W3P85-2115, Forwards Depressurization Capability W/Degraded Auxiliary Pressurizer Spray, Satisfying License Condition 12.Analysis Demonstrates That Sys Can Perform Necessary Depressurization W/Isolation Valve Failing to Open

Forwards Depressurization Capability W/Degraded Auxiliary Pressurizer Spray, Satisfying License Condition 12.Analysis Demonstrates That Sys Can Perform Necessary Depressurization W/Isolation Valve Failing to Open

ML20126E743
Person / Time
Site:	Waterford
Issue date:	06/13/1985
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
W3P85-2115, NUDOCS 8506170205
Download: ML20126E743 (36)

Text

a -I LO U 181 A N A / 44. oeuao~o. .ra. r P O W E R & L I G H T / New cat Ana toVSANA

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70174-800G . (504]386-9345 hiuklNSYS June 13, 1985 W3P85-2115 3-A1.01.04 A4.05 Director of Nuclear Reactor Regulation Attention: Mr. G.W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Waterford SES Unit 3 Docket No. 50-382 Operating License NPF-38 Condition 12 LP&L Report: Depressurization Capability with Degraded Auxiliary Pressurizer Spray

Dear Sir:

LP&L hereby submits the subject report. The report provides the information required to satisfy License Condition 12. The results of LP&L testing and analysis demonstrate that the auxiliary spray system can perform the necessary depressurization to meet the SGTR accident acceptance criteria with a charging loop isolation valve postulated failed open.

Please feel free to contact me or Robert J. Murillo should you have any questions.

Yours very truly,

{.O~.-- (j'Y K.W. Cook )!V Nucicar Support & Licensing Manager KWC/RJM/pc1 Attachment cc: B.W. Churchill, W.M. Stevenson, R.D. Martin, D.M. Crutchficid, J. Wilson, T.A. P11ppo G506170205 050613 ap PDR ADOCK 050003G2

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'. n i DEPRESSURIZATION CAPABILITY WITH DEGRADED AUXILIARY PRESSURIZER SPRAY l

1.0 INTRODUCTION

l While reviewing information to address NRC questions concerning the l

Decay Heat Removal Capability at Waterford 3, a potential single failure vulnerability in the pressurizer auxiliary spray system was identified.

Auxiliary spray would be used to depressurize the RCS during a Steam Generator Tube Rupture (SGTR) event with main spray unavailable (reactor

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' coolant pumps not running). To initiate auxiliary spray the operator must close both charging loop isolation valves, CVC-218A and CVC-218B (See l

Figure 1-1). If one of the valves fails to close (i.e., mechanically sticks open), then some of the flow would be diverted to the charging line instead of to the pressurizer due to differences in elevation and flow resistance. Thus the spray flow might be insufficient to depressurize the RCS and minimize the radiological consequences of the event.

To respond to this concern, LP&L has performed a test at the Waterford -

3 plant to determine the depressurization rate using auxiliary pressurizer spray with one charging loop isolation valve open. In addition, a SGTR analysis with degraded auxiliary spray has been performed to verify that the V~

acceptance criteria given in SRP 15.6.3 and 10 CFR 100 can be met. This report addresses the license requirement imposed by NRC to submit the results of a confirmatory depressurization test and demonstrate that the '

auxiliary pressurizer spray can perform the necessary depressurization to meet the SGTR acceptance criteria (SRP 15.6.3) with a charging loop isolation valve failed open.

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. g Section 2 provides a susunary of the results of both the test and analysis. Section 3 describes the degraded auxiliary spray flow test performed at Waterford 3. Section 4 describes the SGTR analysis performed with a degraded auxiliary spray flow. Section 5 provides the conclusions from this study. A comparison of the analysis given in this' report to the SGTR analysis in Section 15.6 of the FSAR is presented in Appendix A.

