7.1.5) or an inadvertent opening of a secondary safety valve (discussed in Section 7.1.4). Such rapid increases in steam flow result in a power mismatch between core power and steam generator load demand. Consequently, there is a w decrease in reactor coolant temperature and pressure. In the presence of a negative moderator temperature coefficient of reactivity, the decrease in

> reactor coolant temperature causes an increase in core power.

The High Power Level and Core Protection Calculators (CPCs) trips provide primary protection during this event. Additional protection is provided by other trip signals including Low Steam Generator Water Level and Low Steam Generator pressure. The approach to the CTM limit is terminated by either the DNR/ Local Power Density (LPD) related trip, the Variable Overpower Trip (V0PT) or the High Power Level Trip. In this analysis, credit is taken only for the action of the CPC Low DNBR Trip or the V0PT in the determination of the minimum transient DNBR and maximum local linear heat generation rate. The Variable Overpower Trip is described in Reference 7-17.

The following Increased Main Steam _ Flow Events have been examined:

A. An inadvertent increased opening of the turbine admission valves caused by operator error or turbine load limit malfunction. This can result in an additional 10% flow.

B. Failure in the turbine bypass control system which would result in an opening of one or more of the turbine bypass valves. The flowrate of each valve is approximately 11% of the full power turbine flowrate. There are four turbine bypass valves for a total of 451, at full power steam flow.

C. An inadvertent opening of an atmospheric dump valve or steam generator safety valve (see Section 7.1.4) caused by operator error or failure within d Each atmospheric dump and safety valve can release the valve itself.

approximately 5% of the full power turbine flowrate.

l 7-11

q 7.1.3.2 Analysis of Effects and Conseauences As in the Reference Cycle analysis (Reference 7-1), the opening of the four steam bypass valves at HFP produces the most adverse results. The opening of the four bypass valves at full power was initiated at the conditions given in

7. 3-1. A moderator temperature coefficient (MTC) of Table

-3.3 x 10 ao / F was used in the analysis. This MTC, in conjunction with the decreasing coolant inlet temperature, results in an increase in the core heat flux. The most negative fuel temperature coefficient (FTC) with a bias of 25%, was used in the analysis. The minimum CEA worth for shutdown at the time of reactor trip for full power operation is -6.0% ap. The pressurizer pressure control system was assumed to be inoperable to minimize the RCS pressure during All other control systems were the event and reduce the calculated DNBR.

assumed to be in manual made of operation and have no significant impact on the results for this event. The Reference Cycle cited a coincident loss of AC The loss of AC power and power as the limiting single failure for this event.

subsequent reactor coolant pump coastdown occurs such that a coincident CPC Low

! Flow /V0PT occurs. This timing maximizes both the degradation in DNBR and the l

quantity of predicted fuel failure.

l 7.1.3.1 Results The Increased Main Steam Flow Event plus a single failure (loss of AC power) resulted in a CPC V0PT Trip / Low Flow Trip at 4.75 seconds. The minimum DNBR calculated for the event initiated f rom the conditions specified in Table 7.1.3-I was 1.16 compared 'to the design limit of 1. M . This corresponds to a calculated fuel pin failure of less than R*. A maximum allowable initial linear heat generation rate of 16.0 kW/ft could exist as an initial conditicn without exceeding the Acceptable Fuel to Centerline Melt Limit of 21.0 kW/ft during this transient. This amount of margin is assured by setting the linear heat rate LC0 based on the more limiting allowabla linear heat rate for LOCA (13.9 kW/ft, see Table 7.0-6).

NSSS cooldown is two hours in duration resulting in offsite doses of less than 300 REM thyroid and a whole body dose of less than 25 REM. These results are more limiting than those presented in the Reference Cycle for increased Main Steam Flow Events with a single failure.

Table 7.1.3-2 presents the sequence of events for the event initiated at HFP conditions. Figures 7.1.3-1 to 7.1.3-5 present the NSSS response of core power, core heat flux, RCS pressure, RCS temperatures and steam generator pressure. The DNBR response for Cycle 3 as a function of time is presented in Figure 7.1.3-6.

The results of the Increased Main Steam Flow without a single failure would be no more adverse than those presented in the Reference Cycle.

7-12

O 7.1.3.4 Conclusions For the Increased Main Steam Flow Events with a single failure, the radiological doses are less than the 10CFR100 limits of 300 REM for thyroid and 25 REM for whole body. For the Increased Main Steam Flow Event without a single failure, the' DNBR and CTM limits are not exceeded.

7.1.4 Inadvertent Opening of a steam Generator Atmospheric Oumo Valve The results are bounded by the Reference Cycle.

O

,O i f 7-13

)

Table 7.1.3-1 Key Parameters Assumed for the Increased

'~

Main Steam Flow Event Reference Cycle Cycle 3 Parameter Units Value Value J

Total RCS Power MWt 3478 3478

! (Core Thermal Power + Pump Heat)

O Initial Core Coolant Inlet F 560 560 Temperature i

Initial Reactor Coolant System psia 2200 2200 Pressure Initial RCS Vessel Flow Rate gpm 396,000 396,000 Moderator Temperature Coefficient x10 #ap/0F -3.3 -3.3 l

CEA Worth 6t Trip %Ap -4.5 -6.0 Doppler Coefficient Multiplier 1.15 1.25 i

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l 1D 7-14

_ _ _ --- . _ _ . _ _ - _ _ _ _ ~ _ . , _ _ _ _ _ - . _ _ _ . . _ _ , . _ . - . _ _ _ _ _ . . _ , - - , _ _ _ _ . _

Table 7.1.3-2 Sequence of Events for the increased

> Main Steam Flow Event Plus a Single Failure 4

Time (sec) Event Setpoint or Value 0.0 Ouick Open Signal Generated. ---

' Four Bypass Valves Start to Open 1.0 Four Bypass Valves Full Open 145% of full steam flow 4

8.95 Loss of All On and Offsite Power, ---

Turbine Admission Valves and Rypass Valves Start to Close, Feedwater Begins to Coast Down, Reactor Coolant Pumps 9egin to Coast Down CPC V0PT Trip / Low Flow Signal 116% of 3410 MWt, 4

9.75 95% of shaft speed Generated i 10.0 Reactor Trip Breakers Open, ---

! Turbine Trip

10.3 CEAs Regin to Orop in the Core ---

10.7 Maximum Core Power 117.6% of 3410 P4lt 12.1 Maximum Core Heat Flux 110.6% of 3410 "Wt 12.3 Minimum ONBR Occurs (CE-1) 1.16 i

! 12.75 Turbine Admission Valves and ---

Rypass Valves Closed 16.7 Steam Generator Safety Valves Open 1100 psia i

28.05 Feedwater Flow Reaches 5% of Full Power s

O 7-15

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TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 t

INCREASED MAIN STEAM FLOW PLUS SINGLE FAILURE i

O CORE POWER VS TIME FIGURE 7.1.3-1 7-16

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O PLUS SINGLE FAILURE CORE HEAT FLUX VS TIME FIGURE 7.1.3-2 7-17

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NUCLEAR GENERATIN'.i STATION Units 2 & J l INCREASED MA M STEAM FLOW l PLUS SINGLE FAILURE l RCS PRESSURE VS TIME FIGURE 7.1.3-3 7-18

> s 700 g g g g g 650 -

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' PLUS SINGLE FAILURE b)

(./ RCS TEMPERATURES VS TIME FIGURE 7.1.3-4 7-19

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1300 I I 1200 -

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SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 INCREASED MAIN STEAM FLCW PLUS SINGLE FAILURE STEAM GENERATOR DRESSURE VS TIME

, FIGURE 7.1.3-5  ;

e 7-20

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SAN ONOFRE NUCLEAR GENERATING STATION I Units 2 & 3 O INCREASED MAIN STEAM FLOW PLUS SINGLE FAILURE ONBR VS TIME FIGURE 7.1.3-6 7-21

1 7.1.5 Steam System Pioing f]

V Failures in the main steam system piping were analyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.

Steam System Pioing Failures: Inside and Outside Containment Pre-7.1.5a Trio Power Excursions This event was analyzed to evaluate the maximum number of calculated fuel pin failures for the site boundary dose calculation.

7.1.5a.1 Identification of Causes ~

A rupture in the main steam system piping increases steam flow from the steam generators. This increase in steam flow increases the rate of RCS heat removal a decrease in core cool ant inlet by the steam generators and causes temperature. In the presence of a negative moderator temperature coefficient of reactivity (MTC), this decrease in temperature causes core power to increase.

The excursion in core power is terminated by the action of one of the following Core Protection Calculators (CPCs), Low Reactor Protection System (RPS) trips: Linear power Level, or High Containment Steam Generator Pressure (LSGP), High pressure.

7.1.Sa.2 Analysis of Effects anet '.onsecuences

(

Steam Line Rreaks (SLRs) inside containment may be costulated to have begak areas up to the cross section of the largest main steam pipe (7.41 ft ).

Those SL95 occurring outside the containment bg)ilding which have upstream are located break areas of limited by the areas of the flow restrictors (d.13 ft -

the containment penetrations.

Inside containment SL9s may cause env conmental degradation of sensor input to the CPCs and pressure measurement $* stems. Additionally, the higher linear No power level trip undergoes temperatt ce decalibration due to RCS cooldown.

credit is taken for CPC action during this event. Trips which are credited for inside containment SLBs are: LSGP, High Linear Power Level or High Containment Pressure. Additionally, the environmentally degraded value of the Delta Pressure Low Flow trip is used to determine the most adverse timing of a loss of AC Power (LOAC).

Outside containment SLRs are not subject to the same environmental effects as Therefore, the full array of RAS trips the inside containment breaks.

including the CPC Low ONBR trip, are credited for these breaks.

In the Reference Cycle, an extensive parametric analysis in both This MTC parametric and break area was performed on the inside containment SLB event.

analysis identified the limiting inside containment SLB event in terms of fuel ein f ailure caused by the pre-trio power excursion. Table 7.1.5a-1 of the

,' Reference Cycle (Reference 7-1) lists the values of key parameters used in tne parametric analysis.

l Om 7-22

- . . . , . . _ _ _ . __m _ _ . - . . , __m.-

1 l

The inside containment SLB event was reanalyzed in Cycle 3 to accomodate a more adverse pin census changes in other Key Parameters for Cycle 3 are within the ranges used for the Reference Cycle Parametric Study. The Reference Cycle results (heat fl u x , RCS temperatures, pressure and flow rate) were combined with the oin census to yield a value for predicted fuel failure.

7.1.5a.3 Results The outside containment SLRs are bounded by the Reference Cycle, since they are subject to a rapid RPS trip on Low ONBR. This trip provides timely termination a

of the power excursion preventing the fuel design limits from being exceeded.  ;

The radiation release accompanying these outside containment breaks are less severe than the outside containment Double Ended Guillotine Break examined in Section 7.1.5b for the post-trip return to power.

Based on the transient response of the Reference Cycle parametric for the limiting break, the number of calculated fuel pin failures for the inside i containment SLB event is less than Rt.

The inside containment SLB event resulted in site boundary doses less than 300 5

RE$1 to the thyroid and less than 25 REM whole body.

7.1.5a.4 Conclusions The results of this analysis demonstrate that a coolable geometry is maintained during this event as the number of fuel pins calculated to fail is less than 8 Site boundary doses are calculated to be less than the 10CFR100 O percent.

guidelines.

i 4

1 0

9 i

l l T 7-23

_ _ _ . - _ , ~ . _ _

as-*

7.1.5b Steam System Pioing Failure, Post-Trio Return to Power The Hot Full Power (HFP) Steam Line Break (SLR) Event was reanalyzed due to a more adverse moderator cooldown curve and an increase in maximum inverse baron

! worth. The HFP SLB with loss of AC (LOAC) power was reanalyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.

! 7.1.5b.1 Identification of Causes A break in the main steam system piping will cause an increase in steam flow.

This increase in flow results in increased heat removal from the Reactor i Coolant System (RCS). In the presence of a negative moderator temperature coefficient of reactivity (MTC) the cooldown will cause positive reactivity to be added to the core. Highly negative MTCs and large break sizes can combine to degrade shutdown margin and may cause a return-to-power.

1 This approach to criticality is terminated by the addition of safety injection boron and the increase in temperature following either, l

i 1. Termination of steam flow and heat removal by the action of the MSIVS in both steam lines, or i 2. Termination of steam flow from the unaffected steam generator by the MSIV action and dryout of the affected steam generator.

j Q

V The Hot Full Power (HFP) and Hot Zero Power (HZP) Steam Line Break (SLB) Events were analyzed to determine that critical heat fluxes are not exceeded during this event and site boundary doses do not exceed 10CFR100 guidelines.

7.1.5b.2 Analysis of Effects and Consecuences.

< The analytical basis for the HFP simulation are discussed below.

A. A Double-Ended Guillotine break (7.41 ft 2) causes the greatest cooldown

- of the RCS and the most severe degradation of shutdown margin, i

B. A break insice the containment building, upstream of the MSIVs causes a non-isolatable constion in the affected steam generator. This results in continued shutdown margin degradation until the affected steam generator blows dry.

! C. A reactor trip is initiated by either Low Steam Generator Pressure, Low I Steam Generator Water Level, High Linear Power level, Low DNBR, or Delta-Pressure Low Flow Trip (Loss of AC Power).

O. The cooldown following a steam line break results in contraction of the reactor coolant. For this analysis, if the pressurizer empties, the reactor coolant pressure is set equal to the saturation pressure corresponding to the highest temperature in the reactor coolant system.

