ML13330A897

Reload Safety Evaluation,San Onofre Nuclear Generating Station Unit 1,Cycle 8,Revision 2

ML13330A897
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Site:	San Onofre
Issue date:	04/30/1981
From:	Arlotti M, Skaritka J WESTERN NUCLEAR, INC., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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RELOAD SAFETY EVALUATION SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1, CYCLE 8 REVISION 2 April, 1981 Edited by J. Skaritka Approved:_

M. G.(Arlotti, Manager Fuel Licensing and Coordination Nuclear Fuel Division REQULATORY DOCKET FILE COPY

TABLE OF CONTENTS Title Page

1.0 INTRODUCTION

1 2.0 REACTOR DESIGN 3

2.1 Mechanical Design 3

2.2 Nuclear Design 3

2.3 Thermal and Hydraulic Design 4

3.0 ACCIDENT EVALUATION 6

3.1 Power Capability 6

3.2 Accident Evaluation 6

3.3 Incidents Reanalyzed 8

4.0 TECHNICAL SPECIFICATIONS 10

5.0 REFERENCES

12 APPENDIX A LOCA Analysis for 15 Percent of A-1 Steam Generator Tubes Plugged 3849A

LIST OF TABLES Table Title Page 1

Fuel Assembly Design Parameters 14 2

Core Physics Parameters 15 3

Shutdown Requirements and Margins 16 4

Safety Evaluation Parameters 17 5

Rod Ejection Parameters 18 LIST OF FIGURES Figure Title Page 1

Core Loading Pattern 19 2

F Total Versus Axial Offset 20 3

Technical Specification Figure 2.1 21 Safety Limits:

Temperature, Power, Pressure RCS Flow -

201,900 GPM 3849A

1.0 INTRODUCTION

AND

SUMMARY

The San Onofre Nuclear Generating Station Unit 1 is shutdown for Cycle 7/8 refueling and repairs of steam generator tubing. Cycle 8 startup is estimated for May 1981.

This report presents an evaluation for Cycle 8 operation which demon strates that the core reload will not adversely affect the safety of the plant. It is not the purpose of this report to present a reanalysis of all potential incidents. Those incidents analyzed and reported in the FSA(1) which could potentially be affected by fuel reload have been reviewed for Cycle 8 design described herein. The results of new analy ses have been included, and the justification for the applicability of previous results from the remaining analyses is presented. These analy ses.assume that: (1) Cycle 7 operation is terminated between 10030 and 11030 MWD/MTU, (2) Cycle 8 burnup is limited to the end-of-full power capability*, and (3) there is adherence to plant operating limitations given in the technical specifications, and (4) the proposed technical specification changes in Section 4 are implemented.

The San Onofre 1, Cycle 8 core loading pattern is shown in Figure 1.

The one Region 6 and 51 Region 7 fuel assemblies from Cycle 7 have been removed and replaced by 52 Region 10 fuel assemblies. A Region 7 fuel assembly will be reused in the central core position.

Nominal design parameters for Cycle 8 are 1347 Mwt (100 percent rated core power), 2100 psia system pressure, nominal core inlet temperature of 528 0F, 4.64 kw/ft average linear fuel power density, and a 201,900 gpm RCS Thermal Design Flow (96.7 percent of Cycle 7 TDF).

The Cycle 8 reduced TDF accounts for up to an equivalent 15 percent steam.generator tube plugging.

• Definition:

Cycle 8 design power and temperature (approximately 551oF Tavg), control rods fully withdrawn, and zero ppm of residual boron.

3849A This report replaces an October 1980 Cycle 8 RSE(2 ) which evaluated 100 percent rated power with a reduced TDF (93.4 percent of Cycle 7) to account for an equivalent 20 percent steam generator tube plugging.

Changes from the October 1980 RSE are noted by bars in the margins.

3849A '

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 10 fuel assemblies is the same as the Region 9 assemblies. Table 1 compares pertinent design parameters of the various.fuel regions. The Region 10 fuel has been designed according to the fuel performance model in Reference 3.

Clad flattening will not occur during Cycle 8. All fuel regions have a predicted clad flattening time equal to or greater than 50,000 EFPH. No fuel region will receive this exposure.