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SUMMARY

This report confirms that if a charging loop isolation valve failed open during a SGTR event, the' acceptance criteria (radiological dose consequences) would be met. Test results showed that an auxiliary spray flow rate of about 37 gpm was achieved with a charging loop isolation valve open. Analysis of a SGTR with an auxiliary spray flow rate of only

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10 gpm showed that the offsite radiological dose released.was well below the

-acceptance criteria given in 10 CFR 100. Thus, since the actual spray flow rate was much greater than the acceptable spray flow rate from the analysis,

. the auxiliary spray system can perform its function during a SGTR event with one charging loop isolation valve failed open such that the acceptance criteria are' met.

The analysis showed that depressurization during a SGTR with a degraded spray flow rate of 10 gpm was controlled predominately by adjusting HPSI flow.

The spray flow served to initiate the drop in pressure that allowed HPSI flow to increase and fill the pressurizer. With an adequate level in the pressurizer (and subcooling), the operator would throttle the HPSI flow according to the Emergency Operating Procedures and depressurize the RCS. .

Shutdown cooling conditions were reached with the radiological dose consequences well below the acceptance criteria given in 10 CFR 100, e-This report confirms thar *he acceptance criteria for radiological release during a SGTR with a tharging loop isolation valve failed open can be met with the existing Waterford 3 design. The results show that the public health and safety are satisfactorily protected ,ithout any plant

. modifications.

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s 3.0 AUXILIARY SPRAY FLOW TEST A test was performed at Waterford 3 to measure the depressurization rate using auxiliary spray with one charging loop isolation valve open. The test was performed during the startup test program with the RCS at 545'F and 2250 psia. Two charging pumps were used to provide spray flow. This is based on the Technical Specification requirement that two charging pumps be operable and the operator uses both pumps to maximize the spray flow rate during the postulated SGTR. The charging isolation valve CVC-218B was opened because this line has a slightly lower flow resistance than the other charging line.

.This minimizes the auxiliary spray flow. The reactor coolant pumps were tripped and not running during the test. The initial test conditions are given in Table 3-1.

The test was initiated by starting auxiliary spray flow from two charging pumps (88 gpm) with both charging loop isolation valves closed.

After the pressure decreased about 100 psi, one charging loop isolation valve (CVC-218B) was opened. The depressurization with the degraded spray flow continued until the pressurizer pressure decreased to 2000 psia.

The depressurization from this test is shown in Figure 3-1 and Table 3-2.

The initial depressurization rate with full flow from two charging pumps was about 60 psi / min. When a charging loop isolation valve was opened, the e depressurization rate dropped to 24 psi / min.

A calculation was performed using the RETRAN-02 computer code to deter-mine the degraded auxiliary spray flow rate corresponding to the measured depressurization rate.

For a flow rate of 88 gpm, the calculated depressur-Tzation rage was 59 psi / min. This agrees well with the measured rate for the

. a initial part of the test and therefore verifies the adequacy of the calculation. A depressurization rate of 24 psi / min. was calculated with an auxiliary, spray flow rate of 37 gpm. Thus, with a charging loop isolation valve open, an auxiliary spray flow rate of about 37 gpm was achieved. This matches the average spray flow of 36 gpm measured by a sonic flow measurement device during a portion of the time with degraded spray.

The test was performed with a pressurizer level of about 35% (10.5 ft.).

A lower level would reduce the elevation head seen by the charging flow. This would result in a higher charging flow and lower spray flow when the charging isolation valve was open. A difference of 10.5 ft. however would reduce the auxiliary spray flow rate by only a small amount (less than 5 gpm).

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TABLE 3.1 DEGRADED AUXILIARY SPRAY INITIAL TEST CONDITICNS Pressurizer Pressure, psia 2243 Pressurizer Level, % 34.6 Average Cold Leg Temperature. *F 535 Average Hot Leg Temperature. 'F 535 Regenerative' Heat Exchanger Outlet (Charging / Spray)

Temperature. *F' 354 Reactor Coolant Pumps Not Running Charging Pumps B, A/B Running 1

. o TABLE 3.2 DEGRADED AUXILIARY SPRAY TEST DEPRESSURIZATION Time (Min.) Pressurizer Pressure (PSIA) 0 2242.7 1.0 2241.7 2.0 2241.0 3.0 2239.5 3.5 2214.0 4.0 2181.7 4.5 2154.0 5.0 2135.2 5.5 2128.5 6.0 2121.0

. 6.5 2107.5 7.0 2094.0 7.5 2080.5 8.0 2067.0 8.5 2054.2 9.0 2041.0 9.5 2029.5 10.0 2016.7 10.5 2004.7 Auxiliary Spray valve opened at 3.1 minutes.