O E. A safety injection actuation signal (SIAS) is actuated when the pressurizea pressure drops below the setpoint. Time delays associated with the safety injection pump acceleration, valve opening, and flushing of the unbarated

• r r

7-24

- - - -- - - ~. _-

safety injection lines are taken into account. Additionally, the event was

[V} initiated from the highest pressure allowed by the technical specifications to delay the effect of safety injection baron.

F. The cooldown of the 1CS is terminated when the affected steam generator blows d ry . As the coolant temperatures begin increasing, positive reactivity insertion from moderator reactivity feedback decreases. The decrease in moderator reactivity combined with the negative reactivity inserted via boron injection cause the total reactivity to become more negative.

The conservative assumptions included in the HFP simulation are discussed below.

The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to the most negative value allowed by the Tecnnical Specifications. This negative MTC results in the greatest positive reactivity addition during the RCS cooldown caused by the steam line break. Since the 4

j

reactivity change _ associated with moderator feedback varies significantly over the range of moderator density covered in the analysis, a curve of reactivity insertion versus moderator density rather than a single value of MTC is assumed in the analysis. The moderator cooldown curve usad in ' the analysis was conservatively calculated assuming that on reactor t ri p, the highest worth control element assembly is stuck in the fully withdrawn position.

The reactivity defect associated with fuel temperature decrease is also based on a most negative Fuel Temperature Coefficient (FTC). This FTC, in

( conjunction with the decreasing fuel temperatures, causes the greatest positive

( reactivity insertion during the steam line break event. The bias on the FTC assumed in the analysis is given in Table 7.1.5b-1. The del ayed neutron fraction assumed is the maximum absolute value including uncertainties for end-of-life conditions. This too maximizes subcritical nultiplication and thus enhances the potential for Return-to Power (R-T-P).

l The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power level is -8.2Ruc . This available scram worth corresponds to the moderator cooldown curve and stuck cod worth j

used in the analysis.

Ouring the return-to-power, negative reactivity credit was assumed in the

analysis. This negative reactivity credit is due to the local heatup of the i inlet fluid in the hot channel, which occurs near the location of the stuck CEA. This credit is based on three-dimensional coupled neut ronic-thermal-hydraulic calculations performed with the HERMITE/ TORC code (References 7-11 and 7-12) for Calvert Cliffs Unit 1 Cycle 7 (Reference 7-13). Only a fraction of the negative reactivity credit justified for Calvert Cliffs Unit 1 Cycle 7 was used.

The analysis assumed that, on a safety injection actuation signal, one high pressure safety injection pump fails to start. A maximum inverse boron worth of 110 ppm /% ac was conservatively assumed for safety injection. A conservative l MSIV closure time of 10.0 seconds was assumed in this analysis.

O Q

7-25

1 l

7.1.Sh.3 Results The Hot Zero Power SLB events are not presented since the Reference Cycle The results bound Cycle 3.for radiological releases and post-trip criticality.

Hot Full Power SLB with no Loss of AC results are bounded by the Hot Full Power SLB with concurrent LOAC presented herein.

Table 7.1.5b-2 presents the sequence of events for the HFP SLR with concurrent LOAC. The key plant parameters of core power, core heat flux, RCS pressure, RCS temperatures, steam generator pressure and reactivity are shown in Figures 7.1.5b-2 through 7.1.5b-7 The minimum post-trip DNBR experienced during the transient was 1.36 using the Macbeth low flow ONBR correlation. This value results in no calculated fuel i failure during the course of this transient.

i ,

7.1.5b.4 Conclusions The results of this analysis demonstrate that since there is no calculated fuel failure, a coolable geometry is maintained, and the Cycle 3 radiological release is bounded by the Reference Cycle. In addition, since the return-to-power is negligible, sufficient shutdown margin exists to terminate the event.

i

=

I l

O 7-26 i

'g Table 7.1.5b-1

{Y Key Parameters Assurred for the Steam Line Rreak Event Reference Cycle Cycle 3 Parameter A Uniti Value Value Hot Full Power Total RCS Power, MWt 3478 3478 (Core Thermal Power +

Pump Heat)

Initial Core Coolant Inlet 560 560 Temperature. OF -

Initial RCS Vessel Flow Rate, 356,400 356,400 GPM Initial Reactor Coolant 23nn ?300 System Pressure psia Doopler Coefficient Multiplier 1.15 1.15 m 7.5 -3.3 Moderator Tecoer3ture0 Coefficient, 10 M/ F

! CEA Worth at Trip, % M -6.9 R . 2.8 Inverse Baron Worth, pon/1 2 95 110 i

Initial Steam Generator 476 976 Pressure, psia Steam 9ypass Control System Incoerable Inoperable Pressurizer Pressure Inoperable Inoperable Control System High Pressure Safety Injection One Pump One Puno Pumps Inoperable Inoperable Break Area, ft 7.41 7.41 Moderator Cooldown Curve Figure 7.1.5b-1 Figure 7.1.5b-1 i

h 3

7-27 ( ,

f f

n -. -- , n. - --, ..,,n --~ ,

i Table 7.1.5b-2 2

Seouence of Events for the Hot Full Power, 7.41 ft , ,

Inside Containment Steam Line Break with Loss of Offsite Power l

Setooint or Value 4

Time (sec) Event 0.0 Double-Ended Guillotine Break 7.41 ft 2 in a Main Steam Line with Concurrent LOAC, Reactor Coolant Pumps Begin to Coast Down Reactor Trip Signal Generated on 675 psia 2.4 Low Steam Generator Pressure, Main Steam Isolation Signal 2.8 Trip Breakers Open ----

i 3.1 CEAs Begin to Drop 3.3 MSIVs Begin to Close ----

_ 13.3 'MSIVs are Completely Closed ----

17.3 Pressurizer Empties ----

Safety Injection Actuation Signal 1560 psia 17.7 7

Generated on Low Pressurizer Pressure

! 18.9 Safety Injection Pumos Reach Full ----

Speed

< 109.8 Affected Steam Generator Empties ----

132.6 Maximum Post-Trip Power R.3% of M10 MWt l

139.1 Minimum Post-Trip McBeth DNBR >1.10 141.2 -\ Maximum Post-Trip Reactivity .07A% ,-

i 1R00.0 Plant Cooldown Initiated by Manual ----

' Control of the Atmospheric Steam l

i Dump Valves for the Intact Steam i Generator

\ ,

l

-.[

t f!

E

\

s 7-28

i O ,

i 8 i i i e i i i i 7 - CYCLE 3 m HFP 4

. 6 -

CYCLE 2 HFP + j -

5 -

i 2

• • CYCLE 3 -

4 - .r ,

" HZP 3

t:

h3

~

CYCLE 2 HZP W

w "2 -

5 e

Q1 5

0 - ,

-1 -

-2 -

-3 40 42 4,4 46 48 50 52 54 56 58 60

' MODERATOR DENSITY, lbrn/ft l, i

(

SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3

.,-STEAM.LINE BREAK O, , ,

COMPARATIVE MODERATOR REACTIVITIES

FIGURE 7.1.5b-1

\\ 7-2's , i l

0 0 g g  ;

100 -

E go __

2 t 5

y 60 --

t Os 3 m.

g 40 --

S 20 --

k _.

I I I I 0

O 40 80 120 ;6C 200 TIME, SECONOS SAN ONOFRE NUCt. EAR GENERATING STATION Units 2 & 3 FULL POWER STEAM LINE BREAK sd WITH LOSS OF AC POWER

! CORE POWER VS TIME -

l FIGURE 7.1.5b-2 j 7-30

O 1

120 j I I l I l

1 l

100 E

3 80 -.

5 E

n 5 60 -

J 5m O s= 40 -

E O

20 -

1 I I I I 0

O 40 80 120 160 200 TIME, SECONDS f I

! SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3

~

FULL POWER STEAM LINE BREAK, WITH LOSS OF AC POWER O- CORE HEAT FLUX VS TIME FIGURE 7.1.5b-3 ,

7-31

l l

2400 i l g i 2100 5

E

1. 1800 -

5 0

f '

x W -

sm 1500 -

5 5

8 E 1200 -

O 5

000 -

I I I I I

E00 80 120 160 200 O 40 TIME, SECON05 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 I

FULL POWER STEAM LINE BREAK j WITH LOSS OF AC POWER RCS PRESSURE VS TIME FIGURE 7.1.5b 4 7-32

650 g l l l 1

l

~

, 600 -

C vi u

iE 550 z

5 G

~

500 -

=

O e 5

T -

g 450 -

OUT y u

i T 4yg

~

400 -

T

!g 350 O 40 80 120 160 200 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 O FULL POWER STEAM LINE BREAK WITH LOSS OF AC POWER RCS TEMPERATURES VS TIME i FIGURE 7.1.5b-5 I

7-33

i D

4 V 1200 , ,  ; i I

1000 -

UNAFFECTED SG

=: -

, @' 800

~

1 w

u 600 --

E t2

, a:

Y a

@ 400

=

200 -

, AFFECTED SG 7 i .

40 80 120 160 200 0

TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3

! FULL PCWER STEAM LINE BREAK WITH LOSS OF AC POWER STEAM GENERATOR PRESSURE VS TIME FIGURE 7.1.Sb-6 7-34

~ .-. .- . - - , . - . .

^~ _

i O

9 I l i i MODERATOR ~

6 _

l 3 -

DOPPLER d

SI BORON g

o x 7L -  ;

O =

5

" TOTAL 3-0 FEEL >tfACK

-3 -

-6 .

SCRAM I I i l

-9 0 40 80 120 160 200 TIME, SECONOS 1

i SAN ONOFRE NUCLEAR GENERATING STATION Units 2 8 3 l FULL POWER STEAM LINE BREAK WITH LOSS OF AC POWER REACTIVITIES VS TIME FIGURE 7.I.5b.7 7-35

7.2 Decrease in Heat Removal by the Secondary System 7.2.1 Loss of External Load The results are bounded by the Reference Cycle.

7.2.2 Turbine Trio The results are bounded by the Reference Cycle.

7.2.3 Loss of Condenser Vacuum The.results are bounded by the Reference Cycle.

7.2.c Loss of Normal AC Power The results are bounded by the Reference Cycle.

7.2.5 Loss of Normal Feedwater The results are bounded by the Reference Cycle.

b 7.2.6 Feedwater System Pioe Rreak Event The feedwater system pipe break event is analyzed for Cycle 3 to demonstrate that the RCS pressure faulted stress limit of '1000 psia is not exceeded during the transient. This event was reanalyzed on the basis of an assumed increase in the numoer of plugged steam generator tubes and a change in the Doppler multiplier.

7.2.6.1 Identification of Causes The rupture of a feedline will cause rapid reduction of the liquid inventory in the affected steam generator and therefore partial loss of the secondary heat sink. This leads to the heatup of the RCS and an increase in primary pressure.

Depending on initial conditions, break size, break locations and steam generator inventory, any of the several Plant Drotective System (PPS) actions may occur. A decrease in the steam generator water level will initiate a reactor trip on low steam generator water level. The decrease in the steam generator pressure may result in a low steam generator pressure trip signal and cause the main steam isolation valves and the main feedwater isolation valves to close. The partial loss of the secondary heat sink causes the RCS to heat up. This may_ result in a high pressurizer pressure trip. Additional protection against complete loss of secondary heat sink is provided by automatic initiation of emergency feedwater to the intact steam generator.

O (Ji 7-36

3 7.2.6.2 Analysis of Effects and Consecuences The feedwater line break analyzed was assumed to occur during full power operation with concurrent loss of non-emergency AC power at time of trip.

This is limiting from the standpoint of potential RCS pressure increase, since this results in the maximum initial stored energy and minimum steam generator i nventory. In addition, in response to loss of non-emergency AC power upon trip, the following were assumed to occur to maximize the RCS pressure increase:

1. Turbine stop valves close immediately:

2. Reactor coolant pumps begin to coastdown; and

3. Dressurizer control systems are lost.

4 The limiting break size was established by the parametric study reported in the FSAR. The initial RCS pressure and initial steam generator inventory are selected such that the low steam generator water level trip and the high pressurizer pressure trip occur simultaneously. This results in the maximum f

peak RCS pressure after trip. A MSIV closure time of 10.0 seconds is consarvatively assumed for this analysis.

7.2.6.3 Results The feedwater line break event was initiated at the conditions shown in Table 7.2.6-1. This combination of parameters maximizes the calculated RCS peak

, pressure. Table 7.2.6-? presents the sequence of ever}ts for this event.

L Figures 7.2.6-1 through 7.2.6-6 present the NSSS response for core power, core heat fl u x , RCS temperatures, RCS pressure, pressurizer pressure, and steam generator pressure.

The results indicate that the reduction nf the secondary heat sink due to the discharging of saturated water through the feedwater line break and the subsequent emptying of the affected steam generator cause the RCS pressure to

' increase to 2943 psia compared to the Reference Cycle reported value of 2930 psia. Following reactor trip on high pressurizer pressure / low steam generator water level, the decay in core power and the action of the prima ry and secondary safety valves result in a reduction of the RCS pressure. The RCS pressure continues to decrease until low steam generator pressure initiates the closure of Main Steam Isolation Valves (MSIV). The MSIV closure terminates the blowdown of steam through the break thus causing the RCS to heat up once more.

Eventually, the heatup is termigted by the. opening of secondary safety valves.

7.2.6.4 Conclusions The results of this analysis demonstrate that the Feedwater System Pipe Rreak Event will not result in a peak RCS pressure which exceeds the faulted stress pressure limit of 3000 psia.