2.2 NUCLEAR DESIGN T

Cycle 8 core loading statisfies an ECCS analysis limit of F x P

<2.89* as shown in Figure 2. The limitations on FT of 2.89 include the effects of the local power peaking of Figure 3.1 in WCAP 8131(4) to assure that the allowable value for LOCA is satisfied. The points plotted on Figure 2 include maneuvers typically done at San Onofre Unit 1 and variants on these maneuvers done at a number of control rod insertions, times and burnups.

The limiting FT has been determined for the combination of the most adverse FXY and the most adverse FN that will be experienced during operation in Cycle 8. The most adverse Fxy occurs at beginning of life and the most adverse FN occurs at end of life. The results shown for F in Figure 2 include uncertainty factors of 15 percent Q

for conservatism and 4 percent for manufacturing tolerances.

The xenon transient analysis has been evaluated similarly to analyses of previous cycles. The most limiting FN, including an uncertainty of N

Z 10 percent on Fz is 1.87 at 84 percent of core height. With the Cycle 8 FN o1.5anN l AH of 1.55, an FZ of 1.96 at 84 percent of core height would be required to reach a DNBR of 1.30 at this elevation and 118

• The new FQT x P - 2.89 limit results from LOCA analysis for 15 percent equivalent steam generator tubes plugged, as shown in Appendix A.

3849A percent power. This margin exists assuming a control rod withdrawal occurs with the rods moving to the fully withdrawn position.

Table 2 provides a comparison of Cycle 8 kinetics characteristics with the current limit based on previously submitted accident analysis. The effect of the Table 2 parameters, including those that fall outside the current limits, are evaluated in Section 3. Table 3 provides the end of-life control rod worths and requirements at the most limiting condi tion during the cycle. The required shutdown margin is based on a previously submitted accident analysis.(5) The available shutdown margin exceeds the minimum required to meet the accident analysis.

2.3 THERMAL AND HYDRAULIC DESIGN A reduction in thermal margin for Cycle 8 will result due to a 3.3 per cent reduction in thermal design flow. New DNB Core Limits (See Section

.4) are generated using the Cycle 5 through 7 design axial power shape with a DNB design FN of 2.07 at 85 percent core elevation. The zN difference between the DNB design 2.07 FN and the 1.92 limiting FN at 85 percent core elevation results in a DNBR margin of 5.5 z

percent. Reference 7 identified this margin as available to offset rod bow DNBR penalties. However, since the stainless steel clad fuel has less than a 50 percent gap closure, there is no rod bow DNBR penalty(6 ), and the 5.5 percent margin is available for other uses.

The above design 2.07 FN is too restrictive for DNB limiting acci z

dent analyses. Therefore, a DNB design axial power shape with an FN of 1.95 at 85 percent core elevation is used for DNB limiting accident analyses. This 1.95 value provides a 1.1 percent DNBR margin when compared to the 1.92 limiting FN.

This margin is part of the Cycle 8 total DNB margin of 4.4 percent shown below, which is a reduc tion from the previous 8.8 percent margin defined in Reference 7.

-4 3849A

DNBR margins available for Cycle 8:

Pitch reduction 3.3 percent Adverse design axial power shape 1.1 percent Total DNBR margin 4.4 percent For all DNBR analyses, the local power spike due to fuel densification is not included, as justified in Reference 8.

-5 3849A

3.0 ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability is evaluated considering the consequences of those incidents examined in the FSA,(1) using the previously accepted design basis. It is concluded that the core reload and operating constraints due to steam generator tube repairs will not adversely affect the ability to safely operate at 100 percent of rated power and 201,900 gpm RCS Thermal Design Flow (96.7 percent Cycle 7 TDF) during Cycle 8. For Condition II overpower transients, the fuel centerline temperature limit of 4700 F can be accommodated with margin in the Cycle 8 core. The time dependent densification model(8) was used for fuel temperature evaluations. The LOCA limit for the Cycle 8 operating conditions is satisfied by maintaining FT at or below 2.89.*

This Q

limit is satisfied by the power control maneuvers allowed by the tech nical specifications, which assure that the Interim Acceptance Criteria (IAC) limits are met for a spectrum of small and large breaks.