Charging loop isolation valve opened at 4.8 minutes.

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FIGURE 3-1 DEPRESSURIZATION FOR DEGRADED AUXILIARY SPRAY TEST (MARCH 11, 1985)- ,

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4.0 SGTR ANALYSIS An analysis of a single double ended Steam Generator Tube Rupture (SGTR) event with degraded auxiliary spray flow has been performed.

Degraded spray flow would result in a reduced RCS depressurization capability 1

and, therefore^ would increase the time needed to minimize the primary to secondary' leak rate. It would also increase the time necessary for the RCS to reach shutdown cooling conditions. The analysis was performed to determine that, with degraded auxiliary pressurizer spray, the acceptance criteria (radiological dose released) for the SGTR event can be met.

4.1 Analysis Method and Assumptions The SGTR analysis presented in this report assumed that with two charging pumps operating, an auxiliary spray flow rate of only 10 gpm was achieved. The remaining 78 gpm flowed to the RCS cold leg through the failed 1,

open charging loop isolation valve. The transient simulation was run for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. ,

The radiological dose to the thyroid at the exclusion area boundary (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, EAB), and the low population zone (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, LPZ) were calculated. The dose for both the Preaccident Iodine Spike (PIS) and the accident Generated i Iodine Spike (GIS) were calculated as identified in Standard Review Plan Section 15.6.3 e.

The analysis was done using the CESEC-III computer code prior to and shortly after reactor trip. The CEPAC code, which is a long term cooldown

, algorithm based on CESEC-III, was used after reactor trip during plant cooldown. CESEC-III is used by Combustion Engineering for licensing analyses presented in Chapter 15 of the FSAR and support of plant start-up tests.

CEPAC was used for the cooldown because of its ability to simulate operator actions and calculate the offsite dose release.

The method and assumptions of the analysis are given below:

1. Reactor trip was initiated by the Core Protection Calculators (CPCs) on the approach of the RCS hot legs to saturation conditions. Reactor trip could also have been initiated by CPCs on a pressurizer pressure out of range low (1845 psia) signal with no change in results since the two trip signals occurred at

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almost the same time.

2. Offsite power was assumed to be available.

3. The steam bypass system, although available, was not credited in the calculation of radiological doses. This maximized the radiological consequences for the event since steam was released directly to the atmosphere.

4. The Waterford 3 Emergency Operating Procedure, OP-902-007, (EOP) for a SGTR (which is based on the NRC approved guidelines developed by the CE Owner's Group in CEN-152) was followed in simulating e

operator actions during the plant cooldown. A concise summary of the operator actions is given in Figure 4-1.

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5. Subsequent to reactor trip, it was assumed that five minutes passed (wherein the operators perform the Emergency Entry Procedure) before the operators determined that a SGTR event occurred.

Also,'a minimum of two minutes was assumed between a set of related operator actions. For instance, if the operator detected that the subcooling margin was too high he may do one or all of the following simultaneously (which are considered a related set of operator i

actions): throttle HPSI flow, turn on auxiliary spray, or turn off pressurizer heaters. These " timing" assumptions are consistent with ANSI-N660.

6. The RCPs were tripped by the operator when the RCS pressure dropped to 1621 psia. The RCPs (and main pressurizer spray) were 4

not available for the duration of the event.

7. Consistent with the E0P, the RCS hot legs were cooled to 550*F (at 100*F/HR per the Technical Specifications) using the Atmospheric Dump Valves (ADVs) for both steam generators before isolating the affected steam generator.