J 7-37

i l

l Table 7.2.6-1 i

Key Darameters Assumed for the Feedwater System Pioe Break Event 4

+

Reference Parameter Cycle Cycle 3

'1 Initial Core Power Level, MWt 3478 347R 560 560 Initial Inlet Coolant l Temperature, OF 132.2 132.2 IngtialCoreMassFlowRate, 10 lbm/hr Initial Steam Generator Pressure, psia 971 040 s '

Initial RCS Pressure, psia 2240 2240 Modegator Temoerature Coefficient 0.0 0.0 (10~ ao/ F)

Fuel Temperature Coefficient 0.R5 0.75 ,

. Multiplier Minimum CEA Worth at Trip, tac -6.00 6,00 Steam Bypass Control System Inoperative Inoperative ,

l Pressurizer Pressure Control System Automatic Autonatic Mode Mode Pressurizer Level Control System Inoperative Inoperative 2 0.2 Feedwater Line Break Area, ft 0.2 Initial Intact Steam Generator 169,830 169,830 Inventory, ihm Auxiliary Feedwater Capacity 700 700, assuming one failed pump, gpm i

Nunher of Assumed Dlugged Steam 200 1000 Plugged Steam Generator Tubes l

f, l

O 7-38

i l

l Table 7.2.6-2 Sequence of Events for the Feedwater System Pioe Break Event d

Time (seci Event Setpoint or Value i 0.0 Rupture of Main Feedwater Line ---

34.8 Affected Steam Generator Empties ---

35.0 Low Steam Generator Water Level 27.03 ft Trip Condition Occurs in Intact Steam Generator s

High Pressurizer Pressure Trip 2475 psia Condition Occurs 35.6 Pressurizer Safety Valves Open 2525 psia 35.0 Trip Breakers Open; ---

Normal Onsite and Offsite Power Lost ---

36.2 CEAs Begin to Drop into Core ---

40.0 Steam Generator Safety Valves Open 1100 psia 1 40.2 Peak RCS Pressure Occurs 29a3 psia Peak Steam Generator Pressure Occurs llan psia 43.1 AA.a Pressurizer Safety Valves Close 2400 psia 48.6 Maximum Pressurizer Liquid Volume 1270 't 3 f

70.4 Steam Generator Safety Valves Close Ib56 psia 88.9 Emergency Feedwater Enters Intact Steam ---

Generator 212.7 Steam Generator Lcw Pressure Trip 675 psia Condition and MSIS Initiated l

223.6 Complete Closure of Main Stean isolation ---

5 Valves Terminating Blowdown from the

- Intact Steam Generatnr 23a.Q Minimun liquid fiass in the Steam Generator 7527 lbn Connected to Intact Feedline

( 1800. Operator Opens the Atmospheric Steam Dunp ---

Valves to begin Plant Cooldown to Shutdown l

Cooling l ,

7-39

O 0 g g g g g 100 -

f3 80 -

l

'~' l D

y 60 -

O i s

y 40 -

8 20 -

I I I I I 0

O 50 100 150 200 250 300 TIME, SECON05 SAN ONOFRE NUCLEAR GENERATING STATION Unita 2 & 3 FEECWATER SYSTEM PIPE BREAK O ' CORE POWER VS TIME FIGURE 7.2.6 1 7-40

i l

120 l l l l l l

4 100 -

3 E

S 80 -

5 i

E 60 g -

J 3

1 O M 40 -

U ,

! v i l l

! 20 -

l I I I I i 0

0 50 100 150 200 250 300 4

TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION

' Units 2 & 3 i

FEEDWATER SYSTEM P!PE BREAK CQRE, HEAT FLUX VS TIME FIGURE 7.2.6-2 7-41

l 1

1 l l 300a'  ; l g i l I

RCS PRESSURE 2800 1 l 5 E

i .

~"

W 2600 -

S E

z Y -

O m

2400 -

n 8

j 8 2200 -

U 5

cc 2000 -

PRESSURIZER PRESSURE 1800 O 50 100 150 200 250 300 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3

FEECWATER SYSTEM PIPE BREAK

RCS PRESSURE VS TIME FIGURE 7.2.6-3 7-42

O g

660 , , i i 4 640 -

o' A

m 620 -

T OUT w

, 3 -

^ =

1 600 -

1 5 E

8 i O , 580 -

O v -

T;g 3"

560 t

i l

i I I ' '

I 540 200 250 300 O 50 100 150 TIME, SECON05 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 FEECWATER SYSTEM PIPE BREAK RCS TEMPERATtJRES VS TIME f

FIGURE 7.2.6-4 7-43

. _ - - - - _ . . . _ . - - - . - . - . . - _ . - - . - _ . . - - . - - - - - - _ _ .~.

.O 1200 g g g g g 1000 --

4 UNAFFECTED SG

@ 800 -

w E

u 600 -

5 E

ac

, O Y u

h 400 -

l  :

l AFFECTED SG 200 -

l I

I I I I I O

f 0 50 100 150 200 250 300 TIME, SECONDS f

i l

~

SAN ONOFRE NUCLEAR GENERATING STATION l

Units 2 & 3 O FEE 0 WATER SYSTEM PIPE BREAK STEAM GENERATOR PRESSURE VS TIME FIGURE 7.2.6-5 -

7-44

l

! l 1

7.3 Decrease in Reactor Coolant Flowrate

! 7.3.1 Partial loss of Forced Reactor Coolant Flow The results are bounded by the Reference Cycle.

l 7.3.2 Total Loss of Forced Reactor Coolant Flow 1

i. The loss of Coolant Flow (LOF) Event is analyzed to determine the minimum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) such that in conjunction with the Reactor Protection System (RPS) the DNBR SAFDL will not be exceeded.

This event was reanalyzed due to a

reduction in CEA worth at trip. The method used to analyze this event is the same as the method described in Reference 7-14, Appendix A.

7.3.2.1 Identification of Causes A loss of normal coolant flow may result either from a loss of electrical power l to one or more of the four reactor coolant pumps or from a mechanical failure, I such as a pump shaft seizure. Simultaneous mechanical failure of two or more pumps'is not considered credible. If the RCP shaft speed reduction from either cause is greater than the CPC low pump speed trip setpoint, a reactor trip is

! initiated,

i. Reactor trip on loss of coolant flow is initiated by the CPC's on low RCP shaft speed. For a loss of flow at full . power operating conditions, a trip will be

( initiated when the RCP shaft speed drops to 95 percent of its initial speed.

For conservatism, the safety analysis assumes that the CPC's initiate a reactor trip when the reactor coolant flow reaches 95 percent. The reduction in core j

4 flow lags the decrease in RCP shaft speed.

7.3.?.2 Analysis nf Effects and Consequences

. The transient is characterized by the flow coastdown curve given in Figure 7.3.2-1. Table 7.3.2-1 presents the initial conditions assumed in this event.

3 7.3.2.3 Results Table 7.3.2-2 presents the sequence of events for the 4-pump Loss of Flow e Event. This is a representative case and is initiated at a shape index of i

zero. The low flow trip setooint is reached at .R0 seconds and the scram CEas

start dropping into the core 0.52 seconds later. A minimum CE-1 ONBR of 1.31

! is reached at 2.7 seconds. Figures 7.3.2-2 to 7.3.2-5 present the core power, heat flux, RCS pressure, and RCS. temperatures as a function of time.

7.3.2.4 Conclusions i

The event initiated from the Technical Specification LCOs in conjunction wi*.h

) the CPC los RCP shaft speed trip will not exceed the ONBR SAFDL.

!, O

' 7-45

Table 7.3.2-1 Key Parameters Assumed for the Total loss of Forced Reactor Coolant Flow Event Reference Cycle Cycle 3 Units Value Value 4

Parameter MWt 3478 3478 Total RCS Power (Core Thermal Power

+ Pump Heat)

F 560 560 Initial Core Coolant Inlet Temperature gpm 396,000 306,00n i Initial RCS Vessel Flow Rate I

psia 2325 2325 j Initial Reactor Coolant l System Pressure Moderator Temperature Coefficient x10*#co/0F +.5 +.5

.85 75 Doppler. Coefficient Multiplier --

Low Pump Speed Trip Setpoint 0.95 0.95 (Reactor Coolant Pump Shaft Speed Setpoint) 4 Low Pump Speed Trip Response Time sec 0.22 0.27 CEA Holding Coil Delay sec 0.3 0.3 CEA Time to 90% Insertion sec 3.0 3.0

(Including Holding Coil Delay) i

! CEA Worth at Trip (all rods out) too -6.25 -6.0 4-Pump RCS Flow Coastdown Figure 7.3.2-1 Figure 7.3.2-1 i

i l

O 7-46

i Table 7.3.2-2 Sequence of Events for Total loss of Forced Reactor Coolant Flow Event

\ *

Time (sec) Event Setpoint or Value 0.0 Loss of Power to all Four Reactor --

Coolant Pumps i

! 0.R0 Low Reactor Coolant Pump Shaft Speed 95% of shaft speed Trip Signal Generated 1.02 Trip Rreakers Open --

4 1.32 CEAs Regin to Orop into Core --

i 1 2.70 Minimun CE-1 DNBR > 1.31 4.7 Maximum RCS Pressure 2523 psia i

F

)

I i

l l

1 t

i l

l l

l 7-47

. . _ . . . . , _ _ _ . . . _ . _ , _ _ . . _ _ _ _ . _ _ _ - . _ . _ . - . _ _ _ _ _ _ _ _ . . _ _ _ . -_ _,__ _ _ _ _ _ _ . _ _ . _ . . _ ~ . _ _ _ _ _

O 25000 i i  : l 1  :

i i

i 20000 5 ,

i "w

1 R i E

a

- 15000 -

3 l d i

w i 8 v

i 10000 O

5000 O 2 4 6 8 10 12 4 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 i

' TOTAL LOSS OF FORCED REACTOR j j COOLANT FLOW

! COPE FLOW VS TIME FIGURE 7.3.2-1 l

7-48

- . . . , - _ . - . _. _ - ,- . = - . _ . , _ - _ _ . - .. - . _ - - - _ - _ . -

i 4

!O t

I 125 g g g l t 1

100 -

i

! S \ - t

!; \ \

~

l b 75 -

\\

E 0

5s i

. t 5 50 - 1 8 \

O 8

\

\

25 -

\ q  ;

j g t

l

' i i

i

! t j 0-4 2 4 6 8 10 12 i 0 TIME, SECONDS 1

i l

i SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 TOTAL LOSS OF FORCED REACTOR COOLANT FLOW CORE POWER VS TIME FIGURE 7.3.2-2 I

' 7-49

125 1 1 1 l 1

l

- 100 -

E I

i E

e -

75 -

I 5  !

M E

l $

d 50 - ,

j E

! W W

l 8

25 -

t I  !

t

~

0 3

4 6 8 10 12 0 2 i

TIME, SECONOS k

i SAN ONOFRE l NUCLEAR GENERATING STATION Units 2 & 3 TOTAL LOSS OF FORCED REACTOR COOLANT FLOW CORE HEAT FLUX VS TIME FIGURE 7.3.2 3  ;

\

7-50

2625 i i  :

i t 1

2500 - N, 1 3 E \

i E \

' \

S w -

i i

"g 2375 -

5 G

f 5

l E a

2250 -

8 v

l

=

, S

. W ,

i E 2125 -

l

, t I

I l 2000 8 10 12 2 4 6 0

TIME, SECONDS 1

4 i

SAN ONOFRE

• NUCLEAR GENERATING STATION l Units 2 & 3 i

TOTAL LOSS OF FORCEO REACTOR l

! COOLANT FLOW PCS PRESSURE VS TIME FIGURE 7.3.2-4 7-51

I i

i l

j t 650 g g i  ;

4 T

620 OUT

~

i I E i

i2 I

• T X AVG

~

{

5 w

590 -

j 5

! G T

$ IN G, c.60 N,

!O i

8 8

0 l

-1

$ 530 -

i  !

t  !

' I

. 0 2 4 6 8 10 12 l TIME, SECONOS SAN ONOFRE NUCLEAR GENERATING STATION 1

Units 2 & 3 TOTAL LOSS OF FORCED REACTOR I i

COOLANT FLOW

RCS TEMPERATURES VS TIME

\ FIGURE 7.3.2 5 7-32

1 m

7.3.3 Single Reactor Coolant Dumo Shaft Seizure / Sheared Shaft (v)

The single reactor coolant pump sheared shaft (SSI was reanalyzed due to a change in the fuel failure pin census. The SS was reanalyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.

7.3.3.1 Identification of Causes A single reactor coolant pump sheareo shaft is caused by mechanical failure of the pump shaft. Following the shearing of a reactor coolant pump shaft, the core flowrate rapidly decreases to the value that would occur with only three reactor coolant pumps operating. The reduction in coolant flowrate causes an increase in the average coolant temperature in the core and may produce a departure from nucleate boiling (ONB) condition in some portions of the core.

A reactor trip is generated when the rapid flow reduction across the steam generator in the affected loop decreases the delta-pressure belowFollowing the trip setpoint. The reactor trip produces an automatic turbine trip.

turbine trip, offsite power is available to provide AC power to the auxiliaries. The operator can initiate a controlled system cooldown using the turbine bypass valves any time af ter reactor trip. The steam release to the atmosphere, even if operator action is delayed for 30 minutes following first indication of the event, would be no more than that following a loss of all normal AC power.