3.2 ACCIDENT EVALUATION The effects of the reload and the reduced TDF on the design basis and postulated incidents analyzed in the FSA(1) are evaluated in this section and Section 3.3. Most of the non-LOCA incidents are accommod ated within the conservatism of the initial assumptions used in the previous applicable safety analysis. For those incidents which were reanalyzed (Section 3.3), it was determined that the applicable design bases are not exceeded, and, therefore, the conclusions presented in the FSA are still valid.

The new F limit of 2.89 is the result of the Cycle 8 reduction in TDF (96.7 percent Cycle 7 TDF) which accounts for up to 15 percent equivalent steam generator tube plugging. Section 3.3 presents the LOCA analysis which establishes the 2.89 FQ limit. Cycle 7 had a 2.95 FQ limit.

-6 3849A

A core reload can typically affect accident input parameters in the following areas:

core kinetic characteristics, control rod worths, and core peaking factors. Cycle 8 parameters in each of these three areas were examined as discussed below to ascertain whether new accident ana lyses were required.

A comparison of Cycle 8 core physics parameters with current limits is given in Table 2. The kinetic values fall within the bounds of the current limits.

Changes in control rod worths may affect differential rod worths, shut down margin, ejected rod worths, and trip reactivity. Tables 2 and 3 show that the maximum reactivity withdrawal rate, and the shutdown margin with the worst stuck RCCA are within the current limits. The ejected rod worths and trip reactivity curve are within the bounds of the previous Cycle 7 evaluation.

Peaking-factor evaluations were performed for the rod out of position, dropped RCCA bank and dropped RCCA, to ensure that the minimum DNB ratio remains above 1.30. These evaluations were performed utilizing Cycle 8 transient statepoint information and peaking factors. In each case, it was found that the peaking factor for Cycle 8 was lower than the value for which DNBR equals 1.30. The peaking factors following control rod ejection are within the limits of previous analysis for the EOL zero power and full power cases. Peaking factors for the Cycle 8 BOL zero and full power incidents exceed previously analyzed values, and these cases are reanalyzed in Section 3.3.

The repairs to be steam generator tubes (plugging/sleeving) results in three effects:

reactor coolant flow is reduced due to increased steam generator flow resistance, steam generator heat transfer is reduced, and primary reactor coolant mass.is reduced. The impact of these changes were examined to determine whether new accident analyses were required.

-7 3849A

The parameters listed in Table 4 were used for the safety evaluation.

In general, all of the transients are sensitive to steady-state primary coolant flow. The impact of the reactor coolant flow reduction was assessed for each FSA Accident(1 ). Based on this assessment, the following accidents most affected by reduced TDF were reanalyzed in Section 3.3:

control rod ejection, uncontrolled RCCA bank withdrawal at power, loss of reactor coolant flow, steamline break and loss of reactor coolant.

3.3 INCIDENTS REANALYZED The control rod ejection analysis is affected adversely by increased peaking factors following rod ejection for the beginning-of-life cases and by the primary flow reduction for all beginning-of-life and end-of life cases. The four cases shown in Table 5 were reanalyzed, and the hot spot fuel rod does not exceed the limiting fuel criteria(') for any case.

Uncontrolled rod withdrawal at power was reanalyzed due to a change in reactor core limits caused by the reduction in reactor coolant flow.

Accordingly, the variable low pressure reactor trip setpoint equation has been recalculated. The spectrum of reactivity insertion rates was examined. The adequacy of the revised setpoint equation was verified; the minimum DNB ratio remained above the 1.30 limit for all cases, and the FSA conclusions remain valid.

The loss of flow accident was reanalyzed to determine the effect of reduced initial reactor coolant flow and increased loop resistance due to steam generator tube repairs (plugging and sleeving).

For the most severe loss of flow transient the minimum DNB ratio remained above the limit of 1.30, and the conclusions given in the FSA remain valid.

-8 3849A

The steamline rupture accident was reanalyzed using conservative Cycle 8 physics parameters, and conservative feedwater assumptions. Results for the failed dump valve confirm that the core remained subcritical throughout the transient; results for the hypothetical break cases con firm that the DNB ratio remains greater than the 1.30 limit, and the conclusions given in the FSA are still valid.