8. High Pressure Safety Injection (HPSI) flow was throttled only

.when all the termination criteria were satisfied (Subcooling 1 28*F, and PZR level 1 28%, and available SG wide range level is y both 1 50% and constant or rising).

9. Control of the affected steam generator level was performed with '

the steam generator blowdown system to maintain level at 85-90%

4 Wide Range. The unaffected steam generator level was maintained r

using the main feedwater control system at about 68% Narrow Range.

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10. It was assumed that the operator did not stop the depressurization in order to collapse the steam void formed in the reactor vessel upper head.

11. The discharge.of fluid through the blowdown system was accounted for in the radiological release calculation in the same manner as the leak from the RCS. The flashed fraction of steam in the blowdown flash' tank is treated as steam released directly to the atmosphere with a partition coefficient of 0.01.

12. The assumptions used for the radiological dose calculation are given in Table 4-1 and are the same as that specified in the FSAR and Standard Review Plan Section 15.6.3

13. The enthalpy of the auxiliary pressurizer spray was held constant at a value based on normal operating conditions with letdown.

The decrease in spray temperature as letdown (RCS) temperature decreased was not accounted for. The high spray temperature reduced the effectiveness of the spray to depressurize the RCS.

14. Nominal full power values were used for the initial conditions as listed in Table 4-2.

t 4.2 Analysis Results The radiological thyroid dose results are given in Table 4-3. As can be seen, the dose released for a SGTR with only 10 gpm auxiliary spray is

well below the 10 CFR 100 limit for both GIS and PIS assumptions. The major contribution to the dose is due to steam released from the affected steam generator. This is a conservatively high dose release due to the analysis assumptions and operator actions.

The thermal-hydraulic response for a SGTR with 10 gpm auxiliary spray flow is shown in Figures 4-2 to 4-8. Table 4-4 lists the sequence of events.

The reactor coolant system began to depressurize when the tube rup-ture was initiated (Figure 4-2). The pressure continued to decrease until a reactor trip signal was generated at 13 minutes by the CPC on the approach to hot leg saturation conditions. After reactor trip, the RCS pressure decreased rapidly and initiated safety injection. The pressurizer emptied (Figure 4-3) shortly af ter reactor trip due to the leak flow and RCS shrinkage. The RCPs were tripped by the operator due to low RCS pressure

• resulting in a natural circulation flow. The steam generator secondary pressure increased to open the main steam safety valves .due to the reactor / turbine trip and unavailability of the steam bypass system (Figure 4-8).

At about 20 minutes the operator started to cool the RCS until the hot leg temperature (Figure 4-4) was below 550 F using ADVs in both steam generators. The affected steam generator was then isolated and the cooldown to an unaffected loop hot leg temperature of 500*F was continued using the e unaffected steam generator. During this time the pressure remained i

relatively constant with pressurizer level near zero and subcooling increasing. At 80 minutes, RCS hot leg temperature reached 500 F and auxi-liary spray was initiated to depressurize the RCS.

Pressurizer pressure decreased due to the auxiliary spray flow. This allowed HPSI flow (Figure 4-5) to increase and begin filling the pressurizer.

The rise in pressurizer level compressed the steam space causing pressure to increase slightly. The pressure leveled out at a value where compression of the steam space was balanced by condensation from the auxiliary spray.

Pressure remained constant until pressurizer level increased to the value that the HPSI flow could be throttled according to the termination criteria in the E0P. The reduced HPSI flow caused the pressure to drop, but also caused the pressurizer level to decrease below the safety injection termination criteria.

The HPSI flow was therefore increased to reestablish the pressurizer level

, which caused the RCS pressure to increase. Subsequent throttling of the HPSI flow, governed by the need to maintain adequate pressurizer level and sub-cooling (Figure 4-6), controlled the depressurization rate. The oscillations in the pressurizer pressure and level were due to these adjustments in the HPSI flow rate. The primary to secondary leak rate (Figure 4-7) decreased as RCS pressure decreased. Shutdown cooling conditions were reached at 5.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the tube rupture.