7.3.3.2 Analysis of Effects and Consequences L

V) The sheared shaft was assumed to occur at hot full power and at core thermal hydraulic conditions such that the minimum thermal margin is being reserved by the Core Operating Limit Supervi sory System (COLSS). Table 7.3.3-1 contains the initial conditions for Cycle 3 and Reference Analysis (Reference 7-21. No credit was taken for heat flux decay upon reactor trip. This method essentially trades the initial reserved margin off against a reduction of core flow to 75% of its initial value. This method is extremelf conservative. The minimum DNBR for this event was calculated with the TORC computer code.

7.3.3.3 Results The sheared shaft results in a minimum calculated ONBR of 1.12 compared to the design limit of 1.31. This results in a predicted fuel failure of less than 91 The Acceptable Fuel to Centerlina Melt of 21 kw/f t is not violated. The resultant offsite doses are less than 300 REM thyroid and less than 25 REM whole body. Additionally, the peak RCS pressure is less than 2750 psia.

7.3.1.4 Conclusions For the sheared shaf t the radiological doses are less than the 10CFR100 limits of 300 REM thyroid and less than 25 REM whole body. As in the FSAR (Reference 7-2), the consequences of the sheared shaf t are more limiting than the seized rotor event.

s \

v 7-53

_ . _ . _ - . _ _ . _ _ _ _ - - _ _ . . _ . ~ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . . _ . _ _ . . . _ . . . _ . _ _ _ . . _ . - . _m _

f 4

1.

i Table 7.3.3-1 l Key Paracteters Assumed for the

! Single Reactor Coolant Pumo Sheared Shaft Event I

I i

i Reference Cycle 3 i Parameter Cycle Value Value Initial Core Power Level, mit 3478 3478 j

Core Inlet Coolant Temperature, OF 560 560  ;

i 6 136.8 156.1 l Core Mass Flowrate, 10 lbm/hr

! Reactor Coolant System Pressure, Ib/in 2a 2,0n0 235n Maximum Radial Power Peaking Factor 1.59 1.7 i

)

l ,

f 1

i ,

l

! t i

i  :

i l

l l

l l

I 7-54

7.4 Reactivity and Power Distribution Anomalics 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or low power Condition n

(dI The uncontrolled CEA withdrawal (CEAW) from subcritical or low power conditions is analyzed to ensure that the departure from nucleate boiling ratio (DNBR) and the fuel centerline melt (CTM) specified acceptable fuel design limits (SAFDLs) are not violated. Additionally, the CEAW from subcritical and low powers is analyzed to verify that the peak RCS pressure is less than the design limit of 2750 psia.

7.4.1.1 Identification of Causes .

An uncontrolled withdrawal of CEAs is assumed to occur as a result of a single failure in the control element drive mechanism (CEDM), control element drive mechanism control system (CEDMCS), reactor regulating system, or as a result of operator error.

7.4.1.2 Analysis of Effects and Consequences The withdrawal of CEAs from subtritical or low power conditions adds reactivity to the reactor core, causing both the core power level and the core heat flux to increase together with corresponding increases in reactor coolant temperatures and reactor coolant system (RCS) pressure. The withdrawal motion of CEAs also produces a time dependent redistribution of core power. These ,

transient variations in core thermal parameters result in the system's approach to the specified fuel design limits and RCS and secondary system pressure limits, thereby requiring the protective action of the Reactor protection System (RPS).

The reactivity insertion rate accompanying the uncontrolled CEA withdrawal is dependent primarily upon the CEA withdrawal rate and the CEA worth since, at suberitical and lower power conditions, the normal reactor feedback mechanisms do not occur until power generation in the core is large enough to cause changes in the fuel and moderator temperatures. The reactivity insertion rate determines the rate of approach to the fuel design limits. Depending on the system initial conditions and reactivity insertion rate, the uncontrolled CEA withdrawal transient is terminated by either a high logarithmic power trip, high power level trip, high pressurizer pressure trip, low departure from nucleate boiling ratio (DNRR) trip, high local power density trip, or variable overpower trip (V0PT).

A CEA withdrawal from suberitical was initiated from the conditiops in 0Table ,

7.4.1-1. A mc?erator temperature coefficient (MTC) of +0.5x10" ao / F was used in this analysis. This MTC, in conjunction with the increasing core coolant temperatures, yields an increase in core heat flux. The least negative fuel temperature coefficient (FTC) with a bias is used in this analysis. The minimum CEA worth assumed for shutdown at time of reactor trip for zero power operation is 5.15%Ao and 4.0%ap for subcritical (Mode 2) operation.

7.4.1.3 Results The uncontrolled CEA withdrawal from subcritical conditions resulted in a O

'V reactor trip on high logarithmic power at 75.2 seconds. The minimum DNRR calculated for this event initiated from the conditions of Table 7.4.1-1 was greater than the design limit of 1.31. The peak linear heat generation rate 7-55 1

i l

i  ;

(PLHGR) was calculated to be 26 kw/ft which is in excess of the steady state

, acceptable fuel to centerline melt (CTM) limit 0 of 21 kw/ft. However, the fuel

/i centerline temperature does not exceed 4900 F and the fuel is not predicted

( ,) to melt. Additionally, the peak RCS pressure is less than the design limit of 2750 psia. Table 7.4.1-2 presents the sequence of events for this event.

Figures 7.4.1-1 through 7.4.1-5 present the NSSS response for core power, core heat flux, RCS temperatures, RCS pressure and steam generator pressure.

The results of the uncontrolled CEA withdrawal from low power is presented due to a change in the RPS. The V0PT added to the CPCs is credited to mitigate the consequences of this event. The low power CEAWs were analyzed to maximize the RCS pressure increase and to maximize the potential for fuel degradation. The initial conditions for the CEAW that maximizes' peak RCS pressure are listed in Table 7.4.1 3. A parametric on the reactivity addition rate was performed to yield a coincident V0PT/high pressurizer pressure trip in order to maximize the peak RCS pressure. A high pressurizer pressure /V0PT is generated at 151.4 seconds and the scram' CEA's begin to drop at 151.7 seconds. The peak RCS pressure is 2640 psia and occurs at 152.9 seconds. The sequence of events is presented in Table 7.4.1 4 Figures 7.4.1-6 through 7.4.1-11 present the NSSS response for this event. Since the CEAW from low power is a CPC Design Basis Event (DBE) core thermal limits are not exce,eded.

7.4.1.4 Conclusions An uncontrolled CEA withdrawal from either subcritical or low power conditions will not exceed the ONBR or CTM limits. The RCS pressure limit of 2750 psia will not be exceeded during this event.

() 7.4.2 Uncontrolled CEA\ Withdrawal at Power The results are bounded by the Reference Cycle.

i

7.4.3 CEA Misooeration Event The results are bounded by the Reference Cycle.

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1 Table 7.4.1-1 Key Para 7?ters Assumed in the'CEA Withdrawal From .Suberitical Con itions Event Reference i Parameter Cycle Cycle 3 1

! Initi 1 Core Power (evel, MWt 347R x 10*IO 3478 x 10-10 ,

530.5 520 ,

Initial Inlet Coolant '

Temperature, OF 12R.6 150.2 IngtialCoreMass clow Rate, 10 lbm/hr '

L Initial RCS Pressure, 'sia 2000 2000 Modegator Tenperature Coefficient 0.5. 0.5 l

0

?

, (In-*to/ F)

Fuel Temperature Coefficient n.85 0.75 c Multiplier 1

Minimum CEA Worth at Trip, daa . ed.45 4.0 MaximumReacjivityAddition. 0.R 1.9 '

l Rate, fx 10~ AC/sec) +

I l 5 l

}(

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i _ .__ _ _ _ _ _ . . . _ . _ . , _ - _ . _ _ _ _ _ _ . . _ _ . _ _ - . _ . - . . . _ _ . _ . _ , . _ _ - _ _

, O Tabl e 7.4.1-2 Sequence of Events for the CEA tlithdrawal From Suberitical Conditions Event Time (sec) Event Setpoint or Value l 0.0 Initiation of lincontrolled ----

1 Secuential CEA Withdrawal 53.0 Reactor Reaches Criticality ----

74.8 Reactor Reaches High Logarithmic 2", of Rated Power Trip Setpoint

, 75.2 Reactor Trip Generated ----

75.5 CEAs Begin to Orop ----

75.6 Peak Reactor Core Power Reached 65". of 3410 P'Wt 75.7 Peak Reactor Core Heat Flux Reached 9.7% of 3410 MWt 75.7 Minimum DNBR Occurs y,1.31 i l

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Key Parameters Assumed in the CEA Withdrawal From Low Powers Event Reference l

Parameter Cycle Cycle 3 Initial Core Power Level, MWt 0.347R 34.7 A 530.5 520  ;

Initial Inlet Coolant Temperature, OF Initial Core fiass Flow Rate, 12R.6 150.2  ;

l 10" lbm/hr 2000 '!

Initial RCS Pressure, psia f

2000 0.5 0.5

! Modejator 0 Temperature Coefficient j (10~ ao/ F) i Fuel Temperature Coefficient 0.85 0.75 Multiplier Minimum CEA Worth at Trip, t1: a.45 -5.15 i i

MaxinunReacjivityAddition 0.8 1.1

]'

Rate, fx 10~ ac/sec)

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150.5 High Pressurizer Pressure 2475 psia Trip Condition 151.4 High Pressurizer Pressure /V0PT ----

Reactor Trip Occurs 151.7 Scram CEAs Begin to Orop ----

l t

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j 155.7 Pressurizer Safety Valves Close 2400 psia

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l

7.4.4 CVCS Malfunction (Inadvertent Roron Dilution)

The Inadvertent Baron Dilution event is analyzed for Cycle 3 to demonstrate that sufficient time is available for an operator to identify the cause of and to terminate- an approach to criticality for all subcritical modes of operation. The results of the analyses establish corresponding shutdown margin requirements for Modes 3 through 5. This event was reanalyzed on the basis of an increase in critical boron concentrations as shown in Table 7.4.4-1.

7.4.4.1 Identification of Causes During operation at power (i.e., Modes 1 and 2), an inadvertent baron dilution adds positive reactivity and can cause an approach to the ONBR and CTM limits.

The Core Protection Calculator (CPC) trip system monitors the transient behavior of pertinent safety parameters and will generate a reactor trip if neces'sary to prevent the ONBR and CTM limits from being exceeded. The high pressurizer pressure trip will prevent reaching the RCS pressure upset limit.

The trip which is actuated depends _ on the rate of reactivity addition. For a

?,

1 boron dilution initiated from the low power portion of Mode the power transient resulting from the reactivity insertion would be terminated by the

, high logarithmic power level trip prior to approaching these limits. For the subcritical modes (i.e., Modes 3 through Fi) , the time required to achieve criticality due to boron dilution is dependent on the initial and critical boron cnncentrations, the inverse boron worth, and the rate of dilution.

7.4.4.2 Analysis of Effects and Consecuences f Table 7.4.4-1 compares the values of the key transient parameters assuned in each mode of operation for Cycle 3 and the Reference Cycle. The analysis j conservatively assumed higher critical boron concentrations and lower inverse i

boron worths than expected for Cycle 3 These choices decrease the calculated times to criticality in initially subcritical modes. The time to criticality was determined by using the same mathematical expression as in the FSAR.

(Reference 7-2, Section 15.4.1.4.3).

7.4.4.3 Results i

Table 7.4.4-2 compares the results of the analysis for Cycle 3 with those for the Reference Cycle. The key results are the minimum times required to lose the prescribed negative reactivity in each operational mode. The Cycle 3 results are bounded by the Reference Cycle analysis for Modes 1 through 4 The time to criticality for Modes 5 and 6 have decreased due to an increase in critical boron concentration.

7.4.4.4 Conclusion The results of this analysis demonstrate that sufficient time exists' for the operator to take appropriate action to identify t'd mitigate the consequences j of the Inadvertent 9eron nilution Event.

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Table 7.4.4-1 0 xe, Parameters assumed in t*e inadvertent eeren Oiietien e e t Reference Cycle Cycle 3#

Value Value Parameter Critical Boron Concentration, PPM (All Rods Out, Zero Xenon)

Power Operation (Mode 1) 1500 2050 Startup (Mode 2) 1500 2050 Hot Standby (Mode 3) 1500 2050 1500 2050 Hot Shutdown (Mode 4) 1300 2050 Cold Shutdown (Mode 5)

Refueling (Mode 6) 1150 1650(1445)##

Inverse Roron Worth, PPM /%Ao Power Operation 70 80 Startup 60 80 Hot Standby 60 80 Hot Shutdown 60 65 Cold Shutdown 60 80 Refueling N/A N/A Minimum Shutdown Margin Assumed, tao Power Operation 5.15 5.15

?

Startup 5.15 4.0 i

Hot Standby 5.15 4.0 Hot Shutdown 5.15 4.0 Cold Shutdown 3.0 3.0 Refueling

• For Cycle 3, Technical Specification minimum refueling concentration of 1720 ppm with uncertainty is assumed. Extended Cycle Program (ECP) analysis assumes a refueling boron concentration of 2000 ppm.

1. Values assumed are ECP bounding values unless otherwise indicated.

l

1. 1. Cycle 3 specific.