A spectrum of the LOCA DECLG breaks with discharge coefficients of 1.0, 0.8 and 0.6 were analyzed for the reactor maximum permissible power level of 1347 Mwt and 201,900 GPM TDF. The analyses assumed a uniform steam generator tube plugging level of 11 percent plus an additional resistance equivalent to 4 percent to account for tube sleeving. The most limiting fuel parameters for Cycle 8 were assumed. The analyses results showed the 0.8 discharge coefficient DECLG break was limiting with a PCT of 22720F at a design peaking factor of 2.89. This limit ing result satisfies the 1971 AEC Interim Acceptance Criteria (IAC) which has a 2300oF PCT as a limit. Appendix A presents additional supporting information for the worst-case LOCA analysis break (DECLG of 0.8).

-9 3849A

4.0 TECHNICAL SPECIFICATIONS This section contains the technical content of proposed changes to the San Onofre Plant Technical Specifications. These changes are consistent with the plant operation necessary for the design and safety evaluation conclusions stated previously to remain valid.

Due to the reduced thermal design flow, a number of changes to the present technical specifications are required. These changes are sum marized below:

Section 2.1, Item (2);

Replace:

The combination of reactor system pressure and coolant temperature shall not exceed the locus of points established for the power level in Figures 2.1.1. If the actual pressure and temperature is above or to the left of the locus of points for the appropriate power level, the safety.

limit is exceeded.

With:

The combination of reactor power and coolant temperature shall not exceed the locus of points established for the RCS pressure in Figure 2.1.1. If the actual power and tempera ture is above the locus of points for the appropriate RCS pressure, the safety limit is exceeded.

Figure 2.1.1 - Replace with Figure 3 in this report.

Section 2.1, Table 2.1 Item 4 Variable Low Pressure Replace > 14.45 (1.3130 AT + T

) -

7298.7 With >26.15 (0.894 AT + Tavg) 4341 Section 3.5.2:

Control Group Insertion Limits - Basis (Item 1, 2nd para.)

-10 3849A

Replace:

A more restrictive limit on the design maximum value of specific power, F H and F is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F NH and F are 13.97 kW/ft., 1.55

'A Q

and 2.95, respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.

With:

A more restrictive limit on the design maximum value of specific power, F NH and F is applied to operation

'A Q

in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F N and F are 13.7 kW/ft., 1.55 AH Q

and 2.89 respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.

Section 3.11A., Incore Axial Offset Limits; For positive offsets:

Replace IAO = 2.95/P - 2.1225 -3.0 0.03084 With TAO = 2.89/P - 2.1225 -3.0 0.03021 For negative offsets:

R 2.95/P - 2.1181 Replace IAO =

-.03132

+3.0 With IAO = 2.89/P - 2.1181 +3.0

-0.03068

-11 3849A

5.0 REFERENCES

1. Docket Number 50-206, "San Onofre Nuclear Generating Station, Unit 1, Part 2, Final Safety Analysis".

2. Skaritka J., Editor, "Reload Safety Evaluation - San Onofre Unit 1, Cycle 8 -

Revision 1," October 1980.

3. Miller, J. V. (Ed.), "Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations", WCAP-8785, October 1976.

4. "Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station Unit 1, Cycle 4", WCAP-8131, May 1973.

5. "SCE Report, "Steamline Break Accident Reanalysis, San Onofre Nuclear Generating Station Units 1, October 1976", Attachment to letter, K. P. Baskin to K. R. Goller, December 29, 1976.

6. Letters, J. F. Stolz (NRC) to T. M. Anderson (Westinghouse);

Subject:

staff Review of WCAP-8691; April 5, 1979.

7. Letter from K. P. Baskin (SCE) to K. R. Goller (NRC);

Subject:

Rod Bow Margin, San Onofre Unit 1; Docket No. 50-206; February 10, 1977.

8. Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8218-P-A, March 1975 (Proprie tary) and WCAP-8219-A, March 1975 (Non-Proprietary).

9. Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.

-12 3849A

10. Skaritka, J., Editor, "Reload Safety Evaluation -

San Onofre Unit 1, Cycle 7", August 1978.

