ML20155B128

Loca/Eccs Analysis W/11% Steam Generator Tube Plugging

ML20155B128
Person / Time
Site:	Saint Lucie
Issue date:	03/20/1986
From:	Gottula R, Holm J, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17216A474	List: ML20155B128
References
XN-NF-86-23, XN-NF-86-23-R01, XN-NF-86-23-R1, NUDOCS 8604100252
Download: ML20155B128 (56)

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I XN-NF-86-23 I Revision 1 Issue Date: 3/20/86 9

ST. LUCIE UNIT 1 LOCA/ECCS ANALYSIS

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WITH 11% STEAM GENERATOR TUBE PLUGGING I

A B Contributor: C. E. Slater (ITI) 8 Written by: M 0. Udb i

• R. C. Gottula, Team Leader PWR Safety Analysis

, Reviewed by: h - J[Lo //4 J.g . Holm, Manager '

PWR Safety Analysis Reviewed by: k#7 4.'N. Ward, Manager 8 Reload Licensing 4 Concur: ~ YG1 /91L laL $b5 3-tc-SL J. N. Morgan / Manager / y Customer Sectices En'gineering Prepared by: h H. E. Williamson, Manager 3/u/rc Licensing & Safety Engineering Approved: N./

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G. L. Ritter, Manager Fuel Engineering & Technical Services

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I NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This tec'inical report was derived through resea ch and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize ' Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for liant water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.

Without derogating from tne foregoing, neithe Exxon Nuclear nor any person acting on its behaM: W A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infrings privately owned rights; or B. Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

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I Section TABLE OF CONTENTS Page

1.0 BACKGROUND

AND

SUMMARY

.............................. 1 i 2.0 LOCA/ECCS ANALYSIS - ASSUMPTIONS AND METHODOLOGY ......................................... 5 3.0 LOCA/ECCS ANALYSIS RESULTS .......................... 11 i 4.0 SETPOINT ANALYSIS RESULTS ........................... 44

5.0 CONCLUSION

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6.0 REFERENCES

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LIST OF TABLES l l

Table Page 2.1 St. Lucie Unit 1 - Changes in Analysis . . . . . . . . . . . . . . 7 2.2 St. Lucie Unit 1 System Analysis Parameters . . . ..... . 8 2.3 St. Lucie Unit 1 Core and Fuel Design i 3.1 Parameters, ENC Fuel ................................

Cal cul ated Event Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

9 14 3.2 LOCA/ECCS Analysis Results .......................... 15 4.1 Uncertainties Applied for the LCO Based on LPD ...... 45 6

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LIST OF FIGURES Fiaure Page 1.1 St. Lucie Unit 1 Axial LHR Limit . . . . . . . . . . . . . . . . . . . . 4

k. 2.1 Axial Power Profile for 14 kW/ft at y X/L=0.81 ............................................ 10 3.1 RELAP4-EM Blowdown System Nodalization 1.- for St. Lucie Unit 1 ................................ 16 3.2 Total Break Fl ow Rate vs Time . . . . . . . . . . . . . . . . . . . . . . . 17 3.3 Pressurizer Surge Line Flow Rate vs Time ............ 18 3.4 Containment Pressure vs Time ........................ 19 3.5 Normalized Power vs Time ............................ 20 3.6 Upper Core Vol ume, Pressure vs Time . . . . . . . . . . . . . . . . . 21 8 3.7 Lower Core Volume, Average Core Inlet Fl ow R a t e v s T i me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.8 U per Core Volume, Average Core Outlet F ow Rate vs Time ................................... 23 3.9 Lower Core Volume, Hot Channel Inlet Flow Rate vs Time ........................................ 24 3.10 Upper Core Volume, Hot Channel Inlet 5 F l ow Ra t e v s T i me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.11 Upper Core Volume, Hot Channel Outlet Flow Rate vs Time ................................... 26 3.12 Upper Core Volume, Hot Channel Cross Flow vs Time ........................................ 27 3.13 Upper Core Volume, Hot Channel Fluid Temperature vs Time ................................. 28 3.14 Upper Core Volume, Hot Channel Quality vs Time ..................................... 29 3.15 Peak Power Node, Cladding Surface Heat 1 Transfer Coefficient vs Time, X/L=0.81 . . . . . . . . . . . . . . 30 I