7-72

. . _ . __ _ _ _ .. .._ _ _ . ._. _ . _ _ _ __._ _ _. _ . _ _ _ _ _ _ _ _ . _ . ~ . . _

h Table 7.4.4-2 Results of the Inadvertent Roron Oilution Event

(::) ,

Time to lose Acceptance Criterion Minimum Shutdown To Terminate the Event Mode Margin (Minutes) (Minutes) 4 Reference Cycle Cycle 3 Startup (Mode 2) >73 >60 15 )

Hot Standby (Mode 3) >73 >60 15 Hot Shutdown (Mode 4) >73 >60 15 i

,! Cold Shutdown (Mode 5)-

J

) RCS Full >60 >60 15 RCS Partially Drained * >60 >60 15 Refueling (Mode 6) * >60 >60 30

)

• Assumes only one charging pump is operable.

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7.4.5 Startuo of an Inactive Reactor Coolant Puno Event The results are bounded by the Reference Cycle.

7.4.6 Control Element Assembly Ejection The results are bounded by the Reference Cycle.

i 7.5 Increase in Reactor Coolant System Inventory I 7.5.1 Chemical and Volume Control System 4

The results are bounded by the Reference Cycle.

7.5.2 Inadvertent Doeration of the ECCS Ouring power Ooeration The results are bounded by the Reference Cycle.

7.6 Decrease in Reactor Coolant System Inventory i

7.6.1 Pressurizer Pressure Decrease Events (N

The results are bounded by ta Reference Cycle.

I j 7.6.2 Small Primary Line Dice Rreak Outside Containment ,

The results are bounded by the Reference Cycle, i

7.6.3 Steam Generator Tube Ruoture l

l The results are bounded by the Reference Cycle, j 7.7 Miscellaneous f 7.7.1 Asymmetric Steam Generator Events l

The transients resulting from the malfunction of one steam generator ara analyzed to determine the initial margins that must he maintained by the LCO's such that- in conjunction with the RPS (CPC high differential cold leg temperature) the .DNBR and Fuel Centerline Melt (CTM) . limits are not exceeded.

This event- is presented due to a change in moderator temperature coefficient and a change in analytical methodology.

i r

7-74

i f

7.7.1.1 Identification of Causes The four events whtctr affect a single generator are identified below:

a) Loss of Load to One Steam Generator (LL/ISG) b) Excess Load to One Steam Generator (EL/ISG) c) Loss of Feedwater to One Steam Generator (LF/1SG) d) Excess Feedwater to One Steam Generator (EF/ISG) f Of the four events described above, it has been determined that the Loss of Load to One Steam Generator (LL/ISG) Event is the limiting asymmetric event.

Hence, only the results of this transient are reported.

The event is initiated by the inadvertent closure of a Single Main Steam Isolation Valve (MSIV), which results in a loss of load to the affected steam generator. Upon the loss of load to the single steam generator, its pressure and temperature increase to the opening pressure of the secondary safety valves and its water level decreases. The core inlet temperature of the loop with the affected steam generator increases resulting in an asymmetric temperature tilt across the core. The intact steam generator " picks up" the lost load, which causes its temperature and pressure to decrease, and its water level to increase, thus causing the core average inlet temperature to decrease and enhancing the asymmetry in the reactor inlet temperatures. In the presence of 1

a negative moderator temperature coefficient the radial peaking increases in the cold side of the core, resulting in a condition which potentially could l

l cause an approach to DNB and CTM limits. The CPC nigh differential cold leg i temperature trip serves as the primary means of mitigating this transient.

Additional protection is provided by the steam generator low level trip.

7.7.1.2 Analysis of Effects and Consequences f The most negative value of the moderator temoerature coefficient is assumed to maximize the calculated severity of the asymmetry.

The LL/ISG is initiated at the initial conditions presented in Table 7.7.1-1 and is analyzed parametric on axial shape . index to determine the maximum initial margin needed to ensure the SAFDLs are not violated.

The NSSS response is generated with the CESEC code. The resulting core parameters (core flow, RCS inlet temperature, RCS pressure, and reactor trip time) are the input into a 2-0 simulation of the core using the HERMITE code.

HERMITE is used to model both the effects of the temperature tilt on radial power distribution and the space-time impact of the scram. The thermal margin changes are evaluated with the CETOP code. Information from both HERMITE and CESEC is used to determine the resultant DNBR.

7.7.1.3 Results A reactor trip is generated by the CPC's at 6.0 seconds based on high differential cold leg tempecature between the cold legs associated with the steam generators.

f 7-75

\

O Table 7.7.1-2 presents the sequence of events for the loss of load to one steam generator. Figures 7.7.1-1 to 7.7.1-5 show the NSSS response for core power, core heat flux, RC5 temperatures, RCS pressure, and steam generator pressure.

The minimum transient DNBR calculated for the LL/ISG Event is greater than 1.31.

A maximum allowabl'e initial linear heat generation . rate of 17.0 kW/ft could exist as an initial condition without exceeding the Acceptable Fuel to Centerline Melt Limit of 21.0 kW/ft during this transient. This amount of margin is assured by setting the linear heat rate LCO based on the more limiting allowable linear heat rate for LOCA (13.9 kW/ft, see Table 7.0-6).

7.7.1.4 Conclusions This event initiated from the Technical Specification LCO's will not exceed the DNBR and CTM limits.

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Table 7.7.1-1 Key Parameters Assumed for the loss of Load to One steam Generator Event Reference i Cycle 3 Cycle Units Value Value Parameter MWt 3478 3478 Total RCS Power (Core Thernal Power

+ Pump Heat)

OF 553 553 Initial Core Inlet Temperature Initial Reactor Coolant Systen psia 2250 2250 Pressure 0 -2.5 -3.3 Moderator Temperature Coefficient x10~#t.o / F Doppler Coefficient Multiplier 0.85 0.75 Radial Distortion Factor for a 1.158 1.130 O

1R F Core Inlet Temperature Asymmetry j

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Table 7.7.1-2 Seouence of Events for the loss of Load to One Steam Generator Event Time (sec) Event Setooint or Value 0.0 Spurious Closure of a Single Main ----

Steam Isolation Valve (MSIV) 0.1 MSIV on Affected Steam Generator is Closed 0.1 Steam Flow from Unaffected Steam Generator Increases to Maintain Turbine Power CPC' Delta-T Setpoint Reached 180 p 6.0 (Differential Cold Leg Temperature) t Safety Valves Open on Isolated Steam 1100 psia 6.1 Generator

~

6.25 Trip Breakers Open ---

6.55 CEAs Regin to Drop into Core ----

7.15 Minimum DNBR Occurs > 1.31 10.9 Maximum Steam Generator Pressure 1135 psia b

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m 4.0 ECCS Analysis l I v

821 Introduction and Sumary s

An ECCS performance ' analysis was performed for SONGS-2 Cycle 3 to demonstrate compliance with 10CFR50.46 which presents the NRC 2

Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1). The analysis justifies an allowable i

• peak linear heat generation rate (PLHGR) of 13.9 kw/ft. This PLHGR is equal to the existing limit for SONGS Unit 2. The method of analysis and detailed results which support this value are presented herein.

8.2 Method of Analysis The . method of analysis is identical to the Reference Cycle large break LOCA ECCS performance analysis 1(Reference 8-21. As in the Reference Cycle, the calculations performed for this evaluation used the NRC approved 'C-E large break ECCS performance evaluation model which is dhs'cribed in References 8-3 through 89 Blowdown and refill /reflood hydraulics and hot rod temperature calculations were performed with the fuel ' parameters which bound Cycle 3 at a - reactor power level of 3458 Mwt. The blowdown hydraulic calculations were performed with the CEFLASH aA (Reference 8-5)- code while the refill /reflood hydraulic calculations were performed with the COMPERC-II (Reference 8-6) code, i

The hot rod clad temperature and clad oxidation calculations were performed with the STRIKIN-II. (Reference 8-71 and PARCH (Reference 8-4) codes. Fuel performance calculations were performed using the FAlES 3 A

version of C-E's fuel performance code (Reference 8-9 and 8-10) as approved by the NRC (Reference 8-11) with the fuel grain '4 29 ,

restriction. Core wide clad oxidation calculations were also per d 4

in this analysis. ' '

i The significanticore and system parameters for Cycle 3 and the Reference Cycle are shown in Table A-l. ~ The Reference Cycle usad the C-E generic blowdown' analysis for the' 3400 Mwt class plants which conservatively bound the SONGS blowdown characteristics. However, a SONGS specific binwdown t, analysis was performed for Cycle 3 to account for the stean generator tube plugging. This resulted in additional input parameter

.J e

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p . _ , _ _ _ _ - _

differences between Cycle 3 and the Reference Cycle as shown in g) Table 8-1. The major differences between the Reference Cycle and the i Cycle 3 analysis are the fuel performance characteristics, steam generator tube plugging, lowering of minimum initial containment pressure, initial core inlet temperature and the core bypass flow. The other ECCS analysis input parameters are essentially the same as those of the Reference Cycle.

The Cycle 3 analysis accounts for steam generator U-tube plugging of 1000 average length tubes per steam generator. Steam generator U-tube plugging increases system resistance to flow and hence the ability of the Reactor Coolant System (RCS) to vent steam during reflood. The analysis also accounts for the decreased heat transfer area and primary side coolant volume caused by the tube plugging.

J, Additionally, to provide operationel flexibility the minimum containment pressure used was lowered from 14.40 psia to 13.7 psia.

8.3 Results Table 8-2 presents the analysis results for the 1.n DEG/PO* break which produces the highest peak clad temperature. For comparison the resul*.s of the Reference Cycle are also presented. The results of the evaluation confirm that 13.9 kw/ft is an acceptable value for the PLHGR LC0 in Cycle 3. The peak clad temperature ar.d maximum local and core wide clad oxidation values, as shown in Table 8-2, are well below the 10CFR50.46 acceptanca- limits of 22000 F, 17s, and 1%,

respectively. Table 8-3 presents a list of the significant parameters displayed graphically for the limiting 1.0 DEG/PO break.

• 0EG/PO = nouble-Ended Guillotine at Pump n ucnarge I

v 8-2

Burnup dependent hot rod calculations were performed with STRIKIN-II to i determine the initial fuel conditions which results in the highest peak clad temperature (PCT). This study demonstrated that the burnup with i

' the highest initial fuel stored energy results in the highest PCT. This occurred at a hot rod burnup of 1000 MWD /MTU.

The 1.0 DEG/PO break produced the highest peak clad temperature of 21160 F. For the 1.0 DEG/PD break the peak local oxidation (PLO) was calculated to be = 10.08%. The 1.0 DEG/PD also resulted in the highest core wide clad oxidation of less than 0.68% which is well below the 1%

NRC acceptance criterion.

A review of the effects of initial operating conditions on these'results i

was performed. It was determined that over the ranges of initial operating conditions as specified in the Technical Specifications (Section 10), operation of th? plant at a plHP,R of 13.9 kw /ft is acceptable for Cycle 3.

4 O 8.4 Conclusion J

The results of the ECCS performance evaluation for SONGS Unit 2, Cycle 3 demonstrated a peak clad temperature of 21160F, a peak local clad oxidation percentage 'of 10.08% and a peak core wide clad oxidation percentage of less than 0.6R% compared to the acceptance criteria of 22000 F, 17% and 1%, respectively. Therefore, operation of SanGS i Unit 2 Cycle 3 at a core power level of 3458 Mwt (102% of 3390 MWt) and a PLHGR of 13.9 kw/ft is in conformance with 10CFR50.46.

P i

l O

8-3 l

O Table 8-1 1 l

SONGS Unit 2 Cycle 3 Core and Systen Parameters Reference Parameter (Units) Cycle Cycle 3 Average Linear Hear Rate 0102% of of Nominal (kw/ft) 5.6 5.76 Peak Linear Heat Generation Rate (kw/ft) 13.9 13.9 4

Core Inlet Temperature (OF) 557.5 553 Core Outlet Temperature (OF) 618.6 613.5 6 6 System Flow Rate (lbn/hr) 148.0X10 148.0X1n 6

Core Flow Rate (lbm/hr) 142.8X10 143.6X10 0 Gap ConductanceUI (BTV/hr-ft2 OF) 1590.0 1639.0 Fuel Centerline Temperaturef1I (OF) 3411.0 3424.0 Fuel Average TemperatureIII (OF) 2154.0 2155.8 Hot Rod Gas PressureIII (OF) 1131.0 1111.40 Hot Rod Rurnup (MWD /MTU) 998.0 1000 Number of Stean Generator Tubes Plugged per S.G. lon 1000 Minimum Initial Containment Pressure (psia) 14.40 13.70 l

(1) Initial value at the limiting hot rod burnup as calculated by STRIKIN-II at 13.9 kw/ft.

1 O

I 8-4

. .. _ _ _ __ _ . . m _. _ __ _ _ . _. _ _ _ _ . _ . .. . . _ _

i

'l Table R-2 SONGS Unit 2 Cycle 3 4

1 limiting Break Size (1.0 DEG/PD) l t

Reference Cycle Cycle 3

Peak Linear Heat Generation 13.4 13.9 Rate (kw/ft)

< Peak Clad Temperature (OF) 2015.0 2116.0 1

' Time of Peak Clad 257.0 264.0 4 Temperature (Seconds) i .

Time of Clad Rupture (Seconds) 70.50 6R.80 i

i Peak local 10.46 10.08 i

Clad Oxidation ("i Total Core-Wide <0.6R <0.6A Clad Oxidation (%)-

8 O

8-5

f Table 8-3 SONGS Unit 2 Cycle 3 Variables Plotted as a Function of Time for the limiting large Break Figure Variable Designation Core Power 8-1 Pressure in Center Hot Assembly Nod'e A-2 Leak Flow 8-3 Hot Assembly Flow (below hot spot) 84 Hot Assembly Flow fahove hot spot) 8-5 Hot Assembly Ouality 8-6 Containment Pressure 8-7 Mass Added to Core During Reflood 8-8 Peak Clad Temperature 89 Hot Spot Gap Conductance 8-10 Peak Local Clad Oxidation 8-11 Clad Temperature, Centerline Fuel Temperature, Average Fuel Temperature and Coolant Temperature for Hottest Node A-l?