11. "Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 5", Attachment to letter from Jack B. Moore to Edson G. Case, March 7, 1975.

-13 3849A

TABLE 1 SAN ONOFRE UNIT 1 -

CYCLE 8 Fuel Assembly Design Parameters Region 7

8 9

10 Enrichment (w/o U-235)*

4.00 3.99 3.98 4.0 Density (percent Theoretical)*

94.65 94.59 94.66 95.0 Number of Assemblies 1

52 52 52 Approximate Burnup at start of 29050 20800 8750 0

Cycle 8 (Mwd/Mtu)

• All fuel region except Region 10 are as-built values. Region 10 values are nominal.

-14 3849A

TABLE 2 SAN ONOFRE UNIT 1 - CYCLE 8 Core Physics Parameters Current Limit Cycle 8 Moderator Temperature

-4.0 to 0(1)

-3.3 to -0.3 Coefficient, (Ap/oF) x 10 Doppler Coefficient,

-2.75 to -1.4(10)

-2.6 to -1.4 (Ap/oF) x 105 Delayed Neutron Fraction, 0.50 to 0.70(l) 0.55 to 0.62

$eff' Maximum Prompt Neutron 26(11) 11.4 Lifetime (P sec)

Maximum Reactivity Withdrawal 40(10)

<40 Rate, (pcm/sec)*

• pcm 5 AP 3849A TABLE 3 SAN-ONOFRE UNIT 1 - CYCLES 7 and 8 Shutdown Requirements and Margins Cycle 7 Cycle 8 BOL EOL BOL EOL Control Rod Worth (% Ap)

All Rods Inserted 6.7 7.3 6.8 7.4 All Rods Inserted Less Worst 5.5 6.2 5.8 6.4 Stuck Rod (1) Less 10%

4.9 5.6 5.2 5.8 Control Rod Requirements (% Ap)

Reactivity Defects (Doppler, Tavg, 1.9 2.6 1.8 2.6 Void, Redistribution Rod Insertion Allowance 0.8 0.8 0.9 0.9 (2) Total Requirements 2.7 3.4 2.7 3.5 Shutdown Margin (1)-(2) (% Ap) 2.2 2.2 2.5 2.3 Required Shutdown Margin (% Ap) 1.25 1.9(5) 1.25 1.9

-16 3849A

TABLE 4 SAN ONOFRE UNIT 1 STEAM GENERATOR TUBE REPAIR (PLUGGING AND SLEEVING)

SAFETY EVALUATION PARAMETERS Maximum core thermal power, MWt 1347 Thermal design flow, gpm/loop 67,300 S.G. effective tube plugging level, percent 15 T

at 100 percent power, oF 551.5 avg AT at 100 percent power, oF 47 FHN 1.55 AH Tnoload0 F 535 RCS pressure, psia 2100 3849A TABLE 5 SAN ONOFRE UNIT 1 -

CYCLE 8 Rod Ejection Parameters Previous Value*

Analysis Used In values(9 )

Reanalysis HZP -

BOC Max. Ejected Rod Worth, % Ap 0.68 0.68 Max. F 8.47 8.95 Oeff 0.0055 0.0055 HFP -

BOC Max. Ejected Rod Worth, % Ap 0.21 0.21 Max. F 5.48 6.11 Seff 0.0055 0.0055 HZP -

EOC Max. Ejected Rod Worth, % Ap 0.58 0.58 Max. F 9.17 9.17 Seff 0.0050 0.0050 HFP - EOC Max. Ejected Rod Worth, % &p 0.15 0.15 Max. F 7.13 7.13 Seff 0.0050 0.0050 HZP -

Hot Zero Power BOC - Beginning of Cycle HFP - Hot Full Power EOC -

End of Cycle

• These values bound the Cycle 8 values.