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3.16 Peak Power Node, Cladding Surface Heat Fl ux vs Ti me , X/L=0. 81 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.17 Peak Power Node, Fuel Rod Average Temperature vs T'me, X/L=0.81 ....................... 32 3.18 Peak Power Node, Cladding Sarface Temperature vs Time, X/L=0.81 ....................... 33 3.19 Double Intact Loop, Accumulator Fl ow Rate vs Ti me . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 3.20 Double Intact Loop, Accumulator Flow Rate vs Time after E0BY ............................. 35 .

3.21 Double Intact Loop, SIS Flow Rate vs /

Time after E0BY ..................................... 36 3.22 Single Intact Loop, Accumulator Flow Rate vs Time ........................................ 37 3.23 Single Intact Loop, Accumulator Flow Rate vs Time after E0BY ............................. 38 3.24 Single Intact Loop. SIS Flow Rate vs Time after E0BY ..................................... 39 3.25 Downcomer, Mixture Level vs Time .................... 40 '

3.26 Core Fl oodi ng Rate vs Time . . . . . . . . . . . . . . . . . . . . . . . . . . 41 3.27 Core Mixture Level vs Time .......................... 42 3.28 T00DEE2, Cladding Temperature vs Time, as 0.8 DECLG, 14 kW/ft at 0.81 X/L ..................... 43 g 4.1 Veri fi cati on of LCO on Local Power Densi ty . . . . . . . . . . 46 I

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1.0 BACKGROUND

AND

SUMMARY

In November 1985, Exxon Nuclear Company (ENC) reported revised LOCA/ECCS Enalyses for St. Lucie Unit I which corrected errors in previous analyses and increased the level of analyzed steam generator tube plugging to 157.. (1) Reference I reported the first phase of the analysis and provid-

'i ed axially dependent linear heat rate (LHR) limits for St. Lucie Unit I operating with ENC 14x14 fuel for Cycle 7 and future cycles. The LHR limits were established by LOCA/ECCS analysis calculations for the

limiting break (0.8 double-ended cold leg guillotine). The calculations were performed at the exposure at which the peak stored energy occurred.

Analyses confirming that the conditions for the analysis in Reference 1 are limiting 3re documented in Reference 2. These results confirmed the k

0.8 double-ended cold leg guillotine as the limiting break. These results also confirmed that the peak cladding temperature occurs at the exposure where the stored energy is maximum (1.8 MWD /kg peak rod average g burnup). The axially dependent linear heat rate (LHR) limits resulting a from the analyses presented in References 1 and 2 for St. Lucie Unit I are shown in Figure 2.1 of Reference 1. These limits, provide for an allowable LHR of 15 kW/ft up to a relative core height of 0.6, and decreasing linearly to 13.4 kW/ft at a relative core height of 0.81 and to 10.07 id/ft at a relative core height of 1.0.

The purpose of the analysis presented in this report was to support an increase in the allouable LHR at relative core heights above 0.6; specif-ically, 14 kW/ft at a relative core height of 0.81. This analysis g assumed an average steam generator tube plugging of 117., and included W several other modifications in plant data and operation that are listed in Section 2.0 of this report. A top peaked axial power profile was used which bounded the predicted profiles for EOC conditions.

I The results of this analysis showed the peak cladding temperature to be 2183*F for the top peaked profile analyzed. These results support I

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operation at up to 14 kW/ft at a relative core height of 0.81. Combining these results with those in References 1 and 2, the axially dependent LHR limits can be defined as shown in Figure 1.1. This provides for in-creased operating margin compared to the current Technical Specification limits. These results bound expected conditions for Cycle 7 and future cycles using the current ENC fuel design, and are valid for up to an gi average of 11% steam generator tube plugging. k i 1

Operation of the St. Lucie Unit I reactor with ENC 14x14 fuel at or below the LHR limits of Figure 1.1 assures that the U.S. NRC acceptance crite- l ria for Loss-of-Coolant Accident breaks up to and including the double-ended severance of a reactor coolant pipe (specified by 10 CFR 50.46(b)) will be met with the emergency core cooling system for the St. l Lucie Unit I reactor. That is: '

(1) The calculated peak fuel element cladding temperature does not exceed the 2200*F limit.