Hot Spot Heat Transfer Coefficient 8-13

, Hot Rod Internal Gas Pressure A-la i

O 6

. ~ . . .

FIGURE 8-1  ;

. CORE POWER  ;

1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG l

i 1.2001 .

1 0000 I

1,.

$ 8000 5

_a E

O ce E

j

~

6000

. oc .

LLI Z

O t Q_ .

.4000 i i

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.8

, 000bo o

o a

o o

o a

o o o  !

l o o a o o o l c.

o o -o o o .

i 1

- N M v W I TIME IN SEC 4

+

6-7

(

FIGURE 8-2 PRESSURE IN CENTER HOT ASSEMBLY NODE 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG J

2400 0 2000.0 1600.0 \

n'v cc N

w C_

1200.0 m 4 w o

W \

w 800.0 i

w L

\

\

400.0 N

e e o~

ru<a tx sec 8-8

l FIGURE 8-3 ,

LEAK FLOW 1.0 X DOUBLE ENDED GUILLOTINE BREAK  :

i IN PUMP DISCHARGE LEG 120000 100000 PUMP SIDE

--- RV SIDE

, 80000 .

e "i

60000 N

5

\

40000

\

\

\

\

20000 s N s N

y~/~-.

0 5 10 15 20 25

>Q v TIME IN SEC 8-9

FIGURE 8-4 FLOW IN HOT ASSEMBLY-PATH 16, BELOW HOT SPOT 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 30.000 20.000 l

i u 10.000 ]b p W

(n O. N 000 t

'W

- [

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r -10.000 j i O

-20 000 i

-30.000- o o o o o o o o o o S S 9 9 9 9 1 o m a w  !

i m - -

m N l O TIME IN SEC

=

8-10

'IGURE F 8-5 O

~

FLOW IN HOT ASSEMBLY-PATH 17. ABOVE HOT SPOT 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 30.000 20.000

\

u 10.000 i $

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8 9 9 9 9

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1. IME IN SEC l

l

8-11

k l

l J

O FIGURE 8-6 HOT ASSEMBLY QUALITY 1.0 X DOUBLE ENDED GUILLOTINE BREAK a

IN PUMP DISCHARGE LEG

)

NODE 13. BELOW HOTTEST REGION

_ - NODE 14. AT HOTTEST REGION

_ .- NODE 15, ABOVE HDTTEST REGION 1 0000 .

1 g

l i

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/

bl \

g i \1

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,O' TIME IN SEC i

8-12 ..

r , M--- ---

1. w+ 'W-T w y +-r+- y7 m9 -- - *-  :e e +--w T * #m*=r-

f 4

! FIGURE 8-7 CONTAINMENT PRESSURE 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG l

i 60 , , , ,

50 t

l- 40 -

5 E

g 30 - -

5!

C 2

' 20 - .

10 - _

l l

f i

i Q l l t  !

0 120 240 360 480 600 O

TIME AFTER BREAK, SEC 8-13

l l

FIGURE 8-8 MASS ADDED TO CORE DURING REFLOOD 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 150000 125000 100000.

O 75000 '

g cm 1

50000 Time (Sec) Reflood Rate 0-9.00 2.9017 in/Sec 9.00-54.00 1.2529 in/Sec 25000 -

/ '

54.00-600.0 0.6596 in/Sec 8 8 8 8 8 l 8 a a a a a

~

l S Z 8 8 O TIME AFTER CONTACT, SEC

~

i l

l 8-14 l .. - - _ . , - _ .

PEAK CLAD TEMPERATURE  !

1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG FIGURE 3-9 l

2000

/ '

,,-~-

' s

\

.- [ ' -

1800 / s

/ /

's ss

/

! N 4

1600 N

/ \

i

\ / N . . ..

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s 1400 i

O y 1200' j PEAK CLAD TEWERATlRE NOCE

, ui z

--- PEAK OXIDATIOil NODE i

S

< q 5 1000 '

\

k i #

e a

800 -

\

l l

600 400 70 100 200 300 4.00 500 600 T I t1E , SECONOS 8-15

FIGURE 8-10 1800 HOT SPOT GAP CONDUCTANCE 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 1600

~

1400 1200 1000 u.

Oi

~

5 800

\

g - -

a t~600 b

t/ '

N '

t E

5 b 400 o

I t

200 L

100 200 300 400 500

' 600 700-TIME, SECONDS -

l

\

• 8-16

1 FIGURE 8-11 s 18 LOCAL CLAD OXIDATION 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 16 l

14 -

PEAK CLAD TSPERATURE NCDE

--- PEAK OXIDATION NODE 12 10 - - - - - - -

O e '- -

/

/

/

w 0 /

$ /

1 .

/

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l 2

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100 200 300 400 500 600 7CC i TIME, SECONOS i 8-17

i FIGURE 8-12

~

l 2700 CLAD, CENTERLINE, AVERAGE FUEL AND COOLANT l

, l

. TEMPERATURE FOR HOTTEST NODE 1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG

! 2400 i

/

2100 g s -

1 FUEL CENTERLINE N

• CLAD i

j / AVERAGE FUEL 4

f Q 1500 I

1200 i Y ,

i y J

\"

E 900 i

500 , 3 i

C00tAta I

300 / t t I

l 100 200 300 400 500 500 7C TIME, SECONOS

~8-18 ,

_ FIGURE 8-13 HOT SPOT HEAT TRANSFER COEFFICIENT j 1.0 X DOUBLE ENDED GUILLOTINE BREAK j

IN PUMP DISCHARGE LEG 160.

140 of 120 4

~,

t i

Es y 100

u E 80 E

v 5

i b

y 60 E

a E

40 e

20 i

LA 100 200 300 400 500 600 700 TIME, SECONOS -

8-19

. - - . . - - - . . . - . . - . - . ~ . . . - . , . - .

.~,7 , , , . - - , . . . - . ...

i .

FIGURE 8-14 HOT ROD INTERNAL GAS PRESSURE 4

1.0 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 1400 -

2, f

150'O - P

!NITIAL=

/ 1141.4 PSI A i

1000 - j RUPTURE =

68.8 SEC G 800 O

a.

s N

l $

a.

600 -

i J

400 -

1 I

'i

! 200 -

I i t e i e r 0

0 20 40 60 80 100 I

i TIME, SEC l

l 8-20 1

_ _ _ . . _ . _ . _ _ _ _ . - . . _ _ . . ~ . , . . . . . . _ _ . , , _ . - . . _ . , , , . . . _ - - . , _ . . , _ . . . _ _ , _

1 9.0 Reactor Protection and Monitoring System O Introduction 9.1 The Core Protection. Calculator System (CPCS) is designed to provide the low DNBR and high Local Power Density (LP01 trips to (1) ensure that the specified acceptable fuel design limits on departure from nucleate ,

boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (A00s) and (2) assist the Engineered Safety Features System in limiting the consequences of certain postulated ,

accidents.

The CPCS in conjunction with the remaining Reactor Protection System

(RPS) must be capable of providing protection for certain specified design basis events, _provided that at the initiation of these

! occurrences the Nuclear Steam Supply System, its sub-systems, components i and parameters are maintained within operating limits and Limiting 1

. Conditions for Operation (LCOsh

!O

9.2 CPCS Software Modifications i The CPC/CEAC software for- SONGS Unit 2 and 3 is being modified for

. operation in Cycle 3. This modification is being made by taking the

SONGS Cycle 2 CPC/CEAC software (Reference 9 4) as a . basis since it is  ;

{

. the latest NRC - approved software. The modifications for SONGS Units 2

!' and 3 Cycle 3 relative to the Reference Cycle software include algorithm 1

i changes derived from the implementation of the CPC Improvement Program  !

~

(CIP).. These modifications have been presented in detail in References 9-1 and 9-5 and are summarized in Table 9-1.

1 l

In addition to -the algorithm changes, the CPCS data base and i uncertainties will be updated ' f rom the Reference Cycle. All changes being made to the CPCS will be done in accordance with the NRC-approved software change procedure (References 9-1 and 4-2). >

/

l l

E 9-1

9.3 Addressable Constants Certain CPC constants are addressable so that they can be changed as required during operation. Addressable constants include (1) constants that are measured during startup (e.g., shape annealing matrix, boundary point power correlation coefficients, and adjustments for CEA shadowing and planar radial peaking factors), (2) uncertainty f actors to account for processing and measurement uncertainties in DNBR and LPD calculations (BERRO through BERRA), and (3) miscellaneous items (e.g.,

pre-t ri p and trip setpoints, CEAC inoperable flag, calibration constants, etc.).

9.3.1 Changes to Addressable Constants As a result of the CPCS software modifications discussed in Section 9.2 above, changes have been made to the list of addressable constants.

These changes are listed in Table Q-2 and summarized as follows:

j a. Addressable constants for maximum value of Variable Over Power N Trip (V0PT) setpoint and offset between V0PT setpoint and FOLLOW will be needed.

b. As a result of the simplification of the flow calculations, addressable constant FC2, core coolant mass flow rate calibration constant, will be deleted. The pump speed trip setpoint will be made addressable.

c. An ASGT trip setpoint will he added as an addressable constant.

d. As a result of the CEAC densensitization changes in UPDATE, a CEAC penalty factor time delay will be added as an addressa!.le constant.

e. Combination of the penalty f actor multipliers for DNRR and LPf) into a single multiplier will result in the deletion of addressable constant 0F"LTL.

O U

9-2

l f. As a result of power synthesis algorithm changes in the POWER program, addressable constants ARM 6, ARM 7, EOL, ASM6 and ASM7 will be deleted.

g. - The DNBR trip setpoint will be made an addressable constant. l 9.4 Digital Monitoring System (COLSS) t i

The Core Operating Limit Supervisory System (COLSS) is a monitoring j

system that initiates alarms if the LCO on DNBR, peak linear heat rate, core power, or core azimuthal tilt are exceeded. The COLSS data base and uncertainties will be updated to reflect the Cycle 3 core dasign.  :

I 4

i ~

!O 1

1 l

i i

I O ,

9-3

l l

Table 9-1 O CPC System Software Algorithm Changes for Cycle 3 A. FLOW Program

1. Simplification of flow calculations.*

2. Removal of the DNBR flow projection modules.

B. UPDATE Program

1. Addition of variable overpower trip.*

2. Removal of redundant thermal power compensation filters.

3. Enhancement of ASGT delta-T compensation filter.*

4 Changes for CEAC desensitization.*

i O S. Removal of pressure projection.

I-i 6 Combination of pFMLTD and DFMLTL into a single penalty factor

j multiplier.*

l i

l l

• Require ariditions to or modification of Aridressable Constants.

O 3

9-4

~...-m .._-__.,...m -

i.,-..----_---,- . ~ ~ . - - - _ . . _ . _ _ . . . . - - - - . . . , - . . . . . - . - - _ - - , . . . . _ - _ , . . - . . - . - - - - . - - . - - - . - . . - . - - , . - . - . -

- = . .- - - -. -- -. .;-

1 Table 9-1 (c.ntinued)

CPC System Software Algorithm Changes for Cycle 3 4

C. POWER Program

1. Base low power ASI calculation on actual axial shape.

2. Revise power synthesis calcuations.*

j

3. Removal of flow projection calcualtions and DNRR operating limit.

1 4 Incorporation of an ASI dependent power peaking adjustment.

5. Changes for CEAC desensitization - CEA Withdrawal prohibit (CWP) flag for misoperation.

D. TRIPSEO Program I

(

i 1. Renoval of comparison to flow projected DNBR and pressure

} projacted DNRR.

i 2 Redefinition of J trip'

3. Changes for CEAC desensitization.

4 Addition of DNRR trip setpoint to addressable constants.*

E. CEAC Program j 1. Changes for CEAC desensitization - Set flag to initiate CWP.

t

q

• Require additions to or modification of Addressable Constants.

y/

9-5

1 l

l j Table 9-2 1

CPC System Addressable Constant Changes for Cycle 3 f 1

Point 10 Previous A/C New A/C i 061 FC2 RCP Speed Trip  ;

Setpoint i

l 073 EOL ONRR Trip Setpoint f

Maximum V0PT  ;

079 ARM 6 Setpoint t

080 ARM 7 V0PT Setpoint 1

Offset 091 PFMLTL CEAC PF Tine Delay f 006 ASM6 ASGT AT Trip i t

Setpoint [

t 097 ASM7 ---

t i

t i

~

l i 1

I  !

1 i .

L t

f e.

9-6

.- . .. - - . - . _ - _ . . . . _ . - .=. . _ . . _ - .- . . .

i I

10.0 Technical Specifications t

l This section provides a surrnary of recommended changes that should be made 1

to the SONGS-2 Technical Specifications in order to update the Technical f Specifications for Cycle 3 operation. A description of each change and the l

corresponding technical specification change pages are presented in l Reference 10-1.

4 i

i i,

i l

lO i

1 i

l 1

t i

i I

)

i.

4 i

i

10-1

SONGS UNIT 2 CYCLE 3 PROPOSED TECH SPEC CHANGES ,

CHANGE PACKAGE NUMRER: 1 ,,

SECTION NATURE OF CHANGE FOR CYCLE 3 3.1.2.1 Boric acid concentration range and associated heat tracing reqJirements to change.