Throughout the natural circulation cooldown (as specified in the E0P),

the optimum temperature difference between the secondary sides of the steam generators was maintained. This was done in a conservative manner by assuming that the operator periodically opened the ADV on the affected steam generator to reduce its temperature (Figure 4-8). The increased steam ,.

release from the affected steam generator causes a higher radiological dose to be calculated than if the affected steam generator ADV were kept closed.

This is a conservative analysis assumption since the Waterford 3 E0P does not call for the affected steam generator ADV to be opened in this situation.

Instead, the operator would use feedwater and blowdown to feed and drain the affected steam generator, thereby reducing its temperature.

4.3 Analysis Conservatisms The SGTR analysis described above includes conservative assumptions and methods. Some of these conservatisms are identified below.

1. The auxiliary spray temperature was maintained constant at 375'F.

Not accounting for the decrease in spray temperature as RCS letdown temperature decreased resulted in a reduced depressurization capability for auxiliary spray as shutdown cooling conditions were approached.

2. The auxiliary spray flow rate used for the analysis (10 gpm) was well below the calculated spray flow rate achieved during the test (37 gpm).

3. Steam was periodically released during the cooldown from the affected steam generator through the ADV to cool the steam generator and maintain the optimum temperature difference between steam generators. This resulted in a higher radiological dose release than if the affected ADV were kept closed.

4. The Steam Bypass Control System, which directs steam to the condenser, was not credited. This would reduce the radioactivity e released due to scrubbing and dilution with the condenser water.

Releasing steam directly to the atmosphere through the ADV maximized the radiological dose release.

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5. Conservative atmospheric dispersion factors were used that are

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consistent with the SGTR analysis in FSAR Section 15.6 and SRP 1

15.6.3. Best estimate values would significantly reduce the cal-4 culated offsite dose.

6. Initial activity levels in the RCS and steam generator were taken to be at the Technical Specification limit. Typical values would be much lower.

7. The Shutdown Cooling (SDC) System was not used as soon as the RCS conditions would allow SDC to be entered. The additional steam released from the steam generators during the two hours af ter SDC conditions were reached resulted in a higher radiological dose release than if SDC had been entered as soon as possible.

4

Table F1 ' -

Radiological Considerations PIS - Coincident (existing) 1317 spike 60 uCi/gm 2

GIS - Spiking Factor (Increase in Iodine Release Rate From Fuel Rods Due to Reactor Trip) 500 131 Ao -

Initial Primary Fluid Activity 7 2

Dose Equivalent 1.0 uCi/gm 131 A

gg Initial Secondary Fluid Activity 7 2

Dose Equivalent 0.1 uCi/gm H -

Partition Coefficient (ratio of iodine concentration in vapor to iodine concentration in water within affected steam generator) 0.01 BR -

Breathing Rate -4 3 M

3.47 x 10 sec X/Q -

Exclusion Area Boundary (0-2 HR) Atmospheric -4 Dispersion Factor se 6.3 x 10 3 M

X/Q -

Low Population Zone Outer Boundary (0-8 HR) -5 Atmospheric Dispersion Factor 7.1 x 10 sec M

Leak Rate - Primary to Secondary Leak Rate Within yM Intact Steam Generator min

+

,, _ _ _ _ _ _ ,y.

Table 4-2 Initial Conditions for the Steam Generator Tube Rupture Analysis with Degraded Auxiliary Spray Assumed Parameter Value

' Core Power Level, MWt 3410 Core Inlet Coolant Temperature, 'F 553 Reactor Coolant System Pressure, psia 2250 6

Core Mass Flow Rate, 10 lbm/hr 143 Steam Generator Pressure, psia 900 Moderator Temperature Coefficient, 10-4 ao/*F -0.68 ,

Pressurizer Water Level, percent 55 Core Protection Calculator (CPC) Temperature Margin Necessary to Generate Reactor Trip Signal, *F 13 Trip Delay Time from CPC Generated Signal to Time When Reactor Trip Breakers Open, seconds 0.70