(2) The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.

(3) The cladding temperature transient is terminated at a time when the core is still amenable to cooling. The hot fuel g

rod W

, cladding oxidation limit of 17% is not exceeded during or after quenching.

(4) The system long-term cooling capabilities provided for the g initial core and subsequent reloads remain applicable to ENC W fuel.

The Local Power Density Limiting Condition for Operation (LPD-LCO) vas determined as part of this analysis since it is a function of the axially dependent LHR limits shown in Figure 1.1. The proposed LPD-LCO barn is 1 l

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I shown in Figure 4.1. The barn limits power to 88% over an ASI range of 0.02 to 0.08.

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2.0 LOCA/ECCS ANALYSIS - ASSUMPTIONS AND METHODOLOGY The LOCA/ECCS analysis used the EXEM/PWR ECCS evaluation model(3) as described in Reference 1. The St. Lucie Unit I system input was the same as that in References 1 and 2 with the exceptions noted in Table 2,1.

These changes were justified by actual plant operation and comparison of the system input values to plant data. The revised analysis supports g higher LHR values than supported by the analyses reported in References 1 f 3 and 2.

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The axial power profile used in the analysis is shown in Figure 2.1.

This profile has a peak at a relative core height of 0.81 and bounds the g predicted EOC shapes, including uncertainties. The axial peaking factor W was 1.17. A radial peaking factor of 1.84 was used in the analysis.

This value is larger than the Technical Specification value of 1.7 to h account for inasurement uncertainties and control rod peaking factor augmentation. An assembly local peaking factor of 1.1 for the hot assembly wr used.

Primary system measured flow rate (395,877 gpm) and pressure drops at the current steam generator tube plugging level were used as a basis to establish the initial primary system flow rate used in the analysis.

Best estimate system loss coefficients were determined from the measured flow rate and pressure drop. The loss coefficients were then adjusted for asymmetric steam generator tube plugging of 9 and 13 percent. The primary system flow split to each steam generator was then calculated.

The resulting best estimate loop flow rate was determined to be 386,121

gpm. Since a best estimate loop flow rate is used, no additional limit is placed on the core flow rate over those imposed by DNBR considera-tions. The analysis is applicable only up to a steam generator plugging level of 11%.

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St. Lucie Unit I system parameters used in the analysis are shown in Table 2.2. Table 2.3 shows the core and fuel design parameters used in the analysis.

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Table 2.1 St. Lucie Unit 1 - Changes in Analysis Ref.1 & 2 This Analysis Average steam generator-I 15% 11%

tube plugging (17% broken loop, (13% broken loop, 13% intact loop) 9% intact loop)

I Steam generator secondary side initial liquid mass 90,367 lbm 131,745 lbm Accumulator line resistance 7.5 5.94 Initial containment 90*F 100*F -

temperature I Secondary steam flow and feedwater flow Instantaneous 100% steam flow for 1.4 sec after break initiation, followed by a linear ramp to 0.0 flow in 0.3 sec.

Linear coastdown in feedwater flow to 0.0 in 2 see follow-I ing break initiation.

Core cross-flow resistance 10 30 Core average LHR, kW/ft* 6.42 6.35 l 1

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5 The number of active rods in Cycle 7 was already considered in References 1 and 2 except for the system blowdown calculation. This I change only applies to the system blowdown calculation. I I

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Revision 1 1 5

Table 2.2 St. Lucie Unit 1 System Analysis Parameters Primary Heat Output, MWt 2700*

Primary Coolant Flow Rate, lbm/hr 1.452 x 108** (386,121 gpm)

Primary Coolant System Volume, ft 3 19217***

Operating Pressure, psia 2250 l Inlet Coolant Temperature (hottest loop), 'F 549 g'