-18 3849A

FIGURE 1 CORE LOADING PATTERN SAN ONOFRE UNIT 1 CYCLE 8 1 2 3

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APPENDIX A LOCA ANALYSIS FOR 15 PERCENT OF STEAM GENERATOR TUBES PLUGGED A-1 3849A

APPENDIX A TABLE OF CONTENTS Title Page A.1 LOCA Summary A-3 A.2 RCS Blowdown Calculation A-3 A.3 Lower Plenum Refill/Core Reflood Calculations A-4 A.4 Hot Rod Thermal Transient Calculation A-4 A.5 Results and Conclusions A-4 LIST OF FIGURES Figure Title Page A.1 LOCA Transient Values:

Quality, Rod Film Coefficient, Core Flowrate A-6 A.2 LOCA Transient Values:

Core AP Core Pressure, Break Flowrate A-7 A.3 LOCA Transient Values:

Refill Water Inlet Velocity, Film Coefficient, Core Flow Rate A-8 A.4 LOCA Transient Values:

Peak Clad Temperature, Hot Spot Fluid Temperature A-9 A-2

A.1 LOCA

SUMMARY

The Emergency Core Cooling System (ECCS) performance following a Loss of Coolant Accident (LOCA) has been reevaluated for the San Onofre Nuclear Generating Station (SONGS) Unit 1. For this analysis a uniform tube plugging level of 15* was analyzed for the following case:

(Full power primary T&P) 100% license core power, 2100 psia system pressure, 201,900 gpm RCS flow, 13.7 kw/ft, 2.89 FQ.

Previous analyses have demonstrated that SONGS Unit 1 was in compliance with the AEC Interim Policy Statement, "Criteria for Emergency Core Cooling Systems for Light Water Reactors," publishes in the Federal Register June 29, 1971.

To accomplish the reanalysis for increased steam generator tube plugging and reduced RCS condition, the Interim Acceptance Criteria (IAC) assumptions and the 1971 IAC analytical models were used.

The following is a discussion of all the pertinent methods and results of this latest analysis.

A.2 RCS BLOWDOWN CALCULATION The reactor coolant system blowdown hydraulic transient was calculated using the SATAN-V computer code. The temperature of the fluid in the reactor vessel upper head was assumed to be equal to the hot leg temperature.

In the analysis, it was assumed that offsite power was lost coincident with the pipe break and that the reactor coolant pumps tripped. It was also assumed, in the analyis, that 15 percent of the tubes in each steam generator were plugged.

• 11% S.G. tube plugging plus an additional resistance equivalent to 4%

plugging to account for steam generator tube sleeving.

A-3 3849A

A.3 LOWER PLENUM REFILL/CORE REFLOOD CALCULATIONS Consistent with the assumption that offsite power is lost coincident with the break, the ECCS will effectively deliver full safety injection flow (720 lb/sec) to the reactor vessel at 26.7 seconds. The "free fall" time required for the safety injection water to drop from the cold leg nozzle to the lower plenum is approximately 0.9 seconds. It was also assumed that the lower plenum was empty (no liquid water) at the end of blowdown. The lower plenum is therefore filled, and bottom of core recovery occurs, at 68.4 seconds after the break.

The core reflood calculation was performed the same way as in the previous analysis with the exception that the resistance in the steam generators was increased to reflect the equivalent plugging of 15 percent of the steam generator tubes. As in the blowdown calculation (SATAN-V), the plugged tubes were assumed to be uniformly distributed among the three steam generators.

A.4 HOT ROD THERMAL TRANSIENT CALCULATION The hot rod thermal transient calculation was performed using the LOCTA-R2 computer program. There were no differences in the application of the LOCTA-R2 code between the previous analysis and the new analysis.

A.5 RESULTS AND CONCLUSIONS Figures A.1 through A.4 show the transient behavior for key parameters during the accident for the limiting break size of 0.8 DECLG. As shown in Figure A.4, the peak-clad-temperature is 22720F, which satisfies the Interim Acceptance Criteria PCT of 23000F. Therefore, even with an equivalent 15% tube plugging, the ECCS IAC PCT safety limit is satisfied by not exceeding an F of 2.89.

A-4 3849A

The 0.8 DECLG break size is limiting since the 1.0 and 0.6 DECLG break size resulted in lower PCTs (2174 and 21820F respectively). The change in the worst break size from the 0.6 DECLG for the 20% tube plugging case(*) is primarily the result of a reduction in the vessel inlet temperature from 5530F to 5280F.

• Reference 2 of Section 5 in this report.

A-5 3849A

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