Figure 3.1-1 to change.

3.1.2.2 Boric acid concentration range and associated heat tracing requirements to change.

Figure 3.1-1 to change.

3.1.2.7 RWST minimum water volume cf 9970 gals above the ECCS suction connection may change.

3.1.2.8 RWST requirements: 2300 ppm may increase.

3.5.1 SIT requirements: 2300 ppm may increase.

3.5.4 RWST requirements: 2300 ppm may increase.

REASONS FOR CHANGE: Boric acid concentration reduced to allow elimination of heat tracing. RWST and SIT concentration ranges increased for added flexibility.

I i

0 10-2 1

i

l SONGS UNIT 2 CYCLE 3 FRCPCSED TECH SPEC CHANGES CHANGE PACKAGE NUNDER : 2 SECTION NATURE OF CHANGE FOR CYCLE O 5.3.1 Enrichment limit of 3.7 w/o must be raised REASON FCR CHANGE : Longer fuel cycles require higher fuel enrichment.

Cycle 3 will contain 4.05 w/o fuel.

i 10-3 l

h I

t l

l l

l l

1 Ol ,

I I

i l

i le l

4

^

l l 1-l l

l

'l I

I

= 1 1

.I P

. L i

I i

i r

i b

P b

i k

O

j .

i

! SONGS UNIT O CYCLE 3 FROPOSED TECH SPEC CHANGES 4

CHANGE PACKAGE NUMBER : ,

SECTION NATURE OF CHANGE FOR CYCLE O ,

i 3.1.1.3 Neq. MTC limit will get more neg.

1 REASON FOR CHANGE : Negative limit will be close for cycle 3 and is expected to be more negative than current Tech Spec

- value in later cycles. The MIC range is used as an I shalysis ground rule. '

1 f

.I l

1 I

P 5

=

i 1

10-5 i

._. , . , . - . . . . - . - . - . . - - - - - . , , . ~ - . , - - - - , - . - . . - - - . . -- . . . - . . . - - . - . - - - - - - . - - , .. . . - . _ . . - - . - , .

i SONGS UNIT 2 CYCLE 3 FRCPCSED TECH SPEC CHANGES CHANGE PACKAGE NUMBER : 2 i

i SECTION NATURE OF CHANGE FOR CYCLE O

- 5.3.1 Enrichment limit of 3.7 w/o must be raised i

l REASCN FCR CHANGE : Longs. fuel cycles require higher fuel enrichment.

Cycle 3 will contain 4.05 w/o fuel. ,

l 1

i l

e 1

1 i

i i

10-3 I

? .

f SONGS UNIT CYCLE 3 FROPOSCD TECH SPEC CHANGES 4

CHANGE PACKAGE NUMBER : ,

SECTION NATURE OF CHANGE FOR CYCLE O i  ;

I j 3.1.1.3 Neg. MTC limit will get more neg.

REASCN FOR CHANGE : Negative limit will be close for cycle 3 and is expected to be more negative than current Tech Spec

- value in later cycles. The MTC range is used as an ahalysis ground rule.

I 1

1 1

l i

i i

i f

4 4

i i

6 i

e i -  :

i e

10-5

's

- - . . . . , - , , , - . . . - - ..r.- .__,,.,,_-.m_.. , ., -

3. . . , <_ e --, , , , ,, , , , .. .-- .. . , , w,, , .. e,-- . . - . < - - . - . . . . mr -

l

.i 4 .

i SCNGS UNIT 2 CYCLE 3

FROPOSED TECH SPEC CHANGES 1

l CHANGE FACKAGE NUMEER : 5 i

I i SECTION NATURE OF CHANGE FOR CYCLE.3 i

3.3.1 Revise Note in' Table 3.3-2 to change RTD response i

time to 8 sec. Delete Table 0.3-Ca&b.

2 REASCN FOR CHANGE : Safety and CFC analyses will be done using RTD

• response times of 8 sec. Fenalty factors for greater response times will not be verified. Thus i Tables 3.0-Cate will not be supported and must be deleted.

i

+

I i

I

?4 em

}

10-6

. . . ~ . - -.-. . - . . _ - . . . . . - . . . . . . - . - - . . . - - . . - . . . . . _ . . - . - - . . ,

t.

)

i SONGS UNIT 2 CYCLE 3 i

PROPOSED TECH SPEC CHANGES

- CHANGE PACKAGE NUMBER : 6 -

1 f- .

SECTION NATURE OF CHANGE FOR CYCLE ";

i i Tabla 2.2-1 LPD Trip 11mst w/o dynamic terms: 21 kw/ft i

i i, REASON FOR CHANGE : Install generic LPD trip limit, ,

i l -

I ,

I' i  !

l l

J I  !

i t

( e I

! I l

i i i

l l l

i 1

0 0

10-7

. _ . . - - = _ = . .. - -. _-- ._ .- . . . _ _ _ _ _ - . . _ ~ . -

I 1

i SONGS UNIT CYCLE 3 I PROPOSED TECH SPEC CHANGES 1

7 CHANGE PACKAGE NUMBER :

i j SECTICN NATURE OF CHANGE FCR CYCLE 3 1

l 2.2.2 Delete the T/S Section. -

, Table 2.0-2 Delete the Table containing the Addressable Constants.

REASON FOR CHANGE CIP will change, add and/or delete addressable constants 1

I i

i l

f I

k 1

i d

i I

i i

! 10-8

O 11.0 Startup Testing The planned startup test program associated with core performance is outlined below. These tests verify that core performance is consistent with the engineering design and safety analysis. Some of the tests also provide the data needed for adjustment of addressable constants in the. Core i Protection Calculators (CPC's) and in determining constants for the Core Operating Limit Supervisory System (COLSS).

2 11.1 Precritical Test 11.1.1 Contro1' Element Assembly (CEA) Trip Test Precritical CEA drop times are recorded for all 91 CEA's at hot, full I flow conditions hefore criticality following refueling. Acceptance criteria state that the CEA drop time from fully withdrawn to 90%

j inserted shall be less than 3.0 seconds at tha stated conditions.

11.2 Low Power physics Tests 11.2.1 Criticality Criticality is obtained by withdrawing tne Shutdown CEA Groups, diluting to the estimated critical boron concentration, then withdrawing the Regulating CEA Groups to the estimated critical position corresponding to the boron concentration already established.

11.2.2 Critical Roron Concentration Once criticality is achieved, the equilibrium, all CEA's withdrawn boron concentration is obtained. Comparison to the reference critical boron concentration is performed by adding the boron equivalent of the residual CEA worth (from the actual CEA position to the reference CEA position) to the actual boron concentration. Acceptance criteria states that the critical boron concentration shall be within the

equivalent of + 1% AK/K of the design prediction.

11-1

. - - , - , - . .~ . - ._ - ... .- -

l l

Q 11.2.3 Temperature Reactivity Coefficient l (O

The isothermal temperature coefficient is measured at the Essentially

All Rods Out configuration and at a partially rodded configuration.

The average coolant temperature is varied and the reactivity feedback associated with the temperature change is measured. Acceptance criteria state that the measured value shall not differ from the predicted value by more than + 0.3x10"#aK/K/0F.

The moderator temperature coefficient (MTC) of reactivity is calculated by subtracting a predicted value of the fuel temperature coefficient of reactivity. The moderator temperature coefficient (MTC) value is then verified to be within the following Technical Specification criteria:

l

-3.3x10~4 AK/K/ F < MTC < 0.0x10~# AK/K/0 F; Power > 70% Rated Thernal Power

-3.3x10*# AK/K/ 0F < MTC < 0.5x10"# AK/K/0F; Power <

70% Rated Thermal Power 11.2.4 CEA Reactivity Worth CEA worths will be measured using the CEA Exchanga technique. This technique consists of measuring the worth of a " Reference Group" via standard boration/ dilution techniques, then exchanging this group with other groups to measure their worths. Due to the large differences in relative CEA group worths, two reference groups (one with very high worth and one with medium worth) will be used. The groups to be measured by exchange will be " assigned" to a specific reference group, depending on their predicted worth. This measurement technique provides verification that individual group CEA reactivities are within the engineering design safety analysis prediction for all CEA groups. Acceptance criteria state that the measured individual group h

w/

worths shall be within +15% or +0.1% AK/K (whichever is larger of predicted values, and the total worth of all the groups shall be within +10% of the predicted values.

11-2 4

11.3 Powar Ascension Tests

( Following complesion of the Low Power Physics Test sequence, reactor power will be increased in accordance with normal operating procedures. The j l

i power ascension will be monitored by an off-line NSSS performance and data processing computer algorithm. This computer code will be continously executed in parallel with the power ascension to monitor CPC and COLSS performance relative to the processed plant data against which they are normally calibrated. If necessary, the power ascension will be suspended while necessary data reduction and equipment calibrations are performed.

Thus the monitoring algorithm continuously ensures conservative CPC and COLS3 operation while optimizing overall efficiency of the test program.

11.3.1 Reactor Coolant Flow Reactor coolant flow will be measured by calorimetric methods t.t steady state conditions in accordance with Technical Speci ficatio'is.

Acceptance criteria will require that the measured flow be within allowable limits and that both COLSS and the CPC's reactor cool ant flow rates are within calibration requirements relative to the measured calorimetric flow rate.

11.3.2 Core Power Distribution Core power distribution data using fixed incore neutron detectors is used to verify proper core fuel loading and consistency between the as-built and engineering design models. This is 3ccomplished using

! measurement data from three power plateaus.

The first power distribution measurement is performed after the turbine is synchronized. The objective of this measurement is primarily to identi fy any fuel misloading which results in power assymetries or deviations from the reactor physics design. Because of the decreased signal to noise ratio at low powers and the absence of 11-3

xenon stability requirements, radial and azimuthal symetry criteria are emphasized whereas pointwise absolute and statistical acceptance criteria are relaxed.

At the intermediate power plateau (between 40 and 70% reactor power) a core power distribution analysis is performed to again verify proper fuel loading and consistency with design predictions. The intermediate power acceptance criteria ensure that the power distribution is consistent with predictions and that reactor power may be increased to 100% and remain within the design limits.

The final power distributions comparison is performed with equilibrium xenon at approximately 100% power. At this plateau axial and radial power distributions are compared to design predictions as a final verification that the core is operating in a manner consistent with its design within the associated design incertainties.

The measured results are compared.to predicted values in the following manner for the intermediate and full power distribution analysis:

A. The measured radial power distribution is compared to the predicted power distribution utilizing a root mean squared statistical error comparison of the relative radial power density distribution for each of the 217 fuel assemblies. The acceptance criteria states that the comparison of the measured radial power distribution shall satisfy the following exprestion:

~ '

217 1/2 RMS = 2 Zj

< n.05 i=1 I

l . 217 .

l l

where 2 4 is the difference between the predicted and measured relative power density distribution for the i th fuel assembly.

I

8. The measured radial power distribution is additionally compared to the predicted power distribution utilizing a box-by-box comparison of the relative radial power density distribution for each of the 217 fuel assemblies. The acceptance criteria states 11-4

that for each assembly with a predicted relative power density

>=0.9, the measured and predicted relative power density values n

v must agree within + 10%, and for each assembly with a predicted relative power density <0.9, the measured and predicted relative power density values must agree within + 15%.

C. The measured axial power distribution is compared to the predicted power distribution utilizing a root mean squared l

statistical error comparison of the relative axial power

, distribution for each of the 51 axial nodes. The acceptance criteria states that the comparison of the measured axial power

distribution with the predicted axial power distribution shall I satisfy the following expression

51 1/2 RMS =

) h 9

2

< 0.05.

51 1

where h j is the difference between the predicted and measured relative power density distribution for the ith axial % of core height.

D. The measured values of total planar radial peaking factor 1

(Fxy), total integrated radial factor (F e,), core average ,

axial peak (Fz ), and 3-D power peak (Fq ) are compared to predicted values. The acceptance criteria states that the measured values of F xy, F, F, g and F q shall be within e

~

+ 10% of the predicted values.

1 11.3.3 Shape Annealing Matrix (SAM) and Roundary Point Power Correlation Coefficients (RPPCC) Verification i

p The SAM matrix and BPPC coefficients are determined from a linear V regression analysis of the measured excore detector readings and corresponding core power distribution determined from the incore detector signals. Since these values must be representative for a 11-5

= _ _ . _ _ _ _ _ . _ _ _ . _ _ - .