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TABLE 4-3 RADIOLOGICAL DOSE TO THYROID, REM 2 HR, EAB 2 HR, EAB 8 HR, LP3 8 HR, LP3 GIS PIS GIS PIS SGTR with Degraded Spray-(10 GPM) 3.9 15. 2.8 3.7

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Acceptance Criteria from 10 CFR 100 and Standard Review Plan 15.6.3 GIS = 30 REM PIS = 300 REM

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TABLE 4-4 SEQUENCE OF EVENTS 10 GPM AUXILIARY PRESSURIZER SPRAY TIME SETPOINT (sec) EVENT OR VALUE 0.0 Tube rupture occurs 15 Letdown Control Valve throttled to minimum position ---

28 2nd Charging Pump On, pER level, feet below programmed level -0.75 28 Full Output Proportional Heaters, PER pressure, psia 2225 42 3rd Charging Pump On, PER level, feet below programmed level -1.16 57 Back-Up Heaters On, Pressurizer pressure, psia 2200 445 All heater cut off, pressurizer liquid water volume, Ft % 334/28 774 CPC Hot Leg Saturation Trip Signal Approximately 13*F away from sr.turation temperature --

775 Turbine Stop Valves closed, CEAs begin to drop Minimum SG 1evel, % Wide Range 70 Main Feedwater Control System begins to ramp down to 5% (Reactor Trip Override Setting) ---

780 Main Steam Safety Valves (MSSVs) open, psia 1070 783 Maximum steam generator pressure, both SGs, psia 1100 789 Pressurizer empties ___

790 RCPs tripped manually on low RCS pressure, psia

• 1621 794 Main Feedwater Control System supplying 5%

Flow 827 MSSVs closed, psia (Cycle thereafter until ADVs are opened) 1017 840 SIS starts to supply flow, psia 1400

TABLE 4-4 (C:ntinued)

SEQUENCE OF EVENTS 10 GPM AUXILIARY PRESSURIZER SPRAY TIME (sec) SETPOINT EVENT OR VALUE 1195 Operator starts to use ADV's to cool RCS temperature (hot legs) to 550*F. Cooldown rate = 100*F/HR ---

2808 Operator isolates the damaged steam generator, RCS hot leg temperature. *F '

i 550 4810 Unaffected SG hot leg temperature equals 500*F. Operator re-establishes letdown flow (letdown control valves in "AUT0") . Operator initiates auxiliary PER spray to depressurize RCS to 1000 psia ---

8946 Operator detects RCS pressure equal to 1000 psia ---

Operator starts to cool RCS at 50*F/HR using unaffected steam generator Operator controls auxiliary spray flow, pro-portional heater output, and HPSI flow to reduce RCS pressure, control subecoling, and maintain primary system inventory 20,808 Operator detects shutdown cooling conditions, unaffected hot leg temperature, *F/RCS pressure, psia 343/391 SDC Entry Conditions = cold leg temperature unaffected SG g 350*F and PER pressure between 392 psia and 350 psia.

Subsequently, operator tries to maintain these conditions 28,800 End of Transient #

FIGURE 4-1  ;

, OPERATOR ACTIONS FOR A SGTR  !

FROM WATERFORD 3 OP-902-007 Cooldown RCS to below 550*F with both SGs u

Isolate Affected SG h

Cooldown RCS to 500*F (at 100*F/hr) using ADV on unaffected SG n

When RCS at 500*F, depressurize to 1000 psia uring auxiliary spray, Re-establish letdown flow.

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Cooldown and depressurize RCS to Shutdown cooling conditions. Maintain optimum 4T between SGs.

Too Rapid: Too Slow:

Close ADV on  :

Control Cooldown to650*F/hr x Open ADV on Unaffected SG Unaffected SGl Too High: y -

Too Low:

Aux. Spray on Aux. Spray off Heaters off  : Control subcooling  ?!: 28'F n Heaters on Decrease HPSI Increase HPSI Close ADV Open ADV Too High: '

Too Low:

Decrease HPSI -: Control pressurizer level 2: 28%  :-- Increase HPSI h

Control SG 1evel to 77% - 94% WR using the blowdown system for the affected SG and 60% ~70% NR using main feedwater for ,,

the unaffected SG.

v Achieve Shutdown cooling conditions at 350*F and 392 psia

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5.0 CONCLUSION

This report confirms that if a charging loop isolation valve failed open during a SGTR event, the acceptance criteria (radiological dose consequences) would be met. Test results showed that an auxiliary spray flow rate of about 37 gpm was achieved with a charging isolation valve open.