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Reactor Vessel Volume, ft 4522 W Pressurizer Volume, Total, ft 3 1500 Pressurizer Volume Liquid, ft 3 888 Accumulator Volume, Total, ft3 (one of four) 2020 i Accumulator Volume, Liquid, ft 3 1090 Accumulator Pressure, psia 230 Steam Generator Tube Plugging 137. - 97, split SteamGeneratorSecondarySigeHeat 78474 Transfer Area,137. SGTP, ft SteamGeneratorSecondaryS{deHeat 82082 g Transfer Area, 97. SGTP, ft g Steam Generator Secondary Flow Rate, lbm/hr 137.SGTP)

(49-517. power split) 5.868x10f(97.SGTP) 6.108 x 10 ( E Steam Generator Secondary Pressure, psia 823 3 Reactor Coolant Pump Head, ft 272 Reactor Coolant Pump Speed, rpm 886 2

Moment of Inertia, 1bm-ft 101,900 Cold Leg Pipe, I.D., in. 30 g Hot Leg Pipe, I.D., in. 42 E Pump Suction Pipe, I.D., in. 30 l

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• Primary Heat Output used in RELAP4-EM Model - 1.02 x 2700 = 2754 MWt.

• • Best Estimate Flow (37. flow measurement uncertainty not subtracted).

• • • Includes total accumulator and pressurizer volume,11Y. SGTP.

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Table 2.3 St. Lucie Unit 1 Core and Fuel Design Parameters, ENC Fuel I

i Fuel Assembly Rod Diameter, in. .440 Fuel Assembly Rod Pitch, in. .580 Fuel Assembly Pitch, in. 8.180 Fueled (Core) Height, in. 136.7 Fuel Heat Transfer2 Area for Cycle 7 48,967 I (heated rods), ft 2

Fuel Total Flow Area, ft 53.19 Fuel Total Flow 52.70 (excludes areaArea due toforspacer Reflood Calculat{on volume), ft I

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E 3.0 RESULTS

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h The significant event timings for the Loss-of-Coolant Accident (0.8 -

DECLG) analyzed are shown in Table 3.1. The event timings are similar to those given in Table 3.5 of Reference 1 except for the start of reflood I time. The start of reflood time for this analysis was 38.78 sec, as compared to 39.71 sec in the previous analysis.(1) This is due to the lower accumulator line resistance which allows slightly higher accumula-fl tor flow rates and the lower plenum and downcomer to fill more quickly. >

Also, the end-of-bypass time was 21.64 sec, as compared to 21.25 sec in I the previous analysis. The difference in the heatup time during refill of 1.32 sec had a significant effect on the temperature of the ruptured node during reflood. A metal-water reaction excursion was prevented for 3 the case with an LHR of 14 kW/ft at the 0.81 elevation. -

I The effect of the other changes in the analysis listed in Table 2.1 are -

discussed below. The majority of these changes resulted in small effects  ;

I on the calculated PCT. -

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The reduced steam generator tube plugging level relative to the previous =

analysis had the effect of increasing the reflood rate and reducing the _

peak cladding temperature (PCT). This effect is not large since the

_ B change in resistance in the steam generator is not large compared to the -

total line resistance and the resistance due to the locked pump rotor. '

8 The larger initial secondary liquid mass has a small effect on the results. During the reflood portion of the transient when heat is I transferred from the secondary to the primary system, the larger secon-dary mass tends to maintain higher steam temperatures on the primary side, which results in a reduced reflood rate and higher PCT. However, an opposite and compensating effect occurs during blowdown when heat is transferred from the primary to the secondary system. The overall effect I _

I 12 XH-NF-86-23 m Revision 1 g of increased secondary liquid mass on PCT is estimated to be less than a g 10*F increase. 5 The change in the initial containment temperature from 90*F to 100*F also had a small effect on PCT. The higher initial containment temperature ,

caused the containment pressure to be slightly higher during the trai-sient, which resulted in a higher reflood rate. This higher reflood rat'e -

re:ults in a small reduction in the PCT, about 15'F. '

The change in the modeling of the operation of the secondary feedwatcr and steam valves had a significant effect on the transient during the blowdown period as compared to the previous analysis with instantaneous isolation of the secondary system. The increased feedwater flow is g estimated to have a minor effect on the transient. However, the in- 5 creased steam flow over a period of 1.7 sec removed more heat from the primary system. This caused the fluid in the cold legs to remain subcooled slightly longer and resulted in increased pump head and mare l

flow for a few seconds. The additional flow to the core resulted in a g higher heat transfer coefficient for a short period of time and more 5 energy removal from the rods. The overall result was a reduction in the average fuel temperatt.re of 50*F for the hot node at the end-of-bypass.