J 8

L

j.  !

rodded and unrodded core throughout life, it is desirable to use as l wide a range of core axial power shapes as are available to establish their values. The spectrum of axial shapes encountered during the ,

power ascension has been demonstrated to be adequate for the calculation of the matrix elements. Incore, excore, and related data l

} are recorded and incore analysis is ' performed which relates the incore I detector signals to power distribution. and sununarizes the necessary j power distribution and excore detector data in a form and format which j l can be easily input to programs used to perform the least squares  !

fitting. The data is processed and compiled throughout . the power ascension by the off-line NSSS performance and data processing [

). -

algorithm i I

)

j The analysis results include: ,

I t j A. Core peripheral power fractions for the upper, middle, and lower i third of the core for each quadrant; B. Core average power fractions for the upper, middle, and lower 4

third of the core; and i i C. Upper and lower core boundary average power.

l Appropriate CPC constants are modified, if needed, based upon the .

f measured values.  ;

} .I j 11.3.4 Radial Peaking Factor and CEA Shadowing Factor Verification i

! i

{ T5e performance of this test involves establishing the following CEA j configurations:

i j All CEA's Out t

1 i

1 Group 6 at LEL (Lower Electrical Limit) 3 lO Group 6 at LEL, Group P at 37.5 inches withdrawn- ,

i i

j Group P at 37.5 inches withdrawn 11-6

I As the various CEA configurations are established, incore detector data and excore detector data are taken after allowance of sufficient time for stabilization of the incore instrument signals. This data is analyzed and planar radial peaking factors (Fxy) and CEA shadowing factors are determined for each CEA configuration. Appropriate CPC ,

and/or COLSS constants are nodified, if needed, based on the measured l

1 values.

i I 11.3.5 Reactivity Coefficients 1001. Full Power (1) Isothermal Temperature Coefficient - With the reactor at steady state and near equilibrium Xenon, CEA's are moved a specified

! amount. This reactivity change produces a change in reactor power which in turn causes a change in coolant temperature. The change in coolant temperature results in a reactivity feedback to counter the rod movement if the ITC is negative. The system

! eventually stabilizes at a new coolant temperature. Core power f is kept essentially constant by adjustments made to turbine ,

loading. ITC is calculated knowing the power and temperature changes along with the CEA integral worth and by using the i prediction for the Power Coef ficient. The MTC is calculated as

} described previously.

1 (2) Ooppler Power Coefficient - Reactivity changes are made using l

j CEA's, resulting in a change in reactor power. Average coolant temperature is held constant by changing turbine load. The

! reactor stabilizes at a new power when the reactivity feedback due to change in power is equal and opposite to the CEA

reactivity insertion. The Doppler power coefficient is ;

i calculated in a manner similar to the ITC calculation.  !

Acceptance Criteria state the following:  ;

a. The measured ITC shall agree with the predicted values j

within +0.3x10*#4K/K/0F;

(

b. The measured power coefficient should agree with the j predicted values within +0.3x10*#tK/K/t power; and 11-7

! i

c. The HTC shall satisfy the following criteria:

-3.3x10*#aK/K/0F < MTC < 0.0x10"#aK/K/0F;

[ Power > 70% Rated Thermal Power

-3.3x10 ' AK/K/0F < MTC < 0.5x10** a K/K/0F; Power < 70% Rated Thermal Power i

l 11.4 procedure If Acceptance Criteria Are Not Met t

! If the acceptance criteria for any test are not met, an evaluation is j performed before the test program is continued. The results of all I tests will be reviewed by the plant's core analysis engineering group. If the acceptance criteria of the startup physics tests are j i

not Mt, an evaluation will be performed by the plant's core analysis

]

j engineering group with assistance from the fuel vendor, as needed. ,

j The results of this evaluation will be presented to the Onsite Review Comittee. Resolution will be required prior to power escalation. If

=

an unreviewed safety question is involved, the NRC wnuld he notified.

1 l

i 1

f i

I

$ I t

)

i

!n 4 v 1

11-8

. , _ . _ _ _ _ ,. _ _ _ _ _ - _ _ - ~ ,

i i

i O

id 12.0 References i

4 12.1 Section 1.0 References (1-1) " San Onofre Nuclear Generating Station Units 2 and 3 Final Safety I l Analysis Report," Southern California Edison Co., Docket Nos. 50-1 361 and 50-362. L 1

(1-2) Letter from M. O Medford (SCE) to G. W. Knighton (NRC), " Reload i Analysis Report, San Onofre Nuclear Generating Station Units 2 and

] 3," September 2R, IQAA.

i (1-3) Presentation to NRC on May 31, 19R5 i

l 12.2 Section 2.0 References l None I

12.3 Section 3.0 References i

1 j None 1 i i

. 12.4 Section 4.0 Deferences 4

l (4-1) letter f rom M. O. Medford (SCE) to G. W. Knighton (NRC), " Reload j Analysis Report, San Onofre Nuclear Generating Station Units 2 and 3," September 29, 1084 (4-2) LO AA n54, "CEA Guide Tuh* Wear Sleeve Modification," August 3, i 19A4 (4-3) NRC Letter, C. G. Thomas to A. E. Scherer. "Acceptanca for i

Referencing of Licensing Special Report LO-R4-041, CEA Guide Tube i

I Wear Sleeve '4odificatinn," September 7,184 i

(4-4) CENPD-187, "CEPAN Method of Analyzing Oval Cladding," June '1975

{

I 12-1 t

(4-5) CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," July, i 1974 i

! (4-61 San Onofre Nuclear Generating Station Units 2 and 3 Final Safety  ;

i Analysis Report." Southern California Edison Co., Docket Nos. 50-361 and 50-362  ;

t  !

(4-7) CEN-182(8)-P, " Statistical Approach to Analyzing Creep Collapse of ,

]

Oval Fuel Rod Cladding using CEPaN," September 1981.

I j (4 8) CEN-161(B)-P, "Irnprovement to Fuel Evaluation Model," July 1981.

i  !

]

(4 41 Letter, R. A. Clark (NRC) to A. E. Lundvall, Jr. BGAE), " Safety J Evaluation of CEN-161 (FATES 3)," March 31, 1983.

i ,

i  !

j (4-10) CENPD-264 P Revision 1 P, " Extended Rurnup Operation of Combustion 1

4 Engineering PWR Fuel," July 19R4

- i (4-11) E. J. Butcher (NRC) to A. E. Lunduall (BG4E), " Safety Evaluation i of Topical Report CENPD-264-P " To be issued.

! I (4-12) CEN-QA-( A).P Rev. 1. "ANO-2 Reactor Operation with Modified CEA 1 Guide Tubes and Lengthened Upper Guide Structure Flow Channels,"

I July 12, 147R.

i I 12.5 Section 5.0 References ,

(5-1) Presentation to NRC on May 11, 14RS.-

i' (5-2) "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Voluna

5

Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," EPRI NP-3966-CCM, Volume 5. Project 2061-4, .

j Computer Code Manual, April 14RS.

i  ;

i (5-1) A. E. Lundvall (RG4E) to J. R. Miller (NRC), "Calvert Cliffs Ip Nuclear Power Plant Unit 11'00cket No. 50-317 Eighth Cycle License  !

!V Application," February 22, 1985.

i 12-2

(5 4) " Safety Evaluation of the Office of Nuclear Reactor Regulation h Related to Amendment No. 104 to Facility Operating License No. OPR-53 Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant Unit No. 1, Docket No. 50-317."

(5-5) CENPD-153-P, Rev. 1-P A, " INCA /CECOR Power Peaking Uncertainty,"

May 19A0.

(5-6) CENPD-266-P A, "The ROCS and DIT Computer Codes for Nuclear Design," April 1993 12.6 9ection 6.0 Referencas (6-1) CENPD-161 P, " TORC Code, A Computer Code for Determining the Themal Margin of a Reactor Core " July 1975.

(6-2) CENPD-162-P-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1. Unifom Axial Power Distribution," April 1975.

(6-3) CENOP-706-P, " TORC Code. Verification and Simpli fied Modeling Methods," January 1977 (4 A) CEN-160(S)-P, Rev.1-P "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Units 2 and 1," September 19%.

(6-5) CEN-283($1-P, " Statistical Combination of Uncertainties Part I:

Combination of System Parameter Uncertainties in Themal Margin Analyses for San Onofre Nuclear Generating Stations Units 2 and 3," June 19A4 (A 6) CEN-155-(S).P. "CE-1 Applicability to San Onofre Units 2 and 3 HID-2 Grids, Response to NRC Ouestions," March 19R1.

(6-7) CEN-165(S) P " Response to NRC Concerns onD A plicability of the CE.

1 Correlation to the SONGS Fual Design," March 19R1.

12-3

c (6-8) NUREG-0712. Supplement 4. " Safety Evaluation Report Related to the Operation of San Onofre Nuclear Generating Station Units 2 and 3,"

i Docket Nos. 50-361 and 50-362, January 1982.

1 (6-9) CENPD-255-P-A, " Fuel and Poison Rod Bowing," June 1983.

! (6-10) Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 24 to Facility Operating License No. NPF-6, Arkansas Power and Light Company, Arkansas Nuclear One, Unit No. 2, Docket No. 50-368.

12.7 Section 7.0 References 1

(7-1) Letter from M. O. Medford (SCE) to G. W. Knighton (NRC), " Reload Analysis Report San Onofre Nuclear Generating Station, Units 2 and 3." September 28, 19R4

, (7-2) " San Onofre Nuclear Generation Station Units 2 and 3. Final Safety

( Analysis Report".

(7-3) "STRIKIN 11, A Cylindrical Geonetry Fuel Rod Heat Transfer

! Progran," CENPO-135-P, August 1984 1

(7-4) " Loss of Flow - C-E Hethods for loss of Flow Analysis," CENPD-183, July, 1975.

I (7-5) Macbeth, R. V., "An Appraisal of Forced Convection Burnout Data "

Proc. Instn. Mech. Enges., Vol. IRO, Pt3C, PP 37-50, 1965-66.

(7-6) Lee, D. H., "An Experimental Investigation of Forced Convection Rurnout in High Pressure Water - Part IV, Large Diameter Tubes at about 1600 psia," A.E.E.W. Report R 479, 1966 (7-7) "CETOP-0 Code Structure and Modeling Methods for Calvert C1tffs 1 J and 2," CEN-191(R1-P, Decenher 19A1.

O v (7 8) CPC/CEAC Sof tware Modifications for Arkansas Unit 2, CEN-141 A-P, Appendix A, December 19RO.

12-4

(7-9) Not used.

O (7-10) "CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981 Enclosure 1-P to LO-82-001, January 6,1982.

(7-11) CENPD-1RA-A, "HERMITE Space-Time Kinetics " July 1976 ,

(7-12) CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin for a Reactor Core," July 1975.

(7-13) Calvert Cliffs Nuclear Power Plant Unit 1 Docket No. 50-317

" Amendment to Operating License DPR 53 Supplement 1 to Seventh Cycle License Application," September 1, 1983.

(7-14) CENPD-161-P, " TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core," July 1975.

(7-15) CENPD-206-P, " TORC Code -

Verification and Simplified Modeling Methods," January 1977 (7-16) Presentation to NRC on May 31, 1985.

(7-17) CEN-308-P Revision 00-P, "CPC/CEAC Sof tware Modifications for the CPC Improvement Program." July 19R5.

12.8 Section R.0 References .

1 (8-1) Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3. Friday, January 4,1974 i (8-2) Letter from M. O. Medford ($CE) to G. W. Knighton (NRC), " Reload Analysis Report San Onofre Nuclear Generating Station, Units 2 and 3." September 28, 1984 O

I o '

12-5 ,

i

l 4

" Calculative Methods for the C-E Large Break LOCA O

(8-3) CENPD-132 Evaluation Model," August 1974 CENPD-132, Supplement 1, " Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model," December 1974 i

(8-4) CENPD-132, Supplement 2. " Calculational Methods for the C-E Large I Break LOCA Evaluation Model," July 1975.

2 (8-5) CENPD-133 "CEFLASH 4A, A FORTRAN IV Digital Computer Program for

' Reactor Blowdown Analysis," April 1974 CENPD-133, Supplement 2 "CEFLASH 4A, A FORTRAN IV Digital l

Computer Program for Reactor Blowdown Analysis (Modification),"

1 December 1974

)

(R-6) CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Cora," April 1974 iO CENPD-134, Supplement 1, "COMPERC-!!, A Program for Emergency i

Refill-Reflood of the Core (Modification)," December 1974

! (R-7) CENPD-135 "STRIKIN, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1974 I CENPD-135, Supplement 2 P, "$TalKIN-!!, A Cylindrical Geometry 4

Fuel Rod Heat Transfer Program (Modification)," February 1975.

CENPD-135-P, Supplement 4P, "STRIKIN !!, A Cylindrical Geometry fl t

Fuel Rod Heat Transfer Program " August 1976

] (A-8) CENPD-13R, and Supp1* ment 1. "DARCH, A FORTRAN-IV Digital Progran j to Evaluate pool Rolling, Axial Rod and Coolant Heatup," February 1975, CENPD-13A Supplement 2 (P), " PARCH - A FORTRAN IV Digital progran to Evaluata Pool Rolling, Asial Rod and Coolant Heatup," January, lo77.

12-fi

1 1

(8-9) CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report,"

]

J July, 1984 (8-10) CEMPD-161(B)-P, " Improvements to Fuel Evaluation Model Topfcal

- Report," July 1981.

i

}

i (8-11) letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (RG4E), dated i March 31, 1983.

12.9 Section o.0 References f

l (9-1) CEN-39( A)-P, Revision 07, "The CPC Protection Algorithm Software Change Procedure," Oecember 21, 1978.

f i

(9-2) CEN-30fA)-P, supplement 1-P, Revision 01, January 5, 1479.

I (9-3) CEN-30A-P Revision 00 0, "CPC/CEAC Software Modifications for the CPC Improvement Program." July 1985.

i j (9-!) Safety Evaluation Related to Amendment No. 32 to NPF-10 and

! Amendment No. 21 to NAF-15 for San Onofre Nuclear Generating i Station, Units 2 and 3, Docket Nos. 50-361 and 50-362, Southarn i

California Edison Company, March 19R5.

l (4-5) CEN-310 D, "CPC Nethodology Changes for the CPC Inorovement Program " September 14AS.

l 1

l 12.10 Section 10.0 References (10 1) To be issued.

i j 12.11 Section 11.0 References i None n . ,

12-7

%