This flow rate is much greater than the 10 gpm assumed in a SGTR analysis that showed the radiological dose released was well below.the acceptance criteria limits. Thus, the auxiliary spray system can perform its function

'. during a SGTR event such that the acceptance criteria are met.

This report confirms the inherent safety capability of the Waterford 3 plant to cope with various operational events and accidents.

k

Appendix A COMPARISON TO FSAR RADIOLOGICAL RELEASE RESULTS The radiological consequences calculated for the SGTR analysis presented in this report are compared to the results given in Section 15.6 of the FSAR in Table A-1. The results for this analysis show a higher radiological dose release then was calculated for the FSAR. This is not due to degraded auxiliary pressurizer spray, but is caused by.other analysis assumptions related to operator actions. Although a meaningful comparison between an FSAR analysis done for the purpose of licensing the plant and a more realistic analysis can not be made, the following describes the differences from the FSAR calculation that result in higher radiological dose.

The SGTR analysis presented in the FSAR is for a " stylized" event done with a prescribed set of assumptions. An important assumption for this comparison is that the operator takes no action for the first 30 minutes of the event. This means that the steam released to the atmosphere from the affected steam generator (the major cont * ".bution to the radiological dose) comes from a brief period where the tea safety valves popped open to limit steam pressure. After 30 minute. it i. s;sumed that the operator isolates the affected steam generator so there is no more steam released from it.

In the SGTR analysis presented in this report, operator action as identified in the SGTR Emergency Operating Procedure is accounted for. The key action for this comparison is that the operator opens an Atmospheric Dump Valve (ADV; assuming that the turbine steam bypass valves are not available) in both steam generators to cool the RCS hot legs before the

affected steam generator is isolated. The ADV in the affected steam generator was also opened periodically during the cooldown to maintain the optimum steam generator temperature differential (AT). This increases the amount of steam released to the atmosphere from the affected steam generator which increases the radiological dose consequences. Also, this analysis conservatively accounts for the release of activity from water removed from the affected steam generator through the blowdown system.

The basis for the first operator action is to quickly cool the RCS below the temperature at which heat transfer from the RCS would cause the steam generator safety valves to open. If the safety valves opened, there is a risk that one or more might fail open. With no block valve in this release path, the radiological dose consequence would be much more severe.

T,hus, the safer action is to cool the RCS quickly, using the ADVs if necessary, even though this results in a higher dose than calculated for the FSAR event. This action is consistent with CE's Emergency Procedure Guidelines that have been reviewed in detail and approved by NRC.

The optimum steam generator 6T is maintained to assure that natural circulation continues in the RCS loop with the affected steam generator.

In order to maximize the radiological dose released, this analysis assumed that the ADV was opened to cool the affected steam generator within the optimum AT. This is conservative analysis assumption since the Waterford 3 e-E0P does not allow the affected steam generator ADV to be opened af ter the initial RCS cooldown. Although these actions result in more steam being j released from the affected steam generator, the calculated dose is still l well below the acceptance criteria given to 10 CFR 100.

1 1

, o In conclusion, the_ higher radiological dose calculated for this analysis compared to the FSAR SGTR analysis is due not to degraded auxiliary spray, but rather, analysis assumptions related to operator actions.

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. ._ . ._. _ .m - . . _ _ _ . _

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_ TABLE A-1 SGTR THYROID DOSE RESULTS, REM Y

2 HR, EAB 2 HR, EAB 8 HR, LPE '8 HR, LPE GIS PIS g73 pyg d

Degraded Spray 3.9 15 2.8 3.7

- FSAR 0.8 0.51

. 03 .06 4

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