The reduced stored energy at the end-of-bypass led to a Icwer PCT.

The sensitivity in the results due only to a change in the core cross-flow resistance from 10 to 30 was not quantified in this z.nalysis.

However, little difference in the cross-flows between the hot channel and h

average core volumes was seen between this analysis and previous analy-ses. It was estimated that the change in core cross-flow resistance did not.have a significant effect on the results.

The core average LHR for Cycle 7 is less than for pre <ious cycles due to a larger number of active rods (37,316 versus 36,932). This effect was accounted for in the previous analyses described in References 1 and 2 in all but the system blowdown calculation. The larger number of active E

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8 13 XN-NF SG-23 Revision 1 rods and lower core average LHR were included in the system blowdown calculation for this analysis. The effect of this change alone was not  !

quantified. However, it is estimated that this effect on the system response during blowdown is minor and that the system boundary conditions placed on the hot channel calculation are unaffected.

The peak cladding temperature for this analysis was calculated u be

, 2183*F at a relative core height of 0.875 (9.97 ft). Significant results l of the analysis are tabulated in Table 3.2. Plots of various system l paruneters are shown in Figures 3.2 through 3.28.

In the break spectrum calculations reported in Reference 2, the only break size with a PCT relatively close to the 0.8 DECLG case was the 1.0 DECLG. The 1.0 DECLG case blowdown and hot channel calculations were rerun with the system parameter changes mentioned in Section 2 to verify that the 0.8 DECLG ca:.e was still the limiting break size. The average fuel temperature at the peak power node at the end-of-bypass was found to be 92*F lower for the 1.0 DECLG case, thus verifying that the limiting break size is still the 0.8 DECLG.

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E Table 3.1 Calculated Event Times Time of Event Event (sec) 1 Start 0.0 g Break is Fully Open 0.05 W Safety Injection Signal 0.90 Pressurizer Empties 9.1 Accumulator Injection begins, Broken Loop 12.5 Accumulator Injection begins, Single Intact Loop 16.7 Accumulator Injection begins, Double Intact Loop 16.7 End-of-Bypass 21.64 i Safety Injection Flow, SIS 30.90 Start of Reflood 38.78 Accumulators Empty, Broken Loop 59.09 Accumulators Empty, Single Intact Loop 62.54 Accumulators Empty, Double Intact Loop 62.44 Peak Clad Temperature is Reached 180.0 I

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Table 3.2 Analysis Results l

Peak LHR, kW/ft 14.0 0 X/L = 0.81 Hot Rod Burst Time (sec) 41.74 Elevation (ft) 9.22 E -

Fraction of Flow Area Reduction .441 Peak Clad Temperature Temperature (*F) 2183 180.0 Time Elevation (sec)(ft) 9.97 Metal-Water Reaction

- Local Maximum, % 4.5*

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Elevation of Local Max. (ft)

Core Maximum, %

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,I *At 200 seconds.

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4.0 SETPOINT ANALYSIS PES'JLTS The setpoint analysis results for Cycle 7 were reported in Reference 4 except for the Local P:wer Density Limiting Condition for Operation (LPD-LCO). The LPD-LCO is a function of the allowa'cle LHR. This report justifies a revision in the allowable LHR and thus a change in the I LPD-LCO.

The plar.t technical specifications ellow plant operation for limited periods of time with the in-core deteccors out of service. In this I situation, the LPD-LCO barn provides );otection in steady state operation against penetration of the LHR limit established by LOCA considerations.

ENC statistical methodology used to define the LPD-LCO is described in References 5, 6 and 7. The axially dependent LHR limit shown in Figure 1.1 was used to determine the allowed power versus ASI. The statistical I analysis included the effect of appropriate uncertainties. These uncer-tainties are listed in Table 4.1. The points in Figure 4.1 are calculat-ed as described in Refer ance 6. The proposed LPD-LCO barn is shown in Figure 4.1.

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1 45 XN-NF.86-23 Revision 1 l

Il Table 4.1 Uncertainties Applied for the LCO *dased on LPD Source Value*

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, Poyer measurement 0.02 of rated ASI uncertainty i 0.05 P

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5.0 CONCLUSION

S l A LOCA/nCCS analysis was performed for St. Lucie Unit I with ENC 14x14 I

fuel using the EXEM/pWR ECCS Evaluation Model in conformance with Appen-dix K of 10CFR50, The purpose of the analysis was to support an increase I-in the allowable LHR limits at relative core heights above 0.6 and specifically 14 kW/ft at a relative core height of 0.81. The analysis was performed for the previously determined limiting break size (0.3 DECLG) and considered an average 11% steam generator tube plugging along I with several other changes in system parameters. The analysis was performed fcr bounding exposure conditions, and applies for St. Lucie i

Unit 1 Cycle 7 and beyond using ENC supplied fuel.

Axially dependent LHR limits, shown in Figure 1.1 and supported by the calculations documented in References 1, 2 and this report were deter-mined,and assure conformance with NRC criteria. The limits calculated for ENC fuel are equal to or conservative with respect to those currently in place for C.E. fuel. Additionally, the C.E. fuel will operate with I LHRs significantly below those of ENC fuel due to the C.E. fuel exposures relative to the ENC fuel exposures. It is therefore conservative to monitor the C.E. fuel to the limits established for the ENC fuel.

Operation of the St. Lucie Unit I reactor with ENC 14x14 fuel within the I limits given in Figure 1.1 assures that the St. Lucie 1 emergency core cooling system will meet the acceptance criteria as required by 10 CFR 50.46. That is:

I (1) The calculated peak fuel element cladding temperature does not exceed the 2200*F limit.

(2) The amount of fuel element cladding which reacts chemi:: ally with water or steam does not exceed 1% of the total amount of zircaloy in the core.

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48 XN-NF-86-23 '

P.evision 1  !

(3) The cladding temperature transient is terminated at a time when I

the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or i after quenching.

I (4) The system long term cooling capabilitics provided for previous cores remain applicable to cores containing ENC reload fuel.

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6.0 REFERENCES

(1) "St. Lucie Unit 1 Revised LOCA-ECCS Analysis with 157. Steam Genera-i tor Tube Plugging," XN-NF-85-117, Exxon Nuclear Company, Richland, WA, November 1985. ,

1 (2) "St. Lucie Unit 1 Revised LOCA-ECCS Analysis with 157. Steam Genera- l Plugging - Break Spectrum and Exposure Results,"

! tor Tube 1

XN-NF-85-117. Supp. 1, Exxon Nuclear Company, Richland, WA, December 1985.

3 (3) "St. Lucie Unit 1 LOCA Analysis Using the EXEM/PWR ECCS Model , "

g XN-NF-82-98. Supo. 1. Revision WA, January 1983.

1, Exxon Nuclear Company, Richland, I (4) "St. Lucie Unit 1 Cycle 7 Safety Analysis Report," XN-NF-85-73.

Revision 2 Exxon Nuclear Company, Richland, WA, October 1985.

(5) " ENC Setpoint Methodology for CE Reactors," XN-NF-507. Revision 1, I Exxon Nuclear Company, Richland, WA, July 1980.

(6) " ENC Satpoint Methodology for CE Reactors, Statistical Setpoint Methodology," XN-NF-507(P). Sucolement 1, Exxon Nuclear Company, I Richland, WA, September 1982.

(7) " ENC Setpoint Methodology for CE Reactors, Sample Program,"

XN-NF-507(P1. Supolement 2, Exxon Nuclear Company, Richland, WA, November 1982.

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I I XN-NF-86-23 Revision 1 Issue Date: 3/20/86 ST. LUCIE UNIT 1 LOCA/ECCS ANALYSIS 11% STEAN GENERATOR TUBE PLUGGING I

Distribution I a MJ Ades KA Bryan RA Copeland l NF Fausz LJ Federico RC Gottula I TJ Helbling JS Holm JW Hulsman SE Jensen WV Kayser JN Morgan GL Ritter l- g T Tahvili g GN Ward HE Williamson CE Slater (ITI)

FPL/TJ Helbling (25)

Document Control (5)

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